ML20199G177
| ML20199G177 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/20/1997 |
| From: | Graham P NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLS970194, NUDOCS 9711250114 | |
| Download: ML20199G177 (195) | |
Text
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. Nebraska Public Power District
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NLS97034 November 20,1997 ,
U. S. Nuclear Regulatory Commission
' Attention: Document Control Desk I
Washington, D.C. 20555-0001 Gentlemen:
Subject:
10 CFR 50.59(b) Summary Report Cooper Nuclear Station NRC Docket No. 50-298, DPR-46
.In'accorduce with the provisions of 10 CFR 50.59(b)(2) and Paragraph 6.5.1.C.2 of the Cooper Nuclear Station Technical Specifications, the Nebraska Public Power Distr:ct hereby subrits a summary of facility changes, tests, and experiments completed in accordance with the, requirements of 10 CFR 50.59(b). This report covers the time period from March 1,1996, to Augu.it 31,1997, in accordance with 10 CFR 50.4, the original report is enclosed for your use, and copies are being transmitted to the NRC Regional Oflice and the NRC Resident Inspector foi Cooper Nuclear Station.
Should you have any questions or comments regarding this report, please comact me.
Sincerely, PD714- '
)
P, D. Graham Vice President of Nuclear Energy
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U PDG:lb
' Attachment cc: . Regional Administrator USNRC - Region IV Arlington, Texas NRC Resident Inspector- CNS NPG Distribution w/o enclosure 9711250114 971120 A O \,
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l AT'fACHMENT 3 LIST OF NRC COMMITMENTS l Correspondence No: NLS970194 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 cy the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE COMMITMENT OR OUTAGE None I
l PROCEDURE NUMBER 0.42 l REVISION NUMBER 5 l PAGE 9 OF 13 l J
REPORTABLE DESIGN CllANGES (DCs). MINOR MODIFICATION PACKAGES (MMPs).
MQDjfLCATION PACKAGES (MPgl. ENGINEERING EVALUATIONS (EEs)
AND SOFIWARE DESIGN CilANGES (SDCs)
DC 89 284 -
TITLE: Control Rod Drive (CRD) Rebuild Room Modifications DESCRIPTION. This DC provided for renovation of the CRD Rebuild Room. Plant design and pemannel safety are improved and radiological exposure is lowered by implementation of this modification. DC 89-284 included the following: 1) replacement of jib crane,2) removal of existing concrete shield wall,
- 3) replacement and relocation of CRD storage racks,4) enlargement of south door opening and installation of new motorized steel roll-up door,5) installation of a new steel man door in east wall to allow access into hallway,6) construction of a new containment wall in the hallu ay cast of the rebuild area to exp4md the rebuild room storage area, and 7) installation of a new 120/208 VAC power supply.
SAFETY ANALYSIS: This modification does not increase the probability of occurrence or consequences of an accident or >
malfunction of equipment important to safety previously evaluated in the USAR because the CRD Rebuild Facility and a9 rebuild equipment located within it are considered nonessential. The implementation of this modification does not create the possibility of an accident or malfunction of a different kind than previously evaluated. The materials and installation procedures meet or exceed those quahty standards established by the original instrdlation of the interfacing systems. The margin of safety as defined in the Technical Specincations is not reduced as the operation of any system listed in the Technical Specifications is not changed or challenged.
MP 91-l l4 TITIE Steam Leak Detector Switch Field Wiring DESCRIPTION: This MP reconfigured the mountmg of the Steam Leak Detector (SLD) temperature switches to improve their resistance to damage during maintenance activities.
SAFETY ANALYSIS: This MP only alTects the mounting of the SLD switches. The affected equipment is not an accident iniuator and does not impact any equipment that could initiate an acciden.; This modification improves the durabihty of the SLD switches and does not alTect the performance of their safety related function, Acc.aent detection or accident mitigation systems are not adversely affected. This MP does not alter the setpoint or logic of the SLD switches. No new equipment functions are added and no electrical loading limits are impacted Equipment qualification of the switches is maintained. No new failure modes are introduced. The failure mode associated with damaged wire insulation is much less lixely after the modification, therefore, the probability of occurrence of a malfunction is reduced. Physical mounting is impnad with regard to::liabihty. Installation of the mahfied switches is per existing SORC approved procedures. This MP does not reduce the margin of safety as defmed in the basis for any Technical Specification.
- MP 92-155 TITLE: Drpvell F and O Sump Upgrade DESCRIPTION: This modification was implemented to improve reliability and climinate unnecessary maintenance for canponents :ceated in the drywell F and G sumps. The scope of this MP included replacement oflevel switches, addition of pump strainer shields, pump control and alarm setpoint changes, and procedure changes to throttle pump outlet valves.
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SAFETY ANALYSIS: Installation and testing of this modification w as performed during a refueling outage when operation of the sump systems was not required for leak detection and involved no changes to piping or Primary c Containment boundaries. The Drywell Equipment and Floor Drain Systems are not initiators for any -
accidents discussed in the USAR. None of the components located in the sumps are used to safely shut down the reactor or mitigate the consequences of any accidents. The only "important to safety" equipment that the sump systems interface with are the containment isolation valves located on the discharge piping from both drywell sumps. As a part of this inodification, the pump outlet valves will
, be throttled to reduce flonTates and decrease required submergence for the drywell sump pumps. This modification improves the reliability of the primary system leak detection and will not have a negative impact on operation of the containment isolation valves. The primary system leak detection systems are used only during power operation , are isolated during accidents, and have no impact on the ability to safely shutdown the reactor. This MP installs new level switches with the same basic function, but with improved reliabihty. The new level switch assemblics and pump strainer shields are nonessemial and will not be hwated near any safety related equipment when installed in the drywell sumps. New sump level setpoints will ensure existing interfaces with leak detection alarms are consistent and controlled in the future. The drywell sump components are not relied upon for any Technical Specification margins of safety.
MP 93 003 TITLE: Turbine Lube Oil (LOGT) and Turbine Generator El! Fluid (TGF) Modification (Salem Event)
DESCRIPTION: This nnhfication packagn was prepared in response to the turbine trip functional testing event accident at the Salem Nucle 2r Plant. The purpose of MP 93-003 is to provide local indication of turbine speed at the frcnt standard, to trnlify the turbine tnp logic to provide increased an areness of the circuit status, and to provide increased protection upon initiation of an emergency trip via TGF-SOV 20ET.
SAFETY ANALYSIS: This modification alters the turbine overspeed trip logie, but will not change its function. The changes to the trip logic do not have the potential to initiate a turbine overspeed condition because this logic does not open any valves that adnut steam to the turbine. The changes to the turbine trip circuit do proside a new way to trip the turbine electrically, but this condition is present only during testing situations and will be minimized /controlloi by steps in applicable testing procedures. The changes to the turbine trip kpc will not redum the kwl an the turbine in any way that has not already been described in the USAR.
ne panelloadmg and fuse protection for the AC and DC loads have been evaluated and found to provide adequate protection for the power sources to present loss of power. The alteration to the turbine trip systan will not impact the ability to trip the turbine locally. This modification will not impact the ability of any system to minimize the radiation (k>ses associated with the plant's response to an accident or malfunction. MP 93 003 is designed such that it will not impact the availability or operation of the TO tnp system under normal or abnormal operating conditions. The turbine speed indication in the Control Room is used for indication and provides no automatic functions. The sensors that detect turbine overspeal conditions are not changed by this modification. The mechanical overspeed trip mechanisms are not impacted by this modification and will still be able to trip the turbine if the electrical trip sywem fails. The intallation of this modification will not reduce the margin of safety as dermed in the basis for any Technical Specification.
MP 93 061 TITLE: Nitrogen Solenoid Operated Valve (SOV) Over-pressure Protection
- DESCRIPTION: This MP documented changes made under SORC approved MWR 93-1375 and defined additional changes needed to correct a design problem which the SORC approved MWR failed to resolve. The origmal MWR was pertinned to resolve issues highlighted by Generic Letter 91 -15 which alerted utilities to various factors that could ptentially resuh in solenoid valve maloperation. The additional work under tius MP replaced and reclassified regulators IA PRV-PRV740 and 74 I as essential components. This 2-
4 d
MP also replaced PC PRV PCV631 with a smaller n gulator, as well as replacing and reclassifying -
PC-RV-12RV arx! PC-RV 16RV as essential components. In addition, PC RV 17RV was removed as it no longer performed a useful function. i SAlliTY ANALYSIS: This activity does not alter the design, function, or configuration of the primary coolant pressure 2 boundary, reactor monitoring sy.tems, or any reactor control systems. The components afTected by this activity cannot, in themsekes, be a becane accident initiators. Work has been restricted to a time when
, the affected containment isobtion valves can be placed in their isolated conditions, ensuring that the consequences of an aa:ident cannot be increased during the nulification's implementation. Rather, this
. activity is intended to prevent a single failure from potentially increasing the consequences of an accident by ensunng containment integrity. His activity does aher the setpoints of one pressure regulator and two ~
relief vuhes, but does no to ensure that the operating parameters of equipment important to safety using this header as a pressure source will continue to ope ;te within design limits. The probability of equipment malfunction has been reduced since the new regulators will be procured, installed, and -
maintained as essential cunponents A fault analysis was performed to determine if any potential causal relationships were created as a result of this activity and none were found. This activity is specifically intended to reduce the cansequences of an equipment malhnetion that was not adequately resolved, ensuring that ovnali plant safety will be increased By preventing a single active failure from potentially~
affecting contairunent integrity, this activity ensures that the existing margins of safety remain unchanged. ,
a MP 93-107 TIT!.E: Safety Relief Valve (SRV) Stellite / Platinum Pilot Disc Replacement
< DESCRIPTION: ne purpose of this MP was to rglace the stelhte pikit discs in four safety relief valves (MS-RV 7 i ARV, MS-RV 71BRV, MS-RV 71CRV, and MS-RV 7 liiRV) with stellite / platinum pih>t d.scs and make the stelhte/platiman pikit &se replacement performed by Special Test Procedure 93 107 (MS-RV 71DRV, MS RV 7iERV, MS-RV 7IFRV, and MS-RV 7 IGRV) pennanent. The original material, Stellite 6, i was nnhfied by the addition of0.3% platinum. Operating data indicates that use of the modified stellite improves the reliability of the popping pressure for the subject valves.
SAFETY ANALYSIS: The adhtion of the 0.3% platimun alkiy will mt mercase the probability or consequences of any accidents or equipment malfunctions evaluated in the USAR. This imxlification is not considered to adversely e
impact the design, material, or construction standards applicable to the sysicm or equipment being modified Review of available information indicates that under excess oxygen cond tions, the electro <
chemical potential of the stellite / platinum pilot disc could be slightly higher than that for the original
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Stelhte 6 pikit asc. Ilowever, Ocneral Electnc hasjudged that the possibility of significantly increasing ihe corrosion lxnl strength is low ne nnhfication does not change t'Te SRV accuracy or response time, s
. and the moditication dies not place the SRV outside its & sign nvelop The purpose of this
' nuhtication is to increase the pmbability of the proper functioning of the SRVs. It does not result in the operation of any system outside of an analyzed condition and does not increase the radiological
- consequences of a malfunction of an SRV. The margin of safety related to SRV actuation contained in the Technical Specifications is based upon the analysis of the accidents a td transients discussed in the USAR. Based upon testiv,by General Electric, it is determined that the !ikelihixxl of the modification to contribute to increased corrosion bonding is low and hence will not reduce the margin of safety as defined in the Tecimical Specification basis for the operation of the SRVs.
MMP 93-179 TITili: Plant Management Information System (PMIS) Data Concentrator Replacement DESCRIPTION: This MMP replaced the two data concentrators of the PMIS with a s3 stem which utilizes VAX 4000-100 computers and a VME chassis. The previously existing data concentrators used hard disks that am no langer supported by the manufacturer. The new computers also replaced the PMis VAX 11-785 computers.
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ANALYSl& flhis mahfication imuhrs exm-safety rulated hardware which has no direct interface with any equipment -
important to safety. During installation, lie-PNL-UPS) A will be removed from senice resulting in the
- loss of de Rai Positim Infamation System (RP!S)/ Rod Worth Minimizer (RWM) computers for a 6 to 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perial, however, this outage will only be performed when the RPIS/RWM computer is not -
required PMIS is a passive data acquisition, processing and trending system which does not have direct _
interface s ith any safety related system or component, and the outage of EE-PNL-UPSI A will not increne the consequences of an accident or increase the probability of a malfunction of equipment i important to safety, The non-safety related equipment imtalled for the minor mmlification will not directly or indirectly cause degradatkm of any safety related system or component; therefore, no Technical Specification margins of safety are afTected -!
MP 94-006 TITLik Condemate and Feedwater System Sample Points Upgrade Dl!SCRIPTION: Ihis mahlicatim provided for the installation of two additional sample points on the Feedwater System for nxmitaring azul wrificatim of fm! water metals omeentrations entering the reactor, two new sampling points on the Condensate Demineralizer System to pmperly measure the concentration of trace metals, i a irw valve to provide leakage protection for the sample point on the South Condensate Storage Tank, and documentation for seven existing Condensate Drain System sample point valves which were previously installed by DC 86-137.
SAFETY
- ANALYSIS: 1he new sampling points on the Fmlwater and Condensate systems will enhance the performance of the Process Sampling System. 'lhis installa6an conforms to the USAR design basis of the Process Sampling Systent The new samphng pmbes are designed to withstand scismic forces and the forces from process flowi The design also took into account the etTects of cormsion due to pmcess fluid stagnation at the interface of the sampling probe and the inside surface of the pipe. The additional sample point isolation vahr on the South Condemate Storage Tank is an enhancement to the existing configuration and will not alTect the design basis of the Condensate Storage System. The installation of the new sampling points and the sample point iselation vahc is classified as non-sufety, non-seismic, and nonessential. It will not affect the safety function of any equipment in the plant. Th rethre,it will not increase the pmbability of occurrence or concluences of an amident or malfunction of equipment important to safety. The Pmeess Sampling System and the Condensate Storage Tank sample valves are not defined in the Technical Specifications and no margin of safety is established for the feedwater metals concentration or the Condemate Storage Tank leakage protee' ion. Therefore, the installation of this MP will not reduce the margin of safety us defined in the tiasis for any Technical Specification.
MP 94 072 TITLE: Residual Ilent Removal (RllR) Loop "A" Drain Path to Radwaste DliSCRIPTION: Prior to this mahfication RIIR Loop "A" could not be placed in senice with the reactor coolant
, temperature prater than 212*F. The existing piping configuration allowed a drain path to Radwaste for Loop "A" only thmugh the division crossover valve. Opening the crossover valve prior to cooldown below 212'F would make both low Pressure Coolant injection (LPCI) loops (RilR "A" and "11")
inoperaNc. Also, RI1R loop "A" could not tu ilushed with loop "11" operating in the Shutdown Cooling Male because there is no path to Radwaste. This mahlication installed a 4" drain line from the RIIR crosstic piping to the existing R1IR piping leadmg to the Radwaste systent This will enable either loop of Rl!R to be drained to the waste surge tank without afTecting the other loop of RIIR.
SAFETY-ANALYSIS: 'lhis imtallation will take place during normal plant operation when the Rl_ IR Shutdown Cooling Mode is not being used The new piping will meet the appropriate design imd fabrication standards applicable to the RilR system. The pressure boundaries of the R1IR system are maintained and this change will allow more flexibihty in preparing the system for senice or maintenance. Following imtallation of this MP, RIIR-V-101 will be maintained /contmtled as a " scaled closed" valve per Operations procedures to 4
minimize the possibility of an operator error that might result in the mispositioning of RIIR-V.101. If both RilR V 101 and Rl[R-V-102 were open at the sarne time, both RilR loops would be connected, and ifCl would be inoperable. The installation and testing of the new drain piping can be performed witixiut afketing the availability of either loop of the RilR system. Ilowever, if a loop of RIIR has to be taken out of service for any reason, the other imp of R1IR is capable of supporting the design functions of the system. His wuuld require entry into a Technical Specification Limiting Condition for Operation.
This MP increases the availability of the RlIR loops by allowing them to be drained / warmed up independently, No existing automatic functions will be impacted by this change and no new automatic functions are added. No new accident scenarios are created as a result of this MP. The core and containment cmling systems are discussed in the Technical Specifications and bases; however, the installation / testing of this MP will not impact Technical Specification requirements and will not reduce the margin of safety as defined in the basis for any Technical Specification.
MP 94-162 1TIR Main Steam Pressure Transmitter (MS-PT-81 A&B) Replacement DESCRIPTION; This modification replaced the existing main steam line pressure transmitters, MS-PT-81 A&B, with Rosemount model i 151GP transmitters. The existing pressure transmitters were manufactured by the ElectroSp Corporatin and neither parts rur new replacement transmitters are carrently n ailable. This MP also installed new redundant power supplies to the transmitters.
SAlTITY '-
ANALYSF This nxxhfication potentially affects the probability of the pressure regulator failure event. Ilowever, the new pressure transmitters are as reliable or better than the old transmitters. The new redundant power supplies to the tramtrutters will reduce the chance of pressure regulator failure caused by losu o! electrical power. No systems or components needed to mitigate or control the consevences of any accident previously evaluated in the USAR are otherwise affected The main steam pressure transmitter and the pressure contmller are classified as nonessential and non-seismic, but these components are important to safety. ne pressure controllers are used to help control the position of valves that admit steam to the Main Turbine. Potential malfunctions associated with the pressure controller are discussed in Chapter XIV of the USAR; however, this modification restdts in no change in potential failure ;aodes or effects of a pressure controller malfunction. The functional and dynamic performance of the pressure regulator system with the new transmitters and power supplies will be the same as before; therefore, given a malfunction of equipment important to safety, the systcm will perfonn as it did previously to control reactor pressure. The pressure regulator and electrical power supplies are physically and electrically separated from any equipment important to safety. Cntical safety parameters for any analyzed event where the pressure regulator is assumed to function will not be alTected; therefsre, there is no change in margin of safety as defined in the basis for any Technical Specification.
MMP 04-316 TITLE: Permanent Documentation of Plant Temporary Modification (PTM) 94-25 DESCRIPTION: 'Hus MMP documented, by paperwurk only, the installation of PTM 94 25 as a permanent change. This PTM installed a 1/4 stainless steel tube through a spare conduit between the Cable Spreading Room and the Cable Expansion Room. The tubing was installed for use as a pressure sensing line to facilitate the perfonnance of the Control Room Emclope Pressurizat;on Test. The conduit was rescaled during the installation to meet fire banier rating requirements. No additional work was pertbrmed under this MMP.
SAFETY ANALYSIS: Since no addithud wtuk was performed under this MMP, the 10CFR50.59 analysis performed for PTM 94-25 is applicable. There are no accidents related to this spare fire barrier penetration seal described in the USAR. No important to safety equipment passes through the affected conduit. Fire seal components meet or exceed applicable requirements. A continuous fire watch was maintained during installation which ensured that the nxxhfication would not alTect the consequences of an accident. Except J
fa the fire seal, this PTM had no interaction with equipment important to safety. The modification was performed when the C<mtrol Rmrn Emergency Filtration System was not required to maintain a
= pressurii.ed Control Roorn Envekipe.
MP 94-324 -
TITill: Altemative Powei to II% MOT 01F-C-ID)
DESCRIPTION. MP 94 324 pmvides attemative power for i1% MOT-(BF C 1B)in the event of a loss of backup power for Division 1. The modification was initiated due to a concern regardmg operability of equipment
. supporting the Control Room Envekpc. This MP allows the selection of power source so that fan . ,
operability is possible when either division of power is inoperable.
SAFETY ANALYSIS: _ Tic affcet d equipment is not an accident initiator and does not impact any equipment that could initiate an accident Procedure changes are made to address changes in operation requirements so that the new equipment can be operated as required by Technical Specifications and the USAR. The MP improves the reliability of the Control Room Envelope, The Control Roorn Envelope is not a single-failure proof sysicm and its failure before or aller the MP implementation will result ie a Control Room habitability issue. This modification climinates the single failure mode involvmg loss of Division 1 power and subsequent loss of til fan. All other equipment failure probabilities are unchanged and no new equipment malfunctions are introduced. A seven day Limiting Condition for Operation is applied while the equipment is iroperable due to implementation of this malification The credible accidents requiring =
the operability of this equipment are bounded by the fuel handling accident during shut <k)wn and Loss of Coolant Accident (LOCA) during power operations. No new equipment functions are added and the reconfiguration will not impact the load capability of afTected equipment. The Technical Specification margin of safety is unafrected by this MP, MP 94-365 TITLE: Turbine Exhaust I kul Temperature Contrdler Upgrade DESCRIPTION: This malification replaced temperature controllers MC-TC-19A and MC-TC-IP13 for the Turbine Exhaust flood Spray Valves with reliable thermocouples and electronic temperature controllers, Operation of the system with the existing Leslie mechanical temperature controllers has been erratic and prone to failure due to vibration induced calibration drift.
SAFETY ANALYSIS: While new temperature control components from different manufacturers will be used after the moddication, the temperature control function, operating conditions, and system parameters for the Turbine Exhaust I hui Spray dystem will remain unchanged. Improved system reliability will reduce demands on operating personnel dunng low power operations and will not increase the pr > ability of a:cidents evaluated in the SAR. The instrument air tie-in for the new controllers will no e made into air lines serving equipment important to safety. The Turbine Exhaust I kul Spray System has no impact on the ability to safely shutdown the reactor. Temperature controller components are not actively used to mitigate the consequences of any design basis accidents or abnormal transients. The only important to safety equipment thm this system interfaces with is the Main Condenser This installation will have -
no impact on operation of the Main Condenser, ne new components are nonessential and will nu be installed near any safety related eqmpment power fcr the new temperature indicating controllers will be from a previously used and vacated source. The impact on electrical loading and breaker / fuse '
canhnation will be insigmficant. Removal of abandoned in place equipment under the mochficction will climinate unnecessary interfaces and decrease the probability of malfunctions for the Oxygen injection System. The Turbine Exhaust Iloal Spray temperature control components are not referenced in any ,
Technical Specifications and are not used to provide any Technical Specification margins of safety.
Setpoints for operation of the spray valves will be unchanged by the modification.
MP 95-042 TITLII: Turbine Eqtuptnent Cooling (TEC) Pumps Mechanical Seals Upgrade
- DESCRIPTION: Ris mahficatim package was generated to auArize replacement of the mechanical seals in TEC Panps A, B, and C. De rotating face material is being changed from carbon to reaction bonded silicon carbide (RSC) and the stationary face material is being changed from ceramic to RSC. He seal 0-ring is being changed from Viton to Ethylene Propylene. Work will be performed when any one of the TEC pumps mechanical seals is declared by the system engineer to be leaking excessively. To date, only TEC Pump C has been upgraded, SAFETY-
, ANALYSIS: The TEC system is not an Essential system and does not cool any Essential equipment. It is not an accident initiation or mitigation system for USAR evaluated accidents. The TEC system, ahhough important fa nmnal operatxm,is not needed for safe shutdown of the reactor, if the modific-1 seals leak, the TEC head tank will make up the loss of water innmediately. It would only cause leakage to the Turbine Buddmg which would be contained by the floor and equipment drains and would be processed by radwaste treatment systems. Seals typically fail gradually with a slow leak; however, a major loss of water would cause an alarm to sound in the Control Room on low head tank level. If a pump seal leak is determined to be excessive, the alTected pump will be shut down, valved out, and another pump will be placed in service without the km of the TEC system. The modified seals should be more reliable and decrease the likelihax! of problems with the TEC sostem. The TEC system is not mentioned in the Technical Specifications and there are no Technical Specifications that the TEC system directly affects.
If the seals were to fail, the leakage wotdd be less than the leakage from evaluated pipe failures.
QC 95-101 1 TITLE: Pipe Support Mahlications for iligh Pressure Coolant Injection (HPCI) Vacuum Breaker Replacement DESCRIPTION: This DC Amendment provided for pipe support modifications associated with the llPCI turbine exhaust vacuum breaker piping as a result of reanalysis of piping due to the vacuum breaker and pipng replacement performed per DC 95-101, in order to maintain code qualification of piping. It also providd for contingency pi: . airport mahtications in the event thet water was found in the iIPCI exhaust piping aller iIPCI system testing; however, implementation of these contingencies was not required.
SAFETY ANALYSIS: The pipe support modifications have been designed to Code requirements for all pertinent IIPCI load combinations. De malifications ensure that the IIPCI turbine exhaust vacuum breaker piping remains within the allowable stresa. No adverse etTect m the llPCI system safety ftmetion wdl occur. All existing operational, transier.t, and accident parameters of any systems or components remain unchanged.
Implementation of tlus modification is in accordance with approved Cooper Nuclear Station (CNS) prouhres and within the Technical Specification defined Limiting Conditions for Operation. All existing mitigating ihnctions, actions, and barriers remain unafTected. Therefore, this activity cannot increase the probabihty ofoccurrence or consequences of an accidertt or malfunction of equipment important to safety.
All margins of safety as defined in the Technical Specifications remain tmchanged.
MMP 95-116
'lITLE: Utilities Ad& ion for Lunch Trailer DESCRIPTION: This Minor Modification provided additional electrical power to a trailer which was converted into a lunch cafeteria trailer. Additional [x>wer was required to support the kitchen appliances being added to the trailer.
SAFETY ANA! YSIS: This Mmor Mahfication aflects the 12.5 KV system only. Loss of the 12.5 KV system would not impact the plant safety systems and is not an accident evaluated in the USAR. Therefore, this modification does not increase the probability of occurrence or consequences of an accident or malfunction of equiprnent important to safety. No new accident scenarios are created by adding additional loads to the 12.5 KV 7
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. system. The system has sullici,:nt capacity to support the additional loads. The 12.5 KV system is not safety related and the addition ofloads to the system does not impact any margin of safety as defmed in the Techmcal Specifications.
EMP 95-118 TITIR Fire Protection Main Tie-in for the New Technical Support Building (TSB) .
DESCRIPTION: This MMP installed a UL listed mechanical joint tec connection in an existing 81 underground fire -
protection main running bett een the Training Center and the West Warehouse. A new post indicator isolation valve was also added at the branch to permit the isolation of the main from the branch. The branch was installed downstream of the valve to supply fire protection water for the new TSB sprinkler system and hydrant. The installation of the tee and valve required entry into a seven day Limiting Condition for Operation per Technical Specifications to permit filling / flushing oflines.
SAFETY ANALYSIS: This activity does not afTect any equipment whose malfunction is postulated in the USAR to initiate an accident or prevent an eccidem fmm occurring. This installation was perfonned to the same Codes, Standards, and design paranaters as Ge system being modified, this maintains the original design basis to supply the largest single emand and hose stream allow.mce. Appropriate isolation devices are provided to isolate this new system during construction as well as after completion to allow for flow to other areas requiring fire suppression. This activity does not change the operation, function, or design basis of any system, stmeture, or component important to safety and no new hazards are created. The Technical Specification requirements and bases applicable to the Fire Protection system are not affected.
hwtallation activities are performed in accordance with the Technical Specifications; theref,re, the margin of safety is not reduced.
MMP 95 118.2 TITLE: Tic-in of the New Technical Support fluilding (TSB) Fire Alarm arid Commmications to the Existing System in the Emergency Operations Facility (EOF)
DESCRIPTION: This MMP provided the tie in of the Gai-tronics, fire alarm, and potable water for the new TSB to the EOF. It installed a new circuit breaker in an existing electrical distribution panel to provide 240 VAC for the new battery charger kicated in EOF Raxn 103. It expanded the EOF Conununications Room 104 into Room 103 by creating an opening in the wall between the two rooms. An automatically activated wet sprinkler system was inttalled in the two rooms to replace the IIalon system. The existing EOF 1IVAC was changed by disabhng in the open position the Room 104 automatic ventilation damper which was previously required to close to minimize the spread of Halon. This MMP also included the documentation requl red to convert the A/E drawings for the new TSB mto CNS site drawings.
SAFUTY ANALYSIS: This MMP installs non-safety related components for fire protection / detection, communications, and potable water fbr the TSB, which will not have an effect upon the probabdity of an accident or occurrence of a fire. It will not change the accident mitigating function of any safety related equipment, The new Gai-tronics end electrical circuits are properly protected and coordinated so that a local fault will have no adwtse impact on existing systems, The existing level of fuse coordination hetween OPA-PNL-CEOF and its emergency feeder MCC-T is not changed by this mahfication. The TSB Gai.Tronics will be fused so that a locallitult will be isolated and not impact the Communications system. The addition of the wet sprinkler in EOF Rooms 103/104 has been evaluated and determined to have no impact on the existing fire protection design. The wet sprinkler system will extinguish the same kinds of fires as the Halon system; hoivever, it is expected that the equipment in Rooms 103/104 .may be damaged? The-Communications room does not contain equipment important to safety. The wet sprmkler system is acceptable per American Nuclear Insurers. The types of accidents that might be created (e g., floodmg due to sprinkler system operation) are similar to those already evaluated in the USAR and remain tunted by USAR asramptions If the EOF becomes unusable as a result of fire, the Alternate EOF will be used per the Emergency Plan Procedures. The non-safety related equipment installed for this MMP will not directly or indirectly cause degradation of any safety related system or component. The margm !
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i I
I of safety described in the basis fir any Technical Specification is not alTected by this activity. Engineering Judgement 96-012 documented that the additional demand of the new sprinkler system would not adwnicly affect the operability of the fne supptession system or impair the capability of other sprinkler systems serving safety related areas.
Ignporary Desien Change (TDC195-129 and Amendment 1 TITLE: Temporary Security llarrier for Exterior Reactor BuilJmg Door RI 15 DESCRIPTION: Thi., TDC documents the inallation of a tem;vxary security barrier on the extenor of the Reactor Buikhng Door R115 locatal at elevation 903'6' et the southwest comer of the Reactor Building. During an Operational Safeguards Response Evaluation, the NRC detennined that the subject dcor did not provide un adequate security barrier for preventing unauthorized personnel from entering the Reactor Building with explosives. Nebraska Public Power District (NPPD) responded to this concern by installing an additional security barrier (concrete slabs) on the exterior of this door under Maintenance Work Request (MWR) 95 1918. llowever, this MWR was implemented wi'hout documenting a 10CFR50.59 Rgxtrability Analysis. TDC 95-129 was subsequently issued to appropriately document the 10CFR50.59 Reprtability Analysis and Sately Evaluation for the current installatiott The in talled barrier is currently meeting security barner requkements and will remain in place until a permanent nxdfication is implemented Amendment I was issued to include provisions for removal of the barrier in the event of a floal and to require a ir.adification package to document permanent removal of the barrier.
SAFETY ANALYSIS: The placement of the concrete slabs on the crenor side of Reactor Building Door R115 does not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. The slabs caanot cause a malfunction of equipment importe.nt to safety. No new features or equipment are added by this TDC which would require a now accident analysis No new typea of design loads are added directly a irdrectly. The placement of the concrete slabs does not reduce the margin of safety as defined in the basis of any Technical Specification. The performanen and reliabihty of Secondary Containment and the Reactor Building have been shown to be unaffected by the installation of the concrete slab security br.rrier 1-:r Nuclear Engineering Departmern Calculation 95-177. This calculation shows that during a seismic or tornadic event, the concrete slabs will not load the Reactor Building walls in excess of the design seismic and tornado generated missile loads-MMP 9MA2 TITLE: Plant Temponuy Modifications (PTMs) 93-33 and 93-37 Documentation DESCRIPTION: This Mmor Modification documented work already performed by PTMs 93-33 and 93-37 as permanent changes. P fM 93-33 replaced flow irxhcatcr REC-F1-475B with a spool piece and PTM 93 37 replaced flow irxhcator REC-F1-467B with a spool piece. Justification for operation without flow indication was previously pmvided in the PTMs on the basis that the indication function is not important t< safety. For REC-F1475B, flow irxlication is only required during cycle surveillance testing; dunng these times, flow indication is pmvided by a portable ultrasonic flow meter. For REC-F1-467B, flow indication is not required since there are no associated surveillance testing requirements. In both cases, an existing flow switch continues to provide the low flow alarm functiott SAFETY ANALYSIS: Since no additional work was performed tmder this MMP, the 10CFR50.59 reportabihty analyses perfooned under PTMs 93-33 and 93-37 are applicable. The probability of an accident is not increased due to the spool piece components meeting or exceeding the original piping material and strength requirements. The spool pieces are also less susceptible to breakage. Accidents related to the REC system described in the USAR remain v.aatTected. Replacement of flow indicators does not affect the consequences of an event. The original flow indicators had no safety design basis in the Technical Specifications. REC system features described in the Technical Specifications have not been altered and the margin cf safety is not affected.
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Mal:fication Packnec (MP)95-182 TITLE: Video Camera Installation ,
DESCRIPTION: Ris MP imtalled 22 vidm cameras and monitoring equipment in the Turbine and Reactor Buildings and allowed for the temporary installation of six cameras in the drywell. This modification was performed.
to pnnide a met}xxl of real-time monitoring inside high radia6cn areas of the plant without unnecessary radiation exposure to plant personnel.
SAFETY ANALYSIS: There are no accUent initiators rasociated with this MP, No process or ekctrical conditions will be l changed as a t k af this installation. There are no seismic concerns introduced by this MP.
Components of tha n tm will not be mounted above any safety related components in the Reactor Building. The design anu function of existing plant equipment are not being changed. There are no equipment interfaces with existing plant systems, structures, or components, except for deriving power from existing convenience receptacles and attaciunent of condwt to non-seismic structural steel on elevation 1001' of the reactor bmlding. The video system does not directly or indirectly interface with equ!pment important to sar cty. Therefore, the consequences of equipment failure remain unchanged by this installatiert Cameras are mounted such that if they were to fall they would not come in contact with anything safety related This installatiou. not affect the margin of safety as deGned in the basis for any Technical Specification I?E 96-001 TITill: Rigid Components Spanning Seismic Gaps DESCRIPTION: Three ProHem Identification Reports were written which identified installation discrepancies across wismic gaps between various pent buildmgs. A seismic gap between two buildings is provided to ensure independent itxnement ofeach buildmg dunng a seismic event. This Engineering Evaluation documents the acceptability of the existing configuration.
SAFETY ANAL.YSIS: The non-safety related components spanning the seismic gaps are ficxible with small masses and may potentially fail during a significant seismic event. The failure of this equipment is not part of the accident analysis These comjonents were not designed to mitigate an accident in the USAR. The plant has been designed for safe shutdown viithout the function of these components and no failure mode exists for these items that could prevent the performance of any safety related equipment. The potential failure of these components / supports during a scismic event will not be a safety concern and will not cause a malfunction of any equipment important to safety. The building structures will function as designed, i.e., the nxntment of the twu buildmgs will not be restricted The existence of the non-safety related components spanning the seisuic gap between buildings does not reduce the margin of safety as defined in the basis for any Technical Specification. This configuration was evaluated to ensure that the required design basis has been met.
EII 96-002 TITill: Concrete Cracks on Turbine Nonh Exterior Walt DESCRIPTION: Cracks were identified on the north wall of the Turbine Building that allowed water to seep into the building during periods of rain. This EE repaired and sealed the cracks per a vendor recommended nrocess.
l SAFETY ANALYSIS: The repair of the cracks on the north wall of the Turbine Building will restore the facility to its original design intent while maintaining the same configuration. Furthermore, the Turbine Building is not an accident initiator. The cracks do not affect the structural integnty of the wall and the repair will prevent further degradation of the wall. There are no activities during the implementation of the repair work wluch will affect any equipment important to safety. Neither the Turbine Building nor its north wall are imuhrd in the perflumance of a safety related function. The margin of safety as defmed in the basis for
- any Techmcal Specification is not afTected by this repair a::tivity.
MP %0li 7 TDMU Main Steam Isolation Valves (MSIVs) Packing Replacement Dl!SCRIPTION: This MP w as generated to replace the packing in the MSIVs with ARGO packing and to install a new stem material which is more resistant to ge"ing. MS AOV AOS6C was previously modified with the new etem material and packing configur. 2 under MMP 95163. The new packing configuration provides better scaling and helps mitigi .ne stem damage experienced with the old packing. The nnhfiol vah>e was ins;xcted aller several months of senice and determined to be sealing properly with rn signs of stem galling. Wrefac, this MP authorized the installation of the new packing / stem material in the remaining seven MSIVs The stufling box was enlarged to allow the use of standard size packing, thejunk ring was renoved, and a difTerent packing material installed. This modification was completed on MS-AOV 86A, MS-AOV A08613, and MS-AOV-A086D. Modifications to MS-AOV A080A, B, C. and D were not perfarmed during Rlil7, but are expected to be completed in a future outage.
SAFETY ANALYSIS: 'lhis nnhiication 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 haw any etTect on any other analyzed accident. It simply imptoves the scaling of the MSIV's stem packing. A minor amount of metalis machined from the valve bonnet but has a negligible cfrect on the strength of the bonnet. Secondary Contairunent will be maintained when required durmg the installation The stem fiiction will be reduced with the new seal design, so valve stroking will not be adversely afTected. This activity does not affect the accident mitigating capabilities of any system. The only_ equipment afTected by this modification is the MSIVs Improving the stem senhng and reducing the stem friction will not increase the probability ofinadvertent closure ef a MSIV nor will it increase the probability of the MSIVs not closing during a containment isolation During the modification installation, the valve work wili proceed until Secondary Containment is a!Tected. Waen Secondary Containment is afrected, 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 tle 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 venfied by satisfactory completion of post modification testing. No new and unanalymi failure modes ara created. Changing the packing of the MSIVs does not hase any etTect on the margin of safety defined in the Technical Specifications. Secondary Containment will be maintained w hen fuel is being moved as required by Technical Specifications.
EII %012 and Rev. I l
TITLIh Intake Structure Ice Deflector livaluation l
DESCRIPTION: 'this EE documents the acceptability of the intake structure ice deflector. The ice deflector is a seasonal installation that is phmt in the Missouri River in late fall and removed in the spring. The current design did not comply with the design requirements and installation as specified in Mmor Design Change 76-111. This EE provides an neceptable design basis for the ice deflector which is supported by Nuclear l Engineering Department Calculation 96-047 and NPPD drawing CNSBLDG421.
SAFETY i ANAL YSIS: A proposed fabrication, installation, and orientation of the ice deflector or its hence atTects only the ice deflector and the intake structure guide wall. The changes in the ice deflector will have no efTect on any intake system or component. The ice deflector is not required to rappert any safety celated equipment and its absence will not result in an accident or abnormal transien Changes to the ice deflector or its absence will not increase radiation exposure to personnel inside the plant or at the site boundaries. In response to prnious safety questions, it was determined that the service water pumps are unafTected by not having the ice deflector in position The changes to the ice deflector will have no elTect on equipment malfunction since no safety related equipment is atTected The ice detiector is not addressed in the basis
!br any Technical Speedication Thus, any change or absence of the ice deflector will have no impact on the margm of safety as defined in the basis for any Technical Specification.
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-EE E017 TITLE: Change in Pipe Schedule for One Inch Diameter Piping DESCRIPTION: The one-inch diameter piping dowmtream of MS-RO-130 was determined to be the wTong schedule piping. This piping is the drain line for the Main Steam piping and is classifial as Seismic Class 11S.
DC 86 101 had replaced the subject piping with Schedule XXS Stainless Steel, but the as-installed condition shows a Schedule 80 Stainless Steel pipe is installed. This EE evaluated the as-installed conditioa and found it to be acceptablec SAFETY
- ANALYSIS: An analysis was performed that concluded the change in pipe schedule is acceptable and the pipe will perfoim to its original design requirements The schedule change of the pipe will not alTect any.
equipment important to safety. Therefore, this activity does not increase the probability of occurrence tu consequences of an accident or malfunction of equipment impanant to safety. The pipe will maintain its structural and pressure integrity and satisfies code requirements The margin of safety as dermed in the basis for any Technical Specification is not reduced due to this schedule change since the pipe will perform to its original design intent and is within code compliance.
- EE %018 .
TITLE: Repair of Oiler / Filter Station #M DESCRIPTION: Lubricator SA LUB-10 was permanently removed from the plant as it no longer perfumed any useful function and was also damaged it was replaced with % inch Schedule 80 pipe. The lubricator was part of Filter / Oiler Station #52 which prmides pressurized air to the Fuel Pool Filter /Demineralizer whose resins do not tolerate oil contamination very wcit.
SAFETY ANALYSIS: De Fuel Pool Filter /Demineralizer and the ponion of the Service Air sycem supplying pressurized air
- to it have been previously analped in the USAR as not being a potential accident initiator. A loss of the Fuel Pool Filter /Demineralizer as the result of this activity will not increase the olTsite dose, nor would the faihire of the Fervice Air system. Therefore, the probability of occurrence or consequences of a plant ewnt evaluated in the USAR are not increased. This activity does not involve ary equipment previously determined to be impatant to safety, nor does it alter any operational or accident parameters of any plant equipment; ther::forv,it dces not increase the probabihty of occurrence or consequences of a malfunction ofequipment important to safety, This activity does not create any new failure modes in either the Fuel Pool Filter /Demineralizer or Senice Air system. Removal of the lubncater does not reduce the reserve capacity or design margins of either the Fuel Pool Filter /Demineralizer or the Senice Air system; therefore, this activity is incapable of decreasing the margins of safety of any plant equipment.
EE 06-019 TITLE: 1IPCI-LS-90 Seal Weld DESCRIPTION: nis activity added a seal weld to the 11 oat chamber / enclosure tube threaded joint ofIIPCI-LS-90. The seal weld was added to correct a potential steam cutting of the gasket seating surface.
- SAFETY ANALYSIS. This activity only adds a seal weld to a threaded joint which is permitted by the system design and fabrication codes. The addition of the seal weld cannot alTect any of the plant events evaluated in the USAR. The consequences of a plant event remain the same because this activity does not alter any -
assumptions or parameters used in the analysis of the events The only malfunction relevant to this activity is the has of pressure boundary ne seal weld will be installed in accordance with system design and fabrication nxtuirements; therefore, the pressure integrity is maintained at the same level of strength and quality. The weld adds negligible mass to the switch so its seismic characteristics are unchanged.
The consequences of a loss of pressure bounday integrity at the joint where the seal weld is to be instalhxl are not increased De kiss of the joint uouhl result in HPCI being isolated and the presence of the seal weld would not cause the mr.lfunction to propagate to other systems. The mechanical strength yy g-w 7 3 - w.--. --r -
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of the joint is still maintained by the threads. De weld ordy provides a better leak-tight joint. The initiation setpoints, tinw to rated performance and performance capabilities (e g., flow and pressures) are not altered by the addition of the seal weld.
MP %022 TITLik Nash float Valve Modification
- DESCRIPTION: his nxhficatia. imtalls strainers in the inlet to the Nash float valves in the condenser vacuum priming -
system. They are designed to allow the vacuum pumps to vacuum all of the air out of the water boxes, -
but filter debris and mud that could cause the Nash float valves to stick open.
. SAFETY-ANALYSIS: This modification has no efTect on any analyzed accident occurrence. The only systent being modified is the condenser water box vacuum pnnung system, which is a non-safety system. If the vacuum priming system failal,it could cause some air to build up in the water boxes which could cause a slightly higher corskmser back pressure on tha turbine, which would slightly reduce plaut etliciency. Ilowever, this will not cause a lost af condenser vacuum. No safety system nor its accident mitigating capabilities are afTected by a slight reduction ofcondemer vacuunt The manner in which the main condenser is operated will not be altered by MP %-022. The main condenser is used to mitigate the con.wquences of some accidents described in Appendix 0 of the USAR; however, this MP will have no impact on the ability of the main condenser to perform as descnbed in Appendix 0. This modification does not alter the design, configuration, or operating parameters af any system important to safety. Installation af strainers in the vacuum priming system does not alTect any safety margins.
SDC %023 TIT!E Installation of 3 Dimension Core Monitor System (3D MONICORE) and Stability On-Lit. 5 ODYSY Monitor System (SOLOMON)
DESCRIPTION: His Sollware Design Change pnnided instruction for installation and testing of the General Electnc 3D MONICORE and SOLOMON systems. Tne new 3D MONICORE system replaces the existing 2D MONICORE system arxl provides are accurate calculations of thermal limits. The SOLOMON system is integratalwith the 3D MONICORE system's software and hardware. It uses on-line plant data from the 3D MONICORE system to determine the power-flow operating state relative to the stability butter region and exclusion regim It also combines the data wnh the ODYSY program to evaluate and predict core and hot channel decay ratio.
SAFETY ANALYSIS: The 3D MONICORE is strictly a monitoring system. It does not provide a safety heiction and does not support any plant equipment. The SOLOMON system uses data from the 3D MONICORE system and does not provide input to any equipment. The USAR accident analysis does not address the existing core monitoring system. The replacement system does not change the function of core monitoring. No intrusion or interference wnh any equipment important to safety or considered in the accident analysis is imulml with this mochficatim The new system does not directly support the plant's operation and there is no potential to alTect a malfunction of equipment important to safety even with a complete system failure. The 3D MONICORE system is a passive system; , '.oes not provide an automatic action or function to any plant equipment. The SOLOMON system uses dua from 3D MONICORE but does not interfere with the 3D function. The 3D MONICORE system does not provide input to any equipment considered in the Technical Specifications and its data is not used to determine any margin of safety.
MP %C34 TITIE Main Condenser Structural Repairs DESCRIPTIONu During Main Condenser internal inspections conducted during the 1995 Refueling Outage, numerous structural problems were noted This MP repaired cracks on the bafile plate of the Main CorWnse .
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replaced the stifTener plate of the Main Steam Bypass Valve Sparger, replaced the pipe supports of the one inch vacuum seming lines, and restored the operability of conductivity probe MC CE 6B.
SATTITY ANALYSIS: The Main Condemer is used to mitigate the consequences of some accidents described in the SAR.
Ilowever, the changes being made by this MP will not impact the ability of the Steam Jet Air Ejector -
(SJAE) system to function and will have no impact on the operation of the Main Condenser. No new failure modes are being introduced. Due to the large size of the Main Condenser, this MP will have no -
impact m the usable volume in the Main Condenser or hotwell storage capacity. The ability of the Main Condemer support systems to detect abnormal discharges of radioactive gases in order to minimize the radiation doses associated with the plant's response to an accident or malfunction will not be impacted by this mahricatiort ne stmetural repairs made by MP %-034 are designed to strengthen the internals of the Main Condenser. The structural and weld materials will be ksigned and installed such that they will be aHe to withstand the normal and transient conditions in the Main Condenser. No new functions are being added by this MP. The Main Condenser is mentioned in the Technical Specifications with reganL to the functions of the SJAE. This MP will have no impact on the SJAEs or on the basis for any other Technical Specifications MP 96-035 TrfLE: Installation of Elevated Release Point (ERP) OH Samplar Sample Line Ileat Trace DESCRIPTION. This MP added a comtant wattage type heat trace on the sample line up to and on the ERP GE Stack Gas Sampic Rack 17 18. The ERP GE Stack Gas Sample Rack servet, as the alternate sampler when the ERP Kaman Normal Range is not capable of particulate and iodine sampling. This MP was initiated due to a problem with water accumulation in the particulate / iodine filter on this sampicr several times during the year, SAFETY-ANALYSIS: The ERP GE Stack Gas Sample Rack is not associatea with any accident initiators discussed in t'ie USAR. The installation of the new electrical load will not have any adverse impacts on the electrbal distribution system The ERP GE Sample Rack has no automatic functions while monitoring gaseous ellluents. The equipment is used as a backup to .he prunary radiation monitor. The installation of the heat trace on the backup ERP sample unit will not impact the ability of the prima:7 rediation monitoring unit to perform its normal or post-accident functions. This equipment will not be used to mitigate the comequences of any accident or transient discussed in the USAR. Since the sample rack is not required for redundancy, the installation of this MP will not create the possibility of a difTerent type of accident than previously evaluated The new components will be installed on nonessential equipment that is not located near any other safety related equipment. The Stack Gas Sample Rack is not addressed by any Technical Specification and is not used in the basis for any Technical Specification.
EE96-016 TITLE: Installation of Emergency Operating Procedure (EOP) Wrench for CRD-AOV-CV33
~ DESCRIPTION. This Engineering Evaluation documents the acceptability of the field installation of a ten inch crescent wrench which is utilized in the EOPs to loosen kicknuts on valves CRD-AOV CV33 and CRD-AOV-CV35. This wrench is hanging over instrumentation tubing for the Control Rod Drive (CRD) solenoid pilot valve drain valve. Even though the toal is secured, it constitutes an unauthorized modification that must be evaluated SAFETY ANALYSIS: The wrench is securely attached to a lanyard which prevents it from impacting any essential components.
The only equipment within reach of the wrench or under it that could potentially be damaged in the event of an earthquake is IA PI 229,IA PS-230 and their associated instrument valves and tubing. Although worst case failure could potentiatly impact the Instrument Air (IA) system, the IA system is nonessential
- and ' act credited for actions or features used to mitigate or prevent the consequences of an accident as l discussed in the SAR. The 1A system is not relied upon to provide a safety action in support of CNS special events. Since the IA system provides clean, dry, oil-free compressed air, there would be no
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' increase in radioactivity even if the system was breached. Where air is needed to support systems or . +
equipmem important to safety, accumulators and other design features ensure an nacquate supply of r indgedent of the IA systent Neither the wrench, the loss of the lA system, nor the " alternate means of ns! insertim procedure" are addressed in the Technical Specifications or bases. Therefore, this activity does not reduce thc margin of safety as defined in the basis for any Technical Spaification.
IM 94039
- TITLE: Ioop Seals for Drywell Fan Coil Units
- DESCRIPTION: Ris !!E added hxp seals in the drairJines offof the drywell's fan coil cooling units. These loop seals are required to ensure tne proper drainage of condensate fmm these cooling units.
SAFETY ANAINSIS: Without these kup seals, the fan coil units will drain in a step function manner, crecting the illusion that an identified leak in excess of Technical Specification limits has developed within the drywell This activity will decrease the pmbability of a plant event by ensuring that draining occurs at a rate egal to the rate of condensate geireration Whether or not these loop seals are installed, the consequences of a loss of drywell cooling or an accident remain unchanged. A failure of the drainline will not cause a malfunction of the drywell coolers, nor will it affect any rearby safety related equipment. The loop seals .
- being added do not have the sim and mass to become a missile with a greater energy level than those 3 previously essumed in the loss of Coolant Accident analyses. The addition of the loop seals does not create any new failure modes for either the drainline or the coolers, n r will it alter the drainline's requiral design characteristics. This aethity does not change the design, function, or configuration of any safety related equipurnt, nor does it create any new equipment interfaces. This activity does not c .ange any design criteria, design limits, or assumptions used to establish any margins of safety.
. MP 94034 TITIR Upgrade of Residual Ileat Removal (RIIR) Motor Operated Valves (MOVs)
DESCRIPTION: The following valves were identified as marginally performing MOVs with a small operating. margin:
RilR-MOV MO34A, RIIR MOV MO348. RIIR-MGV-MO39B, RilR MOV-MOl7, and
, _ RIIR-MOV-MO39A. As a result of this problem, the affected MOVs require more frequent testing under dynamic conditions to demonstrate operability. In addition, diagnostic testing of RIIR-MOV-MO39B indicated a possible misalignment of the valw bonnet, stem, and disc u hich could be detrimental to valve performance. Therefore, the following modifications were implemented.1) replaced motor on RIIR MOV-MOl7,2) rei aced d motor gear sets on RIIR MOV M014 A and B,3) replaced spring pick on RIIR-MOV-MO39A, and 4) replaced valve, yoke, and spring pack on RIIR MOV-MO398.
SAFETY ANA13 Sis: Replacement components will have the same or higher performance qualities as previously existing _
car.ponents ami will meet the same system design requirements and safety classifications. The operating _
conditions for the systems which will interface with the valves to be modified will not be atiected The cmtml logic for these vahes will not be altered. The modiiications will be implemented when the safety related function of the system will not be atTected. The RilR trains containing MO39A MO39B, MO3&, and MO34B will be out of service during installation and performmg no safety related function:
MOl7 is a Prtmary Containment isolation valve and one of tuu (in series) RHR shutdown cooling suction valwa Cmstraints in the M." will ensure that installation and testmg is completed within all applicable limiting conditions for operation. Any change to the performance of the atTected components has been evaluated with no new failure modes being identified.1he resulting increase in operating margin will 4
result in a reduced frequency of dynamic valve testing and will ensure that th : valves will perform their
. designated safety functions. The DC distribution system load will increase slightly due to the increased motor size of RilR-MOV MOl7; however, this change has been evaluated and it was determined the
. additional load will not impact the ability of these systems to perform their safety related functions as
' discussed in the Technical Specifications. The modification to RIIR MOV MOl 7 will slightly decrease its stroke time, however, the new stroke time is below the Technical Specification limit. This modification does not reduce the margm of safety es defined in the basis for any Technical Specification.
MP 96 057 TITI E: 1IPCI-MOV Mol4 and ifPCI MOV Mol6 Motor Operated Valves (MOVs) Dear Change DESCRIPTION: Ris MP replaced the nx*r gc.ar set and valve stem on IIPCI-MOV-M014 and replaced the motor gear set on IIPCI-MOV Mol6. The higher ratio gear sets will increase the torqueAhrust capability of the actuators which will in turn increase the operating margin of both MOVs. These modifications were -
implemented because both of the subject MOVs had only a small operating margin in the closing direction.-
SAFETY ANALYSIS:- %e replacement components will meet Se same system design requirements and safety classifications as the existing components and will have the same or higher performance. The configuration and operating characteristics of the systems which interface with the valves to be modified remain unchanged.
He nxxhfications will be implemented during cold shutdown. Although the stroke time of the valves will be changed, the new times haw been evaluated as b:ing acceptable and will not impact the ability of11igh Pressure Coolant injection to perform its design functions. The resulting increase in operating margin will result in a rethiced fnxiuency of dynamic valve testing and will ensure that the valve will perform its designated safety functions = The added thrust / torque will be addressed and controlled through existing CNS procedures. This MP does not increase the probability of occurrence or consequences of an accident or malfunction of eqaipment important to safety and does not reduce the margin of safety as defined in the basis for any Technical Specification.
MP 96-059
'ITfLE: REC-MOV 702MV and REC-MOV 709MV Valve Rep!x nent and Operator Upgrade DESCRIPTION: Ris mothfication removal and replacal the existing motors, motor operators, valves, vah e yokes, spring packs, and steni nuts for REC-MOV 702MV, Reactor Equipment Cooling @EC) Drywell Supply isolutmn Valve, and PEC-MOV 709MV, REC D,Twell Return isolation Valve. It also required replacement of nx*r overlomi heaters aad modification of pipe supparts that interfered with installation of the larger replacement motors. The existing valves had high valve factors and were marginal performance motor operated valves (MOVs) in regard to thrust / torque output. This modification will improve plant design by increasing torque / thrust margin for 702MV and 709MV.
SAFETY ANAL,YSIS: Existing valve components will be replaced with components having the same or higher performance qualities. The replacement components meet the same design requirements and safety classifications as the existing components and will be scianically and environmentally qualified for their intended functiott The operating conditions for the systems which interface with these valves are not affected. The new valves will have a slightly increased stroke time, but the new stroke time is well below the Inservice Testing stroke time of 54 seconds. The motor geanng capability will be increased which will produce more thrust / torque capabihty am! meets the design basis requirements for these MOVs. The modification of pipe supports will not change the fur.ction, operation, or reliability of the supports which will be designed, constructed, and inspected to the requirements of the original components. The modification will increase tle operating margin between the thrust / torque produced at the torque switch trip and thrust hmitation of the motor operators. These vahrs' accidctit mitigating funetions and sMK to perform their design basis safety related functions will be improved by these modifications and there wf. be no impact on the consequences of en accident previously evaluated in the SAR and no new failure modes are introduced There is no stroke time requirement for REC-MOV 702MV and REC-MOV 709MV in the Technical Specifications. The AC distribution system load has increesed slightly due te the increased motor size, however, these changes have been evaluated and it was determined that the additional load will not impact the ability of these systems to perform their safety related functions as discussed in the Technical Specifications.
MMP 96-079 -
TITLE: liigh Pressure Coolant injection (IIPCI) Steam Leal Detection (SID) Test Switches Installation .
DESCRIPTION: Surveillane Procedure 6.llPCI 301, llPCI Steam Line Space Temperature Switch Functional Test,
- required the deenergization of IIPCl MO M015 and M016 in the valve open position in order to prevent their automatic isolation during the testing of she SID switches. This isolation is undesirable due to the resultant isolation of Augmented O!T Gas (AOG) which is fed by the llPCI line downstream of tin two Primary Containment (PC) isolation valves. Deenergization of the two PC isolation valves ir the open position is also under t rable since the PC isolation fune' ion would be defeated. This modilbation -
adds test switches to the 9-39 and 9-41 panels in the Auxiliary Relay Room to permit testing of the SLD switches in tach channel without disabhng the other three SLD channels in order to preserve PC isolation capabihty, ADO operation, and prevent excessive motor operated valve operation.
SAFETY ANALYSIS: This modification adds a test circuit that is completely deenergized (except for the essential test switch itself) durinF power operation other than -luring surveillances. The SLD test switches and associated logic are not accident initiators. The changes do not afTect the logic or the performance of any safety function. The same SLD function and response will be maintained. There is no added mode of
- equipment malfunction associated with the addition of these test switches. The addition of the switches adds an extremely small possibility that switch faihire can affect the performance of one of the four channels of SLD (single failure). This leaves three of four channels functional. Only ore channel is required to perform a Group 4 isolation; therefore, the 7nodification is acceptable. The proposed modification does not have the potential to create a new type of accident since it only alTects the SLD subsystem which is not associated with any accident initiators. The margin of safety is not reduced because the modification does nt t impact the ;wrformance of the SLD subsystem's safety function of detecting a steam line break and initiating a Group 4 isolation.
MP 96-081 TITLE: RCIC-MOV-Mol6, MS MOV-M074, AND MS MOV-MO77 Motor Operator Upgrade DESCRIPTION: The subject motor opetuted valves (MOVs) have exhibited marginal performance. The closing thrust / torque for MS-MOV-MO74 and MS-MOV-MO77 has been marginal. Therefore,-the operating margin was increased by replacing the existing SMB-000-5 motor operators with SMB-00 10 motor operaters As part of the scope of the original moditication, the motor operator on RCIC-MOV-Mol6 was also to be replaced. Ilowever, the mahlication planned for RCIC-MOV-MO16 was subsequently suspended because approximately 35% udditional margin was established for this MOV. (Reference NLS970069 dated May 7,1997, from P. D. Graham (NPPD) to the NRL,"Redsion of Commitment; Motor Operated Valve Upgrade")
SAFETY ANALYSIS: Existing valve components are being replaced with components having the sanw or higher performance qualities. The replacement components meet the same design requirements and safety classifications as existing components and will be seismically and environmentally qualified for t!dr intended function.
The op.reting conthtions for the sptems which interface with the valves will not be atTected The motor -
operatoni will haw adequate thrust / torque capability to open and close the valves against tne worst case
, design basis conditions.1herefore, these vahrs' accident mitigating functions and ability to perform their design basis safety related functions will be improved by these mahf: cations. This modification does not alter the function of any components, equipment, structures, or systems and does not introduce any new failure modes lhe stric time for MS-MOV-MO74 and MS-MOV-MO77 has been increased slightly.
Ilowever, um times are well below the Techrucal Specification limit. The AC and DC distribution system loads haw men ased slightly due to the increased rnotor size. However, MP 96-081 has evnluated these changes and determined that there are no adverse impacts and the additional load will not impact the ability of th se estems to petionn the;r safety related functions as discussed in Technical Specifications.
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MP E083A
-TITLE: Upgrade of Conduit and Cable Tray llanger RBil 1089 DESCRIPTION: ne purpose of this nnhfication was to upgrade conduit and cable tray hanger RBil.1089 (located in the NE quadrant of Reactor Building 903) to meet design requirements. This hanger must be upgraded as part of CNS' commitment to address Unresolved Safety issue (USI) A.46," Seismic Qualification of Equip nent in Operating Nuclear Plants."
SAITITY ANALYSIS: During implementation, the existing support will remain intact with external stiffening added. Sufficient precautions will be taken to ensure ths.t no operable components are adversely impacted during the installation process. Following completion, the new configuration will have increased capacity to resist seismic forces. This activity ensures that the cable tray support will fulfill the USAR requirement for Class I Seismic. Because the implementatim is wholly external to the existing support, at no time during implementation is the existing factor cf safety decreased. This activity increases the capacity of the hanger and can only decrease the probability of occurrence or consequences of an accident or malfunction of equipment important to safety, and can only increase the margin of safety.
Mi 4 083H TITLE: Anchorage Upgrade of Control Re: a Panels, Auxiliary Relay Room Panels, and the Reac'.or Protection System (RPS) Motor-Generators DESCRIPTION: ne purpose of this modtfication was to upgrade the seismic ruggedness of equipment identified during seismic walkdowns to address I mresolved Safety issue A-46. Included in this MP are the addition of anchorages to Control Room Cabinets and Avxiliary Relay Room Cabinets, the addition of" bumpers" to restrain RPS Mar-Generator Setc., ar.d the securing of Control Room ceiling ditTuser panels.
SAFETY ANALYSIS: This nuhfication meases permanent safety fuerrs for the Safe Shutdown Earthquake (SSE), thus the probability of SSE related accidents will be the same or lower. During implementation, work will be performed only externally, thus preserving existing safety factors. During the Auxiliary Relay Cabinet work, safety factors will temporanly decline during installation; however, this work will be donc during a division outage. Precautior,a will be taken to ensure that no operable components are adversely impacted during the installation process. At all times, the probability of occurrence or consequences of an accident or malfunction of equipment important to sCety will be the same or less. This activity increases the capacity of the equipment for the design loads; therefore, it only serves to increase the marein of safety.
DC %087 TITLE: Cycle 17 Fuel Leaker Replacement DESCRIPTION: This Design Change removed leaking fuel and symmetric btuuties frcm the core. New fuel installed duttng the last refueling outage was shifled to these locations and the core load was completed using fuel bundle previously discharged to the Spent Fuel Pooi.
SAFETY ANALYSIS: The core and fuel design haw a direct impact t n the plant response to a number of transients and amidents evaluated in the USAR. Ikwvever, the chrages themselves do not modtfy any of the equipment malfunct ons or procedural err,rs that are analyzed as accident initiators in the USAR. The changes do not Jmpact the integnty of the fuel cladding, the first barrier to the release of radioactivity from the fuel.
The core design has been analyzed in accordance with NRC approved methods (described in the General Electric Standard Application for Reactor Fuel [GESTAR] II). The fuel bundles meet all of the fuel licensing acceptance criteria of GESTAR 11. The core changes also do not adversely impact the integnty of the reactor vessel and reactor coolant system, the second barrier to the release of radioactivity from the plant. Analysis of the Emiting overpressurization event demonstrates that the peak calculated
_ pressures are less than those allowed by the ASME Code. The changes do not directly impact the
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a operatim of any systems necessary to mitigate the consequences of an accident or transient, or to safely shut down the plant. Margins of safety potentially afTected by the changs are related to fuel limits .
(Minimum Critical Power Ratio [MCPR], Maximum Average Planar Linea lleat Generation Rate
_ [ MAPL 1IGR}) ni to the reactor wsselheactor emlant system (maximum pressure). Analyses oflimiting
_USAR transients for the cycle establish operating limit MCPR values that ensure that the safety limit MCPR is not violated Therefore, the taargin of safety to fuel cla6.w failure due to insuflicient cladding heat transfer during transient events is not reduced. For the lws of Coolant Accident (LOCA), the MAPL 110R limits are devchiped to ensure that the limits of 10CFR50.46 are met. By meeting these limits, the consequences of a LOCA are not increased and the margin of safety is not reduced. Analysis of the linutmg overpressunzation event for the cycle demonstrates that peak reactor vessel pressure is less than the ASMH Code limit of 1375 psig. Thus, the margin of safety between the ASME Code sa'ety limit and the actual failure point of the vessel and reactor coolant system piping is not reduced.
MP %-092 TITI.E: Replacement of Senice Water (SW) Motor Operated Valve (MOV) SW 37MV DliSCRIPTION: Ris mothfication replaces SW-37MV (SW Pumps Crosstic) with a new valve, replaces the existing gear box on SW-37MV, and reorients the valve operator. This MP will improve plant design by providing the necessary shaft exposure on SW 37MV to support Valve Operation Test and Evr.luation System (VOTES) testing required by the CNS MOV Program.
SAFETY-ANALYSIS: The installation of this MP is being performed during shutdown, vehich will minimize its impact on plant operation. The "B" division of SW must be out of senice, however, the "A" division of SW will remain in senice at all times during installation. Temporary cooling will be provided to the Turbine Equipment Cooling (TEC) heat exchangers and a temporary supply to the Screen Wash system will be provided if necessary. Other equipment seniced by the isolated petions of SW and 13C are atrected, but are requinxi only dwing operation. Replacing SW 37MV and its gear box does not alter operating conditions or system parameters or intmduce any new failure modes. SW 37MV is not opable of initiating a USAR evaluatal accident. De valw's safety function, which is to close on low SW "B" header pressure or hip water level in the Control Duilding, has not been reduced. This MP does not alter the function or configuration of any components, equipment, structures, or syrems required to safely achieve shutdown of the unit. Since the primary function of this modification is to allow VOTES testing, reliability of the valve is enhanced. Replacement components base been evaluated for the design basis functions of this valve and meet the design reqmrements. Feilure of SW 37MV would have the same consequences before or aller this MP.
MP 96-095 and AmendmenL1 TITLE: Installation of IONICS Water Treatment System DESCRIPTION: nis modification installed a new leased water treatment system manufactured by 10NICS in the area of the Water Treatment Building. Some components were installed inside the building and some were installed in a mobile trailer kicated outside of the building. The previously existing Plant Makeup Water Tn atment System and components have become obsolete and costly to maintain and operate. In addition, it pmes a significant hazard to personnel due to recurring acid and caustic leaks in the water plant. The new 10NICS system was installed in parallel with the existing system. Operations personnel will have a choice of using the new system, the old system, or both systems concurrently. After this modification has been proven to be reliable and capable of sening the needs of the station, the old system may potentially be abandoned.
SAFETY ANALYSIS: - The new 10NICS system is a nonessential system that will not be installed near any safety related equipment. This equipment will not be required to be in-senice to safely shutdown the reactor or to minimize the consequences of any accident or abtwrmal operational transient.1he connection to the 12.5 KV ring bus will have adequate protection to ensure that a fault at the load will not impact the
.12.5 KV ring bus and the connection will not increase the probability of a loss of oft-site power. The new
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system will rxt impact the operatim of the Fire Protection system. The equipment used to generate fresh water is rux required in order to nunimize the radiation doses associated s,ith the plant's response to an accident or malfunction 1hc modification installs a new 10,000 gallon water tank inside the Water Treatment Building. Failure of this tank could cause the loss of electrical and mechanical equipment inside the buikling. I kmmr, rxme of the equipment inside the building is relied upon to safely shutdown the reactor or maintain the reactor in a safe shuthwu com! tiort The Water Treatment system is a nonessential system that is not discussed in the Fasis for any Technical Specification
- MP %102 TITIR Digital Microwave Upgrade DESCRIPTION: Tbis Muhrication Package provided electrical power to a new Ccmmunication Building kicated on the north side of the Technical Support Buildmg which houses the equipment required for the microwave upgrude, 'This modification added a 15 KVA transformer supplied via the 12.5 KV systent SAFETY ANALYSIS: 1his nxaficatim afrected the loading of the 12.5 KV sy stem; however, the small load increase does not have an adverse impact m the operation or stability of the system. The loss of the 12.5 KV system is not an accident evaluated in the USAR. The 12.5 KV system is not relied upon to support accident mitigation functions and no systems required for accident mitigation ere impacted by this modification during installation or operatiort The ability of the 12.5 KV system to provide plant event response power was not afrected during installation or following completion of this MP. Consequences of failure of the 12.5 KV system are not changed by this MP. The kul ackh! will tax cause any system or equipment to exceed its design capacity. An internal fused transformer was used which will prevent any fault from atTecting the n maitxler of the 12.5 KV system which provides power for equipment defined as important to safety.
'lhere is no nuirgta ofrafety dermed in the Technical Specifications regardmg the 12.5 KV system. The installation of tins MP does not impact the performance of any safety related equipment nor any margins of safety as defined in the basis of the Technical Specifications.
MP %107 rnd Amendment i TITIR Emergency Operations Facility (EOF)/fechmcal Support Center (TSC) Upgrade f
DESCRIPTION. The purpose of this MP was to nxufy the EOF arx! TSC to panide a more organized and etlicient facility from which to manage emergency response activities. The initial modification package provided the necessary design change information to modify the EOF; Amendment I provided the necessary infonnation to modify the TSC.
SAFETY ANALYSIS: 1he EOF nudfication irsults in a rnhetion to the electrical load of the EOF. Tne EOF has the capability to be swvervd from the Emergency Diesel Generators Engineering Judgement 96-086 documents and quantifies the load reduction and its conservative impact on Diesel Generator loading. Therefore, this MP nill rxx have an adverse impact on the Diesel Generators or any other equipment whose malfunction is posnilatalin the USAR to initiate or prevent an accident. The EOF modification does not impact any other equipment that functions to mitigate the consequences of an accident and does not create any new spatial or functional interaction with equipment important to safety; No new hazards are created that could be putulated to cause a ditTerent type of accident or equipment malfunction than those previously analyzed ~he Technical Sprtfication requirements and bases for the Emergency Diesel Generators are not alTeeted by this nuaticatiort No other equipment important to safety is atTeeted by this moddication.
The TSC moddication adds electncal load to Panel LPTSC which has t!.c capabihty to be powered by the Emergency Diesel Generator. The new kiad on Panel LPTSC is enveloped by the existing load axnumx! fbr the panel in the Diesel Generator kaling tabulations Details are documented in Engineering
.hidgement 96M *the TSC tmatication will not alTect the ability of the Diesel Generators to perform '
their safety related function or affect any other equipment whose malfunction is postulated ta initiate or prevent an accident. The 1SC mochlication does not impact any other equipment that functions to mitigate the consequences of an accident and no new failure modes are added that could alTect the 20 i,.
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reliability of the Diesel Generators - The nodification does not create any new physical or functional-
, _ interaction with equipment important to safety and does not create any new hazards that could be postula51 to catue a ddTercrd type of accident or equipment malfunction than thw presiously analped JIhe Tcchnical Speedicatim requirements and bases for the Ernergency Diesel Generators are not affected by this nxxlification: No other equipment important to safety is alTected by this modification.
MP %112 - ,
TITLE: Reactor Core Isolation Cooling (RCIC) Appendix R Modifications DESCRIPTION: 1his modification removed RCIC system Appendix R vulnerabilities to assta e avadability of the RCIC system to support hot shutdown in the everd of an Appendix R fire in the Reactor Building 93 l' north and
' trutheast areas. This modification installed a t cw one amp fuse into the RCIC "A" logic circuit for the Reactor Iligh Water Level trip, an associated relay to monitor the fuse integrity, and another fuse on the negative side of the RCIC "A" logic panel to enhance the protection of the power to the RC;C logic panel.
SAFETY
' ANALYSIS: During installation and testing of this modification, RCIC will be electrically isolated to preclude any impact on other systems. Sizing. selecting, installing and testing the fuses, relay, and conductors per the applicable codes and standards will casure that the RCIC logic circuit operation remains unchanged. The parts will be designed and installed to meet the requirements of an Essential system. The changes made by this MP will not impact the availability of RCIC under normal or accident conditions. Except for the fire scenario for which this moddication is being implemented, the installations made by this MP are transparent to RCIC system operation for all modes During a fire, the high water level turbine trip function is not assumed. This modification will not increase the probability of oceurtence or consequences of an accident or malfunction ofequipment important to safety. The operation of RCic and its trip and isolation instrtunentation is addressed in the Technical Specifications. Ilowever, this -
nuidication does not change the vaha: of any RCIC parameter. Thus, the margin of safety is not changed
- MP %113 TITLE: D.etel Generator No. 2 Appendix R Modifications DESCRIPTION: This modification package removed Diesel Generator No. 2 (DG2) Appendix R vulnerabihties. The existing configuration of the DG2 logic and control circuitry was vulnerable to various tire-induced hot shorts. This MP added a second redundant power supply for the DG2 control power circuit under Attemate Shutdown tire conditions. The modification utilized an existing isolation switch and installed a new fuse and conductors into the control circuit far DG2.
SAFETY ANALYSIS: 1his modification is being installext to meet the requirements for an Alternate Shutdown scenario. The backup power source added by this modification will be isolated from the DG2 control power circuit at all Smes except when operating DG2 locally from the Diesel Generator room There will be no impact to the performance of DG2 under normal and accident conditions. Following the installation of this MP, there u a pctential (only when DG2 is isolated for local operation from the DG room) that the Digital ,
Reference Unit (DRU) for the govemor may fail to its lower limit aller a loss of DC power. This potential impact will be runowd by adding appropriate procedural controls. The parts installed by this MP meet or exceed the same quality requirements as the parts currently installed in the DG control circuits. The abihty of DG2 to supply the necessary standby AC power needed tc upport loads that are required in enter to minimee the radiation doses associated with the plant's response to an accident or malfuretion will not be impacted by this mochlication. The operation of the Diesel Generators and the DC power system are addressed by the Technical Specifications and Bases _ he Techtreal Spec .ication requtrements for power availability as supplied fnun DG2 will not be impacted by this modification. The
- battery and bus kwhng calculations are not impactal by this change because no new DC louds are being ackled. Therefore, the installation of this MP will not reduce the margin of safety as defined in the basis i for any Technicel Speettication. -
MP %116 -
TITIJk 1leating ik>iler Blowdown Piping DESCRIPTION: This MP involved replacing portions of the auxiliary condensate drains blowdown piping from the Auxiliary Steam (AS)Ileating Boilers to the blowdown tank. Specifically, the 3 inch AS 1Icating Boiler blowdown header piping that connects to the AS Ileating Hoiler blowdown tank was replaced, as well as sections of I 1/2" piping that connect to the 3" header. The replacement piping is heavier schedule piping to pro;ide the additional corrosion allowance that the piping system requires for the service conditions Additionally, two unused, capped connections (originally installed for blowdown from the fuel oil boilers which have been remomi) wcre deleted 1his modtfication will also install insulation onto the boiler blowdown piping from the boilers to the blowdown tank; however, this portion of the modification has not yet been completed SAlliTY ANALYSIS: These changes are to a section of non-safety, non-seismic piping that is pressurized during boiler blowdown The modification increases the wall thickness of the afTected piping and climinates unused connections, making the p; ping less likely to leak than before. None of the accidents or transients analyzed in the USAR involve this piping, either directly or indirectly. Thercibre, this MP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety. This modification does not create any new failure modes and has no etTect on any equipment important to safety. 1his MP does not affect the results of any safety evaluation or calculation; therefore, there is no efTect on the margiu of safety for any critical safety parameter. Thus, this change does not miuce the rargin of safety as defined in the basis for any Technical Specification.
MP % I21 TITIR Scram Dischage Instrument Volume Second Auto isolation Drain and Vent Valves DESCRIPTION: The pucpose of this modification was to install redundant automatic isolation valves in series with the i existing Scram Discharge Instrument Volume (SDIV)/ Scram Discharge Volume (SDV) drain and vent isolation valves. This modification addressed a design deficiency with the previous modification to the SDV system per Minor Design Change 81 010-1. The existing configuratio. consisted of a manual valve in series with an automatic air-operated valve (AOV) on the drain line, and an automatic AOV in series with two parallel check valves on the vent line, ihr each of the two SDIV/SDV installations. This configuration failed to meet single failure criteria without operator intervention as required by Safety Cntenon 2 of the Generic Safety Evaluation Report of NRC Bulletin 80-17. Safety Criterion 2 states "No single active failure shall result in an uncontrolled loss of reactor coolant."
SAFIITY ANALYSIS: This moddication will not alter the basic function of the SDV or degrade the function of any equipment.
The new arrangement per MP %121 will meet the single-failure criterion with regard to containment of reactor water coolant and will have a positive impact in preventing the release of reactor coolant water.
The installation of the new vahrs and piping will not impact the abihty of the SDV system to perform its design function to limit the loss of reactor w ater discharged from all the control rod drives during a scram The vahrs and piping will be designed and installed to meet the design requirements of the SDV and Control Rod Drbt systems MP 96-12 I will be installed w hile the reactor is shutdown. The release ofreactor water during the installaf ,will be kept to a minimum by draining the SDIVs and the vent and drain pipes r ore to beginning work. The SDVs will be monitored during installation to detect any l
leakage into ha vnlumes that may require draining. Provisions will be provided to manually drain the suhimes,if required. The existing SDIV drain valves and SDV vent valves are not Primary Containment l isolation valves. 'fhe new valves will be installed outboard of the existing valves. Technical l Specifications contain a surveillance requurment for the SDV vent and SDIV drain valves to close within -
30 seconds upon receipt of signal for a control mi scram. This will be maintained with the new design.
The function of the SDIV level instnunentation as defined in Technical Specifications will not be impacted by the temporary use of the instrument drain valves for emergency draining of the SDV during the installation phasec The installation / testing of MP %121 will not reduce the margin of safety as defined in the basis for any Technical Specilication.
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MP 96-121 TITLE: Steam Tunnel Roof Replacement -
DESCRIPTION: his nnhficatim doctanents the replacemmt of the existing roofmg material on the 932 ft level roof area of the Reactor Building. This area includes the area outside of the Critical Switchgear Rooms on which are located the nonessential llVAC units ihr the Control Building and the Critical Switchgear Rooms.
This area is also referred to as the steam tunnel roof. The existing roofmg material in this area is in a !
degraded condition SAFETY ANALYSIS: Ris activity affa:ts only the exterior ronf of the Reactor Building. No accidents evaluated in the USAR are alTected in any way by this activity. Tbc proposed roormg replacement does not affect the function of any equipment important to safety No equipment that is important to safety is located in the work activity area for this modification. He roofmg material can only be postulated to fail in the event of extreme winds which are currently esaluated in the USAR. The new installation method prmides superior adhesion to the concrete when compared to the existing roof and thus is less likely to fail. No safety margins as dermed in the Technical Specifications are affected by the replacement of the steam tunnel roofmg material with a different type of material. The roormg m arial weight decrease wdl not reduce the margin of safety for the structure, MP % 123 TITLE: Technical Support Building (TSB) Trailer Addition DESCRIPTION: This Mothfication Package provideo electrical power for the trailers positioned west of the TSB. While the placement of the trailers is temporary in nature, installing the required power safely requires that the electrical installation be made permanent. This MP utilized some ofic spare capacity which exists on the 12.5 KV system and transformed it to the required voltage. The modification installed a 100 KVA transformer, ten disconnect switches, and a Gai-Tronics communication power feed from an existing terminal box. -
SAFETY ANAL,YSIS: This mothfication affects the kiadmg of the 12.5 KV and Gai-Tronics systens The loss of these systems is not an accident evaluated in the USAR. The availability of the 12.5 KV system for mitigation of accidents described in the USAR is not impacted by this change. Also, the ability of the 12.5 KV system to provide plant event reslonse power is not a!Tected. The load being added will not cause any system or equipment to exceed its design capacity. A fuse will be added on the primary side of the new .
transformer to prevent any fault from affecting the remainder of the 12,5 KV system, which prmides l power for equipment defined as important to safety. An Engineenng Jwigement evaluated the load addition on the Gui-Tronics system and found it to be acceptable. Therefore, the probability of occunence or consequences of a malfunction of equipment important to safety is not increased. No Technical Specification margin is afTected.
MP %-132 TITih Emergency Core Cooling Systems (ECCS) Suction Strainers Modification DESCRIPTION: This nnhfication replaced the existing ECCS suction strainers of the Residual Ileat Removal (RilR) and Core Spray (CS) systems with new General Electric high performance stacked disk design strainers. The new strainers are larger capacity passive-type strainers. This modification will improve the etliciency of the ECCS during a postulated Loss of Coolant Accident (LOCA) since tne new strainers will reduce the head loss across the strainers, maximize the pumps' Net Positive Suction llead (NPSII) margin, and allow an increase in allowable debris loading during a LOCA. This modification was implemented in response to NRC IE Bulletin W'3.
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SAFETY ANALYSIS. This modification will have no direct or 'ndirect impact on the probability of occurrence of an accident previously evaluated in the SAR because this type of strainer is not an accideat initiator and the 'rainer itself is a passive component. The new strainers will improve the etliciency of the ECCS during a -
postulated LOCA. They meet the same system requirements as required for the ECCS design and do not a fTect CS and RIIR system operating characteristics. The new R1IR and CS strainers have been analyzed in regard to debris loading, sludge loading, tmus shell dynamic loading and the hydrodynamic loadmg experienced on the tcuus,ile penetrations, and attached piping. The calculated stresses _in structures and penetratims meet code alkmable limits. The strainer replacement will further minimize the potential for a ek>gging malfunction of the ECCS because thotrainers are passive components and require no operator intervention the tvw stminers reduce head kus acrow the loaded surfaces, maximize the NPSII margin, arxl allow fu nxxe debris and sludge k>admg during a postulated LOCA. Thus, this activity will improve the margin of safety. Deaig strainer removal, the plant will be shutdown; therefore, no hydrodynamic loadtng conditions eun occur, which is the significant factor in design basis loading conditions. Technical Specification requirements of one operable R1IR pump and one operable CS pump will be met. Under the corxhtims of a design seismic ewnt with strainer removed, loads on the torus are bounded by existing analysis.
MP %134 TrfLE: Fuel Prep Machine Upper stop DESCRIPTION: This hkxlification Package adjusted the upper stop on the fuci prep machine to maintain the Technical Specification limit of 8 % feet of wate. above the top of fuel.
SAFETY ANALYSIS: This MP only changes the location of the upper stop on the fuel prep machine. It hm etTect on any sptem or component which can initiate an accident or is relied upon to rmtigate the consequences of an accident. The fuel prep machine is not relied upon to mitigate design basis or special events nor is it relied upon for post accidet,! monitoring or mitigating the consequences of a malfunction of equipment important to safety. This MP does not afTect any other equipment. This modification is being implemented to maintain comphance with Technical Specification 310 C. No other Technical Specification or bases are afTected by this MP.
- TITIE Cellular Phone and Dosimetry Installation DESCRIPTION: his MP was implemented to enhance existmg conununications in the plant through the installation of a cellular telephone system. It also provided a means of routing a computer cable to be used by teledosimetry units to feed information from inside containment to the Radiological Protection control station This caole was routed with communications cable through an existing conduit penetrating Secm dary Containment. An amendment to this MP was processed to change the type of post-modification testing to be done for the cellular phone installation.
SAFETY ANALYSIS: There are r.o accident initiators associated with th4 modification. All components requiring mounting to Category I structures are designed as Seismic Class I and will not interact with existing plant systems, c,mponents, or structures during a design basis carthquake (DBE). No process or electrical conditions or equipment fimetions will be changed as a result of this modification. All safety related equipment will be able to perform its intended function in mitigating accidents. No equipment operating parameters are changed- The ordy credible malfunction would occur if the mounted components became dislodged during a DBE and struck equipment important to safety. This is unlikely to occur due to the seismic characteristics of the component mountmg. The light weight of the base stations makes it unlikely that any damage would occur if they became dAlged. An existing conduit containing communications wiring that penetrates Secondary Cantair nent will be breached to facihtate the new commtmications cable. Breaching of Secoixlary Contairunent will occur during an outage. Secondary Containment seals arxt fire seak by design will maintain the integrity of the walls that are penetrated. The consequences of
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equipment failure remain unchanged by this installation This installation does not interface with any existing plant systems except awer sunplied by convenience receptacles. Any fault would be cleared locally at the distribution panel. Therefore, the margin of safety defined in the basis for any Technical Specification is una!Tected.
