ML13228A268

Licensing Basis for Protected Service Water System - Updated Responses to Request for Additional Information Item Nos. 107, 109(a) and 109(b); License Amendment Request (LAR) 2008-07-Supplement 6

ML13228A268
Person / Time
Site:	Oconee
Issue date:	08/07/2013
From:	Batson S Duke Energy Carolinas, Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13228A268 (91)

Text

DUKE

ENERGY, SCOTT L.

Vice BATSON President Oconee Nuc/earStation Duke Energy ONOI VP / 7800 RochesterHwy Seneca, SC 29672 864-873-3274 10 CFR 50.90 864-873-4208 fax Scott.Batson@duke-energy.com August 7, 2013 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852-2746

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287, Renewed Operating Licenses DPR-38, DPR-47, and DPR-55 Licensing Basis for the Protected Service Water System - Updated Responses to Request for Additional Information Item Nos. 107, 109(a), and 109(b);

License Amendment Request (LAR) 2008 Supplement 6

References:

1. Letter from T. Preston Gillespie, Jr., Vice President, Oconee Nuclear Station, Duke Energy Carolinas, LLC, to the U. S. Nuclear Regulatory Commission, "Tornado and High Energy Line Break (HELB) Mitigation License Amendment Requests (LARs) - Responses to Request for Additional Information," dated December 16, 2011.

2. Letter from T. Preston Gillespie, Jr., Vice President, Oconee Nuclear Station, Duke Energy Carolinas, LLC, to the U. S. Nuclear Regulatory Commission, "Tornado and High Energy Line Break (HELB) Mitigation License Amendment Requests (LARs) - Supplemental Responses to Request for Additional Information (RAI) Nos. 61, 62, and 107," dated January 20, 2012.

3. Letter from T. Preston Gillespie, Jr., Vice President, Oconee Nuclear Station, Duke Energy Carolinas, LLC, to the U. S. Nuclear Regulatory Commission, "Tornado and High Energy Line Break Mitigation License Amendment Requests -

Response to Request for Additional Information (RAI) Item No. 109," dated March 16, 2012.

By letter dated December 16, 2011, Duke Energy Carolinas, LLC (Duke Energy),

submitted a consolidated License Amendment Request (LAR) for the Oconee Nuclear Station (ONS) proposing revisions to the High Energy Line Break (HELB) licensing bases (Ref. 1). This submittal included the initial Protected Service Water (PSW) and Battery Cell Parameter Technical Specifications (TS) and TS Bases as part of the response to RAI item No. 107.

The Staff subsequently requested Duke Energy consider revising its response to RAI item No. 107 to include a number of risk-reducing contingency actions which would provide assurances that key plant equipment is available to safely shutdown the units in

U. S. Nuclear Regulatory Commission August 7, 2013 Page 2 addition to the changing of the PSW battery parameter format to comply with the information given in Babcock and Wilcox Standard TSs. This request was addressed in a January 20, 2012, Duke Energy RAI response submittal letter (Ref. 2).

Based on a number of follow-up discussions with the Staff, the response is being further refined. Notable changes were made to the Applicability and Required Safety Analyses sections of the TS Bases as well as the inclusion of TS 5.5.22 associated with the PSW Battery Monitoring and Maintenance Program. Enclosure 1 to this submittal contains updated versions of the proposed PSW System and PSW Battery Cell Parameters Technical Specifications and Bases.

Similarly, for the previous responses to RAI item numbers 109(a) and (b), updated versions of the UFSAR markups previously provided to the Staff on March 16, 2012, (Ref. 3) are included. The updated UFSAR pages are provided in Enclosure 2. The most notable change was to identify proposed changes using the 2012 version of the UFSAR currently in effect at the station. The UFSAR markups are intended to highlight the key changes necessary to support installation of the PSW System. Additional changes identified during the ONS design change process will be incorporated into the UFSAR in accordance with 10CFR50.71(e).

If you have any questions in regard to this letter, please contact Stephen C. Newman, Regulatory Affairs Senior Engineer, Oconee Nuclear Station, at (864) 873-4388.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 7, 2013.

Sincerely, Scott L. Batson Vice President Oconee Nuclear Station : RAI Item No. 107 Supplemental Response : RAI Item Nos. 109(a) and (b) Supplemental Responses I

U. S. Nuclear Regulatory Commission August 7, 2013 Page 3 cc: (w/enclosures)

Mr. John P. Boska, Senior Project Manager (by electronic mail only)

U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 11555 Rockville Pike Rockville, MD 20852 Mr. Victor M. McCree, Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. Ed Crowe NRC Senior Resident Inspector Oconee Nuclear Station Ms. Susan E. Jenkins, Manager Radioactive & Infectious Waste Management SC Dept. of Health and Environmental Control 2600 Bull St.

Columbia, SC 29201

Enclosure 1 Supplemental Response to RAI Item No. 107 LAR 2008 Supplement 6

- Supplemental Response to RAI Item No. 107; LAR 2008 Supplement 6 August 7, 2013 Page 2 RAI 107 To ensure licensing-basis clarity and component operability, Technical Specifications (TSs) need to properly address the PSW system in a manner that is consistent with the Standard TS requirements that have been established for the functions that are being performed by similar systems. For example, the minimum required mission time should be 7 days and the completion times should be limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in most cases. The proposed TS for the PSW system allow the system to be out of service for up to 45 days while maintenance is being performed on the system. The proposed TS does not put restrictions on other diverse systems (SSF) that are also used for tornado, HELB and fire mitigation while the PSW system is out of service. The suction source for the PSW system and the SSF are the same (Unit 2 circulation cooling water piping (CCW)). When the Unit 2 CCW piping is dewatered both the PSW system, and the SSF are out of service and cannot perform their intended functions. The proposed PSW TSs does not address this situation. Please address each of the above concerns.

Duke Energy Response This response supplements the prior January 20, 2012, response by providing an updated version of the PSW System and PSW Battery Cell Parameters Technical Specifications and Bases (Attachment 1).

Attachment 1 Revised PSW System and PSW Battery Cell Parameters Technical Specifications and Bases

PSW System 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Protected Service Water (PSW) System LCO 3.7.10 The PSW system shall be OPERABLE Not applicable to Unit(s) until startup from a refueling outage after completion of PSW modifications and after all of the PSW system equipment installed has been tested.

APPLICABILITY: MODES 1 and 2.

ACTIONS

---

NOTE -------------------------------

LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. PSW system is inoperable. A.1 Restore PSW system to 14 days OPERABLE status.

B. PSW system is inoperable. B.1 Restore PSW system to 7 days OPERABLE status.

AND The Standby Shutdown Facility (SSF) is inoperable.

C. ---------- NOTE -----------------

Condition may only be entered when contingency measures have been implemented.

Required Action and C.1 Restore PSW system to 30 days from discovery associated Completion Time of OPERABLE status. of initial inoperability.

Condition A or B not met.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition C not met.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.10-1 Amendment Nos. xxx, xxx, & xxx

PSW System 3.7.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Verify the required PSW battery terminal voltage In accordance with the is greater than or equal to the minimum Surveillance Frequency established float voltage. Control Program.

SR 3.7.10.2 Verify the required Keowee Hydroelectric Station In accordance with the power supply can be aligned to and power the Surveillance Frequency PSW electrical system. Control Program.

SR 3.7.10.3 Verify developed head of PSW primary and In accordance with the booster pumps at flow test point is greater than or Inservice Testing equal to the required developed head. Program.

SR 3.7.10.4 Verify PSW battery capacity of the required In accordance with the battery is adequate to supply, and maintain in Surveillance Frequency OPERABLE status, required emergency loads for Control Program.

the design duty cycle when subjected to a battery service test.

SR 3.7.10.5 Verify the required PSW battery charger supplies In accordance with the a 300 amps at greater than or equal to the Surveillance Frequency minimum established float voltage for > 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Control Program.

OR Verify the required battery charger can recharge the battery to the fully charged state within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding PSW event discharge state.

SR 3.7.10.6 ---------------- NOTE --------------

Both HPI pump motors are individually tested although only one (1) HPI pump motor is required to support PSW system OPERABILITY.

Verify that the required PSW switchgear and In accordance with the transfer switches can be aligned and power both Surveillance Frequency the "A" and "B" HPI pump motors. Control Program.

SR 3.7.10.7 Perform functional test of required power transfer In accordance with the switches used for pressurizer heaters, PSW Surveillance Frequency control, electrical panels, vital I&C chargers, and Control Program.

valves.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.10-2 Amendment Nos. xxx, xxx, & xxx

PSW System 3.7.10 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.10.8 ---------------- NOTE--------------

Cooling water flow to the HPI pump motors are individually tested although only flow to the HPI pump motor aligned to PSW power is required to support PSW system OPERABILITY.

Verify PSW booster pump and valves can provide In accordance with the adequate cooling water flow to HPI pump motor Inservice Testing coolers. Program.

SR 3.7.10.9 Verify developed head of PSW portable pump at In accordance with the the flow test point is greater than or equal to Surveillance Frequency required developed head. Control Program.

SR 3.7.10.10 Verify the required PSW valves are tested in In accordance with the accordance with the Inservice Test Program. Inservice Testing Program.

SR 3.7.10.11 Perform CHANNEL CHECK for each required In accordance with the PSW instrument channel. Surveillance Frequency Control Program.

SR 3.7.10.12 Perform CHANNEL CALIBRATION for each In accordance with the required PSW instrument channel. Surveillance Frequency Control Program.

SR 3.7.10.13 Verify for the required PSW battery that the cells, In accordance with the cell plates and racks show no visual indication of Surveillance Frequency physical damage or abnormal deterioration that Control Program.

could degrade battery performance.

