View source for ONS-2017-074, Proposed Amendment to the Renewed Facility Operating Licenses Regarding Revisions to the Updated Final Safety Analysis Report Section Associated with Tile Standby Shutdown Facility License Amendment Request No. 2017-03

{{Adams
| number = ML17299A125
| issue date = 10/20/2017
| title = Proposed Amendment to the Renewed Facility Operating Licenses Regarding Revisions to the Updated Final Safety Analysis Report Section Associated with Tile Standby Shutdown Facility License Amendment Request No. 2017-03
| author name = Ray T
| author affiliation = Duke Energy Carolinas, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000269, 05000270, 05000287
| license number = DPR-038, DPR-047, DPR-055
| contact person = 
| case reference number = ONS-2017-074
| package number = ML17299A114
| document type = Letter type:ONS, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Updated Final Safety Analysis Report (UFSAR)
| page count = 42
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:J...~DUKE          -
~   ENERGY                                                                                             Thomas D. Ray Vice President Oconee Nuclear Station Duke Energy ON01VP 17800 Rochester Hwy Seneca, SC 29672 o: 864.873.5016 ONS-2017-07 4                                                          10 CFR 50.90                f. 864.873.5791 Tom.Ray@duke-energy.com October 20, 2017 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, Maryland 20852

==Subject:==
Duke Energy Carolinas, LLC Oconee Nuclear Station (ONS), Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 Renewed Facility Operating License Nos. DPR-38, DPR-47; and DPR-55 Proposed Amendment to the Renewed Facility Operating Licenses Regarding Revisions to the Updated Final Safety Analysis Report Section Associated with tile Standby Shutdown Facility                                      '
License Amendment Request No. 2017-03 Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) proposes to amend Renewed Facility Operating Licenses for Oconee Nuclear Station (ONS) Units 1, 2, and 3. The License Amendment Request (LAR) proposes to revise the Updated Final Safety Analysis Report (UFSAR) to allow off-nominal success criteria for a Standby Shutdown Facility (SSF) mitigated Turbine Building (TB) flood event occurring when the Oconee Unit(s) are not at nominal full power conditions. In addition, the LAR requests approval to use the Main Steam (MS) Atmospheric Dump Valves (ADVs), when available, to enhance SSF mitigation capabilities.
The Enclosure provides background information related to NRC approval of the SSF design and Technical Specifications, a description of scheduled plant improvements, the proposed change to the UFSAR, and a technical evaluation that justifies the proposed change. A regulatory evaluation (including the No Significant Hazards Consideration) and environmental * .
considerations are provided in Sections 4 and 5 of the Enclosure. Attachment 1 contains the marked-up and retyped UFSAR pages. Attachment 2 describes the thermal-hydraulic codes and models used to perform analysis of SSF-mitigated TB flood scenarios in support of this LAR. No changes to Technical Specifications are proposed.                      -
The description of the thermal-hydraulic codes and models in Attachment 2 contains information that is proprietary to Duke Energy. Within Attachment 2, Duke Energy proprietary information is identified by brackets. In accordance with 10 CFR 2.390, Duke Energy requests that this information be withheld from public disclosure. Attachment 4 contains an Affidavit attesting to the proprietary nature of the information in Attachment 2. The proprietary information is owned by Duke Energy and has substantial commercial value that provides a competitive advantage.
Attachment 3 contains a non-proprietary [redacted] version of this content.
Attachment 2 to this letter contains proprietary information.
Withhold from Public Disclosure Under 10 CFR 2.390.
Upon removal of Attachment 2, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission October 20, 2017 Page 2 In accordance with Duke Energy administrative procedures that implement the Quality Assurance Program Topical Report, the proposed changes have been reviewed and approved by the Plant Operations Review Committee. A copy of this LAR is being sent to the State of South Carolina in accordance with 10 CFR 50.91 requirements.
Duke Energy requests approval of this amendment request by December 31, 2018. Once approved, the amendment will be implemented within 90 days. There are no new regulatory commitments being made as a result of this proposed change.
Inquiries on this proposed amendment request should be directed to Boyd Shingleton, ONS Regulatory Affairs Group, at (864) 873-4716.
I declare under penalty of perjury that the foregoing is true and correct. Executed on October 20, 2017.
Sincerely,
:!f:)>'.~
Vice President Oconee Nuclear Station

==Enclosure:==
Evaluation of Proposed Change Attachment   1     UFSAR Pages (1-A Marked-Up Pages, 1-B Retyped* Pages)
Attachment   2     Thermal-Hydraulic Models for SSF Transient Analysis [Proprietary]
Attachment   3 -   Thermal-Hydraulic Models for SSF Transient Analysis [Non-Proprietary]
Attachment   4     Affidavit Attachment 2 to this letter contains proprietary information.
Withhold from Public Disclosure Under 10 CFR 2.390.
Upon removal of Attachment 2, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission October 20, 2017 Page 3 cc w/enclosure and attachments:
Ms. Catherine Haney, Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Ms. Audrey Klett, Project Manager (by electronic mail only)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8G9A 11555 Rockville Pike Rockville, Maryland 20852 Mr. Eddy Crowe NRC Senior Resident Inspector Oconee Nuclear Station
* Ms. Susan E. Jenkins, Manager, (by electronic mail only: jenkinse@dhec.sc.gov)
Infectious and Radioactive Waste Management, Bureau of Land and Waste Management Department of Health & Environmental Control 2600 Bull Street Columbia, SC 29201 Attachment 2 to this letter contains proprietary information.
Withhold from Public Disclosure Under 10 CFR 2.390.
Upon removal of Attachment 2, this letter is uncontrolled.

ENCLOSURE EVALUATION OF PROPOSED CHANGE LICENSE AMENDMENT REQUEST 2017-03

==Subject:==
Proposed Amendment to the Renewed Facility Operating Licenses Regarding Revisions to the Updated Final Safety Analysis Report Section Associated with the Standby Shutdown Facility
: 1.     

==SUMMARY==
DESCRIPTION
: 2. DETAILED DESCRIPTION 2.1. SSF System Design and Operation 2.2. Current SSF Licensing Basis Requirements 2.3. Reason for the Proposed Change 2.4. Description of the Proposed Change 2.5. UFSAR Changes
: 3. TECHNICAL EVALUATION
: 4. REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria 4.2. No Significant Hazards Consideration Analysis 4.3. Conclusions
: 5. ENVIRONMENTAL CONSIDERATION
: 6. REFERENCES

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                            Page 1 -
1    Summary Description This License Amendment Request (LAR) proposes to revise the Updated Final Safety Analysis Report (UFSAR) to provide off-nominal success criteria for maintaining the reactor in a safe shutdown condition when using the Standby Shutdown Facility (SSF) to mitigate a Turbine Building (TB) flood occurring when the Oconee Unit(s) are not at nominal full power conditions.
In addition, the LAR requests approval to use the Main Steam (MS) Atmospheric Dump Valves (ADVs), when available, to enhance SSF mitigation capabilities.
This enclosure provides background information related to NRC approval of the SSF design and Technical Specifications, a description of the current issue, proposed change to the UFSAR, and a technical evaluation that justifies the proposed change. A regulatory evaluation (including the No Significant Hazards Consideration) and environmental considerations are provided in Sections 4 and 5 of this enclosure. Attachment 1 contains the marked up and retyped UFSAR pages. A description of the thermal-hydraulic codes and models used to perform analysis of SSF-mitigated TB flood scenarios in support of this LAR is provided in Attachments 2 and 3 (proprietary and non-proprietary versions, respectively). No changes to Technical Specifications are proposed.
Modifications to the plant are also being made to provide a larger capacity SSF reactor coolant letdown line and an improved pulsation dampener for the positive displacement SSF reactor coolant makeup pump that will allow sufficient reactor coolant system letdown and makeup capability over the full range of system pressure required for TB flood mitigation. These modifications are being performed under 10 CFR 50.59;. their approval is not a part of this LAR.
The combination of these modifications and the proposed change to the licensing basis will resolve the existing nonconforming conditions for each Oconee unit. The nonconforming conditions for the SSF are described in Section 2.3 below.
2   Detailed Description 2.1 SSF System Design and Operation 2.1.1 Original Design The three units at Oconee Nuclear Station (ONS) were designed in the late 1960's and the construction permits were issued prior to the development of many of the presently existing regulations and requirements. Oconee Unit 1 received its initial Operating License (OL) in February 1973, Unit 2 in October 1973 and Unit 3 in July 1974. Each unit features a Nuclear Steam Supply System (NSSS) designed and supplied by Babcock and Wilcox (B&W). The Standby Shutdown Facility (SSF), which was not a part of the original plant design, was installed in the early 1980s to address NRC concerns related to plant security, fire protection and Turbine Building flooding.

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                             Page 2 2.1.2   SSF System Description The Standby Shutdown Facility (SSF) is designed as a standby system for use under certain emergency conditions. The system is not credited for the mitigation of design basis events, but provides additional "defense in-depth" protection for the health and safety of the public by
* serving as a backup to existing safety systems. The SSF is provided as an alternate means to achieve and maintain the unit in MODE 3 with average Reactor Coolant System (RCS) temperature ~ 525&deg;F (unless the initiating event causes the unit to be driven to a lower temperature) following a fire, station blackout (SBO), or TB flood event. The limiting turbine building internal flooding event occurs as the result of failure of a Condenser Circulating Water (CCW) piping expansion joint. As the flood height in the turbine building increases, the flooding results in a reactor/turbine trip and a loss of both main and emergency feedwater systems. In addition, the SSF may be activated as necessary in response to events associated with plant security. Because the SSF is a backup to existing safety systems and is not credited in the accident analyses for design basis events, single failure criterion is not required to be met.
Failures in the SSF systems will not cause failures or inadvertent operations in other plant systems. The SSF requires manual activation.
The SSF is.designed to maintain the reactor in a safe shutdown condition for a period of 72 hours following a fire or turbine building flood, and for a period of 4 hours following an SBO. The capability of the SSF to maintain the reactor in a safe shutdown condition is also credited for certain security-related events.
2.1.2.1 SSF ASW System The SSF ASW System is a high head, high volume system designed to provide sufficient steam generator (SG) inventory for adequate decay heat removal for three units during a loss of normal AC power in conjunction with the loss of the normal and emergency feedwater systems.
One motor driven SSF ASW pump, located in the SSF building, serves all three units. The SSF ASW pump, two HVAC service water pumps, and the Diesel Service Water (DSW) pump share a common suction supply of lake water from the embedded Unit 2 Condenser Circulating Water (CCW) supply piping. CCW is the ultimate heat sink. Main Steam pressure is controlled automatically by the main steam relief valves.
2.1.2.2* SSF Reactor Coolant (RC) Makeup System The SSF RC Makeup System is designed to supply makeup to the RCS in the event that normal makeup systems are unavailable. An SSF RC Makeup Pump located in the Reactor Building of each unit supplies makeup to the RCS should the normal makeup system flow and RCP seal cooling become unavailable. The system is designed to ensure that sufficient borated water is provided from the spenffuel pools to allow the SSF to maintain all three units in MODE 3 with average RCS temperature~ 525&deg;F (unless the initiating event causes the unit to be driven to a lower temperature) for approximately 72 hours. An SSF RC Makeup Pump is capable of delivering borated water from the Spent Fuel Pool to the RCP seal injection lines. A portion of this seal injection flow is used to makeup for RCP seal leakage while the remainder flows into the RCS to makeup for other RCS leakage (non-LOCA) and to support RCS cooldown to nominal post-trip conditions.

