ML081840139

Submittal of 10 CFR 50.59 Annual Report

ML081840139
Person / Time
Site:	Oconee
Issue date:	06/30/2008
From:	Baxter D Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML081840139 (19)

Text

Dukerg Duke POconee DAVEPresident Vice BAXTER Nuclear Station Duke Energy Corporation ONO1 VP/7800 Rochester Highway Seneca, SC 29672 864-885-4460 864-885-4208 fax dabaxter@dukeenergy. com June 30, 2008 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1, 2, and 3 Docket Nos. 50-269, 50-270, 50-287 10 CFR 50.59 Annual Report Attached are descriptions of Oconee facility changes, tests, and experiments which were completed subject to the provisions of 10 CFR 50.59 between January 1, 2007, and December 31, 2007. This report is submitted pursuant to the requirement of 10 CFR 50.59 (d) (2).

If there are any questions, please contact Graham Davenport at (864) 885-3044.

Very truly yours, Dave Baxter Site Vice President Oconee Nuclear Site Attachment www. duke-energy com

U. S. Nuclear Regulatory Commission June 30, 2008 Page 2 xc: Mr. Luis Reyes Regional Administrator, Region II U. S. Nuclear Regulatory Commission Atlanta'Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303 Mr. Andy Hutto Senior NRC Resident Inspector Oconee Nuclear Site Mr. Lenny Olshan Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555

U.S. Nuclear Regulatory Commission June 30, 2008 Page 1 of 17 Type: Nuclear Station Modification / UFSAR Revision (OD300221 / UFSAR Change #

06-22)

Title:

Restore Reactor Coolant Wide Range Pressure to Diverse Scram System (DSS) via Integrated Control System (ICS)

==

Description:==

Train 1 of DSS has failed, due to failure of the Inadequate Core Cooling Monitor (ICCM) train A pressure signal from 3RCPT0224, due to capillary line failure. This failure has rendered DSS inoperable and Selected Licensee Commitments (SLC) 16.7.2.B is entered. OD300221 has been initiated to restore the DSS Train 1 to service by providing another Reactor Coolant System (RCS) pressure input to the Automatic Trip Without Scram (ATWS) Train I Programmable Logic Controller.

(PLC). In order to declare DSS returned to service and subsequently to operable status, positive assurance must exist that the new RCS pressure input can meet all of the design criteria for DSS. The design requirements include diversity from the Reactor Protective System (RPS) and sufficient accuracy to assure that the RCS is protected from overpressure from an ATWS event.

Evaluation: This 50.59 evaluation is to consider the use of different instrumentation as input to the DSS assumed in the UFSAR to determine if use of these instruments is acceptable to meet the operability requirements of Selected Licensee Commitments (SLC) 16.7.2.

The DSS is described in some. detail in the UFSAR including discussion of the particular RCS pressure transmitters used for input. This design change will affect the RCS pressure inputs to the DSS; therefore, the UFSAR will require revision due to this change.

The ATWS PLC equipment and associated power supplies and output devices are not affected in any manner by this design change, and all descriptions and assumptions regarding the ATWS equipment, excepting the RCS WR pressure input, are completely unaffected by this change.

The NRC SER for Oconee ATWS (11/29/1989) describes the equipment diversity requirements for RPS and DSS and details the features of DSS that demonstrate conformance to the ATWS rule. None of the assumptions or descriptions of equipment details in the SER are made invalid by this change. This change is made under 10CFR50.59 in accordance with the SER.

Based on the above, there were no safety concerns. No Technical Specification, Bases or Selected Licensee Commitments need to be

U.S. Nuclear Regulatory Commission.

June 30, 2008 Page 2 of 17 changed due to this Design Change. There is an UFSAR change required.

Prior NRC review,and approval is not required.

Type: Nuclear Station Modification / UFSAR Revisions (OD300444 / UFSAR Change #

07-20 & 07-21)

Title:

Control Room and Plant Chart Recorder Upgrade.

==

Description:==

The purpose of this design change for Unit 3 will remove and replace, where required, obsolete chart recorders from the control boards. With the removal of obsolete recorders, and with many of the inputs routed to the Operator Aid Computer (OAC), additional monitoring capability of OAC points is needed. To provide for this, a new multi-system monitor is installed in both 3UB2 and 3AB2A. A new OAC workstation is installed in 3AB2. The existing 3AB2 and 3AB2A bench boards are replaced with new cabinets that are more functional for the keyboard and monitor installation and usage.

Some of the obsolete recorders slated for replacement are designated as QA-1 post accident monitoring (PAM) devices and are not isolated from their safety related PAM signal source. Replacement recorders are not QA-1 but are isolated from safety circuits with qualified isolators.