OCONEE UNITS 1, 2, & 3 3.7.10-3 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 B 3.7 PLANT SYSTEMS B 3.7.10 Protected Service Water (PSW) System BASES BACKGROUND The Protected Service Water (PSW) system is designed as a standby system for use under emergency conditions. The PSW system provides added "defense in-depth" protection by serving as a backup to existing safety systems and as such, the system is not required to comply with single failure criteria. The PSW system is provided as an alternate means to achieve and maintain safe shutdown conditions for one, two or three units following postulated scenarios that damage essential systems and components normally used for safe shutdown.

The PSW pumping system utilizes the inventory of lake water contained in the Unit 2 Condenser Circulating Water (CCW) piping. The PSW primary and booster pumps are located in the Auxiliary Building (AB) at elevation 771' and take suction from the Unit 2 CCW piping and discharge into the steam generators of each unit via the Emergency Feedwater (EFW) system headers. The raw water is vaporized in the steam generators (SGs), removing residual heat, and is dumped to atmosphere via the Main Steam Relief Valves (MSRVs) or Atmospheric Dump Valves (ADVs). For extended operation, the PSW portable pump with a flow path capable of taking suction from the intake canal and discharging into the Unit 2 CCW piping is designed to provide a backup supply of water to the PSW system in the event of loss of CCW and subsequent loss of CCW siphon flow. The PSW portable pump is stored onsite.

The PSW system is designed to support cool down of the Reactor Coolant System (RCS) and maintain safe shutdown conditions. The PSW system is designed to maintain SG water levels to promote natural circulation Decay Heat Removal (DHR) using the SGs for an extended period of time during which time other plant systems required to cool the RCS to MODE 5 conditions will be restored and brought into service. In addition, the PSW system, in combination with the High Pressure Injection (HPI) system, provides borated water for Reactor Coolant Pump (RCP) seal cooling, RCS makeup, and reactivity management.

The PSW system reduces fire risk by providing a diverse power supply to power safe shutdown equipment in accordance with the National Fire Protection Association (NFPA) 805 safe shutdown analyses (Ref. 4).

OCONEE UNITS 1, 2, & 3 B 3.7.10-1 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES BACKGROUND The PSW system consists of the following:

(continued)

1. PSW building and associated support systems.

2. Conduit duct bank from the Keowee Hydroelectric Station underground cable trench to the PSW building.

3. Conduit duct bank and raceway from the PSW Building to the Unit 3 AB.

4. Electrical power distribution system from breakers at the Keowee Hydroelectric Station and from the 100 kV PSW substation (supplied from the Central Tie Switchyard) to the PSW building, and from there to the AB.

5. PSW booster pump, PSW primary pump, and mechanical piping taking suction from the Unit 2 embedded CCW System to the EFW headers supplying cooling water to the respective unit's SGs and HPI pump motor bearing coolers.

6. PSW portable pumping system.

The mechanical portion of the PSW system provides decay heat removal by feeding Lake Keowee water to the secondary side of the SGs. In addition, the PSW pumping system supplies Keowee Lake water to the HPI pump motor coolers.

The PSW pumping system consists of a booster pump, a primary pump, and a portable pump. Other than the portable pump, the pumps and required valves are periodically tested in accordance with the In-Service Testing (IST) Program.

The PSW piping system has pump minimum flow lines that discharge back into the Unit 2 CCW embedded piping.

The PSW primary and booster pumps, motor operated valves, and solenoid valves required to bring the system into service, are controlled from the main control rooms. Check valves and manual handwheel operated valves are used to prevent back-flow, accommodate testing, or are used for system isolation.

The PSW electrical system is designed to provide power to PSW mechanical and electrical components as well as other system components needed to establish and maintain a safe shutdown condition.

Normal power is provided by a transformer connected to a 100 kV overhead transmission line that receives power from the Central Tie Switchyard located approximately eight (8) miles from the plant. Standby OCONEE UNITS 1, 2, & 3 B 3.7.10-2 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES BACKGROUND power is provided from the Keowee Hydroelectric Station via an (continued) underground path. These external power sources provide power to transformers, switchgear, breakers, load centers, batteries, and battery chargers located in the PSW electrical equipment structure. There are two (2) batteries inside the PSW Building. Either battery is sized to supply PSW DC loads. The battery banks are located in different rooms separated by fire rated walls. A separate room within the PSW building is provided for major PSW electrical equipment.

PSW building heating, ventilation, and air conditioning (HVAC) is designed to maintain transformer and battery rooms within their design temperature range. The HVAC System consists of two (2) systems; a non QA-1/non credited system designed to maintain the PSW Transformer and Battery Rooms environmental profile and a QA-1/credited system designed to actuate whenever the non QA-1 system is not able to meet its design function.

The hydrogen removal fans are designed to maintain the hydrogen in the Battery rooms below 2% in accordance with IEEE-484 (Ref. 1). The multiple thermostats in each Battery Room ensure temperatures are maintained within acceptable limits.

APPLICABLE The function of the PSW system is to provide a diverse means to achieve SAFETY ANALYSES and maintain safe shutdown by providing secondary side DHR, RCP seal cooling, RCS primary inventory control, and RCS boration for reactivity management following scenarios that disable the 4160 V essential electrical power distribution system.

To verify PSW system performance criteria, thermal-hydraulic (T/H) analysis was performed to demonstrate that the PSW system can achieve and maintain safe shutdown following postulated fires that disable the 4160 V essential power distribution system, without reliance on equipment located in the turbine building. The analysis evaluates RCS subcooling margin using inputs that are representative of plant conditions as defined by Oconee's NFPA 805 fire protection program. 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 and 4160 VAC power 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 PSW system is capable of meeting the relevant NFPA 805 nuclear safety performance criteria.

OCONEE UNITS 1, 2, & 3 B 3.7.10-3 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES APPLICABLE During periods of very low decay heat the PSW system will be used to SAFETY ANALYSES establish conditions that support the formation of subcooled natural (continued) circulation between the core and the SGs; however, natural circulation 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 minimum HPI flow through the Reactor Coolant (RC) vent valves. 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 and 2 other than operation at nominal full power, the duration of operation in these conditions is insufficient to result in an appreciable contribution to overall plant risk. As a result, T/H analysis was performed assuming full power initial conditions, as described above and in the Oconee Fire Protection Program, Nuclear Safety Capability Assessment. The plant configuration examined in the T/H analysis is representative of risk significant operating conditions and provides reasonable assurance that a fire mitigated by PSW during these MODES will not prevent the plant from achieving and maintaining fuel in a safe and stable condition.

The PSW system is not an Engineered Safety Feature Actuation System (ESFAS) and is not credited to mitigate design basis events as described in UFSAR Chapters 6 and 15. No credit is taken in the safety analyses for PSW system operation following design basis events. Based on its contribution to the reduction of overall plant risk, the PSW system satisfies Criterion 4 of 10 CFR 50.36 (c)(2)(ii) (Ref. 3) and is therefore included in the Technical Specifications.

LCO The OPERABILITY of the PSW system provides a diverse means to achieve and maintain safe shutdown by providing secondary side DHR, reactor coolant pump seal cooling, primary system inventory control, and RCS boration for reactivity management during certain plant scenarios that disable the 4160 V essential electrical power distribution system.

For OPERABILITY, the following are required:

" One (1) primary pump, one (1) booster pump, and one (1) portable pump.

• A flowpath taking suction from the Unit 2 CCW piping through the PSW pumping system (including recirculation flowpath) and discharging into the secondary side of each SG and the required HPI pump motor bearing cooler.

OCONEE UNITS 1, 2, & 3 B 3.7.10-4 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES LCO (continued)

" TS 3.8.3 required number of 125 VDC Vital I&C Battery Chargers.

Note: The Standby battery chargers cannot be credited for PSW OPERABILITY because they are not supplied with PSW power.

• One (1) of two (2) PSW batteries and the associated battery charger.

• PSW building ventilation system (QA-1) consisting of ductwork, fans, heaters, fire dampers, tornado dampers, motor-operated dampers and associated controls of the Transformer room AND in-service battery room.

" A PSW electrical system power path from the Keowee Hydroelectric Station.

For OPERABILITY, PSW supplied power is required for the following:

• Either the "A" or "B"HPI pump motor.

" PSW portable pump (unless self-powered).

• HPI valve needed to align the HPI pumps to the Borated Water Storage Tanks (HP-24).

• HPI valves that support RCP seal injection and RCS makeup (HP-26, HP-139, and HP-140).

• Pressurizer Heaters (150 kW above pressurizer ambient heat loss).

" Reactor Vessel Head Vent Valves (RC-159 and RC-160).

• One (1) RCS Loop High Point Vent Pathway (RC-155 and RC-156 or RC-157 and RC-1 58).

• Required 125 VDC Vital I&C Normal Battery Chargers.

For OPERABILITY, the following instrumentation and controls located in each main control room are required:

" Two (2) high flow controllers (PSW-22 and PSW-24).

• Two (2) low flow controllers (PSW-23 and PSW-25).

• Two (2) flow indicators (one per SG).

• One (1) SG header isolation valve (PSW-6).

" One (1) HPI seal injection flow indicator.

" One (1) "A" HPI train flow indication (from ICCM plasma).

The LCO is modified by a Note indicating that it is not applicable to Unit(s) until startup from a refueling outage after completion of PSW modifications and after all of the PSW system equipment installed has been tested. Certain SRs require the unit to be shutdown to perform the SR.

OCONEE UNITS 1, 2, & 3 B 3.7.10-5 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES (continued)

APPLICABILITY In MODES 1 and 2, the PSW system provides a diverse means to achieve and maintain safe shutdown by providing secondary side DHR, reactor coolant pump seal cooling, primary system inventory control, and RCS boration for reactivity management during certain plant scenarios that disable the 4160 V essential electrical power distribution system.

As a result of the system's contribution to overall plant risk in mitigating transients initiated during these operating conditions, PSW is required to be OPERABLE in MODES 1 and 2. In MODES 3 and 4, the PSW system can provide a diverse means for secondary side DHR (while the steam generators remain available), reactor coolant pump seal cooling, primary system inventory control, and RCS boration for reactivity management.