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                                  Page 3 The SSF RC Makeup Pump is a positive displacement pump. To protect components located downstream of the SSF RC Makeup pump from being damaged, a pulsation dampener is provided to dampen pressure surges. The currently installed dampener is a bladder with nitrogen overpressure; a design which is effective at nominal RCS operating pressures. This dampener is being replaced with a passive design which is effective over the full range of RCS pressures. This change is being implemented under 10 CFR 50.59.
Each Unit's SSF RC letdown line is independent of the normal RC letdown line. Currently, the SSF RC letdown line originates off one of the RCS cold legs and discharges to a fuel transfer tube via an orifice and isolation valve arrangement; flow through the line cannot be throttled. At nominal RCS operating pressures, the SSF RC letdown flow is limited to approximately 41 gpm and is reduced at lower RCS pressures. The SSF RC letdown line is being replaced with a new line that originates off one of the RCS hot legs and discharges to a fuel transfer tube via an isolation and throttle valve arrangement. The new SSF RC letdown line will be capable of providing approximately 300 gpm at nominal RCS operating pressures and is effective at low RCS pressures. This change is being implemented under 10 CFR 50.59.
2.2 Current SSF Licensing Basis Requirements In the late-1970's and early 1980's, Duke Energy developed the conceptual and final design of the SSF to augment existing plant capabilities relative to mitigating postulated occurrences such as fires, turbine building flooding and security-related incidents. The ONS SSF design and associated criteria were approved in an NRC Safety Evaluation (SE) dated April 28, 1983 *
(Reference 1). The NRC SE was based on submittals made by Duke Energy in References 2, 3, 4, 5, 6, and 7. Within these communications, the SSF is described as a "bunkered" facility which houses the systems and components necessary to provide an alternate and independent means to achieve and maintain a hot shutdown condition for one or more of the three Oconee units. At that time hot shutdown was defined in the ONS Technical Specifications (TSs) as the reactor subcritical by 1% b. k/k and Tavg ~ 525&deg;F.
Duke Energy performed analyses to support the SSF design during this time frame that evaluated and confirmed the effectiveness of the SSF in controlling and mitigating the response of the RCS based on initial conditions of full power (Reference 2). The two concerns of importance, the possibility of return to criticality and the ability to maintain natural circulation, were addressed by the analyses. During the licensing of the SSF design, Duke Energy and the NRC acknowledged the SSF would have the capability of maintaining hot shutdown conditions in all three units for approximately three days following a loss of normal AC power            *
(Reference 1).
The April 28, 1983, SE (Reference 1) states that the SSF design meets appropriate requirements and acknowledges that the SSF was designed to provide an alternate and independent means to achieve hot shutdown conditions. In the SE, the NRC requested that Duke Energy continue work on the TSs needed to ensure the operability of SSF components agrees with assumptions used in the design. The SE acknowledges that the SSF RC Makeup System was designed with a capacity to account for normal primary system leakage and shrinkage which results from transitioning from a hot power operating condition to hot shutdown.

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                                 Page 4 The SE also states that the SSF Systems required for safe shutdown are designed with adequate capacity to ensure safe hot shutdown conditions for all three Oconee units.
*.Duke Energy submitted a License Amendment Request (LAR) on July 26, 1985 (Reference 8),
proposing TSs requiring the SSF to be operable at any time an Oconee unit is in the hot shutdown condition, hot standby or power operation. At the time, hot shutdown was defined as when the reactor is subcritical by at least 1% b.k/k and RCS Tavg is ;::: 525&deg;F. By letter dated January 23, 1987 (Reference 9), NRC stated the proposed TSs should be revised to incorporate a Limiting Condition for Operation (LCO) Applicability comparable to that in Standard Technical Specifications (STS) for the Emergency Feedwater (EFW) system and other safety related systems since the SSF ASW System was being credited as a backup to the EFW System.
Duke Energy supplemented the July 26, 1985, LAR on August 14, 1987 (Reference 10), to address these concerns and extend the SSF operability requirements down to Tavg ;::: 250&deg;F; further expanding the applicability of the SSF TS rather than just expanding the applicability of TS requirements for the SSF ASW System. The TS definition of OPERABILITY is that systems, structures, and components (SSCs) can perform their specified safety function. The safety function of the SSF is to achieve and maintain the unit(s) in MODE 3 with average RCS temperature 2 525&deg;F for up to 72 hours. The analysis performed to support the SSF design assumed an initial condition of 100% nominal full power conditions.                         \
After the SSF design was approved by the NRC in 1983 (Reference 1) and prior to approval of the initial SSF TS in 1992 (Reference 11 ), Duke Energy requested and received approval to credit the SSF to mitigate Station Blackout (SBO) (References 12 and 13), and to address EFW system equipment vulnerabilities associated with single failures, tornado missiles, and seismic design (Reference 14).
Per UFSAR Section 9.6, for fire and TB flood events, the SSF is designed to:
* maintain a minimum water level above the reactor core, with an intact RCS, and maintain RCP seal cooling;
* assure natural circulation and core cooling by maintaining the primary coolant system filled to a sufficient level in the pressurizer while maintaining sufficient secondary side cooling;
* transfer decay heat from the fuel to an ultimate heat sink; and
* maintafn the reactor 1% shutdown with the most reactive rod fully withdrawn.
Following the transition to NFPA 805, the requirement to assume the most reactive rod remains fully withdrawn was eliminated for fire events and the initial conditions assumed for evaluating fire events mitigated by the SSF was clarified (References 16 and 17). As a result, the turbine building flood is the only remaining SSF-mitigated event associated with TS operability where ONS current licensing basis does not specifically limit initial conditions for the evaluation to nominal full power conditions.
Oconee thermal-hydraulic analyses demonstrate that the SSF is capable of meeting these success criteria for events initiated from nominal full power conditions.

License Amendment Request No. 2017 Enclosure October 20, 2017
* Page 5 2.3 Reason for the Proposed Change 2.3.1 SSF Comprehensive Review During an SSF comprehensive review performed by Duke Energy in 2012, the review team identified a concern that thermal-hydraulic (T-H) analyses for some SSF-mitigated events (Appendix R fire and TB flood) did not consider all initial operating conditions, especially lower operating modes and lower decay heat (Reference 15). Rather, thermal-hydraulic analyses had only been performed for nominal full power conditions. Technical Specification (TS) 3.10.1 requires the SSF to be operable in Modes 1, 2 and 3. The primary concern was whether the current success criteria (assure natural circulation and core cooling by maintaining the primary coolant system filled to a sufficient level in the pressurizer while maintaining sufficient secondary side cooling) for certain events initiated with low decay heat or from lower temperatures could be met for the 72-hour mission time. These events are described in more detail below.
2.3.2 Initial Condition - Low Decay Heat Early in the event, low decay heat may result in challenges to the pressurizer level (low) success criterion. In such conditions, heat removal by normal Main Steam (MS) system loads will likely exceed available decay heat; resulting in cooling and shrinkage of the RCS. This decrease in RCS volume will be seen in pressurizer level which may go off-scale low. The potential for pressurizer level to go off-scale low due to low decay heat exists during the operating cycle until there is an adequate core power history. This plant response is not impacted by the modifications to the SSF RC Makeup System described in Section 2.1.2.2.
Late in the event, rather than heat rejection to the steam generators, ambient heat losses to containment and heat removal via SSF reactor coolant makeup and letdown become the dominant methods of decay heat removal. This would become apparent as SSF ASW flow to steam generators is continually reduced and ultimately stopped in an attempt to maintain stable RCS temperatures. However, RCS temperatures continue to slowly decrease following cessation of SSF ASW flow. Although the core remains covered and cooled, natural circulation flow may periodically stagnate. This method of decay heat removal is not currently described in the UFSAR. The potential for natural circulation flow to stagnate due to low decay heat exists during the operating cycle until there is an adequate core power history. This plant response is not impacted by the modifications* to the SSF RC Makeup System described in Section 2:1.2.2.
The current configuration of the SSF reactor coolant letdown line controls flow by means of an orifice; the line cannot be throttled. In order to properly balance SSF reactor coolant makeup and letdown, operators maintain RCS pressure in a band of approximately 1950 to 2250 psig.
With the current plant configuration, water-solid conditions (without water relief) may develop in the pressurizer later in the event as insurges of relatively cold water overcome the ability of the SSF controlled pressurizer heaters to maintain RCS pressure in the required band. As the RCS hot legs cool due to insufficient decay heat, pressurizer level must be periodically increased to compress the pressurizer steam bubble to restore RCS pressure to the required band. As pressurizer level is increased, the cooler water from the RCS hot leg subcools the pressurizer fluid. Eventually, these cold insurges overcome SSF controlled pressurizer heater capacity and the RCS evolves into a water-solid condition. The potential for a water-solid RCS due to low decay heat exists during the*operating cycle until there is an adequate core power history. The

License Amendment Request No. 2017-03:.. Enclosure October 20, 2017                                                                             Page 6 modifications to the SSF RC Makeup System described in Section 2.1.2.2 will eliminate this condition by providing the ability to balance SSF reactor coolant makeup and letdown by throttling letdown flow (regardless of RCS pressure).
2.3.3 Initial Condition - Low Initial Temperature with High Decay Heat During postulated SSF events, Main Steam system pressure is controlled by the Main Steam Relief Valves (MS RVs). The lowest MSRV setpoint pressure is -1050 psig which corresponds to a saturation temperature of -552&deg;F. SSF events that initiate at lower temperatures with high decay heat will result in RCS heatup (and expansion) to -552&deg;F until a secondary side steam flowpath is established through the MSRV's. This increase in RCS volume will be seen in pressurizer level which may go off-scale high and in some cases may result in water relief from the pressurizer safety valves for the current plant configuration. Although the RCS may become water-solid, the modifications to the SSF RC Makeup System described in Section 2.1.2.2 will eliminate the potential for water relief through the pressurizer safety valves by providing the ability to significantly increase SSF reactor coolant letdown flow.
2.3.4 Current SSF Operability Requirements Duke Energy entered this concern into the Corrective Action Program on March 29, 2012, and subsequently determined that the SSF could not meet current success criteria for an Appendix R fire event and TB flood event during certain periods within plant startup and shutdown. Since then, ONS has transitioned to NFPA 805 and can meet the current success criteria for the NFPA 805 fire. Duke Energy determined that current success criteria for the TB flood event may not be met for two cases:
: 1) an initial condition with low decay heat; and
: 2) an initial condition with a low initial RCS temperature and high decay heat.
This condition was reported as an unanalyzed condition that significantly degraded plant safety and as a condition prohibited by Technical Specifications (TSs) in Licensee Event Report (LER) 2012-001-00 dated June 4, 2012 (Reference 15). At the time of LER submittal, the planned action to resolve the unanalyzed condition was to provide appropriate analysis or licensing*
changes for SSF operability during off-nominal conditions. Currently due to the unanalyzed condition, during periods of startup and shutdown when current success criteria cannot be met, the SSF is declared inoperable. SSF Operability during periods other than nominal full power conditions is dependent upon the operating history of the reactor core and its resultant decay heat profile following event initiation. Sufficient decay heat is required to meet the current success criteria associated with SSF events, with some of these criteria being challenged early in the event (pressurizer level) while others are challenged at the end of the event (maintaining natural circulation for the 72 hour mission time).
* During the low decay heat conditions of startup after a refueling outage, the SSF is declared inoperable (upon entry into the mode of applicability) and the applicable TS 3.1 O Conditions are entered until the required decay heat level is achieved to allow current success criteria to be met. The startup is permitted to continue as the SSF Technical Specifications (TS 3.10.1) include an LCO 3.0.4 exception that allows entry into a MODE or other specified condition in the