Evaluation: Design change OD300444 replaces selected obsolete chart recorders. A total of eleven new multi-point graphics recorders are installed in the Control Room. The existing bench boards are replaced with new cabinets that are more functional for keyboard and monitor installation and usage.

Direct and immediate trend of the information recorder by obsolete recorders was never considered essential for operator information or action, nor is the recording device the primary indicator for the recorded parameter. Regulatory Guide 1.97 requirements for recording of parameters are met post-modification.

New safety to non-safety (or QA-1 to non-QA-1) interfaces are isolated using QA-1 isolation devices. New devices/components installed by this design change are not required to be environmentally qualified since they are located in a mild environment. Field work is performed using QA-1 procedures when required (Control Room panel modifications are considered QA-1). Mounting of new recorders is QA-4; therefore, there are no seismic interactions between seismically qualified and non-seismically qualified structures, systems or components. A control board seismic review for the control board changes/deletions associated with this

U.S. Nuclear Regulatory Commission June 30, 2008 Page 3 of 17 design change has been completed with a determination that the control boards are not adversely affected. A 10CFR50 Appendix R fire review was performed for the design phase with no adverse affects identified.

The components installed by this design change are not initiators of any analyzed accidents do not introduce new failure modes, and are not relied upon to mitigate an accident. The failure of these components does not prevent safety-related SSCs from performing their designed functions.

Based on the above, there were no safety concerns. There are Technical Specification Bases, Selected Licensee Commitments and UFSAR changes due to this Design Change. Prior NRC review and approval is not required.

Type: Nuclear Station Modification (NSM ON-13098 / UFSAR Change # 07-30)

Title:

Upper Surge Tank Inventory Protection

==

Description:==

This NSM will make some modifications to eliminate single active failures associated with the Upper Surge Tank (UST). These modifications involve both "active type" and "passive type" isolation.

"Active type" isolation requires a signal to be sent to a valve to close to assure flowpath isolation. "Passive type" isolation uses a check valve to prevent reverse flow. The changes for the active and passive types are described later in this section.

This modification is to add four air-operated valves (AOV) that automatically close when the UST level drops below 7.5 feet. New valves 1C-903 and 904 are to isolate flow to the hotwell and to the Auxiliary Boiler Feedwater (FDW) pump. A bypass valve, 1C-912, is provided around valves 1C-903 and 904. Valves 1C-906 and 907 isolate flow to the Powdex Backwash Pump. Both the Auxiliary Boiler FDW pump and Powdex Backwash pump will have pressure switches in the suction line that will trip them on low suction pressure. These pressure switches are to be installed on the first unit that implements the automatic isolation valves C-903, 904, 906, and 907.

The modification is also to add a new Condensate Recirculation path to the UST. This path will allow flow from the Condensate Booster Pump suction line to the UST Riser. A manual throttle valve, 1C-899, and flow indication (locally and on the Operator Aid Computer) are provided. This new Condensate Recirculation path should provide significant operational flexibility during unit start-ups.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 4 of 17 Evaluation: There are currently three somewhat independent paths of water from the UST to the condenser. The modification will combine the supplies to these paths with a common header. With the existing design for normal operating conditions, one of the AOVs can fail closed (e.g.,

loss of air to valve operator) and this failure will notnecessarily prevent another of the AOVs from opening (or remaining open) to supply water from the UST to the condenser hotwell if there is a low hotwell level. In the new design, a failure of one of the new AOVs could prevent water from transferring to the hotwell from the UST through this pathway. The safety function of the valves is to close on low UST level. During certain accidents/events, the valves in both the current and proposed new design will receive a close signal on low UST level. In the new design, the failing closed of one of thenew AOVs (1C-903 or IC-904) can stop the flow of water from the UST to the condenser. The UFSAR describes the three separate pathways from the UST to the condenser hotwell, but the context of this wording is in relation to the pathways being automatically isolated on a low UST level. In the unlikely event that one of the new AOVs does inadvertently close, manual actions could be taken to open the bypass around the new valves. If this action was not taken and the level dropped significantly, a trip of the unit could occur. The loss of air to the new AOVs could occur due to a loss of the non-safety related Instrument Air System or due to a loss of air locally at the valve (e.g., loss of the safety related power supply to the solenoid valve on one of the new valves. A loss of the Instrument Air System would result in the same effect in both the existing and new design. The effect would be that the flowpath(s) from the USTs to the hotwell would be isolated due to fail closed AOVs. If the air supply is lost to one of the new valves (1C-903 or 1C-904), then all three flowpaths would be isolated. But, there are manual actions that could be taken to bypass the failed closed AOV. If the hotwell level is not replenished over time, a trip of the unit could occur.