MP 97-001 '
TITLE: . Reload 17 Analysis for Cycle 18 DESCRIPTION: The analysis and design review documented in this MP addresses the reload core changes associated with Reload 17 fuel bundles for Cycle 18 operation at CNS. The prpose of the Safety Analysis is to review and document the results obtained from the reload licensing and safety analysis perfomied by General Ekctric fa Rekmd 17 fuel bundles Reload 17 new fuel bundles are Qc same design type as the Reload 16 fuel bundles. This MP does not modify any systems or components and there is no work or maintenance activity associated with this MP.
SAFETY 4.NALYSIS: The reload core and fuel designs have a direct impact on the plant response to a number of transients and uccidents evaluated in the SAR. The reload safety analysis performed by General Electric for Cycle 18 with Reload 17 fuel bundles is based on the safety enteria established in the General Electric Standard Application fa Reactor Fuel (GESTAR) for the specified abnormal operational occurrences and design basis accidents of SAR Section XIV. Therefore, this MP does not increase the probability of an accident previously evaluated in the SAR. The Reload 17 fuel bundles and the associated core changes do not adversely affect the integnty of the fuel cladding, reactor press .re vessel, or reactor coolant system. Core operating limits have been established in the reload licensing analysis to ensure wiat the safety limit critical power ratio is not exceeded. The Reload 17 licensing analysis of the limiting overp essurization event shows that the peak calculated pressures in the reactor pressure vessel and reactor coolant system are less than those allowed by the ASME Boiler and Pressure Vessel Code. It is concluded that the Reload 17 core changes do not adversely affect radiation release barrier integrity and will not increase the radiological consequences of any accident evahtated in the SAR. The Reload 17 core changes do not directly affect any plant safety equipment or safety systems necessary to mitigate the consequences of accidents or transients, or to safely shut dosm the reactor. No physical changes or modifications are being made by this MP. De Cycle 18 core fuel assemblies are passive components, and are not direct accident initiators. The margins of safety potentia"y atTected by the Reload 17 core changes are related to fuel thermal limits and maximum pressure limits in the reactor pressure vessel.and reactor coolant system.
Analyses of limiting SAR transients with Reload 17 core changes have established operating limit 1 Minimum Critical Power Ratio (MCPR) values that ensure that the safety limit MCPR is not exceeded.
Therefore, the margin of safety to fuel cladding failure due to departure from nucleate boiling during transients has not been reduced The Reload !? safety analysis performed for the Loss of Coolant Accident (LOCA) meets the limits of 10CFR50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors. Therefore, the consequences of a LOCA are not increased and the margin of safety is not reduced. The margin of safety between the ASME Code safety h, nit and the actual failure point of the reactor pressure vessel and reactor coolant system piping is not rco ed. The Cycle 18 fuel load will result in plant operation in accordance with the nuclear safety l opera kmal requirements as specified in the Technical Specifications, .vith all existing margins of safety maintainco.
EE 97-008 and Rev.1 TITLE: Reroute ofAuxilia:y Condensate Duplex Unit Vent Line DESCRIPTION: This EE provid.d for rerouting of the vent line for the Auxiliary Condensate Duplex Unit in the Ser ice Water (SW) Pump Roont De existing wnt line vented steam to the atmosphere of the SW Pump Room l in the intake Structure below an existing heat detector which caused frequent false indications of a fire l in the SW Pump Room. This EE rerouted the vent line by extending it to a tie-in with an existing equipment drain which leads to a floor drain.
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i e
SAFETY:
JANALYSISL - Ris activity will rxt change the overall design, function, or reliability of the Auxiliary Condensate Drain .
L system The vent line will still vent non radioactive steam to the atmosphere of the SW Pump Room.
This system is nonessential and does not perform any accident mitigation functions; therefore, thic EE .
will rxt result in any increased radiological effects? This activity will not change or affect any essential -
equipment 6 the SW Pump Roorn, rur will it induce any equipment malfunctions or failures. Tit &ies not _
change any existing interfaces between syttems. The design of this modification has properly considered
. scianic, thennal, structural, material, and system interaction concerns This activity does not afTect any assumptions, calculations, procedures, or design specifications used to establish the ' asis for defiaing the plant's margin of safety. The existing margins of safety as defined in the basis for any Technical - <
Specification will remain una'Tected.
I
- EH 97 009
' TITLE: Evaluation ofIIPCI MOV-M015 and llPCI MOV M016 Closing Time DESCRIPTION: This"use as is" Engirm:cring Evaluation estabhshes the acceptability of the present design and establishes
.tn upper limit valve stroke time for meeting the design requirements for containment isolation, internal ,
ikxximg, building pressuraation, and emironmental qualification of electrical equipment including flood -
depths, temperature, and pressure. The 1ligh Energy Line Break (llELB) study per FSAR Arr.:ndment -
25 assumes that breaks in the 8' steam line supplying the Resiant !Icat Removal heat exchangers are isolated by closure of!!PCI MOV M015 nr 1IPCI-MOV Mol6. The existing surveillance procedure specified an operabihty limit for valve closure time of less than or equal to 51 seconds for iIPCI-MOV M015 The FSAR Amendment ilEIE analysis assumed a valve closure time of 40 seconds.
Le results of the analyses perfanned fa this EE show a 50 second valve closure time is acceptable. The correct stroke time of 50 seconds is stated for llPCI-MOV-M016 SAFETY
- ANALYSIS: This activity does not alTect the physical cordiguration of any pla: . m em or coroponent, nr.d does not atreet the manner in which any system is operated. It dies not alTes s precursors to any plant event or equipment malfunction; therefore, it cannot increase the probability of occurrence of a plant event or malfunction ofequipment important to safety previously evaluated in the SAR. The design basis closing time is being clarified as 50 seconds. - The valve continues to meet the containment isolation design requirement. Calculatians haw detemuned that the environmental conditions due to a 50 second closing time do not exceed the values of pressure capacity of any room, temperatute or pressure established by the Equipment Qualification pogram, or the ikxximg lewis established by the Internal Flooding Program.
Since the longer closing time d ies not cause any of the post accident parameters to exceed the values previously established, the consequencea of a previously evaluated plant event are not increased. This change does not intmduce the pusibility of a difTerent type rf plant event or equipment malfunction.
There are no specific Technical Specifications related to the isolation time for this valve. This change does not affect the isolation circuitry or the ability of the valve to perform its Technkal Specification function. Therefore, the margin of safety as delined in the basis of the Technical Specifications is not decread MP 97-014 TITLE: Drywell Closed ircuit Television (CCTV) System DESCRIPTION: This MP pmvided for the installati_on of a CCTV system in the drywell to prm .de Operations with the ability to visually monitor the drywell at elevations 890' and 923', The CCTV iystem consists of three cameras to assist in detennining the source of water draining into the F and 0 s imps.
. ANALYSIS: This activity does not inmtw any equipment or systems credited as event initiatoi s nor does it adversely -
i= _
ntTect any systems credited with terminating transients whose failure could renlt in a plant event. This
' actisity does not atfeet the ability te shut dow1: the reactor or to maintain it in cold si utdowTi and &ies not atTect the plant's d ility to contain radioactive materials either during normal opvation or post event.
Adequate protective features have been included in the design to ensure that any credible failure of the
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I equipment installed by this activity will rxt adversely impact equipment important to safety This activity does not affect any parameters whose margin of safety is addressed in the basis for any Techrucal '
Specification Mp 97-015 TITIJI: Chamfering ofIIPCI MOV M016 Disc, Seat, and Guide DilSCRIPT10N: This maltlication implemental an Electric Powu Research Institute rearnmendation for iIPCI MOV-M016 to chminate the potential Ar unpredictr'. ', stroke behavior (i c., possible inannplete imlation). The out ide diameter cJges of the valve disc and the it ide diameter edges of the valve seat
- wue charnfaal, along with the dac guide skits 1he cinnfaing pmvides a dull edge contact of the valve internals iesulting in a more predmtable stroke behavior to ensure the design basis margin of IIPCI-MOV Mol6 is achieved This nahfication enhances the abihty of Mol6 to isolate during maximtun expected differential pressrc (i c., Iligh linergy 1.ine lhesk cmditions).
SAlliTY ANAL,YSIS: This activity will tot change the overall design, function, or reliability ofi!PCI MOV Mol6 or of the -
Iligh Pressure Cmlant injectim (lIPCI) system. The chamfered intemals will still meet the same system design requirements as bcRec the mahlication The work will be implemented during coki shutdown when Primary Cmtainment is tot regtured Chamfering of the valve intemals will not affect any accident mitigatim functions, nor will it result in any increased radiological efTects. The valve will still function as required upon receiving IIPCI isolation signals; this mahfication only improves the valvo's stroke ,
predictabihty. This activity will not adversely affect any equipment important to safety nor will it introduce any new failure modes. The probability of occt.trence e consequences of a plant event or malfunction of equipment important to safety are not increased This activity does not afTect any auumptims, calculatima, procahires, or design specifications used to establish the basis Re defining the plant's margin of safety The existing margins of safety as defined in the basis kr any Technical Specification will remain unchanged Mp 97-016 T111JI: Z sump loss of Power Annunciation DESCRIPTION: 1his nnhfication prtnides unslusive imlication in the Control Room and kically at the respective panels of sma availabihty tbr the Z sump puu mntrol powet circuits which previously required additional canpematory measures to be performeo tiy the operators to ensure operability, MP 97-016 installed a km-of pmvr tvlay fir the anenuitir.g citetutty to the existing ili-lli lli armunciation circuit which is also a pution of the omtrol circuit y Ibr the Z sump pumps. In addition, the starter circuitry for both pumps was mahlied to indicate un overload trip via a complete loss of contna power. This MP provides unprtnw! acurance that the Z stunp pumps are maintaining levels below the point at which Standby Gas T catment (SOT)is atiected SAFETY
. ANAL,YSIS: l'aihar of the systena, structins, or canpoents (SSCs) bemg nulified by this MP (Z smnp control logic and Z ump starter circuitry) cannot itutiate a phint ewnt. The only potential to iacrease the consequences of a plant event is to cause the Z sump pumps to become inoperable such that water would go beck up into the SGT piping and reduce flow, thereby causing SGT to become inoperable. During MP imtallation, craht is taken for the actions implemented by an Operability Assessment (OA) related to this caution Thew actian include hourly observation of pump control lights, pump operation verification via daily obsavatian of punp run time / event counters, and the redundancy pmvided by the two Z sump I. umps. Since operability / redundancy of both pumps is currently being credited by the OA, the MP rattures that the new has-of pmu relay be imtalloi first. After this portion of the MP is completed, one of the Z sump pumps can be taken out of senice to perfonn the required modificatiott The new relay / light'winno being alled is similar in design to that which is existing ami was purchased as Esser.Lal and installed ps all applicable codes and standard ( The k>ad added to EE PN1 LPOBl(8) was shmm to be negligible by Engineering Judgement. The testing performed by the MP verifics the circuitry operates per design. The annunciation of the kas-of-pmu relay indicates that the 6.7 hout vahie credited 27-I
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-____ _ ___ _, - , - _ , - . . - - - , - - .)
in existing analysis fbr time available to restore power is being maintained If power is hist, the applicable procedures provide adequate direction to opershirs to identify / respond to potential failures.
1heref(re, the acuequences of a plant e mt presiously evaluated in the SAR cannot be increased The SSCs being afkrted by this MP will be maldial with the SSCs tagged out. In this condithm, the tagged out SSC cannot be caused to malfunction. In addition, at least one pump remams operable at all times to avure that the Z sump does not overfill 'the relay / light installed atxt the wiring changes made do rot rnlace the icwl of pretection pmvidal to any SSC. The SSCs afTected by this MP cannot fail in any new w my atal the severity of a malfunction is also not adversely afkrted. There are no margins of safety apphcable to the Z sump starters or the control / annunciator logie defined in the basis for any Technical ;
Specificatmn UH 97-021 TITidt Radwaste fluildmg Unauthorized Moddications Dl!SCIUPTION: Varsus unauduziral nodificnns uvre kkntifialin the Radwaste and Augmented Radwaste llutkhngs. l The modifications made were personnel safety enhancements such an installing enclosures armmd autananatal areas with attaclunents to the flats and walls of the building. Although these nxxhfications are mm aafety related, they should have been te iewed and autixtived by Engineering. This EE
! thicummts the acceptance of the pre iously installed mahlications.
FAMITY ANAL,YSIS: The mmhficatmn compments are non safety related The components could fail without affecting any safetuelatal equinment. llo do not afTect any plant event presiously evaluated in the SAR nor alTect ,
the rathological releases of any plant event evaluated in the SAR. The malificathms do rmt increase the pmbabihty ofoccunence tv ansequences of a malfunction of att, equipment important to safety. There ,
are to margms of safety specified in the Tecimical Specificat:nns which are related to these modifications.
hE2Bul TITidt Motir Opetutal Vahe (MOV) Upgnales fbr !!PCI MOT M015 and Chamfering ofIIPCI Mol5 Disc, Seat, and Guide DESCRIPTION: This mahlication replaced the 25 filb motor installed on IIPCI.MO Mol5 with an equivalently quahfied 40 fbib motor to increase the torque output of the actuator rad to ensure IIPCI MOV-MOIS will meet Omenc letter 8910 chwure requirements ibliowing tlw issuance of the impending 1.imitorque revisal motor sizing criteria (refererne NRC Infbnnation Notice 96 48) In addition, the protective thermal overload relay heaters were replaced to accommodate the larger motor. This MP also implemental an Ekttric l\mur Research Insutute recommendation fbr iIPCI MOV Mol5 to climinate the lutential Ibr unpredictable stroke behavkir (i c , incomplete isolation). The outside diameter u! gas of the vahr dise and tiie imide diameter edges of the valve seat rings were chamfered, ahmg with tie disc guide slots. This chamfering pmvides a dull edge contact of the valve internals resulting in a more predictable stroke behavior.
SAFliTY ANAL.YSIS: The svplacement motor and overload heaters will meet the same system design requirements and safety classifications as the existing notor and overload heaters The control logic for llPCI Mol5 is not aheralby this mahficathn 1hc notor. heater replac ment does not alter operating conditions or system parameters or introduce any new faihire males. The modification will be performed during cold shutthmn with the iligh Pressure Coolant hvetion (iIPCI) system out of service. This MP does not alter the function or configuratkm of any compenents, equipmen*, stnicture, or systems required to mitigate
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postulated accidents or required to safely achieve safe shutdown. The modification will increase t(rqueAhn4 margin availaNe for the valve and actuator to perform its safety function Chamfering skies not aher the design bases of MOI S and will improve its reliability durini: the perfbrmance ofits safety design basis functions. This activity will not change the overall design, function, or reliabinty ofilPCI-MOV M015 or of the llPCI system. The chamfered internals will still meet the same system design sequirements as before the mahlication. Chamfenng of the valve intemals will not affect any accident nutigatko functions, nor will it result in any increased radiological efTects. The valve will still function 28
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f r
as required upon receising IIPCI isolation signals, this mcalication only improves the valve's stroke [
predictability, Chamfering will not adscrsely affect any equipment im;xrtant to safety not wdl it ;
introduce any new Iailure modes. %e probability of occurren.c or consequences of a plant event or !
malfunction of equipment important to safety are not increased. This activity dies not affect any assunptims, calculatims, pmeeduresar design specifications used to establish the basis for derming the plant's margin of safety. %e existmg margms of rufety as defined in the basis for any Technical Specification will remain unchanged
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MP 97-036A TITI.E: Installation of Auxilia7 elays R and l'uses to Alleviate Distribution System Undervohage Concerns DESCRIPTION: his nudricatim irovided for the installation of two auxiliary relays to resolve 120 VAC safety related contml pown margmal mitage problems identified by an Engineering analysis. Two existing fast timed fuses were relacul with delayed time fuses to ensure the ability of the circuits to have available control power aller a degraded voltage condition.
sal'ETY .
ANAL.YSIS: ne component circuit changes asociated with this MP will be perfonned on a divisionel basis during cold shutamu cmdttions After completion, all components will appear to function exactly as they did prior to the nxafication. Testing of the indisidual associated components will not result in a function difTnent than any already analyeal lir in the SAR. Therefore, there is no ine* case in the frequency of any 1 plant event previously evaluated in the SAH The modification provides no new release pathways and
- does ruit aher the function or performance r "any event mitigation system. The new auxiliary relays and ;
fuses meet r excml the same starxlard rd requirements as the onginally installed devices Testing will validate that the nudficatam does ot afket the operation of the pump motors, nor afrect the Appendix R islations ne nxdfication does not alter the funetion of the Reactor Equipment Cooling (REC) pump motor control circuitry, nor dies the installation and testing provide any new failure modes. The nnhfication will be installed and tested prior to plant startup from RE l 7 and will not alTect the ability of the REC pumps nor the diesel generator exhaust fans to provide their safety related functions aller installation as the pumps and fans design characteristics are not ahered. The equipment function will not be changed and the same operabihty requirements remain aller implementatioa Therefore, the margin of safety as dermed in the basis for any Tec'.nical Specification is not reduced.
MP 97-039 TillR Thermal Overpressure Protection for Containtnent Penetrations X 18, X 19, and X 20 DESCRIPTION: An evaluation determined that the demineraheal water lux (penetration X-20) and the drywell equipment and floor drain sump discharge lines (penetratmns X 18 and X 19) were susceptible to thermal overpressuritation under a 1.oss of Coolant Accident condition in the drywell. The d ywell floor drain arxl drywell equipment drain systems were mahfied by the addition of a relief valve inside containment in the non safety portion of the syatems to provnic overpressure protection A nmdification to the deminerali/ed water system rehicated the primary Containment lxxindary valve and installed a manual isolation valve, a drain valve, and a spectacle flange upstream of the new boundary valve to provide a means of assuring that the deminerab/ed water ime remains drained during formal operation, thereby precludmg the possibility of thennal overpressurization in this line.
sal'ETY ANALYSIS: The malification to the daywell sump lines will allow trapped water to expand without adversely impacting the piping. 1he mahficution to the dcmineralized water line will ensure the line remains drained dunng nomial operation, thus precluing the possibility of thermal overpressurization. The matmals umlin the nulification will meet the emTent piping cale requirements and future testing will be in necordance with the appropriate codes %c mahlications do rot alTect the function of the systems and do not affect the manner in which the systems are operated. %e malificatians do not afTect any precurars to plant events or eqmpment malfunctions, arid will not increase the probability of occmTence of a plant event or equipment malfunction evaluated in the SAR. The drywell equipment sump arxi drywell flar drain sumps do not perform any essential function in respome to plant events evaluated in 29-
t the Salt The demineralieed water line has no safety function other than the containment isolation furetxn, armi(kcs rx4 interface with other safety systems.1his MP will not afTect the severity of a plant ewnt and willrut &ct tic :bibty of any plau system tolufism its required wtion m response to a plant ewnt, therefire, tlw ancituus of a plant event will not be increased. This enange does not create the pubihty of a n w release teint fnun the containment and does tu afTect the ability of the containment isolation system to perform its intended function. Although new valves were noded to various piping s>*ms, no new operational modes are ercated There is no specific Technical Specification related to the operation of the sumps. This change will not affect the ability of the plant to meet the containtnent integnty arul kcalleak rate test requirements specified in Technical Specifications and will not afTec the coolant leakage rates or the capability of the sump flow measurmg systems emered by Tects.ical Specifications _ There is no Technical Specification related to the operat.on of the demineralizal water system. The margin of safety as dermed in the basis for any Technical Specification is not decreased.
Im 97-0$2 and Rev.1 TITIJ!: FP RV.16RV's DischarFe 1.ine Size increase D1? SCRIPT 10N: This lili replaced FP RV 16RV with a larger 1% inch valve and also installed a new 2 inch discharge kne. The original design called for a i inch relief valve with a i inch dischepr 'ine. This modification will enharce the reliabihty of the Fire Protection (FP) system by assuring adequate mmtmum flow under low flow or dead-head conditions SAFl!TY ANAL,YSIS: This activity only afkets the FP system which has been previously analyred as not being an accident initiat(t by the CNS Fire Ilazards Analysis. It (kes not reduce the abihty of the FP system to respond and deliver the minimum amounts of water to a fire, nor does it alter any operational or maintenance pnoodures asseinted with this system The minimum flow requirements of the fire protection pump are not being reduced Rather, this activity is intendal to increase this minimum flow rate, thus enhancing pump reliability. The consequences of the relier valve, its discharge piping, or even tie main pump ,
i failure ter;inin unchanged This activity does not add new equipment, create any new interfaces, or chanFe the design function of any systems, structures, or compments. It does not alter the normal or emergency design parameters of the fire suppression system, nor does it change its design or amfigurutun, negatmg the possibihty of creating any new failure modes The fire suppression pum;. has sullicient margin to easily withstand the small increase in system losses incurred by the larger relief capacity. This malification withiot reduce any existing margins of plant safety.
hip 97-056 TITIJI: Neutron Monitor Voltage Regulator Replacement ;
DliSCRIPTION: This nnhficatian rephced two voltage regulators iu Neutron Monitor System Flow Units 81 A and Hill with a new moil. The previous models are no longer available as replacement units. This MP also changed documentation to allow optional use of eithen an ohl model voltage regulator or new model voltage regulator at other h> cations in the Neutron Monitor System.
SAFl!TY ANALYSIS: 'lhe equipment containmg the subjcot ampments does not initiate any previously evaluated plant events.
'the replacement voltage regulators are the vendor's recommended current replacement matels for use in the Neutnm Mautor System and perfirm all the functions u hich were perfonned by the original units.
The replacement of the voltage regulators will have no efTect on radiation dose to the public. The eqmpment being mahlied is important to safety; however, this MP does not increase the probability of occuntnee or ansequences of a malfunction of equipment important to safety. The mt.rgin of safety as dermed in the Technical Specifications is not reduced by the replacement of the voltage regulators.
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MP 9748
! l Tilld!: Ihgh Point Venting of 48' Air RemovalIloldup Drrin Line DliSCR!l' TION: This rnodificathm was implemented as a result of a pr# , Anth.% h% e 4 n Air ;
Removal hn' iup Ime to the Z sump. Unan@ipated ,a a o8 5 e- . e y ; z:L.ntified '
which, dunts kiss of ofTsite power, haw the raientd 4 s o;W / - 4 - Treatment (SOT) lues by baking up the SOT drain lines that are ah w eM s e e a ' r assure differential between the holdup line and Z sump slows or prtvesa a Mi:n 4 N% em the holdup line ;
- to the Z sump during normal operation of AugmenteM%,@0). Changes in plant operation or opa ation of the A00 system reduces the ditTntentini pressure between the holdup Ime and the Z sump alkming the tullatal water in the holdup Ime to drain into the Z sump. This matification added a high point vent fit the 48' Air Renoval holdup drain line, two new penetrations with isolation valves through the X sump cover, and an additional isolation valve in the Off Oas lluilding. An isolation vahc and 5 tanp vanly capped sectkm of stainless steel tubiog installed outside the sump were to be used at a lat er time to conmd a vent hne to the ADO allerfilter fit purposes of equalizing the prtusures between the 48' holdup hne arx! the drain smup hiop seal hicated in Z stimp to ensure steady drainage at the holdup line.
In addition, wiring and a fuse bhick were installed in the OIT Oas lluilding in preparation fir future solenoid operated valve (SOV) connections tux!ct MP 97 068A. Ilowever, MP 9748A eventually abandoned the SOV concept.
sal 1?TY ANAL YSIS: This mahlication thic not afTect any system, equipment, or component uhich could be an accident or transient initiator. The Z sump will fulfill its safety function in support of the operation of the SGT system. 1he W line installed inside the Z sump is nonessential and rmn seismic which is the same classincation as the 2" drain line to w hich it is connected The essential class Ocation and testing of the isolation vahrs on the new penetratun ensures Out the valves will fu' fill their functmn to serve as passive pressure landaries for the Z sump. The ramhfication does not afIect the operation of the essential equipment in the Z sump arx! no new failure mechanisms are introduced The accident analysis takes no credit for the equipment, both electrical and mechanical, in the Off Oas I?uilding. The portions of the unhfication in the OIT Uns fluildmg have nonessential functions and do not interface with any essential equipnent. 'lhe previous safety evaluation for DC 95 033, Sump Z Mali 0 cations for SGT Operability, denustrated that the malfunction of essential equipment had no adverse consequences since redundant equipment could fulfJ1 the safety functions The Z sump will perform its safety functions in the same manner as it would prior to this modification. Installation of the modification was performed while the plant was ,n the cold shetdown condition so that Secondary Containment integrity was not required and thus there are no safety wncems associated with this aspect of the modi 0 cation installation MP 97-068A TITI.l!: Z Sump Modt0 cation Phase 11 DliSCRIPTION: This MP was performed to assist in resolving problems aswiciated with the drainage of the 4r Air Renoval toldup hne to the Z sump (reference MP 97-068). The scop of improvements smder this MP w as: 1) installation of a vent line for gas removal at the high point on the 2* drain line into the Z sump for the 48' Air Removal System holdup pipe,2) installation of a sensing line for Z sump pressure between the sump cover an i oft & 'luilding, and 3) installation of monitoring instrumentation in the OIT Gas lluikling to ensure any excessive buildup of hquid in the holdup pipe can be identified and resohyd in a timely nmnner. MP 97-068A was installed betw een existing tie-in connections into the OIT Uus System. Z sump cover, and the 48" hoklup drain piping made under Phase 1 of this project per MP 9748 SAFliTY ANALYSIS. The materials ami components installed by this MP are rated for the temperature and pressure / design requirements of the interconnecting systems. The components mstalled are passive components and are not associated with any of the accident / plant event initiators discussed in the USAR The addition of the clwtrical loads will not impact the operation of any emergency busei The addition of the instrumentation and vent / equalizing lines will not impact the Z sump pumping sysient The instruments, piping, and
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valves that are comwctal to the Z sump atmosphere fonn an extension of the Z sump *a Secmdary Containraent integrity and were installed as " essential" c unponents Control of the "rmrmally closed" valves is under strict control of Operations to ensure that Semndary C(mtainment integrity is not umprmund by deir rnarupulatim No autcanatic functions or alarms were installed or impacted by this )
Mp.1herrlise, the pobability ofmurrence of a plant event is not increami 1his installation does not impact the 30 minute holdup time of the 48' Air Removal line. It does rmt impact the ognation of any plant equipment that is required to operate in order to mitigate the conwquences of an accident. This installaton &cs not prevent the kop neals in the drain hnes frten performing their design functions and aen not impact the functions of the Off Gas or Air Renovel systerns. The equalizing line connected to the 2* disin line on the 4M' holdup line allows air flow to exhaust from the drain line only.1his allows a relcaw path for any gaws that may mne out of solution, but will not allow the Z surnp to take a suction fnun the oft Oas filter inlet piping. Therefore, the probabihty of a previous'y evaluated malfunction of ,
1 eqmpment important to safety is tot increased 1hc installation of Mp 97 068A does ret impact the abi'ity to pump water out of the Z sumps under any conditions. This Mp does not impact any Radiation Morutonng niuipment that is used to mitigate the consequences of an accident. The maintenance of the kop neals asociated with Standby Ons Treatrnent (SOT)is an important to safety function of the Z sump.
Mp 97 06MA does ext impact the abthty to maintain the SOT hiop seals and it does rmt hapact the ability of tie Z sump pumpmg system to prevent u ater from backing up into the kop seals. The installation of this Mp des rx4 i-upact the operation of the 01T Oas filtert 1he function / significance of the Z sump is not defined or discussed in any Technical Specification or basis. 1he description of Se:undary Containnent integnty arxl the requircreents for the maintenance of Seumdary Containment as defined in the Technical Specifications are not impacted. The installation of Mp 97 068A maintains the same current design requirements as the existing Z aump cover. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced ,
IE 97-074 i
TITI.!!: Removal of Insulation on Standby 1.iquid Control (St.C) piping in the Drywell DESCR!pTION: The SI.C piping in the drywell is insulated with calcium silicate insulation,1his insulation must be rewml due to a potential problem of plugging the Residual Ilent Removal and Core Spray strainers in the event of a loss of Coolant Accident. An evaluation was perfonned to determine the acceptabihty of
. removing the calcium silicate insulation and leaving the SI.C piping in the drywell uninsulated it was determined that the insulation is not required and could be removed SAFl?TY ANALYSIS: Removal of the insulation on the St.C piping wdl increase the heat loading in the drywell. Ilowever, the ;
increased heat load has an inconsequential efTect on the total drywell heat load. This change ars not alicet the prveursors to any plant ewnt and does not alTect the abihty of any system to perform its safety function 1.imits previously evaluated in the SAR are not exceeded This change &cs not affect the probability of occurrence or consequences of a malfunction of equipment important to safety since the drywell tenperature will not exceed the specified limit due to this change. The increased heat load will rut afTect the operation or perfonnance of any Equipment Qualification (EQ) equipment in the drywell due to kical temjuaturv ineseaw. The insulation on the St.C piping has no elTect on the operation of the SLC system Since this change &cs not cause the drywell temperature to exceed specified limits, there is no reduction in the margia of safety as defined in the basis for any Technical Specificatioti im 97-075 TITili: Diesel Generator (DO) Main llearing 1 ube Oil Connectors DESCRIPTION: While perfomung enaintenance on DO #1 during REl7, pipe couplings were found between the lube oil ties inw and the nuun Ivarings which are not shown on the applicable Cooper Bessemer drawing. This EH allows the notn ued uw of the existing fittmgs or removal of the fittings, as either configuration was detamined to be awcptaNe.1he albtimal fittings were installed during original manufacture. 1his EE applies to teth DO #1 and DO #2.
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SAll!TY . _ l ANAL.YSIS: 1he opsating e uxhtuns, system parametern, and control logic of the Dos will ixt be changed No new failure modes will be intialuced and the probability of an accident will txt be increased. -The undocumented pipe fittings will not alter the function or cmfiguration of the DUs or any related equipment, structure, compuwnts, or systems required to mitigate postulated accidents (e required to achiew safe slaitamn The fittings were installed by the DG mamifacturer and were part of the original equipment. The DGs still meet the same design requirements and their ability to gwrform design basis safety related functions is not afTected.
I?E 97-087 ,
TIT 111: Steam Relief Valve Support Repair in Drywell DESCRIPTION: While perfisming a walkd mn in the drywell, it was documented that the upper radial beams at elevation 921%10* were attached to the drywellliner using a lubrite plate on a beam set with slotted holes. This :
usux ction is desigms! to allow for difTerential movement between the line and the radial beams. It was iawfied that neveral Steam Relief Valve Discharge 1.ine pipe supputs were rigidly attached to both the radial beam and the drywell liner, thereby restricCng thermal and pressure mduced movements and pdentially increasing stresses in um supputs and attachment points. A detailed resiew of these supports revealed duit the <vigtnal design did ixd account fir the difTerential movement when detenni' ting stresses arnlicactims All subject suppsts were evaluated and it was concluded repairs were required. This EE authorval minor weld repairs to supports VR il 55A/59A and VR ll-47A/51 A to ensure that the support weld stresses will remain within code design allowable stress limits and emure the supports perform their intended function SAFUTY ANAL.YSIS: The function of the Steam Relief Valve supports is to provide adequate support and restraint to the radioactive vent piping in the event of Steam Relief Valve actuation and subsequent closure during the ocemtence of a plant event The Nuclear Pressure Relief System is directly affected by the Steam Relief Valve discharge piping in that the vacuum breakers and piping prevent excessive thxxiing in the event of Steam Relief Valve actuation The repairs to supputs VR ll 55A/59A and VR ll 47A/51 A will
- ensure the support weld stresses remain within code allowable design stresses and consequently will ensure the integrity of the Steam Relief Valw piping in the event of an accident that requires Steam Relief Valve actuation Therefore, since the design basis for the supputs will be maintained and the supports will remum functional,the repair wo:L will not increase the probability of occurrence of a plant event or ,
equipment malfunction Since the Steam Relief Valve system will remain functional in the event of actuation and closure in a plant ewnt, the consequences of a plant ewnt evaluated in the SAR ill not increase. No other systems, components, or equipment important to safety are alTected by the implennitatim of the subject repair, Since the repairs will ensure the supports and support annponents remain within cale allowable stresses and perfonn their intended function, any margin of safety associated with the Steam Relief Valves or Nuclear Pressure Relief System m the Technical Specifications will not be affected.
EE 97-000 TITili: Reacto: Vessel Metal Temperature Tenywrature Element Repair DESCRIPTION: Reactor wssel metal temperature element N111 TE-69111 provides erratic operation during normal operatim lherefive, this EH utilves spare temperatme element Nil! TE-69112 to provide indication of reactor wssel metal temperature to Nlll TR-89, Reactor Vessel Metal and Flange Temperature Recorder, aux! to the Plant Management Information System (PMIS). No changes are required within the drywell or at the thennoeouples, only wiring changes within Tenninal llox 632 are required to facihtate implementation This EH was implemented during plant shutdown SAFETY ANAL,YSIS: The use of the spare thermocouple to provide input to the recorder and PMIS will act increase the probability of occurrence or consequences of a plant event previously evaluated since these themxumples pmide indication only, are not an event imtiator, and prmide no safety related function.
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'lhe use of the spare llamocouple will provide an indication that will be reliable and accurate while the current themmoouple is erratic and unrchuble at rated pressure and temperature. The thermocouples uwd to monitor reactor vessel metal temperature are passive c4vuponents. The failure mode of the thermocouple will not change; the spare is identical in every way to the original thermocouple and prmides the same nusutanng parameter for vencel temperature. The use of the spare themiocouple will I not reduce the margin of safety as defined in the basis for any Technical Specification The original design is not changmg,just the hication (azimuth) of where the temperature is bemg sensed.
l M.92:lfl Evaluation of RWCU MOV-Mol$ and RWCU MOV MolH Closing Time l TIT!Ji:
DESCRIPTION: This lie established the acceptability of the prescat design and established an upper limit valve stroke )
time for snecting the design requirements for coo!ainment isohtton, internal flooding, building prepuriration, and envimnmental qualification of electrical equipment inchiding flood depths, temperature, and pressure.1his EE corrects the Reactor Water Cleanup (RWCU) liigh Energy 1 ine lireak (!!ELII) analysis due to incorrect valve closure time and incorrect signal delay time used in the original RWCU IIELil analysis. The !!ELil design basis closing time for piping downstream of RWCU MOV Mol$ and RWCU MOV Mol8 is being clarified as 46 seconds. This includes a 30 we4md valve stroke time and a 16 sec4md signal delay.
SAITTY ANALYSIS: 1his change does not afrect the physical configuration of any plant system or cornponent, and does not afTect the manner in which any s>s'em is operated 1he valve continues to meet the containment isolation design requirements and also the lilitil requirements. Calcul.tioni show that the environmental conditions due to the revised closing time do not exceed the values of pressure capacity of any reom, temperature or pressure established by the equipment qualification program, or the flooding levels established by the internal flooding program Since it does not afrect the precursors to any plant event or malfunction of equipment important to safety and does not cause the post accident parameters to exuul the values pres iously established by the programs and documented in the USAR and/or DesiFn Criteria Dneuments, the probabihty of occurrence or consequences of a plant event or equipment malfunction previously evaluated in the SAR are not increased. The powibility of a plant event or equipment malftmetion of a difTerent type than any previously evaluated in the SAR is not created Technical Specifications require that valves RWCU-MO 15 and RWCU MO 18 have a maximum operating time of 60 seconds in order to ensure the core is not uncovered. This change does not alTect the isolation circuitry or the abihty of the valve to perform its Technical Specification or containment isolation functiort This valve still meets all Technical Specification functions and the margin of safety as defined in the bagis of any Technical Specificatmn is not decreased EH 97-108 TITI.li: Evaluation of RCIC MOV Mols and RCIC MOV M016 Closing Time DESCRIPTION: This EE estabhshed the acceptahihty of the present design and utablished a valve closing time, including delay time for diesel generator startmg, for meeting the design requirements fbr intemal limximg, building pressurization,- and ennronmental qualification of electrical equipment includmg flaxi depths, temperature, and pressure, This EE documents the additional analysis that has been performed to accommalate: 1) a longer signal delay time, and 2) a longer stroke time for the isolation valves in the three inch Reactor Core Isolation Cooling (RCIC) steam supply line.
SAFETY ANALYSIS- This change does not atTect the physical configuration of any plant system or component, and does not alket the manner in which any system is operated The valves continue to meet all design requirements.
Calculatims have detennined that the eminmmental conditions due to a longer valve closure time do not emed the values ofpressure capacity of any nuun, tanperature or pressure established by the Equipment Qualification program, or the 11msling levels established by the internal flaxling program. Since the longer closing time does not affect the precursors to any plant event or cause any of the post-accident
. parameters to exceed the values previously estaNished by the programs and skicumented in the USAR
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arx! Air Design Criteria Documents, the probabihty of occurrence or consequences of a plant event or rnalfunction of equipment imprtant to safety presiously evaluated in the SAR are rmt increased. De pumbihty of a plant event or equipment malfunction of a different type than any previously evaluated in the SAR is rat created his Eli cnsures that the existing I Lgh Energy Line Ilreak calculations bound the time that valve RCIC-MO 15 will close after a line break. nis change does not affect the isolation circuitry or the abihty of the valve to perfbrm its Techmcal Specificution or containment holation functkn Tic valve still mwts all Technical Specification functions and the margin of safety as def'med in the basis for any Technical Specification is not reduced.
EH 97 109 TITI H: 1:quipment Quahfication (EQ) Splices for RilR MOT M02$A DliSCRIPTION: his lili imtalled EQ splices for the motor terminations in RilR-MOT M025A, replacing the existing terminal bhick. The replacement was performed due to the inabihty to verify the qualification of the terminal bhick in this DC motor, SAFliTY ANA13 SIS: N oc of de cquipment invohtx!in this work i2 a credible event initiator. R1IR 25A is the low Pressure Coolant injectan (IPCI) and Shutdown Cooling (SDC) loop A Injection Motor Operated Vahe which has no potential to imtiate an event, The work will be perfamed during a IfCl Division 1 outape.
l'adure of R1IR 25A (includmg failure during perfinnance of this wc k) is of no comequence during the I.PCI Division I outage and the SDC injection aligned through RllR 2$lt With the plant in cold shutdown and 1.PCI Division I outage in effect, no ifCl requirements involving RilR 25A exis!.
!*olathm requirements are satisfied since the valve will be deenergized in the chised position. This work establishes a quahfied anfiguration so that equipment operability and system functions are ensured. This utsk does rut impact any other equipment in the plant. Once the splicing is completed, qualification of the nutor interface is established and operability can be reestablished No credible failure modes exist during the work and no new failure modes are created. There is no reduction in the margin of safety as defiru! in the basis fix any Technical Specification because ifCl Division I will be in outage status azul tic aunpletal wwk will emure R1IR 25A will respond es required to perform its safety related function as designed I.E 97-l1S TIT 1.li: Hvaluation of Jet Pump Sst Screw Gaps DESCRIPTION: Dunng REl7,in Vessel Visual Inspection of the reactor internals revealed gaps between the restrair.er bracket set screws and the inlet mixer pipe for jet pumps #15 and #20. Per the design drawings for the jet punps, the set screw should be in amtact with thejet panp's inlet mixer, it was detennined the screw gaps have existed in their current state since June 1985, Analyses were perfbnned to demomtrate the adequacy of the pumps in their current condition The Sately Resiew for this IIE documents the acceptability of the jet pump set screw gaps basal upon the evaluations performed SAFETY ANAL YSIS: liased on the results of the structural analyses performed for the CNS jet pumps, operation of the jet pumps with set screw paps does not change the overall system function or reliability of the jet ptunp awmbly. De analyses show that design basis stresses remain below allowable limits, thus assuring jet pump integnty. The structural integrity of the jet pumps is not a precursor to any plant event previously analyzed in the SAR. Since jet pump integrity will be maintained during all design basis normal, transient, and accident cornhtiom, the jet pumps meet design and licensing bases in the current configuratkut ne abihty of the jet pump to maintain the two thirds core height floodable volume is not arnpmmisal, arxl any accidents previously evaluated remain valid. Operation of thejet pumps with the set screw gaps pnwent will not increase the consequences of a plant event, nor will it cause any increased nwhological efRvts. It also will not induce any equipment malfunctions or failures. Structural analyses demonstrate that the stmetural integrity of the jet pumps is very similar with or without gaps at the set screw kications fix design basis conditions Operation of thejet pumps with set screw gaps does not affect any plant equipment or systems important to safety, ruir does it alter their accident mitigation 35
i capabihty, change thetr failure nodes, or create any new failure naxles. It does riot riffect ariy i assunptims, calcuMns, pnmlures,(r design specifications used to establish the basis for defming the ,
margins of safety M CNS !!xisting margins of anfety were demmstrated to le met in the structural j analyses perftsmal kr tiejet ptanp assembly. Thus, de existing margins nf safety as dermed in the basis ,
for any Technical Spmfication remain unafrected. ,
!!I!97 133 TillJI: Weld Repair of Moisture Separator "C" Inlet Steam l'ipe Dl! SCRIPT 10N. !Aning the littnimounnion (!!/C) inspection of a Main Steam pipe unnponent, it was discovered that the wall thickness of the piping was at a minimtuu of 0 313' The 36 inch Main Steam piping has a ;
ranninal wall thickness of 0.750' and the li/C acceptable nunimum wall thickness in 0.450; Since the nrasured thickness u as less than the !!/C acaptable nunimum and less than the code allowable (0.316'),
tie piping requiral repair. The wall thirming was healized and confmed to approximately 3" above the expansion joint. This activity involved the installation of weld build-up on the exterior of the thin area of the pipe. l SAFIITY ANAL.YSIS; This activity restores the piping wall thickness to its original required design and meets code j requirements. The line functhm is unaffected and the confiFuration is unchanged Therefore, the ,
probability ofoccenerx:e <r comequences of an accident or malftsiction of equipment important to safety are not increased The piping sepair is performed in accordance with the original supplier's accommendation. This activity does not afket the margin of safety as dermed in the basis for any Technical Specification.
E!! 97 167 4 TITI.!!: RFC R-LRPR07 Chart Drive Recorder (Replacement)
DliSCRIPTION: This !!!!crahtatal an unautixviminahfication which replaced RFC R l.RPR97. This recorder is a two pen recorder that is med to document reactor water level and reactor pressure. The recorder was (wiginally a single speed two pen chart drive necorder. Maintenance Work Request 74 3 185 replaced the single speed chart carriage assembly (recader intemals) with a two speed two pen chart carriage
. assembly. This lil! sk*tennirnt that the imtallation meets the design requirements for this apphcation and is acceptable for continued use.
SAFl!TY '
ANAL.YSIS: 'll.e subject recorder is used as an Operator aid to contintaiusly record reactor water level and reactor
- presmire and almi pnwides alarm inputs to inftsm Operators of"high or low reactor water level" or *high reactor pressure" This recorder does not initiate any automatic controls required to shutdown and
< maintain the reactor in a safe icnditmn or to mitigate tlw comequences of a plant event. Additional Control Room indicators are available to monitor reactor water level and reactor pressure and this recorder is not required to support any Regulatory Guide 1.97 instrumentation requirements. The nurder (kies not perform any safety related functions. The two speed two pen chart carriage assembly in nunier RFC-R IRpR97 mes the same housing as the or;ginc.1 single s;wed chart drive assembly The inputs to the two s;wed recorder are being obtained from the same hications as the inputs to the single spen!tccorder. The installation of the two speed feature has no impact on the annunciater functions of this nunder. The recorder is a nonessential component and the replacement parts were obtained from an original equipment manufacturer. Thus, this installation does not increar, the probability of occunence of a malfunction of equipment imponant to safety. This recorder is not discussed in the Technical Specifications or bases; therefore, this installation will not reduce the margin of safety as
' defined in the basis for any Technical Specification.
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!!!! 971tV) l Ii11E DenriFit Drsinimp Sealllanges l
DESCRIPTION: Maintenance Woik Request 821631 installed flanges on the loop seal drain piping for the Radwaste '
System aum pit.11c flanges wae installat in mkr to gvovide a means for periodic cleaning of the loop scal. This installation cmstituted a Station modification, but was completed without the appnpriate i engineermg &cumentation in place. This Unc As-In IIE was written to evaluate this unlification and authorire Inth the moddication and regmred drawing revisions. The nnhfication was found to be i acceptable fnnn a design standp(Jnt SAFIITY ANAL,YS:S: This activity will not affect system perfonnance and reliabihty or any sy stem interface in any way that could lead to an accident. Systems and equipnent will not be affected such that they are degraded or operated outxiJe their design tv test knuts Ths netivity aes not increase the possibility of operator crmr
<r inkl amplexity to human faeke etexhtions such that the probability of an accident is incicased.1here is no efTect or increase in the radiation dose associated with the plant's response to any accident or equipment malfunction The Radwaste System will retain its design function and the decon pit and its drain piping will also function as designed installation of the flanges, bolts, and gaskets meet the applicable gnaterial aral design specifications far this class of piping and no sei.cnic cmcern is intnxtuced Ikith the &um pit and the dram piping are considered nonessential.1his activity aes not decrease the ecliabibty of a systenur equipment im;xrtant to safety assumed to function in the accident analysis. This nnhfication has ext a!Tected the station's ability to keep radioactivity in elliuents ALARA. 0ms there is .
no reduction in the margin of safety as defined in the baws of One Technical Specifications.
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37
1 EMfr TFMPORARY M01)lFICAT10NS fPTMa) l l
PTM %(6 TITIE. Installation of Off Ons Sample Chamber j DESCRIP110N. This P1M was issued to prwide for the tempvary installation of an oft gas sample chamber between valves (XbV 77 a*xl OG-V 78 in onkt to perform on line fuel integrity m;nitoring. lhe purpose of the morntoring was to 1.elp kicate a fuel claddmg failure.1he sample chamber was removed aner all necessary nxstitoring u as perf smal SAFliTY ANAL,YSIS: There are no accidents in the USAR involving the O!TGas Monitoring System in addition, the sample chamber will nxxt or exem! Oc existing design of the Off Gas Systent 1he installation of the tempirary sample chamber does not aficct tx.nnal operation of the OlT Ons System, nor does it change its original desy.n intent. The OfiOas Monitoring System does not perfmm an accident mitigation function This PTM i as no effect upon the safety design basis of radiation monitoring components descrituiin the USAR.1hc existing equipment will continue to monitor normal system parameters and will not be adveracly alTectul by this installation The consequences of an Off Oas Syrtem component faihire will not be inercasal The installation of the ternpaary sample cha uber will utili/c existing sample connectxus llecause dec vahrs currently exist as sample points, the possibihty of e new accident has txit lurn ercated The installation and removal of the temporary chamber will la controlled so that there is no effect on the nxmitoring capability of equipment normally in service. The margin of safety is not inkml as dw temprary san.ple chamber does not atiect the ability of the oft Gas umnitor to detect fuel pin claddmg failures because normal flow will be inMntained through the existing sample chamber.
PTM %07 TillE: Fire Protection (FP) System 12 DliSCRIPTION: This PTM was issued to disable the "C" llent Actiuted Device (11 A D ) for FP System 12 on the Ttubine Generator to allow automatic operation m the ranammg nye ll. A D. hiops and amid inadvertent system actuation fnen an air leak on the "C" hop ALARA concerns prevented repair of"C" hiop until plant shutdown SAFliTY ANAL,YSIS: This PTM only alTects the automatic operniion of FP System 12 which is utilized to mitigate a Turbine lluikhng fire. Tis probabihty of fue occurrence is unchanged.1he consequences of a Turbine lluilding fire have hun evahtated vrtiniut any supp ession system actuation. The inadvertent actuatior, of System 12 has been evahiated and its urnequences are unchanged. Inadvertent actuation of FP systems in safety related areas and the subsequent malfunction of safety related equipment has already been evaluated.
No ddferent types of failures have been introduced This PTM only affects the probability Cnadvertent suppression system actuation which is decreased FP System 12 is not a Technical Specification suppression system,0 crefwe. the mar gin of safety as defined m the basis for any Technical Specincation is not reducal.
PTM %08 TITIE: Installation of Support Clatap on Ilypass Valve # 1 DIiSWIPTION. 1his PTM installed an enckwurc support clamp on byper valve #1. Two screws are designed to secure the encbure to de contml bhick I knvever, one of the screws was broken and could not be repaired due to Al. ARA concerns The instalkd clamp perfonns the same function as the mounting screw, so the function of the enclosure, control cunponents, and bypass valve uas not altered The screw was rubsequently repaired during RH17 and the clamp was removed.
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i SAWTY Ah LYSIS: 1he probabihty of an axident is not increased due to the enclosure support clamp meeting or exceeding material ard strength requirements Physical shape of the compment will not alter the component or its function The probabihty of steam line breaks or turbine missile generation are not affected The l i
conseqtux:es of vanous accidmts (steam lite bicaks) are unchanged because they credit the MSIVs and steam line flow restrictors. 1hc bypass valve and associated compments are outside Primary ,
Cmtainment ard do rud aflect quatam of tie Main Stemn isolation Valves or steam line flow restnctors. :
The MS Il0V IIVI ard associated piping and compments have no safety design basis. Presence of the PTM will not affect operation of the bypass valve. Dose omsequences of steam line break accidents runnin loundmg and are unaffected. Presence of the clamp cannot cause any other kind of accident or malfunctim lhe logic of the potective features described in the Technical Spaifications fbr steam line Imk nutigation have not been altered by this PTM. The existing margins of safety remain unchanged.
11e MS llOV-IIVI ord anciated piping can control cooldown rates but are not relied upon for nuclear safety. Malfunctions of the turbine bypass system and the efTects of such failure on other components are evaluated in Section XIV of the USAR. In addition, pnicedures are in place which prmide guidance for controlling cooldown rates in the event the rates cannot be controlled utilizing bypass valves.
PTM 9610 TillJ1: Removal of Contsol lud Drive (CRD) Pump ilatch Cover DisSCRIPTION: This PTM suttuiri/cd the removal of the hatch cover on the 903' level of tb Reactor fluilding over the CRD pumps in order to facilitaic removal of a CRD Pump for repair.
SAFl!TY ANAL,YSIS: 1he only essential equipment which the plug renxwal could afrect is Core Spray Pump !! in the Southeast Quad. The efTect on limergency Core Cooling System (!!CCS) equipment functions as a result of the removal of a plug on the 903' level was previously analyzed when Design Change 93-062 was implemented to permanently remove the plugs above the Northwest and Southwest Quads which house the Resi' taliIcat Renxwal(R)1R) pumps. The one exception with this PTM is the need to consider the effect of thxdmg due to the 18' feedwater line break. A dam was not etccted to prevent flooding; however,it was determined that a dam was not required Ibr the !!CCS system to maintain the plant in a safe shutdown condition fbilowing a break in the fecl water hne in the steam tunnel. Assuming a loss of OITsite Power (1,00P), in cordunction with a feedwater line bre k in the steam tunnel, ard the worst pssible single faihire of the Division i diesel, results in a total I!CCS capability of the two Division 11 R1[R pumps lhe mitigation sequerxx for the less of Feedwater i! vent, which has the same cfTect on the reactor ave as the feedwater line break, indicates that a successful sequence is for one RIIR pump to be in tic Tmmohng nude and the other R1IR pump to be in the injection mode. Thus the configuration where the plug is rmNd fnun the 903' level above the Southeast quad is within the CNS design basis.
Since Core Spray Pmnp til is available to fulfill its safety function for those events for which it is required,it is considered operable during the time period of this PTM. This PTM does not increase the pmbabihty of an accident since the work des not increase the probability of a LOOP, a Loss of Shutdown Cooling, nor an inadvertent IICCS initiation because there is no interaction with any of these systems There is uo increase in the cont races of an accident or transient since there are suflicient I!CCS pumps available to ensure safe shmoowlt This activity does not violate any Technical Specifications since all safety equipment would be available when required by Technical Specifications.
A tevision to the above Safety livalur. ion was subsequer.tly issued u hen a concern was raised that the original Safety I! valuation did not address Technical Specification 3.5 AA, which allows a 30 day 1.imiting Condition for Operation if one R1IR pmnp is inoperable. The revised evaluation showed that there is ample time available for the operator to maintain core coolmg and tor us cooling even with only one RilR pump available. (The revised Safety livaluation also pertains to PTM 95 37 which implemented a similar temponny modtlication in 1995)
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l PTM %11 and PThM616 1111JI: Installathm of Tempvary Pipe Supput for Control Rod Drive (CRD) Pump liquahrer 1.ine DliSCRIPTION: 1his P1hiinwalled a tempvary pipe suppet ibt the equaliang line betueen the A and 11 CRD pumps The hne was uncoupled from the A CRD pump, w hich was replaced.1he remaining pipmg, cortaccted to the il CRD pump, requiral tempor ary support.
sal 1 STY ANAL.YSIS: 1hc nnhfication was required to support the piping connected to tic 11 CRD Pump to prevent the pipe from failie g it can only afTect the 11 CRD pump. No fadure of this pump can cause or increase the probabdity of an accident prniously evaluated in the USAR While CRD flow is mentiotal in some accident scenarios as a contnbutor to reactor vessel iriventory, the CRD pump is not usal to mitigate the consequences of an accident prniously evaluated in the USAR. The new suppvt providee equivalent support to thet which was supphed by the removed CRD pump. The pipe supput or its failure canrot l create an accident.1he design of the CRD System anticipates the loss of CRD pump pressurc/ilow by prmidmg fte accurnulators or reactor pressure for the scram function 1he CRD system is not required or rehed upon to mitigate systern leakage. i P1M 961$
TITLl! 1.cak Repair of Main Tuibine Stop Valve Number One OliSCRIPTION- 1his PTM authori/cd the iryation of leak repair compiund as a temsvary repair of the main stop valve number one bonnet steam leak. A s;weial leak repair funge and leak repair imnnet nuts were installed around the existing lonnet which were injected with leak repair compound SAITTY ANAL,YSIS: 1he probabihty of an accident is not incicased due to the lerk depair Gange, nuts, and scalant meeting or exceedmg the origir.al piping material and strength requirements Leak scalant is compatible with the Main Steam piping innterial, m the probability of steam line breaks is not affected The consequences of various accidents (steam hne breaks) are unchanged because they credit the Main Steam isolation
- Vahrs (MSIW) and steam ime flow restrictors The stop valve and associated piping is outside Primary Containment and dies not afTect operatim of the MSIVs or stearn line flow restnctors The stop valve and associated piping have no safety tiesign basis Presence of the PTM will not afTect operation of the drain valve. Dose consequences of steam line break accidents remain bounding and unaffected Materials ofconstruction and the leak acalant utenal are compatible with the pressure loundary The logic of the protective features described in the Technical Specifications for steam line break imtigation have not been altered by this PTM. The seJarit material has been selected to ensure that the coolant chemistry requirements of the Technical Specifications are not excenkd when propenly utilized The esisting margin of safety remains tmchanged P1 M %20 and P1 M %21 TITI.!!: Disabling of Reactor Rtrirculatum Moks Omerator (RRMO) Scoop Tebe Positioner Internal lilectrical Stop Dl?SCRIPTION 1hese PTMs disabksithe RRMe scoop tube mtemal electrical stop L2 UP on RRMO Sets A and B. This stop pwtite hmital kwal core flow to approximately 98% of rated, thereby scatneting power operation 1he !.S-6 extemal electrical stop will provide the required semp tube stop function l&6 is set to limit maximum core flow to less than 102.5% of rated, as required by the Core Operating Limit Report and
. the plant design transient analysis for internal pressure drop. The P1Ms were removed ad the stops secalibrated in Rill 7.
SAll!TY ANAL.YSIS: The RRMO sets haw tluce redtniant senp tube stops, two electncal and one mechanical The transient analysis assumes only one, set to limit total core flow to 102.5% The L2 UP electrical stop, whether b3passalor not,is rxulesignal to prochide an accident. Its only hu,ction is to stop scoop tube positioner nxtim at a preset point. The !.S-6 esternal electrical stop will provide un equivalent core flow limiting 40-
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i capabihty as the L2 UP intemal electrical stop.1he semy tube stops do not interact, directly or !
irahrectly, with any aanpment dint is a radmactive material barrier. The LS 6 external stop will prmide tLe ruluiral hmit on are Ikyw to maintain stresses on vessel intemalr. during upset umulitions to less than their design values.1he stops, whether disabled or not, cannot cause a reactivity insertion or loss of cmhng. The recirculation pump trip furx:txo is unaffected and is available to prevent pump damage that auki result in has ofomlant invent ny and release of radioactivity to the Primary Containment.1he L2 UP stop, whetimr disabled or not, is tot relied upon to nutigate the consequences of any accident previously evaluated in the USAR and does not interact with any equipment required to mitigate the i unsequences of an accident The L2 UP stop does tot interact uith any other equipment in such a way as ki cither cause er prevent a malfunction of equipment important to safety. 1hc scoop tube p>sitioner is txt irnputarit to safety arid its riialfuriction has not beeri previously evaluated in the USAR. Using the external electrical stop LS-6 to limit total core flow can res.ilt in scoop tube lockout, inhibiting recirculaton pump naback. If there were a feedwater pump trip uhile the scoop tube is kcked out, the !
recirculation pumps would not runback as designed. A reactor scram on low water level would occur, This transient has been previounty evaluated with acceptable consequences without credit for runback.
11e saxy tube stops are dnetasalirxhrectly m the bases of Techrucal Specifications 3 11.A and 3.1 l .C.
1hc purpose of the st ps is to Imut total core flow and power in case of a recirculation system runout that is too slow to be termmated by an Average Power Range Monitor high power reactor scram.1he LS-6 st ps are set to hrrut kital are ik>w toless than 102.5% of rated. The transient analysis does not assume, rug is it requiral to asstune, the faihtte of the electrical stop (a redundant stop is not required). Limiting coie ;10w to 102.5% of rated in order to hmit stresses in vessel intcmals during upset cc iditions is not a Technical Specification requirement.
PTM 96 23 T111JL Replacement Valve for DO AOV Mll2 DESCRIPTION; 1his PTM alkiws the temprary installation of a rephecment valve for DO-A0V MH2, Diesel Generator Na 2 Muf11er Hypass Valve. The replacement valve has a dnllirig pattern for a 25 lb. flange instead of a 125 lb.11enge (original design), preventmg the installation of 4 of the 28 tolts. A Technical livaluation was perfismed that detennined that the remaining telts ensure adequate structural intepity. This PTM will remain in place until the final disposition can be implem:nted SAFliTY ANALYSIS: 1he eflict of the Ibur missing luhs is msigruficant because the reduction in joint strength is less than 15%
arxl thex jomis ski not suppet any significant pressure loadmg. These aspects ensure a suflicient margin of strength remains for this cmfiguratiort The valve replacement does not increase the probability of occunence or consequences of a previously evaluated accident. The non-standard flange imhing arranganent does ruit afliet the functimality, operating parameters, or integrity of the valve, and thus does not increase the probability of occurrence or consequences of a malfunction of DG AOV-MB2 or any other equipurnt, Installatim of the replacement vahr witii a difTerent bolting arrangement will not create any new faihre nules, rur increase the prubabihty of existing failure modes. lhe integnty, functionality, auxlichabihty of the Diesel Generatas remain unchanged, negating the possibihty of any decrease in any specified or unspecified margins of safety.
PTM o6 24 T11ML Drain Line lietween Ser ice Air (SA) Moisture Separator and Trap DESCRIPTION: llus PTM installed a drain line between SA Moisture Separator SA-MSEP-A and trap SA TP ACDA to provide an interim solution to moisture separator SA MSEP A drain line pluggmg and to allow bkmkiwu of the drain line without losing drain trap SA TP-ACDA seal The P1 M replaced an cibow between SA MSEP A arxl SA V 49 with a tee and installed additional piping and a ball valve.
sal'HTY -
ANALYSIS: W tanprary nxoficatim will be anstruetal to the design standards of the SA system The SA system is not a contnbutor to any of the accidents or transients described in the USAR and does not perfe,nn an accident mitigatwo functiort The PTM (Lies not alTect the normal operation of the SA system, nor does 41
i at change its onginal design intent. Control of the new valve is established to ensure that nonnal system [
- operation is not affected The adition of this temprary mahfication does init present any ditTerent falure nudes for the SA systnn Malfunction of the SA system is not evaluated in the USAR.1he SA
sptem is tot described in the Technical Specifications.
p1M %25 TITLE Installation of Test Equipment on timerpency Transformer Undervohape Relay !
l DESCRIPTDN: This PTM authorimi the connection of test equipment to EE REL-(27 ET3) 1his relay is the .
Emergency Transfonner second level undervohnge relay.1his undervohage relay previously actuated !
for unknown reawns The test equipment was installed to mortitor the incoming signal for potential tr ansients that could have caumi the original problem ,
SAFETY ANAL,YSIS: Installing the test equipment will not affect any of the factors that contribute to the probability of an accident since the Emergency Transfonner F llus, and the potential transfonner that supplies the relay are not accident initiators.1he operation of the Emergency Transformer will not be ahered and it will be able to perfirm its intended function 1his will be assured by isolating the test equipment with fuses that will open prior to it alTecting the existing circuit in the event of a failure of the nonessential test aluipnuit. The fuse bkwk, fuses, and associated wiring which will connect to the 27 ET3 relay will be classified f<r use in an Essential application The installation of the fuse bhick and placement of the test equipment on the floor in front of the breaker unnpartment has been seismically evaluated by an ,
Engineering Judgement and it was detennined that the installation has no adverse effects on the seismic ,
quahfication of the Ibential mnponents lhe slight additional burden that the test equipment will have on the potential transformer has been evaluated and determined to be well within the capability of the transfonner The test equipment will not change or prewnt any actions that are used to mitigate the consequences of an accident. In the ; vent of a malfunction of equipment important to safety, all other l oluipnuit will still perfism as designed ihe fused test monitoring equipment being in parallel with the tak tvohage relay prvents tir failure of the test equipment from atTecting the performance of the safety related function Installation, operation, or failure of the test equipment will not afTect the operation of the Enwrgency Transformer, F llus, or the potential transformer. The design of the system has not lven aheral in any way to change the operation of the system and no other systems will be alTected, therefore, the margin of safety will not be changed PTM %26 and PTM 97 05 TITIA Tems rary Repair to Fire Protection (FP) Jockey Pump Piping DESCRIPTION; PTM %26 was an limergency PTM implemented in Septemtwr 19% that provided f(n the temporary repair of four pinhole leds in the discharge piping of the FP Jockey Pump. PTM 97 05 was implemented in February 1997 for repair of one additional pinhole leak ia tins piping. Temporary patches were installed to prevent w ater impingement on FP components The Camaged jockey fire pump piping was subsequently replaced SAFETY ANALYSIS: The probabihty of a fire or other USAR pwtulatal accident is not increased as a result of these temporary patches.1he patcles are to restore leak tight integnty of the piping. Consequences of accidents are not mitiputed by the jockey fire pump and its associated piping, llowever, the FP system and its auto start features are provided fit nutigatim of spwial ewnts,inchdmg spent fuel pool makeup following a design basis carthquake. These PTMs do not affect the ability of the FP system to supply water to mitigate these special events. An Operabihty Assessment concluded that adequate water is available for event mitigatim even with completejockcy ptanp piping failure This bounds the impact of the failure of these PTMs. The temporary patches reduce the probability of a failure of FP equipment by preventing water damage to components in the vicinity of the leaking piping Also, spurious fire pump auto starts are -
prewnted The capabihty of the fire pumps to supply water for fire suppression activities is not afTeeted by these PTMs Failure of thejockey pump aaJ electne fire pump have beia previously evaluated. The diesel fire pump could auto start, but would not fail as a result of these PTMs. No additional failures are
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i i
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intaducal by the PTMs ard isolation capability is pmvi&d 1he capabihty of the lhe pumps to sequentially auto start arul provide flow and pressure is unaffected by these PTMs. L PTM %27 and PTh4 %28 TITIE: Diesel Generstor (DO) Muflier Ilypass Valves Failed Open [
Dl!SCR!pil0N: 1hese PTMs placed DO#1 and 0,0#2 Mullier Ilypass Valves, D0.AOV.Mlli and DO AOV Mil 2,in ;
a faikd open puitim due to design anecins with tic present muflier b> pass valves The PTMs removed ;
control air from the mullier bypass valve actuators which failed the mullier bypass valves open. !
I lingineering Pndect Request %I03 has been initiated to modify the existing muffler bypass system.
SAlliTY ANALYSIS: Placing the dicsci muffler bypass valves in the failed open position &cs not affcot the operation of the 1 cmergency diesels.1hc diesels will perform their required actkos during accidents (loss of Coolant Accident wnhloss of Ofiute Power) and the probability of these accidents is not incrensd Placing the diencis' rnullier b>pms valves in die failed open condition emures an exhaust path will exist for any peible accident scenarios Radiological release consequences assumal for a design basis less of Coolant Accident will not be affected by this activity. This PTM emures the aesel exhaust path is ;
available fir operation to mpply pnver to systems which mitigate ra&ological release consequences 1he i only direct safety related equipment afrected by the pngesed activity is the diesel engine itself. The !
vendor has stated that placing the muflier bypass valves in the failed open position does not have any ;
short term or long term effects on the operation of the diesel engines. It also climinates any failure ,
mechanisn m the diescl's exhaust path system.1hc dicscis will be able it supply power to designated safet3 equipment lle pmbabihty ofequipment mah' unction is unchanged and no new types of accidents tr eqtiipnent malfunctkos are created 1his activity does not reduce the margir. of safety for the diesels by ensuring an exhaust path always exists for the diesel engine exhaust, thereby allowing the diesel engines to perform their design function to supply power to safety related equipment under accident omditions.
PTM %29 TITII: Remote Camera Imtallation for Cooper Opca llouse Dl!SCRIPil0N: 1his P1 M &cumented the temporary addition of eight remote cmtrolled cameras and their associated conununicatans ard pnm cables to supprt 19% CNS Open Ilouse activities Cameras were installed in the following areas Control Rotan Reactor fluildmg 100P lxvel, Turbine Generator lluilding 932'6" 1.evel, Turbine Generator fluilding Main Crane, and the intake Structure. The installation on the overhead Turbine lluilding crane results in a temporary inoperable condition for the cranc due to the '
unummicatkos cables which are routal to the floor nica, Ilowever, the crane can be returned to senice in an emergency by simply cuttmg the small communications cables SAFl!TY ANALYSIS: 1hc temprary imtallations have tu efTect on any safety related systems and have no impact on any of the accidents disemsal in the USAR. Therefore, there is no increase in the probability of occurrence or anupnees of any of these accidena lhe tempontry cameras are imtalled such that they cannot cause the malfunction of any equipment important to safety. TI.c only postulated accident scenario is tl.at of a seisnue interaction omeem which is already direselin the design enteria of safety systems as desenbed in the USAR. None of the camera imtallatims pysent a seismic 11 over I concem. No margins of safety in the Technical Specifications are affected by the imtallation of the temporary sideo equipment.
PTM 96 30
. TIT 1.ll: Imtallation ofIr.sulation on Z Sump Discharge Piping DliSCRIPTION; 1his PTM authoriecd the temporary instrJlation of Armallex msulation on caposed scetions of Z sump discharge piping to prevent frecting. 1his imtallation was required until proper replacement material was procured and received.
43 l
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i i
SAll!TY :
ANALYSIS. 11us activity dies not alter any plant parameteis, margins of safety, or operating enteria. he these ,
~
reasons, il carmot increase the probabihty of accident occunence. This PTM is intended to prevent freeting of the Z sump's discharge lines, ensuring that the accidmt mitigating functions of tir Standby Gas Treatment system remam intact and unabated. Tle weight of the temporary insulatim, plus the weight of the current insulation, is stn! lighter than the weight assimed in the seismic analysis.1he use i of nonessential material is acceptable on a temporary basis beamse liberal conservatisms have been swumed in specifying the type and minimum thickness to be und 1he etwmical composition of the material has been carefully rniewed to ensure that no detrimental ef fects will occur when in contact with carbon steel. These factors combine to ensure that the probability of equipment malfunction is not ,
inervaul 1he potential failure nodes associated with Z sump's disc harge piping remain unchanged and no trw failure modes are introduced. The basic fit, form, and fun tion of the sump's discharge piping ranain unaffected Since this activity does not alter Oe function of the Z sump or afTect any interfacing systems, the margin of safety is not reduccd PTM %31 1111,11: Ileat Tracing of Standby 1.iquid Control (SLC) System Piping bliSCRIPTION: This PTM installed manually energized heat tracing on the Sic pwnp discharge relief valves !
SLC.RV.10RV and ST-RV.!IRV and aneciated relief valve discharge piping. The occurrence of an area ambient temperature of 10* F or more below that reflected in To:lmical Specificatum Figure 3.4.2 ;
Penxnt Sodium Pentabc.n.te by Weight of Solution) could result in lxton precipitation and subsequent ;
line laickage. This potential for hne bhickage during low temperature conditions alTeets both relief valves and aiuld cunpmmise their abihty to satisfy t}wir intended functions. This section of piping is not in the primary process flow path. Consequently, the occunence of a postulatcd line bhickage due to luon buikhip does tot directly asnpnunise the function of the SLC system in terms of its ability to irpect a neutron absorber into the reactor vessel. This moafication is a reliability enhancement as these lines am mt required for operability of the SLC system The PTM rennins in place and is expected to be followed up with a permiuwnt modification.
SAFl!TY ANALYSIS: 1he a&htion of heat tracing and insulation to the SLC system relief valve piping has no impact on the function or opern' ion of the SI C system or any other system and, therefore, does not contnbute to the ;
probability of occunence of any accidents, The adation is an enhancement to the system protection which will decicase the pouibihty of system failure without impacting the function or operatum of the SLC systen a othemise contnbutmg to equipment malfunctions. The heat tracing and insulation cannot impact any systems required for accident mitigation This PTM increases system reliability, thus it will not reduce the margin of safety.
PTM % 32 TITW: triled I.ead for "A" Main Transfonner Oil Pump Alarm ;
Dl!SCRIPTIONr 1his P1M hfled a lead to asable a nuisarme alarm in the Control Rmm "A" Main Transformer oil pump
- 4 had an inwmal grmind and the pump could not be repaired with the transfmmer in senice. As an interim solution to remove the ground frmn the system, the pump w as electrically discoru.ccted. As a result of disconnectmg the pmup, an alarm window alanned each time a transformer oil pump cycled.
1herefore, this PTM lilled the lead (br pump #4 alann input to chminate the resulting nuisance alann-All other pump low oil fiow alanns reinained in operation. The pump was subsequently repaired in Rlil7.
SAF1?TY ANALYSIS: 11e Main Step Up Transforme is not inchided as an irutiator or contributor to a design basis event as desenbed in the USAR. Defeating an alarm fa the out of senice pump has no impact on the transfbrmer, i 1he alann funetam associated with the oil pump, transformer cooling, or transformer function in general
-44 l
l l
l have no impact on the consequences of any event described in the USAR. The Main Step Up Transformer supplies power to the grid, Interface with systerr : important to safety is provided by the stanup trandnner and the emergency transfamer. Therefore, defeating an alarm on the Main Step Up Transformer has no impact on the probabihty of occunence or omsequences ef a ~ .,dfunction of 1 cquipuent imputant to safety.11e Main Step Up Transfinrar is net c cc::tnbutta nor is it used for event mitigation as delirniin the statim bladamt analysit Disabling this alann has no impact or interiace with l r ety.1hcidain Step Up
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any functions asociated with safety equipment <t equipment hnportant b sa Tramfintar is not discussed in the basis for any TechnicalNS eif:cann lida actisity does not teduce ,
the mat gin of safety as defined in the basis for any Tecimical Sec@mathm (
PTM %33 '
Tilt!!: Disabling of Diesel Fire Pump Remote Stop Capability i D13 SCRIPT 10N: The Appendis R revalidation efrwt identificJ that the dicael fire pump, FP P-D, msy not perform as t sequiral fit a fire in tic Cable Spreading Room or Cutarol Roont Fire inducchlar,1sgo to cable FP.104 could r.puriou4 energi/c the renote stop relay, and thereby prevent the puty %om starting, or stop the pump af already running This PTM installed a jumper in the diesel fire ptem curarol panel to disabL the remote stop capabi.ity and as.rure automatic pump operation for rtres in 'he C& Sp ending Room ard Contml Ravn Engiruxting Pniject Request 97-044 is being generated fa permanent redution of ;
this concern.
SAFl?TY ANALYSIS: Disaining of the remote stop capabihty for 1T-P-D w111 not affect the probabihty of fire occurrence because FP P-D is designed only to respond and mitigate the cmsequences of a fire and the P fM does sut represent an igrution sotam lhe fire nutigation capability of FP P D is unafTected by this PTM. The abihty to sugyly water to the spent fuel pool is likewise unaffected.1he fire suppression effects analysis does rut rely on a srveific time of inadvertent system actuation and FP system rupture is bounded by the flooding analysis for Main Steam Lme and I cedwater Line breaks in vanous plant a cas. The PTM raluces the probabihty of FP P-D failure in specific fire events, while the remaining failure probabilities are unchanged. It disables the remote stop capability only and does not intrw*,:ce aMitYnal failw nudes Fire pump grrabihty in restored by installation olthis PTM, thus increasing the salcty margins relative to fire ptunp availability.
PTM %34 TITLli: Unplugging of Cooling Fans on Plant Management Information SyMem (PMIS) Terminals in Control Room Dl!SCRIPTION: 'ihis PTM authorind unplugging of the cooling fans on PMis termin .ls in the Control Room in order to >
tahxx reise levels to meet NUREG 0700 puidelines and enhance Control Room communication. The e PTM is expected to remain in place uroi! a mahfication package is developed to permanently remove the fans.
SAFETY ANALYSIS: The PMis terminal coohng fans were originally installed without design and salety evaluation documentation The fans were intended to reduce heaiiclated terminal failures. Ilowever, installation of the f.ms has had no street on the number of heat related terminal failures. No cooling fans were imtalkd when plant ewnts ard equipment malfunctions wcre originally evaluated in the SAR; thmefore, the removal of there fans cannot allixt the consequences of any previously evaluated events or eq iipment malfuncuxu The terminals prmide visual data only end do not directly afTect operation of the pmcess canputer. Process computer terminals are not Technical Specification equipment, not do they fulfill a safety ftmetion I
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PTM 97-01 TITlJI: Installation of Tunnel on Turbine Deck DliSCRIPTION: The Ttabinelluildmg 932' level u as part of the RtJiologically Controlled Area (RCA). As a result, no internal pathway exided f<r pammnel to walk fnun t}c Maintenance Shop to the Adm.nistration lluilding without going out& ors 1his PTM provided for the erection of a temporary tunnel on the turbine kk to create a non RCA corridor letween the Instrument & Control (l&C)/Electne Shop unridor and fne Machine Shop A partial ctatain was also placed on the east end of the I&C/ Electric Shop corri&ir. The installitana was fabricated inen non-combustible framework and translucent, fire retardant plastic shecting. The PTM was removed when the plant shutdown far REl7.
sal 1?TY ANALYSIS: Installation of the tunneUcurtain will not change any scismic analysis Tunnel structural materials are rurxxsnbustible and non-flammable and plastic materials are fire retardant and do not adversely affect the fire loads in the Turbine lluilding The structure is passive and does not directly impact or alter current plant systems The turmel will be made such that it may is entered or exited quickly for plant personel safety. Operator access to various equipment will not be impted, nor des the existence of the tunnel create a new release path for radmactive material. Air flow in the Turbine lluilding and
!&C/ Electric Shop (xtridct willIc unaffected Tunnel covenng is translucent, allowing egress with only emergency lighting. 1hc tunnel / curtain acs not impair or alter any fire &dection or protection niuipment.1his actaity acs not change the Turbine lluildmg IIcating and Ventilation Sy stem from its eunent configuration, not will it cause the system to fail. It dues not afyect the control or :lonitoring of radioactivity. De tunnet/ curtain is tot located near any equipment important to safety. Failure of the ttamel will rut cause equipment to malfunction. I!xistence of the tunnel will not change the analysis for tornado generated missiles. The tunnel will not impair plant response to emergencies or equipment function The turmel! curtain will not obstruct any critical ventilation paths and will be kicated such that it ass not cover any floor drains. Separation of plant equipment will remain unchanged. This activity will not interact with any equipment that would reduce the margin of safety. Existing margins and analyses remain unchanged PTM 97-02 anu PTM 97-08 TITI.li- Connection of Memtoring Equipment to isolation Relay Cabinet A DliSCRipliON: These PTMs nonected an oscillographic recorder to contacts within the Pnmary Containment (PC)
Isolation System Train A. The temporary modifications connected two channels of an oscillographic ruxrder to three points within the Isolation Relay Cabinet A. Each of the ch:.nnel circuits was emnected through two fuses and a resistor and temporary jumpers were installed during the installation / removal pncess of the PTMs to maintain system operability.
SAFETY ANAINSIS: Le woik performed during the installation of the PTM and the equipment left in place while the PTM is active do not aficet the initiators of plant events evaluated in the SAR as no connections are made to any system that could initiate such events. The functions of the system will be preserved during the installatim' removal process via the use of temporaryjumpers. The design features of the emnection of the nxmitoring equipment provide for the isolation of the equipment from the system and precludes the propagaton of nuutaring equipment faults into the PC isolation System. The function, operatility, and unfiguratim of the PC isolation System will be maintained The mstallation of the equipment has been resiewn! for seismic interactions and found to be acceptable 'Jnplanned actuatmas of phnt equipment during the mstallatim' removal wuuld be the result of a single failure or operator error ihr which the plant is designed and evaluated. No systems or components are prevented from performing their deMgned action. All safety system actions will be performed if called upm The potential int.Jvertent actua! ion of the Group 6 isolation will not decrean the margin of safety sinec the actmn is the safety state.
Capabilities will remain to reset the isolation signal. Therefore, the margin of safety is not decreased.
46
l PTM 97-01 ,
TITLE: Tempvary Renxwal of RF POST ll! Due to Gathng Resulting in Adverse Effects on Reactor Feed Pump Turbine (RFPT) Operation DESCRIPTION. This PTM autluri/cd the temporary remcwal of 16 POST Ill,16PT 11 Rwle lilock Position Trarwmitter, due to galling severe enough to impact operation of the turbine. The transmitter was ,
terrunul to msure nonnal turbine operation until a replacement transmitter was procured and installed.
sat 1?TY ANAI.YSIS: 1he Pni maintams equipnwnt reliabihty and functionality by alkming the RFPT to ressmd to changes in demand signal as designed. The nonessential Reactor Feed system does not perform any accident mitigation functions, not does it provide support to any system or component performing mitigation functions No change in radiological efTects will occur as a result of the PTM. The transmitter has no impact on operation or control of the RFPT and, therefore, can have no irnpact on the probability of oceturence of any evaluated malfunction. No new malfunction mechanisms are introduced by removal r,i the transtnitter. Any possible malfunction scenario is botmded by the loss of feedwater transient,
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fm! water omtroller failure maximtun demaral transient, and the turbine missile analysis Lack of position irxlication f(r the RFPT runle block does rut create any accident scenarios different than those currently evaluated and no new faiSrc modes are introduced. The PTM maintains the reliability of the RFPT to respond by clindnating the impact of the failed transmitter on turbine respome. This maintains the validity of any assumptions, calculations, pacedures, or design specifications used to detennine the plant's margin of safety.
PTM 97-06 TITI.ll: 'ulling of Annunciator Card for Augmented Off uas (AOG)llydrogen Analyzer Alann DESCRIPTION: his PTM autluri/al pulhng of the annunciator card for the AOG 1lydrogen Analyzers downscale alarm to eliminate a nuisance olann. The downscale alarm is intended to provide an indication of analyzer problems baul m a loss of analper output signal Ilowever, due to tlw extremely low hydrogen cmtent runnally fotn!in the process, the alann setpoint is very close to the normal analyre output level which resuhs in frequent spurious trips. Engmeering is evaluating a permanent resolution to this problem.
SAFl!TY ANALYSIS: The AOG hydrogen alarm is not an accidem initiator and cannot increase the probabihty of occurrence of a piant event.1he alarm functam is acceptably perfonned by other checks, cahbrations, and functional tests. No increase in the cmsequenas of a plant event will resuh fmm this PTM smee the analyzers will be fully capable of perfmning their intended function The analyzers ne nonessentiel and have an alarm function m the event of high hydrogen in AOG The downscale alann function that was rernoved has no ciTat on equipment reliability. This PTM does not affect any safety related equipment, The possibility of arudyrer malfunctim will be the same before and aller this PTM. The Technical Specification margin of safety is unchanged since the high hydmgen alann setpoint is unchanged and since the existing checks, functional tests, and cahbratior.s are sullicient to provide analyter operational status that was originally perfonned by the downseale alarm PIM 9711 TIT 1.E: - Installation of Recorder on Digital Electro ilydraulic (del 1) Analog Circuiuy DESCRIP flON: This PTM attached a chart recorder to eight test points in the Dell cabinet to facihtate monitoring of DEII analog ciremtry. The DEII analog pressure controller circtutry was monitored for abnonnal signals to aselin determming the cause of a 5 psig reactor pressure drop and the cause of controller "ll" drin.
While the PTM did not determine the cause of the pressure drop, it did determine the cause of the controller dnf1 and provided other useful information The recorder was removed during RE17.
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sal 1?TY ANALYSIS: Test p>ints are isolated from the Dell emtrollers govenxv valve control signals 1he recorder will be in a postal area and access wdl be restricted to persormel requirol to monitor the recorder and change the chart pairr.1his PThi dxs not impact or alter the function of the Dell pressure controllers and the system will function as it did before the PTM was installed No systems or components needed to mitigate or control the consepences of any accident or malfunction of equipment important to safety previously analyred in the USAR are affected DIUl is classified as nonessential and ram-seismic, but is important to safety.1his P1 M will monitor Didi test points in order to investigate the cause of the l abnormal guvenxv valve movement and reduce the chance of a plant trip. The test points are isolated I by a 10fX0 ohm resistor to prevent the test equipment from affccting omtroller input or output signals. ,
lhe Dell system carnt initiate an accxkst that is not bounded by the USAR and (kes tot introduce any j failure mode that would cause a DidI failure that has not been analyzed DEII is not a safety related system arulis not a safe shutdown system. This PTM does not reduce the margin of safety as dermed in the basis for any Technical Specification PTh19712 TITLli. Tempnary Test Connections for ilV A0V 261 AV i
DESCRIPTION = 1his PTM instal'ai temptury test connections and isolation valves to the air lines ofilV A0V 261 AV, a Reactor thakhng Ventilatni Exhaust Inboard Isdation Valve, in order to be abb to quickly attach and remove diagnostic test equipment- lhe temporary test connectmns were installed to facilitate the attachment and removal of AirCet Diagnostic Test Equipment required for troubleshooting a periodic valve stroke time failure. Engineering is evaluating the pasibihty of making the test connections pennanent.
SAFl!TY ANALYSIS: IIV-AOV 261 AV and the anmciated commments do not alTect the initiators of any analyzed accidents.
The test tees are made of stainless steel and are coinpatible with the adjacent components of the Instrument Air system lhe test tecs do not restrict the flow of air to the air operator. Post installation testing of the tocs includes leak testing, stroke testing, and accumulator testing. In addition, the test tees are equipped with an isolation valve and a quick disconnect Therefore, the system will operate as designed with the test tees installed All components used in the PTM will be installed Essential.
