NL-04-0194, Report of Facility Changes, Tests and Experiments, Safety Evaluation Summaries

Report of Facility Changes, Tests and Experiments, Safety Evaluation Summaries

ML040440165
Person / Time
Site:	Hatch
Issue date:	02/10/2004
From:	Sumner H Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-04-0194
Download: ML040440165 (18)

Text

H.L Sumner, Jr. Southern Nuclear Vice President Operating Company, Inc.

Hatch Project Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.7279 February 10, 2004 SOUTHERNNA COMPANY Energy to Serve Your World" Docket Nos.: 50-321 NL-04-0194 50-366 U. S. Nuclear Regulatory Commission ATITN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Report of Facility Changes. Tests, and Experiments.

Safety Evaluation Summaries Ladies and Gentlemen:

Enclosed is the 24 month report of facility changes, tests, and experiments safety evaluation summaries in accordance with the requirements of 10 CFR 50.59(d)(2).

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely, H. L. Sumner, Jr.

HLS/IL/daj

Enclosures:

Report of Facility Changes, Tests, and Experiments, Safety Evaluation Summaries cc: Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch Document Services RTYPE: CHAO2.004 U. S.Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. S. D. Bloom, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch

ENCLOSURE EDWIN I. HATCH NUCLEAR PLANT - UNITS 1 AND 2 NRC Docket Nos. 50-321 and 50-366 Operating Licenses DPR-57 and NPF-5 REPORT OF FACILITY CHANGES, TESTS, AND EXPERIMENTS, SAFETY EVALUATION SUMMARIES

GLOSSARY ACRONYMS AND ABBREVIATIONS ABN as-built notice AC alternating current ADS automatic depressurization system AHU air handling unit ALARA as low as reasonably achievable APLHGR average power linear heat generation rate APRM average power range monitor ARI alternate rod insertion ARM area radiation monitor ARTS average power range monitor, rod block monitor, and Technical Specifications ASME American Society of Mechanical Engineers ATWS anticipated transient without scram ATWS-RPT anticipated transient without scram-recirculation pump trip BHD bottom head drain BOP balance of plant BWR boiling water reactor BWROG Boiling Water Reactor Owners Group CFR Code of Federal Regulations COLR Core Operating Limits Report CRD control rod drive CS core spray CST condensate storage tank DAS data acquisition system DBA design basis accident DBE design basis earthquake DC direct current DCB double cantilever beam DCR design change request DCS dry cask storage DHR decay heat removal dP differential pressure ECCS emergency core cooling system ECP electrochemical potential EDG emergency diesel generator EFCV excess flow check valve EFPD effective full power days EFPH effective full power hours

GLOSSARY ACRONYMS AND ABBREVIATIONS EHC electrohydraulic control ELI Equipment Location Index EMI electromagnetic interference EOC-RPT end of cycle-recirculation pump trip EOF Emergency Operations Facility EPA Environmental Protection Agency ERFDS Emergency Response Facility Display System ETS Environmental Technical Specifications EQ Environmental Qualification FHA Fire Hazards Analysis FPC fuel pool cooling FSAR Final Safety Analysis Report GE General Electric GL Generic Letter GPC Georgia Power Company HCU hydraulic control unit HNP Hatch Nuclear Plant HPCI high pressure coolant injection HVAC heating, ventilation, and air-conditioning HWC hydrogen water chemistry I&C instrumentation and control IE inspection and enforcement IGSCC intergranular stress corrosion cracking ILRT integrated leak rate test IRM intermediate range monitor ISFSI independent spent fuel storage installation ISI inservice inspection IST inservice testing LAN local area network LCO limiting condition for operation LDS leak detection system LDCR license document change request LLRT local leak rate test LLS low-low set LOCA loss of coolant accident LOSP loss of offsite power