Thus, there is a possible'increase in the potential for a turbine/reactor trip if makeup to the hotwell is not able to be achieved. But there are other means for a reactor/turbine trip to occur. UFSAR Section 15.8 provides a number of means for a turbine trip. The potential cause for a turbine trip is described as including a generator trip, low condenser vacuum, loss of turbine lubrication oil, turbine thrust bearing failure, turbine overspeed, main feedwater pump trip, high steam generator level, or a reactor trip.

The loss of hotwell level could, if low enough, cause the main feedwater pumps to loose suction pressure and ultimately trip. The trip of the main feedwater pumps are listed in the UFSAR section described above. This small potential for a localized loss of air to one of the new valves is considered to be a negligible increase to the overall turbine trip potential and thus is not a

U.S. Nuclear Regulatory Commission June 30, 2008 Page 5 of 17 "more than a minimal increase" in the frequency of occurrence of an accident previously evaluated in the UFSAR.

The two AOVs (1C-903 and 1C-904) in the common header from the USTs to the hotwell will isolate the UST if a low UST level is detected or if air is lost to the valves. These two AOVs are to be used as the QA-1 Class F boundary so that the UST tank contents will be isolated even in the event of a single failure. The potential for "more than a minimal increase" in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR was investigated with respect to the UST's makeup going into a common header before going to the three separate pathways. Equipment important to safety affected by this modification includes the UST (assured source of Emergency Feedwater (EFW), condenser hotwell (one of the potential long term sources of EFW), and EFW System (provides feedwater in the event of a loss of main feedwater). There are currently three somewhat independent paths of water from the UST to the condenser. The modification will combine the supplies to these paths with a common header. Although flow from the UST to the condenser hotwell is not required to be designed to withstand a single failure, the potential for "more than a minimal increase" in likelihood of occurrence of a malfunction of an SSC important to safety needs to be considered. With the existing design for normal operating conditions, one of the AOVs can fail closed (e.g., loss of airto valve operator) and this failure will not necessarily prevent another of the AOVs from opening (or remaining open) to supply water from the UST to the condenser hotwell if there is a low hotwell level. In the new design, a failure of one of the new AOVs could prevent water from transferring to the hotwell from the UST through this pathway. The safety function of the valves is to close on low UST level. During certain accidents/events, the valves in both the current and proposed new design will receive a close signal on low UST level. In the new design, the failing closed of one of the new AOVs (1C-903, 1C-904) can stop the flow of water from the UST to the condenser. The UFSAR describes the three separate pathways from the UST to the condenser hotwell, but the context of this wording is in relation to the pathways being automatically isolated on a low UST level. The "important to safety" aspect of the condenser hotwell is its function of supplying an EFW supply of water after the UST source has been exhausted. The flowpath from the condenser to the EFW pumps is not adversely affected with the new valves since that flowpath is not used when supplying the EFW pumps via the hotwell. The "important to safety function" of the existing AOVs and the new AOVs in the new design is considered to be their closure on low UST level. This function is enhanced in the new design. Supplying the hotwell from the UST is considered more of an

U.S. Nuclear Regulatory Commission June 30, 2008 Page 6 of 17 operational issue versus an "equipment important to safety" issue. Thus, the use of a common header with two AOVs in series is not considered to cause a "more than minimal" increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

The EFW System is used to mitigate accidents involving the loss of main feedwater. The modification will not change the design function of the EFW supply sources as evaluated in the UFSAR. Thus, in an accident involving loss of main feedwater, the EFW System will still be able to mitigate the event as currently described in the UFSAR. There is no adverse effect on containment integrity and no new release paths are created.

The UST will be designed to provide a source of water to the EFW System even in the event of a single failure. The hotwell backup source is not designed to provide the additional EFW water supply in the event of a single failure. The flowpath from the UST to the hotwell is not required to be designed to withstand a single failure for the function of allowing water to flow. This path is designed such that a single failure does. not allow UST flow to be depleted to th& Hotwell. Thus, the EFW function is not adversely affected with respect to mitigating loss of feedwater scenarios previously evaluated in the UFSAR.

Based on the above, there were no safety concerns. No Technical Specification or Selected Licensee Commitments are required. UFSAR changes are required due to this Design Change. Prior NRC review and approval is not required.