IIV AOV 261 AV will be inoperable when the nonessential diagnostic test equipment is attached. The test equiprnent will be contmuously monitored when attached After the removal of the test equipment and the chwure of the test tee isolation valves,1IV.AOV 261 AV will be operable. These changen have no logic (r electncal interfaces.1his PTM tkes not affect the olwration of1IV A0V 261 AV or any other safety telated equipment. If the redundant valve (llV MOV 260MV) would fad,ilV AOV 261 AV would still function to isolate secondary contamment. The PTM cannot create a failure of the adjacent nx*r operated valve and does not intraluce any new event initiators. The components installed in the PTM excas! the pessure rating requirements of the instrument air tubing. During the connection of the tempvary test tees, llV-AOV-261 AV will be closed and inoperable. Aller the connection of the test tees, the stroke test ard the accumulator test will verify that the valve operates per the current station res,uirements Thetcliire, the margin of safety is not reduced PTM 97 14 TITLE: Temprary Power for Torus Work DLSCRIPTION: This PTM authorved the installation of tempirary electrical pewer fit outage hiads inside the Reactor 13uikimg to acumutudate wurk in the krus, specifically deshdging and the installation of new Emergency Core Cooling System strainers. It also supported in Vessel Visual Inspection activities. Temporary power supplies were routed through a spare penetration. llis PTM required the installation of a temporary fire neal and an essential seal to ensure Secondary Containment operability. The PTM was in place only during cold shutdown conditions.
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SAll!TY ANALYSIS: Design is pt tle Natimal Electric Code (NEC), Orrefore, the pnibability of fire from temporary cables is rnd increased The electrical loads are supplied fmm rmnessential IESWilD SLDC which has no impact on accident analynca The fire rical and Secondary Containment seal are Ibr accident mitigation arxl do rxt create tie reibihty of an axident Scustdary Containment will remain intact when required.
Retesting of the rescaled Seemdary Containment penetration will te by visual inspection with smoke or lo wire anenxeneter. The electrical loads are being supplied from nonessential 12.5 KV power and do rad impact c41uipment imputant to safety.1his coupled with NEC design of the distribution cabling arxl fusing will prornt any impact to safety equipment. The fire barrier will be sealed with an aproved seal design with a fire rating commentirate with the rating of the fire barrier. The Secondary Containnx:ut seal will utdve cuential materials to preclude rnalfunction The operability of the seal will be venfied dising PTM imtallatiert 1he avvcquences of a Secondary Containment or fire barrier failure are unaffected 1he 12.5 KV system is not governed by Technical Specifications. The Limiting Cornhtion for Operations for Technical Specifications 3.7.C and 3.19.A are being observed during imtallation and removal of tic PTM. Oper ability will be verified following installation.1herefore, there is no reduction in the margin of safety as dciined in the basis fbr any Technical Specification l'1M 97 15 TITLl!: Installation of Portable llattery Chargers for Diesel Fire Pump DESCRIPTION: The battery charger for the Diesel Fire Pump (FP P D) failed. therefore, temporary battery charging pnnisiam were required in tyder to supprt FP-P.D operability. Portable battery chargers were installed on each of the 24 VDC batteries for the diesel fire pump. This PTM was in effect until the original chargers were rcpatted SAFI!TY ANALYSIS' The fire pump battery charger suppirta a safety feature, but is not an accident initiator.1his activity ensures that the fire pump will perfbrm as designed 1he portable chargers are self regulated and will maintain the battenes fully charged in order to ensure that the diesel fire purnp will be operable on denuux! No impact m dm pnhabihty of a malfunction of equipment will result from this PTM. Failure of the subject njuipnent either with tr witinut this PTM ultimately results in the loss of the dicsci engine dnven fire panp 1his PTM is limited to the diesel fire pump and cannot afTect any equipment that could cauw a plant event. No new types of equipment malfunctmns are created by this PTM that could affect equipment outsule de dxsci fire pump system 1here is no elTect on the Technical Specification margin of safety. Diesel fire pump operabihty w111 be preserved and fire pump system operation will be unaffected-pTM 97-17 TITLE: Diesel Generator #2 lhiildmg Crane Mothlication DESCRIPTION: This PTM imtallal an ekttrical hoist / trolley onto the 2-ton crane hicated in the Diesel Generator Room.
This tempranty repheed de mechanical hoist'imiley mechanism normally imtalled The PTM provided for a more cilicient and quicker means of mnoving heavy diesel engine parts during diesel overhaul during the outage. The electrical hoist w as subsequently removed and the mechanical hoist reinstalled following completion of diesel engine overhaul.
SAFl?TY ANALYSIS: The technical evaluation performed for this PTM determined that kiading/scismic and electrical mpirenuits are nd alTeeted by this activity, and that operation of the diesel engine will not be arfected.
The PTM will not prevent the diesels from providmg emergency power needed to mitigate radiological release requirements for a less of Coolant Accident (l OCA) or Loss of OITsite Power (LOOP) The potential failure scenario of the hoist fathng on the diesels will not be changed by this PTM. No new fadure axiuuios are created lius P TM is specific to the equipment in the diesel engine room. No other safety related equipment would be atrected Any equipment failures would be bmnded within the i OOPLOCA design basis nvnts The luist/ trolley unit is not identified in Technical Specifications and will not alTect Technical Specification equipment.
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PTM 97 19 and PTM 97-21 Till.!!: Deminei ahied Water (DW) Supply to Condensate Pressure Maintenance System i DESCRIP110N: The Condensate Systen supplies the pressure maintenance system during rmrmal plant operations;
- howeser, when the condensate pumps are not operating, the Reactor Building Auxiliary Condensate Pump is the rmtmal backup supply to the pressure rnainte.aance system. Due to two planned Division i ekttrical bus outages, the Reactor fluilding Auxiliary Condensate Pump was not available. Therefore, these PTMs installed a temporary cross-connect between the DW system and the Condensate Makeup system to maintain a pressure source to Emergene;' Core Cooiing System (ECCS) and Reactor Core ,
Isolation Cooling (RCIC) system pressure maintenance piping.
SAFETY ANAL.YSIS: .b statal in the Technical Spectricatims, all ECCS and RCIC systems will still perform their design basis functions without the pressure maintenance system being operational. Utilizing a ddTerent pressure i muce for the systems will rxt afket alamur operatimal functions described in the USAR. The pressure
- nource being changed from the nonessential Condensate Makeup system to the imnessential DW system does not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR. The existing isolation check valves remain in the flow path,
' maintaining the safety related pressure boundary intact. The use of a hose as a cross connect does not ercate the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evahiatalin the USAR. loss of pressure maintenance is already considered, and no ,
new faihre modes are added The intent of the PTMs is to tanintain pressure maintenance in operation in order to minimire damage to the discharge piping of ECCS systems. The margin of safety as dermed in the bases of the applicable Technical Specification is being maintained.
1 PIM 97-24 TULIL Temporary Patch on Riverwell Piping Dl! SCRIPT 10N. This PTM installed a temporary patch over a pinimle leak in riserwell piping near Circulating Water (CW) Pump D, near the outside wall of the Senice Water (SW) Pump Room. This patch remained in place until the piping was replaced SAFETY ANALYSIS: This PTM dxs txt incruse the probability of a CW or SW failure. Installation of the patch will enable operation of the CW and SW system to suppst outage operation. The chance of failure of the SW system is not increased as the pumps are designed to operate without gland water supplied by the riverwell .
I system in all design basis accidents SW and Residual 1leat Removal SW (RilRSW) do not rely upon tiverwell to perform their safety function Neither riverwell nor CW are relied upm to mitigate the unsequences of an accident. The patch functions as a diversion to prevent water from spraying on the CW pumps Faihtre of the patch uill not cause a plant event or malfunction of equipment important to safety. The riverwell system will function without the patch lloth a CW pump trip and the loss of riverwell to SW and R1IRSW have been previously analyzed No new faihire modes aic introduced.
'lhe existmg Tothnical Specification margin of safety is not reduced as the fmetion of SW is not changed by this PTM.
PTM 97-27 TITLit llackup ilatteries for Cellular Phones and Microwave Digital Muhiplexer DESCRIPTION: MP 96-153 cmnected the new cell phme system to the exirting PBX power .,upply w hich increased the loaamg such that the backup batteries were unable to provide the required six hour battery backup capabihty to the PilX telephone systan as regmred by the CNS Emergency Plan. In order to address this ancem, a temprury 48 VDC battery bank was emnected to the existing charger in the Communications Roan to power the new site cell plune systcm and miemwave digital multiplexer, The cell plone system
< and microwaw digital multiplexer were separated from the PBX power system. This PTM was removed u hen the PllX bettedes were replaced with larger capacity batteries.
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l SAF1ITY ANAL YSIM: Ris aethity does rxt intaface with or involve any components or systems credital w ith the initiation of a plant event ne cite cell phone sptem and microwave telephone line are not cre hted in any accident w transient analysis This PTM does not change the function of any event mitigation system, structure, or component. The potentie.i for increased dose is not increased and the ability to mitigate the nosequerxes of a plant ewu is not redtmi nis activity has no c!Tect on equ;pment important to safety.
%e unnplete faihre of the equipment installcJ by this activity will have no cfTect on the nuclear system a plant safety systern his PTM does sut involve any parameter included in the basis of any Technical Specification.
PTM 97 31 TITI.": Installation of Jumper on Main Turbine Governor Valve 1.imit Switch DliSCRIP flON: This PTM povided fm the installatior.of an electricaljumper on the main turbine governor valve #2 closed li. tit switch. This switch failed to actuate during performance of Surveillance Pmcedure 6 RPS 302, Main Turbine Stop Valve Closure and Steam Valve Functional Test. This failure prevented the main turbine stop valve #2 from closing. His PTM was implemented to suppret the main turbine stop valve monthly testing by instalhng ajumper to simulate a governor valve #2 c% sed signal to the main turbine st(p valve #2 test solmoid circtut. The jumper remained in place until the closed limit switch on govemor valve #2 was fixed.
SAFl!TY ANAL,YSlS: The only function of the open or closed limit switch on the govemor valve is to provide valve position inxlication in the Control Room and an interkick ontact to the stop valve test solenoid circuit. The limit switches do not pruvide any safety related fur ai and a*e tiot discussed in the USAR. This PTM does not alter the function of the Digital lilectro-llydraulie (Dlill) pressure hxy, turbine trip circuit, or any safety relatal systent it interfaces only with the main turbine stop valve test solenoid circuit. No systems or compoents nmlal to mitigate or hmit the consequences of a plant event or malfunction of equipment important to safety prniously evaluated in the USAR are affected This PTM does not intraluce any fuihue nusle that would cause a turbine tiip circuit or DIIII faihne that is not analyzed in the USAR. The margin of safety as dermed in tic basis for any Technical Sprification is tot reduced. The closed or olen limit switches on the main turbine governor valves arc not discu sed in the Technical Specifications.
P'l M EU TITI.!!. Disabling oflilevated Release Point (IIRP) Kaman Radiation Monitor Sample Flow Control Velve Dl! SCRIPT 10N: The purpose of this PTM was to electrically disable the !!RP K.anan nonnal range radiation monitor sainple flow notml vahr such that it will not respond to variations in liRP stack flow, i c., utilize a fixed sample flow rate. This changed the valve from automatic to mnnual operation. It also changed the particulate sampling methodology of the !!RP normal and high rar.ge radiation momtors from isokinetic to fixed flow samphng The PTM is expected to remain in place until Design Change 92-021 is implemented to convert the isokinetic sampling of the !!RP monitors to facd flow sampling.
SAFl!TY ANAL YSIS: Ec liRP Kamaninvitars mly prmide indication and alarm ftmetions. They do not provide any control functions. The monitors are not identified as a potential sotrxe of a plant event. The indicatiorvalarm function wdl not impact on-site or otT site radiation releases. Therefore, these monitors will not increase the consequences of plant events or equipment malfunctiom desenN:d in the USAR. The !!RP Kaman nusutors are not imulmi with initigating any plant event descr%: in the USAR imd do not prmide any supprthotml of any equipment imputar.t to safety used k :ne mitigation of a plant event. Bis change imuhrs unplugging an electrical connector at the sam (.e flow ec.itrol valve at the !!RP Kaman normal range monitor which changes the particulate sam,)lir.g methak31ogy from isokinetic to fixed flow sampling. The act of unplugging the electrical cor.neetor cannot create a plant event. The change in particulate samphng at the !!RINoly impacts (reduces) the ermr in particulatchaline release rate records.
The monitors cannot cause a malfunction of any equipment important to safety. The Technical Spmfication fa the !!RP high range radiation vnonitor is unalTected by this change. Also, per rniew of 51
l Technical Sl uificaixo liases 3.21 A 2, there is no impact on the baws for the !!RP Kaman nonnal range i radiation monitor. Since the change w111 result in the reduction of error in particulate / iodine sampling, l tic margin of safety in the basis for Tcchnical Specifications cannot be reduced Per rniew of Technical Specification Ilases 3 21 C, Gaseous Elliuents, the change will not impact this bases P1h4 97 36 TITil?: lilevated Release Point (liRP) Kaman Radiation Monitor Ground Noise Spiking DliSCRIPTION.11e ERP rannal:ange radiation trxsut(v had been expenencing intertaittent ground noise s;,ikes. It was suspx tal that a capacitor between the signal and chassis grounds was causing the pound noise spikes.
1hereftsv, this PTM renxwed tic capacitor fcr 2 to 3 weeks to detennine ifit was in fact the cause of the spiking. During prnious troubleslmting, it was verified that the removal of this capacitor did not affect the taliatim measurements axu at the skid mounted italication and wntrol unit or recorder.1he removal of the capacitor did not solve the spiking problem therure, the capacitor was reirutalled ANAIMSIS: 1hc ERP Kaman nxnitors mly panide indication and alann functions. They do not prodde any contml functions,1he monitors are not identified as a potential sou.cc of a plant event. The indication /alann functim will rut impact on site or oft. site radution releases. Therefwe, these monitors will not increase the consequences of plant events or equipment malfunctions described m the USAR. The ERP Kt 'an truautors are tut invahrd with mitigating any plant event described in the USAR and do not provide any suppvVeuntml of any equipment impstant to safety used in the mitigation of a plant event This change invohrs tenning a capacitor in the ERP rxvmal range radiation monitor Removal of this capacitor does ext change tle function of the monitor. The ERP Kaman radiation monitors cannot cause a malfunction of equipnat impstant to safety. Based m rniew of To;hnical Specification liases for sections 3.21.A.2 arxl 4 21.A 2, the margin of safety is unaffected by this change. During the installation of the PTM, the rahation measurement willle confmned before and after the renoval of the capacitor. This ensures that the calibtution of the detector was not affected.
PTM 97-37 TITI.li. Temporaiv Removal of RF POST 1 A DiiSCRIPTION. This PTM acumented tia removal of RFJOST 1 A, Reactor feed Pump Turbine (RFPT) A Noule likick Positw Transnutter, by Opera %ns personnel The trar,;mitter was removed at the direction of the Shift Supenid ;u .uspected galling. Following funher investigation, it was detennined that RF-POST 1 A was not the cause of the mactor feed pump pwblems Ilowever, Reactor Fent Pump 1 A must be shut down to reinstall RF-POST 1 A.
SAFETY ANALYSIS: 1his tnuratutter panides an indication function only to assist Operations when transferring an operating RIl'T fmm low pressure to high pressure and from high pressure to low pressure steam Removal of the transmitter eliminated Control Room irlication, but the impact is minimal since altemate means of verifying proper operatim of the turbine exists, and this indication is not required by any safety basis documert The PTM maintains equipment reliability and functionahty by allowing the RFPT to respond to changes in demand signal as designed The nonessential reactor feed system does not perform any accident mitigatior, functions, nor provide support to any system or component performing mitigation functions Therefore, the consequences of an event cannot be increased by this PTM. No change in radiological efTects will occur. The transmitter has no impact on operation or control of the RFPT and, therefore, can have no impact on the pmbability of previously evahmed malfunctions No new malfunction mechanisms are intrnhwed by renxwal of the transmitter, Any possible accidenthnalfunction scenario is bounded by the loss of feedwater transient, feedwaer contmller failure maximum demand transient, and the turbine missile analysis Lack of position mdication for the RFPT nonle bkick (kies not create any new accident scenanos Renmal ofindication cannot introduce a failure mode not already ansideral 1he PTM does txt affect the vahdity of any assumptions, calculations, procedures, or design s[veifications used to determine the plant's margin of safety.
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PTM 97 39 TITLli: Installation of Monitonng Equipment for Reactor Feed Pump "Ir' DESCRIPTION: his PTM installed temporary monitoring equipment to support troubleshooting of the feedwater Immp control logic in order to detennine the cause of feedwater pump turbine speed oscillation. The PTM installed two AirCet pressure tramducers and two AirCet laser displacement transducers. The PTM was' removed upon completion of troubleshooting -
SAHiTY ANALYSIS: The pressure transoucers will be connected to redundant current to-ptcasure (I P) modules in order to satisfy a single failure. A failure of the one of the pressure transducers will be no difTerent than a failure of one of the I-Ps The svstem is designed in such a way that a failure of one of the 1 Ps will not afTect system operation.1his PTM will be installed and removed when the feedwater pump "Ir' is out of service h mler to min.mue reactw feedwater ikw oscillation. The noncuential reactor feed system does l ruit pafmn any 4.ccident wtigation furx:tims, nor suppmt any system or component performing accident mitigation functions. 1herefore, the consequences of an event cannot be increased. No change in rathological effects will occur. There is a redundant pneumatic control signal that will prevent a Reactor Feed Control System malfunction due to a failure of one of the 1 Ps. No new malfunction mechanisms are intnxtuced during installation or removal of this PTM. Any possible malfunction / accident scenario that could be introduced by this PTM is bounded by the loss of feedwater event, feedwater controller maximum demand event, and the turbine missile event.1his PTM does not affect the validity of any assumptions, calculations, pncedures, or design specifications used to determine ths plant's margin of safety. ,
PTM 97-40 -
TITI.ll: Monitoring of Z Sump Control Circuit Dl?SCRIPTION: This PTM provided recorder capability to detennine the cause for annunciation of the Z wump trouble alarm Si/115. It autixvval the installation of essential wiring and termination connections funn selected tominal points inside the terminal box to essential fuses which provided isolation between the essential pations of tic contml circuit and the tuviessential recorder. The PTM will be removed upon completion ofmonitoring.
SAMITY ANALYSIS: The loss of the Z sump circuit cannot initiate any plant event it only afTects control circuitry for the Z sump for the period of time required to connect the wiring. Therefore, there is no increase in the probability of a previously evaluated plant event Isolation fuses assure that Z sump will operate in response to all analyzed plant events and therefore will not reduce the ability to limit releases within 10Cm100 hmits lhe essential wtnng arx! isolation fuses preclude interaction between the recorder and Z sump pumps control circuits. Faihire modes (fault, relay faihire, and loss of power) and circuit protection remain the same. No new release pathways are introduced due to this PTM. The functions and operation of the Z sump control circuits remain the same and no other circuits are affected by this change. The PTM does not alter the function of any current plant systems, structures, or components; thercline, the possibility of a plant event of a ditTerent type than previr,usly evaluated in the SAR is not crenied. His PTM does not interface with any other equipment imposic9 to -afety and no new failures imles are intrahiced The margin of safety as defined in the basis of any Technical Specification is not reduced by this PTM 53
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l P1M 97 41 llT!JL Installation of Manual isolation Valve for 11 Sump Pump Instrument Air Supphen D11SCRIPTION: 11e cunoit instrument Air (IA) systan does not allow isolation ofIA for perfonning maintenance on 111 and 112 Sump Pumps without renuiving lA frten other systern components such as Reactor Core Isolation Cmling (RCIC) air operators and Service Water (SW) air operators. Therefore, this PTM installed a manualisilation valve for the 11 Sump Pump 1A Supply to support maintenance of the 11 Sump Pumps in the RCIC quad lingineering IIvaluation 97 307 is being prepared to make this modification permanent SAFl"I Y ANAINSIS: 'lle valve being added to the 1A system upstream of the 11 Sump Pump 1A isolation valves will meet or exceed the system design requirements of the lA system. Failure of the valve would be no more likely than the failure of the tubing it regaces and would have no d iferent efTect on the lA system than would faihre of the tubing it replaces. The addition of the ruanual isolation valve does tot change the function of the IA system since the valve is being added only for maintenance pmposes. Mispositioning of the iwilathe valve is not camidered a credible event. During installation of this PTM, the RCIC system will be declared inoperable and the appropriate Technical Specification Limiting Condition for Operation entered Also, the Reactor !!quipment Cooling (Rl!C) SW cooling weler throttling control will be transfened to safety related valves SW.MOV.650/651 as neecasary och that the cooling water supply to the Rl!C heat exchangers will not le afTected. For these tensoris, the pu,babihty of occurrence or amequences of a plant event ce malfunction of equipment important to safety are not incr :ascA and the margin of safety as defind in the basis for any Technical Specification is not reduced P1 M 97 42 TITLil: Jumper installation on filevated Release Point (liRP) Kaman Radialmn Monitor Dl!SCRIPTION: The pncess flow measured at the IIRP Kaman Normal Range Monitor indicated zero flow, which disagreed with the indications of the !!RP Kaman Iligh Range Monitor and the liRP Process Flow Recorder When the signal path of the liRP Kaman Nonnal Range Monitor is interrupted, the paress flow will work for approximately 25 minutes and then over a five minute time period drop to zero.
Troubleshooting determined thet jumpering out the capacitor between the signal and chassis grounds climir.ated this problem The jum; wing together of the chassis and signal ground systems danng troubleshooting had no elTect on the radiation rate measurement as seen at the skid mounted indication and omtrol unit 1herefore, this PTM ins 1alled a jumper which connects the two ground systems together creating a single point ground system The PTM is expected to remain in place until a replacement detector (of a difTerent revision)is installed.
SAFliTY ANAL.YSIS: 1he liRP Kaman numitors mly panide indication and alarm functions. They do not provide any control fimetions. These monitors have not been identified as a potential source of a plant event. The !!RP Kaman radiation monitors will not limit on-site or otT-site dose resulting from plant events. These nusutors will not increase the consequences of plant events described in the USAR. The liRP Nonnal Range Monitor is not identified as equipment important to safety. Any malfunction of equipment used to mitigate a plant ewn! described in the USAR will result in no changes to radiation releases due to these Kaman nxnutors The act ofinstalling a jumper or a jumper itself will not initiate a plant event. As the
!!RP Kaman radiat:on nxsut<rs have no support /contml functions, they cannot cause a malfunction of any equipment important to safety. llased on a review of Technical iS ecification ilases for sections 3.21.A.2 aisl .121.A 2, the margin of safety is unaffected by this change. During the installation of the PTM, the radiation rate measurement will be confumed befin aint after the instahation of the jumper. This ensures that the cahbration of the detector was not atTected 54 P
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.)
PTM 97-43 TITI.E: Troubleshoothg Reactor Feed Pump Turbine (RFPT) B Control Sysm DSSCRIPTION: This PTM installed temporr.y monitoring instrumentation to support troubleshooting of RFPT U speed er.cursions liigh impedance test equipment was attached to one spod probe output and various points in the Signal Pmeessor Unit and tell in place to capture data during the next excursion The PTM was removed folle ving completion of troubleshooting Repair work subsequently eliminated the speed excursions.
SAFliTY ANAL.YSIS: Test equipment m.1 be electncally isolated from station power, and concunent verification during a onecton and dmonecum pmvides controls to reduce intnxluction of a transient. Protective bamers will orevent inadvertent contact. Due to the highly reliable design of the test equipment once installed, the probability of a control system failure caused by the test equipment resulting in a plant event is not increased This PTM does not impact the function of the RFPT control system, No systems or comixnents txxded to nutigate or control the consequences of an accident or malfunction of equipment important to safety analyzed in the USAR are alTected The RFPT control system is nonessential /non-seiese, but important to safety. Due to the highly reliable design of the test equipment, the probability of a malfunction of equipment important to safety is not increased The test equipment will be adequately isolated and anchored to prevent impact on system operation Ilowever, should the test equipment fail, the failure is bounded by the has of feedwater and feedw ater contmiler maximum demand events already analyzed Failure during installation and removal has been previously addressed by a separate Safety livaluatmn for Control of Test liquipment. This PTM does not introduce any faihire rnxic that would cause a RFPT contml sptem failure not bounded by existing USAR analyses This PTM does not alTect the validity of any assumptmns, cal-lations, pmcedures, or design specifications used to determine margin of safety.
I REPORTABLE SPECI AL PROCEDURES (Spo/ SPEC [AL TEST PROCEDURES (STPs) l I
Note; Some of the SPs/STPs included in the following section wcre performed piior to 1996 but not previously reported.
hPEuL4h TITLE. A-46/IPEEE Screening Verification and Walldown DESCl(IPTION: 1his SP panided the pnoess and enteria to be used to perform the Screening Verification and Walkdown fa plaat equipment and components identdied on the Unresolved Safety Issue A-46 and Individual Plant Examination of External Events (IPEEE) Safe Shutdown Equipment I.ist (SSEL). These walkdowns documented the results of the initial seismic evaluation for each component on the SSE!, and fulfilled commi:ments made by NPPD in response to Genetic Letter 87-02 and Generic Letter 88-20, Supplement 4 SAFETY ANALYSIS: 1his SP provides guidance for conducting a visual inspection of components listed on the SSEL. There will be no physical operatica of plant equipment and no plant hardware or sof1 ware will be altered in any way. Therefore, there will be no increase in the probability of a previously evaluated accident and there will be no etTect on the meration of any equipment relied upon to prevent or mitigate the consequences of an accident. The only potential for a malfunction of equipment is the inadvertent disturbance of a plant canponent. Ilowever, this woukt not represent any increase in the probabihty of a malfunction since the inadvertent disturbance of these components exists any time plant personnel are near the components.
Atkhtional precautions have been taken to lessen the risk ofinadvertent disturbance of plant components by prohibiting any climbing on sensitive plant equipment and by providing formal training that will emphasize the need to exercise care in not disturbing plant equipment, especially instrumentation. Since no equipment is operated, altered, or placed in any condition outside the normal oper ational lineup of the plara, no new types of accidents or equipment malfunctions are created and there is no reduction in the margin of safety as defined in the basis for any Technical Specifications.
STP 90-320 TITLE: liydmpen/ Oxygen Analyzer Pump Performance Monitoring Test DESCillPTION: 1he purpose of this STP was to collect engincenng data on the Ilydrogen/ Oxygen Analyn " umps while insenice. Tlus intivmation was used for performance monitoring to ensure the pumps are not degrading at a rate that could preclude them from remaining continuously operable for a 90 day post-accident time frane lhe S IT' used the pump test methodology fnun an approved CNS procedure. Test results verified that the IlydrogenDxygen Analyzer Pumps are capable of operati:.g for a period of time not less than 90 days SAFETY ANAL,YSIS: The probability of occurrence of an accident or malfunction of equipment important to safety is not mereased because all systems and equipment will be operated within their design basis If an accident or equipment malti.netion were to occur during the STP, the consequences of that event would not be changed since the Primary Containment gaseous concentrations would be monitored by the liydrogen/ Oxygen Analy7er system the same as if the STP were not being perfbrmed During the performance of the STP, the pumps are operated in a manner d tTerent from their normal operating mode since they are operated deadheaded. Ilowever, pump performance momtoring will ensure pump operability during normal operation and post accident. This STP makes no permanent or temporary moddications to any systems or components; therefore, no new accident or equipment malfunction is postulated. STP test frequency should not have an adverse etTect on the llydrogen/ Oxygen Analyzer sysicm since test duration is short and because the STP will ensure the pump can continuously operate for a 90 day post-accident time frame.
56-l
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STP 91 ll5 LT!TII:- ' Installation and Testing of Radiation Monitoring Equipment at PC-PENT XI A and PC-PENT X1B DESCRIPTION:- This STP was performed to test a remote readout radiation monitor installed on each of the Primary _
Containment equipment doors kx:ated at PC-PENT XI A and PC-PENT-X1B for use as a preplanned alternate method of drywell radiation monitoring. The STP collected test data to determine if these monitors would be a succm.ful pieplanned altemative for the failure of the liigh Range Containment Monitors. The twmitors were not found acceptable for long term use.
sal'ETY ANALYSIS: Failure of the temporary monitor would have no elTect on the existing monitors or other safety related -
equipment. This activity does not increase the probability or the consequences of an accident or malfunction of equipment important to safety because the temporary monitors do not interface directly or indirectly with safety related equipment. For the same reason, it does not create the possibility of an unevaluated accident or malfunction or reduce the margin of safety of the plant. The existing monitors that are addresnl in the Technical Specifications are not affected by this activity.
SP 92-043 TITIE: Main Steam Relief Valve (SRV) Removal and installation ,
DESCRIPTION: This SP provided direction for removal and reinstallation of the main steam safety relief valves during an outage for testing / repair / replacement with the vessel head on (Operating State C). Since plant cond tions did not permit the installation of the main steam linc plugs, the piping was temporarily open to the vessel att the potential existed for ikxxling the relief valve piping and the drywell. This SP included requirements / constraints during the SRV removal and reinstallation to minimize the potential forikxxiing SAFETY ANALYSIS: The etTect of this SP on all of the accidents and transients in the USAR was evaluated with the conclusion that the probability of occurrence or consequences of any accident or transient evaluated in the USAR is not increased The following events were determined to be applicable to this SP: Loss of OfTsite Power, Loss of Shutdown Cooling, Shutdown from Outside the Control Room, and inadvertent Emergency Core Cooling System (ECCS) Initiation A total Loss of OfTsite Power event would cause a temporary Loss of Shutdown Cooling; however, station procedures require that the Diesel Generator in the kiop being used for Shutdown Cooling be operable. Therefore, when the dicsci starts, Shutdown Cooling can be reestablished. If, for any reason, Shutdown Cooling cannot be reestablished, calculation has shown that there are 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> before the water in the vessel heats up to boiling. With this much time,
, either Shutdown Coohng can be reestablished, or one of the alternate decay heat removal backup systems can be established. The attemate decay heat removal function can be assured whenever it is required, i c.,
- _ pnor to the water in the vessel reaching 212*F. Since the consequences of a total loss of Olisite Power during the SRV replacement evolution are no different than dunng current plant conditions where the reactor is shutdown and the vessel head is on, there is no unreviewed safety question involved. The Altemate Shutamn Panel has provisions for operation of two SRVs in the event that the Control Room has to be evaeumed (MS RV 7IERV and MS RV 7iGRV). These two SRVs remain available except for a period of time that there could be only one SRV in service during the changeout of 71ERV or 7 IGRV, Ilowever, there is a time period of about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> before the SRVs are required which allows for completing the installation of the spare SRV. There is enple time margin to replace the SRV into the vacant position before it may be required for backup decay heat removal The SP contains precautions that no work will be performed which could possibly cause a ECCS pump start or generate a spurious ECCS initiation signal. It also directs that the Control Room Operator lock out all of the low pressure ECCS pumps except for the Residual llent Removal pumps in the kop aligned for Shutdown Cooling and one Core Spray pump, and to closely monitor vessel water level. The Operator monitoring vessel water level would have alxmt 1.5 minutes to shutdown one Core Spray pump to prevent water from flowing to the dryweh in case of an inadvettent initiation. There are t.o nuclear safety issues associated with the very low possibihty of water from an inadvertent ECCS initiation overflowing into the drywell since the core will be covered and cooled at all times. Ilowever, every precaution will be taken to avoid l' l l
i spreading contammated water into the drywell. The SP is performed such that equipment required to perform a safety functxo is available at the perial in time when that safety function is required. Although the perfmruuxx of this actisity in State C is unique, the work performed on the SRVs has been performed lir the pas 20 years with no advene consecpences to the equipment. The plant is within its design basis duri.ig all of the activity. The only Technical Specification llases which applic.4 to this activity deals with decay heat and the safety evaluation denumstrates that alternate decay heat removal in available when it is required.
SP 92-127 TITI.II: R(actor Recirculation Motor Generator (RRMG) Set Voltage Regulator Adjustment DI!SCRIPTION: This SP was generated to adjust, as necessary, the vohs versus hertz ratio of the RRMO Set voltage regulators to the design value of 70 voltuhertz Adjustments were perfonned for both RRMG "A" and "11" in accordance with the SP, however, post maintenance testing was unable to be completed due to plant conditions At a later date, the tachometer generators for RRMO "A" and "II" were subsequently rephxxd At that titue the proper speed control for RRMO "A" and "B" was ventied. This verification showed that the voltage regulators were working satisfactorily and that there was no adjustment needed for the volts / hertz ratio SAFl!TY ANAL.YSIS: The accident analyses described in the USAR do not take credit for the RRMG sets. Therefore, the probabihty ofoccurrence or conwquences of an accident or malfunction of equipment important to safety presiously evaluated in the USAR are not mercased and the consequences at any RRMG malfunctions are not alTected by this SP. Any potential problems uhich could be created by this SP are bounded by the presiously analyzed Reactor Recirculation pump trip. There is no reduction in the margin of safety defined in the besis for any Technical Specification as the bases only discuss recirculation pump trips.
SP 92-140 Amendment 1 TITI.li: Temporary Placement of AP 101 Cask on Reactor Building 003' DliSCRIPTION: The enginal SP was developed to provide an approved set of guidelines to be used with the loading and handling of the AP 101 cask during the Fuel Pool Cleanup Project. Amendment I was developed to proside mstmetnus for securing the AP 101 cask on Reactor Building 903' w hile it u as temporarily free standing durmg cask ngging evolutions.
SAFliTY ANAL.YSIS. Smcc the cask had the potential to tip over during a Safe Shutdown liarthquake, temporary honzontal restraint was required The cask and restraint system do not interface with any equipment imponant to safety. The restraint system used on the cask prevents it from tipping over and damaging equipment important to safety dunng a Safe Shutdown liarthquake. There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety and the margin of safety as defined in the basis for any Technical Specification is not reduced.
STP 93402 TITill: Determination of Fuel Pool iIcatup Rate DliSCRIPTION: '!he purpose of this STP was to doetunent the heatup rate of the Spent Fuel Pool Storage Pool (SFSP) so an accurate estimate could be made as to how long the Fuel Pool Coohng (FPC) system could be removed from service for maintenance. By isolating the SFSP from the FPC system and installing three thermocouples in the SFSP, a detennination was made as to how long the FPC systc.n could be shutdown before an operating temperature of 120*F w as reached The results of this STP were used to support maintenance work on FPC components.
SAFETY-ANALYSIS- Re Ababihty of occurrence or consequences of an accident or malfunction of equipment important to -
safety previously evaluated in the USAR will not be increased by this STP. This conclusion is based on not allowing the system design temperature to be exceeded nor the SFSP to exceed the maximum temperature used to si/c the Standby Oas Treatment (SOT) heaters. Since system design parameters for FPC and SOT will not be exceeded, the probability of occurrence or consequences of a difTerent type of accident or equipment malfunction are not increased. For the above reasons, there is also no reduction in the margin of safety as defined in the basis for any Technical Specification.
STP 93-107 TITIR Safety Relief Valve (SRV) Stellite! Platinum Pilot Discs DESCRIPTION: This STP replaced the stellite pilot discs in four of the eight SRVs with stellite / platinum pilot discs (MS RV 71DRV, MS-RV 71ERV, MS-RV 711HV, MS-RV 71GRV). The new pilot discs were designed by Ocneral Electric and are made from stellite 6B a!!oyed with 0.3% platinum. The Boiling Water Reactors Outiers Group investigated the industry wide problem of"setpoint driR" with respect to Target Rock Tm>. Stage Model 7567F SRVs and recommended that utilities install the newly developed stellite / platinum pik4 discs into approxunately half of their SRVs to determin whethet the new pilot din material mitigates setpoint drill While operating data was not conclusive, it indicated that ;ne stellite / platinum discs were an improvement. Therefore, MP 93-107 was implemented to nde the installation of these four pilot discs permanent.
SAFETY ANALYSIS: The change of pilot disc material is not considered to adversely impact the design, material, or cmstruction standards applicable to the system or equipment being mo hfied. The new pilot disc will not contnbute to any new rmle ofdegradation. Review of available information indicates that under excess oxygen conthtims, the electro-chemical potential of the stellite / platinum disc could be slightly higher than that for the original Stellite 6B pilot disc. Ilowever, General Electric has judged that the possibility of significantly increasing the corrosion bond strength is low. - The change of pilot disc material does not degrade a 11< .n product barrier, does not change the response time of the valve, and does not increcse any accident i adiation dose calculation or increase onsite or olTsite radiation dose or personnel hazards.
The purpose cf the new piiot disc material is to increase the probability of the proper functioning of the SRVs. The STP will not cause any systems to be operated outside their design or testing limits.
Berefore, the probability of a malfunction ofequipment important to safety is not increased. The change of pikit disc matenal does not change the valves' automatic or manual operation. It does not change the 4 supplier of safety related equipment, seismic or equipment qualification limits, testing time intervals or requirements, or equipment protection features. Therefore, the consequences of ;uipment malfunction are not inercased. The only events related to the SRVs would be either a stuck open SRV or the inadwrtent opening of a SRV. In neither case would such an even' be directly traceable to the addition of a catalyst to the disc material. The change of pilot disc material does not create a new failure mode, does not create a new component or system interaction, and does not cause component / system performance outside the design range. The mat gin 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.
Bis mothfication is intended to increase the potential for the SRVs to actuate at the established setpoint.
liased upon testing by General Electric, the likelihood of the modification to contribute to increased corrosion bonding is low and hence will not reduce the margin of safety as defined in the Technical Specification basis for the operation of the SRVs.
STP 93-146 TITIR Secondary Containment Leak Testing Under Windy Conditions DESCRIPTION: This purpose of this STP was to collect enginecrmg data to be used to develop a correlation between Reactor Huilding ddTerential pressure (dP) and wind conditions. The STP measured Reactor Building negative pressure during Secondary Containment isolation under various atmospheric conditions (wind speed and direction) and compared this to the results of a baseline test with no wind. The Standby Oas
-59 y - m.g.
Taratnx:nt (SOT) subsystem was utilized to maintam the Reactor Building at a negative pressure under windy conditions Multiple tests were perfonned over a two year period during shutdown conditions under a variety of wind conditions. The results of the testing provided justification to perform the Secondary Containment leak test at higher wind speeds. Surveillance Procedure 6.SC.501 was subsequently revised to replace the requirement that the average wind speed during the test be between 2-5 mph with a maximum wind speed of 10 mph.
SAFETY ANALYSIS: All systems and equipment Ull be operated within their design basis. This STP does not introduce any -
conditi(ms which were not considered in the original design of the system. No system or component is nudified by Jtis STP and no new equipment is added. If an accident or malfunction were to occur, the consequences of that ewnt would not be changed since the oITsite dose would be the same as if this STP was not being performed For purposes of this STP, the Reactor lluilding dP will not be allowed to exceed -0.05" to ensure that the Reactor Building is maintained at a negative pressure. The negative ,
pressure ensures that all Reactor Buildmg efliuent is processed through the exhaust plenum and high radiation would be promptly identified by the radiation monitor. Secondary Containment will bc ,
maintained throughout the tests. This STP does not operate any syrtems or components in a manner different fmm their expected operating modes; therefore, no new accident or malfunction is postulated.
Le performance of the SGT system and Secondary Containment system under windy conditions is not addressed by any Technical Specification bases. The Technical Specifications only state that Secondary Containment shall be capable of maintaining 0.25' negative pressure with wind speeds greater than 2 mph and less than 5 mph. %c negatiw pressure requirements at higher wind speeds are unkrunvn and are speculated to vary according to wind direction Therefore, the margin of safety is determined to be unaffected since no systems or components will be modified or operated difTerently from the normal or emergency operating procedures.
STP 93 15813 TITLE: CS-MOV-MO5A DifTerential Pressure Testing DESCRIPTION: ne purpose of this STP was to perform quarterly difTerential pressure testing of motor operated valve (MOV) CS-MOV-MOSA to obtain data to be used to provide assurance that a valve anomaly identified during Generic Letter (GL) 89-10 in-situ testing in the 1993 outage was not degrading. During the referexcd OL 89-10 dynamic flow testing of CS-MOV-MOS A, a short duration loading condition w as observed in the Valve Operation Test and Evaluation System (VOTES) signature of the open to close direction stroke. STP 93-158B was performed three times, twice on the existing valve and once aner the valve was replaced. De first two tests showed that all measured values were within the allowable values and there were no operability concems. Subsequent to these tests the valve was replaced after it failed to completely ckwe during tmubleshooting. STP 93-158B was performed on the new valve to ensure that the anomaly detected on the original valve was not present on the new valve. This test resulted in all measured and extrapolated values being within the allowable target values. The valve successfully opened and closed against the tested difTerential pressure conditions.
SAFETY ANALYSIS: Implementation of specific portions of the STP will be performed when those portions of the affected systems are not required to perform their safety function as govemed by the Limiting Conditions for Operation (LCOs) of the Technical Specifications. Testing and verification in the restoration section of this STP ensures the atrected components will adequately perform their safety and design functions upon completion of this STP. This test will not alTect the design basis of any systems as described in the USAR. The MOV covers that are removed during performance of the procedure do not alTect the ability of CS-MOV-MOSA to isolate containment as the isolation function is required only during a radiation harsh environment. No wires, connections, or mechanical components (other than the covers) are atTected on the actuator or the valve. For these reasons, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety are not increased. The margin of safety as defined in the USAR or basis for any Tecimical Specificatien is not impacted by the use of the test equipmem or the performance of the tests in this STP. All applicable Technical Specification LCOs will be observed during the performance of the STP.
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- STP 93-158B Amendment 1 TITLE; -CS-MOV MOSB Dynamic Flow Differential Pressure Tes;
_ DESCRIPTION: 'ne purpose of this STP was to perform a design basis dynamic flow 2Terential pressure test of motor operated valve (MOV) CS MOV-MOSB. The ditTerential pressure data was analyzed to provide i assurance that CS MOV-MOSB will operate when subjected to design basis conditions. As part of the respume to a CS-MOV-MOSA operabihty concern, NPPD canmitted to inspect and repair, as necessary, .
. both Core Spray Mmimum Flow Valves CS-MOV-MOSB was included in this commitment since it is a similar configuration MOV and the observed anomaly could have been due to a generic concern, A -
signature abnormahty in the open direction was noted on seural of the traces; however, this noted oscillation was not believed to be degrading the cor.dition of the valve, it was concluded that CS-MOV MO5B was capable of performing its design basis function and was considered operable.
Although not a duret result of this STP, CS-MOV MOSB was later replaced with the same type of valve installed in CS-MOV-MOSA.
SAFETY
/ NALYSIS: Implementation of this STP will be performed when the tested portion of the afTected system is not required to perform its safety function as governed by the Limiting Conditions for Operation (LCOs) of the Technical Specifications. Testing and verificatmn in the restoration section of this STP ensures the alTected components will adequately perform their safety and design functions upon completion of this STP. This test will not affect the design basis of any systems as described in the USAR. The MOV covers that are removed dunng performance of the procedure do not affect the ability of CS-MOV MOSB to isolate containment as the isolation function is required only during a radiation harsh emironment. No wires, connections, or mechanical components (other than the covers) are alTected on the actuator or the valve. For these reasons, the probability of occurrence or consequences of an accident or malfunction ofequipment important to salcty are rot increased The margin of safety as defined in the USAR or basis for any Technical Specification is not atrected by the use of the test equipment or the performance of the test. All applicable Technical Specification LCOs will be observed during the performance of the STP.
SP 93-195 TITI.E: Model NLI l/2 Cask Operation and Control Blade Disassembly, Inspection, and Removal from Spent Fuel Storage Pool (SFSP)
DESCRIPTION: This SP provided direction to cut the vehicity limiter olT a used test control blade in the SFSP and load the blade and a nrconium test capsule into a spent fuel shipping cask for transport to General Electric.
The Model NL1-1/2 shipping cask provided a double containment to ship the control blade and test capsule in a dry eminmment. The control blade and test capsule were successfully delivered to Oeneral Electric.
SAFETY ANALYSIS: The cask size, load weight, and path are bounded by previous USAR spent fuel shipping cask analysis.
The cask will be moved by the Reactor Buddmg crane using approved ngging and lilling devices, similar to the previous USAR analysis for spent fuel shipping container movement. The spent fuel cask drop accident analysis encompasses a smaller cask such as the Mcxtel NLI-l /2; therefore, no new failure mode is introduced Since this activity is bounded by the fuel shipping cask analysis described in the USAR, there is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety This actinty does not reduce the margin of safety as defired in the basis for any Technical Specificatan because the cask si7e and path are bounded by the spent fuel shipping cask Technical Specification bases.
fi1P 93 257 and 93 257-1 TITIR Control Romn Pressurization DESCRIPTION: This STP was developed to determine the effects of adjacent ilVAC systems (Turbine Ocnerator fluildmg, Radw aste Buildmg, Contml Baikimg, Muhi-Purpose Facility, and Change Areas) upon the level ofprexamation of the umtrol Room envekipe and to assure that the Control Room pressurization test tuals design basis na:i&nt conditions. Data was gathered to be used in further analysis of the Control Rnun pressurization system Amendment I was written to add steps to temport rily shut off the battery nuun exhaust fans wtule the essential Contml lhiildmg iIVAC is operating and redefired the restriction on wind speeds to a maximum of 35 mph.
ANALYSIS. This S1? does xt unhfy tic operation of any equipment which could initiate rn accident. The Control Room Emergency Filter System (CREFS) is safety related- however, it is used for mitigation of an accident only. This S1/ will not operate the CREFS outside its design intent. Shutting off the battery nxwn exhaust fans will rxt affect the operation of equipment in these rooms due to limitadons on duration ofisolatim arxl ventdation pnwidal by the Essential Contml lluilding iIVAC. Manipulation of adjacen' IIVAC systems will not afTect the conwquences of any accidents as a possible radiological release fro,a these buildmgs, except Radwaste,is not fdiered and therefore it does not matter from where it is released.
- Air flow in the Radwaste lkikhng will be maintained fnun rum-cmtaminated areas to contaminated atras, via manipulation of the dual controllers, to be esbausted through the filtered exhaust train. The only equipment important to safety being alTected is the CREFS and the battery room exhaust fans.
t Maniptdations of adjacent iIVAC will only serve to decrease load on the CREFS. The consequences of a faihire of the CREFS are not changed by this STP and no new equipment failure modes are intnxtuced.
This STi%loes not reduce the margm of safety as defined in the basis for any Technical Specification.
The CREFS will be maintained and operated within its present operation requirement and capability.
STP 94 075 TITI.E: Appendix R Emergency 1.ighting Verification Test DESCRIPTION: The purpose of this test was to venfy the adequacy ofillumination levels of the 10CFR50 Appendix R related emergency battery lighting units installed at CNS. The emergency battery lighting units were testal taler rniuced lighting c(whtions to venfy that adequate lighting exists for station operators to gain access to and perfonn the required equipment operations necessary to safely shutdown the plant in response to a postulated Appendix R fire event This test was also utilized to identify the specific emergency lighting units in each plant area which enable the operator to access required locations and accomphsh tlw required actions. This err igency lighting unit data was used to establish a listing of required Appendix R emergency lighting units and distinguish them fmm other life safety lighting units.
The results of the testmp identified emergency lighting discrepancies which were subsequently addressed by the malification pmcess.
SAFETY ANAL,YSIS: This STP does not alter or require re alignment of any plant systems relied upon to mitigate the cmsequences of accidents evaluated in the USAR. Normal lighting to vanous areas of the plant will be tumed otrwith Shift Supervisor approval. The probability of occurrence or consequences of a previously evaluated accident are not afTected. This test will be performed when the plant is shutdown. Slight discharge ofemergency lighting unit battery charge is otTset by the plant being in a shutdown condition.
The consequences of a malfunctioning energency battery lighting unit during cold shutdown conditions are not alTeeted or increased. No safety related or important to safety equipment is altered or prevented fnun operating. 'the USAR previously es aluated a kiss-of otTsite power. l.oss of normal plant lighting, with Shift Supervisor approval, does not create the possibility of safety equipment malfunction.
Emergency battery lighting units and nonnal nonessential plant lighting are not addressed in the Technical Specitications. All Technical Specifications as<ociated with the plant essential electrical system eqmpment remain vahd and attainable.
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STP 94 100. STP 94 100-1 L and STP 94 100-2 TIT 1.E: Core Sprmy (CS) loop B Flow Tramient Troubleshooting DESCRIPTION: his STP provided for troubleshooting of the unexpected actuation of CS-MOV-MOSB, CS Minimum Flow Valve, during the performance of Surveillance Procedure 6.3.4.2. He test consisted of a flow check of me CS flow transmitters A and B and astrument lines in order to determine if an obstruction existed that had the potential to result in the flow transient which produced the vahe actuation.
Amendment I was issued to increase the scope of the original STP by installing equipment to monitor and record pressure and flow data during the performance of Surveillance Procedure 6.3,4.2 for several ddTerent time constraints. Amendment 2 was issued to further increase the scope by installing equipment to monitor system response to cycling of the A and B CS test line retum valve under various system configurations. Both pressure and vibration data were recorded. Additionally, Amendment 2 was intended to wiify that cycling of CS MOV-MO26B would not cause level perturbations on imtruments associated with tlw Nuclear Boiler Instrumentation reference leg. A modification to the mit.imum flow valve control circuitry was subsequently perfomied which satisfactorily resohed the problem.
SAFETY ANA1.YSIS: The associated CS subsystems will be declared inoperable during testing. Subsystems will be tested sequentially (not simultaneously) in order to comply with Technical Specifications. Amendment I will be performed during scheduled system surwillance testing. Amendment 2 will be performed during cold shutdown, Primary Containment and CS are not required to be operable durmg the performance of Amendment 2. De design basis of the CS system will not be altered during or following this test. STP 94100 and 941001 are considered to be equivalent to authorized surveillance testing. The test will have no c!ket on the function of the CS 110w instrumentation or the CS minimum flow valves following the test. Verifications in the restoration section of the test ensure the affected components are retumed to their original configuration. The removal of one CS subsystem at a time has been previously analyzed in the USAR. This activity does r>ot violate any Technical Specifications and all applicable Technical Specification I.imiting Conditions for Operation will be obsened during the perfbrmance of the test.
Tims, the margin of safety is not reduced.
STP 94-160 TITI.E: 480V Bus IF and 10 Breaker Undervoltage Device Testing DESCRIPTION: This STP prmided guidance for station personnel to perform testing of the undervoltage trip devices installod on 480V Bus IF and 10 feeder breakers. The infarmation obtained was used in the imestyration of previously noted anoraalies associated with 480V undervoltage trip devices identified during the performance of SP 94 208 Tlus STP determined that the undervohage trip devices were unreliable. The undervoltage trip desices were subsequently replaced with shunt trip devices per Design Change 94166.
SAFETT ANAL.YSIS: His STP does not permanently afTect safety related plant systems. The potential transient ofinterest is a kus of shutdan cooling. This STP will test 480V switchgear critical buses one at a time, and will be performed during cold shutdown with shutdown cooling equipment supplied by the unalTected bus. A loss of shutdown cooling due to a loss of the unalTected bm and a single active or passive component failure is considered extremely improbable. Ilourver, the time to boiling in response to a loss of shutdown cooling transient is anticipated to be in excess of 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, which is more than suflicient to abort the test if necessary. Compensatoiy measures will be established to ensure that consequences of an accident will not be increased These measures include an Operations briefing and stationing an Operator kwally to abort the test if requitvd. Appropriate verifications will be employed throughout the activity to ensure that temporary modifications on the critical 480V suitchgear will be removed prior to retuming the mtchgear to senice. Only one 480V critical bus division will be under test at a time and the redundant 480V bus is capable of supplying all kuls necessary in response to any postulated at.cident or transient. This STP will not reduce the margin of safety as defined in the basis for any Technical Specification because it will be performed under cold shutdown conditions and there is more than l sutlicient time to abort the test in the unlikely event of a loss of shutdown cooling.
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l SP 94 163
- TITLil: Circulating Water (CW) Supply Piping Inspection DESCRIPTION: This SP was perlinned to detennine the extent of erosion and corrosion of the underground CW supply piping by visual insp(xtion of the coating and ultrasonic test measurement of the pipe wall thickness.
Results of the testing were satisfactory with no repairs or additional impections required.
SAFETY ANALYSIS: This SP is performed during plant shutdown conditions, the CW system is designed to be shutdown during an outage. lix:re arc isit any accidents desenbed in the USAR that require CW. This SP does not afket any safety relatal systems and results in no changes to pirnt equiprnent. Service Water (SW) will remain operatona! to supply cooling water fmm the river for safety related loads. The riverwell system .
of the CW system pmvides seahng water to the SW pumps llowever, this is not a safety function and this portion of the CW system will remain in operation during this SP. With the CW system shutdown during an outage, there is not an impact on any safety related systems The design and operation of the CW system remains unchanged by this SP. All margins of safety remain unchanged.
SP 94-202 4
TITLE: Primary Containment Walkdown DESCRIPTION: The pinpone of this SP was to assemble walkdown packages for each Primary Contaimnent penetration and conduct a walkdown of each penetration. The walkdowns were performed to assist with the dewlopment of the himary Contamment Design Criteria Document, to support the 10CFR50 Appendix J
' program upgrade, and to ensure that each containment penetration was reviewed as committed in the response to NRC Inspection Report 93-17. The SP consisted of a visual review of the Primary Containment penetrations and piping out to the outtmard containment isolation valve.
SAFETY ANALYSIS: 'Ihis SP dutxts a visual inspection of the containment penetrations only. No equipment is operated and no plant hardware or sonware is ahered in any way. This SP does not authorize any individual to physically alter, marupulate, renove, or duttab any plant equipment. The only potential for a malfunction ofequipment is the inadienent daturbance of a plant comgenent, such as an instnunent. This would not represent any increase in the probabihty of a malfunction, since the inadvertent disturbance of these components exists any time plant personnel are near the com[mnents. Additional precautions have been taken to lessen the risk of an inadvertent disturbance of plant components by prohibitmg any climbing on sensitive equipment and by providmg formal training that will emphasize the need to exercise care in 7 not disturbing plant equipment. The inadvertent disturbance of an instrument cannot create any new malfunction since the only possible resuh is an inadvertent actuation of the instrument or the inadvenent faihac of the instmment. This is within the delinition of a single malfunction of equipment important to safety for w hich the plant has been previously c valuated in the USAR. The walkdowns will not reduce the margin of safety as defined in the basis of any Tecimical Specifications.
STP 94-253 TITIJi: Use of Reactor Water Cleanup (RWCU) for Shutdown Cooling (SDC)
DI! SCRIP ~IION: The purpose of this STP was to pmvide SDC with the RWCU system so that the Residual Ilent Removal (RIIR) SDC subsystem could be taken out of service. The RIIR SDC subsystem was required to be taken out of service to perform a kcal leak rate test of RilR MOV-Mol7 and RI1R MOV htO! 8, to remove RI1R RV 17RV,and to perform electricalinspections on RIIR MOV-MOI 8 Decay heat removal was -
pmvided du.ing this time with the RWCU system in accordance with this procedure to ensure that the reactor coolant temperature was maintained below 212*F.
SAFETY =
ANAL.YSIS. The loss of SDC is not identified as one of the accidents considered in the USAR Ilowever, it is identified as an abnonnal operational occurrence m Appendix G. This STP isolates the normal means of removing decay heat, replacing it with an ahemate system with sutlicient heat removal capacuy to
maintain the coolant temperature within Technical Specification limp.s. Administrative controls in the STP casure that a loss of the alternate system will allow suflicient time to restore the normal means of
- decay heat removal. Furthermore, redundant essential backup systems are available to pro ide SDC through the suppression pool and RIIR Service Water system. The consequences of accidents resulting from a loss of SIX' are less than the consequences calculated in the existing accident analyses published in the USARJ The SW rmofigures the plant to provide SDC through the RWCU non-n: generative heat .
exchangers. Calculatims demonstrate that when operating in this mode, the systems affected by the STP will be operating widun their rmnal ranges orgwration with the following exception. A maximum flow equal to 110% of the tubeside design liow will be allowed through the RWCU non regenerative heat exchangers during the STP to maximize the rate of decay heat removal. This value represents the heat exchanger vendor's acommended mannum tubeside flow for sustained operation without incurring deleterious tube vibration ' The only equipment important to safety afTected by the performance of STP 94 253 would le the Primary Cmtainment isolation desices in the RIIR and RWCU systems. None ofthe automatic safety functions of these components are compromised at i.ny time during the STP, with the possible exceptim of the Ri!R velves subject to kical leak rate testing which are tmder administrative control. Each of these components has a redundant backup available to perform the requiral function.
loss of the RWCU SDC capability would be equivalent to a loss of RIIR SDC. This event has been previously evaluated in the USAR. All equipment important to safety is operated in accordance with existing operating procedures and is used within its normal operating / design limits. Since the function performed by the RIIR SDC subsystem is not considered a nuclear safety function, the Technical Specifications do not address RilR SDC subsystem operability or purameter limitations. Similarly, the filter /demineralizer sections of the RWCU system used to control reactor coolant chemistry which are isolated during the STP are not considered essential and, as such, are not explicitly addressed in the Technical Specifications. It is concluded, therefore, that the margins of safety as defined in the basis of any Technical Specification are not reduced by the performance of thi STP.
STP 94-261 TITLE: Vibration Testing of EE-4160V Breaker DESCRIPTION: & purpose of this STP w as to cycle a nonessential 4160V breaker to obtain seismic test data. Vibration measurements were obtained dunng the closing and opening actuation of a 4160V breaker for the purpose of comparing this normal sen ice load with the loading that would be expected from the knomi response spectrum for the Critical Switchgear Room dunng a design basis seismic event. This data provided input into a seismic evaluation calculation that was performed to assess past operabilisy of a misaligned breaker.
SAFETY ANALYSIS: This STP involves testing a nonessential 4160V breaker during cold shutdown conditions and will not afTect safety related plant systems The breaker will be cycled using existing procedures No changes to plant equipment or procedures are perfonned and all operations are within the plant design basis. The test exercises plant equipment in a manner in which it was intended. All potential equipment malfunauons caused by the performance of this STP will not be ditTerent than that caused by normal operations The added temporary instrumentation will not alTect the operabihty of the breaker. Plant Tecimical Specification requirements will be observed to verify that a Limiting Condition for Operation does not occur.
STP 95-003 i
, TITLE: Operability Testing of SW-MOV 37MV
. DESCRIPTION: This STP provided guidance for testing the operability (automatic function) of SW-MOV-37MV by trippmg pressure switch SW-PS-364B. This STP was initiated in response to a concem that functional testmg of Senice Water (SW) isolation valve SW-37 may not be adequate to assure reliable functioning when actuated by SW-PS-3641L Testing verified that the logic operated as designed.
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1 SAFETY.
ANALYSISr Ris STP involves exercising of equipment in a manner in which it was mtended. It does not modify or permanently affect safety related s>y. ems. All potential equipment failures are coveloped by accidents described in the USAR. All potential equipment malfunctions that may result from the performance of this STP will not be chfferent than those caused by normal operations because the equipment is the same as previously analyval and the malfunctions are the same as previously an.d>7ed. Therefore, there is no increase in the probability of occurrence or comequences of an accident or malfunction of equipment important to safety. The plant Technical Specification surveillance requirements require SW Motor Operated Valve operability testing once per month The Limiting Conditions for Operation listed in Technical Specificaton Section 3.12.C regardmg SW system operability will be monitored during testing.
STP 9%038 TITIE: PC 243AV/244AV Accumulator lloid Time DESCRIPTION: This STP provided instructions to perform a functionallcak test of the Reactor Building to Torus Vacuum Dreaker accumulators PC-ACC-234 AV and PC-ACC 244 AV. The accumulators were tested to ensure they pnnide at least a one hour supply of compressed air which will ensure the Reactor Building to Toms Vacuum Breakers remain closed during the peak Design Basis Accident containment pressure. Initial pc.formance identified the nctd for maintenance, however, tbc STP subsequently detennined that the accumulatcrs provided the required supply of air.
SAFETY ANALYS11 This STP will not nnhfy any system or component which could initiate an accident previously evaluated in the USAR. It will not perform any permanent modifications to equipment used to mitigate the cmsequences of a previously evaluated accident or malfunction of equipment important to safety. The rcrtoratim of temporary modifications will be venfied by post STP testing. The test will be performed during cold shutdown and Primary Containment is not required. This STP does not add any equipment cr.pable ofcreating a ditTerent type of accident or equipment malfunction, nor does it detrimentally alter nny existing equipment's function, operating parameters, service conditions, or accident modes. There is no reduction in the margin of safety as defined in the basis for any Technical Specification since the STP will be performed when Pnmary Containment integnty is not required and no other system addressed in the Technical Specificatiom is n!Tected-STP 90086 TITLE: Ultrasonic Feedwater Flow Assessment DESCRIPTION: This STP was generated to perfonn an ultrasonic feedwater flow assessment under the direction of and usmg equipment supplied by Caldon incorporated. The ultrasonic test fixtures are extemally mounted, non-mtrusive unitt The purpose of the STP w as to detennine if a calibration error exists for the flow nonles. -The installed imtrumentation was dete mined to be sufliciently accurate for continued use in calculating core thermal pov er SAFETY ANALYSIS: Performance of this STP will not increase the probabdity of an accident previously evaluated in the USAR. The pipe surface preparation required to mmmt the ultrasonic transducer will not increase the probabihty of a feedwater line break in the Reactor Feed Pump room, an accident considered in the USAR as part of the " pipe break outside contaiiunent" event. The amount of material to be removed is well within the cale allowaNe margin based m Erosion / Corrosion Program estimates. During the course of the testing, w all thickness will be venfied There will be no interface with, or risk to, any equipment which is safety related or important to safety. All test equipment will be contained in the Reactor Feed thanp nun, which is a Class 11 structure, and contains no equipment which is safety related. Since the STP does not alTect any equipment important to safety, is non-intrusive, and does not interface with any plant contml or indication systems, perfoimance of the STP cannot increase the probability of occurrence or consegmnces of an accident or malfunction ofequipment impoitt.nt to safety. The STP does not atTect any equipment listed in the Technical Specificatiom and these is no reduction in the margin of safety as defined in the basis of any Technical Specification.
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STP 95-110
- TITIR Dynamic Testing of Generic Letter (GL) 89-10 Motor Operated Valves (MOVs)
DESCRIPTION: This STP was generated to,crfbrm dynamic testing of twelve MOVs during REl6 in order to obtain test data that would be used to provide assurance that these MOVs will function when subjxted to their design basis conthtims. This testing was conducted as part of CNS's MOV Program in response to GL 89 10. AITected MOVs were in the following plant systems. Core Spray (CS),lligh Pressure Coolant injection (1 IPCI), Main Steam (MS), Residual I leat Removal (R1IR), Reactor Equipment Cooling (REC),
and Senice Water (SW) All data required by the STP was obtained and the overall test results were satisfactory.
SAISTY ANALYSIS: his STP will only be performed when the afTected system or tested portion of the affected system is not required to perfmm its safety functions as govemed by the applicable Tecimical Specificadon Limiting Conditions for Ogration (LCOs). Performance of this STP will demonstrate conformance with the design bases of the tested components / systems as described in the USAR. Testing and verification performed in the restoration section of the STP casures the tested components are operational upon completion of the applicable test. The applicable test sections of this procedure have been written to ensure testing will rat afket the operation of the spent fuel pool coohng system The brief period of REC flow rahictim to the fuel pool conhng heat exchangers will only result in a negligible increase in fuel pool temperature. Connection of the test equipment requires removal of some MOV covers; however, this does not afket the ability of the tested components to isolate containment, since the isolation function is only requiral during a radiation harsh environment. No wires, connectims, or mechanical components
, (other than the covers) are affected on either the actuator or valve. For the above reasons, there is no increase in the probabihty of occurrence or consequences of a previously evaluated accident or malfunction of equipment important to safety Since the tested components will be tested and verified operational within their applicable design bases, performance of tais STP does not create the possibility of a ddkrent type of accident or equipment raalfunction than previously evaluatal There are no specific Technical Specification requirements for the spent fuel cooling system related to temperature. Pool level is rxx alketed by this test. De requirements of the Safety Evaluation Report for License Amendment 52 are not reduced. Performance of this STP does et reduce the m : gin of safety as defined in the basis for any Technical Specification.
STP 95110 Amendment i TITLE: IIPCI-MOV-M014 Dynamic Test DESCRIPTION: This amendment added liigh Pressure Coolant injection (IIPCI)-MOV-M014 to the scope of STP 95-110 (see above) The results of the testing conducted by this Amendment were satisfactory.
SAFETY ANALYSIS: This test will be performed in accordance with a Station Operations Review Committee (SORC) approvaliIPCI Sun'eillance Procedure and will demonstrate conformance with the design bases of the tested components /systerr.s as desciibed in the USAR. The requirements for Linuting Condit% for Operation (LCO) 3 5 C will be strictly observed prior to entering the LCO to prevent any degradation of the desi;:n safety features of the plant. Testing and venfication performed in the restoration section of the STP ensures the tested component is opurational upon completion of the test. Therefore, this STP Amendment does not increase the probability of occurrence or consequences of an accident or malfimetion of equipment important to safety. Smce the tested components will be tested and ventied operational within their opphcable design bases, performance of this STP Amendment does not create the pomubihty of a duTerent type ofequipment malfunction than previously evaluated. LCO requirements
- will be strictly observed, therefore, performance of this test does not reduce the margin of safety as defined in the basis for any Technical Specification.
STP 95-112 and Revision l_
TITLlh Iligh Pressure Coolant Injection (1iPCI) Exhaust leg Drain Evaluation IESCRIPTION: Operation of the llPCI system indicated that a quantity of water greater than expected may have Ecen accumulatmg in the !IPCI turbine exhaust line. This STP provided for manually draining and measuring the umtent of the !!PCI exhaust hne dripleg The purpose of this test was to: 1) determine the source,
- 2) provide an estimate of the quantity of leakafe into the llPCI steam exhaust dripleg following surveillance testmg, and 3) obtain a rourdmg of the exhaust line pressure and temperature transient. The test w as peribrmed multiple times to evaluate the source and quantity of water present in the exhaust line until appropriate nxxhfications were implemented during kE16.
SAFETY ANALYSIS: The primary concern for perfonnance of this STP is the breach of the Primary Containment boundary.
Ilowever, Technical Specification 3.7.D and Procedure 2.0.2 allow for the manipulation of manual cmtamment botmdary isolation valves within specified administrative controls. The open containment isolation valves will be monitored continuously and closed immediately if plant conditions merit.
Operation of the llPCI system will be unafTected IIPCI and Primary Containment have been demonstrated operable by Operabihty Evaluation 95-031-020, Rev.1. The resultant impact of draining the 1IPCI exhaust ime dnpleg will have a negligible c!Tect on eqmpment operation. Draining the dripleg remotely is a routine operatiort Manually draining the dripleg will have a similar elTect on the system and not impair the 1iPCI pump in any w ay. Primary Containment will be maintained administratively as per Technical Specifications. Any potential accident resulting from this activity is bounded by the USAR.
The bases of the Technical Specifications allow manipulation of Primary Containment valves when administratively controlled Operation of1IPCI will be performed within the bounds of the Technical Specifications.
STP 95-121 TITLE: Testing of Residual lleat Removal (R1IR) and Core Spray (CS) Motors DESCRIPTION: This STP perfonned tests, inspections, and measurements of R1IR and CS pump motors to provide input to the Motor Condition Assessment (MCA) program for the Emergency Core Cooling System (ECCS) motors. The pnmary purpose of the ECCS motor assessment was to evaluate the need for, and potentially defer, the 10 year intmsive inspectiort Static and dynamic inspections and measurements were performed under the dm:ction of General Electne persmnel. Results of the STP wcre used to help assess both short-tenn and long-tenn maintenance requirements lbr the subject ECCS motors.
SAFETY ANALYSIS: This STP does not mothfy or change the operation of any equipment which could initiate an accident or afTect the consequences of an accident. Performance of this test will be done in conformance with the Technical Specnications. Limitmg Conditwns for Operation will be entered for each system that is tested per this STP. Only one system in one daision will be tested at any one time. This test is being conducted to assess the conditmn of the RilR and CS pump motors and does not change any plant equipment or manner of plant operation; therefore, the probabihty of a malfunction of equipment important to safety is not inerensed This STP requires only system lineups identical to those performed by approved system operating procedun:s. The testing is non-intrusive and can be performed with a minimum impact to the motors or plant personnel This STP does not create the possibility of an accident or equipmen' malfunctim of a dmerent type than previoust, evaluated No setpoint changes or equipment changes are introduced As the STP will be performed in accordance with the Techmcal Specifications, there is no reduction in the margin of safety SP %053 TITLE: Model 2000 Transport Package Ilandling DESCRIPTION. This SP prouded operations and mamtenance instructions for the General Electnc Model 2000 Transport Package. The shipping cask was to be used for the transport of material specimens to General Electric l
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for testing. The cask w as loaded in accordance with the SP, however, radiation levels on external cask surfaces prohibited shipment of the cask The cask was subsequently unloaded and shipped back to Ocneral Electric empty.
SAFETY ANAL.YSIS: The Model 2000 Transport Package is well within the rated ha capacity of the main Reactor Buildmg crane. The same restncted mode ofcrane operation that is designed for use when handling the spent fuel shipping cask will be involved when transporting the Model 2000 cask. This SP does not involve any changes to systents/ components or involve any unusual system operation that could initiate an accident.
This SP invokes an approved pmcedure to assure that Secondary Containment integrity is maintained while trampating the Model 2000 Transport Package through the Reactor Building's Railroad Airlock thr. The radiological consequences of a Model 2000 cask drop accident are enveloped by the IF.300 spent fuel cask drop accident which has been previously evaluated in the USAR. This SP does not alTect the function of a,:cident mitigation systems / components, nor does it prevent personnel from performing functions which may midgate the consequences of an accident. This SP operates equipment well within its designed limits and per its intended function. It enforces the Technical Specification surveillance requirements for spent fuel eask handling by invoking the perfonnance of approved pmcedures or by specific instructional steps. In the event of a plant emergency, the SP instructs persorinel to discontinue the evolution by lowering the cask to a safe position which is consistent with the basis for Technical Specification 3.10.11(Spent Fuel Cask llandling) The SP ensures that Secmdary Containment integrity is maintained during this evolution. For these reasons, this SP will not reduce the margin of safety as defined in the basis for any Technical Specification.
SP %-073 TITLE: Spent Fuel Pool irradiated liardw are Pmeessing DESCRIPTION: SP %073 provided an apprmed pacedure to volume reduce and process CNS spent fuel pool irradiated hardw are and other radioactive weste materials stored in the spent fuel pool. Waste processed included control mds, guide tubes, source range momtors, intemiediate range monitors, local power range monitors, shroud head bolts, filters, vacuum equipment, hghts, tooling, and miscellaneous small irradiated components and component hardware. An experienced spent fuel pool waste processing vendor was utih/cd to volume mluce and process the radioactive waste. Removal of the radioactive waste stored in the spent fuel pool enhanced pool cleanliness and housekeeping, thereby minimizing the potential for foreign matenal intrusion during refueling operations. The potential for unplanned or unnecessary nuhation exposure was also reduced by removing the highly activated radioactive waste from the spent fuel pool.
SAFirfY ANAL.YSIS: The trradiated hardware processing equipment is well within the rated lin capacity of the main Reactor Building crane. This SP does not involve any change to system components or involve any unusual system operation that could initiate an accident. The SP invokes an approved pmcedure to assure that Secondary Containment integnty is maintamed while transporting the irradiated hardware processing equipment through tlc Reactor Building's Railroad Airlock Door and while handling this equipment on the refuel fkior and spent fuel pool. In the event of a plant emergency, this SP instructs personnel to discontinue the evolution This SP does not alTect the function of accident mitigation systems / components, nor does it prevent personnel fmm performing functions which may mitigate the consepences of an accident. SP %073 operates equipment well within its designed Inuits and per its intended fauction. The use of approved procedures reduces the probabihty for personnel error and assures proper operation of the Reactor Buildmg crane. The consequences of a Reactor Building crane failure dunng an irradmted hardw are processing equipment lift are enveloned by the crane failing during an IF 300 spent fuel cask hil The conseuuences of a radioactive material release as a result ofirradiated hardware processing equipmtat mahanction or operator error are enveloped by the radiological consequences of the previously evaluated fuel handling accident. This SP does not operate the plant in a manner not described in the USAR. It does not operate equipment in a manner inconsistent with its intentkxl function or beyond its design hmits This SP enforces the Technical Specification surs eillance requirements for spent fuel cask handling by invoking the performance of approved procedures or by specific instruction d steps It ensures that Secondary Containment integrity is maintained dunng this
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emlutiort In additim, this SP has no impact upon the operability of the Standby Gas Treatment system.
Therefore, this evolution will not reduce the margin of safety as defined in the basis for any Technical .
Specification STP 96-084
- TITI.tk Determinatim of Radio Frequency Interference (RFI) by I lello Direct Wireless I leadsets in Control Room DI! SCRIPT 10N: This STP was performed to determine experimentally if any radio frequency induced interference with Contml Room instrumentation would be caused by the use ofIIello Direct wireless telephone headsets, ulnioperated in the Control Room The use of wireless headsets m the Control Room would enhance a day-tulay as well as emergency respee activities. The test was successful as there was no interference with any Cmtrol Room instrumentation. Ilowever, the need for the headsets was subsequently overcome by the mstallation of a cellular telephone system.
SAFETY' ANAL YSIS: 1his test is to be performed during cold shutdown The only credible f'tilure that could result from this testing is a tempinty disruption of electronic component function in the Control Room. This dismption would only be a temporary phenomenon and would not permanently alTect the equipment's perfamanec/ functionality. The disruption of Cont.;l Room equipment operation is very unlikely and due will be Operators stationed at crucial areas to provide corrective actions in the event of equipment problems. This STP will require only a short time to perfonn No process condition or electrical conditions or equipment functions will be changed. The possibility of a malfunction of safety related equipment is extremely remote and is a temporary effect only that can be controlled effectively by aborting the test. Testing will be performed very cautiously so as to detect any imminent equipment operational pmblems prior to actual malfunction ofequipment. Testing of the subject wireless headsets cannot cause an accident based on their potential failure mode and plant status. The temporary disruption of Control -
- Room electronic equipment cannot reasonably initiate a failure of redundant safety related equipment in the Contml Romn or result in any significant reduction in safety. This activity will not alTect the margin
- of safety as dermed in the basis of any Technical Specification All Emergency Core Cooling Systems (ECCS) are redundant designs and thuefbre not subject to a common mode failure due to this testing.
4 SP 06-086 TITLE: Reactor Pressure Vessel (RPV) 1Icad 1.cak Test DESCRIPTION: This SP provided instructions ihr the performance of a nominal operating pressure leakage test of the RPV head llange and head vent pipe flanges which were disassembled and reassembled during the June % fon:cd shutdown The pressure test was performed with the reactor shut down and complied with the pressure-temperature limits described in the USAR and Technical apecifications- Primary Containment identified and unidentified leak rates were below specified limits and there was no leakage identified at the RPV head flange and head vent pipe flanges.
SAFETY ANALYSIS: Ilydrostatic pressure testing is not a precursor to any accident described in the USAR. In addition, the pressure test is performed in accordance with temperatu*c-pressure limits for the reactor vessel specified in the USAR and Technleal Specifications, and at pressures well below the design pressures of attached piping pressurized during the course of the test. The Main Steam relief and safety valves are available to provide pressure relief in the unlikely event that al1 other metimds of pressure control are lost. The worst possible consequence of this test would be a leak in the reactor coolant pressure tvundary. Any through wall Icakage would be bounded by the accident analysis since the test is performed at a lower temperature than operating conditions, with control rods inserted, and with decay heat levels significantly lower than immediately fbliowing a reactor scram. RPV temperature shall not exceed 250*F to ensure that the Shuttkwm Cooling System can be retumed to serview af ter vessel depressurizatio t Although the Reactor Recirculation (RR) pump speed limiter 20% feedwater flow interlock has been defeated, RR pump operation will not be adversely atTected. Test temperatures are much lower than the operating temperatures where RR ptanp cavitation may occur if adequate feedwater tiow is not available to provide adequate Net Positive Suction Ilead to the pump. Iligh Pressure Coolant Injection and Reactor Corv Isolation Coohng kiw steam nressure interkx:ks are defeated to permit draining water from the steam lines
-70 y,. _ _ - -m n _ , _ - r-_o _ _
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I 4
folkming depressurization Since these interkicks are defeated while depressuriicd and in cold shutdown, the probability ofequipment malfunction is not incrersed Precautions are included to specify actions if 3 RR pump cavitation occurs or the RR pumps trip, AtTected systcms are operated within design limits, and all Technical Specificatiom, are met during the test. RR pump speed has been limited to 50% to eliminate vibration problems which may occur at prenter than 50% in relatively cold water.- The pump and vahr logic temprarily defeated by this pnmlure doci not impact the operability of Emergency Core Coolmg Systems required fot this plant condition Sp %090 .
TITI.E: Spent Ftel Pool Irradiated Ilardware Shipping DESCRIPTION: SP 96-090 provided an approved procedure to load and ship CNS spent fuel pool irradiated hardware and other radioactive waste simed in the spent fuel pool to the Barnwell, South Carolina, waste disposal facility. Wr,ste shipped included control rods, guide tubes, s( urr.c range monitors, intermediate range nanitors, kcal pmer range monitors, shroud head bolts, filters, vacuum equiprnent, lights, tooling, and miscellaneous smail trrathated components and component hardware. An experienced spent fuel pool waste processing vemkr was utilized to load and ship the radioective waste. Waste classification was also perfermed 1:y the wrdir using RADMAN sollware codes. Re noval of the radioactive waste stored in the spent fuel pool enhanced pool cleaaliness and housekeeping, thereby minimizing the potential for foreign material intrusion during refueling operations. The potential for unplanned or unnecessary radiation exposure was also reduced by removing the highly activi,ttd radioactive waste from the spent fiel pml SAFETY ANALYSIS: The shipping casks and shielded container are well w; thin the rated lif1 capacity of the maia Reactor Iluilding crane. This SP does not involve any change to system comnments or involve any unusual system operation that could initiate an accident. The SP invokes an approved procedute to assure that Secondary Containment integrity is main'ained while transporting the shipping casks and shielded container through the Reactor Building's Railmad Airlock Door and whik handling this equipment on the refuel floor and spent fuel pool. In the event of a plant emergency, tho SP instructs personnel to discontinue the evolution. This SP does not alTect the function of accident mitigation systems /componente nor does it prevent personnel from performing functions which may mitigate the consequences of an accident = SP %090 operates equipment well within its designed limits and per its intended function '!he use of approved procedures reduces the probability far personnel error and assures proper operation of the Reactor fluilding crane. The consequences of a fteactor Huilding crane fadure dunng shippmg cask or shielded container hft are enveloped by the crane faikng during an IF.300 spent fuel cask lift. The consequences of a radioactive material release as a result of s drorped shipping usk or shielded container, crane failure, or operator emy are envch3 ped by the radiological consequences of the previously evaluated fuel handling accident. This SP does not operate the plat.t in a manner not desenbed in the USAR. It does not operate equipment in a manner inconsistent with its it tended function or beyoniits design hmits. This SP entbrees the Technical Specification surveillance requirements Ihr spent fuel cask handling by inmking the perfbrmance of approved procedures or by specific instructional steps. It ensures that Secondary Contaimnent integrity is maintained during this evolutiott in addition, this SP has no impact upon the operabihty of the Standby Gas freatment system. The efore, this evolution will not reduce the margin of safety as defined in the basis for any Technical Spect'ication SP 96-092 TITLE: Diesel Generator (DO) #2 Starting Nr Receiver Inlet Check Valves Leak Test DiiSCRIPTION. This SP performed a leak rate test of the DG#2 Starting Air Receiver Inlet Check Valms to determtae the air rrce:wrN ability to retain suflicient air for muhiple starts of the Diesel Generator without the att compressor being operable The test was performed when DG#2 was out of service. The information obtained from this SP was to be used to determine if the DG#2 Starting Air Compressors could be reclassified as nonessential. Results of the SP were inconclusive.
W 71
-SAFETY ANALYSik The operability of the Diesel Generators is not a precursor to any accident described in the USAR. In addition, this test is perfonned whea DO#2 is not required to be operable. DG#1 will be operable throughout the test. The two diesels are physically and electrically separated, and a malfunction in the components being tested would not aflect nuclear safety. A loss of o!Tsite power during the test would have the worst ptential omsequerres. Ilowver, the consequences of any accident occurring during tids test are bounded by the existing analyses. W implications of an inoperable diesel have been presiously evaluated Operatior with one diesel out of senice is permitted by the plant Technical Specifications for up to seven days. This test will make the diesel unavailable for less than one shift Afrected systems are operated within design limits. No pcmmnent changes in the system configuration or operating conditions -
are introduced by this test. For these reasons, there is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety and there is no reduction in the margin of safety.
SP 96-133 TITIR Residual Ileat Removal (R1IR) Division 2 Flow Perturbation Evaluation DESCRIPTION: His SP was generated to monitor the RIIR Pump D suctim annidischarge pressure. Pressure transducers and instrumentation were connected to the R1IR system to obtain the required data. This testing was performed to help determine the cause of an apparent flow anomaly identified during surveillance testing.
The previous apparent anornaly could not be reproduced.
SAFETY ANALYSIS: The installation of the test equipment will not adversely alTect the RIIR system or associated instrumentatim. Personnel will be stationed at the kcation of the transducers so that if a leak is detected it can be immediately isolated he test equipment will be used to gather data and will not be used to actuate plant equipment. If an accident were to occur while the test equipment is installed, the system will respnl the same as if the equipment were not installed The test equipment can be isolated and removed while the system is in operation without afTec6ng the system. The point of the pressure taps is outside the Prinwy Containment boundary. The maximum pressure rating of the pressure transducers exceeds the design requirements of the spent The installation does not alter the design of the plant in a way that would intaduce any accid nt initiators or new failure possibilities. An evaluation was performed which detemuned there wcre no seismic concems. The SP nuutors the pn ssun: in a system important to safety, but will not intaduce any new faihire modes that will impact the safety equipment. As the perfonnance of the R1IR system or other systems will not be alTected by this test equipment the margin of safety is not reduced.
SP %1002 TITLE: PC-REL-RMAX Contact 2 and PC-REL RMBX Contact 5 Replacement DESCRIPTION. This SP was pvpaml to obtain as found whage and resistance values for PC-REL-RMAX Contact 2 and PC-REl-RMBX Comact 5 which were suspected to be the potential cause of a Standby Oas Treatment Train"A" initiation without a full Gmup 6 initiation. The procedure also contained steps to replace the subject contact (s) if they were shown to be degraded PC-REL-RMBX was subsequently replaced per this SP.
SAFETY ANALYSIS: The work perfonned by the SP does not alTect the initiators of plant events previously evaluated in the SAR. Safety actions will not be prevented from occurnag as a result of actual inputs or by the maintenance being performed. This activity involves a decrease in probability of a =nurious actuation of equipment imp xtant to safety. Unplanned actuations of plant equipment would be the result of a single failure or operator error for which the plant is designed and evaluated The long term design is not changed Although a portion of the circuit will be disabled, the safety function of that circuit will have been perfonned T he safety functions will not be ahered in such a way that could prevent any systems j_ - fmm peribrmirg their designed action. The maintenance perfoimed by this SP does not introduce any i new or ddTerer.t initiators; Group 2 and Omup 6 actuations are design features of the plant. The potential l
l -7 2 -
isolation of the dqwell radiation monitor is not consideted a redection in margin of safety because ahematiw means are available to detemune drywellleakage. Additionally, the monitor could be returned :
to service as soon as the isolation is reset.