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                              Page 7 Applicability when an LCO is not met. Duke Energy analyses demonstrate that following approximately four days of operation at 100% power there is adequate decay heat to support SSF operability through the entire 72 hour mission time. This does not delay startup provided the required decay heat level is achieved prior to the end of the 7-day Completion Time of the applicable TS Required Actions. The approximate 4-day duration in these low decay heat conditions does not result in an appreciable contribution to overall plant risk. Likewise, for a short period of time during shutdown, when reactor coolant temperature is low (during cooldown to< 250&deg;F) and decay heat is high, the SSF is declared inoperable. The approximate 10-hour duration of operation in these conditions does not result in an appreciable contribution to overall plant risk. In addition, infrequent mid-cycle power reductions below approximately 85% full power may require the SSF to be declared inoperable due to inadequate decay heat.
Unless it is taken out of service for other reasons, such as scheduled maintenance, the SSF equipment remains available during those periods where it is declared inoperable and would be used by station operators to facilitate the mitigation of SSF events.
2.4 Description of the Proposed Change Duke Energy proposes to revise the UFSAR, as shown in Attachment 1, to state that current success criteria applies to a TB flood event occurring at nominal full power conditions and that off-nominal success criteria applies to a TB flood event occurring at less than nominal full power conditions. A nominal full power condition is defined as a unit at 100% power operation for a minimum of approximately 4 days, which provides the decay heat required to meet the current nominal SSF success criteria. The current success criteria for events occurring during nominal full power conditions are to assure natural circulation and core cooling by maintaining the primary coolant system filled to a sufficient level in the pressurizer while maintaining sufficient secondary side cooling with the reactor maintained at least 1% .6k/k shutdown with the most reactive rod fully withdrawn. For off-nominal operating conditions, the proposed off-nominal success criteria would: 1) allow the pressurizer to go water-solid provided that there is no liquid relief through the pressurizer safety valves; and, 2) would consider ambient losses and letdown and makeup of reactor coolant acceptable methods of decay heat removal during periods of low decay heat in lieu of sustained natural circulation with SG heat removal. SSF equipment needed to support SG heat removal will remain available.
T-H analyses were performed for a Turbine Building flood event using Duke Energy's RETRAN-3D ONS thermal-hydraulic model. The RETRAN-3D model is described in Duke Energy's NRC-approved methodology report DPC-NE-3000-PA (Reference 18), and has been modified, as described in Attachment 2, to capture important phenomena in the RCS and pressurizer for longer-duration SSF events, which may transition to a water:-solid condition. The results from these analyses demonstrate the SSF systems and operator guidance can be used to successfully mitigate a Turbine Building flood event, meeting the current success criteria for nominal full power conditions and the proposed off-nominal success criteria for off-nominal conditions.
Duke Energy also requests NRC approval to use the Main Steam (MS) Atmospheric Dump Valves (ADVs), when available, to enhance the mitigation of SSF events. NRC approval is

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                                      Page 8 needed since the ADVs will be used to manually control MS pressure below the main steam relief valves (MSRVs) lowest lift setpoint. Use of the ADVs, when available, will also limit cycling of the MSRVs. During licensing of the SSF design, Duke Energy only credited the automatic opening of the MSRVs to release steam for decay heat removal. Use of the ADVs is not credited in the thermal~hydraulic analyses.
* 2.5 UFSAR Changes Duke Energy proposes to modify the UFSAR, as follows, to describe the T-H analyses performed to demonstrate the SSF can achieve and maintain safe shutdown following postulated turbine building floods and to provide off-nominal success criteria for a TB flood event occurring during off-nominal full power conditions (UFSAR marked-up and retyped pages provided in Attachment 1A and 1B, respectively):
Current UFSAR Section 9.6.2 for Turbine Building Flood Event (Page 9.6-3)
     "TURBINE BUILDING FLOOD EVENT The Turbine Building Flood was one of the events that was identified in the original SSF licensing requirements. The SSF is designed to maintain the reactor in a safe shutdown condition for a period of 72 hours following a TB Flood. No other concurrent event is assumed to occur. The success criteria for this event is to assure natural circulation and core cooling by maintaining the primary coolant system filled to a sufficient level in the pressurizer while maintaining sufficient secondary side cooling. The reactor shall be maintained at least 1% Af</k with the most reactive rod fully withdrawn.
(Reference 1, 10)"
Proposed Changes to UFSAR Section 9.6.2 for Turbine Building Flood Event Replace the current text from the beginning of the fourth sentence ... "The success criteria for this event is" with the following:
    "To verify SSF performance criteria, thermal-hydraulic(T-H) analyses were performed to demonstrate that the SSF can achieve and maintain safe shutdown following postulated turbine building floods.
The analysis evaluates RCS subcooling margin using inputs that are representative of nominal full power end of cycle plant conditions. The analysis uses an initial core thermal power of 2619 MWth (102% of 2568 MWth) and accounts for 24 month fuel cycles. The consequences of the postulated loss of main and emergency feedwater were analyzed as a RCS overheating scenario. For the examined overheating scenario, an important core input is decay heat. High decay heat conditions were modeled that were reflective of maximum, end of cycle conditions. The high decay heat assumption was confirmed to be bounding with respect to the RCS subcooling response. The results of the analysis demonstrate that the SSF is capable of meeting the success criteria for this event"

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                                       Page 9 and then add the following new paragraphs to provide success criteria for the SSF for events occurring during off-nominal conditions for Oconee Unit(s) with the SSF letdown line and SSF pulsation dampener modifications complete.
   "During periods of very low decay heat the SSF will be used to establish conditions that support the formation of subcooled natural circulation between the core and the SGs; however, natural circulation involving the SGs may not occur if the amount of decay heat available is less than or equal to the amount of heat removed by ambient losses to containment and/or by other means, e.g., letdown of SSF reactor coolant makeup. When these heat removal mechanisms are sufficient to remove core decay heat, they are considered adequate to meet the core cooling function and systems supporting SG decay heat removal, although available, are not necessary for core cooling.
Regarding operation in MODES 1, 2, and 3 at other than nominal full power, T-H analyses were also performed to demonstrate that the SSF could achieve and maintain safe shutdown following postulated turbine building floods. A nominal full power condition is defined as a unit at 100% power for approximately 4 days of operation which provides the decay heat required to meet the nominal SSF success criteria. The results of the analyses demonstrate that for this range of initial conditions the SSF is capable of meeting the specified success criteria: 1) to maintain the reactor at least 1% ~k/k shutdown with the most reactive rod fully withdrawn, and 2) maintain a minimum water level above the reactor core and conditions that support the formation of subcooled natural circulation between the core and SGs without water relief through the pressurizer safety valves.
In some cases that initiated with low decay heat, pressurizer level was not maintained on scale low; however, a minimum water level above the reactor core was maintained and conditions that support the formation of subcooled natural circulation between the core and the SGs were maintained. In some cases where the pressurizer did become water-solid, there was no liquid relief through the pressurizer safety valves."
Current UFSAR Section 9.6.3.3 Auxiliary Service Water (ASW) System (5th paragraph)
   "Auxiliary service water enters the steam generators via the normal emergency feedwater ring headers."
Proposed Changes to UFSAR Section 9.6.3.3 Insert the following after the first sentence of the 5th paragraph:
   "Main Steam pressure is controlled automatically by the main steam relief valves or manually by the atmospheric dump valves (ADVs). When the ADVs are operated in this manner, communication with the SSF Control Room is in place to coordinate main steam pressure control with RCS pressure and temperature parameters. Local main steam pressure indication is also available at the ADV s."

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                              Page 10 3 TECHNICAL EVALUATION 3.1     Justification for the Proposed Change The SSF is provided as an alternate means to achieve and maintain the unit in MODE 3 with average RCS temperature<:: 525&deg;F (unless the initiating event causes the unit to be driven to a lower temperature) following a fire, turbine building flood, and station blackout (SBO) events.
Duke Energy recently concluded, as described in Section 2.3.1 above, that SSF events in which the initial reactor power assumptions are not defined should.be considered to initiate from any operating condition during Modes 1 through 3, consistent with the modes of applicability for the SSF Technical Specifications.
For SSF-mitigated fire events, initial reactor power assumptions are defined as 100% power with sufficient decay heat such that natural circulation can be achieved (Oconee NFPA 805 Submittal dated April 14, 2010 (Reference 16), NRG Safety Evaluation for ONS NFPA 805 dated December 29, 201 O (Reference 17) and UFSAR Section 9.6.2, Design Bases - Fire Event). For SSF-mitigated SBO events, initial reactor power assumptions are defined as 100%
power and at least 100 days of operation at this power level (Regulatory Guide 1.155, Station Blackout, Section 3.2.1 ). For SSF-mitigated turbine building flood events, ,initial reactor power assumptions are not defined by docketed correspondence or NRG rules. Therefore, Duke Energy has assumed that the SSF-mitigated turbine building flood event can occur from any operating condition during Modes 1 through 3.
As described in UFSAR Section 9.6.2, the success criteria for the turbine building flood event is to assure natural circulation and core cooling by maintaining the primary coolant system filled to a sufficient level in the pressurizer while maintaining sufficient secondary side cooling. The reactor shall be maintained at least 1% ~k/k shutdown with the most reactive rod fully withdrawn. Using the T-H models described in Attachment 2, analyses confirm that these current success criteria are met for an SSF event occurring from a nominal full power initial condition. However, additional analyses have concluded that some of the current success criteria may not be met for some turbine building flood events occurring with an off-nominal initial condition. For off-nominal initial conditions, the proposed new success criteria would allow the pressurizer to go water-solid provided that there is no liquid relief through the pressurizer safety valves, and would consider ambient heat losses and letdown of reactor coolant an acceptable method of decay heat removal during periods of low decay heat. Additional T-H analyses have been performed using the T-H models described in Attachment 2 that confirm these new success criteria are met for an SSF TB flood event occurring from off-nominal initial conditions.
The following three scenarios envelope the initial conditions for which the SSF is required to be operable for the TB Flood event.
* Nominal full power - Currently the SSF is operable with the existing plant configuration. A nominal full power condition is defined as a unit at 100% power operation for a minimum of
     -4 days, which provides the decay heat required to meet the current nominal SSF success criteria. Thermal-hydraulic analyses *have been performed using Duke Energy's RETRAN-3D ONS model (Reference 18), modified as described in Attachment 2. Boundary conditions, inputs, and assumptions in the analyses are selected to examine these cases as

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                             Page 11 either an overheating event or an overcooling event to verify the predicted plant response is conservatively bounding. Parameters that may vary between the overheating and overcooling cases include decay heat response, timing of reactor trip or trip function, EFW flow rates while pumps are available, RCS ambient heat losses, and modeling of secondary system steam loads. For example, an overheating case may assume maximum decay heat, a delayed reactor trip from the Reactor Protection System, minimum EFW flow rates, and neglect modeling secondary system steam loads and RCS ambient heat losses to minimize primary-to-secondary heat transfer and maximize the post-trip RCS overheating response.
An overcooling case may reduce the decay heat response, assume an early or immediate reactor trip, maximize EFW flow rates and model RCS ambient heat losses and secondary system steam loads to maximize the RCS overcooling response. For both types of cases, these analyses show subcooled single-phase natural circulation flow is maintained in the RCS for the duration of the event, thus ensuring core cooling and decay heat removal through the SGs. After losing MFW and EFW due to the TB flood, operators use SSF ASW and steam relief through the MSRVs for SG heat removal, establishing SG levels that support natural circulation in the RCS. Borated RCS makeup and RCP seal injection flow are provided from the SSF RC makeup pump. The reactor remains at least 1% ~k/k shutdown with the most reactive rod fully withdrawn. Pressurizer level remains on-scale, with operators eventually controlling pressurizer level by balancing makeup and letdown flows using the modified SSF letdown line. Later in the event, operators use SSF-controlled pressurizer heaters to re-saturate pressurizer fluid and control RCS pressure. No UFSAR changes are required to support the nominal full power case.
MODE 1, lower initial power or burnup, MODE 2, or MODE 3 - Current analyses support SSF operability for a small portion of this region with the existing plant configuration. For example, the SSF has been shown to be operable following a unit startup and achieving approximately four days of 100% power operation. Additionally, analyses have been performed to show that the SSF remains operable following power reductions to 85% power after four days of operation at 100% power and power reductions to 70% power after 600 days of operation at 100% power. Outside of these specific off-nominal initial operating conditions, the SSF is currently inoperable in this region.
T-H analyses have been performed using Duke Energy's RETRAN-30 ONS thermal-hydraulic model (Reference 18), modified as described in Attachment 2. The analyses credit the increased capacity and throttle capability of the new SSF letdown line modification. For a scenario where the initial power level or burnup is low, the success criteria for maintaining shutdown margin and sufficient pressurizer level are challenged by evaluating these conditions as overcooling events. Initial power level or burnup conditions greater than those considered in this scenario and less than those considered for nominal full power conditions are bounded by the nominal full power scenario for overheating and the low initial power level or burnup scenario for overcooling. Scenarios with low initial power or burnup conditions may credit the proposed off-nominal success criteria that considers ambient losses and letdown of reactor coolant an acceptable method of decay heat removal during periods of low decay heat in lieu of sustained natural circulation with SG heat removal.
For the range of initial conditions examined through analysis within the low initial power or burnup envelope, the results show subcooled single-phase natural circulation flow is maintained in the RCS for the duration of the event, thus ensuring core cooling and decay