Type: Nuclear Station Modification (OD501924 / UFSAR Change # 07-22)

Title:

Temporary Chilled Water to Reactor Building Auxiliary Coolers

==

Description:==

The Reactor Building Cooling Units (RBCUs) and Reactor Building Auxiliary Coolers (RBACs) are the reactor building cooling (RBC) systems and are normally operating. The RBCU is safety related and QA-

1. The RBACs are non-safety and non QA-1. Both systems are normally cooled by the Low Pressure Service Water (LPSW).

This evaluation is for OD501924, which is a design change that does not involve field work. OD501924 is a design package that will address manually isolating LPSW and using a temporary chiller and associated

U.S. Nuclear Regulatory Commission June 30, 2008 Page 7 of 17 polyethylene piping to provide chilled water to the RBACs during Modes 1-4. The design package contains red-marked Design Basis Document pages to reflect this operation and also includes requirements for the use of the temporary chilled water system. Only one nuclear unit can be provided with chilled water from the temporary chilled water system at any time. Previous design changes addressed the use of a temporary chiller and associated polyethylene piping to provide chilled water to the RBACs and RBCUs in Modes 5,6, and No Mode. The previous design changes also addressed connecting a temporary chiller and piping to the RBACs and RBCUs during any mode. These previous design changes are not part of the scope of the design change being evaluated. The use of a temporary chiller for supplying chilled water to the RBCUs in Modes 1-4 is not being addressed and is not allowed.

Flow diagrams OFD-124B-1.2, OFD-124B-2.2, and OFD-124B-3.2 show the existing allowed hbok-up connections for Unit's-1,2,and 3 respectively. The chilled water will be hooked up downstream of the LPSW supply manual isolation valve (LPSW-1051) to the RBAC and returned upstream of LPSW return manual isolation valve (LPSW-1065).

All aspects of the instrumentation, controls & interlocks regarding the RBACs will not be changed. The RBAC air operated containment isolation valves (in board (LPSW- 1054 & LPSW-1055) and out board (LPSW-1061 & LPSW-1062)) are between the manual isolation valves and will be maintained opened and active in automatic. The RBAC isolation circuitry, flow instrumentation, and alarm will also be maintained operational. The instrumentation and alarm will be used to monitor the functionality of the chilled water circulating through the RBACs.

Recent station mods for each unit have modified the LPSW supply and return lines. These mods physically separated the LPSW supply & return to the RBACs & RBCUs since the RBAC piping was not QA-1. The mods addressed and reduced waterhammer magnitude(s) that could be induced due to concerns presented in G.L. 96-06. Descriptively, the mods separated the LPSW piping associated with the RBACs from the LPSW supply to the "B" Reactor Building Cooling Unit (RBCU). The mods tied the RBAC piping to the LPSW supply and return main headers in the Auxiliary Building by utilizing two penetrations (Pen. 63 & 64). Two air operated valves per penetration (supply header and return header) were provided as containment isolation valves for the RBAC supply and return piping that penetrates the Reactor Building to ensure that the integrity of the containment is maintained following a design basis accident. Valves, flow measurement orifices, instrumentation, test connections and flanges were also provided. The portion of the RBACs piping inside the containment is no longer credited as a containment barrier.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 8 of 17 During times when LPSW temperature is high and greater cooling is desired inside the Reactor Building, a temporary chilled water system may be used to supply chilled water through portions of the LPSW piping this provides water to the Reactor Building Auxiliary Coolers (RBACs) or RBCU "B". The primary intent for using the temporary chilled water system to the RBACs during modes 1-4 is to allow the Reactor Building to be cooled prior to a refueling outage. It could also be used to provide additional cooling in the event of a failure of one or more of the RBAC coils. Though not specifically prohibited, it is not intended as a replacement for the normal LPSW cooling water supply to the RBACs.

Evaluation: The RBACs are used to provide normal containment ventilation. The RBACs and the temporary chiller (thatwill provide cooling to the RBACs) will not initiate a transient or accident. The RBAC is NOT credited for any Oconee Design Basis accidents, accidents and events in UFSAR Chapter 15, and other Design Events and Accidents. This design change does not change the function of the RBACs or affect the operation of the RBCUs such that it affects the containment and/or containment responses during normal or accident condition(s).

The potential for an increase in an Auxiliary Building flooding event was also considered. Based on licensing correspondence, Auxiliary Building flooding was determined to be caused by large continuous sources, not closed loop systems. Since the new piping is a closed loop system that can not flood safety related or important to safety equipment, it is not considered an Auxiliary Building flood initiator and thus does not increase the AB flood frequency.

The temporary chilled water system is a closed loop system. It is isolated from its makeup source. The system is of limited inventory and can not, flood safety related equipment or equipment important to safety.

The RBACs are normally operating during power operation. Their function is to supplement the RBCUs to assist in ensuring the reactor building condition(s) and specified limits aremet during normal operation.