SP 97,001 TITIJL Drywell Fmi Coil Unit (FCU) 1.iquid f loidup Verification DESCRIPTION: This SP provided guidance to validate that the Drywr11 FCUs are holding suffnent quantities ofliquid to create a pulse drain to the Drywell F Sump. This infortnation was gathered to validate the apparent cause of a Condition Adverse to Quality (CAQ %1049) which identified a pulsating phenomena to the F Sump. The CAQ evaluation determined that the phenomena was the result ofliquid holdup in the condensate pans of the Drywell FCUs. The SP determin-d that the FCUs will hoM water up to a point and then dump the u ater to the F Sump. EE E039 was subsequently generated to install loop seals in
<- the drainhnen of the Drywell FCOs.
SAFirrY ANAL.YSIS: Design, configuration, and function of the F Sump and FCUs remain unchanged. F Sump is isolated on a Group 2 isolation signal. Drywell ventilation is nonessential and is designed to trip off during a loss ofpmver,on reactor km low low level, and on drywell high pressure. Response to indications of sump fill and increased leakage are not changed by this SP. There are no accident initiators or contributors associa:cd with this equipment. The F Sump is designed to collect drainage in the drywell and pump it to the Radwaste system. Pumpmg of the F Sump pumps is not a safety related function. Performance of this test requires tripping of a FCU, however, past operation with three FCUs at power during the summer months with outside temperatures frequently in excess of 95'F did not alTect any eqwpment important to safety. This procedure has no impact on any other equipment or components that could increase the probabihty ofoccurrence w consequences of a malfunction of equipment important to safety.
De wurst case scenario that could result from this test would be the loss of the use of one 1 CU. For the expected implementation period, ambient environmental conditions will preclude a significant buildup of heat and humidity in the dqwell. Drywell temperature and humidity are rnonitored. This scenario is bounded by the steam line break event. No permanent changes are being made to plant components or processes. The Operators' response to indications of increased leakage or F sump high level is not changed. The margin of safety as dermed by the basis for Technical Specification 3.6.C. Coolant 1.cakage,is not changed SP 97 002 TITI.E; Vibration Measurement of EE-CV-4160G(SWP1D) Tophat DESCRIPTION: This SP was developed to collect data to determine if the vibration generated as a result of a 4160V Magne Blast breaker operation could have an impact on the relays locaLxi within the tophat, specifically EE-Cll-(27X16-1G). Vibration equipment was connected to the tcphat of EE-CB-4160G(SWP1D).
Based on the results of this test, it does not appear that the operation of the breaker has a significant impact on the operation of the Agastat relays; however, repeated operation over a given peiiod of time n,ay have a cumulative etTect of causing dnft SAFETY ANAINSIS: The change within the Seniec Water (SW) System is restricted to a change in the number of SW pumps in operation over a fixed perial of time of approximately two hours. There will be a temporary pressure increase but this is expected with the start of another pump. This will not adursely impact the operation of the SW system and will temporarily add flow and cooling for this short duration The operation of the SW Pump iD is under normal operatmg proceduns No change in the method of operation, valve lineup, or any other characteristic of the pump operation is affected. Due to the short duration, pressure increase is transitay and no throttling adjustments are riecessary. A two hour wait time has been specified within the SP between the two one-minute durations in w hich the SW pump is actually being operated to allow the SW system and breaker to duipate ey transi:nt efTect generated by the one minute run. This conforms with Operations Procedure 2 2.71. The test involves placing a vibration probe on the tophat back wall which is hekt in place by a flat magnet. Should the probe or mag:,et come loose, it will fall on 73 I
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the tut un of the tophat. Ecre are m components attached to the bottom of the tophat with which it can l come in contact There is also no equipment directly below the probe with which it can come in contact.
The test equipment wiP he located in the 41600 bus switchgear room a suflicient distance from the
. 41600 switchgear to not interfere with operation of other electrical equipmentc The attachment cf the snagnet, probe, ami cord do re impact the operation of the breaker and t.2erefore does not create the pachitty of a ddTerent ty pc of malfunction of equipment importan; to safety. The vibration equipment is a padve desice only for Ihe purpose of data gathering. The margin of safety is not reduced because an additional pump will be providing additional flow thereby slightly reducing operating temperatures
, to Reactor F.quipment Cooling, Iurbine Equipment Cooling, and Dicael Generator heat exchangers for the shcut duration of the test The required natimum numter of pumps will always be available.
SP 9b104 -
TITIJL Vendor Test for Cellular Pbne Coversge DESCRIPTION: his SP performed tesdng of the proposed wireless communication system to determine appropriate base ,
l unit coverage. This required the vsmdor to survey the plant using instrumentation that emits a radio frequency. The SP was satisfactorily performed with no interference encountered.
SAFETY
" ANALYSIS: This testing will not directly interface with any plant systems. The only credible electmmagnetic/ radio '
I frequency interference (EMl/RFI) effect on electronic control equipment in general u ould be a temporary disruptim of electrts component function. The emissive power output of the test units is very low and within the Electric Power Research Institute guidehnes established for equipment susceptibility. In additim. tests haw tren successfully performed at other plants using similar equipment, but with higher emissive power output, with no adverse impacts on equipment operation. No process or electrical cmhtions or equipment functions will be changed as a result of this SP, All safety related equipment will be able to perform its intendal function in mitigating accidents As a precautionary measure, the test units will not be permitted to be kicated within three feet of sensitive equipment. In addition, Operators will eswrt the contractors performing the test, where necessary, to assure this restriction is followed and to perform any requinxi actims that nught result from any EMI/RFI cffects induced on the equipmentc Any dismption would be only a temporary phenomenon and would not permanently affect the equipment's perfmnance and can be controlled effectively by aborting the test. Sullicient confidence exists that the possibility of a malfunction of safety related equipment will not occur. Consequences of equipment failure will not be changed by this SP. No accident initiators are involved with this testing. No testing activity a!Tects any USAR evaluated accident or the plant response to the accident. This activity will not reduce the margin of safety as delbed in the basis for any Technical Specifications. All Emergency Core Cooling systems are redundant dedgns and therefore not subject to a common male failure due to this testing.
SP 97-005 TITile RMV-RM-1 A Time Mode Filter Change DESCRIPTION: The Cmtrol Room Radiatim Monitor, RMV-RM 1 A, uses a digital filtering technique called time mode littering to filter noise out of the radiation measuremen'. The previous time male futer setting for the particulate channel was 5 minutes. Calculations determined that the 5 minute settmg was nonconservative. The purpose of this SP was to lower the time mode filter setting for the particulate channel from 5 minutes to 6,4, and 2 seconds and document the background radiation levels for each setting. Engineering Judgement 97-113 subsequently determined that a particulate channel time male filter setting of 6 secoals would provide adix[uate protection for limiting dose to Control Room personnel, and the setting was adjusted accordmgly. The SP also verifici that the reduced time mode filter setting will not result in spurious trips.
SAFETY t ANALYSIS: The Control Room Radiation Monitor does not impact any of the equipment assumed to initiate or mitigate plant events described in the USAR. The purpos: of the mormt is to limit the radiation exposure to Control Room personnel. Since the monitor will remain operable during the SP, it will
+
ccotinue to prtnide this protective function. This SP only impacts the particulate channel which will be declared irxperable during the test The gaseous and iodine channels will remain operable which causes a tiu nxmita to be operable. The gaseous and iodine channels are not impacted by this SPc The monitor will be able to perform its safety function if called upon; therefore, there is no reduction in the margin of safety as defmed in the basis for any Technical Specification.
SP 97-010 TITIR Testing of Permanent Cellular Phones DESCRIPTION: nis SP performed pre-installation testmg ofcellular phone equipment in order to demonstrate that the 1
ogration of the proposed cellular phones will have no effect on the operation of plant equipment. The -
teleptures were held next to a representative array of the most sensitive plant equipment for a duration of five minutes to check fw any alarms or anomalies. His SP was performed during shutdown conditions ned in constant conununication with the Control Room. All of the equipment tested satisfactorily with the :ellular phone equipment and there were no alarms or anomalies in equipment operation.
SAFETY ANALYSI"' This T *.tl le perfonned during reactor shutdown during which the only credible events involve ron-transient scenarios. No plant event initiators enn credibly be affected by this testing. The potentially affected equipment will te carefully monitored and any effects on the equipment will result in immediate .
cessation of the testing. Suspending this test as a result of the occurrence of a plant event will climinate any effects due to the testing and normal equipment operation will be restored. All originally operable plant equipment will be able to respond a.: required. The testing in this SP cannot credibly initiate the permanent degradation of plant equipment. This activity does not change any existing plant systems, sinntures, or compments. Wurst case, it may induce electrical noise in nearby conductors or components resulting in a spurious tnp or actuation. The response to, and the consequences of, these actions remain tarhangett and are as previously analped No new malfunctions are introduced. It is unlikely that any sperinas act ion will occur during this SP as equipment sensitive to electromagnetic / radio fiequency interference (EMI/RFI) utilizes shielded conductors to minimize the effects of EMI/RFL This SP does not involve any parameter in the basis for any Technical Specification.
i OTIIER REPORTABLE AC'IIVITIES lievi< ion to OITsite Dose Awessment Manual (ODAM) llTLis: Sampic Stations in the ODAM Dl?SCRIPTION: This revision changedredefined sample stations in the ODAM. These changes were due to wasonal conditions and to reflect requirements specified in Technical speifications !br projected sector dose.
Alm, the Semi-Annual Operatmg Report for Radioactive !!!11uents was changed to an Annual Report in accordance with License Amendment 172.
SAFI!TY ANALYSIS: Sample locations cannot initiate an accident described in the USAR. These changes are administrative in nature and do not afTect the accuracy or reliability of elliuent, dose, or setpoint calculationt Changes m samp!c hwations do not affect the consequences of an accident nor challenge /utTect components which have un accident mitigation function 1hese ch,mges cannot alrect the operability of equipment described in the USAR. Tlwse changes comply with 10CFR20,10CFR50, and 40CFR190 w hich are the bases for the Technical Specifications therefore, the margin of safety is not reduced Operntme I icense Chance Request (OLCR196 f04 TITLE Correction oflirrors in Technical Specification llases Dl!SCRIPTION: This OLCR was generated to correct typographical errors in License Amendment 175 as issued by the NRC (Tech Spec pages 131,200,201) and to delete a paragraph which w as inadvertently overlooked during the technical review of the p6oposed change (Tech Spec page 199). Corrections were made to liases Sections 4 5,3 9, and 4 9 only.
SAFl!TY ANALYSIS, Changes to pages 131,200, and 201 are typographical errors and editorial in nature. The paragraph deleted from page 199 is no longer applicable us a result of NRC approval of Amendment 175. This change was addressed by the significant hazards evahution submitted with the amendment per 10 CFR 50 90 Therefore, the probability of occurrence c: consequences of an accident or malfunction of equipment are not increased by this OLCR and the marg'n of safety is not reduced Operatine I icense Change Recuev (OLCRT %a06 TITI.l! Revision to O!Tsite Dost Assessment Manual (ODAM)
Dl!SCRIPTION: This revision changed a sample station m the ODAM This change retlects the requirements specified in the Technica! Specitications for projected sector dose. Last year's envimnmental data was evaluated requirmg a change to the broadleaf vegetation sample station sal'ETY ANALYSIS. Sample hications cannot initiate an accident described in the USAR. This change is admmistrative in nature and does not atreet the accuracy or rehabihty of efiluent, dose, or setpoint calculations. Changes in sample hicatmns do not alrect the consequences of an accident nor challenpe/alTect components u hich have an accident mitigation function Thi* ch;mge cannot alTect the operability of equipment described in the USAR. This clunige comphes with 10CFR20,10CFR50, and 40CFR190 which are the bases for the Technical Specifications, therefore, the margin of safety is not reduced Operatme I Lense Chance Reauest (01 CR) %008 TITI.l! Revision to Process Control Program (PCP) Docunient DESCRIPTION This revisien deleted references that no longer exist or wear not used in the development of the PCP document, and changed the semi annual reporting frequency of the radioactive materials release report to annual per License Amendment 172.
t SAFETY ANALYSIS: The PCP is not assumed in the imtiation of an accident nor is it relied upon to mitigate the consequences of an aa:ident previously evaluated in the USAR. The references being deleted have been incorporated into the remaining PCP references, or were not used in the developmmt of the PCP. Changing the reparting fnxtuency fnen semi-annual to annual was prcviously evaluated under License Amendment 172.
The subject change does not involve any physical, procedural, or programmatic changes to structures, systems, w components (SSCs) or the manner in which these SSCs are maintained, modified, tested, or inspected Therefore, there is no increase in the probability of occurrence or consequences of a malfunctim ofequipment impanant to safety and the probabihty of a new type of accident is not created.
This change does not affect the margin of safety because the PCP is not assumed in any safety analysis margin and current safety analysis assumptions are maintained.
Oocratine Liceme Chance Reauest (01 CR) 96 009 TITLE: Rcvision to OITsite Dose Assessment Manual (00AM)
DESCRIPTION The ODAM was revised to change "MPC" to " effluent concentration" and to correct references to 10CFR20.1302 per Amendment 174 to the Technical Specihcations. This change also deleted the nx:tlux! for calculaung gaseous effluent nxnitor setpoints based on concentration. The nirrent calculation method is based on release rate.
SAFETY ANALYSIS: The ODAM describes actions to be taken in respmse to an accident to assess concentration and dose estimates shoukt a release occur. Changes to the ODAM have no elTect on the probability of an accident or malfunction of eqtupment important to safety. The proposed changes do not affect radioactivity concentratim or dose estimat.: calcu' ions and do not alTect the margin of safety. The attemate method to calculate gaseous elliuent monitor setpoints being deleted is not used and cannot be used without a design change to the monitors Doeratine Liceme Chance Reouest (OLCR)06-011 TITLE. Revision of Technical Specification Bases to Change Testing Acceptance Criteria DESCRIPTION: The Ibes for Technical Specificatiom 4.7.H and 4.7.C (Tech Spec page 183) was changed to allow for detectable leakage amund Standby Gas Treatment (SGT) System access doors to be added to leakage fnun other filter bypass sources to detennine an overall filter efliciency of not less than 99 percent. The previms wtrding indicated that "any detection" of DOP was unacceptable. Modern equipment is much more sensitive than previous testing equipment and can detect leakage less than 0.001% This change reflects a commitment revision and is also included in the Commitment Change portion of this report.
SAFETY ANALYSIS. The SGT system is not an accident initiator. Performance of the testing is not t. hanged by this revision and the likehhood of a plant event is not alTected The SGT litter etTiciency requirements for accident nungauan are not changed by this revision and no increase in dose consequences as previously analyzed will occur. The SGT system is required to be tested to 99% efticiency to ensure mitigabon of the com:qumees of a Design Basis Accident or Fuel Ilandling Accident. Failure of the SOT to perform this rJtigation function is outside the design basis and has not been evaluated in the SAR. The only credible malfunction is the loss of the integrity of the duct and housing of the SGT due to faihne of the gasket amurwl the access door which would result in unacceptable accident consequences. The door and gasket mil continue to be tested and inspected and the possibility of this type of malfunction is unchanged The basis for Techment Specification 3.7.H is preserved by ensuring 199% iIEPA filter efliciency.
l l
Operating i icense Channe Reauest (OI.CR)97-003 TITLE: Revision to OfTsite Dose Assessmcut Manual (ODAM)
DESta!PTION: This revision nuhfied sample stations in the ODAM related to font product-broadleaf vegetation sampling. These changes were made as a result of the annual land use census and emironmental monitorire meteorological program evaluation. Changes were made in accordance with Technical Specification 3 21.F.1.
SAFiffY ANALYSIS: ODAM sample statims are axit initiators to any plant event and are not included in any assumptions used to analyve plant events previously evaluated in the SAR. The changes in sample locations do not impact the plant's n'o ihty to trutigate the cmsequences of a radiological relcas These changes do not introduce any nert mode of plant operation; cause a system, structure, or component (SSC) to be operated in a manner not previously evaluated; result in a new or difTerent type of SSC to be installed; or make any physical changes to the facility which could afTect a SSC. ODAM sampic locatians are established per Tahnical Specification 3 21.F.I. Changes to the sample locations are also made per this specification.
'lhe sample statim kcatkms are not used to establish any margin of safety as defined in the basis for any Technical Specification.
Or, gating 1.icense Chglge Reauest (OI CR)97-004 USAR1ttutge Reauest (UCR)97-079 Eneineering Evaluation WE197-152 TITIE OI.CR 97-004 - Revision of Technical Specitication Bases UCR 97-079 USAR Change, Control Rod Drive Support Gap EU 97-152 Evaluation of Contml Rai Drive 1Iousing/ Support Gap DESCRIPTION: CNS Pmeedure 6 3 10.3 (now 6 CRD.401), Control Rod Drive (CRD) I fousing Suppod Inspection, was previously revised to a!!ow the gap between the CRD housing and the CRD housing support to be in the range of a 0.75" to s 1.25' The previous range was 2 0.75" to s i.0". This change was made, however, without appropriate consideration of its impact on the applicable wording in the USAR aryd Technical Specification Bases An Engineering Evaluation was performed to provide the design basis documentation for accepting the presious change implemented under Procedure 6 3.10.3 (6.CRD.401) as well as appropriate changes to the USAR and Tec!mical Specification Bases to more appropriately discuss the consideration of the gaps. There is no elTect on safety as result of the revised description of the gaps between the CRD housing and the CRD housing support.
SAlTITY ANAL.YSIS: EE 97 152 demonstrates that the expected maximum reduction, due to thermal and pressure effects, in the gap between the CRD housing supports and the bottom contact surface of the CRD housing is less than the muumum pap allowed by procedure. This assures there will be no contact between these areas when the reactor is at full temperature and pressure, and thus climinates the possibility of contact stress.
'lhere was no change to the lower gap limit in the pmcedure so there is no increase in the pmbability of a break which could cause the control rod to be dropped. The expected maximum gap allowed by procedtar with the wsselin a cold conthtion is 125". EE 97 152 indicates that the gap when the vessel is at full temperature and pressure is 1.0* or less. A 1.0" gap is the size assumed in the analysis to determine the design basis for the CRD housing r,upports. Thus the CRD housing supports design basis is preserved by the setpoints established in the pmcedure. The only time that the vessel would be
' presstuized while in a cold condition, and thus with a gap potentially greater than 1.0*, would be during the vessel pressure test; however, this was evaluated and determined not to be a safety concent The pmbabihty of malfunction of the CRD housing supports is unchanged by allowing a gap of 1.25" in the cold condition The malfenction of the CRD housing supports could result in a Control Rod Drop Accident which is already analyzed in the USAR. The dme consequences of this accident are not atTected by gap size since for the accident to occur, the CRD housing is arbitranly assumed to fait regardless of -
gap si/c The nuirgm of safety is the amount of control ral travel beyund a single withdiawal increment up to the pomt where the control tal is sufliciently withdrawn to cause fuel damage EE 9~-152 denstrates that the revised upper knut for the gap is in conformance with the existing analysis and thus 78
l preserves the design basis requiremert that the control nxi movement will still be less than a single ,
control rod withdrawal movement. Therefore, the :nargin of safety as defined in the Tecimical i Specification basis fw the CRD housing support systems is unchanged by the resised upper limit for the i gap.
Con & tion Report (CR) 94-0854 TIT!Er Pressure Setpoint in the Control Buildmg DESCRIPTION: Design pressure fa the Contml Buildmg at the time of preoperational testing was +0.1" Wg with respect .
to atmosphere. Sometime aftrr startep, the setpoint for the ddTerential pressure indicator controller in the Cmtml Buildmg was chantpl fnxn 4010" Wg to 0.10" Wg. The modification package that changed this setpoint cannot be fourxt. Also, the USAR description of the Control Building iIVAC was changed from positive (+0.10" Wg)io negativt. pressure (-0.10" Wg) in 1990 to correct an error, Ilowever, no discussion of safety concems was included for the change. Therefore, this safety evaluation was performed to address the subject changes.
SAIT!TY ANAL,YSIS: The Control Building pressure has been changed from positive to negative to ensure that no inleakage occurs from the potentially contaminated building to the Control Room. 1 he Control lluilding pressure 4 at 010" Wg is at negarm pressure with respect to the atmosphere and the Control Room envelope. This will minimize the uranonitored release of contaminated air to the atmosphere and the inleakage of unfiltered air fmm the Contml Building to the Contro'. Room. Therefore, the probability of occurrace or consequences of an accident are not increased. Changing the setpoint of the dilTerential pressure indicating controller in the Control Buildmg from +0.10" Wg to -0.10" Wg does not afTect any other equipment in the plant nor is any new equipment added Thernre, this change does c,t increase the probability of occurrence or consequences of a malfunction of equipment important m safety. The Control Buildirg pressure is not mentioned in the Techrucal Specifications. Ilowever. Technical Specification Fcction 3.12.A states that the Control Room Emergency Filter System is designed to maintain the Control Room pressure to the design positive pressure so that all leakage should be out leukage. This requirement would not have been met without changing the Control Buildsg pressure to negative. The change in Control Buikhng pressure does not reduce the margin of safety as defined in the basis for any Technical Specitication Condition Renort (CR) 97-1124 TITLE: keinstafiation of Fuses for Primary Containment Isolation System (PCIS) Isolation Valves RiiR-SJV SSV60,61,95, and %
DESCRIPTION: This CR was written to docum-nt a concern that terminal box TB-641 contained cables from both Divinan I and Disision 11 power control circuits supplying PCIS valves RIIR-SOV-SSV60,61,95, and
- 96. The contml powtr fuses for these valves were removed to address the immediate concern of the CR.
The subsequent safety review determined that it was acceptable to return the fuses to the circuits.
MFETY ANALYSIS. The fail safe position of the redundant valves coincident with a single active failure on a division precludes the need for circuit separation. Existing downstream restrictive sample system piping and valves inside Sceondary Containment will provide any necessary isolation betuten the Residual IIcat Removal (R1IR) system and the outside environs. Beyond design basis credibility, failure of the valves to ekwe would result in minimal flow diversion from the R1IR system and is bounded by analpis of the failure of a Diesel Generator te start resulting in kiss of one division of Emergency Core Cooling Systems.
The sample piping is designed to withstand pressues in excess of the maximum RIIR operating pressure of 450 psig and mamtain dose exposures less than 10CFR50 AppendN A limits. The reinstallation of the valves' control circuit fuses will not prmide a new release pathway beyond the sample piping.
Remstallation of the fuses skies not alTect any system's ability to respond to a transient or accident and, therefore, no increase in the dose associated with a plant event can occur. Two redundant fail safe isolation valves are provided in each division of RlIR at the interface piping between the RIIR and samphng systems. The purpose of the redundant valves is to provide assured isolation in each disision 79-
in the event of a single active failure. No new failure modes are intaxiuced The redundant isolation valve pr.ver cmtrol circuits provide assurance of positive cmtainment isolation assuming a single failure.
Failure to maintain separution of t!c power circuits to the solenoids is acceptable due to the diversity and n fundance of the desipt This activity does not reduce the margin of safety as dermed in the basis of any :
Technical Specification Siemens Power Comoration Procedure I'MF-P71.145 TITLE: In-Core Sipping System DESCRIPTION: This procedure provides for the performance ofin-core s;pping to kwate a failed fuel assembly. The pneess is a radiochemical sampling process perfmned by Siemens Power Corporation "in-Core Sipping System" procedure.
SAFETY ANALYSIS: The proposed activity will be performed in BWR Operating State A, reactor vessel he.d removed and shutdown. In this condition, the only previously evaluated accident in the USAR inat needs to be cmsidered is .he fuel handling accident. The hypothesized accident is dmpping a single fuel assembly fnxn its nmximum height onto the core. No fuel wilme moved during the sipping process so this accident is not credible. Therefore, it is concluded that this activity does not increase the prubability or consequences of any accident previously evaluated in the USAR. The rJpping equipment and process do not interact with any plant equipment in such a way as to cause any malfunction of equipment imputant to safety pcViousy evaluatal in the USAR. All plant equipment will be operated in its normal unie and will be used m!y for its inteHed function. The malfunction of the refueling bridge hoist could prevent lifting the sipping desice oc the upper grid, however, this malfunction would not result in claddmg damage and the nxbological consequences of the sipping device remaining seated on the upper grid are nil Fuel sipping will be perfonned with the reactor vessel head removed and the reactor cavity thxxicd. Primary containment and the reactor prest - vessel are not intact. Fuel sipping cannot adversely affect Scankiry tmtainment. Tierefore. on',, ful cladding damage need be considered. The following types af accidents were considered; arbitrary rupture of any single pipe, mechanical failure of various compments leading to the release of radioactive material from the fuel, and overheating of the fuel cladding None were determined to create the possibility of an accident of a ditTerent type than any evaluated in the USAR. Le following parameters which have the potential for deleterious cfTects on the nuclear steam supply system were evaluated: nuclear system pressure increase, reactor moderator temperature decrease, positive reactivity insertion, reactor vessel coolant inventory decrease, cenctor coolant flow decrease, reactor coolant flow increase, core coolant temperature increase, and excess coolant inventory None were detennined to create the pnsibihty of a malfunction of equipment important to safety ditTerent than any previously evaluated in the USAR. The sections of Technical Specifications applicable to this procedtae are 1.1 L1, Core Thermal Limit (Reactor pressure <= 800 psia arWor Corv Flow <10%), and 1.1.D, Reactor Water Level (Shutdown Condition). The margin of safety in these specifications has not been reduced.
GeneralIllectric Fuel Inspection Renair Phm and Procedures TITI.E: Damaged Fuel ilundle Inspection / Repair DESCRh' TION: The subject plan and procedures provide for failed fuel inspection and repair. They direct disassembly of a fuel bundle, inspection for damaged fuel nxis, replacement of defective fuel nxis, and reassembly of the fuel bundle. His activity is considered a repair activity, the CNS License and our Appendtx B program allow repair to safety related equipment. Ths wair of an irradiated bundle can only be performed at a nuclear plant with a containment system and safety related systems to handle a fuel handling accident.
SAFETY ANALYSIS: Dropping a fuel bund!c from maximum height onto the top of the core is evaluated in the USAR. The probability of this accident is not changed Using the refueling bridge instead of the jib crane, as desenixxlin the USAR, to move the fuel bundle does not adversely afTect safety. The refuelmg accident bounds the possible release of pellets and fragmentation in the spent fuel storage pool (SFSP) The
-80
1 radiologicai consequences of this activity are also bounded by tha fuel handling accident. The teconstructed fuel bundle will be mechanically equivalent to the original bundle and the nuclear-composition of the repaired bundle wit be essentially the same as the original bundle. No other
. equipment is being changed. The prd> ability of a malfunction of either the bundle itself or other equipment important to safety is unchanged. This activity will be performed Li accordance with . >
Technical Specification 3.7.C.d. The radmlogical consequences of a dropped fuel nx! because of the malfunction of the tools used to grapple the rod are bounded by the release from the fuel handling
- accident. Subcriticality in the SFSP is assured if the exposure dependent K= of each bundle is s 1.29.
The possibihty ofperforming multiple mispositioning errors that could result in a k- > 1.29 is extremely knv. The bundle is being repaired by the original manufacturer using their procedures, personnel and -
Quality Assurance program, therefore the possibility of enur is unchanged from the original -
numufacturing pacess. 'ihe number of fuel nxis outside an approved storage location will be limited by pacedure to no nore than ten, thus precluding inadvertent enticulity. The possibility of a fuel loading error is unchangal frorn the evaluation in the USAR. All equipment and systems will be operated using existing appmved procedures and in compliance with the Technical Specifications.
~ General filectric Procedures for Fuel Bundle Rework IITI.H: New Fuel Bundle L"sassembly for Return to Manufacurer; Fuel Burxile and Rods Returned from Reactor Sites (Procedure 5 0.9.0);
Packaging of Fuel Bundles and Imse Rods (Proedt e 5.1.10),
New Fuel Bundle Rework at the Site (Procedi:.e 246-GP-6 t).
Dl!SCRIPTION: Fuel bundle YJJ246 was damaged during transfer to the spent fuel storage pool. The bundle was disassembled, fuel pins removed, decontaminated, packaged for shinment, and shipped to General Electric (GE) for inspection / replacement of damaged fuel pins. These activities were performed in accordance with the subject procedures using qualified GE fuel technicians. Movement and control of fuel was per station pmcedures.
SAFETY ANALYSIS: The rewuk of bundle YJJ246 does not atTect any plant events. The refueling accident and the fuel cask drop bound any events involving this work activity. Therefore the probability of occurrence or consequences of any previously evaluated event are unchanged Dropping an unirradiated fuel bundle in the pool or in the Reactor Building is not a new event and is thus bounded. The Reactor Building crane, Secondary Containment, and the Standby Oas Treatment system are equipment important to safety affected by this activity. There are no changes in the way this equipment is operated and existing pnredures are being used to operate them. Thus, there is no change in the probability of occurrence or consequences of a malfunction of equipment important to sefety. This activity does not alTect any Technical Specification or its basis. The margin of safety as defined in the basis for any Technical Specification is not changed.
Feedwater Temocrature Reduction TITLli: Operation with a I.oss of Feedwater iIcating DESCRIPTION: Feedwater temperature is used as an input to the CNS transient analysis. As a result of a failure in an extraction non-return valve (CD-AOV-LCV61 A), feedwater temperature was reduced by approximately 18' from the CNS design value. This safety evaluation was performed to determine whether an unreviewed safety queation exists for continued operation at this reduced feedwater temperature conhtim lhe effect of the rahm! feedwater temperature on the CNS transient analysis and the Reactor Pressure Vessel Fatigue Analysis was investigated ar part of this evaluation.
SAFETY ANAL Y' SIS: Retha:ed fealw ater heating does not involve a physical change of any plant system or procedure and to accident imtiators are alTected. Current analysis supports a Loss of Feedweter ilecting Trans6.t or up to 100*F. Reduced feedwater temperatures afTect the tran icnt analysis state point ass imptions Ilowewr, analyses performed by General Electric haw shown that the Operating Limit MinimunMritical Power Ratio bounds operation with up to a 60*F loss of feedwater heating and ensures that the tafety
~ . - . -- ..-. .- . -
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Limit Minimum Critical Power Ratio will not be exceedal for any Abnormal' Operational Transient Therefore, the proposed change does not inen:ase the probability of transition boiling during any plant
_ event and fuel integrity is not afTectal Raluced fwdwater temperature also impacts the Reactor Pressure Vessel (RPV) fatigue analysis. Ilowever, NPPD Calculation NEDC 96-005 addresses the effect of reductims in feedwater temperatures on the RPV and shows that reductions of feedwater temperature of
- less than or equid to 25'F have an insignificant effect on the RPV fatigue analysis. Thexfore, the loss of up to 25'F of feedwater temperature does not increase the probcbility of occurrence or consequences of a plent event or malfunction of equipment important to safety. Since the limiting transient has not changed as a result of operation at a reduced feedwater temperaturcJhc loss of up to 25'F of feedwater temperature (krs not reduce the margin of safety as defined in the basis for any Technical Specification.
Core Shufilq TITLE; Refuelmg Outage 17 Core Shuille Shutdown Margin Calculation DESCRIPTION: This safety evaluatie9 addresses core shutdown margin during a core shufile and the calculations performed to assure that shutdown margin is maintained for all intermediate core configurations. These calculations were perfonned using the computer sof1 ware program COSMOS and benclunarking calculations performed using an NRC approved 3D Nmlal Simulator. Maintenance of shutdown margin durmg the REl7 core shuffle is necessary to provide assurance that an intennediate core configuration (kes not increase core reactivity sufliciently for the core to achieve criticality with one control rod
! withdrawn.
SAFETY ANALYSIS: The proposed change does not physically change any plant system or procedure and no accident preaam are alTected This change involves only the method by which shutdown margin is calculated .
during refueling operations. A conservative shutdown margin design limit has been established to account for model uncertainties and Liases. Calculations show that the COSMOS calculations of shutdown margin are consen ative for all core configurations. The shutdown margin design limit was not exceeded for any intermediate core configuration thereby precluding an inadvertent criticality during refueling operations Fuel integrity is preserved and the radiological consequences of presiously evaluated accidents are not inreased. The method of calculating shutdown margin has no impact on the operation of any sptem nor ia die control rul system physically impacted. The margin of safety as defined in the basis for any Technical Specification is not reduced.
General Electric Procedures TITLE- Cooper Marathon ControlIlle.c Inspection EOC17 Control Rai Blade Inspection - Procedure 246-GP-48, Rev. 2 -
DESCRIPTION: The above pnmhires pmvide instructions for the inspection of the Marathon control rod blades during 1m17. 'fhis inspection is requaed t)y commitments made by General Electric to the NRC to examine the lead um Marathon Contml Rm! Blades. This activity is bemg performed by General Elcetric using their procedures, personnel, and equipment.
SAFETY ANALYSIS: De examination of Marathon contml blades during the refueling outage does not change any of the plant events that could occur at that time and does not atTect any mitigation systems. Only a visual examination is being perfanned. All other activities aiTectmg equipment important to safety are being performed using existing primlures,includmg moving the blades. No changes are being made to a ry equipment. The is no change in the margin of safety as dermed in the basis for any Technical Specification.
l
. . - =--- - -. - .- - - . - - -
biblem Identifiention kroort (PIR) 2-08641 Ti1 LIE O Sump Operation Witimut Ili arul 1liili Alarms, and Iligh Fillup Rate Alarm (Sussted Stock Th>at Suitch) !
i DliSCRIPTION: This analysis evaluated tie safety aspects of the O sump operathm in the dryuell uidmut Ili and Ili Ili alarms, automatic lh and t h lli pumping (perations, ard the proper operation of the high fillup rate alann ikith O suno purre ccritrol switches are in Os pull to kwk position u hich disables the automatie i pumping and slann functions !!arly leak detection by way of the O sump high fillup rate alarm in the drywell is also daabini due ta its logic tie to the su4pected stuck float switch. Operations pumps de .
O sump every two hours ard recorddtrends Or leakage rate. Procedure 6 LOO 601 was revised to j inungirats this reairement.1his requirement will remain in place until the float switch is repaint sal!!TY ANALYSIS: The impact to safety due to the loss of the automatic functions is minimal as other leakage monitedng methods in the drywell semain unchanged includmg drywell tems rature, pressure, humidity, and radiation detection. Any leakage in exuss of the O sump volume between pump downs would be directal to ard detectal in dw F stanp. 1his leakage would 0 en be detected and indicated in the Control Ram Failure of O manp is tot credited in any analy sis as an event initiator and O sump is not credited as an accident mitigation system O sump does not irnpact the capability of the Primary Containment 1 system. 1hc hm of this Regulatory Guide 1.97 Category 3 annunciation will not cause the malfunction or km ofolla ecptinnent imputant to safety.1he lack of 0 sump annunciation ard automatic pumping functions does not increase the consequences of a malfuncthm of equipment imp >rtant to safety, specifically F or O manp pump failure would result in a Technical Specification 24 Imur shutdown F and O sunps carumt irutiate a leak in the drywell. Other metinis of drywell leakage detection are available.
No cSinges in existing plant hardware lunt lo:n made. The sump equipment is operated more ollen thus iracasing wear rate, however, it is nonessential.1he Technical Specificathm basis states that leakage less than the specified Limiting Qindition for Operation limits can be detected reasonably in a matter cf a few hours lhe O sump fill rate timer and slarm which are disabled by this ctmdition were designed such that Control Rann Operat<rs roccive un early warning of changes in sump fillup rate. This does not te hice the margin of safety because the increased manual pumping frequency will identify smaller leaks that tle filho rate alarm was designal to identify, l.arger breaks are identified quickly by other systems such as radiation monitors and drywell pressure. Themfore, the two hour manual pumpiny performed by Operathms would ak rt operators within the basis described limits.
Mamicmtmr Wort Reauest (MWR197-0163 ITl1R Pesetting of Local Power Range Monitors (LPRMs) to Original Celibrated State DihRIPTION: This MWR provided instructions to reset the LPRM cabbration performed per Proceduies 10.5 and 61J'RM.302 on January 23,1997, to the trigtnal values prior to the cabbration Resettmg of11 e 1.PRMs to their pi<r setting was necessary to seturn the Li'RMs to a known cahbrated state. This was required because Traversing incore Probe (TIP) C failed and a final ODI coald not be perfonned. In this xrdun the cahbrated LPRM settings were not available to ntn a Pl program. As part of this activity, the Average Power Range Monitors (APRMs) were also readjusted to the proper settings based m the reset LPRMs.
SAFETY ANALYSIS: Resetting the LPRMs to their original calibrated state rotund them to a state that can be used by P1 to pnperly calculate thermal limits. Ily ensuring that tiermal hme such as the Minimum Critical Power Ratio, Maximum Average Planar Linear llent Generatbn Rate, et;. are properly calculated by P1 and ensunng that the APRM gains are pmperly rese s this activity did not thange the probability of occurrence or consequences of a plant ewnt. The surveillance interval for LPRM calibration allows the required cahbratien to ha p:rftsmed Ir/tx ard still meet Oc TechnicalFS eification surveillance requirement. The r %bability of occurrence or consequences of a malfunction of equipment important to safety (TIPS, L19Ms, APRMs) is vnatTected This activity ensures that the margin of safety for thermal hmits as deft, al in tlw basis for tir Technical Specifications is maintained
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1 1
Problem Idenuncation Reriort (Pilt) M*AS [
t TillE Minor Mamtenance to Potential 1 ransforurt I
Dl! SCRIP 110N: 1his P!R initiated mumt mamtenance to check phase angles and voltage amplitudes between phases of the l' bus seming potential transfonner frorn the limergency Transformer.
SAlliTY ,
ANALYSIS: 1hc I!merswy 'lrandsma,lillus, and potential transformer are not accident initiators The operation i of the linupicy Trandnro u111 rud le altered armiit will be able to perform its intended functiort 1hc :
test equipment will not change or prevent any action that is used to mitigate the consequences of an :
accident. The text equipmect will set as a loaJ but will tot draw sufliont current to change the characteristics of the circuit. The test equipment does not supply a source of energy to the circuit end tluefac will rud impair or cornpromise operable equipment.1he test equipment being in parallel with the urxicrvoltage relay prevents the failure of the test equiparnt from afTecting the performance of the safety related function.1he test equipment will be electrically isolated slumld a failure of the test equipment occur, lhe design of the system has tot been altered in any way to change the ope:ation of the syst(m.1he margin of safety is not changed Shichhno Regests 9710 and 97 11 11T1.11: Temporary Shieldmg on Reactor Water Cleanup (RWCll) Valves and Piping DliSCRIP1 ION: Temptary lead shieldmg waz,imtallalon vahrs R o CU V 10 and RWCll-MOV Mol 5, and the piping between these two valves lhe shieldmg was temporary and in place only during shutdown to reduce l
dose rates for Rlil7 work actmtics in the area sal'lTY ANALYSIS: With the shieldmg in place, the piping and anociated supports continue to meet design code nnNor operabihty requirements and will not under any design basis conditions lose their ability to perform their intendal safety function in case the shiekhng were to become dashiJged, it would only fall to the grating directly behiw, causmg no damage to nearby equipment 1he shielding will only remain in place during shutamn No iracased radametiw releases are possible as a result ofinstalling the shielding. No other postulated plant events are intraluced by temporarily addmg the shieldmg. The margin of safety as defined in the basis for any Technical Specification is not alTected.
Setpunt Chance Rgauest (SCR)(XuI8 TITI.li: Setpiint Change Request for Reactor Water 1.evel Alann Module Turbine Trips Ri>C 1.A 121 A D.C DliSCRIPTION- The setpiint of RFC-l.A 121 A.II,C was changed from 58 0" 1120 indicated reactor water level to 54.9" l120 it beatal reactor u ater lewl as suppated by Nuclear lingineering Department Calculation 96-027, 11us chnnge takes into account emy terms associated with the instrument hops that provide trip signals to the Main Turbine and Reactor Feed Pump Turbines.
SAFl!TY ANALYSIS: This setpoint change does not increase the probabihty of occurrence of a plant esent since the setpoint is calculaial imng General likrtne Setpoint Methalology u hich cynluates Imth the margm to protect for overfill and spurious tnp avoidance from normcl water level variances The consequences of a plar8 ewnt or nudfunction of equipment are not incream! since the change lowers the setpoint to ensure that proper owrfill pnaceton is maintained This overfill protection is necessary since overtill of the reactor could result in lithng of the steam knes with water. The system will maintam the current configuration with the exception of a lower setpomt that accounts for the emir tenm associated with the instrumeat hop This change which hmers the setpoint increases the margin of safety from the setpoint to the analytical limit to prevent overfill-
. 84 4
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$ctreint Chance Rcouest iSCR) %19 <
Till.lk Setpoint Cimnge Request for Refueling Platform compresms Pressure switch TSE PS-C <
DliSCRIP110N: MM 95 093 replawd the Refuehng Bridge Air Compresww. Ilased on operating experience following implanentation of the nni fication, prensure wttings uere revised to load at 90 psig and unload at 125 psig This reduction of pressures will provide adequate compressor protecthm and provide ample presuure kr all equipment to function as designed SAI!!TY ANAL,YSIS. Resettmg the pressure switch to a lower operating sange will reduce the stress on the omnpressor by running in a continuous male rather than an intennittent nude. Changing de range at which the ,
unnpresst operates will rad increase the probability of a plent event as previously evaluated. l.owering of the operating range will ruit clumpe die o;x rattrig fwietiori of any tools associated widi die Refuel lindge If the cunpesw r were to fail, the grapple would not be able to perform any opening or closing functions Revising of the setpoints at which the cornpressor loads and unloads will not increase the
- probability of occunence or consequences of a n'alfunction of eqmpment, but in fact will decrease the probabihty of occurrence by making 0 e air compressor more reliable by reducing the stress on the aanpreswr. Revisam of das wtpoints wdl not create any new faihnc males The design, configuration, uml operating chanieteristics of the compressor remain unchanged The air compressor is not defined in the basis ihr any Technical Specification, therefine, no Technical Specification margin of safety in reduced.
Sctruint Chance Reouest (SCR) 96-28 TITI.lk Setpoint Change Request for Reactor i bgh Pressure Scram Instruments N!!!.PS 55A,B,C,D Dl! SCRIPT 10N. Nuclear lingmeeting Department Calculatum 92-0501 Rev. I detennined a new operating setpoint cf 1038 5 psig with a head aunttuvi of 13 psig, for a new calculated setpoint of 1051.5 psig. The existing .
operatmg wtpoint is 1035 psig with a 14 psig head urnxtion, for a calculated setpoint of 1049 psig. The wt; mint of 1049 will remam unchanged by this SCR since it is conservative. This SCR only documents the new operating wtpunt and head conection (w hich is used in determining the instrument limit stated in the procedures)
SAFl!TY ANAL.YSIS- 1his setpoint change does rut cluvige the existmg setpunt of the high pressure seram but only documents the new operating setsiint and head correction value, therefine, this activity does not increase the probability of occunence or consequences of a plant event or malfunction of equipment important to safety. this setpoint clumpe is an udnurustratne change only; no recahbratum is required by this activity.
1he custmg setpunt specified in the procedures remains unchanged since it is more conservative than the new calculated setpunt; thereftre, there is no reduction in the margin of safety as defined in the basis for any Technical Specification Setremt Change t{gquest %24. Rev 1 TITI.12 Setpoint Change Request for Drywell Air Morutor Particulate, RMV RM-4A DliSCRIPTION: 1he setivint liv the Drywell Air Mautor Partidde Channel was changed from 780 epm to 7.48E4 cpm.
The previoudy custing setpoint u as causing spunous alarms in tl e Control Room as a result of the fuel pin failure. The alarm limit and setpoint were detennined in Nuclear Engineering Department Calculation 97 021 and !!ngineering Memo DliD97044 sal'liTY ANALYSIS: The monitor alann function provides indication of increasing particulate activity u a result of an occuitence. Adjustments to the setpoint fhr the alarm function do not increase the probabihty of an occunence. The monitor is an alternate system ihr drywell leak detecton Automatic functions to mitigate the comequences of a plant event arc imt affected by this setpoint. The alarm setpoint is based on nonnal reactor water activity with a pre-determined leakage rate. liquipment important to safety 85
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1 within the drywell is qualifial to nunc severe cmxhtims The function of the momtor is not changed No specific setpoint is identified in the Technical Specifications for this instrument, l Setrxiint Chance Reauest 97-05 I 1111Ji. Setpoint Change Request for Drywell Radiation Monike RM%RM 4A,ll,C DliSCR!pilON: 11e kw (kw alann setpe nt fit RM%RM-4 is being changed fnun 1.5 cfm to 1.0 cfm to conform to the mxke's rmannerstaturt 1he kwering of the setpoint to 1.0 cim will climinate nuisance alarms since the normal operating flow is approximately 1.5 cfm.
SAFliTY ANAL,YSIS. lic km ikw alarm function provides indicatxn ofnxsukt optability only and provides no safety related functim lhe nusutor is um! as an alternate metixxl for dowell leakage detection. The changing of the low flow setpoint will not increase the probability of occurrenct or consequences of a plant event or malfunction of equipment imputant to safety. 'I he low flow setpoint will still provide indication to llw Cmtrol Rasn of monitor operability and will not reduce the margin of safety as defined in the basis for ;
uny Technical Specification.
&toomt Chance Request 97,06 11TI.li: Setpoint Change Request for Drywell Air Monitor. Particulate, RMWRM 4A Dl!SCRipTION: The alarm setpoint fit the particulate channel of RMWRM 4A is being changed from 7.841!4 cpm to 5 251!4 epm lhe setpoint is being reducaldue to kwenng the sample flow rate from 2 0 cfm to 1.5 cftn.
SAllITY
- ANAL.YSIS: 1he monitor ahirm function provides indication of incrensmg particulate activity. Adjustments te the setpint fit the alarm function do rut increase the probability of occurrence of a plant event. The monitor is an alternate system for dr>well leak detection Automatic functions to mitigate the consequences of a plant emit are not afTectalby this setpoint. The alarm setpoint is based on normal reactor water activity with a praletennirxslleakage rate. !!quipment important to safety within tic drywcll is qualified to more neme canhtions (Insa of Conlant Accident) 1his activity lowers the monitor alann setpoint, however, the function of the mor,itor is not changed No specific setpoint is identified in the TecHical Specifications for this instrument.
Ijmitoroue Oocrator Maintenance Procedures TITili- Acceptabihty of Past Changes to 1.imitorque Operakir Maintenance Procedures DliSCRil' TION: This evahiation w as perfirmal to document the acceptabihty of the following changes w hich were made to various motor operated valve maintenance pnx.edures m the past without a safety evaluation:
installation of a new style spring cap, replacing 5/16" and 3/H" grade SAi!-5 with 5/16* and 3/8' grade SAll M capscrews, replacing 3/8* hmising covers with %" housing covers, the removal of the kical irabeator drive chain, the replacement of fink lug connectors with ring lug connectors, and replacement of melamine rotors with fibrite rotors. 1he atTecksi procedures are: 7.2.50 3 Rev. 6,7.2.50.4 Rev. 8, 7.2=50 M Rev. 2,7.2 50 9 Rev. 2,7 2.5010 Rev. 21,7.2.50.1 I Rev. 2,7.2.5012 Rev. 2,7,2.50.13 Rev.
> 2,7.2.5014 Rev. 2,7.2.50.15 Rev,2.1,7.2.50.I6 Rev. 2,7.2.5017 Rev. 2, and 7.3 50.5 Rev. 2. The smi to evaluate these changes w as identified when new motor operated valve maintenance procedures were being evaluated SAFl?TY ANAL.YSIS: The changes to the operator are consistent with the manufacturer and or CNS dccumentation The changes do not alter the function of the operator in any way that will affect the mitiator of evaluated accidents Sysk m function and response remain the same and no new faihsre modes are intraluced that coukirnhmx the mitigatim capabihtiss of the systems mvolved The changes are an improvement to the op'rator, M such the probabihty ofoteurrence of a rnalfunction is decreased Since neither the operation nor capabihty of the operater is changed, it cannot increase the consequences of a malfunction of equipment. No new imtiakus or failures have been intnslucal that could lead to a plant event of a
different type.1he changes do m4 alter the performance or requirements of the valves as required m i Technical Spaifications j Safety and Relief Valve Setooint Chance Reauest 96401 ;
T1113!: Setpomt Change for FP HV.16RV,lilectric Fire Pump "l?" Circulatmg Relief Valve Dl!SCRIPTION 1his setpiint change w as irutiated to alknv the circulating relief valve on the filectric Fire Pump to open at or beknv claun (deatheal) perare, which is currently measured at approximately 173 psig. This will protect the pump fnun overheatmg thus helping to ensure its contmued availabihty. The setpiint was clutnged fann 175 psig * $ 25 psig to 167 psig *5 00 psig.
SAFl:1Y ,
ANA!JSIS. The probability ofoxunence of a fire is not afrected by this change. None of the accidents evaluated in the USAR are impacted The ability to mitigate the consequences of a fire requires that water demand for inanual and automatic suppression actiuties be met.1his setpoint change ensures that the lilectne .
Fire Pump can operate at claun presstre and stdlle cooled by the water allowed by the relief valve. This will prevent the pump fnun overheating. There are no adverse consequences fnim the resultant 1%
diversion of flow through the relief valve. This activity does not increase the probabihty of occurrence of a malfunctim of equipment by ensuring that the water in the pump and portions of the fire protection piping do not overheat and flash to steam.1he single failure of one fire pump has been previously evaluated and is mitigated by a redundant dicsci fire pump of equal capacity.1his activity is designed to help mitigate the consequences of a " deadheaded" electric fire pump. 1he new setpoint of Fp-RV.16RV sets the opening pressure of the valve within the operating presswe range of the Fire a
Jockey Pump "F." 1his does tv4 create the possibility of a different type of malftmetion because cach fire panp has a didunge check valve w hich prevents backflow to their respwtive pump.1his actinty will ru 4 re&cc the inargin of safety as the prpisc is to ensure the availabihty of the lilectne Fire Pump under all types ofoperatumal conditions. Technica' Specification capacity requirements for the fire protection w ater suppression system are maintained via redundant piping and diesel fire pump.
Sahtt and l{def Yahe setooint Changd{cauest 97-001
'llTIJi: Setpoint Change for RCIC RV 10RV, Reactor Core Isolation Cooling (RCIC) Pump Suction Relief Valve Dl?SCRIPTION. This setpoint clumpe unensal the setpoint for RCIC RV 10RV to prevent the relief valve from weeping w hile the RCIC system is rat in operation and pressure maintenance in mamtnining the water level within the piping system The setimint was changed from 9012.7 psig to 100 *3 0 psig.
SAFl!TY ANAINSIS: RCIC is not an irutiator for any plant event and this setpoint change wdl not alTect the abihty of RCIC to operate. Cluinging the setpoint for the rehef vahr does rat aficct the abihty of RCIC to perform its desigt functions During ogn ation of RCIC, the rehef valve will continue to provide its pressure boundary functum lhe wt pressure is within cale requirements in order to protect the RCIC pump suction piping fnen overpressurvation increaxmg the set pressure will be nure conservative in providmg the Primary Contamnnit lxunwkuy when RCIC is lined up to take suction from the suppression pool The margin of safety is rut reduccd as the relief valve will continue to meet the design code requirements and abihty of RCIC to perform its design functions is not changed Safety and Rdef Valve Setooint Chance Reauest 97-004 TITIJi. Setpsnt Change for Fp-RV 16RV, Fire Protection Pump li !)ischat ge Relief Valve OliSCRIPTION; The wtpsnt of the Fire Protection Pump 11 Discharge Relief Valve was changed from 167 *5 0 psig to 15014.5 psig This change w as made to allow the relief valve to perform its function of cooling the pump uben tic punp is in a dealhead cormhtxo which is appnmmately 173 psig. The setpoint had been preuously revml to 167 *5 0 psig gu Setpoint Change Request 96 001 to ensure activation of the relief valve 1lowever, flow was not sutlicient in order to cool the pump
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SAlliTY
- ANAL.YSIS: 1he probability of occurtence of a fire is not afTectal by this change. None of tic accidents p>stulated !
in tle USAR are impactal by tin change 1he abihty to mitigate the consequences of a fire requires that wata (kmand f<r manual ard auttunatic suppession actisities be met,1his setpoint change will prevent the pump frorn overheating by allowing cooling flow when the pump is in a no discharge condition. ,
line are in adverse untepiences fann tic diversion of flow (approximately 35 gpm) through ti e relief ,
valve.1he setpoint change ensures that the water in the pump and pvtions of the fire protection piping do not overhent. The flow fiom the relief valve is routed to a fker drain and will rmt affect any other t equipnent Faihire of the pump has lwn previomly evaluated and is mitigated by a redundant die I fire pump of equal capacity. The setsiint of the relief vMvc is below the operstmg pressme of Fire Jockey Pump FJ' I kiwetti, the lilectric Fire Pump has a discharge check valve which prevents backflow into the panp and relief valve. Therefore, the fire jockey pump will not is affected by this setpoint change.
Technical Specification capacity rcquirements for fire protection are maintained and redundant systems are not affected Mokir Docrated Valve Setooint Chanec Reauest 4
1TilJb Setpoint Change for CS MOV MO511 Dl!SCRIPflON: Replacement of CS-MOV M0511 resulted in a change to the hmit switch settings.1he limit switch ,
setting changes are necessary because the replacement valve dimensions are alightly different than the original valve. Thir safety analysis evaluates the change in limit switch settings SAFl!1Y ANAL,YSis: 't he hmit switches will be set in accordance with estabbshed metixslology and the resulting stroke time venfied to meet valve operability limits. Post maintenance testing will be performed to ensure that the required stroke time specifications are met.1he revised limit switch settings will not afTect the ability of the new CS-MOV M0511 to stroke within specified requirements.11erefore, the probability of occurrence or c(msequences of a plant event or malfunction of equipment important to safety are not ,
t increaul and the margin of safety as defined in the basis for any Technical Specification is not reduced Surveilhmce Procedure 61IV 104 TITili: Control Room timergency Fan Charcoal and IlliPA Filter 1.cak Test, Fan Capacity Test, ard Charcoal Sampling DliSCRIPTION. Sisveillance Procedure 6 IIV.104 verifies that the Control Romn limergency Fan, charcoal and !!!!PA filter are capable of performing their intended functions. This evaluation addresses the opening of the system sample ports, and intrusion of test equipment. Based on this evaluation, there is no procedure change required SAFliTY ANAL.YSIS: Opening of sample ports and installation of test equipment has no impact on event initiators used in the safety analysis and identified in the USAR. Opening sample parts does not impact system flow and bypass flow remains less than 1% The eqmpment is intended for use post event to maintain Control Room emironment. An evaluntion for impact on system petformance based on test equipment failure and debris intnsluced into the system was performed it was conchuled that the system will remain operable lhete is no interface with rotating equipment, controls, power supplies, or other components or equipment that could initiate a malfunction of equipment important to safety as defined in the USAR.
11e system is capable of performing its intended function with sample ports open The system's ability to ma t its intended function as defined in the Technical Specification liases is not impacted. This activity does not raluce the margin of safety as defined in the basis for any Technical Specification.
88
l Sun etilance Prtredare 6 I?l! 6n9 TIT 1E: 125V/250V Statuo llattery intercell coninxtion Testing t
Dl! SCRIP IlON; An evaluatxo was performed to allow using the lhgital 1.ow Resistance Olun (DIRO) meter to test the intaccll conmoction of the station batteries The pucedure can be perftsmed with the battery onlme or ollhne.1he interunnxtum is designed as a short circuit and the DLRO does not alter that short circuit.
SAllff Y ,
ANALYSIS: lie battenes we not awciated with any iruuators of piniously evaluated plant events, therefwe, testing i the batteries using the DI.RO will not change any initiating eventa.1he operation and capacity of the batteries is not aticcted No other systems are involved 1hc DIRO will not degrade the capability of the batteries to perform their designed function Checking the resistance of the connections does not increase the probabihty of a malfunction h will however indicate if maintenance is needed and prevent :
pomNe malfunctims in the future No new failure nxthanisms are introduced by the DIRO. Measuring the resistance across a short circuit does not physically or electrically change the circuit or how it will operate.1hc requirements of the battery are tot changed and the margm of safety is not reduced.
Surveillance Procedures 6.1ki!C 302 and t' 71U!C.302 TITIR IU!C Pump Time Delay Relay Testing and Setting (Division 1. Division 2)
DliSCRIPTION. Pntedures 6. lRliC302 anl 6 2 REC 302 prmide instructions for testing the Reactor Equipment Cooling i (RI!C) time delay relays These relays control the sequence time ihr placing the REC pumps onto the D esel Oerrrator.1his renew punides justification tojumper one relay at a time for testing and setting the relay using installed jumper switches.1hc renew also es aluates impact to interfacing systems.
SAFiffY ANALYSIS: The Rl!C system consists of two independent subsystems, each contaimng two pumps and one heat exchanger. One subsystem is capable of supplying the coohng requirements of the essential senices loads ibilowing design accident conditions with only one pump in ti e subsystem. To jumper the time delay relay for any one pump makes that one pump, on that train, unavailabic for automatic resomse w hile the jumper is m place.1his comhtion is within the assumed conditions analyred by the safety analysis uhich assumes only one train of equipment available to respond. The installed jumper switches hnut the impact to mterfacing systems If a fault were to occur, only the comgonent being tested would fail to sequence. 'the REC system and diesel load seqtwncing circuits are not identi6ed as an initiator or contributor for any event identified in the USAR or the safety analysis There is no increase in the probabihty ofoccunumv consequences of a phmt ovnt evaluated in the USAR. Train redumiancy and sparatim ensures that minimum equipment requirements are maintained during the performance of the surveillance. With the redundant train of equipment available, the probabihty of occurrence of a malfunction of equipment important to safety is unchanged The safety analysis and basis fbr Technical Specifications assume that one train of equipment will be available to respond to an event. During this sunedlance, me train of REC, the other train of equipment and its anociated diesel generator will remain available to respond to an event. There is no unpact on the snargin of safety. ;
Sutveillangsfraedure 61IV 105 TIILE. Control Room Envelope Pressuruation Test DESCRIPTION: This pnelure was evahuited as a result of a Significant Condition Adverse to Quahty (SCAQ 96-0763) ameeming installing test equipment with the system operable. The pacedure connects a manometer to test puts aal uses e pitot tube traverse m thmannmuncter to determme flow. This evaluation applies i only to the installation of the test equipment and the opening of sensing lines.
SAFETY ANALYSIS: The connection to the downstream side of ilV-DpT.845 may cause a minor change in llow from the Control lluilding Recirculation Fans These fans do not have a safety function and the connection will not atTect safety. Opening of the sensing ime that connects the cable spreading nom pressure to the Cmtrol Room has no etTect on the Control Room envehipe because the line is small and there is a very ,
89
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large vent connecting these ravns; tlerefore,11 ere is no affect on safety. Opening the test pat to determine flow will only shghtly decrease t}c anxmnt of outside air supplied to the Control Room.1he nduced free air was pc iously evaluntal as rxt creating an unrnicuni safety question 11e connection to the atmospheric reference tap ofIl%DPT 840 may cause a mitxt change m the flow control for the turbine tankimg exhaust fara 11cse fans do not have a safety function and the connection will not affat safety, There will be a temporary breach of the Control Rmun enychipe by opening a 1/4* line that amnects the cable expansion nunn to the cable spreading rann and to the atmosphere. The breach is ternporary and the cable spreading runn is at a higher pressure so tic flow will be out of the Control Ravn tanckipe, this will prevent tle spread of contamination into the Control Rann envekipe. The size of the seming line prevents the has of enough air to impact the pressurizatkin of the Control Room lheref<re, tins is rxd considered to have any cfTect on safety. Opening the sample ports and installation of test equipment has no irnpact on event irutintors identified in the SAR. The testing perfirmed has only a very minor impact on the llVAC systems. These systems will still perform as designed No other systems are allected.1his activity does not increase tic probabihty of occurrence of consequences of a malfunction of equipment important to safety.1he system's ability to meet its design function as described in Techriical Specifications is txd altered and the margin of safety is not changed
' hErdure OCP.10 2.NPPD.7301 T111J!: 11nderwater Coating Repair DESCRIP110N; This is a temprary pacedure to provide instructions and requirements fit performing and d cumenting underwater coating repair in accordance with Contract 96-75 which will repair the conting in the torus immersion aren The coating to be med is UT 1511nderwater Cured lipoxy that is qualified for use in the torus immersum areas and the main condensate storage tank i A.
SAFl!TY ANALYSIS: 1his activity aen rxit invohr any equignent or systems credited as event initiators or with mitigating the ameluences of a plant event rur dies it adversely affect any systems crnhted with terminating transients ulame failure could result iri a plurit event.1hc coating has been shown to satisfactorily withstand the ternperatures arul pressures of the steam environment postulatal during a design basis loss of Coolant Awident for the wetwell Since this coating is quahlied to a design basis accident, there is no efTect on safety lhin actnity does rud affect tle abihty to shutdown the reactor and maintain it in a cold shutdown, ist does it aflixt the ability ki amtsin radioactive snateriais either during normal operation or post event.
1his activity ass not reduce the margin of safety as defined in the basis for any Technical Specification since no equipment functions are impacted.
Pumlare OCI .10 7 NPPD 730:
TITLli: Application and Impection of Unquahlied Coatings DliSCRIPTION: 'this is a temprary pncahre to provide instructions and requirements fbr performing and documenting unquahfied underwater coating repair in accordance with Contract 96 75 w hich will repair the coatmp -
in the tona immersion area The amount of coating being applied is suflicient to nutigate the galvanic cornwion on tic flanges being installed for the new Emergency Core Cooling System (ECCS) strainers; how ever, the coating thickness is not qualified for a design basis accident.
SAFETY ANALYSIS: The coating has been shown to satisfactonly withstand the temperatures and pressures of the steam eminuunent postulatal aning a design basis loss of Coolant Accident for the -twell I km ever, since the coating will not be installed accordmg to the mill thickness as requires per the emironmental qualification, the coating will have to be classified as unqualifwd The unqualified coating is being applial to tic bolt holes of the new welded in flanges for the ECCS suction strainers and the one side of the w ashent Tic annunt is insignificant and also the hication of the coating will preclude it from being dishxiged and becoming sucked up by the suction strainers The mill thickness will be adequate to .
prornt potential galvanic reactaity. This will maintain the integrity of the flange, This activity does not aficct. any equipment or systems credited as event initiators nor dies it aJversely affect any systems credited with terminating transients whose failure could result in a plant event. It does not affect the 90-
J plant's abihty to(xotam radmaets matenals either during memal operation or post event. This act vity I does not reduce the margin of safety as defined in il e basis for any Technical Spmifications since no equipment functions are impacted !
Proedere M f-CNS-100V1 ;
Till.ll: Procedure for Magnetic Particle lhaminaton Using the Yoke Technique ;
DliSCRIPTION: 1his is a Oext allilectric non-destructm exarmnatim (NDl!) pncedure fir the perimmance of magnetic particle examinations on ferromagnetic materials using the yoke technique. Magnetic particle j examinations require temprary contact between the component to be examined and a magnetic yoke.
1his pnsuture has been updated to the requirements of the 1989 Edition of ASME Sections V and XI.
SAll!TY ANAL,YSIS: Inservice inspection is tot a precunor to any plant event descnixxlin the USAR and plays no role in accident mitigation. The examinations and tests perfonned for insenice inspection do not affect the operability of plant equipment The exarninations and tests are required by regulation, perfunned in auxrdance with ASMl! Code, and are controlled by the Maintenance WarL Request process. Insenice inspectim is required by the Technical Specifications De proposed procc/ure is consistent with ASMi!
Ctsic ter 4irements and is, therefore, consistent with the basis for the Technical Specifications.
Preventive Maimenance (PM) Items 027M and 069H7 TillE Seniec Air (SA) Preventive Maintenance (PM) Items Dl! SCRIP 110N: lie subjwt PMs tequire bloding open SA AOV.PCV609, Senice Air Supply Ilemler Isolation Valve, ,
to prevent a senice air system isolation when they are performed An unresiewed safety question evaluation was performed fm these PMs.
SAll!TY ANAL.YSIS: The probabihty of a plant event occuning by not bhicking the isolation valve open and opening it with tic SA hetuler depressuimiis greater tium tic probability of g SA header line break dunng perfonnance of the PMs. He SA and lastnm ent Air (IA) systems are nonessential systems that are not relied upon to mitigate the consequences of plant events. Performing the PMs is required to ensure that the equipment is maintaunt such that it can perlinn the expected isolation should the need a,ise. liquipment imputant to safety fails safe or is supported by accumulators w hich are tested to ensure they contain an adequate supply of air. This accumulator air supply ensures automatic actions wdl occur or the Operator has sufficient time to perform the action before the accumulator pressure is insullicient to provide the required function Operator actions for manipulating compments supplied by these accumulators are specified by timergency Pncedure 5 2 8,"less of Instrument Air." Therefore, the probabihty of a malfunctim of equipment is not increased The only type of plant event or equipment malfunction that pafonning tic PMs can cause is an acceler ated loss ofIA pressure caused by bhicking PCV 609 open The SAR does not rely on the IA system for safe shutdown and has evaluated the effects oflosing the system Tl e salla systems are not relied upon by any Technical Specification, therefwe, this actisity does not reduce the margm of safety as defined in the basis for any Technical Specificatiort conn-etion of Test l'auioment TIT 1.ls: Safety livaluaton to Allow Use of Non-Intrusive Instrumentation during Preventive Maintenance and Troubleshooting Activities DliSCRIPTION: The pro;xwed activities to be procedurally supported by this safety evaluation and incorporated into established maintenance procedures are the use of certain measurmg devices such as 11 uke Digital Muhimetas (DMMs) for witage measurements and clamp-on ampere meters for current measurements on equipment which may be energiicd and considered operable. The application will be limited to mitage nsasurements and clamp.cn ampere meters Tnis limitation ensures that the measunng desice remains passive to climinate interactions with the circuit. The concems discussed in NRC Information Nouce 95-13 related to meter burden and electromagnetic interference related noise is addressul by the 91
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rajunennut that t!c nwier ex4 le instalkd ard left unattended for any length of time and the inhetent high input impedance of the meters SAFl?TY ANALYSIS: The use of these meters will ruit catise the operatiori of ariy circuit comsments or oilerwise actuate ,
system dnices that would lead to dw initiati m of accident events. This activity dies rad unhfy tim functusus opvatun of plant systenn requiral to resporal to accidents The ruiri iritrusivericss of die test equiprnent and the rninunal burden imposed by this equipment will not render any c ungment or ,
equipnent inopsable in addition, the activity will be limital to a single tram of plant systems required to respwd to design basis axi&mts in tir ewnt a anann measuring device failure should occur during ,
tle short time that it is amnected, the faihue would be inunediately evident to the technician and >
appropriate steps would be taken to assure the operabihty of treessary equipment for safe operation based on 1. uniting Con &tions for Operation. In adation, the devices allowed by this evaluation have features such as high inpit impdarxrs and/or fuses which would limit the propagation of device failures into plant circuits.1he measunny doices will only be conr.ected to ciremts for a abort time duration and under continuous supervision of quahlied crafi personnel.11s linutation that such dev ces have high mput impalarne arnt be used only for voltage mensmements or cunent measurements using a clamp <m ampic rneter prechden the triteracti<m of tlw rneters wIth operable circuit components. The activity will not change instnament hiop accuracy, setpoints, or resporne time. The conservative treatment of pstulatal device faan en reqmres condderation of postulated dnice open and short circuit con &tions.
1he unnapnees of such failures wdl ixt result in tim loss of circuit functionahty and therefore does not intn thxx trw types of nudfunctiorn Ikraue this activity does not result in a change in any plant feature or impact on plant equipment amiciated with the scope of the Technical Socifications, there is no change in the margm ef safety.
Revision to Fire Ilantds Anaksis TITI.!!: Fire Ilantds Analysis (Fl!A) Change Notice 95 m - Fire Ihr !! valuation for Door 1007 DliSCRIPTION: Fire Protection !!rmncering livaluation 95-02 was incorpunted into the Fila by this rnision. This evaluatm detennined that fire d nr 1 D07 maintains its operability without a functioning closer or latch because this function is supplanted by the use of dead bolts and security alann switches SAFl?TY ANALYSIS- The abihty of the subjec. Act to prevent the spread of fire to adjacent areas is not affected, only the mearn by which the d or is maintamed closed is bemg changed The probability of fire occurrence is unchangal lhe uswcq unrs of a fire are unchanged beyond what has been previously evaluated in the Fire Ilanuds Analysis aid Safe Shut &mn Analysis. The change in how aior 1007 is maintained closed actually decreases tir probabiht) of malfunction by replacmg the higher faihire rate closures and latches with more reliable dead bolts and alarm switches. Fire is the only accident for which this change is appheable The only failure m<de with an impact on safety is door opening. Changing how the door is closed has no impact on, nor will it change, its failure mode of gc.ng open The rnised bolting is an improwment in maintaining the das closed The margm of safety is not reduced as fire barTier integrity is mamtained Amendis R Procram Chanca TITIli: Appenda R Validation l'roject Documentation Updates DliSCRIPTION: As part of tic Appendis R Validation Project, several station druments required inision to reticct the new analytical results A conumm Safety livaluation was performed for these documentathm uplates. s The changes inchaled USAR Change Request 97 124; Fire Ila/ards Analysis (Fila) Change Notice 97 02; Appendix R Safe Shutdown Change PacLapes %46 and 97-03; and Pncalures 6.FP.606, 5 4 31,and 5 4 3 2. Illa Change Notice 97 02 removed duplicate information specific to Appendix R nmiphan,r smcc all mformation specific to Appendix R comphance is contained in the Safe Shutdown Analysis Repet (SSAR), and also identified new Fire Protection lingineering !! valuations written as a result of the Pnyect Safe Shutdown Change Package E06 d icumented the changes made to the CNS Appendix R arxl Attemate Shutdown Analysis resuhing from the implementation of the Appendix R 92-
t t
l Vahdatxo Pniject, and (kcumented the resuhs of the rnimi analysis for each fire area Safe Shuttkmn ,
Change Package 97 03 documented the chanres to die CNS Apperalix R and Ahemate Shukkmn Analyms endting fann the inclusion of Station Air Comptessor 111 Die USAR was te ised to remove ;
spuric infbrmation regarding Appendix R compliance and to reference the SSAR for this infwmation i Changes to the SSAR will be amtmllal through the 10CFR50.59 safety evaluation process. 1he SSAR w as enhanced funn a hunum faches perspective to prmide a trore ducct link from the analysis resulta to the aunpliance strategies for a given fue noe and the CNS licensing bases Procedures 5.4.31 and ;
$ 4 3 2 were rnimi to relicci the compliance strategies validated by the Project. Procalute 617.606 was iniwd to reacct Appendix A fire baniers which were upgraded to Appendix R requiral fire ;
barriert 1he net cIIcct of all of these chanFes is an increased level of plant anfety as a result of ,
umitk ration of additional fire induced faults and a more conservative overall compliance strategy that will ensure the capabihty of CNS to safely shut down folkming a fire event. ;
SAFl!TY ANAINSIS: 1he subject changes do ruit increase the puibabihty of occurrence of a plant event prninusly evaluated in the SAR twause they are styresentative of the ruults of a tuote conservativa analytical methakilogy arul do not intraluce or change accident precursors or irutinkirs crnhted in the SAR. 1he te ined analytical techniques and their output (k) not increase the probability of a fire or other special event. ,
These changes do not increase the consequences of a pre ic.usly evaluated plant event because they prmide a nave untrvative respme la mitigating a fire event than previously evaluated. The bounding [
nasumption of Appendix R analyses is that all equipment in a given fue tone is subject to fire inducal damage arxhunpensaksy actum are specified to miti Fate these malfunctions no that s3fc shutdown can be achieved The probabihty of occunence of equipment malfunction remains the same. The changes denumtrate the capabihty of CNS to mitigate the consequences of fire induced equipment malfunctions within the op tational capabihties ofranaining plant equipuent assuming a single worst ene fire iruluced spurious operation Therefore, the consequences of fire induced equipment malfunctions remain within the previously analped results for an Appendix R scenario. Ric changes do not intnx!uce any new !
accident precursors or initiators Equipment is being operated post fire within analped system design basis limitations and no new equipment failure modes are intraluced The changes do not rnluce the margin of safety as dermed in the basis for uny Technical Specification because Technical Specification equipment is being operated within analped design basis parameters in ressmse to a fire event. Post fite safe shutdown capabihty is a condition of the I,icense and is the basis for Technical Specification 3 21 lhe n;uipment and comsments hsted in the Technical Specification are unchanged as a result of the tevised analytical restdts.
F.mer cency Puredure Guiddms TITI.th CNS limergency Pmeedure Guideline (EPO)(Revision 3)
Dl!SCRIPTION: Numerous changes were made to the CNS EPO, which is the guidehne used to develop the CNS Emergency Ol uating Pnmlures (EOPs). Some of the more significant changes included: !) direction of a umtrollahk pressuritation to assure core steam flow exists for core cooling while efforts are being made to estabhsh reacht pcssure vessel injection sources 2) euidance to start Core Spray reference leg in etum aller an enugency &pressurizatum if adequate ave omling is not assured. 3) clarification when emerpency depressuritation is required based on potential radiation release rates outside Primary and Secondary Containment, and 4) rnision to relicct changes in Cold Shutdown thron Weight, llot Shutdtmn ikom Weight, and Ikom Injection Initiation Temperature that resuhed from incorporation of Design Change 94 041.
SAFliTY ANA13 SIS: Dunng plant ewnts associatal with the EPO sections being te ised that require EOPs, the CNS licensed design basis will already have been exceeded, and the safe recovery of the plant becomes the matter of paramount impcarm lhe NRC, m its Safety Evaluation Report on the llodmg Water Reactor Owners Group El% fourxl the use of the limits specified in the EPGs rather than those specified in the licensed design basis acceptable during degraded conthtions The implementation of the revised EPO will not increase the probability of occurrence or consequences of an accident or malfunction of equipment imputant to safety prevuusly evaluated in the USAR.1he revised EPO cannot increase the probability ,
ofoccurrence of any event analped in the USAR because the procalute will only be used after the event
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has ctrunnmi The implementation of the revised IIPO will tot create a possibihty for an accident or malfunctum of a ddlertut type than any previmsly evaluated in the USAR because the revised EPO dces not modify the operatim or design basis of the plant. I or plant cmdations uh6h aheady excml the licen cd design basis, the question of a reduced margin of Safety is tot meaningful.
limcgency Procalute Guideline TIT!Ji: CNS Emer rency Procedure Guidehne Desistions/ Justifications (r.PO D/J)(P.evision 3;
" ESCulPTION: Varmus desiatum est between acti(m hsted m the CNS EPO and m the generically written Revision 4 of the lkiihng Watcr React (v Owners Group (ll%1 TOG). The deviations are a result of different installed systems arul equiprunt and standard terminology used at CNS. The CNS D/J prmides a listing of each
) deviation and la justification to show that the intent of the procedural actions in the llWROO EPG is -
maintained Numerous changes were made to the EPO D/J document by this revision. Significant clumges were consistent with the changes made to the EPO, Rev. 3 discussed almve.
h MP Y AN " YSIS: During plant events associated with the lipo D/J sections being revised that require !! ops, the CNS licensed design basis will already have been exceeded, and the safe recovery of the plant bewmes the snetter ofparanunna nputance.11e NRC,in its Safety livaluation Report on the llWROG !!PGs found the use of the limits specified in the EPGs rather than tho::e specified in the licensed design basis acceptable during degraded cmditions 1hc revised CNS EPO D/J will not increase the probabihty of occurtence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR because the procedure will only be used tojustify actions that will be used in comparison to the llWROO !!PGs after plant condnions have reached EPO entry conditions. The implementation of the resised EPO D/J will not create a possibility f r an accident or malfunction of a different type than any previously evaluated in the USAR because it does not modify the operation or design basis of the plant. l'or plant conditions which already exceed the licensed design basis, the question of a raluced margin of safety is not irsaningful.
Gunibcation livnluation Package (Cl!P) 97 100 Tl11.!!- Clip for N111 PS 5i A,11, C, and D DESCRIPTION: This CEP reclassified the low low Set pressure switches Nill PS-51 A,11, C, and D from Environmentally Qualified (EQ) to Essential (E) These switches are hicated outside the drywell.
Changing the c(anpnent classification from *EQ" to "E" identifies that the com;mnents are not required to be quahfied for a harsh emironment.
sal 1?TY ANAL YSIS- The low low Set pressure switches are used to mitigate downcoiner thrust loads of subsequent safety /rehef valve (SRV) actuations during an abnormal transient or a small break loss of Coolant Accideat, These switches arc imt an accident or event initiator. The reclassification of these switches will not increase tW dose to the pubhe. During events when these switches are required to operate, a harsh eminnment will not be created in the area of the low low Set pressure switches Therefore, these is no rea$m fw these switches to be "EQ", ahhough they do perform a safety function requiring them to be "E" The switch function or operation will not be ahered by this reclassification The low low Set logic is designed such that no single faihire in the electrical circuits would cause a spurious SRV actuation, The change in component classification does not affect any setivints for the low low Set pressure swJches lherefore, this actisity will not reduce the margin of safety as defined in the basis for any Technical Specification.
, , - - - , - w. .n- yy ,r, . . , - - .e,m - - - - - - . --
-,-,v .,----. .-
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Safety Evaluation for Plant Restart Ti1LE: Operation with the ADO System Secured and Assessment of the impact on the SOTS DiiSCRIPTION: 'this Safety Evaluatxn was perfmned to evaluate the aweptability of plant rest .rt fnun the July 29,1997, shutdown with the Augmented Off Oas (A00) system isolated The isolation of the ADO system was regtural to assure the operabibty of the Standby Gas Treatment (SOT) systcm until a nahfication could Ic installed to eliminate a cmoem with SOT opcrabihty (refctmce MP 97-068A). This safety Evaluation evalustal SOT operabihty with the A00 system isolated and evaluated the impact on plant operation and offsne dme emuluenxs inun pla it startup and nortnal power operation with the AOG system isolated.
'the uiginal evaluatim was appuval August 2,1997, arxl coveral a one month period of time. A second evaluation was performed to evaluate Se acceptability of cmtinued plant operation until September $,1997.
SAIETY ANALYSIS: The ADO system, whether operating or isolated,is not a plant event initiator. With the AOG system isolutal, quability of the SOT systun is assurcd. 'the only event which cmsiders A00 operation is the Control Rod Drop AccidmL I kmmr the dose consequences of this uent w cre analyzed both with and without operation of the A00 system with the conclusion that the dose consequences for teth cases remain withimeeptable limits It is concluded that while operation without AOG in senice w111 result in a higla netnity lew! in plant efliuent, no adverse c!Tects to safety equipment or personnel result from such operation, the increase in pnijected dose to the public danng operation is insignificant, and such operation is allowed by the CNS Technical Spdfications and within the NRC's current guidance contained in 10CFR50 Appenda 1. The oft 0as system, via operation of the dilution fans, ensures that hydrogen ancentration is maintained below 4% by volume, thereby protecting against a hydrogen burn event, Operation without the A00 does rut impact any plant equipment or structures which could initiate a plant event arxl has to impact on Primary or Secondary Containment operability. Ilypassing the AOG will not impact the ability of the Off Oan system isolation valves to perform their function, ensuring the ability to isolate the Off Ons system is maintained This activity will not exceed the acceptance limits nor the calculatal eme;uences for any previously evaluated accident. Ilami on data ecliccted, the Technical Specification Section 3.21.C.4.c threshold dose reporting limit was reached during mid-August.
I kmewr, gaseous emuent o!Tsite dose projections based on actual emuent releases during this time and assuming operation with the A00 bypassal for an entire year, indicate that such operation would not exccal 30% of any Tedmical Specification limit for dose to the public. While the th eshold for formally reporting such operation to the NRC in accordance with Technical Specification 3 21.C.4 e has been execulal arx! a Special Reput to the NRC is required, operation is expected to remain within the overall gaseous elliuents concentration and olisite donc limit of Technical Specification Sections 3.21.C.1 through 3.2 i C.3. In either case, operation with the ADO secured does not impact the margin of safety with respect to the NRC acceptance limit for any essential system.
1.icense Change Reauest (1 CRs 90-0032 TITI,li USAR Change Emergency Core Coohng System (ECCS) Keep Fill Sptem DESCRIPTION: This change incorporates the plant mahfication implemented by Minor Design Change (MDC) 73 2, which bypassed the presste a control valves for the reactor buildmg auxiliary condensate system ECCS keep till lines. This MDC was implemented in response to a preoperational test in which it was determined that the system, as originally installed, cou!d not perform its intended task. A LCR was initiated in 1990 when it was detmnhed that the plant change had not been incorporated into the USAR.
The LCR incorporated changes to reticet the plant as it was tested dunng the preoperational testing.
Ilowever, a safety evaluation was not performed in conjunction with the LCR at that time.
SAFETY ANALYSIS: MIX' 73 2 was initiated because the original system could not perfonn its intended task The purpose of the keep fill function is to ensure rapid delivery of design flow upon system actuation by maintaining the piping full of water. The capability of the system to perform this function, as mahfied, was denxostratal by the satisfactay cmmleton of preoperational testing. and by the past performance of the system. Alarm indication oflow condensate supply pressure is provided m the main Control Room to 95
indicate a possible loss of water fill in the imes Since this change does txt impair the ability of the system to maintain the piping full of water and does rxt, afTect any precurexrs for plant events pres ously evahiated in the USAR,it cantxt irxrease the probability of occurrence or consequences of a plant event or malfunction oIcquipment important to safety previously evaluated in the USAlt it does not change the function of"he system, and does not afTect the menner in which the ECCS system perfirms its interded function 1he change d es txt affect any plant operating parameters and does not create any new or unusual operat onal nudes fir the plant; therefore, the change car not create the possibility of a plant event or equipu ent malfunction different than any previously evaluated in the USAR. Technical Specification 31.0 requires that the discharge piping for the Core Spray subsystems, low Pressure
, Coolant injectim subsystems, liigh Pressure Cmlant injection system, and Reactor Core Isolation Cooling system he filled when the systems are required to be operable. The keep fill function of the Condensate Syst :m functions properly to maintain the !!CCS discharge piping full of water, therefore, the margin of sal:ty as defined in the basis of the Technical Specifications is rnd reduced Ligense Chnnee Reauest (1 CR193 0001 TITlJh USAR Change Minimum Torus Water Lewl DliSCRIP110N: 'this LCR arrecta d the torus minimum water level in Table V 2 4 of the USAR to correspond with the torus minimum vater volume of 87,650 cu. A. The minimum torus water level listed in USAR Table V 2 4 had a valu s of 2 ft 9 in below torus centerline. Table V 2 4 lists an associated minimum torus watcr mluns vah e of 87,650 c18. These munters are not consistent. Ila ei on a review of the Plant Unique Analysis Report (PUAR),it has been detennined that the USAR water level value of 2 ft 9 in.
should be changed to i ft 9.5 in.
SALT!TY ANALYSIS: Per the USAR Chapter XIV, Station Safety Analysis, the minimum water level in the torus does not impact the hkeldxud ofevaluated plant events. The minimum torus water level as described in the SAR is bemg chanpl to match the minimum torus water level volunw. per the PUAR. The consequences of an analyred plant event have not been increased and no facility changes are being made. Nonnal plant operating level remains the same. The impact oflow torus water level has already been addressed in the lloiting Water Reactor Owners Group limergency Procedure Guidelines, Appendix C. The Guidelines are symptomatic in nature and cover all credible accident scenarias.1herefore, the likelihood of a new plant event or failure unic has rxt been intrahral The Technical Specifications list the minimum torus water level at 87,650 cu A The water level limit of I ft 9,5 in below torus centerline curTesponds with this value. The margins m the Technical Specifications are being maintained License Chance Request (1 CR194-(031 USAR Chnnee Reauest (UCR) 97 130 T!T1Ji: USAR Change . Flood of Record DliSCRIPTION: These changes uplatalinfivmatim in the USAR concerning the flood of ret ord. Changes were required to incorporate the new maximum Missouri River level of 900 8 feet Mean Sea Ixvel (MSL) and other pertinent information related to the dad which occurred at CNS in 1993.