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                              Page 12 heat removal through the SGs and RCS ambient heat losses. The combination of RCS ambient heat losses and secondary system steam loads result in a slow and near-continuous reduction in RCS temperature shortly after TB flood initiation. If decay heat is sufficiently low, RCS pressure may experience a slow and continual decrease throughout the event until SSF-controlled pressurizer heaters re-saturate the pressurizer fluid much later in the transient. Even so, significant subcooling margin and shutdown margin are maintained throughout the transient with the most reactive rod fully withdrawn. Borated RCS makeup and RCP seal injection flow are provided from the SSF RC makeup pump, increasing RCS boron concentration and providing shutdown margin control during the RCS cooldown.
Due to the assumption that some secondary system steam loads remain active for a limited time period, SG pressures decrease shortly after TB flood initiation. Once the majority of steam loads are isolated early in the event, SG pressures begin to increase again.
Eventually, smaller residual steam loads, ambient heat losses, and RCS makeup and letdown flow exceed decay heat, and SG pressures decrease for the remainder of the event.
After losing MFW and EFW due to the TB flood, operators use SSF ASW to establish SG levels that support natural circulation in the RCS. As RCS fluid temperatures decrease, the fluid density increases causing a:pressurizer level reduction as the RCS inventory shrinks. In some cases, although a minimum water level above the reactor core and conditions that support the formation of subcooled natural circulation between the core and the SGs were maintained, pressurizer level may go off-scale low. Eventually, the RCS temperature reduction and resulting RCS inventory shrinkage slows and makeup flow provided from the SSF RC makeup pump is able to recover pressurizer level. Once pressurizer level has increased to the minimum required level, operators use SSF-controlled pressurizer heaters to re-saturate pressurizer fluid and control RCS pressure to maintain RCS subcooling margin.
In addition, during periods of very low decay heat, beyond the range of conditions examined in the T-H analyses, natural circulation. may not occur if the amount of decay heat is less than or equal to the amount of heat removed by ambient losses to containment and/or by other means, e.g., letdown of SSF reactor coolant makeup. When these heat removal mechanisms are sufficient to remove core decay heat, they are considered adequate to meet the core cooling function and systems supporting SG decay heat removal, although available, are not necessaryfor core cooling.
The results of lower power/low burnup T-H cases support a revision to UFSAR Section 9.6.2 to reflect that pressurizer level may not be maintained on scale and natural circulation may not occur as a result of low decay heat and ambient losses. The proposed revision is described in Section 2.5, UFSAR Changes, of this Enclosure.
MODE 3, low initial temperature, high decay heat - Currently the SSF is declared inoperable in this region with the existing plant configuration. The low RCS temperature conditions are achieved through a controlled cooldown of the unit following reactor shutdown. The current SSF letdown line capacity was designed to account for net RCS makeup from the SSF RC makeup pump, with little additional capacity. With the existing SSF letdown line, T-H analysis determined that the pressurizer may reach a RCS water-solid condition with the* potential for water relief through the pressurizer safety valves if the

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                            Page 13 event initiates at a low enough RCS temperature in MODE 3. The modified letdown line with greater capacity and throttling capability will preclude water relief through the pressurizer safety valves. Currently, SSF reactor coolant letdown does not have throttling capability due to the orifice/motor operated valve (MOV) configuration. The modified line will provide throhle valves. Thus, makeup and letdown can be balanced over the full range of RCS pressures and temperatures in MOD Es 1, 2, and 3. Implementation of these modifications is scheduled for the fall of 2018 for Oconee Unit 1, fall of 2019 for Oconee Unit 2, and the spring of 2018 for Oconee Unit 3.
T-H analyses have been performed using Duke Energy's RETRAN-3D ONS thermal-hydraulic model (Reference 18), modified as described in Attachment 2. The analyses credit the increased capacity and throttle capability of the new SSF letdown line modification. For conditions in MODE 3 with low initial RCS temperature and high decay heat, the RCS heats up and thermally expands following TB flood initiation until primary system temperatures approach the saturation temperature of the lowest lifting MSRVs (-550&deg;F). As the RCS heats up and pressurizes, fluid surges into the pressurizer, increasing pressurizer level and, potentially, challenging liquid relief through the Pressurizer Safety Valves (PSVs). As such, these scenarios are analyzed as overheating events to exacerbate the RCS transition to a water-solid condition and challenge the pr:oposed off-nominal success criteria for allowing water-solid conditions with no liquid relief through the PSVs.
For the range of initial conditions examined through analysis within the MODE 3 low initial temperature and high decay heat envelope, the results show subcooled single-phase natural circulation flow is maintained in the RCS for the duration of the event, thus ensuring core cooling and decay heat removal through the SGs. Following the loss of MFW with the TB flood, RCS temperatures begin to increase rapidly. Temporary use of EFW and, eventually, SSF ASW to fill SGs to natural circulation levels slows the rate of temperature increase. Once SG pressures reach the setpoint of the lowest lifting MSRVs, SG pressures and RCS temperatures stabilize. Adequate RCS subcooling margin is maintained throughout the event. Prior to the TB flood, operators maintain shutdown margin requirements during the preceding RCS cooldown. As a result, reactor shutdown margin is not challenged during the event since increasing moderator and fuel temperatures along with borated makeup flow from the SSF RC makeup pump increase shutdown margin throughout the event.
Depending on when the TB flood begins during the RCS cooldown, RCS temperatures and pressures may be low enough for Low Temperature Over Pressure (LTOP) protection to be active. For such cases, as RCS pressure increases early in the event, the pressurizer Power Operated Relief Valve (PORV) may cycle at the LTOP LOW setpoint of 530 psig until RCS conditions permit operators to set the PORV to the HIGH setpoint of 2450 psig. During this
* time period, the new modified SSF letdown line is available to slow the rate of RCS pressurization, reducing the load on the pressurizer PORV. Once the pressurizer PORV is set to the HIGH setpoint, RCS pressure will continue to increase until operators increase letdown flow through the SSF letdown line to balance letdown against makeup flow and the thermally expanding RCS inventory, where RCS pressure will remain stable for the remainder of the event. For TB flood events beginning with RCS conditions above LTOP protection, operators will use the SSF letdown line to stabilize RCS pressure while the systems heats up and expands. For some higher initial temperature conditions, the

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                               Page 14 pressurizer PORV may cycle under steam relief at the HIGH setpoint of 2450 psig until operators open the SSF letdown to reduce and stabilize RCS pressure.
Following event initiation, pressurizer level increases quickly with increasing RCS temperature and pressure. Once operators begin controlling RCS pressure with the SSF letdown line, the rate of pressurizer level increase slows down. Even after RCS pressure has stabilized, pressurizer level continues to increase due to thermal expansion of RCS inventory, the large insurge of subcooled liquid to the pressurizer overcoming SSF-controlled pressurizer heaters, and the collapse of the pressurizer steam bubble due to ambient heat losses and increased interfacial heat transfer in the presence of subcooled liquid. The transition to a water-solid pressurizer condition occurs slowly over several hours after pressurizer level goes off-scale high. It is within operator's capability to control subsequent increases in RCS pressure following the loss of the pressurizer steam bubble given the letdown capacity of the new modified SSF letdown line and operational margin between the PSV or PORV lift setpoints and operational guidance that will be developed for RCS pressure control. Once RCS temperatures stabilize, further changes in RCS pressure are the result of a mismatch between letdown flow and the low capacity SSF RC makeup pump. As a result, RCS pressure changes under water-solid conditions with the new SSF letdown line progress slowly. Results from the T-H,analyses show RCS pressure remains more than 700 psi below the PSV lift setting with a water-solid pressurizer condition for the duration of the event. Changes to the SSF letdown line control valve position is a manual action from the SSF, and the operator is required to maintain a very high awareness of the plant status for RCS pressure. Therefore, Duke Energy considers water-solid operation an acceptable method of maintaining safe shutdown during an SSF TB flood event, with adequate margin to prevent liquid relief through the PSVs.
Depending on the initial RCS temperature condition, SSF-controlled pressurizer heaters may re-saturate the pressurizer fluid and draw a steam bubble during the 72-hour mission time of the SSF. For such cases, as the steam mass in the pressurizer increases, so does RCS pressure. Operators, instructed to control RCS pressure with the SSF letdown line, compensate by throttling open the letdown line control valve(s) to increase letdown flow.
The combination of vaporizing pressurizer fluid and increasing letdown flow causes pressurizer level to slowly decrease. Once the desired pressurizer level has been reached, operators use SSF-controlled pressurizer heaters to control RCS pressure, and maintain a constant pressurizer level using the SSF letdown line. For the lowest initial temperature conditions within MODE 3, the pressurizer fluid may not re-saturate during the 72-hour SSF mission time due to the initial insurge of significantly subcooled RCS fluid overcoming the SSF-controlled pressurizer heaters in combination with ambient heat losses from the pressurizer. For such conditions, the pressurizer may remain water-solid for the duration of the event, with operators using the SSF letdown line to control RCS pressure.
The results of low initial temperature/high decay heat T-H cases described above support the revision to UFSAR section 9.6.2, described in Section 2.5 (UFSAR Changes) of this Enclosure, to reflect that pressurizer level may not be maintained on scale high following an event which initiates from a low temperature in MODE 3 with high decay heat.

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                         Page 15 In addition, although the Main Steam (MS) Atmospheric Dump Valves (ADVs) are not credited in the analyses for SSF events, operation of the ADVs may allow the event to be terminated earlier by manually controlling main steam pressure to a pressure below the MSRV setpoint.
Controlling at a lower MS pressure reduces Tsat for the steam generators and reduces heatup and swell of the RCS to< 550&deg;F. *The ADVs are periodically tested and have been used' successfully for a plant cooldown. The manually operated ADVs are located on the fifth floor of the turbine building near the main control room entrances. Due to their location and dependence on station staffing, they may not be available for all SSF events. ADV use will also reduce the cycling load on the MSRVs over the course of an event. Duke Energy requests NRC approval to use the ADVs, if available, to enhance SSF mitigation capabilities. NRC approval is needed since the ADVs will be used to manually control MS pressure rather than relying on the automatic opening and closing of the MSRVs at a specified pressure. ADV use will rely on manual operator action. Use of the ADVs also allows plant stabilization to occur more quickly and at lower temperatures, and it eliminates repeated cycling of the MS relief valves. Use of the ADVs is not credited in the thermal-hydraulic analyses.
As noted above, engineering changes are being performed to enable the SSF to meet the proposed success criteria for a TB flood event. Duke Energy is modifying the plant to install a higher capacity letdown line and to replace the RC (reactor coolant) makeup pump pulsation dampener in each unit. These changes will be implemented under 10 CFR 50.59.
Current success criteria cannot be met for a TB flood event occurring during off-nominal initial conditions. However, Duke Energy has demonstrated by analyses that the SSF events can be successfully and safely mitigated using the methods described above and meet the proposed off-nominal success criteria. The proposed off-nominal success criteria allows water-solid operation with no water relief through the pressurizer safety valves, and considers ambient losses and letdown and makeup of reactor coolant an acceptable method of decay heat removal during periods of low decay heat.