Because of the relatively short periods of operation in these MODES, the contribution to the reduction of overall plant risk in mitigating transients initiated during these operating conditions is not sufficient to warrant inclusion of OPERABILITY requirements for MODES 3 and 4 in the Technical Specifications.

In MODES 5 and 6, the steam generators are not available for secondary side DHR. As such, the PSW feed to the SGs is not required. Protected Service Water system backup power to some of the HPI components may be relied upon for shutdown risk defense-in-depth associated with primary system makeup. There are multiple means to achieve primary system makeup during these conditions. As a result, the contribution to the reduction of overall plant risk during these operating conditions is not sufficient to warrant inclusion of OPERABILITY requirements for MODES 5 and 6 in the Technical Specifications.

ACTIONS The exception for LCO 3.0.4 provided in the NOTE of the Actions, permits entry into MODES 1 or 2 with the PSW system not OPERABLE. This is acceptable because the PSW is not required to support normal operation of the facility or to mitigate a design basis event.

A.1 With the PSW system inoperable, action must be taken to restore the system to OPERABLE status within 14 days. The 14-day Completion Time (CT) is reasonable based on the Standby Shutdown Facility (SSF) Auxiliary Service Water (ASW) and reactor coolant makeup (RCMU) systems being OPERABLE and a low probability of scenarios occurring that would require the PSW system during the 14 day period.

B..1 With both the PSW and SSF systems inoperable, action must be taken to restore the PSW system to OPERABLE status within 7 days. The 7 day OCONEE UNITS 1, 2, & 3 B 3.7.10-6 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES ACTIONS B.1 (continued)

CT is based on the diverse heat removal capabilities afforded by other systems, reasonable times for repairs, and the low probability of scenarios occurring that would require the PSW system during this period.

C..1 Ifthe Required Action and associated CT of Condition A or B is not met, action must be taken to restore the PSW system to OPERABLE status within 30 days. Operation for up to 30 days is permitted if risk-reducing contingency measures are taken. The 30 days is from the time of discovery of initial inoperability.

The condition is modified by a note indicating that contingency measures are required to be in place prior to entry. The contingency measures provide additional assurance that key equipment is available. For example, the Keowee Hydroelectric Units (KHUs), Emergency Feedwater (EFW) pumps, High Pressure Injection (HPI) pumps, Elevated Water Storage Tank (EWST), and 230 kV switchyard, are key equipment which impact overall risk during the extended outage period. Unavailability of the specific equipment does not preclude entry into the condition nor does it require any action by this TS. Rather the appropriate actions for the specific equipment are specified in the applicable TS or Selected Licensee Commitments (SLC). For example, ifthe 1A HPI pump becomes inoperable before entry or becomes inoperable after entry, only TS LCO 3.5.2 (HPI), Condition A shall be entered for Unit 1 and the appropriate actions taken until the pump is restored. This does not preclude entry into LCO 3.7.10 Condition C.

The strategy for the contingency measures is to defer non-essential surveillances or other maintenance activities where human error could increase the likelihood of a loss of offsite power (LOOP) or remove key equipment that is important to overall plant risk. This does not preclude surveillances required by technical specifications or corrective maintenance to equipment that is important to overall plant risk. Technical specification required surveillances and corrective maintenance are examples of essential activities.

The following contingency measures are applied to available key equipment to reduce plant risk:

" No non-essential surveillances or other maintenance activities, or testing, will be conducted in the 230 kV switchyard.

" No non-essential surveillances or other maintenance activities, or testing will be conducted on the Keowee Hydro Units' emergency power system and associated power paths.

OCONEE UNITS 1, 2, & 3 B 3.7.10-7 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES ACTIONS C.1 (continued)

" No non-essential surveillances or other maintenance activities, or testing, will be conducted on each unit's EFW motor-driven and turbine-driven pumps and associated equipment including the EFW cross connects.

• No non-essential surveillances or other maintenance activities, or testing, will be conducted on the unit's HPI pumps and associated equipment.

" No non-essential surveillances or other maintenance activities, or testing, will be conducted on the EWST.

D.1 If the Required Action and associated CTs of Condition A, B, or C are not met, the unit(s) must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed CT is appropriate to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems, considering a three unit shutdown may be required.

SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltage assumed in the battery sizing calculations. The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.2 SR verifies availability of the Keowee Hydroelectric Station power path to the PSW electrical system. Power path verification is included to demonstrate breaker OPERABILITY from the Keowee Hydroelectric Station to the PSW electrical system. To verify KHU-1 can supply the PSW electrical system, Breaker KPF-9 is closed. To verify KHU-2 can supply the PSW electrical system, Breaker KPF-10 is closed. Breakers KPF-9 and KPF-10 are electrically interlocked such that breakers cannot be closed simultaneously.

OCONEE UNITS 1, 2, & 3 B 3.7. 10-8 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES SURVEILLANCE SR 3.7.10.2 (continued)

REQUIREMENTS Electrical interlocks prevent compromise of existing redundant emergency power paths. To verify either KHU can supply the PSW electrical system, the PSW Feeder Breaker [B6T-A] or [B7T-C and the PSW switchgear tie breaker] is closed. The Surveillance Frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.3 This SR requires the PSW primary and booster pumps be tested in accordance with the Inservice Test (IST) Program. The IST program verifies the developed head of PSW primary and booster pumps at flow test point is greater than or equal to the required developed head. The specified Frequency is in accordance with IST Program requirements.

SR 3.7.10.4 A battery service test is a special test of the battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length correspond to the design duty cycle requirements.

The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.5 This SR verifies the design capacity of the battery charger. According to Regulatory Guide 1.32 (Ref. 2), the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.

The minimum required amps and duration ensure that these requirements can be satisfied.

This SR provides two options. One option requires that each battery charger be capable of supplying >300 amps for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at the minimum established float voltage. The current requirements are based on the output rating of the charger. The voltage requirements are based on the charger voltage level after a response to a loss of AC power. The time period is sufficient for the charger temperature to stabilize and to have been maintained for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

OCONEE UNITS 1, 2, & 3 B 3.7. 10-9 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES SURVEILLANCE SR 3.7.10.5 (continued)

REQUIREMENTS The other option requires that the battery charger be capable of recharging the battery after a service test coincident with supplying the largest coincident demands of the various continuous steady state loads (irrespective of the status of the plant during which these demands occur). This level of loading may not normally be available following the battery service test and will need to be supplemented with additional loads. The duration for this test may be longer than the charger sizing criteria since the battery recharge is affected by float voltage, temperature, and the exponential decay in charging current.

The battery is recharged when the measured charging current is < 2 amps.

The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.6 This SR verifies that the PSW switchgear can be aligned and power both the "A" and "B"HPI pump motors (not simultaneously). Although both pump motors are tested, only one (1) is required to support PSW system OPERABILITY. The surveillance frequency is in accordance with the Surveillance Frequency Control Program. Refer to the SR 3.7.10.7 table below for testing of the HPI power and transfer switches.

SR 3.7.10.7 This SR verifies that power transfer switches (shown in table below) for pressurizer heaters, PSW control, electrical panels, and valves, are functional for the required equipment.

Component 1HPI-SX-ALGN001 (PSW HPI alignment switch) 2HPI-SX-ALGN001 (PSW HPI alignment switch) 3HPI-SX-ALGN001 (PSW HPI alignment switch)

I HPI-SX-TRN001 (IA HPI pump transfer switch)

I HPI-SX-TRN002 (l B HPI pump transfer switch) 2HPI-SX-TRN001 (2A HPI pump transfer switch) 2HPI-SX-TRN002 (2B HPI pump transfer switch) 3HPI-SX-TRN001 (3A HPI pump transfer switch) 3HPI-SX-TRN002 (3B HPI pump transfer switch)

I HPI-SX-TRN003 (1 HP-24 PSW transfer switch) 1HPI-SX-TRN004 (1 HP-26 PSW transfer switch)

OCONEE UNITS 1, 2, &3 B 3.7. 10-10 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES SURVEILLANCE SR 3.7.10.7 (continued)

REQUIREMENTS Component 2HPI-SX-TRN003 (2HP-24 PSW transfer switch) 2HPI-SX-TRN004 (2HP-26 PSW transfer switch) 3HPI-SX-TRN003 (3HP-24 PSW transfer switch) 3HPI-SX-TRN004 (3HP-26 PSW transfer switch)

HPSW-SX-TRN001 (ICA CHARGER auto transfer switch) 1 PSW-SX-TRN002 (1 CB CHARGER auto transfer switch) 2PSW-SX-TRN001 (2CA CHARGER auto transfer switch) 2PSW-SX-TRN002 (2CB CHARGER auto transfer switch) 3PSW-SX-TRN001 (3CA CHARGER auto transfer switch) 3PSW-SX-TRN002 (3CB CHARGER auto transfer switch) 1 PSW-SX-TRN004 (manual transfer switch for 1XJ) 1 PSW-SX-TRN005 (manual transfer switch for 1XK) 2PSW-SX-TRN003 (manual transfer switch for 2XJ) 2PSW-SX-TRN004 (manual transfer switch for 2XJ) 2PSW-SX-TRN005 (manual transfer switch for 2XK) 3PSW-SX-TRN003 (manual transfer switch for 3XJ) 3PSW-SX-TRN004 (manual transfer switch for 3XJ) 3PSW-SX-TRN005 (manual transfer switch for 3XK) 1RC-1 55/1 RC-1 56 power transfer 1 RC-1 57/1 RC-1 58 power transfer 1RC-1 59/1 RC-1 60 power transfer 2RC-1 55/1 RC-1 56 power transfer 2RC-1 57/1 RC-1 58 power transfer 2RC-1 59/1 RC-1 60 power transfer 3RC-1 55/1 RC-1 56 power transfer 3RC-1 57/1 RC-1 58 power transfer 3RC-1 59/1 RC-1 60 power transfer The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.8 SR verifies PSW booster pump and valves can supply water to the "A" and "B" HPI pump motor coolers in accordance with the IST program.