GLOSSARY ACRONYMS AND ABBREVIATIONS LPAP low power alarm point LPCI low pressure coolant injection LPM loose-parts monitor LPRM local power range monitor LPSP low power setpoint MCC motor control center MCPR minimum critical power ratio MCR main control room MCRECS main control room environmental control system MDC minor design change MG motor generator MPC Multi-Purpose Canister MOV motor-operated valve MPL master parts list MSIV main steam isolation valve MS SRV main steam safety relief valve MSL main steam line MSLRM main steam line radiation monitor MSR moisture separator reheater NMA noble metals addition NPSH net positive suction head NRC Nuclear Regulatory Commission NSSS nuclear steam supply system ODCM Offsite Dose Calculation Manual OPDRV operations with the potential to drain the reactor vessel OPRM oscillation power range monitor PAM post accident monitoring PASS post accident sampling system PCIS primary containment isolation system PCIV primary containment isolation valve P&ID piping and instrumentation diagram PLC programmable logic controller PPC plant process computer PRB Plant Review Board PRNM power range neutron monitor PSW plant service water P/T pressure/temperature

GLOSSARY ACRONYMS AND ABBREVIATIONS QA quality assurance RBM rod block monitor RCIC reactor core isolation cooling RCPB reactor coolant pressure boundary RCS reactor coolant system REA Request for Engineering Assistance RES Request for Engineering Services RFI radio frequency interference RFP reactor feed pump RFPT reactor feed pump turbine RG Regulatory Guide RHR residual heat removal RHRSW residual heat removal service water RMCS reactor manual control system RPS reactor protection system RPT recirculation pump trip RPV reactor pressure vessel RRS reactor recirculation system RSCS rod sequence control system RWCU or reactor water cleanup RWC RWCS reactor water cleanup system RWE rod withdrawal error RWM rod worth minimizer SAER Safety Audit and Engineering Review SAT station auxiliary transformer SBGT or standby gas treatment SGTS or SGT SCM stress corrosion monitor SDC setpoint design change SED System Evaluation Document SJAE steam jet air ejector SLMCPR safety limit minimum critical power ratio SNC Southern Nuclear Operating Company SoRA Summary of Required Actions SPDS Safety Parameter Display System SRB Safety Review Board SR Surveillance Requirement SRM source range monitor SRV safety relief valve

GLOSSARY ACRONYMS AND ABBREVIATIONS SSAR safe shutdown analysis report SSC system, structure, or component TBWD thrust bearing wear detector TCV turbine control valve THV torus hardened vent TIL Technical Information Letter TIP traversing incore probe TLD thermoluminescent dosimeter TM Temporary Modification TRM Technical Requirements Manual TS Technical Specifications TSV turbine stop valve

1 0 CFR 50.59 SUMMARIES PLANT PROCEDURES 34SP-03-05-02-BC-1-0 This special purpose procedure describes the necessary actions, precautions, and prerequisites required to allow removal of 230kV Switchyard Bus #2 and each of the four individual incoming 230kV Transmission Lines from service. Section 8.2 of the Unit 2 FSAR describes the configuration of the 230kV Switchyard and offsite sources available to the Unit 1 and Unit 2 4kV emergency busses. During the interval when Buss #2 and the transmission line are out of service, the expected worst case fault with a concurrent breaker failure could result in the loss of startup transformer (SUT) 2D. However, SUT 2C, both of the Unit 2 emergency diesel generators and the swing emergency diesel generator are available to provide adequate power to the 4kV emergency busses. A degraded grid condition could become more probable during a fault condition in the switchyard. However, the degraded grid studies performed indicated that during the Bus #2 outage and with the required precautions and prerequisites of this procedure, the expected worst case fault with breaker failure would result in a worst grid voltage above or equal to the required 101.3% of 230kV in the switchyard.

TEMPORARY MODIFICATIONS (TM)

TM 1-02-10. Rev. N/A This TM places ajumper across thermocouple wires for control rod 38-27 to eliminate a MCR nuisance alarm without affecting the alarm signal from any other control rods. There will not be any temperature monitoring capabilities for control rod 38-27 while this TM is active. The effectiveness of a reactor scram is not impaired by the malfunction on any one control rod. Therefore this TM does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety.

TM 1-02-96. Rev. 0 This TM is to gag closed the minimum flow valves on the Unit 1 RHRSW system.