LPSW flow to and from the RBACs is automatically isolated by containment isolation valves on Engineered Safeguards (ES) signal. The containment isolation valves are also automatically closed on low LPSW supply header pressure to assist in limiting the magnitude of waterhammers on LPSW restart. The temporary chilled water system is a closed loop system which is isolated from the LPSW system. In a loss of

U.S. Nuclear Regulatory Commission June 30, 2008 Page 9 of 17 LPSW or Loss Of Offsite Power (LOOP) voiding will not occur therefore there is no potential for a waterhammer.

The temporary chiller is located in the yard. Temporary piping is routed from the chiller to the flanged pipe stubs in the East Penetration wall.

From there, the pipe is routed to either a connection on the LPSW piping going to the RBACs (via valves 1/2/3 LPSW-1052 and 1064, a pipe connection on the LPSW piping going to the 'B' RBCU (via valves 1/2/3 LPSW-878 and 879), or both. Previous modification packages have documented an acceptable routing for the polyethylene pipe and the acceptability of tying the temporary piping into the seismically qualified LPSW piping and the pipe stubs in the East Penetration wall for all three units. The analysis performed assumed the pipes were filled with water.

Thus, the applied loads will be the same for modes 1-4 as they are in

'modes 5, 6, and No Mode. Requirements are specified in the design package red-marked up DBD to maihtain the seismic qualification of the LPSW piping once the polyethylene piping is installed. Requirements are also specified to inspect the pipe to avoid unacceptable conditions.

Supplying chilled water to the RBCUs is beyond the scope of this evaluation.

The temporary piping was reviewed for seismic interactions. The temporary piping that is not installed on the floor is not routed over any safety related or QA-1 components. The temporary piping that is installed on the floor is not located near any safety related or QA-1 components.

Requirements for installing and supporting the piping are documented on the piping layout drawings. Therefore, with regard to the pipe routing, there are no seismic interaction concerns with the design change.

The temporary chilled water system is not considered as having a design function if it does not allow equipment important to safety to be flooded if it breaks. No safety related or important to safety equipment would be flooded if all the contents of the temporary chilled water system were released to the Auxiliary Building.

The manual isolation boundary valves used during the containment isolation leak test are the valves that are used to isolate the LPSW System from the closed loop created by the temporary chilled water system.

Leakage past these valves is not expected since they are indirectly tested as part of the leakage testing of containment isolation valves 1,2,3 LPSW-1054, 1055, 1061, and 1062.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 10 of 17 Having the manual valves closed will also assist in minimizing the effects of waterhammer that could be caused by filling of the voided line following the restart of the LPSW pump. The closed valves create water filled closed loop system so that voids needed for waterhammer are not created.

The safety analysis does not credit RBACs design functions for meeting the required containment cooling and pressure reduction (to limit radiological releases) and Environmental Qualification (EQ) requirements.

A review by the Safety Analysis Group regarding the impact of chilled water supplied to the RBACs following an accident and any possible

'overcooling' has been performed. The review determined that with the isolation of the chilled water occurring at ES 5 and 6 initiations, any heat removed by the RBACs would be comparably small, and would be confined to the first few seconds of the transient. The isolation will not impact the reactor building conditions at the time of switchover (transfer from BWST to the sump). Therefore, it was concluded that supplying chilled water to the RBACs has no impact to the analyses in ensuring adequate NPSH for the RBS / LPI pumps.

The LPSW required flow to safety related components (e.g., RBCUs, LPI) is not adversely affected due to this change. By design and required accident mitigation responses, LPSW to the RBACs will be isolated.

LPSW Flow to non-required loads (e.g., RBACs, Reactor Coolant Pump bearing and motor coolers during normal operation) is also not adversely affected.

The portable chiller is powered from the 4160 volt, B4T switchgear. This switchgear provides power to non-safety related loads. Any problems encountered with the chiller will not cause a problem with plant safety related power systems.

The maximum allowable pressure differential across the containment isolation valves is 170 psid. The maximum operating pressure of the temporary chilled water system is 100 psi. Since the differential pressure applied to these valves is less than the maximum value, operation of the chilled water system during modes 1-4 will not have an adverse impact on the operation of the containment isolation valves 1,2,3 LPSW- 1054, 1055, 1061, and 1062.

If a feedwater line break were to occur in the East Penetration Room, the temporary piping is not within the target area for impingement. Therefore, the temporary piping will remain intact during this event.

U.S. Nuclear Regulatory. Commission June 30, 2008 Page 11 of 17 A Main Steam Line Break (MSLB) is postulated to be within the target area of the temporary piping in the East Penetration Room per High Energy Line Break (HELB) group.

The steam could damage the piping such that it could empty its entire contents onto the floor of the room.

Cooling water to the RBACs is not required for tornado mitigation. Thus, the temporary chilled water supply to the RBACs is not required and does not need to be protected from a tornado.