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                              Page 16 4 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The applicable regulatory requirements for the SSF are contained in Chapter 9 of the Oconee UFSAR. Specifically, Section 9.6.1 states that "the SSF is designed to:
: 1. Maintain a minimum water level above the reactor core, with an intact Reactor Coolant System, and maintain Reactor Coolant Pump Seal cooling.
: 2. Assure natural circulation and- core cooling by maintaining the primary coolant system filled to a sufficient level in the pressurizer while maintaining sufficient secondary side cooling water.
: 3. Transfer decay heat from the fuel to an ultimate heat sink.
: 4. Maintain the reactor 1% shutdown with the most reactive rod stuck fully withdrawn, after all normal sources of RCS makeup have become unavailable, by providing makeup via the Reactor Coolant Makeup Pump System which always supplies makeup of a sufficient boron concentration. (The stuck rod requirement was eliminated for fire events when NFPA 805 was adopted. See Section 9.6.2)"
The proposed change described in this LAR will not affect Oconee's conformance to these design criteria during the Turbine Building flood event except during off-nominal conditions when the amount of decay heat available is less than or equal to the heat removed by ambient losses to containment and/or letdown and makeup of reactor coolant or when decay heat is high and RCS temperature is low. For the low decay heat case, natural circulation may not occur.
However, as discussed previously in the LAR, the proposed new success criteria allows ambient losses with makeup/letdown flow paths to be credited for decay heat removal. Duke Energy's analyses concluded these decay heat removal mechanism are adequate to meet the core cooling function. For the high decay heat case with low initial RCS temperature, pressurizer level may go off-scale and water-solid conditions may be reached; however, the Rcs*will be maintained subcooled with no water relief through the PSVs.
The regulatory requirements described in Oconee UFSAR Section 9.6.2, "Design Bases" for the fire event, security-related event and station blackout event also continue to apply for the SSF.
The proposed change will not affect Oconee's conformance to any of the design criteria for those events. The Technical Specification (TS) requirements described in Oconee TS 3.10.1, Standby Shutdown Facility, also continue to apply. No changes to TS 3.10.1 are proposed.
The SSF Auxiliary Service Water (ASW) System requirements are described in Oconee UFSAR Section 9.6.3.3. The proposed change will not affect Oconee's conformance to these existing design criteria.

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                              Page 17 4.2 No Significant Hazards Consideration Duke Energy Carolinas, LLC (Duke Energy) proposes to revise the Updated Final Safety Analysis Report (UFSAR) to provide off-nominal success criteria for maintaining the reactor in a safe shutdown condition when using the Standby Shutdown Facility (SSF) to mitigate a Turbine Building (TB) Flood occurring when an Oconee Unit is not at nominal full power conditions. In addition, the LAR requests approval to use the Main Steam (MS) Atmospheric Dump Valves (AD Vs), when available, to enhance SSF mitigation capabilities.
During high decay heat and low initial RCS temperature conditions, the proposed new off-nominal success criteria allows the pressurizer to become water-solid with no water relief through the pressurizer safety valves during SSF event mitigation. Duke Energy has evaluated water-solid operation and confirmed it does not result in liquid relief through the pressurizer safety valves (PSVs) and is an acceptable method of maintaining safe shutdown during an SSF event. During low decay heat conditions, the potential for pressurizer level to go off-scale exists until there is an adequate core power history. Also, there is the potential for RCS natural circulation to stagnate as RCS ambient losses and heat loss from SSF reactor coolant makeup and letdown exceeds decay heat. Under such low decay heat conditions, the proposed new off-nominal success;criteria allows ambient losses with makeup/letdown flow paths to be credited for decay heat removal. Duke Energy's analyses concluded these decay heat removal mechanisms are adequate to meet the core cooling function.
Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
   , 1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
No. The proposed change provides off-nominal success criteria for SSF mitigated TB flood events occurring during off-nominal initial conditions. The proposed change does not impact the current success criteria for SSF events occurring during nominal full power*initial conditions. The LAR also requests NRC approval to use the MS ADVs, when available, to enhance SSF mitigation capabilities. The proposed change does not adversely impact containment integrity, radiological release pathways, fuel design, filtration systems, main steam relief valve set points, or-radwaste systems. No new radiological release pathways are created. During licensing of the SSF design, SSF performance was evaluated assuming the events that were to be mitigated by the SSF were initiated from nominal full power conditions. Duke Energy analyses demonstrate that SSF mitigated Turbine Building (TB) flood events occurring during off-nominal full power conditions can be mitigated acceptably when the proposed off-nominal success criteria are met. As such, the proposed change does not have a significant impact on the dose consequences of an accident previously evaluated. The SSF is not an event initiator; therefore, it does not affect the frequency of occurrence of accidents previously evaluated in the UFSAR. The use of off-nominal success criteria is not a precursor to a TB flood event; therefore, the proposed change does not involve a significant increase in the probability of any event requiring operel:tion of the SSF. The proposed off-nominal success criteria will continue to ensure the SSF can maintain the unit(s) in a safe

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                           Page 18 shutdown condition. As such, the proposed change does not involve a significant increase in the consequences of any event requiring operation of the SSF.
: 2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
No. The proposed UFSAR change requests approval to modify the SSF licensing basis for off-nominal conditions by using off-nominal success criteria for SSF mitigated TB flood events occurring during off-nominal conditions. Duke Energy analyses demonstrate that meeting the off-nominal success criteria is an acceptable method of mitigating the TB flood event and does not create the possibility of a new or different kind of accident. The LAR also requests NRC approval to use the main steam atmospheric dump valves, when available, to enhance the mitigation of SSF events.
The proposed change does not change the design function or operation of the SSF. The SSF is designed with the capability to mitigate a TB flood and meet specific success criteria for the entire 72 hour mis~ion time. These changes do not adversely affect this mission time.
The proposed change does not create the possibility of a new or different kind of accident since the proposed change does not introduce credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.
: 3) Does the proposed amendment involve a significant reduction in a margin of safety?
No. The proposed change requests approval of an off-nominal set of success criteria for SSF mitigated TB flood events occurring during off-nominal power conditions. Duke Energy analyses demonstrate there is adequate margin to prevent lift of pressurizer safety valves while water-solid. The proposed change does not involve operating installed equipment (ADVs) in a new or different manner. The ADVs are periodically tested and have been used successfully for a plant cooldown. Use of the ADVs to enhance the mitigation of SSF events serves to improve plant safety by preventing the pressurizer from reaching water-solid conditions and by reducing the pressure at which the MS system is controlled. ADV use also allows plant stabilization to occur more quickly and at'lower temperatures, and eliminates repeated cycling of the MS relief valves. The proposed change does not involve a change to any set points for parameters which initiate protective or mitigation action and does not have any impact on the fission product barriers or safety limits. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed revision to the wording in the UFSAR and operation of the unit in the proposed manner, (2) the proposed revision will be implemented in a manner consistent with the Commission's regulations and (3)

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                              Page 19 the issuance of the amendment will not be adverse to the common defense and security or to the health and safety of the public.
5    ENVIRONMENTAL CONSIDERATION Duke Energy has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. Duke Energy Carolinas, LLC, has determined that this license amendment request meets the criteria for a categorical exclusion as set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:
(i)   The amendment involves no significant hazards consideration.
As demonstrated in Section 4.2, this proposed change to the UFSAR does not involve a significant hazards consideration.
(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.
The proposed change will not change the types or amounts of any effluents that may be released offsite.
(iii) There is no significant increase in individual or cumulative occupational radiation exposure.
The proposed change will not increase the individual or cumulative occupational radiation exposure.
Therefore, no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment pursuant to 10 CFR 51.22(b ).
6    REFERENCES
: 1. Safety Evaluation by the Office of Nuclear Reactor Regulation Oconee Nuclear Station Standby Shutdown Facility, Docket Nos. 50-269, 50-270, and 50-287, April 28, 1983 (Accession Number 8305200103).
: 2. Letter from W.O. Parker (Duke Power Company) to H.R. Denton (NRC) dated March 28, 1980, Information in Support of Standby Shutdown Facility (Accession Number ML16134A655).
: 3. Letter from W.O. Parker (Duke Power Company) to H.R. Denton (NRC) dated February 16, 1981, Response to NRC Request for Information (Accession Number ML152388273).

License Amendment Request No. 2017 Enclosure October 20, 2017                                                                    Page 20 4:  Letter from W.O. Parker (Duke Power Company) to H.R. Denton (NRC) dated March 31, 1981, Response to NRC Request for Information (Accession Number ML15238B314).
: 5. Letter from W.O. Parker (Duke Power Company) to H.R. Denton (NRC) dated April 13, 1981, RAI Response (Accession Number ML15238B326).
: 6. Letter from H.B. Tucker (Duke Power Company) to H.R. Denton (NRC) dated September 20, 1982, RAI Response (Accession Number ML15238A655).
: 7. Letter from H.B. Tucker (Duke Power Company) to H.R. Denton (NRC) dated December 23, 1982, RAI Response (Accession Number ML15238A727).
: 8. Letter from H.B. Tucker (Duke Power Company) to H.R. Denton (NRC) dated July 26, 1985, Proposed Technical Specifications for SSF (Accession Numbers ML15264A327, ML15264A329).
: 9. NRC letter to Duke Power Company dated January 23, 1987, Inadequacy of Technical Specifications for Safe Shutdown Facility (Accession Number 8702060151 ).
: 10. Letter from H.B. Tucker (Duke Power Company) to H.R. Denton (NRC) dated August 14, 1987, Proposed Revised Technical Specifications for SSF (Accession Numbers ML15264A490, ML15264A492).
: 11. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Nos. 195, 195, and 191, Adds SSF Technical Specifications, May 11, 1992 (Accession Numbers ML012000400, ML012190128).
12 Safety Evaluation issued by NRC, "SER for Station Blackout - Oconee Nuclear Station,"
March 10, 1992 (Accession Number 9203170114).
: 13. Supplemental Safety Evaluation "Supplemental SER for Station Blackout - Oconee Nuclear Station," December 3, 1992 (Accession Number 9212110152).
: 14. Safety Evaluation issued by NRC, "Seismic Qualification of the Emergency Feedwater System," dated January 14, 1987 (Accession Number 8701300192).
15 Letter from S.L. Batson (Duke Energy) to NRC Document Control. Desk dated June 4, 2012, Licensee Event Report 2012-001-00 (ML12157A323).
16 Letter from D.A. Baxter (Duke Energy) to NRC Document Control Desk dated April 14, 2010, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants.
: 17. NRC Issuance of ONS NFPA 805 Amendments Regarding Transition to a Risk-Informed, Performance Based fire Protection Program in Accordance with 10 CFR 50.48, December 29, 2010 (Accession Number ML103630612).
: 18. Duke Energy Methodology Report DPC-NE-3000-PA, Oconee Nuclear Station, McGuire Nuclear Station, Catawba Nuclear Station, Thermal-Hydraulic Transient Analysis Methodology, Revision Sa. (Safety Evaluations for Oconee Nuclear Station dated August 8, 1994 (Accession Number ML16293A840); October 14, 1998 (Accession Number 9810190223); September 24, 2003 (Accession Number ML032670816); October 29, 2008 (Accession Number ML082800408); and July 21, 2011 (Accession Number ML11137A150).