OCONEE UNITS 1, 2, &3 B 3.7. 10-11 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES SURVEILLANCE SR 3.7.10.9 REQUIREMENTS (continued) This SR requires that the PSW portable pump be tested to verify that the developed head of PSW portable pump at the flow test point is greater than or equal to the required developed head. The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.10 This SR requires the required PSW valves be tested in accordance with the IST Program. The specified Frequency is in accordance with IST Program requirements.

SR 3.7.10.11 Performance of the CHANNEL CHECK for each required instrumentation channel ensures that a gross failure of instrumentation has not occurred.

A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel with a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The instrument string to the control room is checked and calibrated periodically per the Surveillance Frequency Control Program.

Agreement criteria are determined based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. Ifthe channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Off scale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled in accordance with the Surveillance Frequency Control Program.

OCONEE UNITS 1, 2, & 3 B 3.7.10-12 Amendment Nos. xxx, xxx, & xxx

PSW System B 3.7.10 BASES SURVEILLANCE SR 3.7.10.12 REQUIREMENTS (continued) CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift to ensure that the instrument channel remains operational between successive tests. CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are within the assumptions of the setpoint analysis. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint analysis.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled in accordance with the Surveillance Frequency Control Program.

SR 3.7.10.13 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.

The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function).

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled in accordance with the Surveillance Frequency Control Program.

REFERENCES 1. IEEE-484-2002.

2. Regulatory Guide 1.32, February 1977.

3. 10 CFR 50.36 (last amended September 24, 2008).

4. NFPA 805 Safety Evaluation Report, dated December 29, 2010.

OCONEE UNITS 1, 2, & 3 B 3.7.10-13 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters 3.7.1Oa 3.7 PLANT SYSTEMS 3.7.10a Protected Service Water (PSW) Battery Cell Parameters LCO 3.7.1 Oa Battery Cell parameters for the required PSW battery shall be within limits.

APPLICABILITY: When the PSW system is required to be OPERABLE.

ACTIONS

---

I V IJIt -----------------------------------------------------------

LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Required battery with A.1 Perform SR 3.7.10.1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> one or more battery cell float voltages :5 2.07 V. AND A.2 Perform SR 3.7.10a.1. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> AND A.3 Restore affected cell 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> voltage > 2.07 V.

B. Required battery with B.1 Perform SR 3.7.10.1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> float current > 2 amps.

AND B.2 Restore battery float 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> current to < 2 amps.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.10a-1 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters 3.7.1Oa ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

........----- NOTE -- ---------------- NOTE --------------

Required Actions C.1 and Required Actions C.1 and C.2 C.2 shall be completed if are only applicable if electrolyte level was below electrolyte level was below the the top of plates. top of plates.

C. Required battery with C.1 Restore electrolyte 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> one or more cells level to above top of electrolyte level less plates.

than minimum established design AND limits.

C.2 Verify no evidence of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> leakage.

AND C.3 Restore electrolyte level 31 days to greater than or equal to minimum established design limits.

D. Required battery with D.1 Restore battery pilot cell 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pilot cell electrolyte temperature to greater than temperature less than or equal to minimum minimum established established design limits.

design limits.

(continued)

OCONEE UNITS 1, 2, & 3 3.7.10Oa-2 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters 3.7.1Oa ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and E.1 Declare associated battery Immediately associated Completion inoperable.

Time of Condition A, B, C, or D not met.

OR Required battery with one or more battery cells float voltage

< 2.07 V and float current > 2 amps.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10a.1 --------------- NOTE ------------

Not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.7.10.1.

Verify battery float current is < 2 amps. In accordance with the Surveillance Frequency Control Program.

SR 3.7.10a.2 Verify battery pilot cell voltage is >2.07 V. In accordance with the Surveillance Frequency Control Program.

SR 3.7.10a.3 Verify battery connected cell electrolyte In accordance with the level is greater than or equal to minimum Surveillance Frequency established design limits. Control Program.

SR 3.7.1Oa.4 Verify battery pilot cell temperature is greater In accordance with the than or equal to minimum established design Surveillance Frequency limits. Control Program.

(continued)

OCONEE UNITS 1, 2, &3 3.7. 10a-3 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters 3.7.1Oa SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.1Oa.5 Verify battery connected cell voltage is In accordance with the

> 2.07 V. Surveillance Frequency Control Program.

SR 3.7.10a.6 Verify battery capacity is > 80% of the In accordance with the manufacturer's rating when subjected to a Surveillance Frequency performance discharge test or a modified Control Program.

performance discharge test.

AND 12 months when battery shows degradation or has reached 85% of the expected life with capacity < 100% of manufacturer's rating.

AND 24 months when battery has reached 85% of the expected life with capacity a 100% of manufacturer's rating.

OCONEE UNITS 1, 2, & 3 3.7. 10a-4 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters B 3.7.10a B 3.7 PLANT SYSTEMS B 3.7.10a PSW Battery Cell Parameters BASES BACKGROUND This LCO delineates the limits on battery float current as well as electrolyte temperature, level, and float voltage for the Protected Service Water (PSW) Power system batteries. In addition to the limitations of this Specification, the PSW Battery Monitoring and Maintenance Program specified in Specification 5.5.22 for monitoring various battery parameters is based on the recommendations of IEEE-450 (Ref. 1).

Each PSW battery consists of 60 cells (nominal) and either battery can meet the PSW DC System design basis duty cycle with up to two (2) cells jumpered out. A minimum of 58 of 60 cells are required for a battery to be considered OPERABLE.

The battery cells are of flooded lead acid construction with a nominal specific gravity of 1.215. This specific gravity corresponds to an open circuit battery voltage of approximately 124 V for 60 cell battery, i.e., cell voltage of 2.07 Volts per cell (Vpc). The open circuit voltage is the voltage maintained when there is no charging or discharging. Once fully charged with its open circuit voltage < 2.07 Vpc, the battery cell will maintain its capacity for 30 days without further charging per manufacturer's instructions. Optimal long term performance however, is obtained by maintaining a float voltage 2.20 to 2.25 Vpc. This provides adequate over-potential which limits the formation of lead sulfate and self discharge. The nominal float voltage of 2.22 Vpc corresponds to a total float voltage output of 133.2 V for a 60 cell battery.

The PSW DC system consists of two (2) batteries, two (2) battery chargers, a distribution center and panelboards. Either battery can be aligned to either battery charger. For PSW DC System OPERABILITY, only one (1) battery and one (1) battery charger is required to be aligned to the PSW DC Bus.

APPLICABLE The PSW system is not credited to mitigate design basis events. No SAFETY ANALYSES credit is taken in the safety analyses for PSW system operation following design basis events. Based on its contribution to the reduction of overall plant risk, the PSW system satisfies Criterion 4 of 10 CFR 50.36 (c)(2)(ii)

(Ref. 3) and is therefore included in the Technical Specifications. Refer to the Applicable Safety Analysis discussion in the Bases for LCO 3.7.10.

OCONEE UNITS 1, 2, & 3 B 3.7.10a-1 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters B 3.7.1Oa BASES LCO For PSW DC System OPERABILITY, only one (1) battery and one (1) battery charger is required to be aligned to the PSW DC Bus. A minimum of 58 of 60 cells are required for a battery to be considered OPERABLE.

PSW Battery parameters must remain within acceptable limits to ensure availability of the PSW DC power system after an occurrence that disables essential systems and components needed for safe shutdown.

Battery parameter limits are conservatively established, allowing continued PSW DC electrical system function even with limits not met. Additional preventative maintenance, testing, and monitoring for the PSW batteries are performed in accordance with the PSW Battery Monitoring and Maintenance Program specified in Specification 5.5.22.

APPLICABILITY The battery parameters are required solely for the support of the associated PSW electrical power systems; therefore, battery parameter limits are only required when the PSW DC power source is required to be OPERABLE. Refer to the Applicability discussion in the Bases for LCO 3.7.10.

ACTIONS The exception for LCO 3.0.4 provided in the NOTE of the Actions, permits entry into MODES 1 or 2 with the PSW system not OPERABLE. This is acceptable because the PSW is not required to support normal operation of the facility or to mitigate a design basis event.

A.1, A.2. and A.3 With one or more cells in the required battery < 2.07 V, the battery cell is degraded. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> verification of the required battery charger OPERABILITY is made by monitoring the battery terminal voltage (SR 3.7.10.1) and the overall battery state of charge by monitoring the battery float charge current (SR 3.7.1Oa. 1). This assures that there is still sufficient battery capacity to perform the intended function. Therefore, the affected battery is not required to be considered inoperable solely as a result of one or more cells in a battery : 2.07 V, and continued operation is permitted for a limited period up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Since the Required Actions only specify "perform," a failure of SR 3.7.10.1 or SR 3.7.1Oa.1 acceptance criteria does not result in this Required Action not met. However, if one of the SRs is failed, the OCONEE UNITS 1, 2, & 3 B 3.7. 10a-2 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters B 3.7.1Oa BASES ACTIONS A.1. A.2 and A.3 (continued) appropriate Condition(s), depending on the cause of the failures, is entered. If SR 3.7.1Oa.1 is failed then there is no assurance that there is still sufficient battery capacity to perform the intended function and the battery must be declared inoperable immediately.

B.1 and B.2 A required battery with float current >2 amps indicates that a partial discharge of the battery capacity has occurred. This may be due to a temporary loss of a battery charger or possibly due to one or more battery cells in a low voltage condition reflecting some loss of capacity. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> verification of the required battery charger OPERABILITY is made by monitoring the battery terminal voltage (SR 3.7.10.1). If the terminal voltage is found to be less than the minimum established float voltage, there are two possibilities: (1) the battery charger is inoperable or (2) it is operating in the current limit mode. Condition A addresses charger inoperability. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, if the charger is operating in the current limit mode, it is an indication that the battery has been substantially discharged and likely cannot perform its required design functions. The time to return the battery to its fully charged condition in this case is a function of the battery charger capacity, the amount of loads on the associated DC system, the amount of the previous discharge, and the recharge characteristic of the battery. The charge time can be extensive, and there is not adequate assurance that it can be recharged within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Required Action B.2). The battery must therefore be declared inoperable.