The loss of the contents of the chilled water system, including the applicable portions of the LPSW system in the Reactor Building and the contents of the chiller skid, into the East Penetration Room will not cause adverse effects to safety related or important-to- safety equipment.

The manual isolation boundary valves used during the containment isolation leak test are the valves that are used to isolate the LPSW System from the closed loop created by the temporary chilled water system.

The Reactor Building Spray (RBS) and Reactor Building Cooling Units (RBCUs) are the credited systems for containment cooling and pressure reduction to limit radiological releases and meet Environmental Qualification (EQ) requirements. None of these functions will be adversely affected by the change of cooling water to the RBACs. It can be concluded that there is not a minimal increase in the consequence of an accident that requires containment integrity or in the consequence of failure of a component or system that could create radiological releases.

Failure of the chilled water system during normal operation is considered no different than LPSW completely isolated to the RBAC due to a failure or a single failure. It should be noted that the RBACs are not required to be designed for the single failure criterion. Failure of an Emergency System (ES) signal to close one set of containment isolation valves will not prevent the second set of containment isolation valves from closing. A single failure will not prevent the low LPSW pressure logic from actuating on low pressure.

The UFSAR states that the LPSW system is a non-radioactive cooling water system that is monitored for radioactivity (section 9.2.2.2.3). Since the temporary chilled water system is in essence an extension of the LPSW system (albeit isolated from the LPSW system), radiation monitoring will be provided for the temporary chilled water system when operated in modes 1-4. Sampling of chilled water will be performed

U.S. Nuclear Regulatory Commission June 30, 2008 Page 12 of 17 routinely and collected at an appropriate location and evaluated for gross radioactivity.

No portion of the temporary piping is located in the Reactor Building.

Leakage of chilled water within the Reactor Building by way of an Auxiliary Cooler coil or LPSW piping during modes 1-4 is of no more consequence than if the leakage were from the LPSW system. Limits are in place that establishes the acceptable leakage in the reactor building (from potential sources). The cooler rupture feature for the RBACs will remain operable with the temporary chilled water system and would detect a sizable leak. Minor leaks would be tracked the same as LPSW leaks which is by monitoring Reactor Building sump rate. Operators would have the option to live with the leak (if small enough), isolate the system by manually closing the containment isolation valves and shutting down the temporary chilled water system.

With the installation and operation of the temporary chilled water system, LPSW flow to the RBACs will be isolated. This LPSW flow will be re-distributed to other loads that the LPSW system supplies. The flows that will be re-distributed are extra flow that will provide some additional cooling capability with no known detrimental effect(s). The discharge pressure at the LPSW pump(s) will be increased slightly due to the isolation of the RBAC flow path. This is not an adverse impact on the pump(s). This is expected when the LPSW is isolated to the RBACs in an event.

Technical Specification 3.6.4 requires Reactor Building pressure to be less than or equal to 1.2psig and greater than or equal to -2.45 psig during power operation. The temperature and pressure in the Reactor Building will be monitored as currently required to ensure pressures are maintained within limits. SLC 16.6.13 specifies the upper Reactor Building temperature. Containment peak pressure analyses are based on a containment average temperature lower limit of 80'.F. The temporary chilled water system is expected to be more efficient in removal heat than LPSW. The upper and lower pressure and upper temperature limits are covered by existing Technical Specifications (TS) or SLCs. The design package specifies the lower Reactor Building temperature limit to be maintained.

The change does not introduce the possibility of a new accident because the isolation of the RBACs will not initiate an accident and/or introduce new failure modes for structures, systems, or components that have design functions. Failure of the chilled water system during power operation will not be different than the loss of LPSW to the RBACs.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 13 of 17 Isolation of the LPSW from the RBACs and chilled water to provide the cooling is not an accident initiator. The alignment will not introduce new failure(s). Installation of the chilled water to the RBACs (during Modes 1-

4) will not introduce a new possibility of a malfunction with a different result because it does not in introduce a failure not previously bounded by those described in the UFSAR.

The fission product barriers are the fuel pellet, cladding, reactor coolant pressure boundary, and containment. The containment isolation function will be maintained by the existing containment isolation valves and the manual valves upstream and downstream of these automatic will be closed to isolate the LPSW. This change in alignment will not alter plant safety limits, setpoints, or design basis limits for fission product barrier therefore does not result in exceeding or altering a design limit for fission product barrier as described in the UFSAR. A review by the Safety Analysis Group determined that the containment analyses were not affected by the temporary chilled water following an accident. Thus, the modification package will not change an evaluation methodology described in the UFSAR.

Based on the above, there were no safety concerns. No Technical Specification or Selected Licensee Commitments Changes are required, but changes to the UFSAR are required due to this Design Change. Prior NRC review and approval is not required.