ATTACHMENT 1-A UFSAR MARKED-UP PAGES

License Amendment Request No. 2017-03 -A October 20, 2017                                                                                                         Page 1 Oconee Nuclear Station                                                                   UFSAR Chapter 9 To verify SSF performance criteria, thermal-hydraulic (T/H) analysis was performed to demonstrate that the SSF can achieve and maintain safe shutdown following postulated turbine building floods. The analysis evaluates RCS subcooling margin using inputs that are representative of nominal full power end of cycle plant conditions. The analysis uses an initial core thermal power of 2619 MWth (102% of 2568 MWth) and accounts for 24 month fuel cycles. The consequences of the postulated loss of main and emergency feedwater were analyzed as a RCS overheating scenario. For the examined overheating scenario, an important core input is decay heat. High decay heat conditions were modeled that were reflective of maximum, end of cycle conditions. The high decay heat assumption was confirmed to be bounding with respect to the RCS subcooling response. The results of the analysis demonstrate that the SSF is capable of meeting Delet d Paragraph(s) per 2012 update.
TURB NE BUILDING FLOOD EVENT The T rbine Building Flood was one of the events that was identified in the original SSF licensin requirements. The SSF is designed to maintain the reactor in a safe shutdown conditio for a period of 72 hours following a TB Flood. No other concurrent event is assumed to occur. ;the success criteria for this event !!&sect;]to assure natural circulation and core cooling by maintaining the primary coolant system filled to a sufficient ievel in the pressurizer while maintaining sufficient secondary side cooling. The reactor shall be maintained at least 1 % ilk/k with the most reactive rod fully withdrawn. (Reference 1, 10)
S CURITY-RELATED EVENT he events that was identified in the original SSF licensing During periods of very low decay heat the SSF will be used to establish conditions that support the formation of subcooled natural circulation between the core and the SGs; however, natural circulation involving the SGs may not occur if the amount of decay heat available is less than or equal to the amount of heat removed by ambient losses to containment and/or by other means, e.g., letdown of SSF reactor coolant makeup. When these heat removal mechanisms are sufficient to remove core decay heat, they are considered adequate to meet the core cooling function and systems supporting SG decay heat removal, although available, are not necessary for core cooling.
Regarding operation in MODES 1, 2, and 3 at other than nominal full power, T-H analyses were also performed to demonstrate that the SSF could achieve and maintain safe shutdown following postulated turbine building floods. A nominal full power condition is defined as a unit at 100% power for approximately 4 days of operation which provides the decay heat required to meet the nominal SSF success criteria. The results of the analyses demonstrate thatJor this range of initial conditions the SSF is capable of meeting the specified success criteria: 1) to maintain the reactor at least 1% tlk/k shutdown with the most reactive rod fully withdrawn, and 2) maintain a minimum water level above the reactor core and conditions that support the formation of subcooled natural circulation between the core and SGs such that there is no water relief through the pressurizer safety valves.
In some cases that initiated with low decay heat, pressurizer level was not maintained on scale; however, a minimum water level above the reactor core was maintained and conditions that support the for~ation of subcooled natural circulation between the core and the SGs were maintained. In cases where the pressurizer did go water-solid, there was no liquid relief through the pressurizer safety valves.
(31 DEC 2016)                                                                                         9.6-3

License Amendment Request No. 2017-03 -A October 20, 2017                                                                                           Page 2 Oconee Nuclear Station                                                               UFSAR Chapter 9 9.6.3.3      Auxiliary Service Water (ASW) System The SSF ASW System is designed to cool the RCS during a station blackout and in conjunction with the loss of the normal and Emergency Feedwater System by providing steam generator cooling.
The SSF ASW pump is the major component of the system. One motor driven SSF ASW pump, powered from OTS1 Switchgear, serves all three units and is located in the SSF. The suction supply for the SSF ASW pump, the SSF HVAC service water pumps, and the SSF DSW pump is lake water from the embedded Unit 2 condenser circulating water piping. A portable submersible pump that can be installed in the intake canal and powered from the SSF is available to replenish the water supply in the embedded CCW pipe if both forced CCW and siphon flow through the CCW pipe are lost.
The SSF ASW flow rate provided to each unit's steam generators is controlled using the motor
        <;>perated valves on each unit's SSF ASW supply header. Manually operated bypass valves, installe9 in parallel with the motor-operated valves, are also available to: ;
: 1. Provide SSF ASW Flow control at low SSF ASW Flow rates.
: 2. Provide more precise SSF ASW Flow control when used in parallel with the motor-operated valves.
The SSF ASW pump is sized to provide enough flow to all 3 Oconee units to adequately remove decay heat from the RCS and maintain natural circulation in the RCS. An SSF ASW pump minimum flow line is provided to ensure that the pump minimum flow requirements are met. The SSF ASW system, pump and valves are operated and tested from the SSF only. The SSF ASW system is shown on Figure 9-36.
Auxiliary service water enters the steam generators via the normal emergency feedwater ring headers. - - - - - - - - - ,
The SSF ASW S stem rovides he motive force for the SSF ASW suction i                   air e"ector. The Main Steam pressure is controlled automatically by the main steam relief valves or manually by the atmospheric dump valves (ADVs). When the ADVs are operated in this manner, communication with the SSF Control Room is in place to coordinate main steam pressure control with RCS pressure/temperature parameters. Local main steam pressure indication is also available at the ADVs.
pass flow. Therefore, full SSF RC Makeup System seal injection flow will be provided to the RC pump seals in time to prevent seal degradation or failure.
Though not a requirement for operability, the SSF diesel generator should be aligned to carry SSF loads and the SSF ASW pump should be operated to provide a large enough load so that diesel souping concerns are not a problem when the Emergency Start pushbutton is used to start the SSF diesel engines and continued operation of the SSF diesel engines is desired.
While continued operation of the SSF diesel engines when they are lightly loaded is* possible (i.e. one, two or three SSF RC makeup pumps operating without operating the SSF ASW pump), lightly loading the engines in this manner is not preferred due to the potential for a fire in the diesel exhaust if a large load is added after souping of the engine occurs.
Portions of the SSF ASW system are credited to meet the Extensive Damage Mitigation Strategies commitments per NEI 06-12 (8.5.b) and NEI 12-06 (FLEX). Some of these commitments have been incorporated into the Oconee Nuclear Station operating license Section H - Mitigation Strategy License Condition.
(31 DEC 2016)                                                                                     9.6-9

ATTACHMENT 1-B UFSAR RETYPED PAGES

License Amendment Request No. 2017-03 -B October 20, 2017                                                                                   Page 1 Oconee Nuclear Station                                                           UFSAR Chapter 9 250&deg;F with a long term strategy for reactivity, decay heat removal and inventory/pressure control. Long-term subcooled natural circulation decay heat removal is provided by supplying lake water to the steam generators and steaming to atmosphere. The extended coping period at these conditions is based on the significant volume .of water available for decay heat removal and reduced need for primary makeup to only match nominal system losses. A stuck rod is not required to be postulated for this event. Initial conditions are 100% power with sufficient decay heat such that natural circulation can be achieved. The hypothesized fire is to be considered an "event", and thus need not be postulated concurrent with non-fire-related failures in safety systems, other plant accidents, or the most severe natural phenomena (Reference 31).
Deleted Paragraph(s) per 2015 update.
Deleted Paragraph(s) per 2012 update.
TURBINE BUILDING FLOOD EVENT The Turbine Building Flood was one of the events that was identified in the original SSF licensing requirements. The SSF is designed to maintain the reactor in a safe shutdown condition for a period of 72 hours following a TB Flood. No other concurrent event is assumed to occur. To verify SSF performance criteria, thermal-hydraulic (T/H) analysis was performed to demonstrate that the SSF can achieve and maintain safe shutdown following postulated turbine building floods. The analysis evaluates RCS subcooling margin using inputs that are representative of nominal full power end of cycle plant conditions. The analysis uses an initial core thermal power of 2619 MWth (102% of 2568 MWth) and accounts for 24 month fuel cycles.
The consequences of the postulated loss of main and emergency feedwater were analyzed as a RCS overheating scenario. For the examined overheating scenario, an important core input is decay heat. High decay heat conditions were modeled that were reflective of maximum, end of cycle conditions. The high decay heat assumption was confirmed to be bounding with respect to the RCS subcooling response. The results of the analysis demonstrate that the SSF is capable of meeting the success criteria for this event to assure natural circulation and core cooling by maintaining the primary coolant system filled to a sufficient level in the pressurizer while maintaining sufficient secondary side cooling. The reactor shall be maintained at least 1% L'lk/k with the most reactive rod fully withdrawn. (Reference 1, 10)
During periods of very low decay heat the SSF will be used to establish conditions that support the formation of subcooled natural circulation between the core and the SGs; however, natural circulation involving the SGs may not occur if the amount of decay heaf available is less than or equal to the amount of heat removed by ambient losses to containment and/or by othe*r means, e.g., letdown of SSF reactor coolant makeup. When these heat removal mechanisms are sufficient to remove core decay heat, they are considered adequate to meet the core cooling function and systems supporting SG decay heat removal, although available, are not necessary for core cooling.
Regarding operation in MODES 1, 2, and 3 at other than nominal full power, T-H analyses were also performed to demonstrate that the SSF could achieve and maintain safe shutdown following postulated turbine building floods. A nominal full power condition is defined as a unit at 100% power for approximately 4 days of operation which provides the decay heat required to meet the nominal SSF success criteria. The results of the analyses demonstrate that for this range of initial conditions the SSF is capable of meeting the specified success criteria: 1) to maintain the reactor at least 1% Llk/k shutdown with the most reactive rod fully withdrawn, and
: 2) maintain a minimum water level above the reactor core and conditions that support the formation of subcooled natural circulation between the core and SGs such that there is no water relief through the pressurizer safety valves.
(31 DEC 2016)                                                                               9.6-3

License Amendment Request No. 2017-03 -B October 20, 2017                                                                                          Page 2 UFSAR Chapter 9                                                                 Oconee Nuclear Station In some cases that initiated with low decay heat, pressurizer level was not maintained on scale; a minimum water level above the reactor core was maintained and conditions that support the formation of subcooled natural circulation between the core and the SGs were maintained. In cases where the pressurizer did go water-solid, there was no liquid relief through the pressurizer safety valves.
SECURITY-RELATED EVENT A Security Related Event was one of the events that was identified in the original SSF licensing requirements. The SSF is designed to achieve and maintain a safe shutdown condition for this event. No other concurrent event is assumed to occur. (Reference 1) The success criteria for this event is to assure the core will not return to criticality, the active fuel will not be uncovered, and long-term natural circulation will not be halted. (Reference 41)
STATION BLACKOUT EVENT This event was licensed after the   design of the SSF was completed and approved by NRC. The SSF was credited as the method      the plant would employ to mitigate a SBO event. (References 38 and 39) The success criteria     is to maintain the core covered for 4 hours. No stuck rod is assumed for this event. Initial      conditions are 100% power and 100 days of operation.
(Reference 40)                                                                                        '
SSF TORNADO DESIGN CRITERIA This is a design criterion for the SSF that livas cammitted to as part of the original SSF licensing correspondence. All parts of the SSF itself that are required for mitigation of the SSF events are required to be designed against tornado winds and associated tornado missiles. This requirement is satisfied through appropriate design of the SSF structure. This requirement does not extend to SSCs that were already part of the plant which SSF relies upon and interfaces with for event mitigation. It is important to note that the SSF was not licensed to mitigate a tornado event or a tornado missile event (Reference 1). Tornado design requirements for the plant itself are addressed in Section 3.2.2. A subsequent issue related to crediting SSF ASW as an alternative for EFW tornado missile protection vulnerabilities is discussed below (see EFW Tornado Missile Design Criteria).
EFW SEISMIC DESIGN CRITERIA (GL 81-14)
During the seismic qualification review of the Oconee EFW system in the 1980s, the N RC postulated that a seismic event could break a pipe and potentially cause a flood of the turbine building thereby submerging and failing the EFW pumps. The NRC wanted to ensure that the EFW System was seismically designed and could withstand a single failure, as well. As an alternative to upgrading the EFW System, NRC credited the use of the SSF ASW System and HPI Feed & Bleed (Reference 34). These two decay heat removal systems are seismically designed and independent from each other. The event postulated by GL 81-14 (a seismic break) was a special condition imposed on ONS to evaluate the EFW design. It was not intended to re-define the SSF mitigated TB Flood (which does not concurrently consider a seismic event, nor does it impose a single failure). Although both "events" are TB Floods, they are two separate licensing actions with different scopes, different acceptance criteria, and different purposes. The GL 81-14 flood does not have specified initial conditions, other mitigation assumptions, or success criteria to be considered because it is not an event, only an EFW design criterion (Reference 34).
9.6-4                                                                                       (31 DEC 2016)