If the float voltage is found to be satisfactory but there are one or more battery cells with float voltage less than 2.07 V, the associated "OR" statement in Condition E is applicable and the battery must be declared inoperable immediately. Iffloat voltage is satisfactory and there are no cells less than 2.07 V, there is reasonable assurance that, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the battery will be restored to its fully charged condition (Required Action B.2) from any discharge that might have occurred due to a temporary loss of the battery charger.

A discharged battery with float voltage (the charger setpoint) across its terminals indicates that the battery is on the exponential charging current portion (the second part) of its recharge cycle. The time to return a battery to its fully charged state under this condition is simply a function of the amount of the previous discharge and the recharge characteristic of the battery. Thus there is good assurance of fully recharging the battery within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

OCONEE UNITS 1, 2, & 3 B 3.7. 10a-3 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters B 3.7.1Oa BASES ACTIONS B.1 and B.2 (continued)

If the condition is due to one or more cells in a low voltage condition but still greater than 2.07 V and float voltage is found to be satisfactory, this is not indication of a substantially discharged battery and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a reasonable time prior to declaring the battery inoperable.

Since Required Action B.1 only specifies "perform," a failure of SR 3.7.10.1 acceptance criteria does not result in the Required Action not met.

However, if SR 3.7.10.1 is failed, the appropriate Condition(s), depending on the cause of the failure, is entered.

C.1, C.2. and C.3 With the required battery with one or more cells electrolyte level above the top of the plates, but below the minimum established design limits, the battery still retains sufficient capacity to perform the intended function.

Therefore, the affected battery is not required to be considered inoperable solely as a result of electrolyte level not met. Within 31 days the minimum established design limits for electrolyte level must be re-established.

With electrolyte level below the top of the plates there is a potential for dryout and plate degradation. Required Actions C.1 and C.2 address this potential (as well as provisions in Specification 5.5.22, PSW Battery Monitoring and Maintenance Program). They are modified by a note that indicates they are only applicable if electrolyte level is below the top of the plates. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> level is required to be restored to above the top of the plates. The Required Action C.2 requirement to verify that there is no leakage by visual inspection and the Specification 5.5.22 item to initiate action to equalize and test in accordance with manufacturer's recommendation are taken from Appendix D of IEEE-450 (Ref. 1). They are performed following the restoration of the electrolyte level to above the top of the plates. Based on the results of the manufacturer's recommended testing the battery may have to be declared inoperable and the affected cell[s] replaced.

D.1 With the required battery with pilot cell temperature less than the minimum established design limits, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed to restore the temperature to within limits. A low electrolyte temperature limits the current and power available. Since the battery is sized with margin, while battery capacity is degraded, sufficient capacity exists to perform the intended function and the affected battery is not required to be considered inoperable solely as a result of the pilot cell temperature not met.

OCONEE UNITS 1, 2, & 3 B 3.7. 10a-4 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters B 3.7.1Oa BASES ACTIONS E._1 (continued)

With the required battery having any battery parameter outside the allowances of the Required Actions for Condition A, B, C, or D, sufficient capacity to supply the maximum expected load requirement is not assured and must be declared inoperable.

Additionally, discovering the required battery with one or more battery cells float voltage less than or equal to 2.07 V and float current greater than 2 amps indicates that the battery capacity may not be sufficient to perform the intended functions. The battery must therefore be declared inoperable immediately.

SURVEILLANCE SR 3.7.1Oa.1 REQUIREMENTS Verifying battery float current while on float charge is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state.

The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE-450 (Ref. 1). The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.7.10.1. When this float voltage is not maintained, the Required Actions of LCO 3.7.1Oa ACTION A are being taken, which provide the necessary and appropriate verifications of the battery condition. Furthermore, the float current limit of 2 amps is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.

SR 3.7.10a.2 and SR 3.7.10a.5 Optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer, which corresponds to 2.20 Vpc.

This provides adequate over potential, which limits the formation of lead sulfate and self discharge, which could eventually render the battery inoperable. Float voltages in this range or less, but greater than 2.07 Vpc, are addressed in Specification 5.5.22. SRs 3.7.10a.2 and 3.7.10a.5 require verification that the cell float voltages are greater than the short term absolute minimum voltage of 2.07 V. The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

OCONEE UNITS 1, 2, & 3 B 3.7. 10a-5 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters B 3.7.1Oa BASES SURVEILLANCE SR 3.7.10a.3 REQUIREMENTS (continued) The limit specified for electrolyte level ensures that the plates suffer no physical damage and maintains adequate electron transfer capability. The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.1Oa.4 This Surveillance verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (60 OF). Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements.

Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity. The surveillance frequency is in accordance with the Surveillance Frequency Control Program.

SR 3.7.10a.6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as-found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.

Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.7.1Oa.6; however, only the modified performance discharge test may be used to satisfy the battery service test requirements of SR 3.7.10.4.

A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test.

The modified discharge test may consist of just two rates; for instance the one minute rate for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test.

OCONEE UNITS 1, 2, & 3 B 3.7. 10a-6 Amendment Nos. xxx, xxx, & xxx

PSW Battery Cell Parameters B 3.7.1Oa BASES SURVEILLANCE SR 3.7.10a.6 (continued)

REQUIREMENTS The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 1) and IEEE-485 (Ref. 2). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.

Furthermore, the battery is sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80 percent limit.

The surveillance frequency is in accordance with the Surveillance Frequency Control Program. If the battery shows degradation, or ifthe battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity > 100% of the manufacturer's ratings. Degradation is indicated, according to IEEE-450 (Ref. 1), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is > 10% below the manufacturer's rating. These Frequencies are consistent with the recommendations in IEEE-450 (Ref. 1).

REFERENCES 1. IEEE-450-1995.

2. IEEE-485-1983.

3. 10 CFR 50.36 (last amended September 24, 2008).

OCONEE UNITS 1, 2, & 3 B 3.7. 10a-7 Amendment Nos. xxx, xxx, & xxx

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.22 Protected Service Water System Battery Monitoring and Maintenance Program This program is applicable only to the Protected Service Water Battery cells and provides for battery restoration and maintenance, based on the recommendation of IEEE Standard 450-1995. "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications,"

including the following:

1. Actions to restore battery cells with float voltage < 2.13 V;

2. Actions to determine whether the float voltage of the remaining battery cells is >

2.13 V when the float voltage of a battery cell has been found to be < 2.13 V;

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

OCONEE UNITS 1, 2, & 3 5.0-23 Amendment Nos. xxx, xxx. & xxx

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Deleted 5.6.2 Annual Radiological Environmental Operating Report

---

NOTE-------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.

The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

OCONEE UNITS 1, 2, & 3 5.0-24 Amendment Nos. xxx, xxx, & xxx

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Release Report NOTE----------------------------

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR part 50, Appendix I,Section IV.B.1.

5.6.4 Deleted 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

Core operating limits shall be established, determined and issued in accordance with the following:

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Shutdown Margin limit for Specification 3.1.1;

2. Moderator Temperature Coefficient limit for Specification 3.1.3;

3. Physical Position, Sequence and Overlap limits for Specification 3.2.1 Rod Insertion Limits;

4. AXIAL POWER IMBALANCE operating limits for Specification 3.2.2;

5. QUADRANT POWER TILT (QPT) limits for Specification 3.2.3; OCONEE UNITS 1, 2, & 3 5.0-25 Amendment Nos. xxx, xxx, & xxx

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

6. Nuclear Overpower Flux/Flow/Imbalance and RCS Variable Low Pressure allowable value limits for Specification 3.3.1;

7. RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits for Specification 3.4.1

8. Core Flood Tanks Boron concentration limits for Specification 3.5.1;

9. Borated Water Storage Tank Boron concentration limits for Specification 3.5.4;

10. Spent Fuel Pool Boron concentration limits for Specification 3.7.12;

11. RCS and Transfer Canal boron concentration limits for Specification 3.9.1; and

12. AXIAL POWER IMBALANCE protective limits and RCS Variable Low Pressure protective limits for Specification 2.1.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

(1) DPC-NE-1002-A, Reload Design Methodology II; (2) NFS-1001-A, Reload Design Methodology; (3) DPC-NE-2003-P-A, Oconee Nuclear Station Core Thermal Hydraulic Methodology Using VIPRE-01; (4) DPC-NE-1004-A, Nuclear Design Methodology Using CASMO-3/SIMULATE-3P; (5) DPC-NE-2008-P-A, Fuel Mechanical Reload Analysis Methodology Using TACO3 and GDTACO; (6) BAW-1 0192-P-A, BWNT LOCA - BWNT Loss of Coolant Accident Evaluation Model for Once-Through Steam Generator Plants; OCONEE UNITS 1, 2, & 3 5.0-26 Amendment Nos. xxx, xxx, & xxx

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

(7) DPC-NE-3000-P-A, Thermal Hydraulic Transient Analysis Methodology; (8) DPC-NE-2005-P-A, Thermal Hydraulic Statistical Core Design Methodology; (9) DPC-NE-3005-P-A, UFSAR Chapter 15 Transient Analysis Methodology:

(10) BAW-10227-P-A, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel; (11) BAW-10164P-A, RELAP 5/MOD2-B&W - An Advanced Computer Program for Light Water Reactor LOCA and non-LOCA Transient Analysis; and (12) DPC-NE-1006-P-A, Oconee Nuclear Design Methodology Using CASMO-4/SIMULATE-3 (Revision 0, May 2009).