Type: Nuclear Station Modification

Title:

03C24 Core Reload Design

==

Description:==

The purpose of this change is to replace 73 spent fuel assemblies used in Cycle 23 with 56 fresh fuel assemblies (referred to as Batch 25) and 17 reinsert assemblies from the spent fuel pool. The resulting core will allow restart and operation of Oconee Unit 2 Cycle 23 at full design power (2568 MWth) for a cycle length of 508 +/- 10 EFPD.

No new fuel assembly types are being used in 02C23. The fresh fuel assemblies are type Mk-B 11 A; the third such batch of fuel used in Unit 2.

All of the reinserted fuel from the spent fuel pool is of the Mk-B 11 type.

Sixteen of the reinserts from the spent fuel pool are from the 02C21 discharged assemblies and were selected for insertion on the periphery of the core. One reinsert from the spent fuel pool is from the 02C20 discharge assembly which was selected for the core center position.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 14 of 17 The Batch 25 fuel enrichments are comparable to previous core designs.

The cycle design length is comparable to previous cycles, with both a Tavg reduction and power coastdown utilized to reach the end-of-window bumup. Due to steam generator performance issues, the end-of-cycle Tavg reduction may not occur, which would be compensated for by a longer power coastdown.

Other changes to the core design are nominally expected from cycle-to-cycle. Such changes include initial core loading (Mtu), batch average burnups, and LBP B-4C wt % and reinsertion. 02C23 places second bum fuel beneath the APSRs rather than third bum fuel like 02C22, but both cycles successfully completed the Reload Design Safety Analysis Review (REDSAR) process.

The Reload Change Document also indicates that 02C23 will have minor core power distribution changes with respect to the previous cycle. The magnitude, location, and time in life of the maximum fuel assembly and pin peaking changes from 02C22 to 02C23. Changes such as this, as well as marginal changes in rod worth, critical boron concentration and other physics parameters are expected to occur between cycles and are confirmed to be satisfactory through the Reload Design Safety Analysis Review (REDSAR) process.

The reload is designed using NRC reviewed and approved methods and the 02C23 Core Operating Limits Report comply with Technical Specification 5.6.5.

Evaluation: The subject evaluation was performed to install the core designed for Oconee Nuclear Station Unit 2 Cycle 23. The 02C23 Reload Design Safety Analysis Review (REDSAR), performed in accordance with Nuclear Engineering Division procedure NE-102, "Workplace Procedure for Nuclear Fuel Management", and the 02C23 Reload Safety Evaluation confirm the UFSAR accident analyses remain bounding with respect to predicted 02C23 Safety Analysis Physics Parameters (SAPP), and fuel thermal and mechanical performance limits. The SAPP method is described in topical report DPC-NE-3005-PA.

The 02C23 Core Operating Limits Report (COLR) was prepared in accordance with Technical Specification 5.6.5. This 10CFR50.59 evaluation concluded that no Technical Specification or Bases, UFSAR or Selected Licensee Commitments needs to be changed due to this Core Reload Design Change. Prior NRC review and approval is not required.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 15 of 17 Type: UFSAR Revisions (UFSAR Change # 07-08 & 07-09)

Title:

Revision of UFSAR to incorporate use of Regulation Guide 1.76, Revision 1, for future modifications to SFF related SSCs and other new plant systems and structures which are required to withstand tornado loadings.

==

Description:==

This activity is to revise UFSAR Sections 3.3.2.1 (Applicable Design Parameters), 3.3.3 (References), 3.5.1.3 (Missiles Generated by Natural Phenomena), 3.5.3 (References) and 9.6 (Standby Shutdown Facility), to incorporate tornado design criteria requirements published in RG 1.76, Revision 1. The design of future changes to Standby Shutdown Facility (SSF)-related SSCs and the design of new plant systems (and their associated components and/or structures) required to resist tornado loadings will conform to the tornado wind, differential pressure, and missile criteria specified in RG 1.76, Revision 1. This 50.59 is not to be construed as authorizing any future physical plant modifications.

This 50.59 is addressing the UFSAR changes as a methodology change, as defined in NEI 96-07, Revision 1, and as noted in the accompanying 50.59 Screen. Per NEI 96-07, Revision 1, adverse changes to elements of a method of evaluation included in the UFSAR, or use of an alternative method, must be evaluated under 10 CFR 50.59(c)(2)(viii) to determine if prior NRC approval is required (NEI 96-07 Section 4.3.8). Changes to methods of evaluation (only) do not require evaluation against the first seven criteria.