License Amendment Request No. 2017-03 -B October 20, 2017                                                                                     Page 3 UFSAR Chapter 9                                                              Oconee Nuclear Station 9.6.3.3      Auxiliary Service Water (ASW) System The SSF ASW System is designed to cool the RCS during a station blackout and in conjunction with the loss of the normal and Emergency Feedwater System by providing steam generator cooling.
The SSF ASW pump is the major component of the system. One motor driven SSF ASW pump, powered from OTS1 Switchgear, serves all three units and is located in the SSF. The suction supply for the SSF ASW pump, the SSF HVAC service water pumps, and the SSF DSW pump is lake water from the embedded Unit 2 condenser circulating water piping. A portable submersible pump that can be installed in the intake canal and powered from the SSF is available to replenish the water supply in the embedded CCW pipe if both forced CCW and siphon flow through the CCW pipe are lost.
The SSF ASW flow rate provided to each unit's steam generators is controlled using the motor operated valves on each unit's SSF ASW supply header. Manually operated bypass valves, installed in parallel with the motor-operated valves, are also available to:
: 1. Provide SSF ASW Flow control       at low SSF ASW Flow rates.
: 2. Provide more precise SSF ASW Flow control when used in parallel with the motor-operated valves.
The SSF ASW pump is sized to provide enough flow to all 3 Oconee units to adequately remove decay heat from the RCS and maintain natural circulation in the RCS. An SSF ASW pump minimum flow line is provided to ensure that the pump minimum flow requirements are met. The SSF ASW system, pump and valves are operated and tested from the SSF only. The SSF ASW system is shown on Figure 9-36.
Auxiliary service water enters the steam generators via the normal emergency feedwater ring headers. Main Steam pressure is controlled automatically by the main steam relief valves or manually by the atmospheric dump valves (ADVs). When the ADVs are operated in this manner, communication with the SSF Control Room is in place to coordinate main steam pressure control with RCS pressure/temperature parameters. Local main steam pressure indication is also available at the ADVs.
The SSF ASW System provides the motive force for the SSF ASW suction pipe air ejector. The air ejector is needed 1o maintain siphon flow to the SSF HVAC service water pump, the SSF DSW pump, and the SSF ASW pur:np when the water level in the U2 CCW supply pipe becomes too low.
The SSF ASW System provides adequate SG cooling to reduce and maintain RCS pressure below the pressure where the SSF RC makeup pump discharge relief valve, HP-404, begins to pass flow. Therefore, full SSF RC Makeup System seal injection flow will be provided to the RC pump seals in time to prevent seal degradation or failure.
Though not a requirement for operability, the SSF diesel generator should be aligned to carry SSF loads and the SSF ASW pump should be operated to provide a large enough load so that diesel souping concerns are not a problem when the Emergency Start pushbutton is used to start the SSF diesel engines and continued operation of the SSF diesel engines is desired.
While continued operation of the SSF diesel engines when they are lightly loaded is possible (i.e. one, two or three SSF RC makeup pumps operating without operating the SSF ASW pump), lightly loading the engines in this manner is not preferred due to the potential for a fire in the diesel exhaust if a large load is added after souping of the engine occurs.
Portions of the SSF ASW system are credited to meet the Extensive Damage Mitigation Strategies commitments per NEI 06-12 (B.5.b) and NEI 12-06 (FLEX). Some of these commitments have been incorporated into the Oconee Nuclear Station operating license Section H - Mitigation Strategy License Condition.
9.6- 10                                                                              (31 DEC 2016)

ATTACHMENT 3 Thermal-Hydraulic 'Models for SSF Transient Analysis [Non-Proprietary]

License Amendment Request No. 2017-03 October 20, 2017                                                                           Page 1 Thermal-Hydraulic Models for SSF Transient Analysis The thermal-hydraulic analyses for ah SSF-mitigated Turbine Building flood event described in the LAR Enclosure have been performed using Duke Energy's RETRAN-3D ONS model (Reference 1). This model has previously been approved for use in the ONS UFSAR Chapter 6 and Chapter 15 accident analyses (References 2, and 3). The RETRAN-3D model has been further modified for Turbine Building flood analysis to capture important phenomena for longer-duration SSF events. These modifications include modeling ambient heat losses from the RCS and pressurizer, and pressurizer nodalization changes to better model thermal stratification of fluid in the pressurizer for cases that experience large pressurizer insurges I outsurges or eventually fill water-solid. These model changes enhance the evaluation of the proposed off-nominal success criteria. The changes to the RETRAN-3D model and its application for SSF Turbine Building flood analysis are evaluated against the conditions and limitations of the generic RETRAN-3D Safety Evaluation Report (Reference 5) later in this attachment.
Additionally, for those Turbine Building flood cases occurring at off-nominal conditions that eventually transition to a water-solid pressurizer condition, analyses were also performed using Duke Energy's RELAP5/MOD2-B&W ONS thermal-hydraulic model. Duke Energy's RELAP5/MOD2-B&W model has previously been approved for use' in the ONS UFSAR Chapter 6 Loss of Coolant Accident (LOCA) mass and energy release analyses (Reference 2).
For performing transient analysis of an SSF-mitigated Turbine Building flood, the RELAP5/MOD2-B&W ONS model has also been modified to: 1) include ambient heat losses from the RCS and pressurizer, and 2) to improve its capability to model thermal stratification of fluid in the pressurizer region. The RELAP5/MOD2-B&W analyses were used to confirm the RETRAN-3D RCS response for the transition to and during a water-solid pressurizer condition.
For these cases, results from the RELAP5/MOD2-B&W and RETRAN-3D analyses show good agreement. For both analysis methods, the RCS pressure response for the transition to and during a water-solid pressurizer condition is reasonable and well-behaved. Additionally, there are no numerical discontinuities in the code predictions for filling or emptying of the pressurizer.
Finally, for analyzing off-nominal decay heat conditions, decay heat is calculated differently than describe.d in References 1 and 2. The alternative calculation provides a better estimate of core decay heat for the spectrum of initial conditions analyzed for off-nominal operations. The decay heat response used for these cases continues to include conservatisms to bound future operating cycles.
The aforementioned modifications to Duke Energy's RETRAN-3D and RELAP5/MOD2-B&W thermal-hydraulic models are described in more detail below. The modified RETRAN-3D and RELAP5/MOD2-B&W thermal-hydraulic models have been developed specifically for performing transient analysis of an SSF-mitigated Turbine Building flood event. Duke Energy does not intend to apply these models or modifications to the ONS UFSAR Chapter 6 and 15 accident analyses; nor will their application extend beyond performing transient analysis of SSF-mitigated events. Therefore, revisions to Duke Energy's governing methodology reports, DPC-NE-3000-PA and DPC-NE-3003-PA (References 1 and 2, respectively) will not be made.

License Amendment Request No. 2017-03 October 20, 2017                                                                            Page 2 Should the need arise, Duke Energy intends to use the modified RETRAN-3D ONS thermal-hydraulic model in future SSF Turbine Building flood analyses to evaluate changes to the plant and operator guidance.
Model Modifications Section 3.1 of the LAR Enclosure describes certain off-nominal initial conditions for which the current Turbine Building flood success criteria cannot be met for the SSF. For these off-nominal initial conditions, indicated pressurizer level may go off-scale, the pressurizer can fill water-solid, or natural circulation in the RCS loops may not occur if the amount of decay heat available is less than or equal to the amount of heat removed by ambient losses to containment and/or by other means (e.g., letdown of required reactor coolant makeup flow). As a result, Duke Energy is requesting approval of off-nominal success criteria for these conditions. Duke Energy has performed analysis to confirm the off-nominal success criteria are met for a Turbine Building flood event using modified versions of their existing RETRAN-3D and RELAP5/MOD2-B&W ONS thermal-hydraulic models. These modifications include use of [
                                        ] to model ambient heat losses, and changes to the pressurizer region nodalization to improve the modeling capability for thermal stratification of fluid in the pressurizer region. These model improvements enhance the evaluation of the proposed off-nominal success criteria.
While these modifications have been developed and applied for the analysis of an SSF-mitigated Turbine Building flood event, Duke Energy considers the modifications to be equally suitable for use in analysis of SSF-mitigated station blackout or fire events.
Additionally, for analyzing off-nominal decay heat conditions, decay heat is calculated differently than described in References 1 and 2 to provide a better estimate of core decay heat for the spectrum of initial conditions analyzed for off-nominal operations.
RCS and Pressurizer Ambient Heat Losses The existing exterior heat structures (or conductors) on the RCS and pressurizer components are originally modeled with [                                       ] The heat structure inputs are selected to [
           ] To model ambient heat losses from the RCS and pressurizer regions, these heat structures are converted to [
                                                                                 ] Plant data is used to determine the total ambient heat loss to the reactor building, as well as the specific heat losses from the pressurizer structures. Within the RETRAN-3D and RELAP5/MOD2-B&W models, ambient heat losses are [
                                                                                         ] This change is considered an enhancement to the existing models since it allows them to more accurately model the impact of real phenomena on the RCS and *pressurizer response for longer-duration events associated with the SSF.

License Amendment Request No. 2017-03 Attachment 3 October 20, 2017                                                                            Page 3 Specific to the RELAP5/MOD2-B&W ONS thermal-hydraulic model described in Reference 2, the reactor vessel upper head region is divided into [                                    ] Due to nodalization limitations, under natural circulation conditions in the RCS, the top-most reactor
,vessel upper head node receives minimal flow and the,node effectively becomes a dead-end volume. With ambient heat losses modeled from this region, the dead-end volume effect causes this top-most node to cool down more rapidly than the adjacent volume that continually receives flow under natural circulation conditions. This is non-physical since buoyancy effects would cause circulation and mixing of the RCS fluid in this region. Therefore, for the RELAP5/MOD2-B&W ONS model, [
Section 3.1 of the LAR Enclosure describes three separate initial condition scenarios for an SSF Turbine Building flood event: nominal full power conditions; lower initial power or burnup in MODES 1, 2, or 3; and, MODE 3, low initial temperature with high decay heat. For each initial condition scenario, cases were performed to examine the Qperating window for this set of initial conditions. For all cases, pressurizer ambient heat losses are modeled to exacerbate the collapse of the pressurizer steam space and challenge the ability of the SSF-controlled pressurizer heaters to restore' and maintain a saturated fluid condition in the pressurizer. RCS ambient heat losses are modeled for cases where maximizing the RCS cooldown is conservative.
Pressurizer Nodalization for Thermal Stratification of Pressurizer Fluid As described in Section 2.1..2 of the LAR Enclosure, the SSF has been credited in the ONS licensing basis for mitigation of a variety of events and conditions. Many of these events can be generically classified as RCS overheating and overcooling transients. The pressurizer plays a significant role in regulating RCS pressure during these events, and experiences several important phenomena for both overcooling and overheating conditions.
In general, for overheating events, there is an initial insurge of subcooled liquid into the pressurizer from thermal expansion of the RCS inventory. If the overheating transient is short-lived, the presence of subcooled liquid in the pressurizer has little impact on the immediate response. This is becaus_e there is little mixing in the fluid region under these conditions and buoyancy (density) effects cause the colder liquid in the pressurizer to remain near the bottom of the vessel, while the hotter (originally saturated) liquid remains near the top of the water column and in contact with the vapor space. Thermal stratification of the pressurizer liquid helps limit the amount of steam condensation that occurs at the steam-liquid interface during these pressure excursions.
For RCS overcooling transients, saturated liquid in the pressurizer flashes to steam, expands, and limits the depressurization rate of the RCS. As a result, insurges of subcooled liquid to the pressurizer can limit the ability of the pressurizer to regulate subsequent depressurizations of the RCS. For more severe overcooling events, the pressurizer may empty as a result of the initial overcooling, but subcooled liquid will refill the pressurizer once operators restore RCS pressure or pressurizer level to the specified operating range. In the longer-term recovery