The COLR will contain the complete identification for each of the Technical Specifications referenced topical reports used to prepare the COLR (i.e.,

report number, title, revision number, report date or NRC SER date, and any supplements).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) and Main Feeder Bus Monitor Panel (MFPMP)

Reo~ort When a report is required by Condition B or G of LCO 3.3.8, "Post Accident Monitoring (PAM) Instrumentation" or Condition D of LCO 3.3.23, "Main Feeder Bus Monitor Panel," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring (PAM only),

the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

OCONEE UNITS 1, 2, & 3 5.0-27 Amendment Nos. xxx, xxx, & xxx

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.8 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.10, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

OCONEE UNITS 1, 2, &3 5.0-28 Amendment Nos. xxx, xxx, &xxx

TABLE OF CONTENTS 3.4.6 RCS Loops - MODE 4 ...................................................................... 3.4.6-1 3.4.7 RCS Loops - MODE 5, Loops Filled ................................................. 3.4.7-1 3.4.8 RCS Loops - MODE 5, Loops Not Filled ........................................... 3.4.8-1 3.4.9 Pressurizer .......................................................................................... 3.4.9-1 3.4.10 Pressurizer Safety Valves ................................................................... 3.4.10-1 3.4.11 RCS Specific Activity .......................................................................... 3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP)

System .......................................................................................... 3.4.12-1 3.4.13 RCS Operational LEAKAGE ............................................................... 3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ..................................... 3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation ............................................ 3.4.15-1 3.4.16 Steam Generator (SG) Tube Integrity ................................................. 3.4.16-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ................................ 3.5.1-1 3.5.1 Core Flood Tanks (CFTs) ................................................................... 3.5.1-1 3.5.2 High Pressure Injection ....................................................................... 3.5.2-1 3.5.3 Low Pressure Injection ........................................................................ 3.5.3-1 3.5.4 Borated W ater Storage Tank (BW ST) ................................................ 3.5.4-1 3.6 CONTAINMENT SYSTEMS ....................................................................... 3.6.1-1 3.6.1 Containment ....................................................................................... 3.6.1-1 3.6.2 Containment Air Locks ........................................................................ 3.6.2-1 3.6.3 Containment Isolation Valves ............................................................. 3.6.3-1 3.6.4 Containment Pressure ........................................................................ 3.6.4-1 3.6.5 Reactor Building Spray and Cooling System ...................................... 3.6.5-1 3.7 PLANT SYSTEMS ...................................................................................... 3.7.1-1 3.7.1 Main Steam Relief Valves (MSRVs) ................................................... 3.7.1-1 3.7.2 Turbine Stop Valves (TSVs) ............................................................... 3.7.2-1 3.7.3 Main Feedwater Control Valves (MFCVs), and Startup Feedwater Control Valves (SFCVs) .............................................. 3.7.3-1 3.7.4 Atmospheric Dum p Valve (ADV) Flow Paths ...................................... 3.7.4-1 3.7.5 Emergency Feedwater (EFW ) System ............................................... 3.7.5-1 3.7.6 Upper Surge Tank (UST) and Hotwell (HW ) ....................................... 3.7.6-1 3.7.7 Low Pressure Service W ater (LPSW ) System .................................... 3.7.7-1 3.7.8 Emergency Condenser Circulating W ater (ECCW) ............................ 3.7.8-1 3.7.9 Control Room Ventilation System (CRVS) Booster Fans ................................................................... 3.7.9-1 3.7.10 Protected Service W ater (PSW) ......................................................... 3.7.10-1 3.7.1 Oa Protected Service Water (PSW) Battery Cell Parameters ................................................................................... 3.7.1Oa-1 3.7.11 Spent Fuel Pool W ater Level .............................................................. 3.7.11-1 3.7.12 Spent Fuel Pool Boron Concentration ................................................. 3.7.12-1 3.7.13 Fuel Assem bly Storage ....................................................................... 3.7.13-1 OCONEE UNITS 1, 2, & 3 iii Amendment Nos. xxx, xxx, & xxx

TABLE OF CONTENTS 5.2 Organization ............................................................................................... 5.0-2 5.3 Station Staff Qualifications .......................................................................... 5.0-5 5.4 Procedures ................................................................................................ 5.0-6 5.5 Programs and Manuals .............................................................................. 5.0-7 5.6 Reporting Requirem ents ............................................................................. 5.0-24 OCONEE UNITS 1, 2, & 3 V Amendment Nos. xxx, xxx, & xxx

TABLE OF CONTENTS B 3.7 PLANT SYSTEMS (continued)

B 3.7.9 Control Room Ventilation System (C RV S) Booster Fans ............................................................. B 3.7.9-1 B 3.7.10 Protected Service Water (PSW) ................................................... B 3.7.10-1 B 3.7.1Oa Protected Service Water (PSW)

Battery Cell Parameters .......................................................... B 3.7.1Oa-1 B 3.7.11 Spent Fuel Pool Water Level ........................................................ B 3.7.11-1 B 3.7.12 Spent Fuel Pool Boron Concentration .......................................... B 3.7.12-1 B 3.7.13 Fuel Assembly Storage ................................................................ B 3.7.13-1 B 3.7.14 Secondary Specific Activity .......................................................... B 3.7.14-1 B 3.7.15 Decay Time for Fuel Assemblies in Spent Fuel P ool (S F P ) ...................................................................... B 3.7.15-1 B 3.7.16 Control Room Area Cooling Systems (CRACS) ........................... B 3.7.16-1 B 3.7.17 Spent Fuel Pool Ventilation System (SFPVS) .............................. B 3.7.17-1 B 3.7.18 Dry Spent Fuel Storage Cask Loading and Unloading ................. B 3.7.18-1 B 3.8 ELECTRICAL POWER SYSTEMS ...................................................... B 3.8.1-1 B 3.8.1 AC Sources - Operating ............................................................... B 3.8.1-1 B 3.8.2 AC Sources - Shutdown .............................................................. B 3.8.2-1 B 3.8.3 DC Sources - Operating .............................................................. B 3.8.3-1 B 3.8.4 DC Sources - Shutdown .............................................................. B 3.8.4-1 B 3.8.5 Battery Cell Parameters ............................................................... B 3.8.5-1 B 3.8.6 Vital Inverters - Operating ............................................................ B 3.8.6-1 B 3.8.7 Vital Inverters - Shutdown ............................................................ B 3.8.7-1 B 3.8.8 Distribution Systems - Operating ................................................. B 3.8.8-1 B 3.8.9 Distribution Systems - Shutdown ................................................. B 3.8.9-1 B 3.9 REFUELING OPERATIONS ................................................................ B 3.9.1-1 B 3.9.1 Boron Concentration ..................................................................... B 3.9.1-1 B 3.9.2 Nuclear Instrum entation ............................................................... B 3.9.2-1 B 3.9.3 Containment Penetrations ............................................................ B 3.9.3-1 B 3.9.4 Decay Heat Removal (DHR) and Coolant Circulation - High Water Level ............................................... B 3.9.4-1 B 3.9.5 Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level ................................................ B 3.9.5-1 B 3.9.6 Fuel Transfer Canal Water Level .................................................. B 3.9.6-1 B 3.9.7 Unborated Water Source Isolation Valves .................................... B 3.9.7-1 B 3.10 STANDBY SHUTDOWN FACILITY ..................................................... B 3.10.1-1 B 3.10.1 Standby Shutdown Facility (SSF) ................................................. B 3.10.1-1 B 3.10.2 Standby Shutdown Facility (SSF) Battery C ell P aram eters ...................................................................... B 3.10.2-1 OCONEE UNITS 1, 2, & 3 iv Amendment Nos. xxx, xxx, & xxx

Enclosure 2 Supplemental Response to RAI Item Nos. 109(a) and (b)

LAR 2008 Supplement 6

- Supplemental Response to RAI Item Nos. 109(a) and (b); LAR 2008 Supplement 6 August 7, 2013 Page 2

[Hi RAI 109 (a)

Provide the following information associated with the protected service water (PSW) system modification:

a. Provide the following UFSAR pages for the PSW system modification including piping and instrumentation drawing showing all pumps, valves, pump design data, and the pump performance curves.

Duke Energy Response for RAI 109 (a)

This response supplements the previous March 16, 2012, response by providing an updated mark-up of the affected UFSAR pages using the 2012 version of the UFSAR currently in effect at the station (Attachment 2) and PSW System Figures (Attachment 3).

rHI RAI 109 (b)

Provide the following information associated with the protected service water (PSW) system modification:

b. Provide the proposed changes to UFSAR Section 3.2.2 for PSW system and the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) classification for PSW system pumps and valves.

Duke Energy Response for RAI 109 (b)

This response supplements the prior March 16, 2012, response by providing an updated mark-up of the affected UFSAR pages using the 2012 version of the UFSAR currently in effect at the station (Attachment 4).

Attachment 2 UFSAR Markups in Response to RAI Item No. 109(a)

UFSAR Chapter 3 Oconee Nuclear Station 3.1.4 Criterion 4 - Sharing of Systems (Category A)

Reactor facilities shall not share systems or components unless it is shown safety is not impaired by the sharing.

Discussion Portions of the following systems are shared as indicated. Where sharing between Oconee 1 and 2 is indicated, a separate system is provided for Oconee 3. Safety is not impaired by the sharing.

System Shared by Units Reference Chemical Addition and Sampling 1,2 9.3.2 Spent Fuel Cooling 1,2 9.1.3 Liquid Waste Disposal 1,2,3 11.2.2 Gaseous Waste Disposal 1,2 11.3.2 Solid Waste Disposal 1,2, 3 11.4.1.2 Coolant Treatment 1, 2, 3 9.3.5 Recirculated Cooling Water 1, 2, 3 9.2.2.2.4 Low Pressure Service Water 1, 2 9.2.2.2.3 High Pressure Service Water 1, 2, 3 9.2.2.2.2 Control Room Ventilation 1,2 9.4.1 Auxiliary Building Ventilation 1,2 9.4.3 Turbine Building Ventilation 1,2,3 9.4.4 Area Radiation Monitoring 1, 2 12.3.3 Process Radiation Monitoring 3.1.5 Criterion 5 - Records Requirements (Category A)

Records of the design, fabrication, and construction of essential componeZ of the plant shall be maintained by the reactor operator or under his control throughout the life of e reactor.