This is to assure that full flow will be received by the heat exchangers in an accident scenario. A low-flow bypass is provided from the pump discharge to the intake structure. The bypass flow is required to prevent the pump from overheating when pumping against a closed discharge valve. A pressure control valve limits the bypass flow. Therefore the minimum flow lines are not required

for this particular system and will not reduce the margin of safety as defined in the basis for any TS.

TM 2-02-17, Rev. 0 This TM is to gag closed the minimum flow valves on the Unit 2 RHRSW system.

This is to assure that full flow will be received by the heat exchangers in an accident scenario. A low-flow bypass is provided from the pump discharge to the intake structure. The bypass flow is required to prevent the pump from overheating when pumping against a closed discharge valve. A pressure control valve limits the bypass flow. Therefore the minimum flow lines are not required for this particular system and will not reduce the margin of safety as defined in the basis for any TS.

TM 2-03-07. Rev. N/A This TM is to install a temporary test box (jumper and two position switch) to allow simulating a control rod withdrawal during the rod interlock and other refueling interlock surveillances. Since the test box will not be in place during power operation, and since removal of the box and the restoration of the box are being controlled and documented per approved administrative control procedures, installation of this test box will not increase the likelihood of some type of failure in the RPIS or RMCS systems. Therefore, the proposed change will not result in a decreased effectiveness of the surveillances.

DESIGN CHANGE REQUESTS (DCR' DCR 1-97-041 Rev.1 This design change is for the replacement of meteorological data recorders with digital recorders. The meteorological data collected from the Primary and Backup towers is used to provide meteorological information to determine the potential influence of the plant and its operation on local meteorology and is used to help assess the plant personnel's state of readiness during Emergency Preparedness (EP) drills. The type of data recorded includes wind speed, direction and standard deviation in direction at various elevations, dew point, precipitation, ambient temperature and temperature difference at various elevations.

The data collected from the two meteorological towers is transmitted to EOF control panel 1Y33-P101 where it bifurcates, sending the signals to equipment in the MCR and to equipment in the EOF.

In the Emergency Operating Facility the paper chart recorders, four Esterline-Angus recorders (IY33-R001, R002, R003, R004) and six Texas Instruments TIGraph 100 recorders (1Y33-R005, R006, R007, R008, R009 and R010) store

the meteorological data. These recorders are located in panels 1Y33-P100 and 1Y33-P101. Additionally, the data is sent to the Meteorological Information Dose Assessment System (MIDAS) computer located inside in the EOF. Parallel wiring from the incoming terminal blocks allows signals to be sent directly to the MIDAS computer. The recorders are unreliable and require more than normal maintenance. They are obsolete recorders with spare parts virtually unavailable from the vendor. Daily readings are obtained from the MIDAS. The EOF recorders are more of a burden to maintain which out weighs the benefit received from them and are essentially not used. During Emergency Preparedness (EP) drills, the existing paper-chart recorders are used as the 4th backup source for data behind the MIDAS, SPDS, MCR recorders and the "Communicator Recorder."

The 10 recorders (1Y33-ROO1 through 1Y33-ROIO) are not relied upon for critical data and will be removed from the EOF control cabinets and not replaced. They will be stored and used as a source for spare parts for other recorders of the same models.

In the Main Control Room 4 Esterline-Angus recorders (IY33-R601, R602, R603, R604) and 6 Texas Instruments (TI) TIGraph 100 multipoint recorders (1Y33-R605, R606, R607, R608, R609, R610) are being replaced by five, six channel Yokogawa paperless recorders. These recorders are located in panel iHI 1-P689 and IHI I-P690. The MCR recorders are Reg. Guide 1.97 category three instruments. EMI/RFI filters will be installed on the power supplies of each new recorder.

The replacement recorders are Yokogawa model DX106 DAQSTATION capable of displaying 6 real-time digital data points simultaneously or individually in a wide variety of display modes, bar graph, historical trend, large font numeric, data table or combination. The user friendly high-resolution color liquid crystal displays with bar graphs or continuous display allow quick and easy assessment of the changes in the radiological data. The' same input data as presently exists will be displayed on the new recorders.