License Amendment Request No. 2017-03 October 20, 2017                                                                              Page 4 phase, operator actions and pressurizer heaters are able to re-saturate the fluid in the pressurizer and restore RCS pressure to a desired range.
For SSF-mitigated events with limited pressurizer heater capacity, the ability to re-saturate the subcooled liquid is greatly diminished. Additionally, pressurizer ambient heat losses can cause condensation of the vapor space on internal structural surfaces. Continued condensation of the vapor space leads to a reduction in RCS pressure and increases in pressurizer level. As the vapor space collapses, the continual insurge of subcooled liquid challenges the ability of the pressurizer heaters to re-saturate the fluid. For some scenarios, the pressurizer can eventually fill to a water-solid condition. Under these conditions, RCS pressure control is provided by balancing makeup and letdown flow with the SSF letdown line.
In order to evaluate longer-duration SSF events, it is important that the thermal-hydraulic models be capable of capturing the effects from thermal stratification and ambient heat losses in the pressurizer. Modifications for modeling ambient heat losses are described above.
Section 2.2.1.4 of Reference 1 describes the original pressurizer modeling approach in the RETRAN-30 ONS model. The RETRAN-30 pressurizer is modeled using [
                                                             ] While suitable for many transient analysis applications, this modeling approach is considered inappropriate for longer-duration SSF events in that the pressurizer liquid region is homogenized to a single lumped temperature, effectively ignoring thermal stratification of the fluid region. As a result, any insurge of subcooled liquid to the pressurizer, regardless of magnitude or duration, immediately begins to reduce the calculated fluid temperature in the pressurizer. Even for small insurges to the pressurizer, the immediate reduction in pressurizer fluid temperature challenges the ability of the pressurizer heaters to maintain the bulk fluid at a saturated condition over a 72-hour period.
To improve the modeling capability for thermal stratification of fluid in the pressurizer region, the
[                                        ] pressurizer volume in the RETRAN-30 ONS model is renodalized by
                        ] The pressurizer wall heat structures are updated to match the renodalized geometry. This nodali.zation change reduces the numerical diffusion (or drift) of energy described above for the [                       ] pressurizer approach, while still allowing for phase separation and thermal non-equilibrium in the pressurizer regions where it matters most (i.e., from the pressurizer heater region and up).
Section 2.1 of Reference 2 describes the original pressurizer modeling approach in the RELAP5/M002-B&W ONS model. The RELAP5/M002-B&W pressurizer is modeled with [
                                                                                       ] In order to increase the spatial resolution of axial temperature gradients that can establish in longer-duration SSF events, the pressurizer volume and heat structures in the RELAP5/M002-B&W model are renodalized with smaller nodes. Finer nodalization of the pressurizer allows better (localized) prediction of ambient heat losses from the steam and liquid regions, as well as better predictions of thermal stratification of the liquid region during insurges, outsurges, and large

License Amendment Request No. 2017-03 October 20, 2017                                                                           Page 5 pressure drops. Therefore, the pressurizer volume and heat structures are subdivided into
                                              ]
Section 3.1 of the LAR Enclosure describes three separate initial condition scenarios for an SSF Turbine Building flood event. For all initial condition scenarios, the RETRAN-30 T-H analyses use the [                  ] pressurizer model. RELAP5/MOD2-B&W T-H analyses are performed for scenarios with low initial temperature and high decay heat to confirm the RETRAN-30 RCS response under water-solid RCS conditions. The RELAP5/MOD2-B&W analyses use the [               ] pressurizer model.
Decay Heat Calculation for Off-Nominal Conditions Section 2.2.6.1 of Reference 1 and Section 3.3.1.1 of Reference 2 describe the method for calculating decay heat in the RETRAN-30 and RELAP5/MOD2-B&W m.odel applications, respectively. For analyzing conditions where high decay heat is conservative, the NRC-approved model applications use the 1979 ANS standard decay heat model (Reference 4) and include 2cr uncertainties. This decay heat model continues to be. used for high decay heat applications when analyzing nominal full power scenarios in Section 3.1 of the LAR Enclosure.
Intermediate power levels with different operating histories are also examined in the nominal full power scenario envelope. These intermediate power levels are non-limiting, but are examined to estimate an operational boundary for nominal full power conditions, where separate SSF TBF success criteria would apply. Since these cases are analyzed as overcooling events and low decay heat is conservative, decay heat calculations are performed to conservatively estimate a representative minimum post-trip power response at the examined initial power level condition and operating history.
For off-nominal initial conditions for an SSF-mitigated Turbine Building flood event, decay heat is modeled differently to provide a better estimate of core decay heat for the spectrum of initial conditions analyzed for these operating conditions. Low initial power or burn up scenarios described in Section 3.1 of the LAR Enclosure are analyzed as overcooling events; as such, low decay is conservative. Therefore, for the limiting conditions in a low initial power or burn up scenario, the long-term decay heat response is modeled as a low constant value that conservatively bounds future cycle conditions.
For scenarios described in Section 3.1 of the LAR Enclosure that examine low initial RCS temperatures with high decay heat, the low RCS temperature conditions are achieved through a controlled cooldown of the unit following a typical unit shutdown at End-of-Cycle (EOC) conditions. The Turbine Building flood event is postulated to occur sometime during the plant cooldown. For analyzing these low initial RCS temperature conditions where high decay heat is conservative, an EOC decay heat profile is calculated assuming continuous operation at 102%
full-power conditions followed by a multi-hour power reduction to 0% power conditions. The power reduction duration is based on a review of shutdown data for all three units over multiple operating cycles. The decay heat profile used in the off-nominal SSF TBF analyses for high decay heat and low initial temperatures reflects 24-month operating cycles lengths, and includes additional conservatisms to bound future operating cycles.

License Amendment Request No. 2017-03 October 20, 2017                                                                            Page 6 Conditions and Limitations of the RETRAN-3D Safety Evaluation Report Appendix C of DPC-NE-3000-PA (Reference 1) evaluates the conditions and limitations in the generic RETRAN-3D Safety Evaluation Report (Reference 5) for the application of RETRAN-3D to the Oconee Nuclear Station with replacement once-through steam generators. The evaluation demonstrates that Duke Energy's RETRAN-3D ONS thermal-hydraulic model, as described in Reference 1, is appropriately justified and within the RETRAN-3D Safety Evaluation Report (SER) conditions and limitations. As described above, modeling changes to Duke's RETRAN-3D ONS thermal-hydraulic model from Reference 1 include: 1) use of [
                    ] on the external heat structures of the RCS and pressurizer volumes that are in contact with the Reactor Building environment; and 2) increased nodalization of the lower pressurizer region to allow limited modeling of thermal stratification of fluid in the pressurizer.
With the exception of these model modifications, application of the modified ONS RETRAN-3D model for analyzing the SSF-mitigated Turbine Building flood event is considered consistent with the NRG-approved use of the RETRAN model in Reference 1, and complies with the conditions and limitations in Reference 5.
Applicable SER conditions and limitations from Reference 5 include Items 18, 31, and 37 for pressurizer modeling, Item 28 for use of the local condition heat transfer model applied to the pressurizer heat structures, and Item 3 for use of the general transport model to track boron concentration in the RCS. Regarding these limitations and conditions, changes to the pressurizer nodalization for thermal stratification are consistent with the NRC staff position and recommendations described in Item 18; namely, normal nodes are added [
                                                    ] and results from cases that fill or empty the [
                   ] pressurizer volume are verified to not have any numerical discontinuities during these transitional time periods. Items 31 and 37 were resolved by NRC staff during the review of RETRAN-3D, and Duke's modeling approach is consistent with the staff's resolution.
For Item 28 in the generic RETRAN-3D SER conditions and limitations, the [
                                                                     ] as approved by the staff in Reference 1. The pressurizer heat structures in Reference. 1 are modeled with [
] For modeling pressurizer ambient heat losses during an SSF Turbine Building flood event, the pressurizer heat structures are converted to [
                ] This approach is consistent with the limitations described in Item 28 and with the currently approved use of the local condition heat transfer model in Duke's RETRAN-3D thermal-hydraulic model, as described in Reference 1.
Regarding Item 3 in the generic RETRAN-3D SER conditions and limitations, boron transport in the TB flood T-H analyses is modeled as a contaminant in the RETRAN general transport model. The calculated RCS boron concentration does not affect the T-H analyses as the boron reactivity feedback effect is neglected in the point kinetics model during the analysis. Instead, results for core boron concentration and fluid temperature are used to externally verify shutdown

License Amendment Request No. 2017-03 October 20, 2017                                                                          Page 7 margin requirements are maintained on a cycle-specific basis. The NRC staff position for modeling boron transport with the RETRAN general transport model identifies the diffusive nature of the general transport model solution scheme, particularly if the Courant limit is exceeded. The natural circulation RCS flow rates in the TB flood T-H analyses are very low and do not approach the Courant limit by a large margin. Additionally, since core inlet temperature changes relatively slowly for the TB flood event, shutdown margin evaluations are largely insensitive to variations in the time-dependent core boron concentration. As such, this modeling approach is considered consistent with the limitations described in Item 3 and with the currently approved application for modeling boron transport in Duke's RETRAN-3D thermal-hydraulic model, as described in Reference 1.
References
: 1. Duke Energy Methodology Report DPC.:.NE-3000-PA, Oconee Nuclear Station, McGuire Nuclear Station, Catawba Nuclear Station, Thermal-Hydraulic Transient Analysis Methodology, Revision Sa. (Safety Evaluations for Oconee Nuclear Station dated August 8, 1994; October 14, 1998; September 24, 2003; October 29, 2008; and July 21, 2011)
: 2. Duke Energy Methodology Report DPC-NE-3003-PA, Oconee Nuclear Station, Mass and Energy Release and Containment Response Methodology, Revision 1b. (Safety Evaluations dated March 15, 1995; September 24, 2003)                -
: 3. Duke Energy Methodology Report DPC-NE-3005-PA, Oconee Nuclear Station, UFSAR Chapter 15 Transient Analysis Methodology, Revision 5. (Safety Evaluations dated October 1, 1998; May 25, 1999; September 24, 2003; October 29, 2008; July 21, 2011; and April 29, 2016)
: 4. American National Standard for Decay Heat Power in Light Water Reactors, ANSl/ANS-5.1, American Nuclear Society, August 29, 1979.
: 5. Letter,.S. A. Richards (NRC) to G. L. Vine (EPRI), Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, "RETRAN-3D -A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems", January 25, 2001.

ATTACHMENT 4 AFFIDAVIT

License Amendment Request No. 2017-03  - Duke Energy Affidavit                                                       Page 1 AFFIDAVIT OF JOSEPH DONAHUE
: 1. I am a Vice President of Duke Eiwrgy Carolinas, LLC (Duke Energy), and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.
: 2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke Energy's application for withholding which accompanies this affidavit.
: 3. I have knowledge of the criteria used by Duke Energy in designating information as proprietary or confidential. I am familiar with the Duke Energy information contained in the proprietary version of Attachment 2 of the License Amendment Request.
: 4. Pursuant to the provisions of paragraph (b) (4) of 10 CFR 2.390, the following is fumisl;ied for consideration by the NRC in determining whether t;he information sought to be withheld from public disclosure should be withheld.
(i)    The information sought to be withheld from public disclosure is owned by Duke Energy and has been held in confidence by Duke Energy and its consultants.
(ii)   The information is of a type that would customarily be held in confidence by Duke Energy. The information consists of analysis methodology details that provide a competitive advantage to Duke Energy.
(iii) The information was transmitted to the NRC in confidence and under the provisions of 10 CFR 2.390, it is to be received in confidence by the NRC.
(iv) The information sought to be protected is not available in public to the best of our knowledge and belief.
(v)    The Duke Energy proprietary information sought to be withheld in the submittal is that which is marked by brackets in the proprietary version of the Attachment 2 to the License Amendment Request. This information is consistent with marked proprietary information in NRC-approved Duke Energy methodology reports DPC-NE-3000-PA and DPC-NE-3003-PA referred to in Attachment 2 to the License Amendment Request. This information enables Duke to:
(a) Support license amendment and Technical Specification revision request for its Oconee reactors.
(b) Perform transient and accident analysis calculations for Oconee.

 **License Amendment Request No. 2017-03 Attachment 4 - Duke Energy Affidavit                                                           .Page 2 (vi) The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.
(a) Duke Energy uses this informa~ion to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.
(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants for the purpose of supporting the operation and licensing of nuclear power plants.
(c) The subject information could only be duplicated by competitors at similar expense that incurred by Duke Energy.
: 5. Public disclosure of this information is likely to cause harm to Duke Energy because it would allow competitors in the nuclear industry to benefit from the results of significant development program without requiring a commensurate expense or allowing Duke Energy to recoqp a portion of its expenditures or benefit from the sale of tpe information.
Joseph Donahue affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.
C'\         r ~ G-.-J-Josepl}I;Qnahue Vdol:kA q 1 ZOL T Date         .
My Commission Expires:       _5i..=;.._0~N....::..>._.-..__l+J~Z--=O"---'-&=--=lQi=----
Lisa Salvador NOTARY PUBLIC State of South Carolina My Commission Expires SEAL                                    June 1, 2026}}