Discussion rE1lace: "9.2.3.1" with "9.7" Duke Power Company will have under its control or will have access to all records of major essential components for the life of the plant. Records maintained by Duke Power Company will include:

3.1 -8 (31 DEC 2012)

UFSAR Chapter 3 Oconee Nuclear Station The Keowee Structures considered are Powerhouse, Power and Penstock Tunnels, Spillway, Service Bay Substructure, Breaker Vault, and Intake Structure.

3. Dams and Dikes The Keowee Dam, the Little River Dam and Dikes, and the Oconee Intake Canal Dike impound the waters of Lake Keowee to provide the source of flowing water for the Keowee hydroelectric power plant.

4. Oconee Intake Structure The intake structure supports the CCW pumps, intake screens, and inlets of the CCW pipes.

5. Oconee Intake Underwater Weir The underwater weir retains an emergency water supply in the event that the waters of Lake Keowee are released by the failure of a dam or dike.

IAdd: "SSF"

6. CCW Intake Piping The CCW Intake Piping conveys water from the CCW pumps on the in structure to the condenser, supplies water to the LPSW Pumps, and serves as the reservior for the Auxiliary Service Water System. Ad: "and the Protected Service Water System."

7. CCW Discharge Piping The CCW Discharge Piping conveys water from tlto the discharge structure and supplements the CCW intake piping as a reservior for tl* Auxiliary *ervice Water System.

8. ECCW Piping The ECCW Piping serves two different functions. 1) It can'sphon the Condenser Circulating Water through the Condenser to be discharged at the treatment pond.) It can be used for recirculation of the Condenser Circulating Water back to the Intake Canal. RPl,, Relae:"uxiliary" Auxiiary"w.th"Protcted with "Protected I

9. Essential Siphon Vacuum System Intake Dike Trench _

The Essential Siphon Vaccum (ESV) System Intake Dike Trench is constructed of reinforced concrete (bottom and walls). The covers for the trench are steel plate except at the roadway crossing.

The covers at the roadway are removable reinforced concrete slabs.

The Essential Siphon Vaccum (ESV) System Intake Dike Trench routes the ESV piping, the Siphon Seal Water (SSW) piping, electrical heat trace cables, and electrical instrumentation cables within the FERC boundary without reducing the integrity of the Oconee Intake Dike.

10. Essential Siphon Vacuum System Building The ESV Building is constructed of a reinforced concrete mat foundation and rigid structural steel frame with metal siding.

The ESV Building encloses the ESV System's pumps, motors and associated equipment, providing protection (from weather & freezing) for that equipment and providing a suitable environment for maintenance activities.

11. Essential Siphon Vacuum System Cable Trench The essential Siphon Vacuum (ESV) System Intake Cable Trench is constructed of reinforced concrete (bottom and walls). The covers for the trench are steel plate except at the traffic crossing.

The covers at the crossing are removable reinforced concrete slabs.

The ESV System Cable Trench routes the cables associated with the ESV System and SSW System from the Radwaste Trench to the ESV Building.

3.8 - 62 (31 DEC 2012)

Oconee Nuclear Station UFSAR Chapter 9 9.2.1.2 System Description and Evaluation 9.2.1.3 Mode of Operation 9.2.1.4 Reliability Considerations 9.2.1.5 Codes and Standards 9.2.1.6 System Isolation 9.2.1.7 Leakage Considerations 9.2.1.8 Failure Considerations 9.2.2 Cooling Water Systems 9.2.2.1 Design Bases 9.2.2.2 System Description and Evaluation 9.2.2.2.1 Condenser Circulating Water System (CCW) 9.2.2.2.2 High Pressure Service Water System (HPSW) 9.2.2.2.3 Low Pressure Service Water System (LPSW) 9.2.2.2.4 Recirculated Cooling Water System (RCW) on Seal Water Systems 29 uxlary Service Water System 9.2.5 Co ol Room Ventilation Chilled ater System (WC) 9.2.5.1 Desig ~asis 9.2.5.2 System D ription and Evaluation 9.2.6 References Ad: "(Deleted- Refer to UFSAR 9.7,

[~Protected Service Water System)" after title 9.3 Process Auxiliaries 9.3.1 Chemical Addition and pling System 9.3.1.1 Design Bases 9.3.1.2 System Description and Evalua n 9.3.1.2.1 Mode of Operation Delete 9.3.1.2.2 Reliability Considerations 9.3.1.2.3 Codes and Standards 9.3.1.2.4 System Isolation 9.3.1.2.5 Leakage Considerations 9.3.1.2.6 Failure Considerations 9.3.1.2.7 Deleted Per 2001 Update 9.3.2 High Pressure Injection System 9.3.2.1 Design Bases 9.3.2.2 System Description and Evaluation 9.3.2.2.1 Mode of Operation 9.3.2.2.2 Reliability Considerations 9.3.2.2.3 Codes and Standards 9.3.2.2.4 System Isolation 9.3.2.2.5 Leakage Considerations 9.3.2.2.6 Failure Considerations 9.3.2.2.7 Operational Limits 9.3.3 Low Pressure Injection System 9.3.3.1 Design Bases 9.3.3.2 System Description and Evaluation 9.3.3.2.1 Mode of Operation 9.3.3.2.2 Reliability Considerations 9.3.3.2.3 Codes and Standards 9.3.3.2.4 System Isolation 9.3.3.2.5 Leakage Considerations 9.3.3.2.6 Operational Limits 9.3.3.2.7 Failure Considerations 9.3.4 Coolant Storage System (31 DEC 2012) 9-2

Oconee Nuclear Station UFSAR U Chapter 9 Ct 9.6.4.6.4 Repairs for H(ot Shutdown 9.6.4.6.5 Fire Protectioin Conclusion 9.6.4.7 Flooding Revi ew 9.6.5 Operation and Testing 9.6.6 References 9.7 Protected Service Water System 9.7.1 General Description 9.7.2 Design Basis 9.7.3 System Description 9.7.3.1 Mechanical 9.7.3.2 Electrical 9.7.3.2.1 Electrical Separation Criteria 9.7.3.2.2 Electrical Testing Requirements 9.7.3.3 Instrumentation and Control (I&C) 9.7.3.4 Support Systems 9.7.3.4.1 PSW Building Lighting System 9.7.3.4.2 PSW Building Fire Protection and Detection System 9.7.3.4.3 PSW Building Heating Ventilation and Air Conditioning System 9.7.3.4.4 PSW Building Underground Duct Bank Drainage System 9.7.3.4.5 Alternate Cooling for the Reactor and Auxiliary Buildings 9.7.3.5 Civil/Structural 9.7.3.5.1 Building Structures 9.7.3.5.2 Subsystem Seismic Analysis 9.7.3.5.3 Dynamic Testing and Analysis of Mechanical Components 9.7.4 Safety Evaluation 9.7.5 References (31 DEC 2012) 9-5

Oconee Nuclear Station UFSAR Chapter 9 Figure 9-27. Auxiliary Building Ventilation System Unit I and 2 Figure 9-28. Auxiliary Building Ventilation System Unit 3 Figure 9-29. Deleted Per 1998 Update Figure 9-30. SSF General Arrangements Longitudinal Section Figure 9-3 1. SSF General Arrangements Plan Elevation 777' and 754' Figure 9-32. SSF General Arrangements Plan Elevation 797+0 Figure 9-33. SSF General Arrangements Plan Elevation 817+0 Figure 9-34. SSF General Arrangements Transverse Section Figure 9-35. SSF RC Makeup System Figure 9-36. SSF Auxiliary Service Water System Figure 9-37. SSF HVAC Service Water System & SSF Diesel Cooling Water System Figure 9-38. SSF Diesel Air Starting System Figure 9-39. SSF Sump System Figure 9-40. SSF 4160V/600V/208V Electrical Distribution Figure 9-41. SSF 125 VDC Auxiliary Power Systems Figure 9-42. Essential Siphon Vacuum System Figure 9-43. Siphon Seal Water System Figure 9-44. Protected Service Water System Figure 9-45. PSW AC Electrical Distribution Figure 9-46. PSW DC Electrical Distribution (31 DEC 2012) 9-9

YT* A* *T L uconee Nuciear Staion I.Replace: ASW with PSW UVIAK Cnapter All cooling systems are designed to be operated and monitored fro the control room. Component design parameters are given in Table 9-4.

The design purpose of each of the cooling water systems is outlined below:

Condenser Circulating Water (CCW) System - This system provides for ooling of the condensers during normal operation of the plant. The system generally uses lake water the ultimate heat sink for decay heat removal during cooldown of the plant. In some events, su as the loss of Lake Keowee, the water trapped in the CCW piping is used as the ultimate heat si CW System is the suction source for other service water systems, including HPSW, LPSW, d SSF ASW.

In addition, CCW provides a heat sink for the RCW system. Following a design basis event involving loss of the CCW pumps, the Emergency Condenser Circulating Water (ECCW) System supplies suction to the LPSW pumps.

Hiah Pressure Service Water (HPSW) System - This system provides a source of water for fire protection throughout the station. In the event of a loss of the normal LPSW supply, HPSW automatically supplies cooling water to the HPI pump motor coolers. For loss of A.C. power, HPSW via the Elevated Water Storage Tank automatically supplies cooling water to the Turbine Driven Emergency Feedwater Pump Oil Cooler and the LPSW Leakage Accumulator for all Units.