The process of recording meteorological data in the MCR on recorders remains unchanged in the DCR and no new activities will be created or initiated by replacing the type of recorder on which data is recorded. The new paperless Yokogawa recorders function to record and display meteorological data, as did the existing recorders. No new circuits are introduced or changed by replacing the recorders and no output circuits are added. The malfunction or mis-operation of any meteorological data recorder can only affect that recorder itself. It cannot create an accident of a different type and will not affect data being sent simultaneously along parallel paths to other data logging equipment. Any undesired data which may be received will not cause the recorders to malfunction.

EMI/RFI concerns have been addressed by use of shielded cables for all inputs to the recorders with EMIIRFI filters placed on all recorder power supplies. No new failure modes or accident scenarios are introduced by this modification.

DCR 2-97-012 The activity to which this evaluation applies involves the addition of a Battery Test Breaker and Battery Test Receptacle to both of the distribution panels 2R25-S004 and 2R25-S006. This design change will install a class 1E breaker and new test receptacle adjacent to each 125VDC distribution panel 2D (2R25-S004) and 2F (2R25-S006). Permanent cables will be routed from each 125VDC distribution panel bus bars to each breaker, and from each breaker to the test receptacles.

Mounting for the new supports and equipment will be done in accordance with Seismic Category I or II/I design requirements for safety-related and nonsafety-related applications respectively. Installing the new breakers, test receptacles and cable will eliminate the need for removing the cabinet cover and exposing field personnel to the energized parts of the cabinet. The class l E breakers will serve as an isolation point between the non-class IE test receptacles and the class IE distribution panels 2D (2R25-S004) and 2F (2R25-S006). This provides protection against short circuits at the new test receptacles connected at those terminals. The test receptacles will provide a safe and convenient location for field personnel to connect test equipment and perform battery testing at the Diesel Generator batteries. If the new Westinghouse 150 AMP breakers should trip while Diesel battery testing is being performed, field personnel will replace the 200 amp fuses located in the battery fuse boxes 2A (2H21-P291) and 2C (2H21-P293). The additional breakers and test receptacles will not add any additional load to their respective class 1E distribution panels.

The new breakers are class 1E qualified, they will provide short circuit protection and serve as an isolation point between the non-class 1E test receptacles and the class 1E distribution panels 2D and 2F. The new and modified electrical equipment and their supports are qualified by analyses or tests to meet Seismic Category I and II/I requirements as described in the FSAR so as to preclude their failure as a result of a seismic event. Neither the new breakers nor the test receptacles affect any structure, system, or component that initiates any previously evaluated accident. This design modification does not alter the intended function of the distribution panels 2D and 2F. No safety limits will be reduced as defined in the Technical specifications.

DCR 1-99-042 New Yokogawa DX104 recorders will be installed in place of the Leeds &

Northrup Speedomax recorders, IT48-R601A / B, in the same locations on the panels and using the existing wiring. The MPL numbers will be re-used. The cable will be de-terminated from the existing recorders and re-terminated according to the requirements of the new Yokogawa recorders. No functional changes are being made with the recorder replacements.

New adapter plates will be required since the Yokogawa recorders are slightly smaller than the existing recorders. The new Yokogawa DX1 04 recorders have

been qualified for Seismic Category I conditions per IEEE 344-1975, for EMI/RFI per EPRI Document TR-102323, and for firmware validation and verification per IEEE 7-4.3.2 Annex D. The existing shielded signal cables will be re-used with the new Yokogawa recorders.

This modification provides for a functionally equivalent replacement of the obsolete L & N Drywell wide range radiation and pressure monitoring recorders with Yokogawa digital paperless recorders. The intended function of the Drywell wide range radiation and pressure monitoring is not altered by this design change.

No safety limits will be adversely affected by this design modification such that there is no reduction in the margin of safety.