The tornado design basis for existing plant structures other than the SSF (as given in UFSAR Sections 3.3.2.1 and 3.5.1.3) is not affected by this activity. However, UFSAR Sections 3.3.2.1 and 3.5.1.3 are revised to note that RG 1.76 Revision 1 tornado design criteria will be used for new plant systems and structures required to be designed for tornado loadings. RG 1.76, Revision 1 is also added to UFSAR Sections 3.3.3 and 3.5.3 as a reference. Auxiliary Building loads with respect to tornado missiles, as given in UFSAR Table 3-23 (Auxiliary Building Loads and Conditions),

are not affected by this activity.

The design tornado for the SSF is discussed in UFSAR 9.6.3.1 (Structure),

which provides information that the existing design tornado used in calculating tornado loadings for the SSF is in conformance with RG 1.76 (Revision 0) with the following exceptions:

1. Rotational wind speed is 300 mph.

2. Translational speed of tornado is 60 mph.

3. Radius of maximum rotational speed is 240 ft.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 16 of 17

4. Tornado induced negative pressure differential is 3 psi, occurring in three seconds.

Tornado generated missiles which apply to the SSF design are given in UFSAR Table 9-17 (Design Basis Tornado Missiles and Their Impact Velocities). The SSF is designed to resist the effects of tornado generated missiles in combination with other loadings. Note that the only reference to Table 9-17 in the UFSAR text is in Section 9.6.3.1.

RG 1.76 Revision 1 specifies (for Region I in which Oconee is located) that the design basis tornado will have:

1. Maximum wind speed of 230 mph.

2. Translational speed of 46 mph.

3. Radius of maximum rotational speed is 150 ft.

4. Tornado induced negative pressure differential is 1.2 psi, occurring in 2.4 seconds (rate of pressure drop of 0.5 psi/sec).

5. Maximum rotational speed is 184 mph.

The methodology change addressed by this 50.59 Evaluation is the proposed use of RG 1.76 Revision 1 tornado design criteria for future SSF-related design changes and new plant systems and structures required to withstand tornado loadings, rather than the tornado design criteria presently given in UFSAR Section 9.6.3.1 and UFSAR Table 9-17.

Evaluation: This activity is to revise UFSAR Sections 3.3.2.1, 3.3.3, 3.5.1.3, 3.5.3 and 9.6, to incorporate tornado design criteria requirements recently published in RG 1.76, Revision 1. The RG 1.76 Revision 1 tornado design criteria will apply to future design changes of SSF-related SSCs and to new plant systems and structures required to withstand tornado loadings. This 50.59 is not to be construed as authorizing any future physical plant modifications. This 50.59 is addressing a methodology change only.

The tornado design criteria applied to the design of existing Class 1 structures remain as is. The activity addressed by this 50.59 does not retroactively apply RG 1.76, Revision 1, tornado design criteria to existing Class 1 SSCs required to withstand tornado loads.

The structural aspects of the SSF design, as addressed in the SSF SER, are not changed by the UFSAR changes addressed in this 50.59 Evaluation.

The existing SSF structure is not modified and remains qualified to the tornado design criteria as given in UFSAR Section 9.6.3.1 and UFSAR Table 9-17.

U.S. Nuclear Regulatory Commission June 30, 2008 Page 17 of 17 NEI 96-07, Revision 1, Section 4.3.8, provides guidance on what changes are not considered departures from a method of evaluation described in the UFSAR. One acceptable change allowed is use of a NRC-approved methodology, provided such use is (a) based on sound engineering practice, (b) appropriate for the intended application and (c) within the limitations of the applicable SER (Safety Evaluation Report). Application of RG 1.76, Revision 1, to Oconeefuture SSF-related design changes and to new plant systems and structures required to withstand tornado loadings does not require an SER. In this particular circumstance, the information provided in RG 1.76, Revision 1, provides licensees and applicants with guidance that the NRC staff considers acceptable for use in selecting design basis tornado and design basis tornado generated missiles that a nuclear plant should be designed to withstand to prevent undue risk to the health and safety of the public. The proposed UFSAR changes addressed by this 50.59 will adopt the RG 1.76, Revision 1, tornado design basis criteria in their entirety, without exception, for future design changes to SSF-related SSCs and to new plant systems and structures required to withstand tornado loadings. The application of RG 1.76, Revision 1, as addressed above, is considered to be appropriate and acceptable, and is not considered to be a departure from a method of evaluation described in the UFSAR. Therefore, the UFSAR revisions can be made without a license amendment.

Duke has informed the NRC that it intends to use the guidance provided in RG 1.76, Revision 1, for tornado design basis criteria. The NRC staff agreed that using RG 1.76, Revision 1 was appropriate.

Based on the above, there were no safety concerns. No Technical Specification or Selected Licensee Commitments Changes are required, but changes to the UTFSAR are required due to this Design Change. Prior NRC review and approval is not required.