SAFETY ANALYSIS: Systems and equipnwnt will not be affected such that they are degraded or operated outside their design or test limits. Plant systems, structures, and equipment, includmg equipment designed to control the
- release of radiation, are protected from flooding b3 upstream dams, a levee constructed by the Corps of
!!ngineers, and by the site itself which is one fmt ateve this levec at a MSL clevation of 903.5 feet. In addition, a station emergency procedure is put into effect at a river elevation of 890 feet MSL, which estabbshes prinusy and secondary barricales to protect plant systems and equipment. A Notification of Unusual !? vent (NOUl!) is declared at 899 feet MSL river elevation which allows mitigating the consequences of any accident which may occur e high river elevations If ricer level is firecasted to execal or reaches 902 firt M3L, the plaat is placal in a sluitdown condition per Technical Specifications.
The 1993 maximum find level of 900.8 feet is lounded by the probable maximum 11ood (PMF) discussed in the USAR which has a projected upper limit of 903 feet MSL If the river level rises atmve 96-
t 0.e existing levee, an urumtrollal spreadmg of a massive volume of w ster througimt the six mile wide valley at the station site will occur and will in all likeliland result in an inunediate i m cring of the river l r
smface elevatmn 1he marpm o safety as dermed in the bases of Tecimical Spec ficathm 3.13 is tot reduced by the 1W3 axd level of 900 8 feet MSI. 1he CNS Fhed Procedure is implemented if the river level reaches 890 feet MSI., which is well below t!w PMF level the Corps of lingineers has i calculated at 903 feet MSI and the PMF the NitC has evaluated at 9012 feet MSL Further, the Technical Sp,cificathm margin of safety is maintamed by the And pnwalure whic i calls for a reacne shutdmn upe rmttlicathm of an upstream dam failure, ihmd u aters reaching or foro *sted to reach 902 fie MSL, water axumulatan in the Diesel Ucnerant noms, Reactor fluildmg quad i. Control lleilding !
bawment, or the intake Structure Service Water Pump Room License Chance lleouest fl CIO 91 rC72 '
USAll Chanue Itcouest (UCit! 97131 1111.1!: USAR Change . Core Spray (CS) Flow Requirements !
DliSCHIPTION: These changes uplatal USAR Tables 17 1 and VI.3 1, and 11gure VI l.2 to reflect the concet CS flow requirements and ihnv paths The Cornhthm I flow path is wi h CS suction from the torus which is the flow path incorporated in the accident analysis. The Condition !! fknv path is Innn the Condensate Storage Tank (CST) and is not tested SAlliTY .
ANAL.YSIS: 1his redske does not irnuhr any cluinges to plant equipment, processes, or procedures lhe pararnetera in the USAR revision are those vahics used in the accident analysis in USAH Chapters VI and XIV. In particular, the CS flow rate used in the analysis is 4720 ppm and the CS suction is fnm the torus. 'lhe
<mly tine that the Ch pump is alkm ed by Technical Specifications to be aligtal to the CS1 rather than i the torus is during refueling with the head off and ikuded up in this condition, a loss of Coolant Accident is outskie tic licensmg basis, therefore there is no neal to test this conditiott The changes are in agreement with the reymrements and basis in the Technical Specifications; therefore, there is no reduction in the margm of safety, l.icense Chance Request i1.C10 94 0122 TITLi!: llSAR Change Testing of Core Spray (CS) Valves bliSCRIPTION: 1his I.CR ievisal USAR Satam VI, Cae Starxiby Coohng Systems, to note that the CS inboard injectk n throttle valves are not equipped with bypass vahrs for equali/athm When these valves are tested at pressures >450 psig, a manual operator is used to equali/c pressure by partially opemng the valve. The i USAR was previously incorrect and this change sellects the as built cork'ition.
SAFIITY ANAL.YSIS. Pressure equehzathm is required to test the intmard CS injection valve, however, the metlal of equah/atim does tot afket piping c; arnponents within the test boundary since all the compments are designed to withstand the pressure sanges. The test boundary is controlled via the test pacedure and
, there inim differvnce in the pntabihty ofcausing a plant event using the motor valre instead of a manual
( by pass vahe. The valves are manually operated vtjle the allowed out of service time (AOT) for the
(; nuntillance pmcahire is applat 1he AOT ensures Technical Specification required action completion
, times for the CS subsystem are not exceedal so the consequences of a plant event during testing are l luuxled by the Technical Specification requiral acti ms for the CS system The motor valves are j designed for manual operation at the ddTerential presswes which will be experienced during testing lilectric motor operation automatically disengages the manual operator so the valve remains functional dunng manual numputathn 1he only type of plant event that could be caml by inappropriate pressure equahiati m is overpressuniatko of the CS low pressure discharge piping resulting in a loss of Coolant Accident, wiuch is luuklal by acewient analysis. The possibihty of creating a different plant event is not creatal 1he only types of malfunctnos camal by inappmpnate pressure equalization are motor operated valve faihue or piping faihre which are analyicd and tvunded by current accident analysis and Technical Specificathin requiral acthms for the CS systent The margin of safety as defineOn the basis for any Technical Speedicatkin remains unchanged ,
97
. - _ ___ _ , .. - . . _ __ _ ___ _ ~ ___ _ - . _-. _
Gee Channe Reauest (11R) 94 0126 lillj!: USAR Change . Control Rai Drive holenoid Valves DliSCRIPTION This 1.CR revised Table Vil 31 of the USAR, Process Pipeline Penetrating Primary Containment, to irxhcate that the mntrol rod drive exhaust and inlet solenoid valves do not require air to oper ate. It also ,
t revimi a artespeximg rude for this table to clarify that the solenoid valves do rot operate on a scram signal arxl to nure accurately reflect the function of the valves.
SArl!TY ANALYSIS: 1his diange ciertets enors in USAR Table Vil 3 1.1he changes to the table tellett other design basis (kuarnits ard actual directional control valve operationt Since the change is merely conecting enors ,
in the table, there is no increase in the probabihty of occunence or consequences of a plant evert or malhinctxo of equipinit important to parcty. There is no reduction in the margin of safety for the lasis of any Technical Specification License Chnnee Rea icst fl.CR) 95 0005 1111J!' USAR Change Standby Natogen Injection (SilNI) and Post-less of Coolant Accident (LOCA) Vents Dl!SCRIPTION: The US AR was revised to indicate the majonty of the SilNI and Post LOCA vents were designed for beyond desiF's basis events and inalled as nonessential equipment, per Design Change 89 272. The SilNI syst m was irwtalled to tving CNS into compliance with the design intent of 10CFR50.44, Generic
- 1. citer 84-09, and Regulatory Guide 1.7, SAll!TY ANAL,YSIS: Per the Design Change, the SilNI and Post LOCA vents are intended Ibr beymd design basis use.
Ticeftve, these systems, by defmition, were installed to mitigate or prevent consequences beyored those previously evaluated in the SAR. No new fadure nxxles have been introduced and no changes to the facility are being snade which could affect reliable plant response during analyzed coinhthnis. These systems do ext intaduce any new accident initiatois which are rud botaaled by existing analysis. Safety e stem dependencies and interfaces also remain unafTected as compared to the original expected plant response. The impact of using SilNI and Post-1.OCA vents has been addressed in the lloiling Water Reactor Oners Group Emergency Procedure Guidelines The trargins in the Technical Specifications are ;at admsely impachi by these systems sirxe, by design, their function is to mitigate or prevent plant events beyond the design basis.
Liteme Chance Request (LCR) 95-0010 TITI.ll: USAR Change . Figure Vll 4 4 Dl! SCRIP 110N: This LCR revised USAR Figure Vil-4 4, Core Spray System Ph!D, to show a spare valve w hich was peviously unlateled, CS V 178 SAll!TY ANALYSIS: 1his vahr is a spare nuunnal valw which is rusmally closed and settes no safety or design basis function it is un!lir calibration only and is outside the Primar, Containment boundary. Therefore, its presence has no elTect on the probabil'ty of occurrence or consequences of a plant event or malfunction of eqmp nent important to safety. This valve does not suppet any Technical Specification bases License Chance Request (LCRi 95 0011 TillJI: USAR Change . Uplate of Calculation Reference DiiSCRIPTION: This LCR revised a calculation reference m Secunn 11 of the USAR fnun NEDC 88112 to NEDC
, 94 271. This referoxx supputs the USAR statement that the plant is designal for a safe shutdown under the most entical low water elevation of 865' for the Missouri River, De design of the plant was not changed, the previous calcula: ion was seperseded by a different calculetion.
98-
- e.,,,,.A..w,.,-.v.n-., n.,. y,-- -
n.- ,n, . . . . -v.,,. .- ,,.---_ee.. ~-, ,- ,--r-
. .- - -- -- _ ~ _. . _ --.- ~_- -
Mall!TY ANAL,YSIS: 11e change in the reference does not alTect the design as described in the USAR tv Os design margins associated with the low water level.11rre is no change to the facihty or any pncess. Therefwe, the pobabihty of twurrence or consequemra of a plant event are rad increased. No equipment irnpmtant to safety is alTected by the change. The CNS Technical Specifications do not address the most critical low w atcy clevatim f(t Oc mit.11crefore, the rnargin of safety as dermed in the basis for any Technical - ,
Specification is rnt afixted by this change.
USAR Chann Reauest (UCR) 96-On$ i T11L1!: USAR Change . Reactor Coolant System f lydmstatic Test :
Dl!SCRIPTION: 1 his UCR removed a requirement from the USAR to perform a hydrost. tic test of the reactor coolet ,
system following each removal and replacement of the reactor vessel head. Such a pressure test is not mquimi by any ade, standard, rule, or regulatkut The USAR will state that insenice pressure tests are ,
performed in accordance with the requirements of ASMll Sectkm XI.
SAFl!TY ,
f ANAL,YSIS: 11e Reactw Presstne Vessel (RPV) test is rx4 a pwurar to any accident scerwio. No credit is asmtmed for the pressure test in accident nutigation or in the calculation of accident probabilities n radmlogical ,
efTects of an accident. lilimination of the subject test d(es net increase the probabihty of an accident.
Any leakage from improper assembly of the reactor head would be detected by the leakage collection system or by the Operations walkdown during startup, leaktight integnty is demonstrated by reactor uulant presmre leurnlary perkdic testing in accordance with ASMl!Section XI. Also, the temioning !
of the RpV studs emures that de pont peload is applied to achieve the calculated gasket scaling stress
- - arx! dms assiars lenktightness of the head Any leakage through the RPV seah would is trapped in the Prinuny Cmtainment 11e effect on offsite doses would be bounded by the hne break ar4alysis. Changing the test description to be consistent with ASMI!Section XI does rad create the posmbihty of a difTerent type of accident or malfunction of equipment important to safety. Technical Specifications require periodic pressure testmg in accordance with ASMH XI. There is no requirement in Tecimical Specificathms for any additional pressure tests when the reactor vessel head is remo'ed and pla
- ed.
USAR Chance Requat(UCR) 96Q21 TITIA USAR Change Alternate Rod Insertion (ARl) and Recisculation Pump Tnp (RPT) Desenptiom OliSCRIPTION. The ARI and RPT deicriptions in the USAR were revised to reflect the as-built condition of the plant.
The as built ARI logic does not provide a time delay for any of the initiation conditions; therefore, ,
references to a time delay m high reactor vessel pressure and low reactor water level were removed The RPT description in the USAR was revised to indicate the reactor low water level imtiation function is time delayed SAFl?TY ANA1,YSIS: Tic safety evaluation for Design Change 86-3411 (AR! ATWS/RPT Mojificatiom) aal the NRC Safety livahiation Reput (S!!R) for CNS Canpliara with Anticipated Tramient Without Scram (ATWS) Rule 10C116062 relating to Akl and RPT systems are not alTated by this USAR change and remain valid.
11 tin change dacs rut alter Or design or operation of any system, structnre, or component. The ARI and RPT systems are pmvided to mitigate the comequences of an ATWS event. The capability of the ARI arxl RPT systems to cope with an ATWS event has been foimd acceptable by the NRC Stir relating to ARI and RPT systems lic ATWS event is beymd the CNS design basis and the ARI and RPT functions are not classified as essential Carecting the US.AR desciiptions does tot affect operation of the plant or impact the cmsequences of equipment malfunctiort Tlese changes do not result in violation of any Technical Specificatxm or reduce the margin of safety as defined in any Technical Specification llases 99-
. ,- _ , _ _ -. _-- - - _ _ __ ___ ._ _ m
h USAR Chance Reouest (UCR1964:65 TITIJi- USAR Change . Correction of Drawing lima DliSCRIPTION. Thi. UCR reviad USAR Figurs Vll.81 to unat valve Comsment identification Code ernes made when the subject drawing w as uovertal to a Computer Ai&d Drafhng (CAD) drawing.
SAll?TY ANAL,YSIS: 1his is a drawing erna unectum m!y al does rus involve aoy physical changes to the plant or have any eficct on plant urnpuus ts, systems, or structures as described in the SAR. Therefore, this change dm not affect the probabihty of owunenu or consequences of a plant event or malfunction of equipment important to sefety. There is also no clTat on the snargin of safety as dermed in the basis for any
- Technical S;wilicatiort USAR Chance Reauest iUCR) %D66 !
Till.lt USAR Change Turbine liquipment Cooling (TI!C) Pump Motors Dl! SCRIPT 10N: lius UCR reviull>SAR 1able L7 2 to provide c4rtect nameplate information for Tl!C pump motors.
TEC pump motor si/e, vuhape, and RPM were proiously incorrect The correct values have teen confirmed through research of the onginal purchase contract aral the llurns A Roe design criterie doemnert SAFl!TY ANAL,YSIS: Actual plant design supports the 11:C punp inotor narneplate values and operating procedures are >
unaffected by the corrections Therefore, this change dxs tot affect the probabibly of occunence or uncluences of a plant event or inalfunction of equipment imp
- tant to safety and the rnargin of safety is una!Tected USAR Chnnee Reouest (UCRi %069 TITI.lt USAR Change . Figure Vill 4-4 DliSCRIP110N. lius UCR rnised USAR Figure Vill 4 4 to bnng the drawmg into agreement with Pncedure 2.2.19A assi the actual plant configuration Cubiele heations IC at MCC 1, and 2D at MCC T were corrected.
SAlfTY ANALYSIS: '4 his change avrects cubicle hientions only. Therefore, it will not increase the probability of occmtence or unuluences of a phuit ewnt tv malfunction of equipment important to safety arnt the margin et safety autelitied m the basis for any Technical Specification is not atiected USAR Chance Reauest fUCR) %070 TITI.lh USAR Change . Figure VI! 12 1 DiiSCRit'T10N USAR Figwe Vll.12-1 was uplated to the etutent sevision of the applicable station drawing. The drawmg had been revised to incorporate a wordmg change to a drawing note as a result of Design Change 91-088, Main Steam 1.ine Radiation Monitor Scram and Group i Function Rernoval. The revision of this drawing was initially overkwed by the modification. When the drawing was i,ubsequently revised, the conesponding USAl; tigure was inadvertently :mt updated.
SAFETY ANAL.YSIS: This USAR ligure ts uplated to re0cet the cunent plant configuration The modification that changed ,
the plant cortfiguration wan evaluated per 10CFR$0 $9 and the design received prior NRC approval (refenmec License Amendment No 158) Since the updated figure is consistent with the modification, this c..ange does not increase the probabihty of occurrence or consequences of a plant event or malfunction of equipment important to safety and the margin of safety is unchanged 100
?
i I
@$ Chaner Recuests (UCRs) 06 071. %072. 96478. %079. 96-0U. %085. %-08H. %089.97-01H. 97 021.
21E11 ,
'llTII: late. rial USAR Changes DliSCRIPTION. 1hese clumpe roquests made variout editorial changes to the USAR. Included in these editorial changes uvre 1)anatun of drawing number on USAR drawins label. 2) correction of typographical error in -
valve number on USAR Figure Note,3) carwtion of title on USAR drawing label,4) two corrections of acrerences to another weton of the USAR,5) two cortections of Reference Numbers,6) catection L of typographical aligmnent enor, 7) typographical correction of "250 volt DC" system which was incottectly identifini as "230 volt DC," N) coricetion of valve number error on USAR Figure, and 9) removal of calculation revision level from a reference.
SAll!TY ANALYSIS: All changes are ahkeial iri nature. There are no changes to any system, structure, or compment (SSC),
or changes to procedures afTecting any SSC which is the initiator of any plant transient or accident or i relied uno to mitigate any plant transient or accident. 'there are no changes to equipment important to safety or changes to pnuxlures alTecting equipment impodant to safety which will increase the probability of occurrence or consequences of a malfunction of equipment important to safety. These editorial changes do not reduce the margin of narety as dermed in the basis for any Technical Specificatiort.
USAR Chanoc Reauest (UCRi 96-084 TIT!.li: USAR Change . Spent Fuel Pool Wator f.evel Dl!SCRIPTION: 1his UCR acceted the USAR description of t!r nunimum spent fuel pool water level required whenever imxhatal fuel is st<mlin the spent fuel pul The USAR previously specified a minimum cover of 9 feet of water above the fuel assemblies; this was changed to 8 % feet. This change makes the USAR consistent with the Technical Specification requirement for level SAIT!TY ANAL,YSiS: Fuelluvullmg axident or seismic event probabihty are not afTected by this change because initiators for these events are not afTected Changing the amount of water requiral from 9 feet above the fuel to 8 %
feet above the fuel does not degrade fud iategrity. Minimum water level for fuel handling accident mitigation is presenul with the Technical Spccification limit and overall pool level is unchanged hxline decontamination factor and halogen hold-up time assumptions are preserved The height of the fuelis limited by the limits of the fuel hanaling ;nachine and is not afTected by this activity. Fuel integrity is unaficctal und consequences remain boundal by the fuel handling accident analysis. This actis ity does no' impact integrity of the fuel assembly or cladding. No other mitigation SSCs ate atTected No new failure mechanisms are intraluced. This change makes the USAR desenption of minimum level emastent with the Technical Specifications, therefore, the margin of safety as defined in the basis for any Technical Specification is unalTected.
~ USAR Chance Recuest (UCR)96-086 TIT 1 li: USAR Changt .FigureIV 3 3
~ DliSCRIPTION: USAR Figure IV 3 3 was revised to show the as-built configuration of the 3/4" Primary Containment (PC) piping downstream of PC.V 59 and PC.V.60. This corrected an error which resulted from a previous draning changC.
SAFl!TY ANAL.YSIS: This drawmg change sixwn the co:Tect as tuilt configuration of the sensing imes' cross-tic and low side dram line which are part of the PC boundary. This activity does not involve any physical change to the PC boundary, has no adverse efTect on any sys:cm interfaces, and does not change the system design function. Lis change will have no radiological effects on the PC system and will not increase the consequences of any accident. The PC system performance and reliability will not be decreased, thus, the probabihty of equipment malfunction will not be increased No new or unanalyzed failure modes are 101
i i
i created. The mm tic arxl drain hre are part of the wrixt emfiguration for the PC rensing lines ard n*e rxt dianged by this drawing change. This activity has no effect on the margin of safety as defined in the 1 basis few any Technical Specification l
Q%R Channe Reauest (UCR)96-095 :
TI11Jh USAR Change Standby Liquid Control (SLC) system DliSCRIPTION 1his USAR change punides clanficatxo of de design bas;s of tbc SLC system and clearly sta'es thst th St.C system is not essential. The text. as prniously wntten, could have twn mi2nterpeted leadmg to
. an ernmeous conclusion that SLC is an essential system. The SLC Design Criteria Document was also revised to refixt changes awcinted with the USAR and to accurately reacct source &cuments. i SAll!TY ANAL YSIN: The SLC system d(es not operate in any way that it can contribute to the initiation of an accident. As ,
such, de probabihty of c.xunence of any accidents is unchanged. The SLC system is not credited in the accident analysis as a nutigatxo system SLC is credited in the evaluation of special events. The subject ;
cluinge Aos rud change se void any of the design basis / requirements used in this evaluation. Therefwe, this chai.ge has no impact en any of the accidents previously evaluated in the SAR. The probability of I aviatence of a malfunction orcquipment impetant to safety is not affected by thinhange as the system t remains as origmally designed 1he fiulure nodes and effects analysie for St.C or any odn.t system is not changed These changes have no efTect on the perfortnance or reliabihty of the SLC system These changes do ret alTect parameters unt in the bases of the Technical Specification requirements for the SLC system; therefoie, the rnargin of safety is not decicant USAR Chance Reauest (UCRi %099 TITIJL USAR Change l'igure X Il 1 Dl!SCRipTION: 1his drawing was uplated to reflect the as-built condition of the coagulant systera in the Makeup Water Treatment lluikhr4 The changes reflect the removal of components from the coagulant system in the pretreatment pation of the Makeup Water Treatment System (MWTS)
SAFliTY ANAL.YSIS: The removed components have never been used 1hc MWTS surveillance and monitormg pncesses ensure the punty lewl of the M%TS is maintained beknv design limits. Thus, the absence of the removed parts has no impact on corrosion of essential systemdcomponents. Other systems (e p , condensate demineraliters, reactor water cleanup demineralucts and tneir asxciated monitoring inst 4umentation) are als in place to further ensure that unter purity is maintained within Tedmical Specification limits lhe rmoved cangunents do rud haw any impact m essential electrical distribution s.,_ . all components rmanul are meelumical All rmoved comp,nents are kcated in the MWTS building and are classified as Class !! Seismic. ily definition, the structure, equipment, and components are important to reactor operation, but are not essential for preventing an accident which would endanger the public heahh and safety, and are not requiral for the mitigation of the consequences of an accident. This system is not associated with any of the accident initiators discussed in the SAR. This chc.npe has no impact on this or any oder system which could increase the dose either during operation, during an accident, or during a malfunction of equipment. The coagulant system is not discussed in the basis for any Technical Sgwification, therefore, these changes do not reduce the margin of safety as defined in the buis for any Technical Specificatioit USAR Chance Request (UCR) %100 TITLli: USAR Change Request Figure X-6-1 DiiSCRIPTION: Durms regeneration of the asuciated station drawing fmm hand drawn to a Computer Aided Dralling i
generated drawing, an erin was nale in that the Residual Ilent Removal (R1 IR) pumps w cre incorrv.ly identified as Iligh Pressure Coolant injection pumps. This UCR corrected the drawmp and associated USAR figure.
102 r
mm z.,. m.- --~ ~_. r- , --
i i
SAFETY [
ANALYSIS: This change shows the correct aebuilt cmfiguration of the idIR pump cmlers, which are part of the Reactor Closed Cooling (RCC) system his change does not involve a physical change to the RCC !
system, has no etrect on any system interfaces, and does tot change the system design funt Lion. His !
activity has no radological effects on the RCC system and will riot increase the comcquences of any l accident. As system performance reliability will not be affecksi, the probabihty of an equipment malfunctmn wdl ruit be increaul No new or unanalyzed 1ailure rmies are created. This activity has no ;
c!Teet on the margin of safety as defined in the basis for any Tahriical Specificatior . ;
WiAR Charwe Reouest (UCR) %103 '
TITLlh USAR Change Request Sewage Treatment Facihty i
DliSCRIPTION; This cimnge remmul diseunion of nunt< ring and chkrination of the Sewage 1 reatment Plant's ellluent j
~
prior to discharge into the plant's Circulating Water discharge canal. This is no kmger done as a result of Design Change %99 which imtalled two laga ns to handle all the Sewage 'l reatment Plant's efiluent.
saltily ANALYSIS: This change does not alter the function of any safety related plant equipment, nor does it alter the inairiteriance or operation of ariy equipment or systerns important to ainfety. It does rmt afroct any equipment or systems used for mitigatmg actions, nor does it alter or alTect any abrunmal or emergency ogsrating pns:cdures Therefore, this activity is incapable ofincreasing the probability of occurrence or consequences of a safety sigruficant plant event. A malfunction of the Sewage Treatment Plant is incapable of creas ing any environmental or pmccan corxliti(ms conducive to or causing the malfunctior ofequipment impatant to safety. No new faihire trusles are it;tnsluced. This activity does not affect any assumptiom, calculatims, design dwumentsar design criteria used to establish the mergins of safety of any safety relakxl equipment or components Thus, no Technical Specification margins of safety are alTected.
USAR Chance Reauest (UCR) 97 001 TITLik USAR Change Turbine liypass System Dl!SCRIPTION: The USAR was revised to state that the turbine bypass system consists of three automatically operated regulatig valves, instead of statmg that these valves operate automatically and sequentially. Tis regulating valves operate at the same time vs. sequentially.
SAFl!TY ANALYSIS: This activity kes not change the overall system ihnetion or reliability as the bypass valve will still perfism its design functian of controlhng reactor pressure. A worst case failure of these valves can only -
result in an abnomial operational transient (ie , generator load rejection or turbine trip) and cannot become an accident initiator because overpressure protection is provided by the Safety Relief Vr,1ves, w hich will prevent this situntum from escalating to an accident. Thus, simultaneous operation of the bypan valves will not increase the prubabihty of an accident previc.usly evaluated in the SAR. The turbine bypass valves do nos perform any accident mitigation functions, theref(re, this change will rmt cause any increased radiological elTects Whether these valves are cperated sequentially or simuhannusly, the resultant pressure changes remain well within the design limits of all equipment and will not induce any equipment malfunctions or failures The simuhancous opening of the three valves could potentially cause a ditTerent rate of pressure or temperature chanpc but this is judged to be within the design limits of all atTected equipment and bounded by other accident scenanos. Any potential accident causing bypass valve faihto is lumded by the previously analyzed generator k>ad rejection witimut bypass and turbine trip without b> pass. The Station Safety Anahtis is tot based on sequential operation of these valves. This change does not alter the function or reliabilinmf the turbiac bypass
, system, nor does it afTect any assmr.ctions, calculations, pacedures, or dest . weifications used to establish the basis for derming the plant's margm of safety. The existing margin. <h.fety are unaffected 103- ,
i
.- ~ _ _ _ _ _ _-.
l i
USAll Chanoc itcouest (UCIO 97 004 .
Tllidt USAR Change . l' ire Protection (ll') System Pip ng Dl!SCRIPTION: 1his change was made to clarify that the 11' system piping is Class !! Seismic.1 km ever, w henever the 4' 11'systcm piring pases over or near safety related Class i Seismic systems or equipment, it is designed to uithstarsi a Class 1 Seismic occunence and maintai structural arxl pressure integnty.
%All!TY ANALYSIS: 1his is a USAR change only, not a physical change to the plant. The subject piping has been analytal ,
to awure that it mil nsntain its Ntructural athi pressure intcKnly during a CIn%s 1 Seismic or barge impact (
L occunence lhe 12P ystem cannot cause the failure of any safety related sptem or component.
1herefore, the probabihty of occurrence or consequences of a plant event or malfunction of equipment important to safety are not increased the 11' piping will function as originally designed and will not rniuce the mat gin of safety as dermed in the basis for any Tec!nical Specification.
USAR Chann Reauest (UCIO 97-006 1
TITidi: USAR Change Tmbine Ocnerator fluildmg Ventilation Systern DESCRil TION. 11e USAR w as revised to reflect tic proper configuration of die Turbine Generator lluilding Ventilation System lhe control circuitry dws rmt include logic to stop the fans when temperature decreases below 40*F;lumever, there is annunciation in the Control Roorn A cacton w as also made to reflect that numitoring mstrumentation in located on the fan supply vs fan discharge SAll!TY ANAL,YSIS: 1his change to the USAR system description scaccts the system conuguration sixiwn oli USAR Figure X-10 2.1here is no change being rmnic to any r' .nt system, structure, or component. The Turbine Generator iluilding llenting and Ventilating System is not required for the safe shutdown of the plant.
Neither the Turbine lluildmg temperature ret the location of the temperature element in considered in any plant event evaluation Procedural controls are currently in place to minimi/c the impact of not tripping the fans prior to reaching the free /ing p> int. This conected configuration has no impact <m radiatum m its numitoring/ logic ami will thus not resuh in any increase of exposure to personnel. The Turbine lluikhng Ileating and Ventilating System is nonessential Adequate controls are in place to prwhkle this configuration from creating the smibihty of a di!Terent type of plant event or malfunction of equipment imprtant to safety.11e Turbine 11ailding llenting and Ventilating System is not discussed in the basis of any Technical Specification and airects no Ttxhnical Speci6 cation margin of safety.
USAlt Chanec Reauest (UCIU 974W TITI.lt USAR Change-l'igure13 2 DliSCRIPTION: The subject USAR figure is a Cow diagram of symbols and abbreviations A note was added to this drawing which states that " Chicago fittings, quick disconnects, hose connections, and other equivalent fittings are slamu lir mfmnatnmoly.1 hey are not required to be as built or sinmn m the drawing, and may or may not be installed" 1his note will allow the use of these fittings without requiring implementation of the plant nadi 6 cation process SAllITY ANAL,YSis: Use of these littingdconnections will not affect the pnmary coolant pressure tvundary or isolation capability of any system Primary bmtairunent valves arwl caps are administratively controlled by procedure arxl are also identified in the licld by being painted )citow. Use or non-use of the subject littings and connections is not an accident mitiator. This change allows for the installation of fitdngdecimeetions for he purpose of hose connection, tool connection. etc. and will not resuh in any uneased radiological effects.1his actisity will not induce any equipment malfunctions or failures and u c or no-uw of these littingdurnstons does not atTat any plant equipment important to safety. Ticir use will not adwrsely aflixt the strucunal mtegrity of the apphcable pipmg systems Appheations of the Chicago littmgs, quia disconnects, Imse connections, etc. will still be controlkxl by normal plant mamtenance una engineering controls Use of these fittings will smt alter the design or function of the 104, t
.. __ _ _ . - - _ . _ ._ ~ ._ . _ . _ . ____ __
eqtupruit they are installed m lhis actaity will not alter the reliabihty or accident nutigation capability of any equipment or systems, run skes it afTect any assumptions, calculations, pnuxlures, or design si ecificathms used to establish the basis for defining the plant'6 margin of safety.
11SAR Chance Pgaevt (UCR)97-014 TITIJi: USAR Change Core Spray System Vahr Control DESCRIPTION: This UCR resises the sectxo of the USAR [utaining to Core Spray System valve control to indicate that tlure are fota pressure switches in a one out-of two twice logic to momtor system pressure and that they i pro'ide an ogut tumissne signal to the asclunge valves.1his change updates the subject text to reflect the as-btult omfiguratkn W previms descriptam of the low pressure opening pennissive was identical to the desenpuon in the FSAR submVted in 1971. The circuit logic was nulified to its current cmfigur atxo dunng the (xnstnetxo p..ase basal on 0:ncral Electric elementary drawings w hich reficct the current circuit logic.
SAFETY ANAI YSIS: 1his change does not involve a physical change to any plant systemn or components and d es not affect tir marner in w hich plant systems perform their intendxl functiort 11 des not afTect any precursor for plant events previously evaluated in the SAR. The current coufiguration will exceed the perfmmance characteristics or the presious two switch arrangement. The logic change enhanced overall reliabihty by meetmg the single failure pnof design ritena lbr both injection arxl low pressure pipmg protection.
1hereftre, this change thies m/ increav the probability of occurrence of consequences of a plant event or malfunctko ofequipment imprtan: a safety.1he current configuration is reflected in the Technical Sgrificatkm 11cef(ve, this dinnge d: es rut reduce the margin of safety as defined in the basis for any Tecimical Specification USAR Chnnec Reauest lUCR197-Ol 5 TIT 1JE USAR Change Technical Support Center (TSC) EmerFency liypass Fdter System Dl!SCRIPTION. The USAR was revised to reficct that the TSC limergency llypass Filter System includes two IIEPA fiers and not just one as previously indicated SAFETY ANAL.YSIS: 1his change does not involve a physical change to any plant system or equipment. it does not alTect any precursor for plant events previously evaluated in the SAR. This change does not affect any plant parameters and thies not n!Tect the emitorunent in which plant equipment functmns. It also d es not afket the manner in which plant systems perfonn their intended function The two filter system which in mstalled will meet or esemi the performance of a single filter system Therefore, this change to the USAR canrut increase the probabihty of occurrence or consequences of a plant event or mal?nction of equipment import. int to safety. There are no Technical Specificatkms related to the TSC Emergency llypass Fdter System, therefore, this change d es not reduce the margin of safety as defined in the basis for any Technical Sncification 1)fAR Chance Reuucst (UCR)97-017 TITI.th USAR Change - Suppression Pool Water Temperature ,
I DESCRIPTION: The L.1AR discussion of suppressmn pool water temperature indicatic, and recording was revised.
There have been several changes made to the suppressim pool water temperature kops over the years I
that were not clearly reflected in this section of the USAR.
SAFETY ANALYSIS- 1hc tmrdmg and indicatim of the suppressum pool water temperature does not have any impact on the initiators of plant events These indicators / recorders provide the Control Roorn Operator with information on how the plant is operating and provide an input to the annunciator system to alert the Operator ofincreasing torus water temperature. liased on this information, the 0;-rator may perform I
l ecstam aesm ile current configuration supplies the Operator with dnisionally separate and qualified l
l -105-l l
instrunnitauan owr tic raguired range in:essary. lluefore, the plant can be operated as designed.1hc ctatent equipment is of a higher safety elassification than was originally installed The system does tot ,
aukanatically initiate equipment importarit to safety nor could it prevent the equipment from operating.
Ib new types ofmalfanctums have been intnslucal. The current configuration is eflected in Technical
! ceifications, therefore, there is not a reduction in the margin of safety.
USAR Chance Reuucst (UCR)97-019 TnLIL USAR Change Containment Temperature Monitoring Dl! SCRIPT 10N' The USAR was rnised to reflect that the range of the containment temperature nomtoring in the area o rthe safetyhclic," valves is 50-600*F, rather than the previously stated range of 50-350*F.
SAFl!TY ANALYSIS: This revisko tellects the rmuutormg range of the installed equipment. It does not afTect the equipment installed in the plant and does not alTect the ability of any plant system or compment to perfonn its intended function. This change also does not alter the temperature or emironment within the antainment.1his revisam does not afTect any precursor for prniously evaluated plant events and does not increase the probabihty of occunence or consequences of equipment malfunction. Technical i S weificatko Table 3 2 F, Pnmary Containment Instrumentation, lists the subject temperature indicators and specifies the range as 50 350*F. The temperature indicators weic reiaced with new instruments which have a range of 50400*F.1hc new instruments allow monitoring of a broader range of tcruperatures in this vicinity of the drywell, and exceed the morutoring range specified in the Technical Specifications. No specific Tnhnical Spec:lication actions are taken for these instnanents The instruments will still be able to be esed by the operators to momtor plant conditions, therefore, this
. :anpc does not reduce the margin of safety as defined in the basis foMhe Technical Specifications.
USAR Chance Reauest (t tCR)97-022 TITIli USAR Change USAR Tcble Vll 12 5 DliSCRIPTION: This UCR iemoved a previous change made to Table Vll 12 5 by 1.icense Change Request (LCR) 95-0025. 1his proious I CR was initiated to change the number of miaimum required operable Reactor fluilding isolation Ventilation Radiation Momtor channels from aae to two Ilowever, this change was prematurely ironistatalinto tic USAR witixmt the appmval of a corresporaling Technical Specification change and witiumt an appanni 10C1 R50 59 cvaluatko 11cefore, this UCR returned the afTected table lak to its conditam prkv to I.CR 95-0025. While the changes made to this table under LCR 95-0025 wcre conservative in natur . they canrot be implemented as a change to the licensing basis widmut a conemtent Tee mical St reification change.
SAFliTY ANALYSIS: 11us UCR umnters a prnious change that was implementa! without an approved Technical Specification change or 10CFR50 59 evaluathm Since the licensing basis was, in efTect, never changed, the restoration to the original configuration has no etTect on the pnibability of occtutence or consequences of a plant event or malfunctkm ofeqmpment important to safety and the margin of safety as defined in the basis for any Technical Specification is not reduced USAR Chance Reauest (UCRi o7-023 TU1.ll: USAR Change low Pressure Coolant Injection (LPCI) Pump low Flow Dl! SCRIPT 10N: TaNe Vll 4 4 of the USAR, low Pressure Coolant injection Instnnuent Specifications, contained an incorrect range and accuracy for the LpCI pump low flow switches. During construction, the instrumentation loF4 was changed from irxinidual pump flow switches to hop flow switches for the operation of tic numu.am flow valves llowe&r, the change in instnunentation was not reflected in the USAR.1he flow range was changed from 0-800 ppm to 0 3600 ppm and the instrument accuracy was changed fmm *2% to *1%
106-
i SAMiTY ANALYSIS: 1he minimum flow functim was nxdified during construction from individual pump dischage switches to hop flow switches. No actual change to the plant will be made by this UCR. This change does not affat any precunm fit plant evmts evaluated in the SAR.1he minimum flow valve is meant to protect the pump funn damage w hen operating at near shutofThead, this change does nct alter the perfwmance of the fur.etion. 1herefore, the Id'CI system is still able to perform its intended function within the capacity as stated in the SAR. This change corrects the listed flow range ard accuracy of the flow switches to make them consistent with the mstalled equipment and the setpoint given in the Technical Specifications. Therefore, there wdl be no reduction in the margin of safety as defined in the basis fbr >
any Technical Speedications.
USAR Chance Reauest (UCR)97-024 11TLlh USAR Change Piping Penetrations DESCRIPTION. This UCR deleted Table V 2 2," .etration Schedule and Testable P ' trations, and Table V 2 7, I Testable Primary Containment
- olation Valves The appropriate intwnation from these tables was transferral to Figure V 4 1, Prirr ey Contairunent Vessel Penetration location and Schedule, and Figure V 4 3, P imary Containment Suppression Chamber Penetration location and Schedule. Incorporation of these tables onto figure drawings streamlines the USAR and makes it more user friendly. No mfmnation was deletal during the transfer ofinformation onto the figure drawings. Ilowever, additional information was added as authorized by various mahfication packages and other approved CNS documents.
SAlliTY LALYSIS: 1his UCR does not imulve a physical change to 'he Primary Containment (PC) system, has no .fTect on any system interfaces, and does not change the system design functiort All physical changes to the PC system were made by either existing station modifications and theit appropriate saf:ty evaluations or during construction and preoperational testing. The changes are to varhus penetration desenptions and/or isolation salves and their kications which were madvertently overkioked by the mahfications.
Thus, this activity will have no radiological elkets on the PC system and will not increase the asnequences of any accident or malfunction of equipment important to safety. PC system perfmmance and reliabdity will not be n!Tected, therefore, the probability of equipment malfunction will not be mereased 1his UCR does not alTect or alter the design or function of any plant eqmpment important to safety, 11 only rnises the USAR to incorporate pertinent PC information in one hication. No new or unanaly/ai faihire males are avata! This activity has no effect on the margin of safety as defined in the basis fbr any Technical Specification.
USAR Chance Reauest (UCR)97-026 TITLli USAR Change Pipchnes Penetrating Primary Containment (PC)
DICCRIPTION USAR Table VII 3 1 and curespnhng text references were revised to reflect actual plant configuration.
1his table is strictly for pipchnes in which tirir respective isolation valves perform an automatic isolation function ihr PC purpmes Sneral vahes that did not trxet this criteria were remm ed frorm Table Vll 31 and others were added as appopriate.
SA WTY-
- ANALYSIS: 1his UCR does not imuhr a physical change to PC, has no etTect on any system interfaces, and does not
- change the system design function All physical changes to the t'C system uue made by either existing station nnhfications and their appnipriate safety evaluations or dunng construction and preoperat ional testmg. The changes are to correctly identify those isolation valves that perform an automatic isolation function and meet the criteria for being listed on Table Vll-3 L Thus, this actnity has no radiological etTects on the PC system and will not increase the consequences of any accident or malfunction of equipnent inputant to safety PC system performance and reliability will not be alTected, therefbre, 'he probabihty of an equipment malfunction will not be increased. This UCR does not alTect or alter the s
- 107-w - -,-- -- - - - . - - -.~-y-, --
y- - - .w-.--
3 -.y-_me- w- + ~ - . , -- y-.--e.---.-f g--- e+4.e-g. g.-
design tv furstion of any plant eqwpment important to safety. No new (* unanalyred failure nxxles are crcated This actisity has no effect on the margin of safety as dermed in the basis for any Technical Specification.
USAR Chance RU ugst (UCR)97-028 TITIJL USAR Change Steam Jet Air Ejectors (SJAEs)
DESCRIPTION: The USAR was revised to remove the statement that the SJAE sinrn supply valves inolt'e upon sensing Ndi ternperature or high pressurc. The applicable sentence now indicatea that only the air ejector inlet (condemer air outlet) valves isolate.
SAFETY ANAL.YSIS: 1his actnity will rd change the owrall design, function, or reliability of the SJAE steam supply pressure antrut valves tr the Main Steam tr oft Uns systems. The function of these valves to control steam flow to the SJ AEs will not be afhtal by currectmg their description in the USAR. 1he twiginal system design did not provide fr an automatic isolation function of the SJAE steam supply pessure control valves; therefore, their non-isoh, tion upon sensing high temperature or high pressure will not be an accident initiattr.1he air ejector inlet valves will an:1 isolate u;xm sensing high pressun or high temperature in the SJAE discharge piping, preventing the possibility of fire or explosion in the main condenser in the emnt of an Off Uns system expknion or burnback. The steam supply valves are txmessential and do not perform any accident mitigation function; therefore, this change will not result in any increased rathokigical efTects This activity will not affect any plant equipment or systems important to safety, nor will it induce any equipment malfunctions or failures. 'lhe original form, fit, design function, and operating parameters of the SJAE stcam supply valves are unchanged. The Station Safety Analysis is not based r
- the isolation of the steam supply valves This change does not afTect any assumptions, calculations, procedures, or design specifications used to establish the basis for defining the plant's margin of safety.
USAR Chnnec Reauests(UCRs)97 032. 97 03197 014.97-035 TIT 1.It USAR Changes USAR Figures XI-61, X-h-6, and X 12 1 DESCRIPTION: The subject USAR figures were updated to relicct revised Inservice Inspection (ISI) boundary flag ,
hications. The revised boundary flag kications are in conformance with CCFMD 1, the ISI 13oundary liasis ASME XI Classification liasis Document.
SAFETY ANALYSIS: The ISI boundary classification is for inspection purposes only and does not afTect the ability of any sys'em, structure, tu aunponent to perfmn its safety function. The boundary classification does not alter the physical or operating characteristics of any component. Therefore, there is no increase in the probabihty of occurrence or comequences of a plant event or malfunction of equipment important to safety. A change to the ISI txumdary does not afTect the basis for the ISI Technical Specification. '
D ul Chance Rcouest (UCR197-043 TITLE: USAR Change Reactor Coolant 1.cakage Into Primary Containment DESCRIP flON: 1his change cluninatal an inconsistency between tla USAR and Technical Specification descriptions of reactor coolant leakage limits. A paragraph in the USAR that was inconsistent with Technical Specificatic.n iequirements was deleted.
SAFETY-
- ANALYSIS
- This USAR change restores consistency betweca the USAR and Technical Specifications concerning
" identified" and " unidentified" reactor coolant leakage rates into Primary Containment. It will not increase the probabdity of a plant event or result in an increase in otTsite dose, nor result in a failure of
- i. the detection capability for determining coolant system leakage. This change does not alte any >
1- operational or accident pa.: meters of any plant equipment. The credible failure mode of the leakage l detection system in the Primary Contairunent remains unchanged and no new failure modes are created.
1 lOS-
t Restoring cmsistency between the USAR and Technical Specifications will not reduce the design margins of the Containment leakage Detection System.
USAR Chance Restest (UCR)97-044 ,
TITIA USAR Change Standby Liquid Contiol(SLC) Solution Tank Alarm Octpoints DliSCRIPTION: This cha ge removed the actual .dann setpoints listed fm the SLC Solution Tant. Alaim setpoints I provide indication the solution volume has changed, which might indicate a change in solution concentration. The actual tank levels are umtrolled by de solution volume vs. cmcentration requirements of the Techmcal Specifications, and not tiy the alann levels.
SA11ITY ,
ANALYSIS- The SLC system operstmg and monitoring capabihty ternains unchanged, therefore, there is no increase ia the anwquences of any plant nutt. Design acuments are in place which strictly control tim specified setpoints for the higMow level alarm points Technical Specifications strictly govern the solution mlume emcentration requirements The volume concentration requirements are controlled and met to enwe proper mlume is available witlout risk of solution precipitation No increased failure probability is intrtsluced This change does not alter the requirement to maintain the system within the !imits seniGed in the Technical Specifications The SLC system will respond as before to shut d nvn the N.ictw. No new failure trusles are creatal The margin of safety is not reduced because thi; change dxs not impact the solution volumc concentration reqmtements that are required for SLC to perfonn its safety function USAR Chnnee Reauest itICR)07-045
'iITlE USAR Change Standby Liquid Control (SLC) System Room Temperature DISCUSSION. Ihis change rnised the normal room temperature for the SLC system. It also re ised the detennination of the minimum nuun temperature basalon the adjusted Saturation Temperature vs Concentration Cu.ve found in Technical Specifications. This change rnised the room temperature since the solution temperature in the piping is now maintained by heat trace.
SAlliTY ANALYSIS: This change does not alter plant equipment or implement any changes in the metlux! of operation of the plant. ihe SLC heatmg system has tem anal >7al for the ce,ceted nxnn tempen..ure range, and therefore there is no incicased probabihty of equipment failure. Clanfication of the governing temperature vs.
saturation curve d ies not change any operatmg characteristics, nor does documenting the previcesly analyzed capability of the SLC heating system to maintain the solution without precipitation within the range of ex1xetalnunn tempe:atures. This change (kx s not reduce the margin o! safety because it chies not change the system capability for which the SLC system has been designed and analyzed.
USAR Chinee Reauest (UCR197-0M TIT 1.l! USAR Change + Standby Liquid Control (SLC) System Capabihties Dl! SCRIPT 10N: 1he USAR was te ised to chtnfy the capabilita of the SI C system during an Anticipated Transient Without Scram (ATWS)cvent. This change clarifies that the SLC system n ould be capable ofinjecting liquid control solution at a higher than nonnal reactor pressure, which would be expected dunng an ATWS event.
SAETY ANALYSIS: ' lhis change does not alter plant equipment or implement any changes in the method of operation of the plant, therefore, it cannot increase the probabihty of a plant event or malfunction of equipment There is no increase in the cmsequences of a proiously evaluated plant event because the changes only provide clarification and desenbe the capability of the SLC system dunng the clevated pressures of an ATWS nunt.1hc change pnwides clantication of the operating condition for which the SLC system has been designed and analyzed There is no change to Inw the SLC system operates or to its associated proculures 1his clumpe &cs tot alter the requirement to mamtain the system within the hmits specified 109 w . _ , - . . . - - ,- . .. . . . -
in the Techrucal Specifications, tv alter any Technical Specification hmits.1here has been no change to any parameter which can reduce the margin of safety.
US AR Chance Reouests (UCRs197-047 and 97 128 TIT 1.II: USAR Change Seismic Quahfication of SmallIkite Piping Dl!SCRIPTION, The USAR was revised to clarify the desip basis for small bore piping used at CNS. The USAR irxxrnrtly stated that dynamic analy=es were performed f(v all Class 1 Seisraic piping systems.1his is only accurate for h rge lxte piping 2 %" and greater in diameter. Small txte piping and supports less tluin 2 %* in diameter wue field routed using span cliarting pr cedures and engmeering judgement 1his nx60d has been recognized by the PRC as an accepted method on a number of occasions. The USAR was updated to tellect the actual design basis for small bore piping.
sal 4?TY ANALYSIS: This 1ange kithe USAR clanfics the original design basis for the small bore piping. It does not change the method used in the design of the piping systems, nor does it change the original safety factors used in the design of piping less than 2 %* in diameter at CNS Therefore, this hange does not have any impact on any structure, iiystun, or component important to safety am! will not irF7 ease the probabihty of occurrence or consequences of a plant event or malfunction of equipment important to safety. This change to the USAR does not in any way n!Tect the original design basis for small bme piping at CNS.
Since the orig %al design parameters are not changed by this activity, the probability of a difTerent t)pe ofplant event <s quipment malfunction is txt created The desinn metints fe small bore piping at CNS have been recognized as acceptaNe practices Since the safeiy factors used in the design of small lxite piping ate not afTected, no margin of safety as d.fmed in the basis of the Technical Spt ifications is alTected USAR Clance Remmt (UCR)97-044 TITI.I!- USAR Change Standby 1.iquid Control (SLC) System injection Times Dl! SCRIPT 10N: This change revised the solution injection tunes as stated in sections of the USAR pertaining to both single pump capabihties and simuhamxus operation of both pumps to achieve ;ompliance with the NRC Anticipated Transient Without Scram Rule. This USAR change was required to tellect SLC system injection times based on the minimum and naximum tank levels calculated in Nuclear lingineering Department Calculation 93142 and stated in the Safety liveluation performed by the NRC ibt Amendment 173 to the Operating License that mercased the required reactor pressure vessel boron conecutation This clarification only indicates the actual solutsn injection times based on the required tank vohanes and uinimum Technical Specification flow rates.
SAMITY ANALYSIS: 1his change does not alter plant equipment or implement any changes in the methal of operation of the plant, therefore, there is no increase in the probability of occurrence of a plant event er malfunction of equipment important to safety. The change provides clarification of the injection timea;uired for one or two pumps to inject the required amount of solution. h cannot increase the consequences of a reactor shutdown witixot cxmtrol rals because the SLC system wih shutdown and maintain the reactor suberitical as requtral The clarified injection rates are within design and Technical Specification limits. There is no change to how the SLC System operates or to its associated procedures This change does not alter 1
.- the requirement to maintain the system within the hmits specified in the Technical Specifications, not does it aher any Technical Specification limits There is no reduction in the margin of safety tw:cause the reactor will be shutdown and ma^mtained suberitical by SLC within the required time frame as before.
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USAR Chance Reauest (UCR)97-055 TITLE: USAR Change - USAR Figure X 10-lH DESCRIPTION: The subject USAR drawing was revised to show the connection of the blowdown piping fron. Elec* ic Hoiler ID to the Blowckswn llender and to indicate the correct valve type. This information was
- inadvertently not reDected on the drawing at the time ofinstallation of the electric toilers under Minor Design Change 30-57.
SAFETY ANALYSIS: 3e mechanical portion of the electric boiler installation was designed and installed in accordance with applicable ASME Cm!cs and addenda. The electrical portion follows IEEE 3841981 for electrical separc*n The subject drawing changes to show the correct valve type and the blowdown connection are n t a:ed witn the non-safety related mechanical porti(m of the electric boiler installation. Therefore, there i, ao irnpact on the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety. The margin of safety as defined in the basis for any Technical Specificatmn is not reduced by this change.
USAR ChpagqLuest (UCR) 97 056 TITLE: USAR Change - USAR Figurc Vll 8-1 DliSCRIPTION: This Nuclear Boiler flow diagram was revised to add a sensing line connectian between LIS-101 A and omkmsing chamber 3A. This change cor % a drawing error and depicts the as-built configuration of the plant.
SAFETY ANALYSIS: This UCR results in no physical changes to the plant and has no effect on plant operation The change
, will mt increase the radiation exposure of plant personnel or the public from a plant event or equip at malfunction previously evaluated in the USAR. No new equipment is being added. The sensing ime connection added to this drawing is not in the Technical Specifications or defineo in the basis for any Technical Specification; thus, there is no etTect on the marFi n of safety.
USAR Chance Reauest (UCR)97-057 TITLE: USAR Change . t'igure 1115 5 DESCRIPTION: This USAR drawing uas revised to change the ASME Class boundary llag d e spa m tr the Control Raiilydraulic System and to atkl a note to define the reactor coolant pressure low ^ y *hese drawing changes conform with the USAR text.
SAFETY ANALYSIS: The luntary classfication is ihr impection purposes only. It does not affect the ability of any component to perfbrm its safety function,nor alter the physical or operating characteristics of any equipment. These changes will not increase the radiation exposure of plant personnel or the public from a plant event or equipment malftmetion previously evaluated in the SAR. There will be no equipment added or any physical or operational changes in the plant. The boundary classification is not in the Technical Specifications or defined in the basis for any Technical Specification; therefore, there is no reduction in the margin of safety.
USAR Chance Reauest (UCR)97-059 TITLE: USAR Change - USAR Figure X-14-3 DESCRIPTION This USAR drawing was revised to corr :ct the Component Identification Code Ihr the Radioactive Laundry Room sorting sink drain vahe from PW-V-1271 to RW V-1271. Radwaste (RW) is a more appropriate system designation for this valve than potable Water (PW)-
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SAFETY ANALYSISL _ This UCR resuhs in no physical change to the subject valve or the system. The valve and system will still meet the same design requirements as before this change. His valve does not serve a safety function.
His activity does not change the function, operation, or reliability of any equipment important to safety, axr will it induce any equipment malfunctions or failures. Changing the system designation of the subject valve will not affect its abihty to function as designed. This activity does not affect any existing interfaces between the radwaste system and any other equipment or systems. h does not affect any assumptions, calculatims, procedures, or design specifications used to establish the basis for derming the plant's nargin of safety, therefore, exinting margins of safety are unafTected.
USAR C. Ngde, ini10CR)97-065 Till.E: Editorial USAR Changes DESCRIPTlJN: This UCR made various editorial changes to the USAR as follows: 1) deleted duplication in two cmsecuthe pages,2) curected misnumbered step,3) corrected error in column alignment,4) corrected typographical error,5) corrected pagination error, and 6) made addition to Table of Contents.
SAFETY ANALYSIS: The changes associated with this UCR are editorial in nature. By definition, editorial changes have no consequential ellect on the allect:d sections /pages. Accordingly, the probability of occurrence or consequences of an accident or malfunction of equipment in,portant to safety are unchanged. These changes have no impact on the margin of safety as defined in the basis for an" Technical Specification.
IISAR Chance Reouest (UCR197-069 TITLE: USAR Change Standby Liquid Control (SLC) Relief Valves DESCRIPTION: An enugency TecJmical Specification change was submitted to remove the SLC relief valve surveillance requirement. In mler to allesiate NRC cmeerns over the licensee controls that will supersede the deleted Technical Specificat;m requirement, this USAR change was made in advance of the NRC's approval of the emergency change. This change deletes the description of the 1450 to 1680 psig relief valve pressure barxicmtained in the Technical Specifications and replaces it with the more restrictive 1540 psig
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band estaldished under the setpoint control program. Additionally, the safety basis of the relief vahe settings is clearly described and the requirement to test the relief valves during each operating cycle is explicitly listed SAFETY ANALYSIS: His change pmvides additional levels of detail to existing USAR text and has no adverse efTect on any of the ATWS precursers described in the USAR. The change captures the testing basis currently used to emurs operabihty of the SLC rehef valves. Tlurfore, tne consequences of an ATWS are not increased by this change. There is no dirn.t etTect on the reliability of the SLC relief valves, or on any SLC component that relies on SLC relief valve operability since the proposed change captures the existing testing requirements This change does not constitute a physical change to any SSC described in the -
USAR or a change to existing analyses described in the USAR. Therefore, the consequences of relief vahr failure will not be increased The information being incorporated in the USAR is already reficcted in existing station procedures and processes. No new failure modes are introduced that are not already txmnded by USAR analyses Technical Specification 4.4.A.2,a ensures that the SLC relief valves will not lift spuriously during SLC system opantion, and will prevent system overpressurization by lifling when required. By replacing the design band with the tighter "as-left" band, the margin of safety is unatTeeted l
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1 DIAR Chance Reauest (UCR)97-075
-TITLIL USAR Change Automatic Depressurization System (ADS) Relief Valve Accumulators DESCRIPTION: This change corrected references to the design criteria for the six ADS relief valve accumulators. The -
USAR presiously stated that these accumulators are sized to contain sullicient pressure for a minimtun of "five valve operations (actuations)" following failure of the nitrogen and/or air supply to each Neumulator. Baml upm an investigation of design requirements, the USAR was corrected to state that the accumulators must be capable of providmg suflicient pressure one hour after loss of the pneumatic supply to provide two valve actuations at 70% drywell pressure, which corresponds to five actuations at atmospheric drywell pressure as the testing requirement. These conclusions are based upon Nuclear Engineering Department Calculation (NEDC)88-306 and an NRC letter from D.13. Vassallo ta l M.
Pilant conceming Quahfication ofADS Accumulators, dated August 9,1985 This clarification was also inwrporated in Design Criteria Document No. 8, Nuclear Pressure Relief System, and Design Criteria Document No.15, Instrument Air System.
SAFETY ANAL.YSIS: This change involves clarification of the USAR and Design Criteria Documents to provide the proper reference for the six ADS acetanulators' pressure requirement. This clarification correctly characterizes the safety function of t.,:sc accumulators regarding required valve actuations. This safety function is already provi<ied by the existing equipmert. This activity does not alter any design or operating parameters, create new interfaces, or afTect any existing assumptions used in operating or accident analysis The margin of safety as defined in the basis for any Technical Specification is not afTected.
USAR Chance Reauest (UCR197-076 TITI.E: USAR Change USAR Update Requirement DESCRIPTION: The Code of Federal Regulations was changed in 1992 to extend the USAR update requirement of 10CFR50.71(c) faun annually to a refueling cycle basis, not to exceed 24 raonths. The requirement for annual uplating was removed frtin the USAR and replaced with a more general statement that the USAR updating requirements will be met in accordance with 10CFR50.71(c).
SAFETY ANALYSIS: The submittal of USAR updates is an NRC reporting requirement. Changes to these reporting requirements have no etTeet on any event precursors evaluated in the USAR, nor do they have any physical or analytical etTeet on any systems, structures, or components described 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. Vese USAR uplate requirements are not credited in the margin of safety of any Technical Specification.
USAR Chance Reauest (UCR197 031 ~
TITLE: USAR Change - Power to Flow Map DESCRIPTION: This change to the Power to Flow Map revises the "Do Not Operate In This Region" shaded area of the curve Etis area was established to prevent cavitation of the jet pumps and reactor recirculation pumps.
'this UCR revised the map to identify a portion of the subject area as "RR Min. Speed Only Region." The map had been revised in 1991 with no justification or discussion of why this area was included as a portion of the "Do Not Operate in This Region" shaded area. The change to the map remains conservative for all previously specified cavitation lines.
SAFETY ANALYSIS: .This change to the Power to Fknv Map climinates the prohibition against operation in a particular region and replaces it with nominal region boundaries within which operation is restncted to minimum pump speed. This change does not physically alter any plant system nor are any of the operator actions or system operations changed as evaluated in the SAR. No accident initiators are alTected nor are the power to flow conditions outside the normally cefined operating region. The two pump minimum flow Ime is detennined by system configuration and is alTected by restrictive forces withm the core and instrument
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calibration ne low feedwater flow recirculation pump runback line is set at approximately 20%
fculw ster flow which equates to approximately 25% core thermal power. Ilowever, both of these lines are norninal lines and arc established by the core conditions that exist at the time the region is entered-nis change does not affect any of the parameters that establish these nominal region tvundaries but does estabbsh that the region boundaries are not exact limits. This change does not afTect core thermal limits bccause the region to be ch nged is below 25% core thermal power. At core thermal power levels s25%,
the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience has indicated that the resultmg Minimum Critical Power Ratio value is in excess of requirements by a considerable margin. TLis change has no effect on reactor recirculation pump or jet pump Net Positive Suction licad limits because operation within the cavitation region is not allowed.
Threfiire, this change does not increase the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety and the marFi n of safety as defined in the baris fa any Technical Specification is not reduced.
USAR Chance Reauest (UCRi 97-082 TITI.!L USAR Change . Description of Circuit Ilreakers Dl! SCRIPT 10N: Certain previously existing 345KV air opc*ated breakers were replaced with spring operated, SF6 gas qiarching breakers. Different corxhtions are now required to ensure proper ooeration and are interlocked with the breaker close circuit. Procedure 2.2.14 was revised to reflect these new interlocks As the eleven breakers in the 345KV switchyard are now a variety of breaker types, the USAR was resised to inchtle a rixire general term of"powv ' circuit breakers, in place of" air" and " air blast" circuit breakers.
USAR Figure Vill 2 1 was also revised to change the description in references to breakers 3310 and 3 312 from "A.C.D " to "P.C.lk"c SAFliTY ANALYSIS: His activity increases the reliabihty and performance of the generator output breakers while maintaining their original function The main generator output breakers are classified as non-safety related and are suit maintained within a 10CFR50 Appendix Il program; however, this activity was intended to install a better ami more reliable component than previously existing. In the event these breakers should fail, the wc;st case result can be asumed to be a loss of olTsite power which has already been analyzed in the SAR. No credit is taken for tie operation of these breakers. The generator output breakers are not credited with any event mitigation and do r.ot interface with any of the event mitigation systems.
Replacement of these breakers does i:ot involve work on any primary coolant toundary nor does it invohr any release paths. This change does not increase the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety. His activity was perfbrmed while the plant w as in a emhtion consistent with Technical Specifications regarding the availability of on and oiT site power sources. The invohed components do not perfonn a safety function and are not included in the basis for any Technical Specification.
USAR Chnnee Request (UCR197-092 l Procedure Chance Reouest (PCR16 HPCI 313 (Revision 5)
TITLIE USAR Change - Iligh Pressure Coolant Injection (11PCI) System Start Time 1IPCI Beginning of Cycle Test (6 IIPCI.313)
, Dl!SCRIPTION: This changt. reflects a revised time for the 1IPCI system to start dunng an automatic start. The USAR previously stated that the iIPCI system controls automatically start the system and bring it to design flow rate within 25 seemis of receipt of the actuating signal This start time was changed to 28 seconds based on an evaluation pr.func i by General Electric that established the acceptability of a 28 second start time in addition to the 5 wwd instniment delay time. This was &termined to be acceptable pr:marily because the capability of the plant to meet the licensing requirements for a postulated less of Coolant Accident is not very sensitive to the actualIIPCI start time and/or flow rate. The 28 second start time will not result in violation of the licensing safety limits.
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SAFETY-ANALYSIS: De change in the acceptance enteria for the start time da:s not introduce any new failure modes or new operating conditions and does not afTect the precursors to any plant event. The CNS licensing basis LOCA analysis result is not very sensitive to the actual IIPCI start time, and the Peak Cladding Temperature (PCT) will not exceed the 2200*F limit, therefore, the consequences of a plant event previously evaluatalin the SAR are not ircreami nis change does not afTect the physical configuration of t!e plant or any plant system, and does not affect tle manner in w hich any system is operated, therefore, it does not increase tie probabihty of occurrence of a malfunction of equipment important to safety. The most limiting event analysis assumed as the single failure component is the disablement of the 1IPCI system (1IPCI or battery failure) Therefore, the increase in the 1IPCI start tin c will have no impact on this analyzed limiting case,11 was judged that the efTect of the slightly longer start time on all other non-limiting LOCA events is small enough so that they da not become the limiting case. Technical Spectfications do not s;xury a start time for the llPCI system As stated above, the capability to main %i:t the PCT below the 2200*F limit is not affected. Le 28 second start time does not result in any unacceptable safety consequences of the plant and the margin of safety defined in the basis of any Technical Specification is not rtduced.
USAR Chance Reauests (UCRs)97-094 and 97-126 TITI.E: liditorialUSAR Changes DESCRIPTION: nese UCRs made editoria! changes to the USAR,includmg the correction of typographical errors and correction of a reference.
SAFETY ANALYSIS: By definitim, editmal changes have no consequential effect on the affected sections /pages of the USAR.
Den:is no increase in the probability of occurrence or consequences of a plant event or malfunction of equipment important to safety and the margin of safety as defined in the basis for any Tecimical Specification is not reduced.
USAR Chance Reonest (UCR) 97-0% and 97-142 TITLE: USAR Change - Residual IIcat Removal (RIIR) Steam Condensing Male DESCRIPTION: These changes revised applicable sections of the USAR which referred to or described the RIIR Steam Condensing Male (SCM) of operation to clearly indicate that the RIIR SCM has been opentionally abandoned Procedures containing reference to this mode of operation were also revised accordmgly.
Operations has not used the RIIR SCM of operation in the 23-year operating history of the statiott SAFETY ANALYSIS: The RilR SCM is an operational option that is not safety related and is not relied upon or required to respond to any analyzed postulated transient or accident. General Electric determined that the deleti on of the RIIR SCM has an insignificant impact on radiological release, Potential add 2tional safety relief valve (SRV) eychng was evaluated to be within the current SRV cycles design basis, The RIIR SCM is not required to assure that the safety related functions of the RIIR system are capable of being fulfilled.
The safety basis of the RIIR system is not alTected. Low Pressure Coolant Injection and Decay lleat Removal capabilities will not be alTected. Additional or new equipment is not being installed. System components will be placed in a cmfiguration that creates a paasive Primary Containment boundary which will continue to be venfied thmugh the station containment integnty program. Eliminating the R1IR SCM as an operating mode at CNS only prechides an operatmg option w hich is manpower intensive, would result in the diversion of operator attention from more important activities, and would impact the availabihty of the RIIR system for Suppn:ssion Pool Coolmg. Operation of the RIIR SCM is not covered by any Technical Spectfications or Bases. RIIR SCM is mentioned in the bases for instrumentation to detect a break in the 1IPCI steam piping. Eliminating RIIR SCM as an operational optian will not afTect this instrumentatiert The climination of the RIIR SCM will have no etTect on any Technical Specification margin of safety.
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USAR Chanue Request (UCR)97-097 TITill: USAR Change Soap Film Testing DESCRIPTION: This UCR deleted reference to the use of soap film for performing leak rate testing. Soap film may be used as an aid to kicate leaks during testing, but is not used to perform testing. Testmg is performed in accordance with 10CFR50 Appendix J. Leak rate testing technology has improved such that scap film is not used. The preferred test methals are makeup flow rate testing or pressure decay testing. These met!xxis are nxre accurate and provide a leak rate used in calculating the total Type B and C leakage as required by the Technical Specifications.
SAFETY ANAL,YSIS. Soap film testing is not a precursor to any plant event ,lescribed in the USAR and plays no role in
- accident mitigation. This change does not alter any equipment or procedures relied upon to mitigate accidents or transients. Soap film testing does not verify the operability of plant equipment, but is performed as an aid to heate leakage. Leak rate testing verifies that components meet their design basis
- allowable leakage rates. The leakage rate tests are controlled by surveillance pnxedures and meet the requirements of Appeixhx J. This change does not alter the inspection methods required by Appendix J or ASME XI. Although leakage rate testmg is required by the Technical Specifications, this change does not affect the basis because the soap film test is not required by Appendix J or ASME XI, and the test methods used meet the requirements of tir applicable Cales and Technical Specifications.
USAR Chance Reouest (UCIO 97-098 TITiji: USAR Change Allowable Valve Seat leakage DESCRIPTION: this UCR deletes a paragraph on allowable seat leakage and revises the preceding paragraph to read that seat leakage testing is performed in accordance with 10CFR50 Appendix J. The deleted paragraph previously identified an allowable valve seat leakage of 2 cc per hour per inch of seat diameter during hydrostatic test at design pressure.1his was a requirement from the procurement specification for valves arx! it was never inteixled that this entena be applied in service. In fact this criteria is in conflict with the 10CFRSO Appendix J criteria and the ASMl! OM-10 criteria for valve seat leakages required by the Technical Specifications Seat leakage measurements are not performed during in-service hydrostatic tests.
SAFETY ANAL.YSIS: Seat leakage testing is not a precursor to any plant event described in the USAR and plays no role in accident mitigation. This change does not alter any equipment or procedures relied upon to mitigate accidents or transients. Testmg venfics that components meet their design basis allowable leakage rates.
The seat lokage tests are nonnally performed w hen the systems are out of service and are controlled by surveillance procedures. This change does not alter the inspection methods required by Appendix J or ASME XI. Although seat leakage testing is required by the Technical Specifications, this change does not alTect the basis because the 2 cc/hr limit is not required by Appendix J or ASME XI, and the tes methods used maet the requirements of the applicable Codes and Technical Specifications.
USAR Chance Reauest (UCR) 97 099 TITLh: USAk Change - Figure IV 7-1 DESCRIPTION: This USAR Figure was revisal correct a drawing enur. R1IR-PT-146 'ias only one process connection instead of the twn shown on the drawing.
SAFETY ANALYSIS: The plant configuration has not been changed and no equipment or components have been modified.
Since no physical changes hase been made, there is no increase in the probability of occurrence of a plant event or malfunction of equipment important to safety Correcting this drawing error will not affect the cmsequences of any ewnt or equipment malfunction. Since no physical changes have been made to the component or to the way the component is operated, no new types of plant events or equipment malfunctions are created Correcting this drawmg error has no etyect on the margin of safety.
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USAR Chance Reauest (UCR)97-101 TITili: USAR Change . Fire Protection Plans and Pmcedures DESCRIPT$N: his UCR redccts- aange in the CNS responsibility for fire dnlls/entiques and the training for the CNS Fire Protection program, it aho deletes the reference to the Reserve Fire Brigade as it is no longer a formally designated group at CNS.
SAFETY ANALYSIS: These changes to the USAR have no impact on the Fire Protection training critena or any equipment surveillances, testing, or any other technical aspect of the Fire Protection training progrant This UCR is not anciated with changes to equipment used for Fire Pmtection training and has no impact on how the Fire Brigade is equipped The USAR sections being revised are not related to any plant event and do not impact the mmecpences of a plant ewnt. The changes are not associated with any equipment that is omsidered important to safety. The changes do not alter the manner in which any desice operates or is controlled and do not affect the storage, failure modes, or operation of any equipment; therefore, the pmbabihty or cmsequences of a malfunction of equipment important to safety are not increased. These changes do not impact the Fire Pmtection system as discussed in the Technical Specifications. The administrative section of the Technical Specifications does discuss the minimum stalling requirements for the Fire Brigade, but there is no mention of a Reserve Fire Brigade. This change to the USAR does not impact the basis for the Fire Brigade minimum stalling, the Fire Brigade training program, or the implancuting procedures for the Fire Protection program as referenced in the Technical Specifications; therefore, the margin of safety is not reduced.
USAR Chance Reauest (UCR197-104 TITill: USAR Change - USAR Figures IV-3-3, !V-7-1, IV-8-2, Vil-4-6, VII-4-1, and X-6 1 DESCRIPTION: The subject USAR Figures were updated to reticet revised Inservice inspection (ISI) boundary classification flag locations The 11ag kications are consistent with the ISI Boundary Classification Document, CCFMD-1.
SAFETY ANALYalS: ISI is not a precursor to any accident or transient and plays no mle in accident nutigation. The ISI boundary classification flags are for inspection purposes only and do not affect the ability of any component to perform its related safety function (s). Although ISI is required by the Technical Specifications, it is not credited with providing any safety margin.
USAR Chanec Recuests (UCRM 97-10A 97-123.97-139. 97-140.97-151 07-152.97-153. 47.I 54.97-155. 97-156. and 9.hl.31 TITill: Editoriall;SAR Changes DESCRIPTION: These change requests made vanous ethtorial changes to the USAR. Included in these editorial ch.nges w ere: 1) correction of typographical errors,2) correction of dratling etTors,3) corection of zone contitmation references m drawings, 4) co:Tection of Component identification Codes on drawings to be cmsistent with plant procedures, tagging, and the Equipment Data File, and 5) clanfication of the location of a transformer.
SAFETY ANALYSIS: By defuution, editorial changes haw no consequennal eta.t on t'ae alTeeted sections /pages of the USAR.
They have no impact on any system, structure, or component (SSC) such that it would increase the probabihty ofoccurrence of an accident or transient evaluated in the SAR. Editorial USAR changes have no etreet on the abdity of any SSC to mitigate a plant accident or transient, thus there is no impact on the radiological consequences of a plant event- Editonal changes do not make changes to equipment impnint to safety that have the potential of increasing the probability of occurrence or consequences of a malfunctim ofeqtupment important to safety. Editorial changes do not create the possibility of a plant ewnt or equipment malfunction of a ditTenmt type than presiously evaluated. These changes do not result in any physical changes to the plant or affect any safety margins.
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USAR Chance Reouest (UCR) 97 107 TIRE: USAR Change Figure V 211 DESCRIPTION: his USAR Figure was revised to ack! rrw Ccmponent identification Codes to two existing Standby Oas Treatment (SOT) onfices and to correct the position of valves SGT AOV-dPCV546A/Il from normally open to closal/ auto.
SAFl!TY ANALYSIS: ne changes to this figure are to enhance the drawing and to correct a discrepancy to make the drawing -
infamation conform with existing plant pmcedures and design documents. The changes will not alTect the design function of the SGT system Offsite dose will not be alTected by these changes. There will be no equipment added or any physical or operational changes to the plant due to this revision. CNS Operating Pncalures reflect the requiral valve position information. For these reasons, this change will rad increase the pmbability of occurrence or cmsequences of a plant event or malfunction of equipment important to safety. These changes will not a!Tect the Limiting Condition for Operation and the Surveillance Requirement of the SGT system specified in the Technical Specifications. The subject restrictmg orifices and difli rential pressure control valves are not specifically addressed in the Technical Specifications. Therefore, these drawing changes will not reduce the margin of safety as defined in the basis for any Technical Specification.
USAR Chance Reauest (UCR) 97 108 TITII: USAR Change Figurc XI 5-1 DESCRIPTION: This figure was revised to remove a " Note" from the drawing which depicted the manual bypass valve (MS-V 136 and MS-V-139) position for steam trap stations #2 and #3 as being throttled. The required puition fbr t%se valves is ftdl open to ensure they do not impede the operation of the air operated bypass valws. MS-V-136/139 are nealle vahts to albw manual control of the steam trap station should the trap fais/ stick shut. Actual fic!d pwition fir these valves is full open. His drawing change only conforms the drawing documentation to actual / normal positioning of the manual bypass vahes.
SAFl!TY ANALYSIS: This is a drawing enhancement actavy. Vahe positions shown in USAR Figures are for information only and are not relied upon for safety related functions. The probability of an accident remains unchanged.
No design requirements are affected. The consequences of various accidents (steam line breaks) are unchangal because they craht the Main Steam Isolation Valves (MSIVs) and steam line flow restrictors fbr consequence mitigation. MS-V-136/139 are located outside Primary Containment and do not atTect operation of the MSIVs nor the steam line flow restrictors. MS-V 136/139 have no safety design basis function and this enhancement does not atTect the operation or design of any important to safety equipment. The dose consequences of Main Steam line break accidents remain botmding and are not affected by this activity. The functional /perfbrmance requirements of Main Steam System steam trap stations haw rux changed Steam trap stations and their bypass flow conditions are not described in the llases of the Technical Specilications. The margin of safety for steam line break mitigation and for the Main Steam System have not tven changed by this documentation enhancement.
USAR Chance Reauest (UCR197-109 TITIE: USAR Change Figure VII-4-1 Dl! SCRIPT 10N: his USAR Figure was revived to change the normal position for valves IIPCI-V-15.1IPCI-V 16, and iIPCI-V-57 fann closal to open as reflected in CNS Operating Procedure 2.2.33 A. The drawing was also revised to show 1IPCI V 109 and 1IPCI V-110 as capped-otTdrain valves in order to correct drafting ernrs and enhance the drawmg. The locanon of the 16" Iligh Pressure Coolant Injection (1IPCI) Turbine vent line was enangd fmm downstream to upstream of the turbine exhet drip leg. This corrects a discrepancy betneen existing plant drawings and makes the drawing ceform with the plant configuration.
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sal-TiTY ANAL YSIS: De changes to this drawing will not afTect the design function of the 1IPCI system and offsite dose will not be afTected There will te no equipment added or physical or operational changes in the plant due to this drawing revision CNS Operating Procedure 2.2.33A reflects the required valve position information. The changes to this drawing correct drafling errors and enhance the drawing, correct discrepancies between the drawing and other existing plant drawings, make the drawing information conform with the existing plant procedure, and make the drawing reflect the plant configuration. For these reasons, there is no increase in the probabihty of occurrence or consequences of a plant event or malfunction , f equipment important to safety; The changes to the drawing will not afTect the 1 imiting Cmhtion for Operanon and the Surveillance Requirement of the 1IPCI system specified in the Technical Specificathm Valws!!PCI V 15,11PCI-V 16,IIPCI V 109,liPCI V 110, and the 16'IIPCI Turbine vent line are not specified in the Technical Specifications. Valve iIPCI V-57 is not specifically addressed in the Technical Specificathms, but it isolates pressure switch IIPCI-PS-84 1 and this pressure switch is disetmxiin the Technical Specifications. The switch monitors the 1IPCI pump suction pressure and provides an automatic slmtofron a low pressure signal. Valve iIPCI V-57 is maintained open per the Operating Pmcedures and this drawing change does not change the physical position of the valve Therefore, these changes will not reduce the margin of safety as defined in the basis for any Technical Specification.
USAR Chance Recuest (UCR197-110 TITLit USAR Change Figure X l1-1 Dl!SCRIPTION: This drawing was revised to show the shutolTvalve numbers for DW-LG-702A/it The shutofTvalves are shown in CNS System Operating Procedure 2.2.11 A. In addition, the normal position for valve number DW V 281 was resised from open to closed.
SAFliTY ANALYSIS: This change is to enhance the drawing by showing the valve numbers that are listed in the existing plant procedure and make the drawing reficct the plant configuration. The change will not affect the design function of the Deminerahzed Water System. The ofTsite dose will not be afTected by this change. There will be no equipment added or any physical or operational changes in the plant due to this drawing change. For these reasons, it does not increase the probability of occurrence or consequences of a plant ewnt or malfunction of equipment important to safety. The shutoff valves for DW-LO 702A/B are not specified in the Technical Specifications. This change does not reduce the margin of safety as defined in the basis for any Technical Specifications-USAR Chance Request (UCR)97-111 TITLE: USAR Change - Figure IX-6-I Dl!SCRIPTION: Tlus drawing was revised to show the instrument valves for DW PI-2404 and DW-FT-2408 in order to enhance the drawing and reflect the as-built configuration. The instrument valves are shown in CNS System Operating Pmeedure 2.2.91.
SAFETY ANALYSIS: This change is to enhance the drawing by showing the valves that are listed in the existing plant pmcedure and make the drawirg reflect the plant configuration. The change will not alTect the design function of the Demineralized Water System. The otTsite dose will not be alTected by this change. There will be no equipment added or any physical or operational changes in the plant due to this drawmg change. For these reasons, it does not increase the probability of occurrence or consequences of a plant event or malfunction ofequipment important to safety. He design functwns of flow transmitter DW-FT-2408 and pnssure indicata DW-PI-2404 are not being altered by this change. This 110w transmitter and pressure indicator and their associated instrument valves are not addressed in the Technical Specificatiora
~ Therefore, this change does not reduce the margin of safety as defined in the basis for any Tecimical Specificatiota
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USAR Chance Reauest (UCR197-112 l TITI.E: USAR Change - Figure X-61 i
DESCRIPTION: This drawing was revised to delete the " throttled" designations from the following eleven Reactor i Eqmpment Cooling (REC) valves: REC 23, REC 25, REC-34. REC-35, REC-36, REC-37 REC-43, REC-200, REC-201, REC 202, and REC 203. This change reflects the as-built configuration and makes the drawing confmn with existmg plant procedures. The normal position of the valves is shown as open m CNS System Operating Procedures 2.2 65,2.2 65 A, and 2.2.68.
SAFETY ANALYSIS: This change is to enhance the drawing by showing that the normal position of REC sy stem valves is open as shown m existing plant pmcedures and to make the drawing reflect the plant configuration. The change will not alTect the design function of the REC system The offsite dose will not be alTected by this change. There will be no equipment added or any physical or operational changes in the plant due to this change For these reassis, there is no increase m the probability of occurrence or consequences of a plant event or malfmetion of equipment important to safety. The REC valves are not specified in the Technical Specifications.11us change does ent reduce the margm of safety as defmed in the basis for any Technical Specifications.
US AR Chance Reauest (UCR)97-113 TIIlli: USAR Change - Figure 13 2 DESCRIPTION. This drawing which reflects flow diagram symbols and abbreviations w as revised to add the symbol and dewnption fa htwmcc inspection (ISI) boundary Dags The added symbol is in conformance with the requirements of the ISIikxnl.uy llasis - ASME Section XI Classification liasis Document, CCFMD l.
A reference to this document was also added to the drawing.
SAFETY ANALYSIS: There will be no changes in the design or operation of the plant and no impact to any plant procedures.
1he change will not increase the radiation exposure of plant personnel or the public from a plant event or malfunction of equipment important to safety. There will be no equipment added or any physical changes to the plant. The boundary classification is for inspection purposes only and does not alTect the abihty of any component to perfomi its safety function. The ISI requirement in the Technical Specifications is per the requirements of ASMi!Section XI The drawing change conforms to these requirements and will not reduce the margin of safety for any Technical Specification.
US AR Chance Recuest (UCR)07-114 TITLE: USAR Change Figure XI-8-1 DESCRIPTION. Tlus drawing was revised to askt instrument valves for DW FT-5 to reflect the as-buih configuration of the plant. The instrument valves are shown in CNS System Operating Procedure 2.2.1 IIL SAFETY ANALYSIS: This change is to enhance the drawing by showing the vahes that are listed in the existmg plant procedure and make the drawing redect the plant configuration The change will not airect the design function of the Demineralved Water System. The otTsite dose will not be alTected by this change. There will be no equipment added or any physical or operational changes in the plant due to this change. For these reasons, there is no increase in the probabihty of occurrence or consequences of a plant event or malfunction of equipment imponant to safety. The design function of the flow transmitter is not being altertd by this change. Tran unitter DW-FT-5 and its associated instrument vakes are not addressed in the Technical Specifications. Therefore, this change does not reduce the margin of safety as defined in the basis for any Tectuucal Specifnations
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4 USnR Cliance Reauest (UCR)97-116 TITI.E: USAR Change . Seismic Instrumentation DESCRIPTION: This change added a statement to the USAR to clarify that data provided by the PAR-400 peak accelerographs may contain recordmps of non-seismh loadings, such as vibratory or shock loadings, which will be considered during evaluations of the data.
SAFETY ANAI,YSIS: No physical change to the facihty is involved with this USAR change. The change is a clarification of the data provi&x! by a scismic instrument. It does no; affect any system or equipment credited as a plant event initiat(r rur des it affect any system or equipment credited with terminating transients and whose faihire could result in a plant event. It has no effect on preventing or mitigating the consequences of a plant ewnt. This change is not associated with nor can it cause any malfunction of equipment important to safety and does not introduce any new failure modes. There are no Technical Specification margins of safety related to this change.
USAR Chance Reauest (UCR)97-117 C
Tri1E: USAR Change . Figure X-12-1 Dl!SCRIPTION: This <irawing was revised to all existing pressure indicator IA-PI-(AD-R 1 A) to depict the existing plant configuration This pressure indicator is locat.J on accumulator IA ACC-(AD R-1 A) which acts as a reservoir for air to be used in air opers. tor PC-AO-(AD-R-1 A) This air operator is used to position damper PC-AD (AD R-I A) the "Drywell and Torus Vent Outlet Damper." The damper fails closed on has of air or power. This revision resulted in no change in the design or operation of any components in the plant.