DCR 1-99-046 Rev.2 This modification replaces the obsolete Leeds & Northrup MSR, RPV, MS SRV Discharge, Reactor Building HVAC and MG set cooling and ARM Systems temperature and radiation monitoring recorders with functionally equivalent Yokogawa digital paperless recorders. This design modification is considered a digital upgrade and has been evaluated for EMI/RFI concerns. Power line filters will be placed on the power supplies for the recorders and the existing shielded signal cables will be used for all signal inputs to the recorders. The new Yokogawa DX200 series recorders will be installed in place of the Leeds &

Northrup model 250 recorders in the same locations on the respective panels. The same MPL numbers will be re-used for the replacement recorders. The existing cables will be de-terminated from the existing recorders and re-terminated according to the requirements of the new Yokogawa recorders. No functional changes are being made with the recorder replacements. A cable review was performed and found that all associated circuit resistances are within the limit regarding the new recorder requirements The intended functions of the MSR, RPV, MS SRV Discharge, Reactor Building HVAC and MG set cooling and ARM Systems temperature and radiation monitoring are not altered by this design change. No safety limits will be adversely affected by this design modification such that there is a reduction in the margin of safety.

DCR 99-049 This DCR is for the replacement of the Unit 1 SPDS system. The staged replacement of obsolete SPDS components consists of replacing, one at a time, the existing Cutler-Hammer multiplexers and Rolm input/output interfaces (and isolation devices where required) with new RTP 2000 input/output (I/O) racks and related ancillaries. The existing Rolm and Cutler-Hammer devices are configured on a type-of-input basis, such as class lE analog signals, non IE analog signals, class IE digital inputs, non IE digital inputs, etc. The removal of a specific existing I/O device will disable its respective inputs from certain SPDS displays.

The replacement I/O racks are manufactured by RTP Corporation, a company known for providing reliable equipment and long-term product support. RTP identifies their racks as "target nodes." For the purpose of this DCR this designation will be shortened to "node" and will be defined as one RTP controller installed in a rack with one or more input cards. The proposed modification consists of installation of eight nodes to replace the functions of the existing Rolm I/O interfaces, the Cutler-Hamnmer multiplexers, and the Rochester 1E to non 1E isolators for the Division I and Division II digital inputs.

For the interim SPDS configurations, the software that currently runs on the Rolm SPDS and ERF processors will be copied to the Data General MV-9600 NRC-ERDS computer. Displays will be drawn on personal computers and monitors installed in the main control room, using data supplied via the Plant Process Computer Local Area Network (PPC LAN). The RTP nodes will also communicate with the MV-9600 via the PPC LAN. As each RTP node is installed, a software change will be installed that will enable the MV-9600 to merge the new node's data into its internal database for display on the personal computer monitors installed in the main control room.

When the complete replacement system is installed, the functions of the Rolm SPDS processor, the ERF processor and the MV-9600 will be replaced by networked servers running software that provides SPDS data gathering, manipulation, calculation, and storage. The replacement SPDS will provide data to the Southern Company "business LAN" to allow remote display of SPDS information to selected users. The Plant Hatch firewall will control the dissemination of SPDS information outside of the Southern Nuclear portion of the Southern Company Wide Area Network.

All software that is being purchased or developed for the replacement system will be verified and validated to applicable standards. The new software will not degrade the functions or operation of the SPDS in either the interim or the final configurations.'

This change constitutes a digital upgrade. EMI/RFI testing as specified in EPRI TR-102323, Rev. 1 has been completed for all RTP-supplied equipment. Review of the test results has proven that the RTP-supplied equipment is not expected to have an adverse impact to operation of any safety-related or important to safety equipment or systems. Power filters for EMI/RFI reduction will be installed as needed to further enhance protection from EMIVRFI.

Implementation of this DCR will result in changes in heat loading in the main control room and the computer room. These have been evaluated and found to be within the capabilities of the existing HVAC systems for these areas.

Although SPDS provides valuable information to the plant operators, it is not safety-related. The replacement system uses completely different hardware.

However, the new displays will present the operator with the same information about the plant as the existing system and storage of events will be maintained.

The system will maintain full functionality during the various interim configurations, except for brief outages of certain displays during installation of new equipment and removal of old equipment.

Both the existing and the replacement SPDS are monitoring systems only; they have no control functions over safety-related or non-safety-related equipment or systems. All interties to safety-related systems for both the old and the new system are isolated by class lE devices.

Neither the existing nor the replacement SPDS has any control functions over plant equipment. Although the system provides important information to the plant operators, alternate means of obtaining that information are available. The SPDS is not credited for any accident mitigation activities. Therefore, the installation of the new system cannot reduce the margin of safety as defined in the basis for any Technical Specification.

DCR 99-050 This DCR is for the replacement of the Unit 2 SPDS system. The hardware used by the SPDS and the ERFDS is obsolete by today's technological standards.

Replacement parts for certain key components are difficult to obtain and the system will no longer support additional data points above the original design requirements. The proposed replacement system retains certain components that still have vendor parts and service support and replaces those that do not. For the purposes of this evaluation, the term SPDS will be used to describe the functions of both the SPDS and ERFDS.

To minimize SPDS downtime a staged replacement plan is being used for replacement of the obsolete components of the SPDS. This approach provides for non-outage integration of components of the new SPDS into the old system.

Custom software, developed by site staff, will provide the interface for the new system components to function in the existing system. As implementation proceeds, more and more of the obsolete equipment will be replaced with components from the new system, until all the new equipment is installed. The last steps of implementation will be the installation of the final display screen software, the viewer software for remote access to SPDS data, and the removal of original system components that are no longer in service.

This installation approach will result in multiple interim configurations of the system, all of which will maintain full SPDS functionality. During the installation of new system components, some SPDS information will be unavailable on the

displays for a few days. However, this will not impact the safety of the plant, since SPDS is not credited for accident mitigation.

The Plant Computer Room door lock assembly is being replaced to enhance the access control to the room. The new lock assembly will consist of a mechanical combination exterior lock and interior panic or push bar latch assembly to ensure exit capability from the room in the event of a fire or manual release of the carbon dioxide fire suppression system. This is a change of like components and does not actually change the door function or fire rating. The door lock replacement does not adversely impact safe shutdown capability for either Unit or the intended function.

The design functions of the existing SPDS will be retained in the new system, although they will be accomplished using different software and microprocessor-based hardware. Therefore, this modification is considered to be a digital upgrade. The Plant Computer Room Door lock replacement does not alter the design function or adversely impact the design function as described in the Updated FSAR.

The functionality of the NRC ERDS communication link for each Unit will be maintained during the interim configurations of the system via the existing connection between the Data General MV-9600 serial port and the NRC modem.

The new SPDS software will provide full NRC ERDS communication functionality; when the new SPDS hardware and software is operational, the NRC modem connections will be moved to serial ports on the new SPDS server computers.

The replacement SPDS has no control functions over any SSC important to safety previously evaluated in the Updated FSAR. All interconnections between the new SPDS components and class lE equipment will be properly isolated using approved techniques. EMI/RFI testing has demonstrated that the RTP racks are not expected to adversely impact any safety-related or important to safety system.

Other replacement SPDS system components have been evaluated for EMI/RFI effects and EMI/RFI power filtering is being applied for key SPDS components to further reduce the possibility of EMI/RFI.

A failure of any component of the new SPDS is not expected to adversely impact the operation or performance of safety-related or important to safety equipment or systems. The replacement SPDS is not credited for any accident mitigation actions.

DCR 00-035 The proposed change provides design for a modified rail spur extending from the east end of the Unit 1 RR Airlock 94 feet to the east. Loaded and unload HI-STAR 100 and HI-STORM 100 casks are transported along the spur via a

fabricated roller pad. The casks are moved along the rail to a transfer point at the centerline of the roadway. At this transfer point the transporter will lift the Hi-Storm 100 or Hi-Star 100 and carry it to the ISFSI. HNP-2 FSAR Section 9.1.5 describes use of the HI-STAR 100 system and movement of the HI-STAR 100 system into and out of the Unit 1 reactor building via a cradle mounted on a railcar or truck. HNP-2 FSAR Section 9.1.5 will be revised to incorporate use of the roller pad and transporter for movement of HI-STAR and HI-STORM casks.