SAFETY ANALYSIS: The pressure indicator is a passive component and it does not provide a signal to indicate, control, or terminate the safety related operation of any system or component. PC-AD-(AD-R-1 A)is not used to mitigate the consequences of any plant event. Failure of the pressure indicator will not impact the operation of any safety related equipment. This change does not install any new components, the pressure indicator is existing. The nahtion of this indicator to the drawing does not impact the ope.ation of the damper or the Instrument Air System. He indicator is connected to a nonessential comp ment that is not relial upm to perform any safety related functions. This change does not alter the failure modes of any equipment. The operation ef the "Drywell and Torus Vent Outlet Damper" will not be impacted by this change. The addition of this pressure indicator to the drawing does not impact any Technical Specification related comp >nents. The indicator and its associated accumulator and damper are not delineated in the Technical Specifications or defined in the basis for any Technical Specification.
Therefore, this change dies not reduce the margin of safety as defined in the basis for any Technical Specification USAR Chance Reggest (UCR)97-118 TITI.E: USAR Change Figure XI 5-1 DESCRIPTION: This drawing was revised to show existing isolaGon, shutoff, and drain valves for instruments MS-PS.103 A, B, C, and D and involved no modification to existing plant equipment. These pressure switches are usal to imtiate a Group i Isolation signal based on decreasing Main Condenser vacuum. The subicct vahrs were documented as extsting components on Design Change (DC) 84-20 I which modified the process hnes to the pressure switches. The pressure switches are functionally tested once per month and calibrated once every three months per Technical Specification requirements.
SAFETY ANA1.YSIS: The valws that are being added to the drawmg are existing valves that are currently listed in Operations valve hneup pmcedures. The requiral puitions for these vah es are not being changed and are controlled by existing Operations procedures. No physisal work will take place in the plant as a result of this update. This activity does not change the existing plant configuration and has no etTect on system 121
performance. The function of MS-PS-103A, B, C, and D is not altered. The abihty of these pressure switches to help mitigate the consequences of a plant event is not affected by this aethity. DC 84-201, Vacuum / Pressure Sensing 1.ines Reroute, was tad as a reference to verify the previous documented uses of the existing vahrs. His activity has no direct or indirect cfTects on important to safety equipment since it does not alter the manner in which any desice operates or is controlled. The subject valves will not be required to be operated any differently than they are in the current Operations procedures. No new failure nxdes to plant equipment or systems are introduced Pressure switches MS-PS 103A, B, C, and D are identified in the Technical Specifications; however, these components are not being modified and their -
functional capabihty is not alteral Therefore, this activity does not reduce the margin of safety as defined in the basis of any Technical Specification USAR Change Reauest (UCR)97-120 TITLE. USAR Change Figure X 8 6 DESCRIPTION: This drawmg was revised to change the position of SW-AOV-854AV and SW-AOV 855AV from rmrmally open to nmnally ckmt Per the Insenice Testing (IST) Program Basis Document, thes;, valves are normally closed and perform a passive safety function in the closed position only. The normally closed valve pmition is supported by CNS Opertting Procedures 2.5.1.6, 2.5.2 3, and 4.7.4. The drawings d> not control plant configuration with regard to valve position; the Operating Procedurn are the omtrollmg acuments for the position of vahrs. It is necessary to change the drawing for clarification since these valves are part of the ASME Section XI Class 3 boundary. This change also reqsed a revision to the Inservice Inspection (ISI) Boundary Basis Document, CCFMD-1, to bring it into agreement with IST Program Basis Document.
SAFETY ANAL.YSIS %c kration of the Senice Water (SW) System ASME Section XI Class 3 boundaries between essential and nonessential portions of the system are not impacted by this change. Changing the position of valves SW AOV-854AV and SW-AOV-855AV as indicated m the drawing will not impact the CNS 0;wrating Procedures or the IST Program Basis Document. CNS procedures control these valves as normally chased Vahrs SW-AOV-854AV and SW-AOV-85SAV an: used to provide a retum padi to the Reactor Equipment Cooling ileat Exchanger to connect the roof drain system and the SW Radiation Monitoring
, acturn lines to the SW discharge header. This discharge path is monitored for radiation by the SW Elliuent Radiation Monitor. These valves are normally closed la isolate the flow path not used. The monitor is nonessential and its function is not impacted by the changes to the drawing or the ISI Boundary Basis Document. These changes do not alter the fluid flow in any system nor create any new fluid or air flow paths. No new equipment is being added to the plant and no existing piant equipruent is being mo&fied Changes associated with this activity do not impact the operation of the SW or Residual Ileat Removal SW systems as discussed in the Technical Specifications. The liquid efIluent radiological monitoring of the SW discharge to the discharge canal is discussed in the Technical Specifications; lowewr, the changes altressal by this activity will not impact the effluent monitoring of the SW system.
Therefore, there is no reduction in the margin of safety as defined in the basis of any Technical Specification.
USAR Chance Reouest (UCR) 97122 TITI.E: USAR Change Main Steam Sample 1.mes DESCRIPTION: The USAR specified that there are four Main Steam sample 'ocations to monitor moisture carryover.
Ilowever, Mmor Design Change 80-086, Amendment 2 (completed June 1981) capped two of the sample hnes. Chemistry does not use the sample lines to take steam samples during routine plant o;wration.- In the event that steam samples were reqmred, the two remaining sample lines would be adequate to obtain required sampics. Engineering Evaluation 97-162 documented the acceptability of the as-installed condition. This UCR subsequently revised the USAR to reflect that there are two Main Steam sample l hwations.
SAFETY ANALYSIS: De renoval of two Main Steam sample points does not contnbute to the occurrence of a plant event or atreet the consequences of such an event. Steam samples can still be obtained using the remaining two
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sample lines. De design parameters of the sample lines are still satisfied. Removal of the two sample lines does not affect the operation or performance of equipment important to safety. There is no Technical Specificatim requirement for Main Steam hne samphng; therefore, this actisity does not reduce the margin of safety as dermed in the Technical Specifications.
USAR Cbange Reauest (UCR)97-132 1TILIk USAR Change Figure XI-6-1 DESCRIPTION: This drawing was revised to add the electrical circuit for pressure switen SW-PS-365B, to change the rumal position for valves SW-V-30 and SW V 31 from " throttled or open" to " throttled open", and to show the correct location of the Insenice Inspection (ISI) boundary flag for SW-MOV 37MV.
SAFETY ANALYSIS: De chance to add the electrical circuit for SW-PS-365B is to show the electrical circuit information that was previously left olithe drawing. ne clatricalcircuit information for SW-PS-365A is already shown on the drawing and this change to add the same information for SW-PS-3658 will make the drawing -
consistent. The electrical circuit information fa the pressure switch is on the General Electric Elementary Diagram for the 4160V Switchgear. Showing the clatrical circuit information enhances the drawing and will not affcet the design functim of the pressure switch. The change in the normal position of SW V40 and SW V 31 makes the drawing information consistent with SW V-23 and SW-V-24 which perform similar functions (Seal Water Supply for Senice Water [SW) Pumps A, B, C, and D). The normal positons of the valves were verified in CNS Operating Proc"ure 2.2.71 A to be " throttled open." The change in kication of the ISI boundary flag for SW-MOV-37MV corrects a drafting enor. The flag was erroneously located before the valve instead of after the valve. The location of the ISI boundary flag is defined in the CNS Inservice Testing Program Basis Document. The boundary classification is for inspection purposes only and will not alTect the ability of the valve to perfcrm its safety function. There will be no equipment added and no physical or operational changes in the plant as a result of these drawing changes. The otTsite dose, normal or accident, will not be affected by these changes. This activity will not increase the probability of occurrence or consequences of a plant event or malfunction ofequipment important to safety. The design function of the SW System is not beirig altered. The SW pressure switch, the Seal Water Supply Valves for the SW Pumps, and the ISI boundary location for SW-MOV-37MV are not specifically addressed in the Technical Specifications. These changes will not reduce the margin of safety as defimed in the basis for any Technical Specification USAR Change Reauest (UCR) 97 133 TITIE USAR Change - Figure XI-6-1 DESCRIPTION: This drawing was revised to atkl a note to indicate that the Service Water (SW) Booster Pump minimum flow valves, SW-MOV-122MV through 125MV, are electrically disab!cd in the closed position. This incorporated changes made to the SW System per the following modifications Minor Design Change (MDC)76-107, Relocation of SW Thmttle Valves for Reactor Building Closed Cooling Water lleat Exchangers, modified the minimum flow lines and justified that the minimum flow valves could remain ckwed withoutjeoiwdmng SW Ikioster Pump operation. Design Change (DC) 88-222D, Panel M and Miscellaneous Ihunan Engineering Ikficiencies, removed closing switches and motor control center starters for the motor operated valves (MOVs) The subject drawing was not revised to reflect the mothfications.
SAFETY ANALYSIS: The change to electrically disable the SW Booster Pump minimum flow valves was evaluated with the mothfication of the SW System per MDC 76-107 and DC 88-222D. He change will not alTect the design function of the SW System. OtTsite dose will not be afTected by this revision There will be no equipment added and no physical or operational changes in the plant due to this drawmg resision.
SW-MOV-122MV through 125MV are in the Safe Shutdown Equipment List with remarks stating that the vahrs haw been electrically disconnected. Design Criteria Document, DCD-3, describes the function of the vahrs and states that they an: clectrically disabled. This is consistent with the modifications made per MIX'76-107 and DC 88-222D. This drawing change tkrs not increase the probability of occurrence
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or comcquenas of a plant event or malfunctxm ofequipment important to safety. He SW Booster Pump mimmum flow vahrs are not specifically addressxlin the Technical Specifications. Dis drawing ch.. ige will not reduce the margin of safety as defmed in the basis for any Technical Specification.
l]SAR Change Reguest (IJCR) 97 134 TITill: USAR Change - Figure X 10-4 DESCRIPTION: This drawing was revised to add motor operated actuators for dampers IIV AD-AD1544 through IIV-AD AD1547. His incorporated changes made to the Control Room Ventilation System (CRVS) per Design Change (DC)91-031, which installed die damper motor operators Note 6 was also added to the drawing to state that "TC-1036, REL-1036D, SS-1036 REL-AC-C-1 A, EP-1035A, and EP-1035D devices serve no control function." his incorporated changes made to the CRVS per DC 86-060 and DC 89 215, which deleted the control functions of these devices. The subject drawing was rat revised to reflect the modifications.
SAFETY ANALYSIS: The change to show that the identified devices serve no control function and the change to install the damper motor operators were evaluated with the modification of the CRVS per DC 86-060, DC 89-215, and DC 91-031. The changes to the drawing will not affect the design function of the CRVS. OITaite dose, normal or accident, will not be afTected There will be no equipment added and no physical or operatimal changes in the plant due to these drawing changes. Design Criteria Document, DCD-10, is consistent with the nxxhfications made to the CRVS per DC 86-060, DC 89 215, and DC 91 031. These changes will not increase the probability of occurrence or consequences of a plant event or malfunction ofeqtupment important to safety. The CRVS devices and dampers alTected by this drawing change are not specifically addressed in the Technical Specifications. The margir of safety as defined in the basis for any Technical Specification will not be reduced.
IJS AR Chance Reauest (I!CR)97-135 TITLE: IISAR Change Figure VII-4-5 DESCRIPTION: This drawing was revised to: 1) revise the description of the Core Spray (CS) pump control switch to change the third position of the switch fmm "open" to " start" for consistency,2) revise CS pump control switch number from 55A to SSA to correct a drafting error, and 3) provide missing information in the drawmg by adding panel number 9-3 to the Check Valve Functional Control Diagram block to identify the Control Room panel where the indicator light for CS-CV 18CV is located.
SAFETY ANALYSIS: This revision corrects tiu drafhnvrrors in the description of the CS pump control switch position and control switch number and provides missmg information in the drawing. The changes to this drawing were verified against the as-built configuration of the General Electric Elementary Diagrams and CNS Pnmlures. These drawing changes will not afTect the design function of the CS Pumps and CS Check Valve CS-CV-18CV. The ofTsite dose, normal or ace ident, will not be afTected by taese changes There will be no equipment added and no physical or operational changes in the plant due to this revision These changes will not increase the probability of occurrence or consequences of a plant event or malfunction ofequipment important to safety. The CS Pump control switch and the CS check valve are not specifically addressed in the Technical .%ecifications. These drawing changes will not reduce the margin of safety as de'ined in the basis for any Technical SpeciFcation USAR Chance R-auest (1JCR)97-137 TlILE: USAR Change Figure VII-4-2C l
j . DESCRIPTION: This drawing was revised to: 1) change the drawing designation for the color of the liigh Pressure l Coolant injection (IIPCI) imtiation signal and reactor vessel high water level rignal seal-in panel j indicatmg hghts fnun"W"(wtute) to "A"(amber),2) change the call-out for the two position turbine test i
selector switch fnun " Auto-Test" to " Normal" " Turbine Test", and 3) add panel designation 9-3 to the
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' *Cmtrol Switch in Stop Position" block of the Functional Control Diagram to identify the panel location of the three positim switch fit the Gland Seal Condenser Condensate Pump. The changes are to correct
~
drafting crTors, and to enhance the drawing by providing missing infonnation and making the drawing descriptions consistent throughout the entire drawing. .
-SAFETY-ANALYSIS: The changes to the General Electric (GE) Functional Control Diagram for the !!PCI System are in accordance with the as-built configuration of the GE Elementary Diagram for the IIPCI System. The - ,
drawing changes will not alTect the design function of the IIPCI System. There will be no equipment added and no physical or operational changes in the plant due to the changes to this drawing. Plant snuxtures are not impacted by this resision. Offsite dose, normal or accident, will not be afTected. The
, revision of this drawing will not increase the probability ofoccurrence or consequences of a plant event
~
or malfunction of equipment imputant to safety. - IIPCI testing as described in the Technical Specifications will not be impacted. Reactor overfill protection as discussed in the Tecimical Specifications is also not impacted. These drawing changes will not reduce the margin of safety as defined in the basis for any Technical Specification if SAR Chance Reauest (ItCR)97-141 TITLE: USAR Change Figure VIII-4-1 DESCRIPTION: 'this drawing was revised to change the sy stem designator for loads connected to Motor Control Center (MCC) compartments MCC A(I A), MCC A(1B), MCC-F(5B), and MCC-F(SC) from "AS" to "MS" and to add the Component identification Code numbers for the motor operators. These MCC compartments supply power to motor operated valves (MOVs) that are associated with the pressure control vahrs used to provide cylinder heating for the main turbine. The valves are Main Steam valves and as such should have "MS" designa+ ors. The valve motors, motor operators, actuators, and MCC compartments are currently tagged with the "MS" system designator and the Equipment Data File
- currently lists the motor operators for these valves with "MS" system designators.
SAFETY 4
ANALYSIS: These MCC loads and changes to the USAR are not associated with any initiators of plant events discussed in the SAR. No physical or operational changes are required to support this change. This drawing change is associated with motor operated valves used fbr the main turbine cylinder heating system. These MOVs are not used to mitigate the consequeaces of any plant event and are not associated with any radiation monitoring equipmer. Therefore, this change does not increase the probability of occunmce er consequences of a plant event. Failure of these valves will not impact the operation of any safety related equipment. These MOVs are located in the Turbine Building and are not kicated near any equipment that is ccasidered imputant to safety. The MOVs do not proside a signa! to indicate, control, or terminate safety related operation of a system or camp nent. The changes to the USAR drawing do not alter the mmmer in which any desice operates or is controlled. These changes do not install any new cunpanera or change any existing cxxnponents and do not impact the operation, fadure modes, or control of any equipment. Therefore, there is no increase in the pmbability of occurrence or consequences of a malfunction ofequipment impxtant to safety. These changes do not impact any Technical Specification rela'ed components The turbine cylmder heating system, the MOVs, and their associated pressure control vahrs, are not dehncated in the Technical Specitications or defined in 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 The following 1.CRs made changes to the USAR in 1995, but were inadvertently not reported at that time:
]ssme Chance Recuest (I CR194-0064
. TITLE: USAR Change - Range Change cf Nuclear ik>iler Instrumentation (NBI) Pressure Switches DESCRIPTION: Tras LCR updated the USA R tables for the Core Spray arrt Low Pressure Coohmt Injection Instrument Specifications to revise the range of the reactor vesel preasure permissive and reactor vessel low pressure permissive to 100-500 psig. This change was associated with Design Change (DC)87-043,
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Replacement of N!!! Picssure Switcles, which replaced 13arksdale switches NBI PS-52A,C with Static-0-Ring switches NBI PS 52Al, A2, C1, and C2.> The range change was inadvertantly ornitted from the safety evaluation of the DC; therefore, a separate safety evaluation was performed to address the range change of the pressure switches.
SAFETY ANALYSIS: The replacement of the subject switches has no efTect on the performetse of their safety fimetions and improves ti-ir accuracy. The probabihty of occurrence of an accident or malfunction of equipment important to safety will not ircrease since the functions of the switches are tmchanged and no new switch functions are added. The improved accuracy of the new switches improves the margin of safety.
Liceme Chance Reauest d.CR) 94-0079 TITLE: USAR ChanFa - Diesel Generators DESCRIPTION: This !.CR made various revisions to the USAR to clarify Diesel Generator requirements in Section Vill-5. These changes ircludal clarification of starting air requirements, engine overspeed limit, engine start times, and corxhtions in Table Vill-5 4, Abnormal Diesel Generator Conditions and Control Room Indication. The changes are based on engine design and operating experience.
SAFETY ANALYSIS: The subject changes are based on design and experience, with no physical or operational changes made.
Therefore, this LCR dxs not increase the probability of occurrence of an accident of malfunction of equipment important to safety and does not reduce the margin of safety as dermed in the basis of any Technical Specification.
I icense Chance Reauest d CR) 94-0087 TITIR: USAR Change Description of Control Room IIVAC DESCRIPTION: This 1.CR amended the USAR to include the Control Rmm habitability requirements in the event of a liigh linergy Line Break (HElB), in particular a Main Steam Line Break (MSLB), in the Turbine Buildmg The basis for this USAR revision is the analysis contained in Amendment 25 to the FSAR, along with the appropnate updatmg by calculations NEDC 94 118 and NEDC 94 157 to accout for the present design of the Contml Rmm IIVAC system.
SAFETY ANALYSIS: This LCR does not increase the probabihty of an accident or malfunction of equipment important to safety smcc no physical changes to the plant are favoh ed and no hardware or pmcedure changes are associated with this change. The increase in the Control Rmm temperature as a result of a !! ELD in the Turbine Building presents no safety hazard to the Control Room Operators or equipment in the Control Room.
The radioactive dose to the operators as a result of this event is bounded by that of the Loss of Coolant Accident. The only active components which were assumed in the safety analysis to have an active function were the Main Steam isolation Vah :s and no change was made tc these components. The only Technical Specification Bases which relates to this LCR is Specification 3.12.A, Control Room Emergency Filter System (CREFS), llowever, no credit was assumed in the subject safety analysis for this system so there has been no reduction in the margin of safety as defined in the basis for any Technical Specification. The CREFS will be capable of fulfilling the requirements of Technical Specification 3 12.A aller the event, including the positive pressure requirement.
1 icense Chance Reauest d CR) 44 0MJ TITLE: tBAR Change Fuel Pool Level Switches Dt!SCRIPTION: This LCR removed the requirement to test fuel pool level switches, FPC-LS-60A/B, once each three months Ibr operability. It has been determmed that the three month calibration frequency is excessive ar.d has no basis Based on established trendmg data, a minimum yearly calibration has been identified for the subject switches.
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SAlli'lY ANALYSIS: The subject switches are not part of any accident initiator and are in a non-safety related system. They will std! be tested, only at a frequency of once a year as established by Procedure 0 38, " Process Instrument Calibration Program." This change does not modify or otherwise permanently affect any safety related plant system; therefore, it cannot impact previously evaluated accidents. Changing the frequency to once a year and removing the reference from the USAR (ioes not impact any system operation, therefwe,it does not increase the probability or consequences of equipment malfunction. The change in calibration frequency does not impact the manner in which the switches operate. The margin of safety as dermed in the Technical Specifications is not reduced by this change.
License Chance Reouest d CR) 92 02 TITI.E: USAR Change - Steam Jet Air Ejectors (SJAEs)
DESCRIPTION: his LCR removed the statement " normal operation involves only one SJAE working at one time" from Section XI of the USAR. This is not correct, as normal operation is with both units, each with one first stage and one smn! stage element wurking to evacuate both shells of the condenser, Specifically stating the normal operation of the SJAEs does not add value to the USAR since SJAEs are designul and analyzed to operate at full capacity in either state.
SAFETY ANALYSIS: No new states ofoperation or new/ditTerent equipment are intnxluced into the system. Furthermore, all states of operation of the SJAE have been analyzed. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety are not increased and the margin of safety as defined in the Technical Specifications is not reduced.
License Chance Reauest d CR) 95-0002 TITLE: USAR Change Reference Leg Injection DESCRIPTION: his LCR revised Section VI of the USAR to provide for valves CS.V.147A/B to be closed during plant shutdown and remain closed until the plant is started up again and the reactor vessel pressure reaches appmximately 310 psig. This valve sequencing change will isolate the Nuclear Boiler Instrumentation (NBI) system from the Core Spray (CS) system when reference leg injection is not required, but will retain the fully functional remote manual operational commitment to the NRC. This change will prevent solenoid vahr wear caused by cyclic pressurization which occurs during CS system surveillance testing when the reactor is depressurized, thereby improving the reliability of the system. In the past, leakage of air through these solenoid operated valves has resulted in reactor water level indication fluctuations.
SAFETY ANALYSIS: ne CS/NBl system interface will be unchanged during any time when the reference leg injection system may be required to function. The system configuration will be unchanged during normal operation at which all design basis accidents presiously evaluated were assumed. Failure of the reference leg injection subsystem would not degrade the CS system Any single failure ofisolation valves CS-V-147A or CS.V 1478 would affect only me train of reactor water level indication. Any failure of components due to this vahr sequencing change is envekiped by existing accident evaluations with loss of one train of CS and one tnun of reactor water level indication. This CS valve sequencing change will not increase vessel level errors. Induced level errors are within Technical Specification limits.
l_icense Chance Reonest d CR) 95-0012 TITLE: USAR Change - Re-alignment cf Reactor Core Isolation Cooling (RCIC), liigh Pressure Coolant injection (IIPCI), Core Spray (CS), and Residual iIcat Removal (R1IR)
DESCRIPTION. This LCR revised the USAR to clarify that while RCIC,llPCI, CS, and Ri!R may re-align from the design flow functional test mode to the operating mode automatically, these systems must be imtially aligned in the standby status for the systems to fulfill their safety design bases. This change brtngs the USAR into agreement with the design specifications for CNS.
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SAIETY ANALYSIS: These USAR changes document the actual design basis for RCIC, !!PCI, IPCI, and CS. . When these systems are not in their standby sta;us, they are not designed to re-ahgn in time frames short enough to meet the injection times assumed in the CNS transient and accident analysis. Du6 to the short duration of these tests, there is no c1Tect on system reliability or degradation of plant safety. No change in actual system operation is involved and, therefore, there is no increase in the probability or consequences of a malfunction ofequipment imputant to safety. These USAR changes reduce the possibility that the plant design will be misinterpreted or misunderstood. The change does not affcet plat safety Neither the design of systems _ to re-align from ihnv functional test mode to the operating mode nor the required system injection times are included in the Technical Specification baser.
I.icenw change Recuest d CR) 95-0014 TITLil: USAR ChanFe . Senice Water (SW) System Description DiiSCRIPTION. This LCR reviel the SW system desenption to include the rmessential header isolation pressure setpoint and indicate the proper number of SW pumps under normal operating and shutdown conditions. This change brings the USAR into agreement with the Technical Specification Hases and Operating Procedure 2.2.71.
SAFliTY AMALYSIS: The SW system is designed to perform its required design basis accident safety function with one ognatmg SW pump degraded to a performance level equivalent to 6000 gpm at 125' TDil. An increase in the number of operating SW pumps in this emdnion will only serve to increase the available safety margin Per the Design Criteria Document for the SW and RIIRSW system, the SW system is designed, evaluatal, and found to be acceptable to be operated during planned operation with one to four operating SW pumps.' The subject change does not incresse the probability of occurrence or consequences of an accident or malfunction of equipment important to safety Calculation NEDC 92-050Ali has evaluated the adeq 2acy of the existing setpoints for the SW system non :ritical kop isolation on low pressure and found them to be satisfactory for lhe most limiting design basis condition.
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REPORTABLE PROCEDURE CIIANGE REOUESTS Procedure Char ce Reauest (PCro 01 (Revision 15)
TITI.E: Intmduction to CNS Operations Manual DESCRIPTION: This rnision added guidance ihr the use of annunciator /abnonnaVemergency pnx:edurn 7his chanF0 added flexibility for personnel to exit oft normal condition procedures in order to prevent unnecessary plant transients (shutdown / scram) from being performed when the condition requiring entry into the specific procedure no kmger exists. Event based procedures aie written with mitigating actions first uhich may resolve the corxhtion and a plant transient may not be required should the initial actions resolve the caxhtiort When exiting these procedures, Operations uses plant knowledge, experience, and normal system operating procedurr and checklists to return the plant to a normal configuration. In addition to this change, numerous acministrative enhancements were ir,corporated into this pmcedure.
SAFETY ANALYSIS: This revision does not alter any plant system or component described in the SAR. Allowing Operations ficxibility when using event based pacedures does not increase the probability of a plant event since tlx se pncedures are used after an event has occurre l. The consequences of a plant event or equipment malfunction are rut increased since the symptom based Emergency Operating Procedures (EOPs) dictate controlhng key plant parameters to hnut the canax]uences of a plant event or equipment malfunction. The incicased Operations flexibility is bour ded by the single Operator enor of not recognizing the event based pacedure is applicable and taking the actions specified by the procedure to rectify the situation.
There is no reduction in the margin of safety of any Technical Specification as a result of this increased Ikxibility.
Procedure Chance Rc3 cst (PCR) 0 4 (Revisica 25)
TITI.E: Procedure Change Pncess DESCRIPTION Vanous enhancements were nale to the pmcedure change process, including the following: 1) addition of guidance to accannxxiate documents uher than those requiring Station Operations Resiew Committee (SORC) appmval and other than the CNS Operation Manual, 2) addition of references to special procedures, special instructions, and vendor pnvedures which will be rnised and approved per this procedure,3) adthten of guidance for assignment of approval authority,4) clarification of placing pmecdures on "admimstrative hold",5) improvements to Verification & Validation process,6) addition of new pmcedures and deletions to list of " intent chtnge" cnteria, and 7) carious enhancements / clarifications based on user feedback.
SAFETY ANALYSIS: Tlcse are administrative changes that do not alter the physical plant or operating procedures. Suitable controls remain to ensure adequate reviews will be obtained for procedures that could afTect the probabih+y of a plant event previously evahiated in the SAR. Adequate contmls remain in place to ensure pmcedure changes with the abihty to alTect the consequences of a plant event, increase the probability ofoccurrence or consequences of a malfunction of equipment important to safety, or alTect the margin of safety as defined in the basis ior any Technical Specification, will still obtain prior SORC review.
Procedure Ch mee Reauest (PCR10 71 (Revision 6)
TITLE: Control of Combustibles DESCRIPTION: 'lhis pmcedure was extensively rnised and rearranged to make it more user friendly and understandable.
Sigmficant changes in this revision included.1) exclusion of NFPA-30 compliant areas and inclusion of roof tops, 2) exclusion of ordmary combustibles inside ekwed metal closures, 3) clarification of requirements for fire triardant lumber, 4) ad<htion of requirement to evaluate existing combustible loads during work planning, and 5) rnision of fire k ading fbr the Multi Purpose Facihty.
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SAFETY
- ANALYSIS: lletter adnunistratm cmtrol of combustibles reduces the probability of fire (scurrence by controlling fire
- sources and reduces the consequer.ces of a fire by controlling fire severity. The probability or consequences of fire induced equipment malfunctions .:re unchanged. The Safe Shutdown Analysis assumes all equipment in a fire amie is dtsabled by a fire and CNS is capable of achieving shutdown goals with equipment outside the fire zone. Administrative control of combustibles does ret form the basis for any Technical Specification.
Procedure Chance Reouest (PCR) 0 7.1 (Revision 81 TITII: Control of Combustibles DESCRIPTION: His pncedure was rewritten to create a Combustible Control Pncedure that is cacy to use, understand, and implement, while still pnwiding the required level af control. The rewrite of the guidelines for controlling transient combustible materials has been incorporated, e well as the dermition of various terms used within the pncedure.
SAFETY ANALYSIS-. This actisity will not change the state or function of any safety related syuems, structures, or ca.ngonents arx! will not aher any of the inputs or assumptions Ibr previously evaluated accidents. This PCR does not change the function, peribrmance, or integrity of any boundaries with which safety related systems fo.m or support the primary protective barriers on which the consequences of an accident are based. This change does not alter the design basis for the fire suppression systems nor does it adversely afTect initiating sequences or starting setpoints of safety related or fire protretion systems. This activity will ensurc ichability of an admirustrative pmcedure to control the amount of transient combu3iible materials entering gxnver bkxi buildings The ansequences of the Fire Suppression EITects Analysis arc bounding arxl unchanged by this activity. His change does not atTect equipment important to safety and creates no new failure modes The margin of safety which is established by the design and performance of safety related systems will not be reduced by this activity.
Procedure Chance Rcouest (PCR10 8 (Revision 0)
TITLE: Safety Assessments and Unreviewed Safety Question Detenninations DESCRIPTION: Previously existmg Procedure 3.3 and other pmcedures involving 10CFR50.59 were combined into a single new procedure that implements 10CFR50.59 et CNS. In addition, this procedure provided three separate elements of analyzing changes (i c., Safety Analysis, Applicabihty Screen, and Unreviewed Safety Question Evaluation), provided clearer and more conservative questions in determining MK'FR50.59 applicability, and expanded the scope of 10CFR50 59 analyses to include special events.
SAFETY ANALYSIS: The proposed changes atrect implementation of 10Ci R50.59 at CNS. Any subsequent changes to the facihty are regulated by the Code and the changes to the alTected pmcedures provide improved gme::e for implementati<n This activity is administrative in natur c Changes implemented through the revised process will be evaluated under clearer and more conservative guidance. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety are not increased and the margm of safety is not reduced.
Procedure Chance Reouest (PCR1016 (Revision 17)
Procedure Chance Recuest (PCR16 FP.604 (Revision 1)
TrTIE: Control of Doors (016)
Fire Door Annual Examination (6 FP 604)
DESCRIPTION: Pruecdtae 6.FP.604 was resised to delete thxximg dimensional criteria and related references. Flooding entena is being deleted because it does not alTect the operabihty of the fire banier fire door. UL tested door gap cnteria is being used to establish operabihty of the fire barrier only. Procedure 0. I 6 was revised to be consistent with 6.FP 604.
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SAIHTY
, ANALYSIS: The subject tes' Joes rot increase the probability of occurrence of a fire or ficod. It ensures operability oferedited fin, barrict door Flood enteria/ analysis of consequences are unaffected as suflicient margin exists in flanhng calculatims fr small dunensional afferences under diors Small dimensional changes in door Faps nll nc increcse the probability of flood occurrence which would lead to equipment malfumtims. Perfirmance of the surveillance test does not introduce any new failure modes or accident precursors. Fire banier rat.ngs and operability are maintained. There is no flooding margin of safety impacted by this change.
Procedure Chance Reauest (PCR10 22 (Rtvision 9)
TITLE: Emergency Operatmg Proc edure (EOP) Maintenance Program DESCRIPTION: his procedure was revised to specifj usis.g Pmcedure 0.4 (Procedure Change Process) to change EOP documents currently revised using this procalure. This change is an administrative enhancement to streamline the two processes. hacedure 0.4 maintains the same level of control currently specified in this procedure Changing the frequency of cyclic reviews to periodic reviews per Procedure 0.4 reduces the frequency of res iews to every five years.
- SAFETY ANALYSIS: This is an admmistrative change to the frequency of review for pmcedures cited in the Technical Specifications. Tecimical Specification 6 2.1.A.4j and NUREO 1358 require EOP documents te be revieuul perialically per administrative control. 'Ihis change does not alTect the physical configuration or operation of any plant system, structure, or component. h d ies not change any procedure required for mitigation of an accident or equipment malfunction. This change dies not involve a reduction in the margin of safety since EOP doctrnents are reviewed on a continuous basis via Operations simulator and perfbrmance training. This change only eliminates unnecersary d,cumentation of reviews being performed during OperMor training.
Procedure Ch.mcc Reauest (PCR) 0 23 (Revision 14)
TITIR CNS Fire Protection (FP) Plan DESCRIPTION: This procedive was revised to clanfy when FP equipment is considered impaired and what should be the corresponding compensaton action (s). Where appropriate, Tecimical Specifict. tion equipment was separated fn m non-Technical Specification equipment to more clearly identify conditions ofimpairment and compensatory measures.
SAFETY ANALYSIS: This activity is designed to help improve FP capabilities and dies not impact accidents previously evaluated in the USAR. By impmving FP capabilities, the consequences of a fire could be decreued.
A fire is not an accident as desenbed in Chapter 14 of the USAR and it is an event already anal >2ed and accounted fir in the Technical Specifications and the Appendix R Analysis The changes that ar: being made do not inen:ase the probabihty ofoccurrence of a malfunction of equipment important to safety, but rather provide Ihr compensatory measures in the event of an equipment malfunctio t Compensatory measures are previously reviewed and accepted practices. This activity clarifies and repeats requirements of Technical Specifications. It does not reduce the margin of safety as defined in the basis for any Technical Specification.
Procedure Chance Reauest (PCR10 23 (Revision IM TITLE: CNS Fire Protection Plan DESCRIPTION. He changes to this pmcedure are designal to enhance the fire protection pregram by ensuring the return of fire fighting equipment to the fire / ambulance house and testoration of the power supply to the fire truck after drills and actual fire responses.
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1 SAFETY ANALYSIS: This change does not increase the probability of fire cecurrence or other special event. It is an cnhancement to ensure. maximum equipment readiness. Tius will enhr.ns e tv abihty of the fire brigade to mitigate a fire. Fire brigade equipment is not considered "important ta safety" as defined in the SAR.
The defeme in depth fire protectica program relies on muhiple layers of fire protection to mit. gate the consequences of a fire. This pmgram design compensates for equipment malfunctions. No additional equipment or feilure modes are introduced by this change. This enhancement increases the availability of equipment credited in the Fire Protection Safety Evaluation Report for License Amendment 56.
Therefore, the margin of safety is not reduced.
. Procedure Chance Recuest (PCR) 0 23 (Revision 17)
TITLE CNS Fire Protection Plan DESCRIPTION: The impairment matrix was revised to include Allowed Out-of-Senice Times (AOTs) for Technical Specification Fire Detection Systems and for Fire Pump FP-P E as is currently described by Procedure 0.26, Surveillance Program. Technical Specification hose stations were separated from non Technical Specification hose stations and compematory measures redefmed.
SAFETY ANAL,YSIS: These changes are administrative in nature and do not affect the probalaility of occurrence of a fire or other plant event, and(k> not a!Icet the ability of the plant to respond to an event. These changes stipulate the compensatory measures required when equipment is o t of senice or being tested. -No new 1 equipment failure mechanisms are introduced. Technical Sp:eification surveillance requirements are based on a presupposed periodici.y which is accounted for in the analyzed margin of safety such that APT 2 may be applied.
Pntedgre Chante Reouest (PCR) 0 26 (Resisions 23. 24. 26. 27. and 28)
TITLE Suiveillance Program DESCRIPTION: 1hese revisions made various atkhtions, deletions, and changes to Attachment 1, Surveillance Procedure Allowed out-of-Senice Tim:s (AOTs). These changes are conservative and within the bounds of the AOT program as evaluated during initialinception. New procedures added to Attachment I meet the critena of tie AOT program screenmg pmcess. Discussion was also added to include examples of AOT appheation and to clanfy how the AOT program screens proecdures for application and criteria for AOT time limits. Revision 27 abo made a temporary change to support the implementation of Plant t Temporary Wxlitication (PTM) 97 12 (see PTM summary reported separately).
SAFETY ANALYSIS: Thc testmg controls enhance the availabihty of the safety systems, while also venf>ing system operabihty, The surveillance program is not an initiator for accidents described in the USAR. I y ensuring system availabihty and operability, the consequences of accidents and equipment malfunctions evaluated in the USAR remain unchanged. Surveillances and their successful performance ensure availability of equipment. The perdrmance of the suntillance procedures and alignment for testing therefore outweigh the risk in increased consequences. The testing controls are consistent with Technical Specification and USAR provisions; therefore, no new types of accidents or equipment malfunctions are created and there is no reduction in the margin of safety as defined in the basis !br any Tecimical Specification, Procedure Chance Reauest (PCR) 0 28 (Resision OT T111R Uncontrolled Vendor Drawing Applicability Determination DESCRIPTION: This new pmcedure was devekiped to ensure that uncontrolled sendor drawings not in the Drawing Contml Pmgram that are utilized for piant maintenance, testing, and mochfications are verified to match the plant configuration prior to use.
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SAFliTY ANALYSIS: Hy ensunng that plant aunponents are mamtained in cmformance with the recommended vendor manual, equipment required for normal plant operation, accident conditions, and abnormal plant transients will function as desenbed in the USAR and othcr design basis documents. Maintaining safety related' equil, ment and components in conformance with the vendor requirements, includmg the appnpriate drawings, will ensure that they will perform their safety function as required by the safety analysis. It also crisures that malfunction of equipment is within the limitations of the plant's design basis and will climinate the possibility of a different type of malfunction By assuring the proper configuration control-of vendor information is maintained, the user pmcess will be able to maintain the plant in its analyzed design The margin of safety as dermed in the Teanical Specification bases assumes that equipment testalin accadance with the Technical Specification requirements will function as required. Using the trxwt uppnpriate vendor information for Technical Specification testing ensures that this assumption is correct.
Pmcedure Chance Reauest (PCR10 39 (Revision I l)
TITII: Fire Watches Dl!SCRIPTION: This procedure was revised to iruxuparate relevant changes made to Procedure 0.23, CNS Fire Protection Plan, to ensure emaisten:y between procedures. It also adds the Administration lluilding Mask Fit Area to the list of plant locations that are exempted from requiring a hot work perrnit for hot work processes pertbrmed within them.
SAFTITY ANALYSIS: This is an administrative pnmlure change to clanfy the fire protection hot work and impainnent tracking processes. These changes do not it. crease the probability of an accident or malfunction of equipment important to safety previously evaluated in the USAR as analyzed in the Appendix R Analysis and the Fire Ilazards Analysis 'this activity does not increase the consequences of a fire since the Safe Shutdown Analysis assumes loss of all equipment in a given fire area and demonstrates adeqcate safe shutdown capability. Compensatory measures (fire watches) are a reviewed and approved method to ensure fire umsequences are not increased 'this activity does not reduce the margin of safety as dermed in the basis for any Technical Specification because it is administrative in nature and does not impact compensatory measures as defined in Techmcal S;weifications.
Pmeedure Chance Reouest (PCR) 0 39 (Revision 12)
- TITLII: Fire Watches DESCRIPTION: Various changes were made to this pmcedure to enhance safety by ensuring better control over hot work and fire watch actmties. Areas were added to/ removed from hot work permit requirements, as appmpriate. Additional gmdance was added for fire watches in respondmg to fires and fire alarms.
SAFliTY ANALYSIS: 1hese cht.nges more etTectiwly cmtrol hot work in safety related areas of the plant, thereby reducing the probabihty of fire occurrence. The consequences of a fire and fire-induced equipment failure have been evaluated in the CNS Fire Ilazarh Analysis and Safe Shutdown Analysis Report. More etTective control of hot work reduces the pmbability of fire induced equipment malfunction. No additio 'l failure modes are intmduced by the subject changes The margin of safety for Technical Specification. dre watches has been mercased thnugh nure ngmms contml of the activity and better human factoring of the procedure.
Pmeedue Chance Reouest (PCR10.39 (Revision 13)
TITil: Fire Watenes DESCRIPTION. This pnxalum was revisal to specify the responsibilities, required equipment, and authority of the three t>1xs of fire watches utihza! at CNS, hot work, impairment continuous, and impairment patrol. Specific guidance for the use ofcameras in high radia'am and hazard areas was provided, as wc!! as the stop work authority for fire watches on hot work. The requirement ihr Operations Manager concurrence for hot I
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work in safe shutdown areas was ,emoved as it is no longer required due to resolution of Appendix R winerabilitics.
SAMITY ANALYSIS: '!he clanfication and gmdance Innided trf this pacedure change do not increase the probability of a fire occurrence and have no impact on any other accident initiators. The change does not adversely alter either fire scwrity or tic rating of fire baniers As such, the consequences of a fire are unchanged. Since the probabihty of a fire is not increased, the probabdity of an equipment malfunction is not affected, and no new failure modes are introduced. Fire induced failures have been presiously evaluated in the Safe Shuklown Analysis Reprt. 'the changes enhance the level of safety provided by fire watches dispatched as Technical Specification reeuired compensatory measures; therefore, there is no reduction in the Teclutical Specification margin of safety.
Pncedure Chance Reauest (PCR) 0 41 (Revision 0)
TITLE: Seismic I k>usekeeping DESCRIPTION: This new procedure provides controls to protect operable Clan I equipment from seismic interaction haards when bringmg temporary items into an aren (i c., equipment or materials) and/or disassembling parts ofpennanent equipment to support work activities at CNS. This pnsedure was devekred as part of the NRC requirements for the SQUG A-46 resolution to maintain Safe Shutdown Equipment List equipment.
SAFETY ANALYSIS: Class I systems, structures, and components are designed for the occurrence of an operating basis carthquake and a design basis carthquake. Implementing seismic housekeeping requirements does not increase the pninabihty ofeithe event smd ensures that the consequences of a seismic event at CNS will be limited to the design basis evskation without the additional potential for temporary items to adversely affect Class I equipment. The pacedure prevents temporary items from creating an equipment failure as a direct result of the temporary item interfering with the operable Class 1 equipment. Implementation of seismic housekeeping requirements ensures that the margins of safety defined in the Tecimical Specifications remain valid by protecting Class 1 equipment from malfunctions caused by seismic interaction with temporary items Procedure Chance Reauest (PCR10 49 (Revision 0)
TITLE; Schedule Risk Assessment DESCRIPTION: This pmcedure panides dec;sim making and sche luling guidance for removing risk significant systems fnxn senice during power operatiort The determination of risk sigmficant systems was derived from the CNS Probabilistic Safety Assesament. The intent of this procedure is to provide a controlled and methodical mechanism to remove vital components from senice to perform various work actni'ics that will improve their reliabihty and thus improve plant elliciency and safety.
SAFETY ANALYSIS: Implementation of Procedure 0.49 will not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety evaluated in the SAR because allowing maintenance to be performed when needed will enhance safety system reliability and preside a mechanism to venfy system insegnty. Atkhtsorudly, this procedure identifies high risk activitics that could potentially imtiate an event or transient and establishes administrauve guidance to minimize the potential of the initiatmg events. Allowed out-of-senice repair times are consistent with Technical Specification pnnisions Implementation of this procedure does not reduce the margm of safety as defined in the basis for any Technical Specification because Allowed Out-of Senice Times for equipment defined in the Technical Specification bases have not been chimged-134-
Procedure Chance Reauest (PCR) O S0 (Resision 01 TITII: Outap Management Program DESCRIPTION: Ris new procalure was devekiped to provide administrative consistency for outage management and to provide an interdisciplinary resiew of the guidelines which control shutdown risk at CNS.
SAFirfY ANALYSIS: This procedure provides recommendations foi system asailability duting outages which are more conservative than Technical Specification requirements; therefore, the probability of occurrence of accidents evaluated in the USAR is actually reduced With the additional safety systems available, the consequences of analyzed accidents would be less severe. Procedure 0.50 and the outage guidelines do rxt specify any abrxirmal equipment linci* , or ccnfigurations The outap guidelines in Procedure 0.50 do not modify equipment importuit to safety. The margin of safety as specified by Technical Specifications is not reduced because the pmealure actually recommends equipment availability in excess of that required by plant Technical Specifications.
Procedure Change Reouest (PCR) 1.1 (Revision 23)
Procedure Chance Reauest (PCR) 1.15 (Revision 14)
Proecdure Chance Reouest (PCR) 91 1.3 (Revision 38)
Procedure Chance Reouest (PCR) 9.1.2 4 (Revision 10)
Procedure Change Reauest (PCR) 91.3 (Resision 21)
TITIE: Station Security (1.1)
Witor/four Station Access (1.15)
Pt sonnel Dosimeter Program (9.1.1.3)
At ; Control Radiolegical(9.1.2 4)
Radu . ion Safety Standards and Limits 'R I.3)
DESCRIPTION: These PCRs implemented changes te radiation exposure monitoiing at CNS. Personnel monitonng devices will no longer be assigned to all station persmnel, but only to those persoru el assigned to work within radiologically conaulled areas (RCAs) at the station. 10CFD2" 1502 requires individual nunutmng devices for adults likely to receive >10% of the limits in 10CFR201201. An evaluation was performed which concluded that non-RCA assigned personnel will not receive dose in an amount that would require the assignment of a Thermoluminescent Dosimeter (TLD).
SAIT?TY ANALYSIS: Personel monitanng devices and the radiation monitoring program cannot initiate an accident described in the USAR This change does not alTect the design or operation of components which may initiate an accident. Personnel nxinitmng devices do not alTect the USAR assumed method of accident mitigation, This change does not allirt the RCA access of personnel u ho perform functions which may mitigate the consequences of an accident or malfunction of equipment important to safety. This change does not challenge /atIcet components which have an accident mitigation function. Personnel monitoring devices and the nahation nunitmng program cannot atTect the operability of equipment described in the USAR.
This change does not atTect personnel access to areas where equipment important to safety is hicated.
This change conforms with 10CFR20 and Technical Specification provisions for personnel monitoring.
Pnxedure Chance Renuest iPCR12.1 201 (Resision 01 TITLE: Restoration From Reactor Pressure Vessel (RPV) Refueling DESCRIPTION: This new proecdure was created to coordmate procedures and activitics involving the recovery of the reactor frtun refuelmg operatims. It also gives directions to Operations to restore the Main Steam Lines from the RPV thuled condition SAFETY ANALYSIS: Guidance has been added to this procedure to recover the Main Steam Lines from a thxxled condition following drain down of the RPV from refueling. This guidance is required because Secondary Containment must be inoperable during this time period This direction is consistent with the
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requirements of the Technical Specifications and ensures refueling floor activities are limited to those when Sewndary Conta:nment is inoperable. Installation of both fuel pool gates prior to drain down of the RPV cavity ensures the additional barrier is in place between the fuel pool and the RPV casity prior to cavity drain down. His pnxnlure has no steps which can directly or indirectly afTect the consequences of an accident. Steps are put in place to ensure activities which may damage fuel or may result in the
- design basis refueling s.ccident are suspended during the period of time Secondary Containment will be inoperable, ne procedure performs no steps which may produce an equipment malfunction import mt to safety. No procedure steps direct activities to be performed that are not within the bounds of the current accident analysis. No activities directed by this pmcedure affect the margin of safety since the pncedure enforces compliance with existing Technical Specificatica and USAR requirernents.
Prtmlure Chance Recuest (PCR) 2 2.3 (Revision 45 C3, Procedure Change Reauest (PCR) 2 2 3 A (Revision 6)
TITI.E: Circu:ating Water System (2.2.3)
Circulating Water System Component Checklist (2.2.3 A)
DESCRIPTION: Four Condenser Backwash Valve Disc Sparger Supply Valves were mistakenly assigned ,mponent identifi.ation Codes (CICs)in the Circulating Water System. These valves are actually Service Water System valves This change corrected the CICs. A change was also made to make the sparging time of the condenser backwash valves less prescriptive.
SAFETY ANALYSIS: Changing of a component's CIC will not change the probability of any plant event. The sparging time of the wndenser backwash vah es is not described in the SAR. There are no safety functions ir volved with the activities or components described in this change. The valves will remain classified as nonessential components No equipment involved with this change is mentioned in any sectiva of the Technical Specifications.
Prxedere Chance Reauest (PCR) 2 2 6 (Revision 37)
TITLE: Condensate System DESCRIPTION: A new section was added to this procedure to provide guidance for condenser draining. No procedural guidance previously existed for this evoluti,n which was accomplished in the past by gravity draining to Radw aste. "he new procedural guidance allows draining the condenser hotwells using a Condensate Pump, a portable pump (s), or draming to the Fhor Drain System. Pmeeduralizing condenser draining will: 1) save time during an outage since direction will be in place,2) provide a variety of methods for drain (kmns uhich will increase flexibility, 3) ensure water that is gravity drained does not get routed to the Eqvipment Drain System, and 4) return mor ' vater to the Condensate Storage Tank (CST) thereby limiting the amount of water that will need to be processed in Radwaste and discharged to the environment.
SAFETY ANALYSIS: The new section will only be performal when the reactor is shut down and the system is not required to be available. The only shutdown accident or transient considered in the USAR is the fuel handling -
accident. Smee the Condensate system is not an iritiator or relied upon for mitigation of a fuel handlmg acetdent, this change cannot affect the probability of occurrence or consequences of a plant ewnt evaluated in the USAR. While shutdown, the Condensate System will not be a contributor to any malfunction ofeqtupment important to safety. Should the tempcrary Condensate Makeup piping rupture, a check 5alve will be installed to prevent loss of inventory from CST A. In addition, an operator will be in attendance when CST A is aligned to temporary pump and piping, in the event of a rupture, these contingencies will ensure CST A lewl remains above the required 150,000 gallons should Low Pressure Coohint injection and/or Core Spray be ahgned to take a suction from that tank. Rupture of the temporary piping is 1,ounded by the Main Steam Line break analysis. There is no impact on assumptions, calculations, procedures, or design specifications used to determine the plant's margin of safety.
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t Procedure Change Reauest (PCR) 2 2 7 utevision 25)
TITLE: Condensate Storage arJ Transfer System -
DESCRIPTION: This procedure has been updated to provide an approved method of transferring makeup water to the Condensate Storage System from the Demineralized Water Systems This procedure will allow makeup as needed in response to im entory loss from the Condense'c System.
SAFETY ANALYSIS: The transfer of Demineralized Water to Condensate Storage is not related to the occurrence of any previously evaluated plant eventi The pmcedure has been written to be within system capability; therefwe, the probabihty of the loss of either Condensate Storage /fransfer or the Demineralized Water System is not increased. Required endensate storage volumes are not affected and any consequences 4
of the loss of Demineralized Water remain unalTected. The Demineralized Water transfer valve fails -
closed on loss of power or air system pressure which will terminate the tramfer. The cha1ge operates compments for the purposes for which they were installed in the plant. The consequen;es of a low Condensate Storage Tank level are not alTected since this procedure defines the method by which the normal makeup is performed to maintain the required volume. No equipment is added or rs moved from the fecihty. Technical Specifications address the available storage volume of the Condersate Storage Tank only. 'this change allows maintaining this lewl to compensate for normal system losses. Therefore, there is no reduction in the margin of safety as defined m the basis for any Technical Specification.
Procedure Chance Reauest @CR 2 217 (Revision 17)
Procedure Chi mge Reauest (PCR) 2 219A (Revision 3)
TITI E: Emergency Station Service Transformer (2.2.17) 480 VAC Auxiliary Power Checklist (2.2.19A)
DESCRIPTION: It was identified that the configuration of the Emergency Transformer cooling fan (s) alternate power feed was not in comphance with the requirements ofnon-class 1 E circuit separation from class 1 E circuits and represented a load that could be nummatically transferred between two safety related buses. This configuration presented a challenge to NRC Regulatory Guide 1.6 and the CNS conunitment in FSAR Question' Answer 8.9 and 8.10. In order to correct this situation, the position of the a' ternate feeder breaker was changed from "on" to " locked otr to disable the automatic transfer feature. Pre etural guidance was also added for the manual transfer of the cooling fan:. should it become ne;essary.
SAFETY ANALYSIS. The km of the Enurgency Transfinner or cooling fans is not an imtiator for any plant event. The subject procedure changes restore the plant configu ation to the 'icensing/ design bases. Lis activity increaxs the reliability of the affected safety related buses by preventing transfer of a potential faulted condition to the opposite division, thereby ensuring divisional separation. The Emergency Transformer is not assumed to operate dunng any event, therefore, its potential loss due to lack of cooling carmot impact any eqmpment important to safety. A loss of olisite power has already been evaluated and restoring phmt configuration to the design basis will ensure that assumptions made regarding independence between rethmdant power sources are apheld. The automatic transfer capability of the Emergency Transformer coohng fans is not discussed or referred to in the Technical Specifications or any ofits bases.
Procedure Chance Reauest (PCR) 2 2 20 (Revision 40 C1)
TrrLE: Standby AC Power System (Diesel Generator)
DESCRIPTION- This precedure was revised to include design basis information which defines the minimum Diesel Generator (DG) starting air receiver pressure necessary to ensure multiple DG start capabihty.
SAFETY ANALYSIS- This change only pro ides design data to the Operator in the discussion section oithe procedure. Steps imuhing actions or acceptance entena haw not been altered. The USAR states that one DG air receiver is" capable ofpmvidmg sullicient air to support multiple starts without immediate replenishment? This pnmlure change prevides a minimum pressure of a 200 psips basis information. Neither equipment l
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f
operability nor function will be alTected by this change. Therefore, malfunction of equipment is not involved in this change. 'Diis PCR adds information that allows the Operator to readily judge that the DO Startmg Air is capable of multiple starts of the associated DO. These requirements are more restrictive than the current Tecimical Specifications. Any change in the margin of safety is an increase.
Pnxedure Chance Request (PCR) 2 2 21 A (Revision 11 T!Tili: Station 1 ighting System Power Checklist DIISCRIPTION. This procedure was revised to reibt changes made to component imeups.
SAFliTY ANALYSIS: Changes made have been validated and ventied to be within the design and plant configuration. These changes do not afTect any of the initial conditions described for evaluated accidents nor do they interact with components which may result in increased accident probability. The design ensures component availabihty as necdal to mitigate the consequences of an accident. Breaker positions changed or deleted by this procedarnk not interact with equipment important to safety. The breakers added by this change have been positioned to supply the respective components within plant design and have been made to ensure any misinterpretation of that component lineup ia orevented. These changes are not precursors to any accidents as the changes are bounded by the elecwal design analysis as verified. This change is not related to any Technical Specification criteria or setpoint data.
Proecdure Chance Reauest (PCR) 2 2 28 I (Revision 15)
TITill: Feedwater System Operution DESCRIPTION: This revision added a section providing procedural guidance for isolating Reactor Feedpump (RFP) mimmum flow linea. This is a change in the way the RFP is operatvi durmg plant conditions requiring minimum flow.
SAFETY ANALYSIS: The events associated with feedwater are concemed with loss of feedwater heating, loss of feedwater, or an increase in feedwater. Isolating the minimum flow line does not aher the initiators of the events or change the possibihty of any of these feedwater events While the feedwater is the normal means of makeup, the Iligh Pressure Coolant injection and Reactor Core Isolation Cooling systems are designed to pnwide the makeup during an event. If for some reason both feed pumps were needed, the minimum flow knes would not be needed. Therefon; there will be no increase in the consequences of a plant event.
Under any transient or accident for which fm!w ater flow is preferred, the feedwater pumps do not require mirumum flow protection since there is a tiow path to the reactor pressure vessel. The loss of feedwater has been analyzed in the SAR and the kus of feedwater bounds the consequences of possiole malfunctions associated mth this change This change has no Technical Specifications associated with it; the thermal parameters associated with feedwater are not affected. Therefore, there is no reduction in the margin of safety.
Procedure Chance Recuest (PCR) 2 2 29 (Revision 30 C2)
TITill: Feedwater iIcaters sud Extraction Steam System DESCRIPTION: IV to this change, a procedural precaution prohibited operation of the feedwater heater (s) dry. This precaution was changed to allow operation of feedwater heater (s) dry if evaluated as acceptable by Engineering The pn: caution states that the evahiation should consider the effect of continued operation at a reduced feedwater temperature, equipment degradation, and personnel safety.
SAFETY ANALYSIS: This change allows can'inued operation with a do heater (s). Steady state operation with dry heater (s) is achieved shortly aller the malfunction w hich caused the dry heater. Operation with dry heaterfs) does not mercase the probabihty of adthtional heaters malfunctioning. The consequence of operation with dry heate%s) is has of feedwater heating Operation with dry feedwater heater (s) still allows a significant
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arruant of fadwata leatmg in cornparium to operation with the name number of heater s bypasci lhe ;
omepw of operating mth heater (v e % lourated by operation with heaters bypasel Operation i with heater (s) dry can result in degrac A 4 the affected heater (s) llowever, Engmeeting evaluation ;
is requiral and aill include equipment degradation considerations prmr to contmtal operation of dry l hentet(s). No other equipment is affected by operation with dry heater (s). Operation of dry heater (s) 7 results in less of a dmp in fmlwater temper ature than if the heaters are bypassed, therefore, emsequence is less. A stakten kms of heater tube integrity could result in a partial loss of feedwater. A totalloss of fmlwater n ent is analyzed and lourds this activity. Feedwater heaters and feedwater temperature are l not addressed in the lechnical Specifications.
Pnienhire Chance ReacesHPCR) 2 2 30 (Revnion 38) ,
i TITIJL Fire Protection (FP) System Dl!SCRIPTION: A new section was added to this pnicedure to alknv placing the old jockey pump into senice without
, dnabhng loth main fire panps ard to prevent entry into Linuting Condition for Operathm (LCO) 3.1 S.C.
1his new secthm tempranly lowers the auto start setpoints of the main fire pumps and allows the old jockey punp to maintain system pressure. Only one fire pump is inoperable at a time and the setpoints are reset following 15 evolution, This allows maintenance of the new jockey pump and its associated :
piping without entry into an LCO.
SAFliTY ANALYSIS. Use of the old jockey pump has tan previously evaluated 1hc use of the Icwer auto start setpoints oceturni pnar to the implernentation of Design change 87 154, FP System Jockey Pump Replacement.
1hercline, the probabihty of a plant event has not been increased over what was pre iously evaluated in the SAR The probability of pressure perturbations is reducal with the revned auto start netpoints vt existing manual pump operation The tse of the old jockey pump and lower setpoints will rot adversely afkot the abihty of the fire suppression system to mitigate a fire or provide Spent Fuel Pool makeup, it does not incicase the probability of FP system component malfunctions Failure of the fire pumps or jockey punp has been previously evsluated and no new failure modes are intnduced The Technical Specification requirement fit both main fire pumps to auto start sequentially and maintain the system pressure above 65 psig has been maintained, therefore, the margin of safety is not alTected Pnienfure Chance Reauest (PCIO 2 2 31 (lievision 12)
Pagdure Chance Reauest (PCIO 2 2 31 A (Revision 1) 11Tlji: Fuct llandling . Refuel Platform (2 2.31)
Fuel 1landling . Refueling Platform Compment Checklist (2.2.31 A)
Dl!SCRIPTION: The cruinge to Pnmlure 2.2.31 incorprates recommendations in response to NRC Inftsmation Notice 94 13 Supplement 1. These recommendathms cmtrol bridge configuration dunng perials when the p!atfmn is rd in use and ensure bridge power is maintained to space heater components.1he resulting change to component lineups is incorporated in Procchire 2.2.31 A. Changes were also mr.de to Procalure 2.2.31 to implement Setpoint Change Request (SCR) 9619 which revised the loading and unkalmg pressure wtpoints of the Refuelliridge Air Compresstr (see Safety livaluation fir SCR 96 19).
SAFETY ANALYSIS. Enhancements have been made w hich improve configuration control of platform components thereby reducing the probabihty of a refuelmg accident. No component lineup has been changed fi r when the platform is in use. Component lineups for when the platform is not in use are designed to further minimite misoperation and prevent equipment failures during perials of operation No components are manipulated or positioned in such a way to cause a refueling accident and do not atTect equipment required to respr.d to this event. Space heaters are left on by this change which will reduce platform component humidity etTects, thereby reduemp failures of control commments. Additional loading to Motor Control Center U has twen evaluated and found to be within component design requirements.
Ihgn of the platform is not being ahened, nor is the operating configuration of the platform. The refueling accident Imunds the mechanical aspects of this change. The design basis loss of Coolant Accident tounds the electncal aspects related to distribution systein loading. All compment 2 139-
f i
marupulatums guiarmed are emsistent within these design basis accident n ents The change &cs not ,
affect any Tcc!mical Specification related criteria or survedlance requirements needed to verify the j
~
equi l> ment is operational as required by Technical Specificatims dunng platform operations.
Procedure Channe Reauest (PCR) 2 2 32 (Revnion 251 l TITidt i uel Pool Cmhng and Deminerah/er Systern DlISCRIPTION: 1his loudure w as in ised m provide instructions for draining the Reactor Cavity and Dryer / Separator Pat via the Residualllent Removal (RIIP.) system to "11' C<alensate Storspe Tank. This draindown rnetisd is tablesenbed in the USAR; h(m ever, it in accannxdmed by cunent plant design of the RIIR arxl C<alensate Makeup (CM) systens This a& led flexibility is desired to allow for reactor pressure vessel (RPV) reassembly during reactor cavity drairxkmn.
SAFIITY ;
ANALYSIS: Two safety evaluations w cre performed for this activity. The first addresses the overall safety impact to the plant, as follows: !
The mly evaluated plate events that apply to this activity are tle less of Shutdown Cooling (SPC) and Imss of Cmlara Accident (LOCA). The SDC system will be manually removed fr nn senice and. if required, be re-estabhshed by Jiuttmg the drain valve and startmg an RilR pump 1he possibihty of a LOCA is not increased because the Reactor Coolant System is ruit pressurized, and human error is minimi/ed by stationing personnel at key kications. If a LOCA occurs due to uncontrolled drairulevn, the drainape would be terminated by a Group 11 Isolation which would isolate the SDC system The leakage damage would be imuted to the drairukmn SDC loop, or the Torus area, and not afTect redundant systems or components. All components will be operated within their normal hmits or functions, with the exception of fous short sections of piphg in the Radwaste and Control lluildmps Although the systern design pressure of 50 psig for these sections will be imtially exceeded, the components will be w ell below the code allowable working pressures lhe RilR pumps associated with the system being drainal will be stopped prior to commencing drainmg to prevent operation beknv Net Positive Suction Ilead linnts lhe Gnup 11 isolation will limit any loss of RPV inventory and maintain the core covered Additionally, at least one RPV injection subsystem witi be operable to reikxx1, as specified in the Technical Specification liases for Section 310 F.
A muni safety evaluation was perfo med to specifically evaluate the establishment of operabihty of the selected RI1R loop abgned in SDC with the condensate supply valve open, as follows' The pmposed draindown activity removes RilR SDC fnun senice per approved pmcedure.1hus, the probability of occurrence or consequences of a Inss of SDC occuning as analy/ed in the SAR is not irucael U wkr cold shutamn conditions, only low energy draindown LOCAs are credible events A 1 OCA occurring on CM piping simultaneously with performance of this proposed activity would be tammated by automatic safdy features and is bounded by current analysis for postulated breaks in RIIR pipmg Acanhtmg operabihty status to the RI1R hiop aligned in SDC is reasonable and consistent with current Technical Specification Hasis 3 5, even with the associated corxlensate supply valve open. An operchir stationed at the vahc in constant communication with the Control Room assures that valve ekwure is uutiated imnnbately at arcetion of the Contml Room Operator if a i OCA occurs in the ewnt the R1IR ankmsate suetwn valw cannot le ekwal, realignment of R1IR fm:n SDC to 1J'Cl would result in CF water combined with torus water being injected into the RPV. Although wat:r frmn outside sem ary containment would be contnbuting to injection flow rate, the safety injection function of the RilR ioop would be achieved Thus, the probabihty of occurrence or consequences of a LOCA as analy/a!in the SAR is not irocased No active RI1R system components will be used No CM or R1IR
( components will be operated outside of their hmits. This activity maintains other limergency Core i
Cooling Systems (liCCS) operable as required per Technical Specification 3.10 F for mitigation of postulated necidents This activity uses the hydmstatic head of a flooded cavity to gravity dram through installed piping, without the aid of active components.1he RilR manual valve being open does not invalidate the To:hnical Specification basis for R1IR system operabihty when aligned in SDC, and &cs not impact the configuration or availabihty of redundant !!CCS systems and safety logic systems tha-l l
140-
lensle auto isolatite capabahty 1hus, this actiuty does rmt reduce the mar gin of safety definn! within de basis fit a injuncal Sjwificatan requitanent, I t eldict the R1IR hiop bemp used, or any redundant I safety system
['nxediarc Change Request (PCR) 2 2 60 (Rnisio+i $5) l'wcedure Change Reuucst (I ')6 [ SOT.301 (Rension 1)
Pnwdure Change Reouest (15 JOT 301 (Revision 11 1
TillJL l'rimary containment Coohng and Nitrogen inesting System (2 2 60) :
Standby Uns Treatment (SOT) Operability Test /OIT Uns llow Morukt l'unctional Test Division 1 i (61 SOT 301)
SOT Oper abihty Test /Off Uns l'knv Momtw l>unctional Test . Division 2 (6 2 SOT 301) !
Dl!SCRIPTION 1hese imedure rnisions tenune guidance for cmling down the Standby Uns Treatment (SOT) train !
after a nm of i 15 minutes by placing the heater contsol switch to " Oft' for approximately live minutes '
before placing the fri switch to " Auto " Justification for the removal of this guidance is provided by
!!nginectinglivaluation 964104 i SAll?TY :
ANAL.YSIS: 1hc SOT sy= tem is un! for nutigatun of de amequenecs of an accident and cannot irutiate an accident. ,
Renaning the fiw ninute cookkvn hme fnen tbc procedures will not aticct the operation of SOT or any of its design functions, not will it affect the abihty of SOT to mitigate the consequences of an accident.
1hese pnudures are used during normal operation, not during accident situations Changing the pnmhuem to remove the cooldown time will emt alTeet any of the components of the SOT system lhe SOT system is currently operated without a cookknen time in other CNS pniculures Technical Spuficatao margms of safety are not affected because the changes to these procnhares do not afket any of the design functions of any of the SOT cornponents.
Pmetdure Chance Request (PCR) 2 2 66 (Revision 41)
TITiJi Reactor Water Cleanup (RWCU) :
i OliSCRIPTION: 1his proccJun e revision alkm s RWCU pump operation without Control Rai Dnvc (CRD) mini-purge for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during cold shutdown conditions lhis practice was prniously implemental successfully in Rlil6 as a tempirary change and is now made permanent !
SAlliiY ANAL,YSis. t he probabihty of a RWCU leak is not increased because the seals can be operated up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without seal flow, and the procedure will restrict operation be> ural that time perini 1he consequences of a plant ovnt are not increaul twause leakage from a seal faihue will be water, and not steam, which will not cause an increase in o!Tsite dose. The RWCU pumps rv not relied upon m any accident or :
transient analysis The procedure testricts the operation of the RWCU pumps without mini purge flow to coki shukkmn perials only.1he consequences of a malfunction are loss of pump on ration or seal leak. These are the same consequences as during other operating males, except that during operatmg onhtions the leakape umid be steam wn: water. The faihire of a pump seal is encompassed by existing pipe brvak evaluatans the RWCU pumps are init akhesel in the basis of the Technical Specifications Proenture Chance Reauest (PCR) 2 2 66 (Resiiton 42)
TITIJL Reactor Water Cleanup (RWCU)
Dl!SCRIPTlON: Omd.uw w as mLksi to this procalure to protut timinal overp;essuritation of RWCU penetration X 14.
This is in re rpme to Geinic Iciter 96416 amsns for overpressuritaten of primary Containment (pC) penettations during a less of Coolant Accident SAlliTY ANAL,YSIS: lhis actiuty will not place penetration X 14's isolation valves into an alignment contrary to that allowed by the Technical Specifications, nor will it result m the operation of any eqmpment, systems, or appurtenances lqm! eur ently analymt limi s. This activit) is intended to minimize the probabihty of 141 I __ _ __
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l emuhng the a de alkm able fit de poccu pipe widun penetration L I 4. Ilow ever, ewn if this activity failed to prevent this from happening, the resultant stress levels will rad exceal those of a fanhed ;
anxhtkn.atsunng totainnat integnty will not be lost. Since the RWCU PC penetration integnty will !
be maintained, the existing safety analysis remains unchanged Deenergitmg one of the valves in the .
qm puntun is tot a safety uncem airne de other vdve is in the closed pomtion, and because the closed l valw rcwives a umfirmatory signal during isolathm, negating the possibility of an operator erroneously quung it when iudathm is requiral The integrity of the Primary Coolant or the Primary Containment h anlary will not be reducal, tnw the abihty to achieve isolation No assumptions uxx! in any analysis hcie been impacted by this activity. Rather, safety will be increased because the possibihty of r overptessuritation of de subject penetration will bc rnlucal Prom!ure Chance Reauest (PCR12 2 6R 1 (Revision 15)
Pmtedge Chance Reducst (PCit) 2 4 2 2 3 (Revnhin 12)
TITul: Reactit Recirculation (RR) System operstmns (2 2 6R1)
Recirculation Pump A or 11 Seal Fathue (2 4 2 2.3)
Dl!SCRIPTION. . Procalural actiorn base been rewritten to suppet either a prompt or planned imdation of a RR neal fadure 1his is required to allow procedure adherence based on a varying set ofinitial plant conditions arul plausible now rates funn seal failures to the drywell atmosphere. Steps were added to address the pubihty that the cookken rate of RR piping between the imdathm boundaries will in most cases exceed the estabhshed unidow n rates lius leak imdation capabihty is desirable regardless of omddown limits in order to imdate art uncontrolled blowdown of coolant to the Primary Containment < Consequences of this isolation and the resulting clTects are stated within the prudure and require evaluation prior to restorstkm of normal system operation.
SAlTil r' ANALYSIS; No manipulathms are Ivtlbrmed which would increase the probability of a RR seal failure.
Connequences of any pusible event are tuunded by the Trip of One Recirculathm Pump and ReciruJathm Pianp Sevute Analysis lhe actions os imdate the seal failure are untien to minimize the impact to drywell parameters and minimize the o msequences of the leakage. Acthms do not introduce mechamsms by uhich failure of addith nal equipment can be gustulated Specified actkms taken are within the design of the components to be operated Isolation of the failed seal prevents a continued challenge to containment systems Technical Speci0 cation enteria are not changed as a result of the revismns to these procedurcs Speci0ed nethms are inchidal to monitor and ensure nunpliance with requirements during the failure and prior to restoration of system mumal operatior if required.
Procedure Chance Request (PCR) 2 2 74 (Revision 26)
T11Lil: Standby 1.iquid Control (SLC) System Dl! SCRIPT 10N: 1his PCR iruspratal changes to pmvuk guidance for controlhng SLC storage tank temperature during
. chemical addition This new guidance established an upper lim.t for St C storage tank temperature of I40*F, alove which avadable Net Positive Suction Ilead (NPSII)is below minimum required for SLC pump operabihty.1his PCR also established a specific band of 100-110*F to be maintained for smaller akhthms lhe pnuxhne change chinfied that SLC is inoperable w hen making large changes in soluth n concentrath n at a temperature band of 135'F to 150*F.
SAFlITY ANALYSIS: 11is change only limits the temperature withic the system design and design basis requie ments while mLhng dunicals wla SI C is ausidered operable. The change only allows the tank temper 6ture above NPSil requirements during chemical additmn when SLC is inoperable. The change has no relationship to initiators of plant events liy ensuring the SLC system is operated within its design basis, the probability of ocetuTence or consequences of a malfunction of equipment important to safety are rot increased 1he mimmum temperature to ensure the chemicals react and are dissolved into the solution is unchanged The changes to this procedure do not change the SLC tank temperature, level, and concentratum requirements of Technical Specificatk ns, nor does the change alter any performance ;
characteristics of the St.C system 142
i hecedure Change Reauest (PCR) 2 2 7611(Revision 1) -
1111JI: Turbine l'quipment Coolmg (Tl!C) Water System Instnunent Valve Checklist DESCRIPTION As a resuh of Design Change 90 326 (TEC/ REC to Plant Air Canpresses M1), TEC.PI 413 and 414 were added to the system Consequendy, valves TEC 86 and 98 (TEC-Pl.413 and 414 Isolation. l respectively) became Instrument and Control valves and were added to ti e lastrument Checklist in Pruunhue 2 2.7611 The rmnal pnitun for these valves w as changed to open so the indicator s (gauges) i can be utdited for systern analyzing while in operation SAFliTY ANAL,YSIS-. 1hese valves are igulatan valves for pressee irxbcahirs fit the station air comp *eswe TEC lines Placing these valves to de open position will not inctcase the probabihty of an occut ence of a platit event since the pressure gauges imtalled will be the pressure bourulary preventing the loss of TEC pressure. The TFC system is rmessential arx! there are rm radiological concerns related to this system. Placing the isolation valves in the open position fa the indicators to function will not increase the probabihty of !
occurrence or consequences of a rnalfunction of equipment important to safety; these TEC system cornpments have no effect on equipment imputant to safety. This equipment is not referenced in the ,
Technical Specifications arnt will rot alTect equipment referenced in the Technical Specifications, dwrefore, it wdl rot reduce the margin of safety as dermed in the basis for any Technical Specificatiott i Procedure Change Reauest fPCR) 2 2.90A (Revhion 3 C2)
TITIJr 12.V KV System Comprnent Checklist DESCRIP fluN: The breaker lineup for panel OPA PNielJ.RWI was revised to agree with the as buih condition per ;
Design Change 91-077, low 1.crel Radwaste Storage Facihty.
SAll!TY I ANAL,YSIS: 1he kas of the 12.5 KV system is ruit evalunted as the initiator of any accidents or relied upon to support any accident mitipation functions. The configuration of OPA.PNieldRWI does not exceed the panel i rating The loads on this panel are protecal by ciremt breakers preventing a fault from alketing the ranairaler of the 12.5 KV system This panel does not supply power to equipment defined as imp stant to safety.1herefore, dwre is rm impact to equipment supplied by the 12.5 KV system lhe configuration of OPA-PNI IJRWI does rud afkrt the stability of the 12.5 KV system This panel does not affect any ,
equipment defined as important to safety and no margin of safety in defined in the Techr'ical Specifications regarding the 12.5 KV system. Therefore, the margm of safety will not be reduced.
Procedure Change Reauest (PCR) 231 (Revision 16)
TITIJL General Alarm Pntalute DESCRIPTION: 1 his procedure u ns revised to defme the method to evaluate the impact of disabling alarms, provide a process to d~able an alarm and appropriately momtor the system or parameter as*ciated with the disabled alann, and desenbe the method (br tracking disabled alanns. The procedure identifies key alanns that are credited or identified in design and licensing basis documents as useful to diagrose significant events A&htionally, alarms identified as Technical Specification acceptance enteria are consilmi Ley ahums if key alarms are disabled, alternate monitoring or compensatory actons will be required Provisions are included for perfbrming a ibliow-up Safety Evaluation, when required SAFETY ANAL,YSis: 1he arunariakt systan is not a amtnbutor nor initiator Ibr any event identified in the USAR. Response to a design basis accident, necessary to limit core damage and any subsequent release, is performed by the automatic ressmse of safety systents lhe consequences of any event previously evaluated in the USAR are not changed by disabling alarms not identified as key alarms Alanus identified in the Probabilistic Risk Assessment contnbute to the diagnosis of plant conditions and may hmit the '
ansequences of an event beyond design basis. Actions based on these alanns may lower frequency of core demage. These alarms are identified as key alanus and require altemate monitwing or compensatory netum Alarms asstintalwith the annunciata system, aral as discussed in the USAR, are isolated from 143
i tic nguigmwnt aral cteitrols ruwssary for tic nafety sptem to perksm its interxled function The Operator ;
uses the annunciator system for early detection of a problem such that prennptive action may be takert 11e USAR Safety Analysis does tot crnht precmptive action by the Opnattes Disabling an alann will not inescase the probability of a plant event rr equipment malfunction of another t3pe than previously evaluatal in Oc !ISAlt.11e Tedinical Sjwificata n margm of safety will not be inlucal because alanns '
identifini as l edmical Specification acceptance entens require alternate monitoring or compensatory actims, arwl Technical Specification acceptance cniteria alarms will require that the appropriate actums ;
be taken as specifint by the surwillance pnwalure containing the acceptance criteria hocedure Chanedcquest (PCR) 2.3 217A (Reusion 14.1 C5)
Proecdure Changslcuuest (PCR) 6 PRM 316 (Revismn 1.0)
Primlare Changdequest (PCR) 6 PRM 317 (Revision 31 hwahire Chance Recuest (PCR) 6 PRM 31N (Revision 2) 1111.10 Par el Q Annunciator Q.1 (2.3 2.17A)
Cmtrol Rmm Air Sampling System Krumn Source Calibration (6 PRM 316)
Control Room Air Sampling System lifectronic Calibration (6 PRM 317)
Control Room Air Sampling System Functional and logic Test (6 PRM 31 H)
DliSCRil' TION: !!ach of these paratures was revised to specify the number of detectors, either noble pas, iodme, or particulate, dust are requirn! to le oper able to ensure that the Controf Rmm Radiatiori Morutor (CRRM) is operable per the requirements of Technical Speci6cetion Table 3 2.D. Prior to this change, the gnrahues required that all three detectors be operable for the CHRM to be considered operable. This i diange siwilies that only Iwo of the tluce detectors are reqmted to be operable for the CRRM to fulfill its safety function nrxl be umidered operable per Technical Specificatmnt in addition,in Pncedures 6 PRM 316 and 6 PRM 317, the acceptable and as-left value for the time averaging functum of the ,
Control Room Ventilation Monitor inime channel was revised This change is basal on !!ngineering Judgement 97-02I which avecifies an acceptable value of 30 seconds SAlliTY ANA13 SIS: 1here have twn ruichanges to de manner in which the CRRM functum and there is no malfunction of dw numitor which can cause a plant transient or accident. The changer to the operabihty requirements made by these pmcedure changes maintain the design basis functions of the CRRM. In particular, the CRRM wdl stdl fulfill its safety functam to initiate the Control Rmm timergency Filter System (CRl!FS) in all of the design basis cuts liir which it is required The arbitrary failute of the CRRM is outside the CNS licensing basis since the CRiiFS thies not have to meet tungle failure entenia. Technical Specification Table 3 2 D asstunes that me radiation monitor is adequate to essure that safety criteria are fulfilled I!ven wth the operability hmit changes proposed by the PCR, the Techmcal Specification requirements are fulfillnl The only effect due to the change in the iodine averaging time is a faster detector resixmse time which will ensure that the consequences of any accident in which the CRl!FS is acquired will remain within dose already reviewed and api oved by the NRC. The only possible conwquence whose probabihty has been incicased, although a very small increase, would be an inikhutent actuatim of the CRllS due to a spurious t ip of the iothne detector. An inadvertent actuation of the CRITS has no athuse cmeperres siiw de Cmtml Rmm Ventdatmn System will still maintain the Control Rmm in a fully functioning cmdition even with the CRiWS in operation With the change m averaging time, the detector will be nore responsive in the event of an accident where it is required to function, therefore, there is no reduction in the margm of safety.
howdure Chance 1(Eguest (PCR) 2 3 2 25 (Revision 32)
Procedure Chnnec Request fPCR) 2 4 21.1 (Revision 13)
Praedure Chance Rcouest (PCR) 61.00 601 (Revision 6)
T1Tl th Panel 9-l Annunciator 9 4 2 (2.3.2.2$)
Small 1.cak inside Pnmary Contairunent (2.4 21.1)
Dady Surveillance log Technical Specifications (6100 601) 144
DliSCRIPTION: A C(nhtum Advero to Quahty (CAQ) w as irutiated wiuch identified a pulsing phernenens to the F sump which pmides one ialicatum ofirucased reactor coolant water leakage. livaluatmn has shown that the phenomena is a result of hold.up in the llVAC units and that leakage when averaged over time is unstart In <tdar to pmide the Oguntar with a method to address this phernenens aral venfy that a real paiblem exists or that the leak ste is a result of a pulse, a reviske to the alarm respir.se process was recornmerated The changes made to these pnicedures serve as a tcanorary measure until a Design Change can be implemented Alarm respose pnwedures and abrusmal procedures wcre enhanced to verify increased leakage by examining drywell temperature, pressure, humidity, and radiation Aner verifying levels are not increasing, the Operator can validate the leak rate by attempted pumping of F sump.
DAITTY ANAL,YSIS: Design, configuration, and functhm of the F sump remain unchanged F sump is isolated on a Group isolatior, signal.1he procedure changes do rot impact the abihty to isolate the F sump. Pncedure guidance is being added to enhance the Operator's abihty to diagnose small leakage prior to group iolatkn Response to indications of sump pulsed fill and increased leakage is enhanced. There are no acektent initiators or contributors associated with this pnicess Primary containment valve reliability remains unchanged The procedure changes have no impact on equipment or munponents that could increase the probabihty of occurrence of a malfunction of equipment important to safety. These pnicedure changes will enhance the Operator's abihty to quantify the leakage. As such, the margin of safety as dermed in the basis for the Technical Specifications is not changed.
Pnitedure Chance Reouest fPCR12 4 2 41 (Revision 16)
TITI.th Residual I leat Removal (Ri IR) 1.oss of Shutdown Coohng Dl!SCRIPTION: The Ri!R system piping is tot analyred for water hanuner loads. Operation of the R1IR system in the Shutdown Coohng (SDC) mode with vessel temperature above 212 *F could result in a water hammer event occurring in the kilR suction piping upon realignment to the low Pressure Coolant injection (1.PCI) nxde. If RI1R should isolate due to a Pnmary Contamment isolation System Group 2 isolation signal and 1.PCI in required, then the kiop needs to be thoroughly vented prior to and after realigrunent to 1.PCI. This change provides guidance to Operations perumnel for operation of the Ri!R system if IJ'Cl injecthm is required.
SAfliTY ANAL,YSIS: Actins alled by tius pmcalure fall within the hws of shutdown cmling event analysis The design basis accident remams unchanged as ucli as the availabihty of the other limergency Core Cooling Systems (1 CCS) w thin this operating nxdc because actions are targeted at event response without afTecting any imtiatmg aahtions of an event or accident. LPCI mode of a kop of RIIR in SDC is operable as stated within the basis of Technical Specifications lhe manual actions described in the procedure place the system in the LPCI lineup as needed to suppnt 1.PCI operation fnun SDC operations. This change secogm/es that 1.PCI may not be inunediately available for restoration based on operatkin within SDC mode and directs to restore I.PCI operation if required within the constraints of system design criteria during SDC operations The consequences of a single failure are not changed as the manual steps uill ensure either IICCS system can be restored for system redundancy. Probability of a malfunction of equipment remains unchanged as the added steps are empkged to prevent a water hammer should the system be realigned to a 1.PCI lineup. The manipulatkins in response to an isolation effectively return the system to Ij'Cl operation if needed to provide core cooling The consequences of a failure are less than should the availabihty of the comnon shutdown coohrg line be lost. Applicable steps are added to alert personnel of potential steam or contaminated fluids. These manipulations do not affect opposite dnision sourtes of equipment or other same dnision liCCS/cornponents important to safety Changes do not imulte any Tecimical Specification limits. System operation is performed to ensure adherence to the Technical Specifications as allowed by the basis of the Techrucal Specifications.
145-l l
hardme Chance Reone<t f PCR) 2 5 2.3 (Revision 39)
TITLlh Radu aste I hgh Conductnity 1.iquid Waste Fkior Drain Sample Tank Fluid Transfer Dl SCRIPTION 1his rnision alkms an altername nxtini ofietating fluid addition to the floor dram sample tank during discharge and establishes a method of tank discharge with the liquid dischar ge monitor irmperable.
5AFl!TY ANAL,YSIS: 1his change ai. mes actkos alknvable urder Techrucal Specifications and does not afTect the operatum of any equipment related to discharge activity limits, allowable rate of discharge, or the ddution flow hnuts of the discharge strcam. Thereftve, the pmbability of occunence or consequences of an accident or malfunction of eqmprnent imp riant to safety are not increased 1his change is limited to hquid daluu ge oper ations and does not afTect existing discharge linuts or interact with additional equipment important to safuy. Malfunction of the liquid radwaste discharge momtor is addressed by Technical Sprificatums and tlus change on rates the equipment under the permissible actions for this malfunction.
1herefore, there is to reduction in the margin of safety as defimed in the basis for any Technical Specification.
Procedme Change Reuuest (PCR) 3 410 (Rnision 7) hwedure Chance Reuucst iPCR) 114.11 (Re ision 12)
Tlillk Station Modification Changes (3.4.10)
Status Repnis(3 4.1I)
Dl:SCRIPIION. The revision to Procedure 3 4.10 dumped how On The-Spot Changes (OSCs), Amendments, and Revisions to Mothfication Packages (MPs) and Design Changes are formatted. All steps which discussed Status Report responsibihties were moved into Procedure 3 4.11.
SAll!TY ANAL,YSiS' The approvals required by the rnisal OSC process are identical to that of the old process, with the exception of the Shill Supenisor signature and the Statim Operations Review Conunittee (SORC) signature Ibr non Amendment and non Rnision types of OSCs.1hese signatures provided no added vahr, all OSCs receive the same level of rniew as the original MP.1hc only instances where an OSC will run nxent SORC approval are those instances where the Safety Review of the MP has not changed in any way, the design basis or design entena of the MP has tot changed, and the OSC does tot im olve systems, stnctures, or components not impacted by the original MP. These iestrictmns, combined with the required reviews, combmc to provide the assurance that changes made to the plant via the OSC pmcess wdl rut increase the probability of occurrence or consequences of an accident or malfunction of eqmpinent important to safety and will not reduce the margin of safety, Procethne Change Reouest (PCR) 3 4.10 (Revision 9)
IIT1.lt Modification Document Changes (New Title)
Dl! SCRIP 110N; 1his procedure inision prmided for an Instant On The-Spot-Change (IOSC) pn(ess for moddication thounents 10SCs are restricted to change mly the installatmn section of a nuhfication document when components are out-of-senice or ino;wrable. All IOSCs are required to be converico into " normal" OSCs prior to submittal of the Ready for Testmg Status Report.
SAFliTY ANAL,YSIS: The new pmcess is limited to periods during the installation phase of a modification document. The Wnk Contml/ Maintenance Wtuk Request pmcesses provide the assurance that IOSCs will only alTect niuigenit that is tot in seniec. As a result, this new paress cannot increase the probabihty of a plant event. Controls in other existing proecsses provide the assurance that only systems, structures, or emiponents (SSCs) not required to respod to an event will be taken out of senice. This new process remains botmded by the "rxumal" OSC pnwess, w hich provides all the reviews reqaired by 10CFR50,
- Appendix 11. Section 111, Design Control, prior to declaration of eqmpment availabihty. As such, the pnwess contmues to provide the contmls necessary to assure that SSCs will not be placed in senice unless. required rniews have been completed As a result, the new procer,s does not increase the 146-
t pmbalnhty of mouncnce cr arnegtusoes of equiptra nt malfunction and provides assurance that margins of safety for SSCs will not be reduced.
[ Pacedure Chance Reauest (PCR) 3 7 (Rnicion 14)
TITLl!: Drswing Change Notice (DCN)
Dl!SCRIPTION: 1his pnmlurc inisim nukle various changes to alknv for more eflicient processing of DCNS, including the folknving: 1) provided a nsam to taxvcile DCNs uhen Procedure 3.7 has been revised. 2) provided a neans by which minos drawing ernes can be corrected without being associated with a Mahfication Package <r linguwnng livaluatim. 3) clanficd that Stand Alone DCNs that are USAR drawings require a safety review,4) clanfied pacessing of vendor drawing rnisions, and 5) made various editorial and administrative changes throughout.
SAFl!TY ANALYS!S: The USAR states that the licemcc shall have a pncess ,%r the rnision of plant drawings. The subject revision does rmt affect this requirement. These changes do not directly afTeet plant equipment.1he DCN pnuns panides the rxxtssuy controls to ensure drawing changes are appropriately reviewed for safety concems. The DCN pucess is used as a vehicle to re isc drawings contained in the USAR; however, appmpriate controls and safety evaluations are performed to address safety issues under the USAR change process These pncedure changes are not related to and cannot cause any types of equipment malfunction.1here are no margins of safety related to drawing changes defined in the Technical Specifications.
['nsedure Chanec Reauest (PCR) 3.7 (Revision 15) l'mcedure Chance Reauest (PCR) 3.8 (Revision 12)
ProccJure Chance Reauest (PCR) 316 (Revision 8)
TITIli: Drawing Change Notice (3,7)
Drawing Control (3.8)
New Drawing Pnparation and Appmval (3.16)
DliSCRIPTION' These rnisions punided clarification fbr pncessing new vendor drawings into the Drawing Control
'rogtm They specified that Drawing Change Notices (DCNs) are imt utihred for vendor drawings which are under the ' endor's revision contml, unless NPPD intends to revise the drawing, thereby accepting revision control of the drawing. Engineering evaluation of vendor drawings for CNS applicability is performed during the pmcurement receipt pacess. The ditTerence between equipment vendor drawings and senice supplier drawmps was also clanfied.
SAFETY ANALYSIS: 1hese changes to the drawing antrol process do not involve a physical change to the plant, do not alTect plant apupment, and do tot alter any system design function Thus, this activity will have no radiological etTects m any system, stmeture, or component and will not increase the consequen:es of an accident or aluipn'ent malfunction No ne'v or unanalyzed failure males are introduced The changes provide for more ellicient processing of venar drawings through the Drawing Control pmcess. There are no margins of safety related to drawing changes in the Technical Specifications.
Procedure Chance Reouest (PCR) 3 WRnision i 1)
TITI.E: ASME Code Testing of Pumps and Valves DESCRIPTION: 1his rnision to Procedure 3.9 documents the methodology and basis for the determination ofinsenice testing (IST) stroke time reference values, IST Retest I.imits, and limiting values of full stroke time (refened to as Operability Limits in surveillance tests) for valves being stroke time tested per the CNS IST Pmgram SAFETY ANALYSIS: This procedure change does not impose new testing, does not delete testing, and is not changing the ankrt ofexistmg testmg. Delineation of the process for establishing IST reference values, IST Retest 147
1 1
l l
l l
1.inuts, arxl Oprabihty Linuts continues to ensure operational readiness in accordance with the purpose of the IST Pmpam Consequences of malfunctions are dictated by component safety function This actaity does runt change the w my in u hich IST testmg is perfirmed and &cs not afTect a cangment's safety function. Since stroke tirne limit changes as determined by this process do not impact the ability of a valve to perform its safety function, this actinty ars not reduce the margin of nafety. In addition, the pusess(kfinalis in amadance with ASMl! Section Xl guidance as required by iOCFR50 $$(aXf). i Procedure Chance Reouest (PCR) 3 16 (Revision 71 11TI.th New Drawing Preparation and Approval Dl!SCRIPTION: 1his resision added albtional restnctions to the initiation of new drawings in that a Drawing Change Notice is required to a&fchange drawings tdin the Drawing Control Program in accordance with Procedure 3.7. Drawing Change Notice Also, the use of sketches has been limited to infonnation a cunnutx This is to mere that the drawings utilved for plant maintenance, testing, and moafications :
have been approved by Design !!nginecting ,
SAll?TY ANALYSIS: Maintaming the drawings in confonnance with design requirements assures that any changes made to equipment requirul for rurmal plant oguation, including detection of upset conditions, will not adversely i aflWt its function as described in the USAR or other Design flasis Documents. Pmper maintenance of drawings ensures that any changes made to safety related equipment or associated pmecdures will not adversely affect their safety function as requiral by the Safety Analysis and that any malfunction of equipment is within the hmitations of the plant design basis. Maintaining the drawings in conformance with the design requirements ensures the plant will operate as required in plant transients, accident cisahtions, aml rugmal operation 1he margin of safety as dermed in the Technical Specification bases assumes that equipment is in accordance with Design approved drawings. Requiring drawing adhtiondchanges be approved by Design F.ngineering ensures that this assumption is correct.
Procedure Chnnee Reauest (PCR) 3 25 tRevision 4)
TITIJi: Replacement Component I! valuation (RCli)
Dl! SCRIPT 10N: this pmcedure change made various enhancements to the RCl! pmcess The most significant changes inchided the folkwing: 1) alkmance of rmidentical replacement componentVparts even if the identical replacement is still available, 2) addition of reviews by alTected pmpam owners, 3) addition of attachments for design input and configuration document uplate identification, and 4) requirement to include !!ngineering hold point sign-otTs in Maintenance Work Request mstructions.
SAMITY ANALYSIS: 1his activity ensurs the equivalency of a relacement component'part by identifying and evaluating the critical specifications to ensure that the component /part will perform the same design and functional requirements of the original component /part. Therefore, this activity dies not increase the probability of occunence or consequences of an accident of malfunctmn of equipment important to safety and does not reduce the margin of safety as defined in the basis for any Tectuucal Specification.
Pro dure Chance Reauest (PCR) 3 25 (Revision 5)
TITLli Replaecment Component !! valuation (RCli)
DliSCRIpTION This proenture change made various enhancements to the RCl! pmcess, includmg the following:
- 1) kk:nufication that apphcable plant pntahires be refen nced in the installation and testing requirements section,2) clanfication that the preparer shall renew the Maintenance Work Request (MWR) Post Maintenance Testing assipiments to ensure adequacy,3) clanfication of requirements for completing amfiguration dwument changes, and 4) alhtion of MWR rmmber and CNS part number to apphcable attachments.
. ] 4 8-
SAll!TY ANALYSIS: 1his acuvity msures the equivalency of a replacement c annnent/part by identifying and evaluating the on6 cal design speificaixos to msure that the component /part performs the same function as the original unenenttait 1herefisc, this acovity d es not increase the probability of occunence or consequences of a plant event <r malfunctim ofequipnunt impittant to safety and the margin of safety as dermul in the s basis for any Technied Specification is rxit rnluced. ,
i Procedure Change Reauest (PCR) 3 281 (Revision 2) lil1Ji insenice Inspection (ISI) Program implementation DESCRIP110N: 1his change captural the CNS anunitment to examine the Rea: tor Pressure Venel Core Spray internal pipmg and spargets in accordance with NRC 18 Ilulletin 80-13 and llWRVIP 18. It aim included the requirement Ibr NRC cynluation of defects exceeding the acceptance enteria in ilWRVIP 1 H.
SAFl!TY ANAL YSIS: ISI is rut a pnxurst to any plant event desenbed in the USAR. ISI plays no role in accident mitigation The subject inision does not alter any equipment or pncedures relied ulxin to mitigate accidents or transimts.1k examinatkos arul tests pertirmed for ISI do not affect the operability of plant equipment and are nonnally perfonnal when the systems are out of service and are controlled by a Maintenance Wcsk Request (MWR), If the examinations are performed when the system or component is in senice, addithnal pntautuns are alled to the MWR to proteet operating equipment This change does not alter the inspection metinis required by the ASME Cale. Although ISI is required by the Technical Specifications, this change d ws not afTect the basis tccause these inspections are not required by the ASME Cale and evaluations of flaw a exccaling the acceptance enteria must be accepted by the NRC.
Prtudurgfhance Request (PCR1415 (Revision 151 11113L Elevated Release Point (liRP) and lluildmg Radiation Monitoring Systems OliSCRIP110N: This procedure was te ised to contain all steps necenary to place the liigh Range Kaman in senice.
Guidance was alm added to allow for Kaman troubleshooting due to a continuing problem with the inabihty to recreate / troubleshoot malfunctions after shutting down the Kaman.
SAFETY ANAL.YSIS: Changes are relatal to operation and maintenance of the Kaman monitoring systems only. These systems do ruit lunt the capability of causing an accident evaluated in the SAR. Consequences of any accidents evaluated in the SAR are rut changal becaux these changes pmvide Ocubihty to maintain these systems operable or to provide additional mettuxis to restore these systems to an operable status folkming equipment malfunctions. Changes do rx>t alTect or interact ulth equipment important to safety. The inoperubihty of these systems is ackrxmledged as not ca.sential to safety within the USAR, with actions arxl time perials fir thme actions set fath within the Technical Specifications Margins of safety remain unafTected as described 1 y the actions ihr inopeubility of this equipment, and no mar gins are alTected dunng the perial of time these systems are operable.
Pnudure Change Reauest (PCR) $ 4 31 (Resision 15) 1ITI.H: Post Fire OperationalInformation bliSCRIPTION: 1he cable route for Apperklix R cable 11476 was identified to be incorrect in the Appendix R analysis databar With the ca rected cable route, it was detmumed that Residual !Icat Rere d (RI1R) minimum ikwv vahr R1IR-MOV-MOl611 wadd be hist for a fire in Critical Switchgear Room IF in an Appendix R fire scenano Theiefore, this pnwedure was rnised to add marmal actions to operate RilR MOV-Mol611 fir an Appendix R fire scenario.
SAFETY ANAL.YSIS' 11ss change to a fire response pnwedure has no efTect on the probability of f:re occurrence. The change is to aid in mitigating the consequences of a fire. The Safe Shutdown Analysis assumes loss of all equipment in a given tire zone. This pnxedure change provides recovery from that loss, i c., manual 149-
action. The adational manual action reduces the consequences of a fire induced malfunction No a&h6 anal equipment malfunctions are introduced Manual operation of the RilR mimmum flow valve has locn gevkusly evalustal and does rut intnsluce any new accident sc4marios tr create the possibility of any trw equipruit rnalfunctxos Manual (quatun is currently desenbed in CNS Pncedure 2.2.69.2.
Safe shutamn procedures do cot funn the basis for any Tecluucal Specifications. Manual operation of this valve will not aficct Technical Specification margins for automatic valve operation.
Pntedure Chance Reauest (PCR15 4 311 Revision 16)
Pntedure Chance Rcouest fPCR) 5 4.3 2 iRevision 13)
TITIJL Post Fire OperationalInftsmatani(5 4 31)
Post. Fire Shutdown to Cold shut &mn outside Contml Rmun (5.4.3 2)
DI!SCRIPTION: 1he CNS Apiuxhx R Validadm eflint i&ntified anne analytical fire zones as requiring additional post-fire manual actions not currently desentui in CNS post fire pmcedures fbr coping with fire induced faults. These pnccdure changes incorporate the new manual actions required to achieve post fire safe shutamn.
- SAFliTY
. ANALYSIS: 1hese pncedure changes involve manual actions that enrure plant equipment is in its fail safe position, is utiliicd as allowed in CNS system operating pmcedures, or eperates systems as described by the USAR. Changes to these procedures are to aid in mitigating the consequences of a fire. Providing operuhes with Instructxvis for agiing with fire iluluced faults results irt a redectiors of consequences. The subject changes ensure that safe shut &mn capabihty is preserved. AdJ tional manual actions allow oguaixo ofequipnuit utme suprut etunponents arxRr cables are directly or indirectly alTected by fire, thus reducing the consequences of fire induced malfunctims. The manual actions do not create the smibihty of a new type of accident or malfunctkm Sullicient eight hour emergency lighting exists fit the Operatims crew to perform these additional actions lhe activities have been individually validated to be capable of being p ribnned relative to the time entical nature of the system /furction as described in the CNS Functional Requirements Analysis. Additionally, Operator training on these activities is perf(rmed to preclinic introdection of numan emr. The manual actions described are not the basis fbr any Technical Specification 1he activities confism to &c technical bases of exemptia for Appendix R and, therefore, the margin of safety as described in the Appendix R Safety I? valuation Reports is unchanged Procedure Chance Recuest (PCR) S 4 3 2 (Revnion 15)
TITI.II: Post Fire Shutamn to Cold Shut &mn Outside Control R(unn DliSCRIPTION: W Apguxhx R Validation etibit identified that a fire nhiced sptaious operation of R1IR-MOV Mol511 could prevent the opening of RI1R-MOV MO39I1 from the Alternate Shut &mn Panel due to a valve interkick This pmcedure u as revised to provide additional instruction to plant operations permumel regarding tne manual actkms treessary to ogwrate RilR MOV-MO3911 lblkming an .dternate shutdown fire.
SAFliTY ANAL,YSIS: 1his pnmhar change imuha a manual action to usure plant equipment is utihred as allowed in CNS sy stem operating pmcedures or operates systems as desental in the USAR. The change to Pmeedure 5 4 3 2 uill aid in nutigating the comequences of a fire requiring shut &mn outside of the Control Roorn Providing operators with instructkms for coping with fire induced faults results in a reduction of consequences. This procedure change ensures that safe shut &mn capability is preserved. The Safe Shutdown Analysis assumes loss of all equipment and cabling in a given fire zone and also assumes ccitam cnuble circuit faults which intnance spurious operations of equipment. This pmcedure change
, pnnides fit recovery fnun that km arxl recovery of positive contml fmm spucious component operation.
Manual action to open R1IR MOV MO3911 will abgn the Residual ilent Removal system as required to support the suppression pool coolind mode. Sutlicient eight hout emergency lighting exists Ibr the quatkos crew to perfonn the addithmal actkut This actiuty has been validated to be capable of being perfirmed relative to the time entical natun of the system as desenbed in the CNS Functional
-150-
.- -. - = - - - _ . - - . _ _ - - _ - . . _
l l
Raprunents Analysis 11e manual action to open R1IR MOV MO3911is required to achieve post. fire safe shuttkmn altressed by the basis of the Tecimical Specifications. The actisities omform to the technical bases of exemptions for Appendix R and therefore the margin of safety as desenbed in the Appendix R Safety Evaluation Reports is unchangal.
Procedure Chance Reauest (PCR) 5 R 2 (Rnision 5)
TITill: Alternate Ernergency Depressurization Systems (Table 2)
DESCRIPTION: 1his rnisim nale changes to labeling desenptions for various Emergency Operating Pncedure (EOP)
Plant Tenyxs ny Mahficatkins (PTMs) to provide additional clarity to the operator and to be consistent with labeling in the plant. It also clarified that liOP PTM 30 does tot allect the Reactor Core Isolation ';
Cooling low oil pressure turbine trip The section order in the pacedure was revised so that the instructans are hstalin ile orda of radiokigical(kwe significance, from least to greatest. The description -
of the reactor pressure vensel head vent was revised to match kical labeling.
SAFliTY ANALYSIS: CNS EOPs acmmrrniate strategics for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report on the lloiting Water Renator Owners Group Emergency Procedure Guidehnes (EPGs) fiund tic me of the limits speciGal in the !!PGs and EOPs, rather than those specified
, in the beenuhkhn basis, acceptable dunng degraded cmditions. The implementation of this rnised
!!rnergency Support Pmeedure (ESP) does not increase the probability of occunence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This rnised ESP cannot increase the pmbabihty of occurrence of any event analyzed in the USAR because tle procalure will mly Ic uul aller the event has commenced. The implementation of this revised lisp does not create a possibihty for an accident or malfunction of a dilTerent type than any previously evaluated in the USAR because the rnised ESP does not modify the operation or design basis of the plant For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
Procedure Chance Reouut (PCR) 5 8 2 (Revision 6)
TITI.E: Alternate Emergency Depressurization Systems (Table P Dl!SCRIPTION: It w as identined that Emergency Operating Pacedure (EOP) Plant Temporary Mah6 cation (PTM) 21 wuuld not u ork as intended Therefore, this procedure re ision changed the PTM from a fuse removal to a jumper installation to accomplish the same goal of oveniding an oft gas isolation signal uhen it is required to be open for Altemate Emergency Depressurization per the EOPs.
SAFETY ANAL.YSli The revised Emergency Sepport Procedure (ESP) conforms to the intent of Revismn 3 of the CNS Emergency Pncedure Guideline (lipo) and Revision 4 of the lloiling Water Reactor Owners Gmup '
EPO The implementation of the te isal ESP wil! not mothfy the design basis of the plant and will only be nul after plant omhtims have reacial EOP entry cmhtmns. No operator actions in the revised ESP conflict with the actions assumed in the analysis of postulated events and design basis accidents in the USAR. Therefore, the re isal ESP cannot increase the probability of occurrence or consequences of a plant ewnt or malfunctim of equipment in.;mnant to safety, Since the ESPs are intended for use dunnit t!xwe plant conditions which are beyund those analyni or required to be included in the plant licensed design basis, the question of a reduced margin of safety is not meaningful.
Procalure Chance Reauest iPCR) 5 8 6 (Revision 4)
TITI.E; Reactor Pressure Vessel (RPV) Floating Systems (Table 6)
DESCRIPTION: This revisioniturporatalchanges from Design Change 94 332, which changed the normal position of the Residualllent Renxwal(RIIR) minimum flow valves frorn ekwed to open The procedure sequence was revised to attempt to use low Pressure Coolant injection (LPCI) A and then LPCI 11 rather than attempt to hne up both systems concurrently. Emergency Operating Procedure (EOP) Plant Temporary 151-
Mahfcatims wuc alled to alku reuxte clostae of the RIIR 1leat Exchanger Senice Water (SW) outlet vahrs aner de SW ikuster Pumps have been started rather than manual ly closing the SW inlet manual I valves, in order to chminate the need to enter the Reactor fluil:Img. Appropriate steps were rnised to l clarify restoration of R1IR, Standby Liquid Control, Feedwater, and Condensate systems. This revision also changed the Nuclear lloiter Instrumentatim continuous backfill isolation valve used for RPV :
injectim with the Control Rod Drive system from CRD 15 to CRD 63 to utilize a valve that is easier to use and accent I sal-liTY ANAL,YSIN: CNS EOPs acommudate strategies for events beyond the licemed design basis of the plant. The NRC, in its Safety Hvaluation Report on the lkiiling Water Reactor Owucts Group Emergency Procedure Guidelines (EPUs) found the use of the limits specified in the EPGs end EOPs, rather than those f
specifial in the licawed design basis, acceptable during degraded conditions. 1he implementation of this revisa! Emergency Support Pmcedure (ESP) does not increase the probability of occurrence or comequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. This rnised !!SP cannot increase the probabihty of oc(utrence of any event analyzed in the USAR becauw the pmcedure will only be used after the event has commenced. The implementation of this revised ESP does not create a possibility for an accident or malfunction of a difTerent type than any prniously evaluatalin Oc USAR bconus the rnised ESP does not nudfy the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a 2 reduced margin of safety is not meaningful, Pncedure Chance Reauest (PCIO S H 7 (Revision 4)
TITLis: Pnmary Containment Fkwang Systems (PC/L 2) -
DESCRIPTION: This rnision to Emcrpency Operating Pmcedere (EOP) Sunport Pmeedure 5 8 7 inchaled the following changes: 1):dksi a step to secure the Reactor Core 190lation Cooling (RCIC) gland seal vacuum pump anct Oc RCIC turbine is seemed and a caution that Primary Containment oxygen levels may rise during '
RCIC plarx! seal vacuum pump operation,2) incorporated changes from Design Change 94 332 w hich changed de turmal positi(m of the Residual 1 lent Removal (R)IR) minimum flow valves fmm closed to open,3) tnimi pmcalare nequence to attempt to use low Pressure Coolant injection (LPCI) A and then LPCI 11 rather than attempt to line up both systems concurrently,4) added EOP PTMs to allow remote chwar of the RIIR Ilent Exchanga Seniec Water (SW) outlet valves aller the SW Umster Pumps have ,
been started rather than manually closing the SW intet manual vahrs, in order to climinate the need to cnter the Reactor iluildmg,5) revised sections uhich utilize gravity draining of the Condensate Storage Tanks with the Core Spray system to flood containment,6) revised steps to indicate that torus spray should be secured when drywcP r vssure drops below 2.0 psig, and 7) changed the Nuclear lloiler Imtrumentation continuous backfilt isolation vahc used for reactor pressure vessel injection with the Control Rod Drive system fmm CRD 15 to CRD-63 to use a valve that is casier to use and access.
SAFETY ANALYSIS: CNS EOPs accomnsdite strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety livaluation Report on the ikiiling Water Reactor Oumers Group Emergency Procedure Guidehnes (EPGs) fowx! the use of the lunits specified in the EPGs and E0Ps, rather than those specified in the licenel design basis, acceptable during degraded conditions. The implementation of this rnised limergency Support Pncedure (lisp) &ies not increase the pmbability of occurrence or comequences of an acci&nt or malfunction of equipment important to safety pre iously evaluated in the USAR. This revised lisp cannot increase the pmbabihty of occurrence of any event analyzed in the USAR because the immlure will mly be nul atkr the event has commenced The implementation of this reused ESP does not create a possibility for an accident or malfunction of a ditTerent type than any previously evaluated in the USAR because the revised ESP does not mahfy the operation or design basis of the plant For plant conditions which already exceed the liceming design basis, the question of a reduced -
- margin of safety is not meaningful.
152
Procedure ohnnee Request (PCR) $ M 14 (Revision 31 1111J!: Suppression Pool Make Up Systems DESCRIPTION. 1his toisim iruspratal various aht(sial and administrative changes, added panel kcations, and made changes to be consistent with the CNS and Emergency Operating Pmcedure (EOP) Writers Guide as identified by an !!OP Verification .nd Validation cfTort.
sal-1 TY ANAL,YSIS: CNS !! ops acastmulate strategies for events beyond the licensed design basis of the plant. %c NRC, in its Safety livaluation Reget on the Doiling Water Reactor Owners Group Ernergency Pmcedure Omdebres (liPos) fiiund the me of tle hmits specificd in the EPGs and liOPs, rather than those specified in de hccrtwd design basis, acceptable during degraded conathms The impkmentation of this raised Emergency Suppet Pmcedure (ESP) does not increase 6e probabihty of occunence or consequences of an accident <r malfunction of equipment imputant to anfety prniously evaluated in the USAR. This revised ESP cannot increase the probabihty of occurrence of any event analyzed in the USAR because the pnudure wendy tw nul aller the event has cornmencal The implementation of this rnised ESP does not create a possibility for an accident or malfunction of a different type than any previously '
evaluated in the USAR because the rnised ESP does not malify the operation or design basis of the plant For plant mnditions which already exceed the licensing design basis, the question of a reduced marpin of nafety is not meaningful.
Pmcedure Chance Reauest iPClO $ N 15 (Revision 3) lill.li: Altemate injection Subsystems (Failure to Scram)(Table 15)
DESCRIPTION. 1he pnadre sequerxx was toised to attempt to use Low Pressure Coolant injectmn (LPCI) A and then LPCI il rather than attempting to line up toth systems concurrently. Emergency Operadng Procedure (1 OP) Plant Temprary Mal ficatims were ackksi to allow remote closure of the Residual Ilent Removal '
iIcat lixchanga Service Water (SW) outlet valves aller the SW lhmster Pumps have been started rather than i.tanually chning the SW inlet manual valves this climinates the need to enter the Reactor fluilding.
This rnision also incorporated various admmistrative changes, added component kicatkus, and made other changes to be consistent with I!OP and CNS Writers Guide.
SAlliTY ANALYSIS. CNS EOPs acarnmodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Repirt on the lloiling Water Reactor Owners Group limergency Procedure (hiitk hnes (EPGs)(bund the use of tle hmits sjuafied in the EPGs and EOPs, rather than those specified in the beened design basis, acceptable during degraded conditions. The implementation of this rnised limer gency Support Procedure (ESP)(kies not increase the probabihty of occunence or consequences of an accident or malfunction of equipment important to safety prniously evaluated in the USAR. This se ised ESP cannot increase the probability of ocemrence of any event analyzed in the USAR because the pnuare will mly le nul alla the event has commenced.1hc implementation of this te ised ESP does not ercate a possibility for an accident or malfunction of a different type than any prniously evaluated in the USAR because the rnised ESP does not modtfy the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced ,
margin of safety is not meaningful.
Needure Chance Reauest (PCR) 5 818 (Revision 5)
TITLil:- Primary Containment Venting fbr Primar) Containment Pressure Limit (PCPL)
DESCRIPTION, 1his rnision to Emergency Operatmg Paredure (EOP) Support Procedure 5 8,18 included the following changer 1) clanfied steps to nxuutor Elevated Release Point (ERP) efiluent radiation versus monitoring only the ERP ellluent radiation nxutitors,2) deleted steps for placing PC-MO-232, PC AO 238, PC MO-233, and PC-AO 237 switches to Cl.OSE position, 3) provided instructions to close 00 AO-254 via the ofIgas timer contml switch versus the valve control switch,4) rnised steps to close PC MO 305 and PC-MO 306 first and then place their respective override switches to normal uhen
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i sectsing Oc torus or drywethuit path, and 5) made various other administrative changes and changes ,
- to conform to dw CNS and EOP Writers Guide.
SAFliTY ANAL.YSIS: CNS EOPs acumunudate strategies for events beyond the licensed design basis of the plant. The NRC, ,
in its Safety livaluation Repet on the lloiling Water Reactor Owners Group Emergency Pnwedure-Guidehres (EPGs) faud the me of tic hmits giecifwd in the EPGs and EOPs, rather than diose specified in tic busmi design basis, acceptable during degraded cond thms The implementation of this rnised limergency Suppet Pr cedure (ESP) &ies not increase the probability of occunence or consequences of an acciant or malfunctum of equipment important to safety prniously evaluated in the USAR. This rnised lihP canimt increase the probability of occurrence of any event analyzed in the USAR because tic pnmlure wdl mly le med after the event has commenced The implementation of this rnised ESP l
, does not create a possibihty for an accident or malfunction of a different type than any proiously evaluated in the USAR because the revised ESP does not nxdify the operation or design basis of the plant. For plant cmdithms which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful. .
Pnicedure Chnnee Rect. cst (PCR) $ 8 21 (Revision 3)
Tl11J!. Primary Containment Venting and liydmgen Control (l. css 1han Combustible 1.imits)
DliSCRIPTION: 1his rnision clanfied steps to morutor Elevated Release Point (ERP) efiluct't raantion, versus monitoring only the ERP elliuent radiation monitors. Steps were revised to require PC MO 305 and PC MO 306 to be closed before placing their isolation override switches to normal uhen securing the turus or dryweihuit path Other changes uere made to provide clanty, provide panel heations, i kntify swific instruments to use, and incorporate guidance from the Ernergency Operating Pmcedure (EOP) ,
and CNS Writers Guide.
SAIETY ANAL.YSIS: CNS HOPS accmurudate strategies for events beyond the I; censed design basis of the plant. The NRC, in its Safety !! valuation Repirt on the lloiling Water Reactor Owuers Omup Emergency Pnicedure Guidehnes (EPGs) found the me of the lunits specified in the EPGs and EOPs, rather than those specified in the hcermi design basis, acceptable during degraded conditions lhe implementathm of this revised limer gency Support Pmeedure (ESP) des not increase the pmbabihty of occunence or consequences of an accident a malfunction of equipment important to safety pre iously evaluated in the USAR. This te ised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the pnmhre wdl mly le med aller the event has conunenced The implementation of this revised ESP dies not create a possibility for un accident or malfunction of a difTerent type than any previously evaluated in the USAR because the re ised !!SP &c. w mmhfy the operation or design basis of the plant. For plant conditions uhich already exceed the licensing design basis, the question of a reduced snargin of safety is not meaningful hanlure Chance Reauest(}QD 6 CSCS 401 (Revision 11 TITlJ!. liigh Pressure Coolant injection (llPCI) and Reactor Core Isolat on Cooling (RCIC) Exhaust Line l Vacuum Ilreaker Inserme Testing Disassembly and Inspection Dl!SCRIPTION: 1his pnwhre w as raiul to memde instructions for the llPCI Vacuum lireakers w hich w cre modified by Design Change 95 101 fmm 2* lill check vah es to 3' swing check valves. The re ised instructions for the new I!PCI vahrs closely follow the intent of the original pmcedure. An as-found and as-left setpoint test was ad&d SAFETY ANALYSIS. The basic steps of this pmcedure remain unchanged for the llPCi check valves The addition of the setpoint test will not intnduce or chnge any accident scenarios or change the overall safety of the pnmlare 1his acti,ity kes not change the intent or limitations of the procedure, has no adverse clTect m any system interfaces, and will not cause any new radiological efTects System reliability will not be decreased, thus the probabihty of equipment malfunctam is not increased Performance of the re ised surwillance pnmhre has Oc same purpose as the earlier te ision m!iafaction of the open and closure
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I exercise requirements of die CNS IST Program.1his change does not create any new or unanalyzed j faihue nules 11s functum and reliabihty of the llPCI check valves are tot changed by performance of !
the re vimi pmcalare, ums this activity has no efTect on the margin of safety as defined in the basis for !
any 1cchnical Specification, j Procedure Chance Reauest (PCR16 l>P 601 (Revision 2) ;
1111.lk 125V/250V Station and Diesel l' ire Pump ilattery Weckly Check i i
DESCRIPTION.1his pamhac was reviel to change the defined requirements of the required test equipment to clarify l tle mpurement kir an in-ime 2 amp fuse. When voltage measurements are taken, the fuse intemal to the Digital Multimeter (DMM) provides rm pmtecti(m SAFir!Y ANAL,YSIS: A has of the applicable batteries is tot an accident initiator. In addition, the construction of the DMM. >
in anyunctum with tis w my it is used, will tot cause the batteries to become inoperable in any way. The ,
"canonumis being taken and tic ;unts at which de measurenrnts are taken canrot afTect plant systems
.ach that plant nvnt response is alTected The hication and metini of voltage measurement does not -
interface with any equipment that couki cause a malfunction of equipment. I!ven when die test equipment is not pmperly configured, and a measurement taken or test lead inadvertently connected to groutunegata/;outive source, bancry/ battery charger malfunctions cannot occur due to the intemal fuse protection provida!. The pmeess of taking voltage measurements carmot rcJuce the margin of safety which is controlled by other devices outside the scope of diis pnicedure.
Procedure Chance Reauests (PCRs1 &FP.301 (Revision 1115 FP.301 (Revision 116 FP 303 (Revision 1115 FP 602 (Revisim 1t 15 Flyo3 (Rnision Ii 15 FP 604 (Revisinn 116 FP 605 (Revision 11 !5 FP 606 (Revision 1115 FP 607 (Revisim 1115 FP 6nM iRevision 1115 FP 609 (knision 1115 FP 610 (Revision 1115 FP.611 (Revision 1115 FP 612 (Revision 1115 FP 613 (Hnision 1115 FP 614 (Revision 1115 FP 615 (Revision 1115 FP 616 (Revision 1115 FP 617 (Re ision 1i 15 FP 618 (Rnision 1115 FP.6I9 (Raision Ii 15 l'P 620 (Revision 11 15 FP.621 (RniSion 1115 FP 622 i (Revismn 1i 15 FP.624 (Revisim 111511%25 (Rnision 1115 FP 626 (Revision 1115 FP 627 (Raision 1115 FP 628 (Revision 116 FP 629 (Rnision Ii 15 FP.630 (Raision 1115 FP.631 (Revision 1115 FP 632 (Rnision 1115 FP 633 (Rninon 111511%34 (Revision i16 FP 635 (Rnision Ii 15 FP 637 (Rnision 1115 FP 63H (Revision 1115 FP 639 (Rnision 1115 FP.610 (Revisim 1115 Fl%41 (Revision 1115 FP 642 (Revision 1115 FP 643 (Revision 1115 FP 644 (Rn ision 1115 FP 645 (Revision 1115 FP 648 (Rnision 1)
TITI.lb Operations Pow er llhick Sprinkler System Testing (6 FP,301)
Operations Out-iluilding Sprinkler System Testing (15 FP301)
Operations Deluge and Pre Action Systems Testing (6 FP.303)
System Number 2 Flow Venfication Test (15 IT 602)
System Number 3 Flow Yeafcation Test (15.FP 603)
System Number 4 Flow Verification Test (1517.604)
System Number 5 Flow Verification Test (6.FP.605)
System Number 6 Flow Venfication Test (15 FP (06)
System Number 7 Flow Verification Test (15 FPlo7)
System Number N Fhiw Venfication Test (15 FP.608)
System Number 9 Flow Venfication Test (15.FPlo9)
System Number 10 Flow Venfication Test (15 FP.610)
System Number 1 i Flow Verification Test (15 FP 611)
System Number 12 Flow Verification Test (15 FP 612)
System Number 13 rkiw Ventication Test (15 FP.613)
System Number 14 Flow Venfication Test (15 l'P.614)
. System Number 15 Flow Venfiention Test (15.FP.615)
System Number 16 Flow Venfication Test (15 FP 616) -
System Number 17 Flow Venfication Test (15.FP.617)
Systerr. Number 18 Flow Venfication Test (15.FP.618)
System Number 19 Flow Venfication Test (15 FP.619) system Nun ber 20 Flow Verification Test (15 l'P.620) 155
I System Numler 21 Flow Verification Test (1511'.621)
System Numler 22 arul 23110w Verification Test (15 FP.622) '
System Number 24110w Verification lest (15 FP.624)
System Numtwr 25 Flow Venfication Test (15 FP 625)
System Number 26 Flow Verification Test (1511%26)
System Number 27 Flow Verification Test (15 FP 627)
System Number 2H ilow Verification Test (1511'.62H)
System Number 29 Ilow Verification Test (6 FP.629)
System Number 30 Fksw Venfication Test (15 F1%30)
Systern Numler 31 Fknv Verification Test (15 FP.631)
System Numler 32 Ilow Venfication Test (15 Il%32)
System Numter 33 Flow Verification Test (15 Fl%33)
System Number 34 Fhew Venfication Test (15.FP.634)
System Number 35 Flow Venfication Test (6 FP.635)
System Number 37110w Verification Test (15 FP.637)
System Number 38 Ilow Venfication Test (15 Fl%38)
Systern Numtwr 39 Flow Venfication Test (15.11%39)
System Number 40 Flow Venfication Test (15 FP.640)
System Number 41 Flow Verification Test (15 FP 64 I)
System Number 42 Flow Verificatiot Test (15f1%42)
System Ntunber 43 Flow Venficatiot Test (15 Fl%43)
System Number i A Flow Verification Test (15 i P.644)
System Nmnber 111 Flow Verification rest (16 Fl%45)
Outside Transfbnner Deluge Systerdlow Test (15FP 648)
DiMClllPTION. Tlee prnulares wac rniul to utilve (le electric fire pump for Dow testmg instead of the fite flushing pump.1his action was initiated due to the inabihty to apply Allowed Out of Senice Times ( AUTs) te non Technical Specification testing. These changes will ensure fire pumps remain in auto so that operability of the fire system is una!Tected during procedure perfonnance.
SAFl!TY ANALYSIS. The probability of a fire occunence is not increased by testing None of the other acidents postulated m the SAll are impacted by these changes and fire system activation is not an accident initiator. These changes only impact tie Fire Protection (FP) system and the availability of the electric fire pump for fire suppression activities in tic event of a fue. The ability to mitigate the consequences of a fire requires that w ater demand ihr numual aral aut anatic suppressi<m activities im met The electric fire pump is capable of supplying that danand arxl the minimal ackhtional danand of system testing concunently in the unlikely ewnt a fue moki occur dunng testing. The Safe Shutdown Analysis lleport demonstrates that the plant is capable of shutting down safely assaning the loss of all equipment in a given fire zone. Therefbre, the nevcquences of a fire would be unchanged na a result of these testing changes, Surveillance testing of non Technical Specification FP systems ensures that equipment malfunctions are minimized by datamtratmg equipment states of readiness. The additional usage of the electric fire gunp is negligible wlni comparn! to its ratal renix hfe. Defense in depth criteria and administrative controls ensure that impaired FP equipment is provided with adequate compensatory measures, thus providmg the needed backup to rnlace the consequences of equipment malfunctions. The FP system has been analyzed for intanctions with safety related equipment and the results are unchanged by these changes The electric fire pump capacity is suflicient during system testing to meet Technical Specification margins for suppression system flow demand, hose stream demand, and testing flow s. Use of the electric fire pump will leave the diesel driven fire pump operable and available, thus satisfymg Technical Specification margins for equipment in senice.
Procedure Change Request (PCR) 6 FP 20l fRevision 1)
TITI.lt Operations Cycling of Fire hiain Valves DESCRIPTION: This procedure was reviel to reflect Plant Temporary hiodification (PThi) 9M3, Disabling of Fire
!; Pump "D" Remote Stop. (PThi reinted separately). Vahrs associated with hiinor hiahfication L
156-l l
l
PacLage (MMP) 95 litt, Fire Pndection Main Tic-In for the Nov Tedmical Suppon lluilding, were also added to the procalurc= P4MP report d separately) This change also included infmmation fran Technical S;wci6 cation Interpretation (1 SI) 97 006 which established which vahes are "in the flow path" fa de papmes of satisfying surveillance and operability requirements.1he action oflubricating vahrs in this immlure has las deletal from the acceptance criteria. 1,ubrication is p rformal for case of operatkui and maintenance only afd des rud constitute acceptance criteria.
SAll!TY ANALYSIS: Changes to this procalute do txt increase 0 e probabihty of a fire or inadvertent fire system operation.
Changes ensure continued operabihty of fue system commments which are relied upm to nutigate the umaitmen of a plant nuit liarly idmtificathn of Fire Protecthm (FP) equipment malfunctions allows artneve actkos tole taken befwe the anninents are challengni by a fire event. Cycling of FP valves '
1 can mly cause two nuits to nocur faihre ofFP system activation or inadvertent system activation lloth of these ewnts have been previously evaluated. Cycling of FP valves can only induce failures in the ;
~
valves thernselves or prevent skmmtream system actuation. These malfunctions have been piniously evaluatol and aangotsotay incasures have bcen specified These changes ensure amtinued cc npliance with Tedinical Specification requirements for FP system surveillance and operabihty by incorpirating die ap;m priate Tcdriical Specifethm bases as described in TSI 97-006. Lubrication of valve parts is not a Technical Specification requirement.
Procedure Chne Reauest (PCR) 6 FP 30i fResision 2)
TITLlh Operaum Out iluildmg Sprinkler System Testing DI! SCRIPT 10N: Varkus dianges were made to this pnredure, includmg the following: 1) addition of step to restore fire punp to rummal hrcup in the nuit ola farc. 2) rniske of steps to minimi/c the pump running at shutofT presstre and km ihms,3) deletion of step to start the jockey fire pump, and 4) additim of steps to keep the Comrol Ihxnn infonned of surveillance status.
SAFl!TY ANALYSIS- ~!his activity will ruit dmnge the state or function of any safety related systems, structures, or components Hrd Will th4 allCr any o[ die knplits W amu!nplioris [or previously Cvaluated accidents This PCR dies not change the functum, perfonnance, or integlity of any lx undaries with which safety relatal systems fonn or support the primary protective bamers on which the cmsequences of an accident are based 1his dmnge does rat aher the design basis for the fire pumps nor acs it adversely a!Tect initiatmg sequences or starting set; mints of the fire pumps This activity will ensure reliability of the FP system to functian uhen nec&d to mitigate the consequences of a rue. The consequences of the Fire Suppression liffects Analysis nie bouahng ard unchanged by this activity. This change &cs not alTect equipment imputant to safety and creates no new failure mo len. The margm of safety w hich is estabhshed by the design and performance of safety related sptems will not be redacal by this actaity.
Pnmture Chance Request iPCR16 FP 303 (Revisim 2)
TITI.ll: Operathms Deluge and Pre-Action Systems Testing DliSCRIPTION. Steps for the manual start of FP P.I! were deleted from this procedure twcause this pump will start autonatically when system pessure is dropped below die pump's s,etpoint Steps for starting the jockey fire paup war also deleted locause assuring 95 tem pressure is almve the auto start r,etpoints of the fu e pumps pra:hden the nmi for another fire pump to be started. In addithm, othcr non-intent dianges were incorporated in this resision.
SAFl!TY '
ANALYSIS: This actnity will not change the state or function of any safety related systems, structures, or compments ami wdl not alta any of the inputs a autanptions for prniously evaluated accidents. This PCR ars not change de functim, performance, or integrity of any Imundaries with which safety related systems form or support the primary protective barriers on uhich the consequences of an accide it are based. This change does txt alter the design basis for the nre pumps nor does it adversely atTect initiating sequences
, or starting setpoints of the fire pumps This activity will ensure reliability of the FP system to function wben needed to mitigate the consequences of a fire. The consequences of the Fire suppressim ElTects Analysis are lundmg an! unchanged by this activity. This change does not afTect estupment imputant
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1 i
i to aa'ity arwi neates no new failure trxdes.1he f. srgin o." safety which is established by the design and
{cformance of safety related systems ws!! not be reduced by this activity.
Procedure Channe iteouest (PCR16 Fl'307 fltevision 11 TITili: Operations Out Duildmg Spnnkler System Testing DliSCRIPT10N. This pnicedure was resised to delete the reference to an alann device wMch &cs not exist in order to Iring procedtre into u rnpliara with tic plant configuration of the W9: Warehouse. A change was also made to establish a limit to allow for minor leakage past the check vahc on the Fire Department unrx:ction in Oc Technical Suppst liutidmg. leakage paa e e check valve accommodates valve design of netal to metal sentmp surfaces aral is supported by the design basis of the fire jockey pump which is supposed to rnake up for mirur system leakages.
SAlliTY ANALYSIS: 11s posibihty of a fire event is not increased because no new ignition sources have been added.1hesc changes alTect testing of fire suppression systems and are not accident initiators. Improved testing 1 improves the ability of fire suppression systems to mitigate a fire. The probability of equipment l malfurxton is rud increased twcause the changes are in accontance with applicable cales and starxlards i and are nondestructive in nature. The testing activities do aot introduce any new failure modes. The changes made are to systems rmt crahted in the Technical Specifications 1he margin of safety for the fire water system is unalTected Pandure Chance Reauest (PCit) 6 FP307 (Revision 2)
TITIJi Operations Out lluildmg Sprinkler System Testing Dl! SCRIPT 10N: Varxo clutnpes were made to this procedure, inchidmp the following: 1) addition of step to retum the ,
fire punp fimv to its nonnat hneup in the event of a firc,2) addition / revision of steps to minimize pump l running at sliuto!T head flows and shuto!T pressure flows, and 3) add tion of steps to enter 1.imiting Condition for Operation (1,CO) 31511 and to place FP-P l! in pull-to-hick (arnt subsequent exiting of 1.CO arni retum of pump to Auto),
SAFl!TY ANALYSIS: 1his actnity will tot change de state or function of any safety related systems, structures, or components arxl will not alter any of de inputs ir asstsuptions for previously evaluated accidents This PCR does not change the functast, perfonnance, or integnty of any toundaries with which safety related systems form or support the primary protective barriers on uhich the consequences of en accident are based This change des rot aher tle design basis for the fire pumps nor does it adversely affect initiating sequences or starting setpoints of the fire plunps lhis activity will ensure reliability of the FP system to function when nenied to mitigate the consequences of a fire. The consequences of the Fire Suppression lificcts Analysis are tou:. ' ng arxl unchanged by this actnity. This chage des not alTect equipment important to safety and creates no new failure nxdes. The margin of safety which is established by the design and performance of safety related systems will tot be reduced by this activity.
Prtre dure Change Reauest (PCR16 FP 604 (Revision 2)
TITIJI: Fire ihr Annuallhamination Dl! SCRIPT 10N: The maximum lher pap for aors 11109 and lit 10 w as increased to one inch. This increase brings neceptance critena in ime with fire test paremeters. These doors do not have thresholds and were inadvertently given gap cnteria for doors with thresholds.
SAFliTY ANAL.YSIS: The test avs not inercase the pmsibihty of fire occurrence and no accident initiators are mtroduced by perfttmaru of tie test.11e pntedure ensures operabihty of credited fire bamer fire awrs by ensuring arrect du gap entesia is apphal Consequences of a fire event remain as analyzed in the Fire llazards Analysis and Safe Shutdown Analysis Report. Testing raluecs the probability of door failure.
Perftvuaru of the surwillance does rxt intnsha any new failure modes or accident precuisors. Failure 3-
of fire acuss has been analymi and enmpensatory measures specified in the Technical S;weifications fire bariier operabihty arx!Iatmgs are nuuntairni ard the margin of safety in the Technical Specifications remains unchanged hocedtn e Change Reauest (PCR) 6 l'P 609 (Resision 2) 111 111: Fire Protecthe System Fh>w Test DESCRIPTION: Pnvedural steps were resi9ed to minimize low flow operation of FP P !! by startmg the pump with the enabh4aliknv path as pre iously approwd in Provedure 6 FP.101. Numerous addithmal administrative changes u ere also made to this procedure. ;
sal 1?TY ANAL.YSIS: Perfienuirus of a sunullance does rxt intaduce anyix w accident initiators or precursort Existing plant equipnuit is ofcatal w1 dun specifn! design parameters to verify functionahty. The pnicedure changes ,
enwuc that mutmtun availabihty of accdent nutigation equipment is rnaintained The changes minit,'i/c de probabihty of eqtulmient fadure as a result of testing by minimihng low /no flow operation of the ft.i pump.1hc uncapuices of a failure of de ekx4nc fire pump remain unchanged. A redurxiant fire pump has tuui panidal to aanpenote he de anJy/ed failure of the electric fire pump.1hese changes do not impact any equipment beymd the Fire Protetion systen. 11e surveillance test requirement establishedmatutains the [ofumuince capabihty of the purnp fnun which the rnargin of safety is derived.
Procedure Chance Reauest f PCit) 6 FP 610 (Revision 3)
TITIJ!: Yard Ilyd ant Flow Check DESCRIPTION Vanians changes woe made to this pnudure,inchalete ine following: 1) addithin of a step to return die fire pump flow to its normal hneup in the event of a fire,2) addition of steps to start fire pump autonatically by pressure drop in header and to recirculate flow back to the storage tank,3) addition of caution statenuit prhir to ch> sing FP 120MV to indicate that closing the valve will cause the fire pump kirun at shutoff head pressure and flow,4) addition of steps for restoring FP-P-E lineup after using the aceirculatmg lineup to the storage tank,5) aJJitam of step to verify water level in storage tank, and
- 6) addition of step to fill water stor age tank per Pmeedure 2.2.30, if needed SAFl!TY ANAL,YSIS: This actisity will ruit dutnge die state or functhm of any safety related systems, structures, or components arxl will ruit alter any of the input.uv assumptions fbr pre iously evaluated accidents 1his PCR does not duinpe de functim, perfbrmance, or integnty of any tomdaries with which safety related systems fbrm or support the pnmary protect vc barricts on which the consequences of an accident are based. This change does not aher de design basis fbr the fire pumps nor does it adversely alTect initiatmg sequences or starting setpoints of the fire pumps This ac ity will ensure reliabdity of the FP system to function ului needed to mitigate the consequences of a ure.1he consequences of the Fire Suppression EITects Arudysis are tunnhng and unchanged by this activity. This change does not affect equipment important to safety ard creates no new fadure nudes The margin of safety which is estabbshed by the design and perfonnance of safety related systems will not be reduced by this actisity.
Pmeedure Chance Reauest (PCR) 61IPCI 103 (Revision 3)
Procedure Chance Reouest iPCR) 14 610 (Revision 6)
'TITilh liigh Pressure Coolant injection (IIPCI) Insemce Testing and Quarterly Test Mode Surveillance L Operation (6 ilPCI.103) 1IPCI Stop '/alve lastrumentation Cahbration and Test Setup DESCRIPTION: In respmse to a concern that installation of test equipment to perform transient recordmg ofIIPCI stop valve position aral balance chamber pressure could impact the functionality of liPCI dunng testing, changes were made to the subject procedures Installation of this test equipment was evaluated per the folkmang safety evahtatkui in conjunction with the safety evaluation ihr test equipment installation, O e l Allowed Out-of Senice Time (AOT) start step was moved befive the step that aligns the test 1
159
imtnanentatxo in Pnonlure 6 IIPCI 103,1his ensures that the pressure trammitter is aligned with the pessure lunlary witlun tic AUT pin! When pykeming surveillunce testing, Procedure 14 6.10 will install and remove the equipment Procedure 6 IIPCI.103 will configure the equipment durine the surveillance test SAFl:TY ANAL,YSIS: 1his evaluatim a<kiresws t e imtallation of the temporary ecorder and the mechanical mntwetion to the r, top valve pmtion arm while the IIPCI system is operable. These activities do not have any tie or connection to initiators of accidents analyrnl in the dAlt, The recorder is physically and electiically -
ietated fnen the 1IPCI system u hich is usal to mitigate accidents presiously analyzed in the SAR. The anechanical connectmn funn the valve stem position. arm to the position Imtentiometer ass make a physical umnection to a compment that is relied upm to mitigate accidents analyzed in the SAll, i14 mort, successful past performance of Pmcedure 61IPCI.103 indicates that the adational resistance to opening the 1IPCI stop vahr irsluccd by the puition potentiometer ars tot prevent the 1IPCI system fnen perfonmng its safety function 1his activity does not have the potential to impact the performance of other systems or equipment impntant to safety.1he installation of the temporary recorder favolves electrical connections to existing non safety relatal equipment nhich is cicetrically islated from the irutiatim arx! untrol furctims of the iIPCI system and &cs not have the cap obility of creating a afferent tge of accident or equiprnent malfunction 1his activity does not alTect a reduce the margin of safety as defined in the basis for any Technical Specification Pnxcdure Cb9pc Request (PCR) 61IPCI 104 fitevision 21 TITI.li iligh Pressme Coolant Injection (llPCI) Cycle (150 PSIO) Test Male Surveillance Operation DiiScillPTION: Various changes were male to this procalure to enhance hurmin performance aspects, improve procedure uwability, aral pmvide consistency with similar pnicedures. Steps conceming Radiological Protection (RP):mtdication of1IPCI nms were revised to climinate confusion regarding w hat areas require pnting as high tal areas while iIPCI is operating. RP is now routinely involved in briefs awxiated with lIPCI nos and umtrols access to the idmtilini areas via a standard Special Work Permit ihr iIPCl runs Steps were added to nalress a commit nent made by 1.icensec !? vent Iteport H5 008 to check the governor system by lowenng flow by 500 rpm and venfymg pump runback and flow stability during the iIPCI operability test at 150 psig. When previous pncedure 6 3.3 I was spht into several pnuslures, the requirement became part of the high pressure test and not part of the 150 psig test independent venfication stqu were mLkst to rnake tlus procedure e msistent with the requirements of Procalute 0 31,
!!qmpment Status Control SAlliTY ANAL.YSIS. This proenture is revised to impmvc overall meabihty of the pmenlute. Since the pacedure will be casier to use, the probability of the pnoxlure causing a plant event has been reduced The probability of equipment malfunction or the consequences of a plant event or equipment malfunction are not irneaul smcc the !IPCI system will still be operated in accordance with USAR desenptions and other iIPCI proentures Required testing steps which prevent iIPCI from perfonning its safety function are prfinnni within appnnut Alkmul Out-of& nice Times (AOTs) w hich are lumded by the Technical Spnification action statement ihr the llPCI system. A ditTerent type of plant event or equipment malfunction is not being created since the procedure is not addmg new components or renuning any components desenbal in the SAR. The changes to the pmeedure de not alTect any analysis apphcable to plant events caused by llPCI system mimperation. The pmcedure still satislies all Technical Sovitication reqmroments that it is interxkxl to satisfy and, therefore, does not reduce the margin of safety in any Technical Specification basis Permimel safety is not rnluced since adequate controls are in place to ensure personnel in areas alTected by llPCI system operation am aw are of the potential for higher f tediation leven
. l(6
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I Procedure Chanc_e Rea_uest (PCR) 6 IIV 101 (Revision 1)
TITI li: Control Room Ventilation DESCRIPTION- 1his change im olves moving lascnice Testing (IST) valve exercise testing fnun Procedure 6 IIV.105 to Pnmlurt 61IV.101.1hc perfmnance frequency of Pncedure 61IV.105 was educed, therefore, IST testing of valves llV 270AV, ilV 271 AV, and !!V-272AV was moved to Pncedure 6 IIV.101 to maintain the desired IST testmg frequency. The metlnl of testing is rmt being altered, but only moved j to another testing procalure. Trus change also added steps for tracking Control Room Emergency Fan run time. i SAlliTY i
ANALYSIS: The Control Runn Emergency Filter System (CREFS)is not an initiator for any desir.n basis accident.
No plant hardware changes are pmpad The IST testing steps b,<ing added to 6 IIV.101 are the same steps that were in 61IV.105. lhe new steps being added to record the initial and fmal readings of the Cmtrol Rasn Enugency Hypass Fan Elapel Time Meter are to ensure CREFS dws not nui more than 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> between performances of Procedure 6.llV.104 (filter testing) as required by recimical Specificatkus 'Iherefore, the cmsequeras of an accident, specifically Control Room operator dose, are not increani 1he ability of CRiiFS to maintain the Control Room at a positive pressure is not affected by this change. !!quipment is operated as presentul by plant pacedures within its design capabilities.
No setpoint changes are proposed and the margin of safety is not reduced-Procedure Chance Reauest (PCR) 6 IIV 103 (Revision 21 liltli: Control Room timerpency Fan Filter Train Differential Pressure Test DESCRIPTION: 1his pncalure was reviel to meet new acwptance entena established per Design Change (DC) 93 257 which imtalled a larger fan in the Control Ravn Envelope Filtration System This DC increased the flow rate from 34I cfm to 900 cfm
- 10% An On The-Spot-Change to the DC revised portions of this procedure, but failed to make all the necessary changes.
SAFliTY ANALYSIS: This change bnngs the acceptance critena mto compliance with the revised Technical Specifications.
1hc test pnxess and equipment conditions are not being changed The test instniments installed ck) not rahice the abihty of the system to provide fdteral air to the Control Room. No changes are being made to the way the plant equipment is being manipulated during the performance of the test The corrected pucess will ensure that actual Technical Specification values are used. Therefore, the margin of safety as defined in the basis lbr any Technical Specification is not reduced Procedure Chance Recuest (PCR) 61IV 105 (Revimon 51 llT1.li: Control Room Envelope Pressurization Test DiiSCRIPTiON: The frequency of procedure perfbmiance was changed from quarterly to an 18 month frequency. The District had made a previous comnutment to the NRC to perfonn this te ting quarterly. Since that commitment u as made, the system has undergone several modifications. This conunitment has been reviel(refeiera N1.S960010 fnun NPPD to the NRC) to now perform testing at the frequency required by the apphenble Technical Specification surveillance requirement. This pmcedure change reflects the revised conunitment Quarterly inservice Testing steps previously contained in Pmcalure 6 llV.105 wcre moved to Procedure 61IV.101.
-SAFETY ANALYSIS: 1he Control Room Emergency Filter System (CRiiFS) is not an initiator for any design basis accident.
No plant hardware changes are propwed and no new processes are created. Extensive test data from perftenung this pnmhire m a quartaly basis has been evaluated it shows there is considerable margin between the administrative CREFS acceptance entena of 1/8 inch Wg and actual test results of appnnimately 0 3 urh Wg Testmg irwhcates the system does not degrade such that the margin of safety is rahm!in 18 rnoths. Beiefate, the consequences of an accident, specifically Control Room Operator dose, previously evaluated in the SAR is not increased. The abihty of CREFS to maintam the Control 161-
Raan at a putive penum is not afTected by this procedure change. No setpoint changes are promised 1hc cturent Technical Specifications state that CRi!FS pressuritation tes:mg shall be done once per operating cycle. CNS is changmg the pocedure frequency to agree with the cunent Tecimical Spwilicatmns.
I'm;nigy Chance Reauest (PCR) 6 MISC 502 (Revishin 11 Trftli ASMll Class I System 1.cakage Test DliSCRIP1 ION: 1his pmxture resisim irmptates the requirements of the 3rd Ten Year Interval insenice Instwetion Program, clarifies the test boundaries, incorporates lessms teamed dunn,t the pressure tests performed l by Special Pnicedures 95128 and 5086, and allows the test to be perfumed at reactor coolant temperatures above 212 *F by ensuring required systems are operable prior to exccalmg 212*F.
SAFETY ANALYSIS: Ilydrostatic testing is not a precurer to any accident desenbed in the IJSAR.1hc pressure test is performed in accordance with the temperature-pressure limits specified in the iJSAR and Technical Speetticatims fw the reactor vessel, and at pressures well below the design pressures of attached piping pessmi/cd during de course of de test.11e Main Steam relief and safety valves are available to pruside pretxe reliefin the unlikely event that all odo methnis of ..1 essure control are lost. The worst possible consequence of the test would be a leak in the reactor coolant pressure tvundary. Any through yll ,
leakage wmla be lunksi by de accident analysis. Reactor coolant temperature shall not excred 2WF to ensure that the Shutdown Coohng System can be returned to senice aller vessel depressurization.
Although the Reactor Recirculation pump speed limiter 20% feedwater flow interkick is defeated. l recirculation pmnp operation will not be adversely alTected Test temperatures are much lower than the operating temperature wine Recirculation pump casitation may occur if adequate feedwater flow is not available. Iligh Pressure Coolant injection and Reactor Core Isolation Cooling low steam pressure isolatim intakicks are defeatal to penrut draining water from the steam imes folkming depressuritatiort Since these interkicks are defeated while depressurized and in cold shutdown, the probability of equipnent malfunct on is not increased Isolating the head vent piping from Main Steam Line C during the pressme test will not have an adverse etTect on the vessel because non-condensible gas paiduction will be neghgible dunng reactor shutdoutt In ackhtion, de Main Steam Lines are isolated at the outboard Main Steam Isolation Vaht; duefme, the constant vent to Main Steam Line C will have no cfTect dunng the pressure test Afrated systems are operatal widun design limits during the test. The pump and valve logic temimrarily defeated by this procedure does not impact the operabihty of the limergency Core Cmhng Systems rajuired for this plant condition 1hc pmcedure requires that all applicable Technical Spwilication requirements be met; tlacfore, time is no reduction in the margm of safety as defined in the basis of any Technical Specification-Procedure Chance Reouest f PCR16 PC 201 (Revision 4 Cl)
TITLI.!: Primary Contairunent isolation Power Operated Valve Operability and Closure Timing Test DlISCRIPTION: This procalare was revised to change the Insenice Testing (IST) Operability 1.imit for valves RWCU MOV MOl$ and RWCU MOV-MOl8 from 28 seconds to 30 seconds. An Engineering Judgement (l!J %106) w as perfonned which evaluated the etTects ofincreasing the allowable closure time of these valves SAFHTY ANALYSIS: 1he stroke time of the Reacta Water Cleanup (RWCU) isolation valves is not a precursor to any accident or equipment malfunction analyzed in the SAR. The stroke time is established to mitigate the consequences of a postulated Iligh Energy Line llreak (llElli) The increase in stroke time will not cause any cuential equipment to fail based on !!J %106. The 10CFR100 otTsite dose is basal on the 60 seconds for containment isolation specif:ed in the Technical Specifications. 'the dose due to I(ELil islunkst by the Technical Specification basis The increase in stroke time will cause a shght increase in the resulting temperatures and pressures due to a llELit This is bounded by the EQ analysis for cuenual eqtupnni Due to the comervatism m the structural iIELil analysim, the increase in stroke time will have no net e!Tect on the pressures in the pump roont The stroke time used for the analysts for the 162
i heat ext. hanger rasn w as 33 mmds which is greata than the stroke time used in this change. Therefore, f the equipment required fit safe shutdan remains operable folkming the RWCU lIElJk The change in simke time des rr4 change ',he design or operating characteristics of the valves The !IElll analysis stroke time of 30 seconds is not part of the basis for the Technical Specification on the RWCU stroke time.
Procedure Chance Reouest (PCR) 6 PC 401 (Revision 2)
T111Ji: Drywell and Torus Surfaces and structural Elements laspection ,
DESCRIPTION: Sections of this procalure have been expanded to include the inspectmn of all reasonably accessible painted surfaces within the drywell and torus envelope. Examples of additional painted items to bc ,
inspected by this procedure include the biological shield liner, floors, gratings, and hangers. This pnmlure was expardal as a result of the identification of several areas of peeling paint between drywell elevations 950' and 963',
SAFETY ANAL.YSIS: This procedure expands the inspection of the drywell coating (s) during plant shutdowit It is not a potential accident initiator and will not result in an increase of offsite (kise, nor a failure of the Primary Cmtamment system it does not alter any oper ational or accident parameters, inchiding emimnmental, of any plant equipment. The credible failure innie of the coatings inside the dr)well will remain unchangal as a resdt of the enharmiinvectmn. Enhancing the inspection requirements will not reduce the design margins of the Pnmary Cmtainnut system lherefore, this activity is incapable of decreasing any margins of safety of any plant equipment.
Procedure Chance Reauest (PCR) 6 PC 501 (Resision 3) 1111J1: Primary Containment lacal I cak Rate Tests DESCRIPTION The changes to this pmecdure result in enhancement and clarification of administrative steps, add ficubihty to better account for individual component testing, clarify interfacing truer ice Testing (IST) ,
acceptance critena step in accordance with IST Program requirements, and identify additional test equipment to ensure that testing delays are minimized.
SAFETY ,
ANAL,YSIS. This change aos rat cause the pnmlure to be performed during a difTerent nxxle of operation than was -
previously authorized in addition, containment leak rate testing is not a precursor to any desilm basis i accident or transient 1his revision ars not change the test pressure or Numdary; as such, therbvill be no deletenous etreet upon the conteWment boundary or to any hardware which may be contained within the test boundary. In addition, sti dard industry accepted test metluxis are used such that the ability to ;
quantify containment leakage and to assess pnmary emtainment integrity are not adversely a!Tected. This change does not atreet the Primary Contairunent leakage acceptance critena specified in the Technical Specifications or the ability to determine if that critena is satisfied Procedure Chance Recuest iPCR) 6 PC 501 (Revision 4)
Procedure Chance Reauest (PCR16 PC 518 (Revision 3 C2)
TITLE Primary Containment local lxak Rate Tests (6 PC.501)
Residual 1 lent Removal (RilR) local leak Rate Tests (6 PC.518)
DESCRIPTION; MP 94 072 RI1R loop "A" Drain Path to Radwaste, installed a new 4* drain line to enable either kop of RllR to be drained to the waste surge tank without atTecting the other hxip of RIIR. Procedure 6 PC.501 requiral a change because the heal leak rate test vohune for RllR MOV-MO67 increased as a result of the piping mahlication Pmeedure 6 PC.518 required a change b,cause the recommended clearance order toperfunn the heal leak rate test for valves RllR MOV MO67 and RilR MOV.MO57 had to be revised due to valve lineup changes i
163
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SAll!TY ANALYSIS: 1his duinge does rut caue the pnmlare to be performal during a difTacnt mode of operation than was 15niously autivrim!, in adhtan, containment leakage rate testmg is tot a precutar to any design basis accident or transient. The test hiundary is changed, but there will be no deletenous effects upm the antainment hiurklary because the attire test haniary nwets or exceeds 1.c 5 pressure retaining capability required for tic containruent. Standard industry accepted test mettuis are used to quantify illIR containment isolation val e leakage; there is no change to the methnis usal to quantify such leakage.
Positive uutmls fit restaathm of tanprary changes to system imeups are provided by the requirements !
of Pnuxlare 0 9. Tagging Orders This pnuslure change does not affect the Primary Containment leakage rate nemtunce enteria spwified in tis Technical Specifications or the ability to determine if that nurptarwe entain is satisfied 1herefiire, the margm of safety as defined in the basis for any Technical Specil'ncation is tot reduced Procedure Chance Reauest (PCR) 6 PC 503 f Resision 11 TITLl!: Drywell to Suppression Cham'er Leakage Test DliSCitipTION: 1hc section of this procedure for testing when containment integrity is required was revised The initial '
prenure ditracutial fi>r this section was changed fnun 1.0 psi t00.5 psi due to a conflict with the initial prensure of 0.75 psig assumed fin the Design liasis Accident (DilA) loss of Coolant Accident (LOCA) analysis.1he acceptance criteria fbr the test when containment integrity is required was also rnised to account ibt the lower initial pressure. These changes will ensure that the contaitunent pressure in the ewnt of a DilA LOCA will be maintainal below the containment design pressure. Steps referring to the Sutorbit compicswus were deleted since they are no longer required with the lower test pressure.
SAFl!TY ANALYSIS: No axident precurses are affccted by this change. The change is being made to ensure that containment integnty will be maminined m the event of a DilA LOCA. Therefore, the previously evaluated consequences are unaffected by this change. The integrity of the test is unafrected by the lower test pressure smcc the acceptance entena has beesulenunstrated to be equivalent to the existing. This change does not involve the operation of any equipment important to safety and no new failure modes are introduced Deleting the use cf the Sutorbilt compressors will have no etTeet on safety since these components perform no safety function and are no longer required ihr performance of the test. The auxcptarxx cnteria fit the kiwer test pressure wdl ensure that the dryu cil to suppression chamber leakage is less than the equivalent leakage through a one inch diameter orifice, as required by Technical Specification 3.7.A 4.c.
Procedure Chance Reauest (PCR) 6 PC '.07 (Revision 1)
TITLII: lilectrical Penetrations local leak Rate Tests DliSCRIPTION: The changes made to this pmenfure are enhancements to better facihtate the testing of electncal penetratkus which ensure the testing is dor.c in accordance with 10CFR50 Appen<hx J. An alternative test metint was a& led to proside 11cxibihty in the event that sigmficant leakage is encountered.
SAll!TY ANALYSIS: 1his change dies not cause the pnmlure to be performed dunng a ddTerent mode of operation than was pre iously authortml in a&htion, containment leak rate testing is not a precursor to any design basis acciant a transient. This rnision does not change the test pressure or boundary; as such, there will l ie tu deleterious etTcet upon the containment humdary or to any hardware w hich may be contained within the test boundary. In adhtion, standard industry. accepted test inctin!s are used such that the abilhy to quantify electrical penetration leakage and its impact upon Primary Containment integrity are not adversely allected This change does not alTect the Pnmary Containment leakage acceptance uiteria specified in the Technical Specifications or the abihty to determine if that criteria is satisfied
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Pmeedure Channe Reaugt iPCR16 PC 508 (Revisum 1) llTIJi Control Rtd Drive (CRD)lxcalleak Rate Tests P
DisSCRIPTION: 1hc changes to this pncedure result in clarification arul enhancement of administrative steps, and impnne clearance mkr rmnunendations. The Recornmendal Clearance Onler was revised to include i
additional tags to specify the control switches requis ed as well as keeping the tegs on the valves in the field Correct vent paths identified during the outage were also included. i sal:1!TY ANAIMSIS: 1his change <kes nc ' cause the proculute to be perfamed during a different rmde of operation than was presiously auta orid in addition, cmtamment leak rate testing is not a pecursor to any design basis accident or transia 1his revision &cs rx4 change the test pressure or lutadary; as such, there will be no deletaious effect ugu de contairunent tourglary or to any compments which may be contained within tic test bourxiary, In nahtion, standard industry accepted test methods am used such that the ability to quautify CRD Primary Containment isolation valve leakage and its impac t upm Primary Containment i integrity are not adversely affected Positive controls for restoration of',emporary changes to system lurups are providal by imposition of the requirements of Pncedure 0.9.1 agging Orders. 1his change does not afTect the Primary Containment leakage acceptance critens specified in the Technical ,
Specifications or the abihty to determine if that enteria is satisfied Procedure Chance Reauest (PCR) 6 PC,ll,4 (Revision 1) e
- TITIJD Post Accident Sarnpling Systcm (PASS)localleak Rate Tests Dl! SCRIPT 10N: 1hc changes to this pn cedure result in enhancement and clarification of administrative steps, and impnwe clearance order recommendations to include additional barriers to mispositiming components during testing 1hc Rmunmended Clearuce Order was revised to include cornponent control switches and local component tagging requirements _
SAlzi!TY ANAINSIS: 1his change aes rx4 cause the procedure to be perfornal during a dif ferent mode of operation than was previously authorized In nahtion, containment leak rate testing is not a precursor to any design basis nocia nt or transient. This revision (kes not change the test pressure or boundary; as such. there will be ruuleletaious c&ct upm the omtamment tourxlary a to any comporwnts u hich may be contained within the test boundary. In addition, standard industry-accepted test metlnis are used such that the abiht) to quantify PASS cmtainment isolation valve leakage and its impact upon Primary Containment integnty arc ant adversely rJTected Positive controls for restoration of tentporary changes to system lineups are provided by imposition of the requirements of Procedure 0.9, Ta gging Orders. This change does not afImt the i rimary Contamment Icalage acceptance critena specified in the Technical Specifications or the abihty to detennine if that enteria is satisfied Pnu _ lure Chance Recuest (PCR16 PC 518 (Revision 3)
TITL11: Residual I lent Removal (RllR) local lxo, Rate Tests DliSCRIPTION. Changes made to this procedure included 1) clarifications arxl enhancements of administrative steps,
- 2) addition of provisions fm individual component testing to increase flexibility given the possible different system conditions uhen testmg utille perfonned, and 3) revision of the reconunenJed clearance order to include component breakers, control switches, and hical component tagging requirements.
sal'liTY -
ANAL YSIS: This change &cs not cause the procedure to be perfonned during a difTerent mode of operation than prevmusly authaired in addition, containment leak rate testing is not a precursor to any design basis accident or transient. The revision to the procedure &cs not change the test pressure or boundary, as such, there will be no deletenous efTect upon the containment boundary or any components which may be containal within the test boundary. In aabtxm, standard industry accepted test methods are used such that the ability to quantify RilR contaimnent isolation valve leakage and its impact upon Primary Contamment integrity are not adversely alTected. Positive controls fm restoration of temporary changes 16k
I to system hneups are provided by impositmn of requirements of Proc 4 dure 0 9. Tagging Orders. This ,
change does amt alTect the Primary Containment leakage acceptance criteria specified in the Tecimical i Specifications or the abihty to detenmne if that criteria is satisfini. i Pmecaue Chance Reauest (PCf!) 6 PC.519 (Revision 2)
TITil!: Reactor Core Isolation Cooling (RCIC) local leak Rate Tests DiiSCRIPTION: The changes to this procedure result in enhancement and clarification of administrative steps, and improve clearance order recommendations to include additional baniers to mispositioning cornponents during testing 11e Remnnmla! Clearance Order w as resised to incts le compment breakers, control switches, aral hical connponent tagging reqmrements.
SAll!TY ANAL,YSIS: This change ass not cause the pmcedure to be perfonned during a di!Terent imde of operation than previously authoriecd in addition, containment leak rate testmp is tot a precursor to any desigt. basis accident or transisnt, lhe resision to the pmeedure &cs not change the test pressure or tvundary; as such, there will be no deleterious efTect upm the containment tuindary or an" components which may be antaunt within the test lundary, in akhtan, standard industry accepted test methods are used such ;
that the abihty to quantify RCIC contauunent isolation valve leakage and its impact upon Primary Cmtairunent integnty are tot adversely alTected Positive controls for restoration of temporary changes to system lineups are providal by imposition of requirements of Procedure 0.9, Tagging Orders. This change des tot afTect the Primary Containment leakage acceptance entena specified in the Technical Specifications or the abdity to detennine if that criteria is satisfied Pmcedure ChanceBeauest (PCR16 PC 520 f Revision 1)
TITI.li: Reactor Water Cleanup (RWCU) local leak Rate Tests Dl!SCRIPTION: The changes to this procalure result in enhancement and clarification of administrative and test pmcedure steps, amt improve charance order recunmendations to inchide additional baniers to mispositioning components during testing The Recommended Clearance order was revised to include component control switches and local compinent tagging requirements.
SAlliTY ANAL.YSIS: This change ars not cause the procedure to be performed during a ddTerent mode of operation than previously authonred in addition, contairunent leak rate testing is not a precursor to any design basis accident or transient. The revision to the procedure does not change the test pre mure or boundary; as such, there Mi be no deleterious efTect upon the containment boundary or any co ents which may ,
be antainal within the te< luntary ln akhtim, standard industry-accepted test me, uis are used such that the ability to quantify RWCU containment isolation valve leakage and its impact usin Primary Cmtauunent integnty are not adversely affected Positive controls for restoration of temporary changes to system lineups are provided by imposition of requirements of Procedur e 0 9. Tagging Orders This change does not afTeet the Primary Containment leakape acceptance criteria specified in the Tecimical Specifications or the abihty to determine if that ciitena is satisfied Pmeedtne Chnnee Reauest (PCR) 6 PC 522 (Revision 21 TIT!li: Standby Nitmgen Injection and Primary Containment Purge and Vent System local Leak Rate Tests DiiSCRIPTION: The changes to this pmeedure result in enhancement and clarification of administrative steps, and impmvc clearance order recon mendations to include additional barriers to mispositioning components dunng testmg 1he Reeminented Clearance Order w as revisal to include component breakers, control switches, and hical componera tagging requirements-SAFl!TY ANALYSIS: This change dies not cause the pmeedure to be perfonned dunng a ditTerent mode of operation than previously authoniecd. In aAhtion, containment leak rate testing is not a precursor to any design basis accident or transient The revision te the procedure avs not change the test pre'sure or boundary, as 166 m
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such, there will be no deleterious effect upm the containmeni boundary or any components which may be cmtained within tix test luntary. In albtim, standard industry. accepted test methods are used such that the ability to quantify containment isolation valve leakage and its impact upon Primary Containment integrity are not adversely alTected. Posiuve controls for restoration of temporary changes to system hneups are provided by impositam of requirements of Pmcedure 0.9, Tagging Orders This change does not afTect the 15 mary Containment leakage acceptance criteria specified in the Technical Specificatims or the abihty to determine if that criteria is satisfied Pmcedure Chance Reauest (PCR) 6 PC 527 (Revi< ion 1)
TITLE Reactor Recirculation (RR) local Leak Rate Tests DiiSCRIPTION: The changes made to this pmcedure result in enhancement and clanfication of administrative and test procedure steps, and improve clearance order recommendations to include additional barriers to mispositioning components dunng testing. The Recommended Clearance order was revised to include component control switches and hical component tagging requirements SAFliTY ANALYSIS' This change does not cause the procedure to be perfonned during a difTerent mode of operation than previously authori/cd In addition, containment leak rate testmg is not a precursor to any design basis accident or transient. The revision to the pmcedure does not change the test pressure or boundary; as such, there will be no deleterious c!Tect upon the containment boundary or any components which may be contained within the test tumndary In additim, standard industry-accepted test methods are used such that the unihty to quantify RR containment isolation valve leakage and its impact upon Primary Containment integnty are not udversely affected Positive controls for restoration of temporary changes h to system lineups are provided by impositam of requirements of Procedure 0.9, Tagging Orders. This change does not afTect the Primary Containment leakage acceptance enteria specified ir, the Technical Specificatiens or the ability to determine if that criteria is satisfied.
Procedure Chance Reauest (PCR) 6 PC 528 (Revision 1)
TITIE Traversing Incore Probe (TIP) System Local Leak Rate Tests y DESCRIPTION: The changes to this procedure result in enhancement and clarification of administrative steps, add
, lieubihty to better allow individual penetratmn testmg, and improve clearance order recommendations 1 to include alhtuud barners to mispositiomeg compon(nts dunng testing. The Recommend Clearance Order w as revised to include component r entrol switches and kical component tagging requirements.
SAFETY ANALYSIS: Tlus change does not cause the procedure to be performed during a dt!Terent mode of operation than previously authonfed In addition, containment leak rate testing is not a precursor to any design basis
- accident or transient. The revision to the procedure does not change the test pressure or boundary, as such, there will be no deleterious etTect upon the containment boundary or any components which may be cmtamed within the test boundary. In uahtion, standard industry-accepted test methods are used such that the ability to quantify TIP containment isolation valve leakage and its impact upon Primary Contammentintegnty are net adversely alTected Positive controls for restoration of temporary changes to system lineups are provided by imposition of requirements of Procedure 0.9, Tagging Orders. This change does ru alTect the Primary Contamment leakage acceptance criteria specified in the Technical Specifications or the abihty to determine if that enteria is satisfied.
Procedure Chanac Reauest (PCR16 PC 529 (Revision 1)
TITLE Primary Contairunent Nitrogen / Air Supply 1.ocal Leak Rate Tests DESCRIPTION Re changes to this pmcedure result in enhancement and clanfication of administrative and test procedure steps, and improve clearance order reconunendations to include additional barners to mispositioning components during testing. The Recommended Clearance Order was revised to include component control switches and kical component tagging requirements.
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SAFETY ANAL.YSIS: This change does not cause the procedure to be perfonned during a different made of operation than previously authorized. In addition, containment leak rate testing is not a precursor to any design basis accident or transient. The revision to the procedure does not change the test pressure or boundary; as such, there will be no deleterious cfTect upon the containment boundary or any compor ents which may be cmtained widun the test tunxiary. In addition, standard industry-acceptal test methods are psed such that the ability to quantify containment isolation valve leakage and its impact upon Primary Containment integrity are not adversely afTected. Positive controls for restoration of temporary changes to system lineups are provided by imposiuon ofrequirements of Procedure 0 9. Tagging Orders. This change does not affect the Pnmary Containment leakage acceptance criteria specified in the Technical Specifications or the ability to determine if that criteria it, satisfied Procedure Chance Reauen~" 'R) 6.PRM 316 (Revision 2)
TITLE: Contrs 4 e o Air Sampling System Knouw Source Cahbration DESCRIPTION: This pmcedure change allows the use of sources of a diflirent isotope to determine various detector etliciencies, in addition to those currently required to calibrate the Control Room Ventilatior. Monitor (CRVM) particulate channel. The monitor calibration and acceptance criteria are not affected by using other radioactise sources to determine instrument efliciencies since this procedure revision does not change the sources used to calibrate the monitor.
SAFETY A NALYSIS: Use of an additional source immediately subsequent to the credited source during CRVM particulate channel cabbration will not aher any equipment lineups. The CRVM is not assumed as an initiator to any plant event. ihe control Room Emergency Filter System (CREFS), which is normally in standby, is initiate? by an actuation signal from the CRVM, and is put in service at the outset of the known source calibration or as a result of the source calibration itself, so that it will already be in senice if and when the attemate source is used Since the CREFS is in senice and nerforming its design function during this pmcedure, the consequences of a plant cwnt are not increased. The additional source, if used, can have no adverse elTect on the CRVM since it is designed for this express purpose. The pmcedure also enters an Allowul Out-of-Senice Time for the CRVM (if required by plant operating mode) during the CRVM cahbration per Procedure 0.26 The consequences of a malfunction of the CREFS or CRVM remain unchanged since in pemunent configuration changes are made by this change and the CREFS will have already been placed in its safety related configuration. The CREFS and CRVM are mitigation systems and cannot create a plant ewnt. The CRVM, by design, is intended to function with an actual or potemial source of activi ty present, Imth during operation and testing. The margin of safety is unchanged.
Performance of this procedure requires CREFS ini9ation; therefore, no possibility of lessening the margin of safety involymg Contml Room calculated dose is introduced.
Procedure Chance Reauest (PCR) 6 PRM 316 (Revision 3)
TITLE: Contml Room Air Samphng System Known Source Calibration DESCRIPTION: During performance of this procedure, the Control Room Ventilation Monitor (CRVM) particulate channel sample averaging time is changed several times for testmg purposes, then set and left at live nunutes A Nuclear Engincemg Department Calculation determined that this setting is non-conservat;ve with respect to the Tcchnical Spectficaton value for CRVM trip setpoint and the functional objective of maintaining the Control Room habitable in all conditions. An Engineering Judgement determined tha' a sample averaging time of six seconds would be appropriate. Procedure 6.PR!d316 was resised accordmgly.