Additionally, the QA classification of the roller pad will be incorporated into HNP-2 FSAR Section 17.2A.

The use of the roller pad and transporter for movement of the HI-STAR 100 or HI-STORM 100 outside the reactor building does not impact the operability of any equipment required by the Technical Specification to be available for any mode of unit operation. All handling outside meets applicable code requirements specified in HNP-1 FSAR 12.4.4. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

LICENSING DOCUMENT CHANGE REQUESTS LDCR 2002-069. Rev. 0 This LDCR makes changes to the Hatch Nuclear Plant (HNP) FSAR to implement the 18 to 24-Month Fuel Cycle Extension Project. Technical Specifications Amendment No. 232 and 174 and separate LDCRs have been issued to change the Technical Specifications and Technical Requirements Manual (TRM) surveillance requirements. A No Significant Hazards Analysis performed the amendment request, and 50.59 evaluations were performed for the TRM changes. These FSAR changes are necessary to implement the changes for the Technical Specifications and the TRM.

Only the changes made to the FSAR that are not specifically addressed in the Technical Specifications and TRM change packages will be addressed by the 50.59 evaluation. The No Significant Hazards Evaluation of HL-6117 or the 50.59 evaluations within the LDCRs for the TRM changes have justified all other FSAR changes.

The first FSAR change to be addressed is a change in the LPCI Valve Select Timer field settings. The LPCI Valve Select Timers keep the operators from throttling or closing either of the RHR outboard injection valves for a minimum of 10 minutes after a LOCA. Because of the observed instrument drift associated with these timers, the nominal field setting is changed from 10 to 11 minutes.

Therefore, initiation of drywell and wetwell sprays is delayed until approximately 11 minutes after the start of the LOCA. The added delay has been reviewed and

found to be insignificant in providing required accident mitigation. The change ensures that adequate LPCI flow is maintained as required during the early phases of a LOCA, and is acceptable.

The second FSAR change is a change in the required frequency to update the Offsite Source Voltage Study. The FSAR states that maximum loadings on the startup transformers IC, ID, 2C and 2D are verified in the Offsite Source Voltage Study, which is updated approximately every 18 months. The intent of this requirement is for the study to be updated after each major design implementation period, which is typically during refueling outages. Therefore, this requirement is changed from "approximately every 18 months" to "on a frequency approximately corresponding to the refueling frequency."

The third change to the FSAR to be made is the clarification of Reg. Guide 1.52, with respect to the required frequency of Ventilation Filter Testing. Reg. Guide 1.52, Rev. 2 recommends an 18-Month surveillance interval. The Regulatory Guide does not discuss any specific failure mechanisms or degradation factors that were the basis for specifying 18 months. ASME N510-1989 specifies a recommended frequency of once per operating cycle. Therefore, the 18-Month surveillance interval is interpreted as once per operating cycle.

The proposed FSAR change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR or of a malfunction of a structure, system, component (SSC) important to safety previously evaluated in the Updated FSAR. It also does not result in creating the possibility for an accident of a different type than previously evaluated or creating the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated. The proposed change also has no impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment.

LDCR 2003-076 This LDCR is to revise the Unit 1 and Unit 2 Technical Specifications Bases B 3.7.5 to remove the discussion of the automatic swap of PSW cooling water to the "B" MCR AC unit; add a description to Unit 1 FSAR section 10.7.6 for the manual action, in place of the automatic action, for providing cooling water to MCR AC unit lZ41-BO08B from PSW Division II; and add component condensing unit," its malfunction and comments to the MCR HVAC Systems Failure Analysis Table as the result of the manual action. The automatic function is a design feature of the B MCR AC train that was included for additional defense-in-depth. Retention of this feature is not necessary for operability of the MCR AC units. Items which may impact the probability of a failure in the PSW system, such as design, material, construction standards and requirements, surveillance and preventative maintenance activities and frequencies, are

unaffected by this change. The system design, construction, maintenance, and overall operation will remain as described in the FSAR. Therefore, the probability of system failure remains unaffected as well.