. SAFETY ANALYSIS: h CRVM does alTeet onsite dose limitations and may affect otTsite dose liniitations due to the Control Room personnel's abihty to manually operate plant equipment. The reduction in the CRVM particulate channel sample averaging time fmn. five minutes to six seconds results in the Control Room personnel exposure being within the analped limits during any plant event with the existing trip setpoint and Technical Specification monitor trip limit. This ensures that the safety function of the momtor is 168%
achiemi, therefbre, the change wih not result in an increase in onsite or ofTsite dose. This change does not modify the design of the Control Rmm Radiation Monitor (software design allows the sample averaging time to be adjusted between two seconds and ten minutes). The dose to Control Rmm
_ personnel resulting from the malfunction of any mitigation equipment will remain within the analyzed hmits. Consequendy, the offsite (k>se caused by the malfunction of any mitigation equipment will not be increased because of the impact nf excessive radiation levels on Control Room personnel during the manual operation of plant equipment. The Technical Specification for the CRVM trip initiation setting lunit is 4000 cpm. The purpose of this setting is to ensure that the monitor initiates the CREFS in time to prevent exposure to Control Room personnel from exc:cdmg analyzed limits. The previous sample averaging time of five minutes did not satisfy this function. The new sample averaging time of six seconds will provide the required protection fbr Control Room personnel. Therefore, this change does not reduce the margin of safety as defined in the basis for any Technical Specification.
PIdire Chance Reauest (PCR) 6 RIIR 306 (Revision 2)
Procedure Chance Reauest (PCR16 RIIR 309 (Revision 21 TillE: Reactor iligh Pressure Calibration and Functional (6.RIIR 306)
Reactor I hgh Pressure Once Per Cycle Calibration and Functional, Reactor in Run (6 RilR.309)
DESCRIPTION: These procedure changes added steps to ensure that the logic circuitry associated with the isolation functon of RR-PS-128All) is correctly configured prior to surveillance testing of the pressure switches.
Specifically, R1IR-MOV-Mol7/18 (Resideal Ileat Removal Shutdown Cooling Supply Valves) are mified to be open to ensure that RIIR MOV MO25M3 (RIIR injection valves) are available to perform their safety function.
SAFETY ANALYSIS: Malfunction of this logic circuitry dunng normal operation or testing decs not represent an initiator of any plant event, since the system atreeted is designed to mitigate the etTects of a plant event. The consequences of a plant event remain unchanged. Use of this test equipment will not change Residual I leat Removal or Primary Containment Isolatmn System response to a plant event during the plant modes in which these tests are peribrmed, since this testing already actuates a portion of the logic involved, Momentary connection of test equipment to this logic circuitry will only occur when this portion of the circuit is electrically isolated from the remainder of the logic circuit. Use of this test equipment at the specified poin's in the surveillances has no impact on the consequences of an electrical fault on the associated logic ctreuitry.1he wurst-case malfunction of this logic circuitry is bounded by single-failure design enteria, which is not comprarr sed by use of this test equipment on one logic train at a time. This actisity does not ir av or indirectly atTect any equipment which is assumed to function in the bases fbr any Technical Specification.
Procedure Chance Reauest (PCR16 RPS 302 (Revision 11 TITI E: Main Turbine Stop Valve Closure and Steam Valve Functional Test DESCRIPTION: This procedure change revised the metintology of verifying bypass valve position to use either Digital Electa >llydraulic analog or digital valve position indication and generator output. The previous revision of the procedure used bypass valve limit switches and associated limit switch on-ofTlight indication to venfy the bypass vakes arv functional. The revised method is a more reliable indication of bypass valve nmement. Acceptance enteria w as also revised to provide contingencies for potential failure of turbine bypass valves.
SAFETY ANALYSIS: Neither the method nor frequency of main turbine bypass valve testmg (cycling) is being altered, only the acceptance criteria is being reused to use more reliable and positive indication to verify satisfactory valve cycling. This will improve the overall quahty of main turbine bypass valve testmg that ensures main turbmc bypaas function availability; therefore, the probabihty of occurrence or consequences of an accident or equipment malfunction will not increase. Conservative contingencies are provided in the event a main turbine bypass valw would fail which ensure appropriate action is taken in a timely manner, consistent with USAR and Tecimical Specification requirements. Enhancing the acceptance enteria 169-l
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contingency actions ensures the consequences of a plant event or equipment malfunction arc bounded by USAR/rechnical Specification analy sis The margin of safety .a defined in the basis for any Technical Specification is not reduced.
Procedure Chance Reauest (PCR) 6 RWCU302 (Revision 0)
TMUI: Reactor Water Cleanup (RWCU) Steam Line Hreak Detection Temperature Switches Change Out for Calibration During Operation DESCRIPTION: Changeout of the RWCU steam line break detectian temperature switches is normally performed during an outage tiawgh the use of Pmeedure 15.PC.S301. In the event that changeout is not performed at that time, this procedure will allow changeout of the temperature switches during plant operation with the RWCU system isolated SAFETY ANALYSIS: The RWCU system will be isolated from the reactor system during temperature switch replacement; .
therefore, the possibility of h>ss of coolant inventory through this path will not occur. Also, no work on RWCU piping or system components in direct contact with reac:or coolant will be performed by this procedure. Isolation of the RWCU system prevents the release of significant amounts of radioactive material through this path in the event of an accident as previously evaluated in the USAR. This proccuure will replace temperature sensors which initiate the system isolation in the event of a pipe break withm the system. Replacement of the sensors will casure calibrated sensors are in place as described in the USAR and Technical Fpecifications This will ensure the RWCU isolation signal will occur as designed. The RWCU system will be isolatal and valves will be verified to be closed, Prior to isolation, Chemistry will verify that reactor water chemistry is conducive to removal of the system for the period of time the system is isolated for procedure performance, if conductivity increases to 0.5umhos, steps will be taken to place RWCU back in ser ice. Changing out the switches with calibrated switches will ensure the Primary Containment isolation System will function as designal.
Procedure Chance Recuest (PCR16 SC 501 (Revision 1)
TITLE: Secondary Containment 1.cak Test DESCRIPTION: This procedure was revised to replace the requirement that the average wind speed during the test be between 2 and 5 miles per hours (mph) with a maximtan wind speed limit of 10 mph during the test. The procedure was also revised to include an adjustment to the ditTerential pressure acceptance criterion to account Ibr the " stack etTect' and remove the admmistrative hmit of 1500 cfm for reactor startup. The
" stack etTect' is a change in relative pressures at ditTerent elevations within the Reactor Building due to ditTerences in air densities between the ins;de and outside air.
SAFliTY ANALYSIS: 'this change invohes the mothfication of testing prerequisites and in no way alters the performance of the test. The system under test is a mitigation system that in and of itself carmot initiate any accident sequences. The ch<mge in the testing acceptance criteria will not alTeet the operation of any component or system and does not reduce the requirements for demonstrated functionahty of the Secondary Containment system. The Technical Specification requirements and SAR descriptions are not changed and all c.uTent requirements are shown to be met by the revised surveillance p oceduref The impact of changing the wind speed requirement has been analyzed and found to provide results that bound the criterit. and requirements of the Technical Specifications. The elTect of higher wmd speeds (up to 10 mph) or less than 2 mph wind on Secondary Containment performance were calculated to be an insignificant pressure difTerential relative to the -0.25' Wg criteria. The addition of the stack efTect correction assures that the established test margins are maintained in all design basis temperature conthtions. The remwal of the administrative limit of 1500 cfm for the reactor startup test is acceptable since the originn lanit of 1780 efm as required by Technical Specifications is maintained. The test process, methodology, and steps will be conducted as before and, therefore, the possibility of event untiatm is unchanged. The Technical Specification requirements for Secondary Containment integrity include an acceptance criteria of-0.25' Wg in the Reactor Building relative to the surroundmg areas at wind speuls between 2 and 5 mph. A system capable of satisfying this requirement can be cvected to
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devchip a less negauve pressure at higher wind spmis due to kcalized pressure changes. The replacarcit of the 2 to 5 mph wirxl speal hmitation with a 10 mph maximum wind speed limitation does not represent a ruhiction in the margin of safety because any error introduced due to the new wind speed limitation is less than the cunently established margm of error included in the current test prccedure associated with the allowed 2 to 5 mph wind speed variation.
Procedure Change Reauest (PCR) 6 SC 502 (Revision 2)
TI'IUi: Secondary Containment Penetration Examination DliSCRIPTION: Pncedure frequency was changed from nxinthly to semi-annually, This test had been previously changed from a semi annual frequenev to a monthly frequency in June 1993 in response to a Secondary Containment failure and resulting corrective actions. This allowed Engineering to closely monitor the emaition of the penetratims' seals Engineering has evaluated the results of this testing and determined that it is an inellicient use of plant personnel and resources to perform this test every nmnth. No useful fadure data has been provided or any adverse trends identified that could not be seen by performing this test on a semi annual frequency.
SAFi!TY ANAL.,YSIS: Reducing the frequency of procedure performance will have no etTect on the probability of occurrence of an accident evaluated in the SAR. No new processes are added and no plant hardware changes are proposed Test data has consistently shown the total inleakage to be less than one-third of the allowed flow. Teatmg indicates secondary penetrations do not degrade such that the administrative limit will be exemlaiin six months Herefore, the consequences of an accident or equipment malfunction will not be inernsed by changing the performance frequency of this procedute. The ability of Standby Gas Treatment to maintain Secondary Containment at a negative pressure is not afTected by this procedure change. No Technical Specification margin of safety is reduced by this change.
Procedure Change Reauest (PCR) 6 lDG 101 (Revision 41 Procedure Change Reauest (PCR) 6 2DG 101 (Revision 5)
TITI.E: Diesel Generator (DG) Monthly Operability Test - Division I (6 I DG.101)
Diesel Generator Monthly Operability Test - Division 2 (6 2D0.101)
DESCRIPTION. This revision allows the installation of motion encaling equipment to be moved outside of the Allowed out-of-Service Time (AOT) steps. The AOT will be entered when the test equipment is connected to the diesel control circuits. This provides more llexibility for the activity and reduces the unevailability time of the DG.
SAFETY
, ANAL.YSIS: The DG and its associated controls are not identified as an initiator or contributor for any plant event previously evahiated in the SAR. There is no impact on the probabihty of occurrence of a plant event by instalhng test equipment to the case or shall end of the diesel. Installation of the test equipment does not impact the diescl's ability to start, accelerate to rated speed and voltage, load, and perform its intended safety function The w eight and kication of the test equipment relative to essential support systems and the redundancy provided in the start system prevent defeating the diesel's ability to start should the
- equipment become a missile hazard The pr4 ability of occurrence or consequences of a malfunction of eqmpment important to safety are not changed. Train redundancy and separation ensure that operating the plant while testing one diesel is within the basis of Technical Specifications and the assumptions of the Safety Aaalysis. Both DGs remain available until the AOT is entered. The margin of safety is not changed by installation of the motion encoding equipment.
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i Pncedure Chance Reouest (PCR161DG 101 (Revision 6)
Procedure Chance Reouest (PCR16.2DG 101 (Revision 7)
TITUh Diesel Generator (DG) Monthly Operability Test - Division 1 (6.1 D0.101)
Diesel Generator Monthly Operability Test Division 2 (6 200.101)
DESCRIPTION: These pnmlure :hanges inc(rpruted Engineering recommendations for testing of DG relays. Monthly testmg oflX1 relays DG-REl.-DGI/DG2(59) and DG-REL-DG1/DG2(62Cl.X) was removed from this pnmhae. The okler style (59) relays weic replaced with solid state relays in 1995 and have not exhibited any need for cahbration since mstallation Also, previous dnft of the (59) relays was not suDicient to affect the perfamance of the relays nor alTect overall plant safety or operation. The (62Cl.X) relays also have a good perfbrmance rturd that does tx= warrant nuothly testing The testing frequency of DG REle DG1/DG2(52T) was changed from monthly to once per six months All of the subject relays are also tested by other surveillance / maintenance pacedures.
SAFETY ANALYSIS: %e subject relays will contmue to be tested The DGs will continue to be tested montidy; if discrepancies occtr, the monthly testing will identify a malfunction. The subject relays have proven to be reliab!c; therefore, decreasing the test frequencies of these relays will not prohibit the emergency DGs from performing thm imervied function. There is no change to the facility; this change only decreases the tes6ng frequency (direct uting of relay ftmetion) of specified relays on the DGs. There is no defined vahm in the Technical Specifkations for these relays. Testing of the dicscis ensures that the emergency function is provided as designed This change does not atTect the margin of safety as defined in the basis for any Technical Specification fmcedure Chance Reouest (PCR) 61RPS 304 (Revision 1)
Procedure Chance Reauest (PCR16 2RPS 304 (Revision 1)
TITI.E; Reactor Protection System (RPS) liigh Reactor Pressure Cahbration and Functional Test - Division I (61RPS 304)
RPS Iligh Heactor Pressure Cahbration and Functional Test Division 2 (6 2RPS 304) g '
DESCRIPTION: These pmcedures were revised to incorporate new head correction value (changed from 14 psig to 13 psig) and supplemental Instnanent 1.imit value fbr instruments Nill PS-55A, B, C, and D per Design Calculation NEDC 92-0501.
SAFETY ANAL.YSIS: The head conection change ensures the instruments perform their intended function as specified in the Technical Specifientions and wdl not increase the pmbabihty of a previously evaluated plant event. The change ensures that the instruments perform their required function in order that the specified function, in tlus case a reactor serum, occurs at or before the necessary value in order to negate an increase in the consequences of a plant event or malfunction of equipment important to safety. There is no change to eqmpnwnt important to safety. The NEDC recalculated the desired setpoint of the instruments, however, the instmment contipuration has r ;t changed The margin of safety defined in Technical Specifications is not being changed or reduced. This change ensures the margm of safety is wrrect as defined in the basis for the Technical Specifications Procedure Chance Reouest (PCR) 61 SGT.501 (Revision 1)
Pmeedure Chance Reouest (PCR) 6 2 SOT.501 (Revision 1)
TITI.E: Standby Gas Treatment (SGT) A IIEPA Filter and Component 1.cak Test - Division 1 (6. lSGT 501)
SOT H 1(EPA Filter and Component 1.cak Test Division 2 (6 2SGT.501)
DESCRIPTION: Wese procedures were resised to allow the acceptable leak hmits for SGT filter housmg to be accounted Ibr quantitatively rather than havmg an absolute entena of no leakage, Modem test equipment is very sensitive and is able to detect shght leaks of less than 0 001% Operating License Change Request (OI.CR)96-011 made a correspondmg change to the Technical Specification Bases.
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'l l
l SAFETY- >
ANALYSIS: ne SOT system is not an accident initiator and the hkelihood of an accident is not increased due to SOT ,
filter efliciency testing changes. SOT ediciency requirements will remain unchanged The testing l rmnally causes a light dispersion of DOP smoke in the SGT room; these surveillances will now measure how much leaks into the SGT train in a quantitative manner. The likelihood of SOT failure is not increased The failure modes of SOT will remain the same after the changes to the testing acceptance methods. The new methodology will simply take measurements in a different manner. No changes are being made to the current plant configuration. The Bases for Techn' al Specifications 4.7.B and 4.7.C are being changed to allow testing in a quantitative manner (reference OLCR 96-011), but the Technical
- Specification requirement will remain unchanged. The margin of safety still resides in the requirement for SGT to filter at least a 99% etliciency. This will be maintained Procedure Chance Reauest (PCR) 61 SW 302 (Revision 01 Procedure Chance Reauest (PCR) 6 2SW.302 (Revision 0)
TITLE: Service Water (SW) Pressure Instrument Calibration and Isolation Logic Functional Test - Division 1 (6. ISW.302)
SW Pressure Instrument Calibration and Isolation Logie Functional Test Division 2 (6.2SW.302)
DESCRIPTION: These procedures will perform a logic system functional test on the design bases function of the SW automatic vahing to provide shutoff supply to the turbine building loop on a predetermined pressure on the SW header. They will also perfbrm calibration of kiop instruments to reduce the cycling of the SW crosstic valves. These procedures are required to assess the operational readiness of the automatic closure function dehneated in the Technical Specification Bases.
SAFETY ANALYSIS: These procedures will verify the automatic closure function occurs as described in the USAR and Technical Specitications. Testing of this function will not increase the probability of an accident because of the valve position moving to the conservative design position during the initial stages of procedure performance. Testing ensures required flow of SW is maintained to essential systems as designed; therefore, no increase in consequences is anticipated. Pressure will be morutored and pumps startedhecured as necessary to ensure malftmetiori of equipment important to safety due to loss of SW supply will not occur. Margin of safety is not reduced since the SW kiop pressures are monitored to ensure adequate SW llow is avadable in each loop and the system is in the conservative safe position upon valve closure.
Procedme Chance Reauest (PCR) 6 2SW.101 (Revision 3)
TITLE: Ser ice Water (SW) Surveillance Operation Division 2 DESCRIPTION: This procalure change uplates the laservice Testing (IST) Retest Limit and Operability Limit values for SW-AOV-TCV451B. Maintenance has been performed m this valve that alTects stroke time. Therefore, the IST Retest Limit and Operability Limit are being updated accordmgly.
SAFETY ANALYSIS: SW-AOV-TCV451B perfmnance is not a precursor for any design basis accident. Reactor Equipment Cooling (REC) flow is required within five minutes follcwing an accident to ensure adequate cooling.
Upon k>ss of air or electrical pm er, SW AOV-TCV451B fails from the normal throttled position to the open pantion; under all conditions, SW AOV TCV45 iB passes adequate flow to REC 1IX-B to remove the requin:d heat k>ad Ilowever, even if the valve wcre required to travel from fully closed to fully open, an open stroke time of live minutes would conservatively meet the cooling demands of REC. The required stroke time is more than a factor of 10 greater than the proposed acceptance limit. The new acceptance limit is established in accordance with the IST program such that a degrading condition of the componen will be detected and corrected before fitilure. The revision of the open stroke time acceptance limit does not impact the abihty of SW AOV-TCV451 B to perform its safety function (i.e., to pass SW tiow to REC-1IX-B and to maintain the SW i.ressure lumdary) and does not affect the required system response to accident or transient conditions.
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l Procedure Chance Notice (pCN16 2 2 514 (Revision 21)
TITLE: Residual Ileat Removal (RIIR) Initiation and Containment Spray 1.ogic Functional Test DESCRIPTION: This procedure was rewntten to include the testing previously perfonned in Surveillance Procedure 6.2 2.5.15, RIIR imps A and 11 Containment Spray logic and Valve Movement Functional Test. A 10CFR50 59 resiew was required for this procedure to evaluate if pressing on iIF \ relays can damage the relays and if pressing on these relays simulates the relay coil being energized. (This summary was madvertently not included in the 199510CFR50.59 Summary Repwt. The number of this procedure has since been changed to 6 RIIR301).
SAFETY ANALYSIS: Pushing on the armature of the relay does not damage or afTect the relay's ability to pick up since the fingers of the relay are spring kiaded A physical stop prevents the operator from depressing the annature of the relay farther 'L n wow ' ravel if picked up electrically. This stop is of durable design and is capable of w4standmg the forc . of pickup whether performed manually or electrically. Manually depressing tbmmature mimics the ontact emditions experienced during actual operations, and this test provides ahquate assurance the re sys will function as designed. In addition, since the relay change of state is r result of a manual actior rather than energizing the coil, there is no heat rise from the coil to pnslaa any degradatun c!kets. T1 refbre, it is concluded that the manual actuation of the relays is less severe o the relays than electrical ictuation Manually depressing the armature of the contact does not improu the , untact conditions of; c relay over u hat is experienced by electrically energizing the coil.
The mi nuahk nressing of the relay will not degrade the relay, r.or will it mask any contact problems. It is cmet ad th t manually depressir.g the relays adequately tests the functions as required by Technical Specificetions od does not reduce the margin of safety.
Procedure Chance Reauest O G17.1.8 (Revision 11 TITLE: Rigging and Lif%ng Over or Near Operable Safety Related Equipment for Irradiated Fuel DESCRIPTION: 'lhis precedure was revised to all new Standard Rigging Anchors that serve as pre-approved pick points,
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and to clanfy the requirements for a custom Rigging Engineering Evaluation. The design of the Standard Rigging Anchors fully meets USAR requirements The Standard Rigging Anchors may be installed in accordance with the procedure without additional Design Engineenng approvals.
SAFETY ANALYSIS: The safety evehiatims of the ongmal design m the SAR were based on buildmg structure and substructure performance govemed by meeting the structural design requirements. The design of the Standard Rigging Anchou applies the full design requirements of the USAR. The anchors themselves are passive components that do rot interact unh any active Class I equipment; they only attach to reinforced concrete v alls and slabs of CNS structures. Installation instructions prevent the cutting of embedded reinforcing steel which would be the only possible impact on Class I systems, structures, or components. Thus, by udhng new substructures that meet the USAR structural design requirements, and which have no direct inteructim with active Class I equipment, the probability of occurrence or consequences of a plant event or malfunction of equipment to safety remain unchanged and the margin of safety is not reduced.
Procedure Chance Reauest (PCR) 7.1.9 (Revision 2)
TITLE: General Rigging and Lilhng Procedure DESCRIPTION: This pmeedure was revised to all new Standard Rigging Anchors that serve as pre-approved pick points, and to clanfy the requirements fbr a custom Rigging Engmeering Evaluation. The design of the Standard Rigging Anchors fully meets USAR requirements The Standard Riggmg Anchors may be installed in i _
accordance with the procedure without additional Design Engineering approvals.
SAFETY ANALYSIS: The scope of the procedure specifically excludes nggmg and lithng over or near Class I equipment or
, irradiated fuel The only interaction between the new Standard Riggmg Anchors and Class I systems, structures, or components is the installation into Clas I concrete The procedure protects against
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weukening the Class I concrete by preventing uncontrolled cutting of reinforced steel. The engineering calculation supportmg the Standard Rigging Anchors documents the acceptability of applying the rigging load within the USAR requirements Thus, by adding new Standard RiFeing Anchors, the probability of arurrence or comequences of a plant event or malfunction of equipment important to safety remain unchanFed and the margin of safety is not reduced Procedure Chance Reauest (PCR17.2 6 (Revision 6)
TITIR Control Rod Drive (CRD) Pump Maintenance DESCRIPTION: 1his procedtae was revisal to allow for the removal of the CRD hatch by CNS procedure rather than by the Plant Temporary Malification process. This revision requires the installation of a flooding / safety raihng annni the CRD hatch w hile the CRD pumps are removed or installed from the Southeast Quad area due to floodmg concems.
SAFETY ANAL,YSIS: Instalhng the flaxiing/ safety railing will prevent the possibility of water from a feedwater line break in the steam tunnel fnun entermg the Southeast Quad area via the CRD hatch and rendering Core Spray (CS) 113 inoperable The installation of a railing will maintain system reliability of CS Pump ill with respect to intemal flaxhng and emt= ' ' main bounded by our design calculations in the event of internal 11oaling. The consequenu silure of the floodmg/ safety railing are no difTerent than the comequences of the failure of the CRD hatett The flaxiing/ safety railing provides a flooding barrier design equivalent to that of the CRD hatch with respect to preventing quad flooding. The potential for flashng of the Southeast Quad is unchangal. This activity will not reduce the margin of safety as defined in the basis for any Technical Specification bocedure Chance Reauest (PCR17 2 34. I (Revision 91 Procedure Chance Request (PCR17.2 34 2 (Revision 9 TITLE: Snubber Inspection (7.2.34.1)
Pipe Snubbers Removal and Imtallation (7.2.34 2)
DESCRIPTION: Various administrative changes were made to these procedures to reflect NRC approval of Imervice Impection Relief Request RI 13 and to enhance inspection requirements to better document resolution ofidentified discrepancies.
SAFETY ANAL.YSIS: These changes do not cause the procedures to be performed during a ditTerent mode of operation than was previously authorized. Snubber inspection is not a precursor to a seismic event or transient. Snubber impection continues to be conducted in accordance with Technical Specification requirements; as such, these changes have no impact on the ability ofinspected snubbers to prevent unrestrained pipe motion, and thus to prevent structural damage to the pressure trundary during design basis seismic events. The population of snubbers to be inspected is not changed and inspection requirements are not reduced.
These changes do not affect the snubber inspection frequency, population, or acceptance criteria speci' icd in Technical Specifications.
hoecdure Chance Reouest (PCR) 7.2 35 (Revision 191 TITI.E: Safety and Relief Valve Test and Setting DESCRIPTION: This procedure was updated to conform to ASME Section XI Code,1989 Edition, requirements.
Spa:ifically,insta:ctiom and appropriate information were added to provide for petformance of a visual impection and an as-found seat leakage test, statements were added to require more accurate pressure measurement gauges, and the code requirement to wait ten minutes between successive setpoint tests was incorporated. Expected etTects of these changes are a more reliable and improved test method for
- relief / safety valve testing
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l SAFETY
. ANALYSIS: This activity enhances relief / safety valve testing and does not delete testing nor change the intent of the existing testing. The alhtional testmg being added includes a non-intrusive visual test and a second seat leakage test which is performed in the same manner as the seat leakage test already contained in the procedure. There is no change to the existing test configuration The testing continues to ensure operational readiness in accordar.ce with the purpose of the Inservice Testing Program and the plant's relief / safety valve testing pMosophy. Therefore, this change will not increase the probability of occurrence or conscomoca. of a plant event or malfunction of equipment important to safety.
Consequences of malfunctions are dictated by component safety function (i.e., to maintain pressure boundary and to lill to relieve overpressure cor itions withm the system pressure boundary). This activity dms not afTect a component's safety fun; tion nor change the intent or efTectiveness of testing.
As the alhtional testing does not impact the ability of a relief / safety valve to perform its safety function, this activity d=s not reduce the margin of safety. In addition, this testing is in accordance with ASME Section XI Code as required by 10CFR50 55(a)(f)
Procedure Change Reauest (DCRT 7 2 493 (Revision 7)
TIII.E: Scram Pilot Solenoid Valve Maintenance DESCRIPTION: This procedure revision adds a section to allow replacement of the exhaust diaphragm end caps and exhaust diaphragms on Scram Pdot Solenoid Valves CRD-sol 18(XX YY). The procedure previously only allawed for complete valve replacement and did not allow for any rebuild. i SAFETY ANALYSIS: Replacement of the exhaust diaphragm and end caps does not alter the functionality, operating parameters, integrity, rehability, or qualitication of the Control Rod Drive (CRD) system or its hydraulic control units. This activity does not alter the trutipatmg capabilities or design margins of the CRD system, nor does it alter any critical performance margins associated with this critical system. The new exhaust diaphragm end cap has been approved by Mmor Modification Package 95 179. Suflicient controls are included to ensure proper i nplementation of the diaphragm and end cap replacement. This activity does not fonn any new connections or interface with any other plant systems or components Since the basic configuration, functionality, integrity, and reliability of the CRD system will not be adversely impacted, the possibihty of reducing any margins of safety as defined in the Technical Specifications is negated.
Pmeedure Chance Reauest (PCR) 7.2 531 (Revision 13)
TITLE: Diesel Engine Mechanical Inspection DESCRIPTION: Changes u ere made to this procedure to al'ow for a more eflicient means to perform Diesel Generator (IXI) mechanical inspections and reduce the amount of time the diesel is unavailable for routine repairs.
These changes included climination of unnecessary Quality Control inspections, a new section for governor oil replacement, and clarification of engine inspection and reassembly steps. It also allows for pre-inspection steps to be perfirmed prior to engine teardown. In addition, a change was made to address torquing requirements for the fuel injection pumps in accordance aith vendor recommendations.
SAFETY ANALYSIS. The changes to this procedure will increase the reliability of the DG and tninimize out of service time during the maintenance activity. This maintenance procedure is performed within the requirements of the Technical Specificatons for Standby AC Power. This activity will not prevent the DG from providmg its emergency power function for Loss of Coolant Accident (LOCA) and Loss of OtTsite Power (LOOP) ewnts This procedure is not used to detennine the operability of the DG, operability is detennined via surveillance procedures. LOCA and LOOP events bound all possible plant events and equipment failures This maintenance activity is specific to maintenance on the diesel engines and equipment in the diesel rooms No other safety related equipment is atTected by this procedure change. These changes will not a!Tect the margin of safety requirements spectfied in the Technical Specifications.
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Procedure Chance Recuest (PCR) 7.2 53 3 (Revision 9)
Pmcedure Change Reauest (PCR) 7 2 53 5 (Revision 1)
TITIA Diesel Generator (DG) Cylinder Maintenance (7.2.53.3)
Diesel Generator Dynamic Balancing (7.2.53.5)
DESCRIPTION: These changes provided new guidance for installmg the high pressure fuel hoses on tbc fuel injection pumps. They also a<kkxl a neans to install a temporary air supply for the turning gear, in addition, other changes wcre made to panide clearer directwn when performing maintenance. Selected Quality Control steps were also removed to be consistent with the revision to Procedure 7.2.53.1.
SAFETY ANALYSIS: ne changes to these procedures will increase the reliability of the DG and minimize out of service time during the maintenance activity. The maintenance procedures are performed within the requirements of the Technical Specifications for Standby AC Power. His activity will not prevent the DG from providing its emergency power function for Loss of Coolant Accident (LOCA) or Loss of OfTsite Power (LOOP) events Rese pmcedures are used to perform maintenance on the emergency d. c cl and are not used to determine the operability of the DG; operability is detennined via surveillance procedures. LOCA and LOOP events bound all possible plant events. This activity is specific to maintenance on the diesel engines and equipment in the diesel room. No other safety related equipment is alTected by these changes. These changes will not alTect the margin of safety requirements specified in Technical Specifications.
Procedure Chance Recuest (PCR) 7 2 53 5 (Revision 0)
TITI.E: Diesel Generator (DG) D)namic Balanctng DESCRIPTION: This new pmcedure provides instructions to direct the dynamic balancing of either diesel engine on the CNS DGs. Dynamic balancing was previously perfinned under the guidance and direction of vendor and dicsci owners group testing guidelines and recommendations.
SAFETY ANALYSIS: Here is no accident caused by ti iuss of a DG analyzed in the SAR. The electrical power system has diverse power supplies and is usigned such that no single failure will prevent it from performing its intended function. The DGs are a backup to the preferred ofTsite power sources. The loss of one DG would not prevent the onsite power system from performing its intended function. Loss of a DG is part of the design basis of the plant (Loss of Coolant Accident with a less of Otisite Power and a single faihtre). The etTect of a total k>ss of AC power (Station Blackout) has been previously analyzed.
Ihredurah/ing the maintenance activity of balancing the DG engines is intended to enhance the overall avadahthry and reliability of the DGs by ensuring that the maintenance is donc consistently. It dees not aher the system's operation or its response to an accident that the Diesel Generators are required to help mitigate. It enhances the DG's ability to respond to st.pport systems that are important to safety.
lYocalures which govern DG operabihty and availability as required by Technical Specifications are not atTected by this new procedure. The operability of the DG will be verif.cd by surveillance testing following successful completion of this procedure. The procedure does not introduce a new maintenance activity, but only pmcedurahzes an existing activity. This activity will not alTect the mode of operation of the DGs, will not create a system configuration or operatmg condition such that Technical Specification Limitmg Conthtion f- Operation or surveillance requirements are no longer adequate, nor will it bypass or invalidate automatic actuation features required to be operable by the Technical Specifications.
Herefore, use of Pmeedure 7.2.53.5 does not reduce the margin of safety as defined in the basis for any Technical Specification.
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Procedure Change Reauest (PCR) 7.2 79 (Resision 0)
TITLE: Securing Control Room Ceiling Diffuser Panels DESCRIPTION: This rrw parature provides the requirements for fastening Control Room ceiling ditTuser pancis to the supporting grid. This procedure implements the fastening requirements authorized by Modification Package (X>-083f3 (Anchorage Upgrade of Control P.com I anels, Auxiliary Relay Room Panels, and the
~
Reactor Protection System Motor-Generators).
SAFETY ANALYSIS: The installation p events the diffusers from becoming dist;dged during a seismic event and, therefore, could only decrease the probability ofoccurrence of an accident or malfunction of equipment important to #cty. Removing the tie wTap seismic restraints of the diffusers during maintenance activities will not increase the consequences of an accident or equipment malfunction because the difTusers are attended by personnel performing the work and controlled in accordance with the Seismic llousekeeping Procedure.
No new credible postulated failure mechanisms are introduced which require evaluation. No margin of safety as defmed in the basis of any Technical Specification is afTected by this activity.
Procedure Chance Reauest (PCR) 7.3 23. ! (Revision 2)
-TITLE 24V Reactor Protection System Battery Equalizing Charge DESCRIPTION: This procedure change allows the loading of the 24 VDC electrical distribution panels to the battery chargers during the equalization charging of the batteries following discharge testing. This change was implemented so the time that instrumentation powered by 24 VDC is unavailable due to battery testing is minimiicd.
SAFETY ANALYSIS: The equipment afTected by this procedure change consists of the 24V batteries, chargers, distribution panels, and the 24V loads associated with the distnbution panels The loads consist of Source Range Monitors, intennedsate Range Monitors, Process Radiation Monitors, and the Safety System Status Panel.
None of these components is an event imtiator. Equipment operability will be maintained by this procedure during battery equalization since the power supplies to the 24 VDC equipment will be available/operaNc. This change will allow fuel movement during the battery equalization process since the required instrumentation is operable This precedure change allows the operation of plant equipment withm its established electncal ratmgs and includes periodic electrical checks to ensure that the electrical ratings are not chal!enged by controlling the seing of the battery charger to within these limits The current battery charger equalization voltage setting will not be changed, but will be verified so that the operability of the 24V h> ads on the electrical distnbution panels can be assured. The result of a fadure of any specific piece of equipment alTected by this PCR is unchanged and no new malfunctions will be introduced All equipment operabihty and design functions are maintained The Technical Specification margm of safety is unalTected by this change since all equipment will operate within its electncal limits arwl no new equipment functions or operational requirements are added. The fuel movement mode will be allow d dunng 24V battery equalizahon, but all required mstrumentation will be operating as designed.
Procedure Chance Reauest (PCR) 7.3 34 (Revision 6)
TITLE: Diesel Fire Pump Batteries Replacement and Maintenance DESCRIPTION. This procedure was revised to delete acceptance criteria for battery replacement. This change requires entry into a Limiting Condition for Operatmn (LCO) thr fire pump operabihty. Exit from the LCO will
. be verified by perfbrmance of Surveillance Proche 6 FP.101 in lieu of procedure acceptance critena as Post Maintenance Testing (PMT). System lince, .H c!carance order charges were also made.
SAFETY ANALYSIS: T1us mamtenance procedure does not increase the probabihty of occurrence of a fire event because it does not intnshice additionalignition sources. Work is perfbrmed within the boundaries of a clearance order l
. and existing plant procedures. This procedure only impacts the dicsci fire pump. The diesel fire r.u:nfs I l
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i ability to mitigate events such as fires and provide makeup to the Spent Fuel Pool is unchanged. These '
changes redae the pmbabibty of equipment malfunction by requiring 6.FR 10i to be performed as PMT to ensure equipment restoration folkming maintenance. Replacement of the batteries ensures continued equipment operabihty. Faihre of a diesel fire pump has been previously evaluated. No additional failure nules are introduced Ilattery replacement is performed in lieu ofinspection of batteries per Technical Specification requirements. This increases the margin of safety.
Pmeedore Chance Reauest (PCR) 7.3 50 4 (Revision 01 TITII: SW-650MV and SW-651MV Flow Adjustment DESCRIPTION: TVs new procedure provides instructions for flow adjustment of SW-650MV and SW-651MV valve operators to values established and evaluated per Design i hange (DC)93-057, Service Water and Reactor Equipment Cooling System Moddication.
sal ETY ANALYSIS: his procedure adjusts hmit switches on SW-650/651 to positions previously evaluated per DC 93-057.
Failure of SW-650/651 is not an initiator for any Design Basis Accident evaluated in the USAR. This pnmhre removes the alTected system components from service per a Clearance Order and applicable Technical Specification Limiting Condition for Operation (LCO). A jumper wire and test switch are installed in parallel with the isolation relay contact and do not affect the isolation relay functions. The automatic isolation functions for the operable Service Water system are maintained operable. Therefore, system accident response n not changed and consequences are bounded by Technical Specification LCO analysis. Ihe margin of safety as defined by the applicable Technical Specification LCO is not reduced.
Procedure Chance Reauest (PCR) 7.3 55 (Revision 0)
TITLE: Raceway Installation DESCRIPTION: This new pmcedure pmvides guidance for the installation of electrical conduit and raceways. The appmpriate moddication document will provide the information related to conduit patWrouting and size information.
SAFE"-
ANAL ..S: This pacedure installs condmt/ raceways / supports per modification package content and will be governed by the Work Plaruung process. As a result of these upper tier controlling documents, this procedure and its use cannot initiate a plant esent. Since the routmg and sizing information is taken from the moddication document, this pmcedure cannot be used to install improperly sized or improperly routed conduit / raceway / supports Provisions to ensure the seismic acceptability of condmt/racewsy/ support installations are included in this procedure. Conduitdraceways/ supports are passive portions of an installation and faihire is not credible based upon the seismic acceptability of the installation. The installations covered by this pmeedure are not related in any way to any margin of safety defined in the basis for any Technical Specification.
Procethire Chance Reouest (PCR) 7.3 56 (Revision 0)
' TITLE: Cable Installation Guidelines and Design Ch.mge Considerauons DESCRIPTION: His new pnmhre provides guidance for the installation of power / control cables and conductors This procedure relics upon the modification document to provide cable size / number intbrmation.
SAFETY ANALYSIS: his pnxxdure installs condta: tors / cabling per moddication package content and will be governed by the Work Planning pmcess As a result of these upper tier controlling documents, this procedure and its use cannot imtiate a plant event. Smce the routing and sizing informanon is taken from the modification document, this pmeedure cannot be used to install impmperly sized or improperly routed conductors / cabling. In addition, this procedure has the correct provisions for mcggering of cable and empkiys the use of conservauve calculations. As a result, this procedure contains the controls necessary to assure pre- an.d post event equipment operabihty, Conductors / cables are passive portions of an 179-
installation Mechanical failure is rot credible based upon the seismic acceptability of the conduit installatim in additim, the routing of conductors / cables, which is defined by the modification package, ensures the site respnse to fire induced failures does not jeopardize required equipment operability. The installation of conductors / cabling by this pmcedure is not related in any way to any margins of safety -
defined in the b. sis for any Technical Specification.
Procedure Chance Reouest (PCR17 4.1 (Revision 19)
Pmcedure Chance Reatrst (PCR17.4 1.1 (Resision 4)
TITI.E: Shield Plug Removal (7.41)
Shield Plug Installation (7.4 I 1)
DESCRIPTION: These procedure changes allow the use of Kevlar slings for installing / removing the Cavity Shield Plugs instead of the onginal wire cable slings and strongback. The Cavity Shield Plugs can be installed'remomi using either the enginal wire cable slings or Kevlar slings The new rigging hardwr.c will enhance the operation of h11ing the Cavity Shield Plugs.
SAFETY ANAL YSIS: This activity does not imulve any equipment or systems credited as event initiators nor does it adversely afTect any systems credited with terminating transients. This activity does not afTect the ability to shutdown the reactor or to contain radioactive materials either during normal operation or post event.
The activity of h!bng the Cavity Shield Plugs will not be changed. A calculation was performed by ABB Combustion Engineering and review ed by NPPD engincenng personnel to detennine the configuration of the riggmg and the appropriate Kevlar slings and standard rigging hardware required for the lift. The use of Kevlar shngs and standard rigging hardware will raluce the time in removing and reinstalling these plugs. The increusal stresses caused by the non-vertical sling angles were ibund to be acceptable per the requirements of NUREG-0612. No new type of malfunction of equipment important to safety is introduced with this change. This activity does not alTect any parameter whose margia of safety is addressed in the basis for any Technical Specification.
Procedure Chance Rcouest (PCR) 7 4 61 (Revision 8)
TIT 1.E: Reactor Vessel Steam Separator and Fuel Pool Gate Installation DESCRIPTION: Dunng RE16 and REl7, several of the 36 total shroud head bolts have had nut retainers which failed to engage when the shroud head assembly was installed into the reactor. Any bolt with an unkicked nut retainer is cmsideral to be nonfunctional since the bolt could become loosened dunng reactor operation.
An unalysis was perfbmied by General Electric to determine the mmimum number of shroud head bolts regtural to maintain the structural integnty of the reactor intemals under all design basis conditions. 1he results of tlus analysis show that the required CNS structural criteria are maintained with a minimum of 22 uniformly spaced fully functional bolts This pmcedure incorporates the General Electnc guidance fbr allowing continued operation with up to 14 nonfunctional shroud head bolts providing the General Electric recommendations are met SAFETY ANAL.YSIS: The structural integrity of the reactor vessel intemals is maintained dunng plant operation for all appheable accident scenarios when at least 22 shroud head bolts remain fully functional per the criteria estabbshed by General Electne analysis The nonfunctional shroud head bolta cannot become dislodged into the reactor pressure vessel coolant due to the geometry of the assembly and the mass of the bolts.
Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety are not increased The reduction in the shroud head boltmg configuration does not degrade the structural integnty of the reactor sessel intemals below the CNS safety analysis structural enteria. The margin of safety as defined in the Technical Specifications is not reduced.
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I Procedure Chance Reauest (PCR17.4 24 (Revisi.cn.5)
TIT 1.E: Reactor iluildmg 1001' Strongbacks Visual and Non-Destructive Testing DESCRIPTION This resision changed the Non-Destructive Examination (NDE) requirements for the Reactor Building stnogbacks fnen an inten al of five yeatuo five schaluled refueling cycles This incorporates the current 18 month operating cycle and results in a comrnitment change (reported in the Commitment Change section of this report)
SAFETY ANAL.YSIS: This actinty does not involve any equipment or systems credited as event imtiators, or with mitigating the consequmees of a plant ewnt a malfunction of eqmpment imponant to safety. This change does not allixt the abihty to shutdtmu the reactor or to contain radmactive material either dunng normal operation or post event. The NDE testmg requirements for the Reactor lluildmg strongbacks remain the same; the testmg inten al is the only change occurring Our previous commitment (reference letter 1.QA8.300177 dated July 25,1983) was every five years for an NDE and a visual prior to each use. This conunitment w as based on the infiequent use of these liRing devices. The lifting devices may be used more oRen if a forced outage were to occur; however, the use that the hRing devices would see would still be considered infrequent. The proposed activity does not alTect any parameter whose margm of safety is addressed in the basis for any Technical Specification.
Procethire Chance Reouest (PCR17127 (Revision 131 TITI.E Main Steam 1.ine Plugs installation DESCRIPTION 'Ihis procedure v as revised to include the installation of a vent hose to each mam steam ime plug and routing it to the refueling floor. This will decrease the chance oflosing volume m the reactor cavity dunng refuehng actmties due to the inadvertent opemng of one of the vent valves on the steam line plugs This revision also provides that the position of the steam ime plug vent valve will be detennined by the Mam'enance Refueling Floor Supenisor.
SAFETY ANAL.YSIS: Tlus change les not allixt the integrity of the main steam hne plug and does not alter its function or the manner m which it performs its function The steam hne plugs are refuchng tools and senicing equipment and are not relied on to mitigate the consequences of any plant event evaluated in the SAR.
The vent hne passes through the water in the reactor vessel and cavity, but has no other interface with other plant systen's or equipment. The possibihty exists for a leak to develop in the vent line which would allow water fnun the reactor cavity to enter into the main steam Imes in the event the vent valve was open This uould result m a loss of vessel inventory. The maximum amount of flow would be limited by the si/e of the hole, and could be no larger than the existing vent valve. The same systems and controls for leak detection and nutigation would be in etTect as currently used during this mode of operation, and this event would be tuunded by the existing event of an imulvertently opened vent valve. This resision has no etTeet on the marpm of safety as defined m the bases of the Technical Specifications. There ce no Technical Specification reqmrements for the mstalle: ion or function of the steam ime plugs or for the control of the vent valves. The vent hnes are being installed 'o prevent the loss of w ater from the vessel cavity in the event of an madvertent opening of the plug vent valve. Pool levelis not alTeeted by these changes, therefore, the margm of safety is not reduced Pnwedure Chance Reouest (PCR17.510 (Revision 0)
Procedure Chance Reauest (PCR17 511 (Revision 01 Pnmhtre Chance Reauest (PCR) 7 5. I 2 (Revision 01 Paredure Change Rcouest (PCR) 7.5 13 (Revision 01 TITI.E SMll.000 Motor Operated Valve (MOV) Refurbishment (7.5.10)
SMll-00 MOV Refurbishment (7 511)
SMll-0 through SMll-4 MOV Refurbishment (7.512)
Sil-0 thmugh Sil-4 MOV Refurbishment (7 5.12)
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.i1 DESCRIPTION: These new procedures were developed to provide direction for the performance of maintenance on MOVs. The pacedures combine previously existing electrical and mechanical pmcedures into one procedure. In addition to performing maintenance, the procedures direct the replacement of old style spring cartndge caps with new style spring canridge caps, the replacement of capscrews, the replacement of housing covers, the renoval of the kcal indication drive chain, the replacement of fork lug connectors with ring lug connectors, and the replacement of melamine rotors with fibrite motors. None of these changes represent a change in the design of the MOV.
SAFETY ANAL,YSIS: The changes to the operator are consistent with the manufacturer and/or CNS documentation. 'Ihe changes do not alter the function of the operator in any way that will alTect the initiator of evaluated accidents. The system function and respmse remain the same; therefore, the system mitigation capability terruuns the same. No new failure modes are intraluced that could reduce the mitigation capabilities of the systems involved or lead to a plant event of a different type than previously evaluated. The changes are an improvement to the operator; as such, the probability of occurrence of equipment malfunction is decreased The changes do not affect or change how the system is configured and the oserall design and function of the operator remains unchanged The changes do not aher the performance of the valves as required in Technical Specifications Procedure Chance Reouest (PCR) 8 4 (Revidon 10)
TITI.E: Routine Sampling and Sample Valve Control DIiSCRIPTION: Tlus revision adds additional sample points for sampling Service Water (SW) to facilitate sampling the SW effluent fmm each Residual lleat Removal (RIIR) and Reactor Equipment Cooling (REC) llent Exchanger when the SW Monitor is inoperable or manual sampling is required.
SAFETY ANAI.YSIS: The removal of a sample does not alTect the ability of SW to perform its function. Since the Control Room is aware of sampling in pmgress, and the sampling location is attended and restored at the end of samphng, there is no increase in the consequences of an accident or equipment malfunction The sample punts do not afTect the abihty of either REC or Ri ta e nerform their functions. The loss of water via the sample pnnts is txxnled by the flooding analysis for ti. se areas. Control Room permission is required prior to samphng. The sample points are being used in a manner consistent with their design intent which is to drain or transmit water. Smcc the sample points do not alTect system operation and the affected instrumentation is not contained in the Technical Specifications, there is no reduction in the margin of safety.
Procedure Change Reauest (PCR) 8 8 8 (Revision 16)
TITili: Particulate, hxime, and Noble Gas Sample Collection for Elliuent Monitors, Control Building intake Monitor, and Drywell Air Mon 6r DESCRIPTION: The changes to this procedurc a,wnhancements that allow leak testing in all modes of Kaman operation and provide clear procedur.1 guidance on the change of filters on the monitors. The providing of clear procedural guidance on wt ich steps to perform when equipment is inoperable or auxiliary sampling equipment is operating allo w better procedural compliance. The addition of correction factors for the Reactor and Turbine Building vents provides conservative corrections for analysis results. A step was also adJed to ensure a coating of vacuum grease is present to Ud in proper scaling of Nuclear Research Corpiration monitors.
SAFETY ANALYSIS: The techniques used for leak testing in this procedure are already employed by other plant procedures and the testing does not change the operational configuration of the monitors. The use of correction factors provides conservative correction to analysis results used to determine dose to the public sia the Oil' site lhe Assessment Manual (ODAM) These pmcess monitors are not assumed as initiators to any plant ewnt and are not assumed to mitigate the otTsite consequences of any plant event, only to detect an abnormal condition The leak testing does not change the manner in which morutoring occrrs, but provides more assurance that the precess monitoring equipment perfonns its design function. A failure
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of the testing is already addressed in the procedure. There is no interface between these monitors and equipment important to safety. The leak testing enhances the margin of safety in that in ensures the nxrutors are providmg occuraic sample collection and monitoring. The use of the correction factors adds additional conservatism to the dose calculation results and improves the margin of safety.
Procedure Chance Recuest (PCR) 8 811 (Revision 17)
TITI.E: 1,iquid Radioactive Waste Discharge Authorization DESCRIPTION: This procedure was revised to require at least one Circulating Water (CW) pump in service during liquid radioactaw waste discharges. It also removed notes that state that the Radiological Manager's approval is required to discharge with < 159,000 gpm dilution flow (historically, i CW pump = 159,000 ppm).
SAFETY.
ANALYSIS: The abihty to discharge liquid radioactive waste is exit assumed in any CNS accident or transient analysis.
The operational and safety design basis of the liquid radwaste system is to prevent inadvertent liquid discharges and to ensure that planned discharges are within requirements. The liquid radioactive waste system is not assumed as a mitigation system for any accident or transient analys.s. The changes do not modify any systems, structures, or components or the manner in which they are operated. This change still utdves one of the liquid radioactive waste dilution methods described in the USAR. The procedure still requires liquid radioactive waste discharge + to be diluted such that the limits of Technical Specification 3/4.21B,10CFR20,10CFR50,10CFR100, and 40CFR190 are not exceeded. Therefore, there is no reduction in the margin of safety.
Procedure Chance Reauest (PR 9.2 4 (Revision 9)
TITLE: Surveying Materials for Release ofTsite DESCRIPTION: This pmcedure revised the physical boundary kration u here nonradioactive materials are released for unrestricted use The boundary was previously the protected area fence. It was changed to the Radiologically Controlled Area (RCA)/ Satellite RCAs boundary. This change also revisen the trash release process by allowing the use of the integrated tool monitor in lieu of a frisker to conduct a preliminary survey of the trash. The tool monitor provides the same level of detection capability as a frisker.
SAFETY ANALYSIS: These changes do not imulve any imtiators to any plant event and are not included in any assumption used to analyze plant events in the SAR. They have no efTect on the plant's ability to mitigate the cmscquences of a radioactive release. The changes do not involve any physical changes to plant systems, structures, or components, or the manner in which they are operated or maintained The revisions to this procedure do not alTect the current radioactive material release criteria, or the final release c:iteria for unrestricted release of nonradioactive materials. The changes do not alTeet ofTsite dose to members of the public or the radioactive waste process control program since the materials released ofTsite are nonradioactive.
Procedure Chance Reauest (PCR19 31.4 4 (Revision 6)
Procedure Chance Reauest (PCR19 31.4 5 (Revision 3)
Procedure Chance Rgquest (PCR) 9 3.1.61 (Revision 6)
Procedure Chance Reauest (PCR) 9 3 2 6 (Revision 7)
Procedure Chance Reauest (PCRT 9 3 3.1 (Revision 12)
Procedure Chance Reauest (PCR19.3 4 I t Revision R)
Procedure Chance Reouest (PCR) 9 5 I (Revision 12)
Procedure Chance Reauest (PCR) 9 5 3.10 (Revision 5) lhcedure Chance Reauest (PCR) 9 6.2 (Revision 8)
TITLE: lon Chamber Survey Instruments - 13icam Models RSO-5 and RSO-50 (9.3.1.4.4)
Dositec Portable Remote Monitors (9 3 1.4.5)
Portable Alpha Meter Ludlum Malel 2 (9.3.161)
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Tennelec Lib 510011 arxl APC 11 Operation (9.3 2.6)
Direct Reading Pocket Dosimeter (9.3.3.1)
National Nuclear Corporation (NNC)) artal Monitors (9.3.4.1)
Receipt of Radioactive Materials (9.5.1)
Waste Stream Sampling (9.5.3.10)
Eberline Model MS 2 and MS-3 Portable lleta-Gamma Counting Instruments (9.6.2)
DESCRIPTION: Le abow procedures were revised to reflect a change in terminology by the Radiological Department.
The term "1Icalth Physics" was changed to " Radiological Protection" and the term " IIP" was changed to"RP" throughout the procedures. In adhtan, title changes due to the reorganization of the Radiological Department were incorporated SAFETY ANALYSIS: Altinigh the title changes reprcent an alternative method ofidentifying the !!calth Physics organization as dncussed in the USAR, the title changes are an administrative change only and do not reflect a change in intent. These title changes have no impact on the probability of occurrence or consequences of an acrident or malfunction of equipment important to safety. There are no formal Technical Specification bases for the subject titles.
Prgedure Chance Reauest 9 4 TPl. DOS (Revision 01 TITLE: Tcledosimeter Installation, Reksatmn, and Removal DESCRIPTION: This new procedure was dewkiped to allow for the installation, relocation, and removal of teledosimetry system components throughout the plant.
SAFETY ANALYSIS: The new process provides the necessary limitations and has adequate controls to assure that plant equipment will not be adversely alTected, includina post-even' periods. In addition, the teledosimetry system will not physically interface with any system, structure or component (SSC) T1.e satisfactory results of Special Procedure 97-009 (RCA Tcledosimetry Testing, provide the assurance that SSCs will not be adversely stTected by the radio frequency signals emitted by the teledosimetry rystem. Therefore, the probabihty of occurrence or consequences of an accident or malfunction of equipment important to safety are not increased Since testing of the teledosunctry system has shown that SSCs, including various electrical and electronic systems, will not be adversely alTected, the new process will not reduce the margin of safety as defined in the basis for any Technical Specification Procedure Chance Reauest (PCR) 9 5 2 (Revision 12)
TITLE: Radioactive Sources Control and Accountability DESCRIPTION: Technical Specificauon Amendment 174 deleted sections 3.8 and 4 8 regarding Source Leak Tests. The infmnatim contained in these sections is now required to be contained in Procedure 9.5.2. Reierences to the previous Technical Specification sections were removed The following additional changes were made: revised a:quirements for Control Room / Shift Supervisor notification durmg transfer of sources,
, clantied enteria for when smoke detector sources require leak testing, added critena regarding inventory and disposal of liquid sources, clarified startup source leak test requirements, and removed duplication / unnecessary details.
SAFETY ANALYSIS: Source leak testing and control are not initiators to any plant ewnt and are not included in any
. assumptions used to analyze previousiv evaluated plant events None of these changes afTect the plant's abihty to miugate the consequences el a radioactive release. These changes do not involve any physical changes to plant systems, structures, or components', or the manner in which they are operated, maintained, nnhfied, or inspected No new modes of plant operation are intaxiuced Improper transfer of a radioactive source could result in plant radiation monitor alarms and, in extreme cases, changes in plant system hneups (isolation to imut potential releases or uptakes). Station procedures already address actions necessary to respond to radiation monitor alarms and system realignments due to these alanns.
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These changes do not afTect any offsite release criteria tv increase the ofTsite chise to members of the pubhc. Source leak testing cnteria previously cmtained in the Technical Specificatices are now rekicated to this pmcedure.
Procedure Chance Recuest (PCR) 10 25 2 (Revision 0)
TITIJI: Refucimg Core Shuille
' DESCRIPTIONI This new procedure was devekped to prmide guidance for core shufiling and fuel movement betv cen the care arxl the fuel pool. This pmcedure dacharges a hmited rumber of fuel bundles (maximum of 8) fnun the react (t to the fuel pol. It alkms movement of fuel in the core to die final core kication, but does int alkiw interim novement of fuel in the ccre.1hc safety evaluation documents that the change to allow fuel o!T kiad to begin prior to the five day emidown period does not create an unreviewed safety question, SMTITY ANAL.YSIS: Inss of spent fuel pool emling is not an accident as described in the USAR. The system alignments described ensure that spent fuel pol temperatures remain less than 150' as asumed in the USAR and the NRC Safety Evaluation Report (SER) for1.icense Amendment No. 52. The safety design basis for the Fuel Pml Cmling System of ensuring fuel damage does not occur is not compromised by this pnmture cimnge. This pnmlure does not nnhfy in any way the performance of systems or components utihud to mitigate the conscquences ofpreviously evaluated accidents. The procedure allows for the oft-load of less than a full core, and permits the off-load to begin in less than the 120 hour0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> cooldown time assumed in the origmal design calculations, and also n!!ows fuel to be moved at greater than three bundles per hour. The NRC SER for License Amendment No. 52 stated that,"For the full core elT load case, if we assume that the postulated Safe Shutdown Earthquake (SSE) causes the complete !oss of spent fuel coolmg just aller a full core olT load, the heatup rate of the water in the fuel pool will be about 13'F/hr.
Since at this time the temperature of the fuel pel wdl be less than 150*F, it will take about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to heat the w ater at the surface of the spent fuel pol to 212*F w ben bulk toiling can conuacnce." The SER 1 also concluded that if makeup water is supplied, that under bulk tuiling conditions the temperature of the fuel will not execul 350*F, and that this is an acceptable emperature from the standpoint of fuel element integnty and surface corrosion This procedure does not atreet the abihty of the Fuel Pool Cooling System to remme the heat pencrated by the fuel, and does not afrect the maximum decay heat loads described in the USAR No new operational or failure modes are intnxtuced and the configuration of the plant is not airceted. 'lhere is no etrect on the margin re safety as defined in the basis of any Technkal Specification There are no specific Technical Specification requirements for the spent fuel pool emling system related to temperature. Pool level is not alTected nor is the number of fuel assemblies capable of being stored in the pol altered by this change.
Pmeedure CSnee Recuest (PCR) 10 25 2 (Revision 2)
TITI.h; Refueling . Core Shut 11e DESCRIPTION: This procedure was revised to allow full core shutiling with appropriate hmitations, restrictions, and centrols. It allows the incore shufile method of refueling with all control nx!s fully inserted in the core es reluired by Technical Specifications Information previously added to this pmcedure to support the replacement of a leaking fuel bundle was removed SAFETY ANAI.YSIS: Movement of fuel in the core or fuel pool is alknved by the Technical Specifications and the USAR. This proecdure change will ensure this activity is pmpetly controlled and meets the reactivity requirements of Technic 4d Specifications for the ccre and fuel pool. Plant events related to shuffling fuel such as a fuel kalmg error, rotated bundle loadmg error, refueling accident, or kmding fuel in an uncontrolled cell are r.ot changed by this pmcedure change. The same pmeedures, equipment, and quahfied personnel are being used to move fuel as was previously done. Thus, there is no change in the probability of
. occurrence or comequences of a plant event Equipment important to safety involved with this pre-%rc change includes the refueling platfonn, fuel grapple, refueling tools, the fuel pool, and the vesset No clunges are bemg made to any equi;> ment or the way equipment is used Thus, there is no change in the probability of occurrence or consequences of a malfunction of equipment important to safety. This l
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procedure change will ensure the reactivity margin dermed in the Technical Specifications and USAR is being maintainal A rekind design change for the new fuel cyc!c will require the reactivity of the core to meet Technical Specification, USAR, and licensing requirements. Reactivity data from this reload design is used by COSMOS, a reacter engineering computer program, to detennine reactivity conditions -
in the ctue and develop shuffle movements that minimize reactivity and allow only movements that meet Technical Specification and USAR reactivity limits. Bus, there is no reduction in the margin of safety as defined in the basis for any Technical Specification Procedure Chance Recuest (PCR) 10 31 (Revision 4)
TIT 1.E: Fuel Reliability DESCRIPTION. Previous revisions to this procedure were improperly screened for 10CFR50.59 applicability and no written safety evaluations were done. Herefore, a safety evaluation was performed for the implementation of this pacedure.
. SAFETY ANAL.YSIS: The use of this pmcedure cannot change the probability of occurrence of any accidents evaluated in USAR Chapter 14. The procedure prescribes actions for various levels of off gas activity which assure that reactor coolant activity remains within the requirements of the Technical Specifications for coolant chemistry. Power suppression testing alTects the way the reactor is operated for a limited length of time.
During the testmg period, all thermal parameters are maintained within limits. Core exposure distribution is an important assumption in the accident analyses. Exposure distribution is a very weak function of control rod pattern over the short period of time required for power suppression testing. Therefore, the etrects of temporaq contml nxi pattern changes during the test period will have an insignificant etTect on any assumed accident analysis exposure distribution The pmcedure allows control rods to be imerted to suppress local power to reduce olT gas during subsequent operation. The long term effects of the exixwure distnbution, fuel hwal peaking factors, and thermal limits are evaluated by General Electric to detennine the acceptability ofoperating with power suppression control nxis prior to long term operation with such a control nxi pattern Therefore, the radiological comequences of these accidents are unchanged. All equipment and systems used to implement this pmcedure are used in their normal manner. 'lhe probabihty of a fuel failure because of a power change by core flow or control nxi motion is unchanged The comequences of any equipment malfunctions are bounded by existing USAR evaluations. This pmcedure does not create the possibility of an accident of a ditTerent type than previously evaluated in the USAR. Accidents caused by mechanical failures, clad overheating, and pipe rupture were considered. The activities performed by this pmcedme are not initiating events for any equipment malfur; tion. He margin of safety is not reduced because the procedure assures compliance with reactor coolant Technical Specifications, perfonns all contral rod moves within thermal limits, and evaluates clIects of any altered control nx! pattem e a result of power suppression to assure the changes to the exposure distribution are within the bounds of the accident analyses.
Procedure Chance Request (PCR) 13 til (Revision 7)
TilLE: Reactor Equipment Cooling (REC)licat Exchanger Performance Analysis DESCRIPTION: A Safety Evahiation was performed to addresses existing procedural steps to install temporary pressure gauges and reposition valves to obtain data to evaluate REC IIcat Exchanger performance. The preparation phase of tins activity involws the installation of non-safety related pressure gauges in existing pressure sample points kicated at the Service Water (SW) inlet and outlet of the REC Ileat Exchanger.
The sample points consist of a root valve and a shutotTvalve nhich are closed during normal operation.
During the perfbrmance stage of this procedure, the opening of the root and shutotT valves intmduces the .
. non-safety related pressure gauge as a pressure boundary for the SW system Additions were made to
- this procedure to ensure an associated pipe stress analysis (IdEDC 96-040) is not invalidated and to ensure the proe: dure ia not performed on both divisions of SW/ REC concurrently, I
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SAIETY ANALYSW: The installation of the gauge will not increase the probabdit) of occun ence of an accident as the gauge has no failure nules connected to any accident initiators, and has a pressure rating which is in excess of the design pressure dthe SW piping to the REC llent Exchariger. The irrtallation does not increase the consequences of any previously analyzed accident as the only credible failure mode would be a minor breach of the SW system pressure boundary. The ikxxiing effects as a resuit of failure of the gauge as a pressure txnmdary is bounded by the Medium Energy Line Pipe Break Analysis (Calculation NEDC 91-066). Engineering Evaluation 96-010 has shown that any possible breach has no efTect on system operability 'Ihe SW wystem pressure boundary would be re-established by closure of the essential root valve. The gauge is used for pressure nxmitoring only and does not provide any controlling functions for equipment imputant to safety. Calculation NEDC 96-040 has shown the installation of the gauge does in impact the pressure txuntary integnty of the safety related portion of the SW system. The gauges are ml) installa! on one d vision for perfamance of this pacedure. The kiss of one division of SW has tx:en analped in the USAR. Equipment kicated in close proximity to the gauge installation which performs a safety function has been qualified for the harsh environment created by a liigh Energy Line Break (llELD). The breach of the SW pressure boundary is bounded by the llELB analysis Since system operabiUty is not alTeeted and required design flows are maintained, the margin of safety is not reduced.
Pmeedure Chance Recuest tPCR) 11151 (Revision 81 TITLE: Reactor Equipment Cooling fleat Exchanger perfonnance Analysis DESCRIPTION: This procedure was revised to capore the c!Tects of tube plugging, as well as to reflect the conclusions reached by a recent revision to suppsting calculation NEIX'94-021 which is used to establish maximum acceptable foutng factor acceptance criteria. This change results in a more conservative test by accounting for present and future tube plugging.
SAFhTY ANALYSIS: No changes in the procedure's method '7y or its time ofimplementation are made. The acceptance enteria tar the maxim:nn allowable levels of fouling are being reduced, resulting in a more conservative newptance enteria TLen: fore, this change cannot increase the probability of occurrence or consequences of a plain event. No changes are being made to the operation, maintenance, or testing requirements of any plant equipment as the result of this procedural change This activity does not change the design basis for any systems, equipment, or appurtenances, nor does it alter any plant equipment or their normal or abnormal operating procahires. No equipnent is being nuhried and no new interfaces are being created.
This revision does not chringe any design margins, design criteria, or assumptions used in any design calculations establishing any margins of safety. Rather, by reducing the maximum allowable fouling factor, the existing margins of rafety are increased.
Procedure Chance Reanest (PCII) 14.1.103 (Revnion 0)
I're<_edure Chance Reauest (PCR) 14.1 1031 (Resision 0) j TITLE Air-operated Valve (AOV) Diaginstic Equipment Setup and Operation (14.1.103)
AOV Test Equipment Connection (14.1.103.1)
DESCRIPTIONn New Procedure 14.1.103 describes th: setup and operation of the ABB AirCet system used to test and diagnose air operated valves. New Pracedure 14.1.103.1 provides guidance for connecting the ABB AirCet test cquipment to an AOV and its supporting Instrument Air (IA) components for the purpose of conducting AOV testing.
SAFETY ANALYSIS: The only plausible failure in the AirCet system is a failure which results in an internal rupture of the tubing and an IA leak internal to the ajuipment. Failures in the 1A System are analyved in the USAR and it is conchided that faihire of the IA System wdl not adversely alTect plant safety. The specific failure of the connection to the lA System, the air line sapplying the AirCet test equipment, or the AirCet test
. equipnnt itselfintroduces, at most, a SM' break in the IA System which would not constitute a cortplete lA System faihire. This fadure is bounded by the complete kiss of IA. The failure modes of equipment important to safety as a result of a complete IA System failure are such that the safety function of this 187
equipment is prexrved or actuated ne possibihty of a malfunction to the 1A system itselfis not created by the fact that the installation / removal of the Aircet equipment and disassembly / assembly ofIA tubing is controlled in accordance with procedures and work practices. The plant transients caused by a complete kxss ofIA bound the limited failure caused by the failure of AirCet equipment or connections.
- IA System failure does not change any limiting safety system settings, safety limits, or result in the failure of safety related equipment or equipment important to safety, IA is not discussed in any Technical Specification basis.
Procedure Chance Request (PCR) 1413 3 (Revision 21 hocedure Chance Reauest (PCR) 1413 4 (Revision 2)
TITLit Reactor Feedpump Turbine (RFPT) A Control System Cahbration (1113.3)
RFPT 11 Control System Cahbrstion (14.13.4)
DESCRIPTION. These pmcedures include acceptance criteria for RFPT control valve position indication and isolation
- transmitter output. 'Dus change increased the range of the isolation transmitter acceptance enteria to be more in line with that of the remote indicator and Plant Management Infonnation System computer acceptance critena SAFETY ANALYSIS: The indication loop has no impact on automatic control system operation. The indication is only used by Operations during transfers between low pressure (extraction) and high pressure (main) steam, and redundant indications exist. The nonessential Reactor Feed system does not perform or suppen any accident mitigation or isolation functions Any possible malfunction scenario is bounded by the loss of feedwater transient, feedwater controller failure maximum demand transient, and the turbine missile analysis There!bre, any variations in RFPT operation which might occur will not increase the consequences of any equipment failure. The tolerance change will not lead to a ddTerent failure mode than those already analyzed Since indication tolerance is unchanged, the abihty of the operators to monitor and contml the RFPT is unchanged Therefore, there can be no impact on the assumptions, calculations, procedures, or design specifications used to deternune the plant's margin of safety.
Procedure Chance Reauest (PCR) 14.171 (Revision 12)
Prot edure Chance Reauest (PCR) 14.17 2 (Revision 11)
TITLE: DG-1 Annual Calibration (14.17.1)
DG 2 AnnualCalibration(1417.2)
DESCRIPTION: Several changes were made to enhance performance of these procedures, including the following:
- 1) added setting of DGSA-PRV-9 and DGSA-PRV 10 regulators which were not previously tested,
- 2) moved perfbrmance of some tests to more applicable section Ihr diesel engine mode of operation,
- 3) added calibration of DGSA-FREG 9 and DOSA-FREG-10 which were previously not tested,
- 4) decreased manipulation of valves and test equipment multiple times,5) added note to clarify steps to perfonn ifengine is already running, and 6) eliminated steps to perform multiple tests on the main engine overspeed trip to prove operabihty as the governor and overspeed tnp mechanisms were replaced last outage with improved components.
SAFETY ANALYSIS: This procedure performs calibration ofinstruments not required for an emergency start of the Diesel Generators (DGs). These instruments ensure the DG operates as designed during normal stans. The ability of the DGs to provide emergency power to equipment required to mitigate the consequences of a plant event or malfunction of equipment important to safety is not affected. Ensuring the DG is pmtected and operational through assurance ofinstrumentation calibration mitigates the possibility of a plant event of a ddTerent type than previously evaluated. Calibration of this equipment ensures the margin of safety associated with the DGs is maintained through assurance that the equipment is within cahbration. Therefore, the margm of safety is not reduced
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- Procedure Change Reauest (PCR) 15 EH 301 (Resision 21 TITI.U: Safe Shutdown Emergency 1,ighting flattery Performance Test DESCRIPTION. Changes made to this pmcedure inchalal 1) addition of a note indicating that when the pmcedure is used for post maintenance testing the 70 hour8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> wait period between lighting gmups is not applicable.
- 2) venfication that tested lamps are bright afler 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in lieu of between 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 20 minutes, and 3) venfication that the time tested samps were on greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and if the lamps were dim or unht after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> nd 20 minutes, the lamps should be retested.
SAFETY ANALYSIS; This aethity will not change the state or function of safety related systems, structures, or components and will not alter any of the inputs or assumptions for evaluated accidents. The changes do not afTect the functim, per formance, or integrity of any boundaries with v hich safety related systems form or support the pnmary pntective barriers The emergency hghting system is utilized for accident / event mitigation (i.e., post. fire manual actions or station blackout) Testing of the emergency lighting system ensures reliability fbr these actions The changes to this pmcedure do not decrease the number of emergency lighting units iur do they decrease the illumination of the emergency hghting tmits. This change will not irxhice failure of any equipment important to safety and will not cause any new failure modes to occur, The margin of safety which is established by the design and performance of safety related systems will not be reduced by tlus activity.
Procedure Chimee Reauest (PCR) 15 FP 307 (Revision 1)
TITI.E: Italon 1301 Computer Room Fire Suppression Surveillance Checks DESCRIPTION. No audible alarm occurs at the contml cabinet because no kical horn is installed in this system; therefore, pmcedural steps referring to the alarm were delete \lso, two steps were reversed to indicate that a manual release switch must be reset before alarms e e cleared.
SAFETY ANAL,YSIS: The changes to this pmcedure do not intnxtuce accident precursors or ignition sources. The changes ensure that the llalon system panel operates as designed and installed, thereby ensuring that the consequeneu of a Computer Room tire are not increased and reducing the pmbability of a panel failure.
The consequences of a panel failure remain unchanged from those assumed in the Fire Ilazards Analysis.
No new I talon system failure modes are intnxiuced The Computer Room 1 talon system (krs not form the basis for any Tecimical Specification margin of safety.
Procedure Chance Reauest (PCR) ! S I SWHP 301 (Revision 01 Procedute Chance Reauest (PCR) 15 2SWHP 301 (Revnion 01 TITI.E: Senice Water Booster Pump (SWBP) Start Interhxk Test - Division 1 (15.lSWBP.301)
Senice Water Ikuster Pump Stait Interkick Test Division 2 (15 2SWBP.30l)
DESCRIPTION: These new procedures were developed to test the interkick between the Senice Water Pumps and the Senice Water Ikiostcr Pumps.
SAFETY ANALYSIS: This actnity interfaces with the SW and RIIRSW systems which are required to support shutdown cooling (SDC), however, this activity alTects only one division of RIIRSW while the other division remains available to support SDC requirements. This activity requires that all SW pumps be available for operation and that at least one SWBP is in a protected status and available during its perfonnance.
Altluugh leads associated with SW pump breaker auxiliary contacts are lined, tF,ae leads n'Tect only the start cueuit ofone division of SWhPs These lead lins in no way airect the SW p ,nps' breaker circuitry and all SW ptunps are available to support SDC. Adequate systems and equipment remain available to nutigate the consequences of any plant ewnt assumed for the condition the plant is in, and ta mitigate the consequences of equipment malfunction This activity involves equipment whose malfunction is envehiped by existing analysis and no malfunction can be postulated that would create a malfunction that has not been prevmusly analyzed This activity enhances the reliability of the SW and RilRSW systems
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by ensuring the interlock between both systems' pumps is functional. Technical Specification 3.5.13 allows for the RilR SWilPs to be inoperable during a cold shutdown condition. This 'actisity is perfwmed on m!y me division of RilRSW and resuires that at least one SWBP from the other disision be readdy available which is cxeservative with respect to the Technical Specification 1.imiting Condition for Operation Procedure Chance Reauest (PCR) 15 TG 302 USAR Change Reauest (UCR) 97-12 i TITLE Main Turbine Trip Functional Test (Revision ! Cl)
USAR Change Main Turbine Trip Functional Testing DESCRIPTION: Main turbine trip functional testing was changed from a monthly to a quarterly freque..cy to reduce the -
probability ofinitiatmg a turbine tnp or turbine omrspeed event at power while performing the functional test and to reduce ALARA concerns. This change also represented a change to the testing interval described in the USAR. The USAR previously stated that an operational tes: of the overspeed trip mechanism oil pressure check device was performed in accordance with the provisions in the vendor's instruction manual.
SAFETY
. ANALYSIS: This change will not increase the probabihty of a turbine overspeed event and resultant missile Feneration since no change tc the testing proccas has occurred. Actual test data conclusively indicates that the change m test frequency will not result in a previously evaluated overspeed event in excess of the 108%
setpomt, The consequences associated with this event are bounded by a turbine overspeed event with missile generation No increase in offsite dose release endangering the health and safety of the public would occur should an overspeed event with missile generation occur. Functional trip testing is performed to detect degradation in main turbine protective trip device operation and predict u hen trip device malfunctim will actually occur. Protective trip setpoints, specifically the overspeed trip setpoint,
- will remain within the existing analped design Quarterly perfonnance of functional trip testing will contmue to ensure that degradation of the mechanical overspeed trip device is detected This actisity does not change the proce for testing the protective trip system; therefore, no new operational scenarios are introduced to create the possibility of an event different from any previously evaluated. No new equipment is bemg akkx! to the protective trip system and no new faihire modes are introduced. Turbine overspeed margin of safety is not desenbed in the basis for any Technical Specification. No other Tecimical Specification margms of safety will be reduced by this change.
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REGUIATORY COMMITMENT CllANGES The following Regulatory Commitments were revised based upon evaluations performed in accordance with the Nlil Guidehne for Managing NRC Commitments.
SOURCII OF COMMITMENT Rl!VI4 ION TO COMMITMf?NT 1.etter 1.Q A8300139 fun J. M. The original commitment stated that NPPD would notify the Regional Pdant(NPPD)to D n Vassallo Administrator of any violations of the NPDES permits by transmitting a copy (NRC) dated Apnl 29,1983 - of the applicable letter either sent to or received from the permitting agency.
NRC Notification of NPDES This commitment was clin.inated such that the subject reports no longer need Permit Changes or Violations. to be sent to the Regional Administrator. The original commitment represented an administrative burden without a commensurate improvement in satety. The original commitment was not tied to a regulation, but to an NRC request.
I.etter inun D E Schaufelberger, The response to liA 82-46 stated that "the bhxxl cell count examination will be (NPPD) to J. T. Colhns (NRC) meluded as part of the annual physical examination for CNS employees in the dated August 15,1983 - Response 1984 cycle " The response to IR 82-32 stated that "a momtoring program has to Order Mahfymg lacense, been estabhshed to identify the date of the last medical physical of each Management Appraisalof employee so that thev will be medically examined by a physician at least every Corporate Management, EA 12 months " These conunitments uere revised to chminate the requirement to 82-46. give all CNS employees physicals and bkxx! cell counts. Regulations drive the requirements for physical examinations. 10CFR, OSilA and ANSI Standards i.etter 1.QA8300012 from J. M. give the requirements for physicals and the required tests to be performed Pdant (NPPD) to G. I Madsen during that physical.
(NRC) dated Apnl i 1,1983 -
NPPD Response to IE Inspection Report 50-298/82-32.
I.etter CNSS866027 from J. M. The response to IR 96-19 stated that procedural guidance would outline the Pdant (NPPD) to R. D. Martin requirements for controlling materials that have been radiologically surveyed (NRC) dated October 7,1986 - for unrestricted use, but not yet directly released otTsite. It would outline the NPPD Response to IE Inspection time perial allowed betw een the performance of radiological surveys for Report 50 298/86-19 matenal to be released olisite and the time w hen the material is actually released olisite for unrestricted use, as well as designate storage hications to CNS 1.ieensee livent Report pmvide control over matenals that have been radiologically surveyed for 86-010 dated May 15,1986 - unrestricted use but have not yet heembrectly released otTsite. The response Release of Radioactive Material to to 1.ER 86-010 committed to the ume 1 3rrective action. This commitment Unrestricted Area w as revised to specify that procedural guidance will outime the requirements for radiological sun ey and unrestricted release of nonradioactive materials from the Radiologteally Controlled Area (RCA) and Satellite PCAs. The physical boundary for the survey and unrestricted release of nonradioactive materials was changed from the protected area fence (restricted area) to the RCA and Satelhte RCAs Tlus improves process etTiciency and ehminates several low value and time consummg activities performed by station personnel, yet maintains the health and safety of the public. The commitments made in 1986 were made as a result of weak RCA access and ceress controls The RCA and Satclhte itCAs access and egress controls hase been strengthened significantly since 1986 and have made the IR 86-19 and IliR 86-010 commitments obsolete.
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I etter CNSS876042 from G. A. The response to IR 87 18 stated that, "The Station Operations Review Trevors (NPPD) to the NRC Committee (SORC) will continue to evaluate those events that appear to be Document Control Desk dated - questionable tmder the reportability requirements of 10CFR50.73 utilizing September 2,1987 - Response to guidance provided in 10CFR50.73, NUREO 1022, Supplements to NUREO IE Inspection Report 1022, and other documents, as applicable, such as Generic Letter 87 09. In 50-298/87.I8. the future, if, aAer initial SORC review and evaluation, the reportability requirement for an event is questionable, the NPPD Nuclear Licensing and Safety Department will be contacted for guidance and interpretation." This commitment was revised to state that Nuclear Licensing and Safety shall e /aluate events for reportability. If a question should arise, SORC will be consulted Due to NPPD reorganization and several process improvements, the Nuclear Licensing and Safety Department has taken over this responsibihty.
CNS 1.icense Arnendment No. 7, The District committed to performing an inspection of the Standby Gas Inchidmg Change No.10 to the Treatmer.t (SGT) System filter housing doors using DOP aerosol sprayed Technical Specifications, dated around the periphery of the door to identify any leakage, with any detection of February 6,1975. DOP in the fan exhaust considered an tmaeceptable test result that requires gasket repain and ictest. This con-nitment was revised such that the District will continue to test for in-leakage, but will add it to the overall SGT filter cliiciency which is required to be 2 99% by Technical Specificati ons. The Surveillance Procedure also requires visual inspection and replacement of gaskets as necessary. Modern testing equipment allows for detection of minute quantities of DOP and current criteria for achieving no detected leakage cannot be achieved. Technical Specification compliance is preserved with this change. Reference Operating License Change Request (OLCR)96-011 lbr a related change to the Bases of the Technical Specifications CNS Licensee Event Report As corrective action for LER 94-034, CNS committed to initiate Preventive 94-034 dated December 12, Maintenance (PM) changes for the Emergency Lighting System to ensure the 1994 - Emergency Lighting manufacturer's recommended life times are not exec ed as well as System Cannot Be Assured of incorporate gaxi industry practices for Emergency Lighting maintenance. This Meeting 8 I tour Operation commitment was implemented, but is subsequently being revised to indicate Requirement Due to Design and that a malification will be performed to replace Emergency Lighting Units Maintenance Deficiencies. with more robust units capable of meeting or exceeding design rrquirements and to subsequently establish preventive maintenance in accordance with the manufacturer's recommendations. It was determmed that the reliability of the emergency lights was not being improved by peribrmance of the original PMs.
The sisting emergency lights are on the A(1) classification list of the Maintenance Rule and are being corrected on an expedited basis until such time as a permanent modification is implemented to replace the units with larger capac;ty models.
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l lxtter CNSS907024 fnen L G. Item 11 of Generic Letter 89 13 required licensees to conduct a test program to Kunci(NPPD) to NRC Document verify the heat transfer capability of all safety related heut exchangers cooled by Control Desk dated January 29, ser ice water. A portion of our response to this item mdicated that the District 1990 Response to Generic Letter currently verifies the heat transfer capability of Diesel Generator Jacket Water 89 13. (DGJW) and Diesel Generator Lube Oil (DGLO) llent Exchangers by testing at least once each operating cycle. This commitment was deleted as it pertains to the DG system. GL 89-13 allows for reassessment of testing frequencies aner three tests have been performed lleat exchanger perfonnance evaluations were performed five times on each of these DG 1leat Exchangers since April of 1990. It is felt that there is little value being added by the continued performance of the pre- and post work performance evaluations.
Frequent regular maintenance in lieu of testing is allowed per GL 89-13.
Therefore, the performance evaluation requirements were deleted from the Preventive Maintenance for these heat exchangers. Visual inspection and cleaning, as required, will still be performed on a once per cycle basis. This is adequate to ensure these heat exchangers are maintained ira good condition and provides reasonable assurance that the heat exchangers can perform their design function Letter CNSS906996 from n A. The response to IR 90-30 stated that," Station procedures will be revised to Trevors (NPPD) to NRC require documentatien of all lapsed craft job specific requalification training Document Control Desk dated and the circumstances surroundmg the lapsed trairsng. This documentation October 24,1990 - NPPD will be approved by the cognizant station departmental manager and forwarded Response to Inspection Report to the Training Manager for inclusion into training records " This commitment 50-298/90-30. was revised such that supervisors wdl be notified of all lapses in cran job specific requalification training to ensure that cran personnel are not performing tasks independently when qualification requirements are not up-to-date. Lapse letters to the Training Manager are not necessary since the system used to track training quahfications automatically disqualifies any j individual who has not completed requalification requirements.
! CNS Licensee Event Report The con ective ac' ion ihr LER 87-004 stated that replacement of the seat rings87-004 dated February 6.1987 - in Reactor Feed Check Valves (16-CV-13CV through 16CV) would be Excessive Primary Containment increased from every three years to each refueling outage. This commitment I eakage Discovered During Local was revised to state that the subject seat rings should be replaced every other i euk Rate Testmg. refueling outage. If subsequent testing and inspection support additional service time, the frequency may be extended further. Changes to the valves made during the 1993 refueling outage have proven successful in improving local leak rate test performance, allowing extension of the soft seat changcout to at least once every other cycle.
l.ctter LQA8300177 from L G The original commitment stated that a program would be developed for Non-Kunci (NPPD) to D. R Vassallo Destructive Examination (NDID of the load beanng welds of the designated (NRC) dated July 25,1983 - special lifting devices on the refueling Door at 5 year intervals. This Control of Ileavy 1.oads - Phase 1. commitment was revised to change the 5 year interval to 5 operating cycles.
The original commitment was made at the time when the operatmg cycle was 12 months. Therefore, the frequency of testmg will not be changed based on the operating cycle enteria. The original commitment was every 5 years for an NDE ud a visual prior to each use due to the mfrequent use of the lining devices. The lining devices may be used more than 5 times tf a forced outage wcre to occur, however, the use that the haing devices wouhl see would still be considered infrequent
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