Low Pressure Service Water (LPSW) System - This system provide cooling water for normal and emergency services throughout the station. Safety related functions served by this system are:

1. Reactor Building cooling units.

2. Decay heat removal coolers.

3. High pressure injection pump motor bearing coolers.

4. Motor-Driven Emergency Feedwater Pump motor air coolers.

5. Deleted Per 2006 Update.

6. Siphon Seal Water.

Recirculated Cooling Water (RCW) System - This is a closed loop system to supply corrosion inhibited cooling water to various components. This system has no direct safety related functions.

Essential Siphon Vacuum (ESV) System - This system supports the Condenser Circulating Water (CCW) system by removing air from the CCW Intake header during normal and siphon modes of operation. The nuclear safety-related functions are:

1. Remove air from the CCW Intake Headers during normal operation to ensure that the operable Intake Headers are primed at the start of an event requiring the siphon mode of operation.

2. Remove air from the CCW Intake Headers during the siphon mode of operation to ensure that the siphon does not fail due to air accumulation during a Design Basis Accident involving loss of power to the CCW pumps.

Paragraph(s) Deleted Per 2000 Update.

Siphon Seal Water (SSW) System - This system's nuclear safety-related function is to support the ESV system by providing operating liquid to the ESV pumps. The ESV pumps are liquid ring vacuum pumps which require a continuous supply of water in order to create a vacuum. Additionally, it has a non-nuclear safety-related function of providing sealing and cooling water to the CCW pumps and motors.

On July 18, 1989 the NRC Issued Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment," requesting holders of operating licenses to supply information about their respective service water systems to assure the NRC of compliance with the recommended actions of (31 DEC 2012) 9.2 - 3

Oconee Nuclear Station UFSAR Chapter 9 vacuum (ESV) system is connected to the CCW inlet header to remove non-condensible gases during normal and siphon operations.

Pursuant to the recommendations of the Oconee Probabilistic Risk Assessment study a pushbutton has been installed in the control room for sending a close signal to the CCW pump discharge valves. The capability to close the CCW valves is needed to protect against the possibility of CCW siphoning into the turbine building basement, causing flooding. -- Delete Text.

The intake canal that supplies water from Lake Keowee to the suction o the CCW pumps contains a submerged weir. The purpose of this weir is to provide an emergency po of cooling water if the water supply from Lake Keowee were lost. This emergency pond could be re rculated through the condensers and back to the intake canal for decay heat removal as long as the inte canal le.velremains sufficient.

~

"r1v, ~~ ~ 1r~ ~ ~r'!~Tir~ ~ ~ T+ -A* ~rI ~ii

+I,- Air ¶1 i cOnAer.'au.ve aessumpuon-A- M---Rnn tnoim - rsea :nin sRonggaak 6961gn. -a29gon ruANO-9-1%ur4  ;@4miyp ria Moaon, 9 concluIded that a rapid drawdown of L sake Keowaa nould caure conidAr-Ahla diplace.m t* of the r*iprap sed t fares the-weir. However., the AECG staff did not require Duke Powsr to reodesign the mvir, since the water. tapped in the condenser intake and dirch ar piping lbelowA elevation 7.91 ft USL is adequate to sRu n'-'IAthe three _-e ---- *r .-' _-_m- ._ .-- oi ! off for s _feshutdo .nfo r a neno d o f 3 7 da-.. -'s Auxiliary ervice Water (ASW) stem is capable of using the inventory trapped in the CCW piping .. 3). Therefore, the licensing basis does not rely on the weir nor recirculation of the take canal for deca eat removal after a loss of Lake Keowee event (Reference 2).

9.2.2.2.2 High Pressure Servic ater Syste (HPSW)

The schematic arrangement of the HPSW sy m is shown Figure 9-10. This system is used primarily for fire protection throughout the Oconee stati . In the e nt of a loss of the normal LPSW supply, HPSW automatically supplies cooling water to t HPI pum motor coolers. For loss of AC power, HPSW via the elevated water storage tank automa ically sup lies cooling water to the turbine driven emergency feedwater pump oil cooler for all units. W is a o used as a backup supply to the SSW system. Refer to Sections 16.9.7 and 16.9.8 for specific re ireme ts to support the LPSW System.

Two full size (6000 gal/min at 117 psig) and one reduced s e (5 0 gal/min at 117 psig) high pressure service water pumps supply the high pressure system. A 1000 Replace: "Auxiliary Service Water provides inventory for a backup supply of water. (ASW)" with "Protected Service Water The 500 gal/mm pump will normally operate to keep pressure on the (PSW)" and "9.2.3" with "9.7 one full size pump provides adequate capacity for automatically maintaining the elevated water storage tank inventory. The second full size pump is an installed spare. The HPSW pumps take suction from the CCW system. The HPSW and LPSW pump suctions are connected to the 42 inch cross-connection between the Condenser Circulating Water inlet headers for the three units. Manual isolation valves are provided so that service water may be supplied from any or all of the inlet headers.

Portions of the High Pressure Service Water system are credited to meet the Extensive Damage Mitigation Strategies (B.5.b) commitments, which have been incorporated into the Oconee Nuclear Station operating license Section H - Mitigation Strategy License Condition.

9.2.2.2.3 Low Pressure Service Water System (LPSW)

The schematic arrangement of the LPSW system is shown on Figure 9-11 and Figure 9-12. Oconee 1 and 2 share three 15,000 gal/min LPSW pumps. The LPSW pumps and the HPSW pumps take suction from the 42 inch crossover line between the condenser inlet headers; two LPSW pumps are supplied by one suction line and the other pump is supplied by the other suction line. The HPSW system is connected to (31 DEC 2012) 9.2- 5

UFSAR Chapter 9 Oconee Nuclear Station non-nuclear safety-related data points associated with the ESV/SSW/ECCW systems are sent to the plant computer.

The SSW System consists of two headers that are supplied water from the Low Pressure Service Water (LPSW) system. Only one header is needed to supply all loads. However, both SSW headers are normally in service so that a single failure in the LPSW system cannot cause a loss of safety function.

The SSW supply water routes from the Turbine Building to the ESV Building, where it is strained. Once strained, SSW routes to the ESV pumps and to the CCW pumps. SSW provides an operating liquid for the ESV pumps and provides sealing and cooling water the CCW pump shaft seal and motor bearing cooler. The nuclear safety-related function of the SSW System is to provide the operating liquid to the ESV pumps. The ESV pumps are liquid ring vacuum pumps which require a continuous supply of water in order to create a vacuum. As the header branches to the ESV pumps and then branches to each ESV pump individually, a solenoid valve is contained at each pump. This solenoid valve is interlocked with the ESV pump control circuitry. The valve opens when the pump starts and closes when it stops. A failure of one of these solenoids would cause a single ESV pump to be inoperable. The SSW system function would not be affected, since it could successfully deliver water to the remaining ESV pumps.

The SSW system contains provisions for connection of a submersible pump to supply sealing/cooling water to the CCW pumps. Both the ESV and SSW systems are designated as QA Condition I systems.

They are seismically designed and designed to continue functioning with a single, active failure.

However, they are not designed for tornado loads. Interfacing structures existing prior to the installation of these systems are designated QA Condition 4. The ESV Building shell is also a QA Condition 4 structure.

/Add: "(Deleted - Refer to UFSAR 9.7, Protected Service title 9.2.3 Auxiliary Service Water System k iWater System)" after The Aiuxiliaay' Sor.'ice Water- System is designed for decay heat re-moaval fohllowing a comncurent lor's of theminf_ feedw~atr systsm, Emergency F.edwater System, and Decay Heat Removal System. -The ortions of the Station Auxiliary Service Water system are credited to meet the Extensive Damage Mitigation Strategies (B.5.b) commitments, which have been incorporated into the Oconee Nuclear 7Stat~ion operating license Section H - Mitigation Strategy License Condition.

_2*'1-F...De e.i.ptio .n.

SYSt

+  :+

faw 900liAg w"OF fer-degay heat Femeval Grer and- wav.ratering lines. A4i AuyiiliaFy ROP4., in the AuxiliaFy Building 2t Riau 2:71 ta-k-e-s- its. Suction from tha ocapaA a jigaka ymcilit into the steam g@A@;:atQN Of 020h J'Ait th@ @ffi@FPRG)' fOO&AQ40F h@Q ;6%. ;044, w2ta; ig vaporized in the stmm 3f r-offiewfig r-esid-IIA-4 law-At. and- d;Umped to U--a- 2-1-wiliar-y sapArae water. pump is an and sue- --- C,-i pump w4h a rated GapaGity of 3QQQ I/ : 4. 4 1 U A jr I On It has beon submitted to the following tests:

_____e__________n * \z:iRevised paragraph replacing PSW for

.. no.Witnessed......... t.st IStation ASW and move to UFSAR 9.7 2-3 Witessed perfsrmance test 9.2 -10 (31 DEC 2012)

Oconee Nuclear Station UFSAR Chapter 9 4, A4ll test certificates; for asing, impeller, and sha'"

the pump powr supply is taken fi rvom th 1160olti e anhdby usi Nos 1.

All altes requirold for operation Of tChe Auiliaed Seric Water System aWe)either chck valves or The WC S emi The pump scation is equpped wi4t a manually oported butefy valve and the dischare Syth a chovk valve iand manually opetred gato valve. The pui p is eqippod with 2 minimumT flow Path to thealier 9.. Certimd ow distcharge crossover lne9, which is islated by a globelvave. Tha individl linesps to HeatSurintk tranh satam gpnergatr auilinay feednatera esider ar equipped wnith check walve syad nem noumaltclosse gate valve- hcnh is used-to n coentl flow, The majorisys onon embedded piping is Dbe Class F.

9.2.4 Ultimate Heat Sink Lake Keowee supplies the Condenser Circulating Water (CCW) tern and the lake water generally serves as the ultimate heat sink for Oconee Nuclear Station. In some ev such as loss of Lake Keowee, the water trapped in the CCW piping serves as the ultimate heat sink CCW system is described in Section 9.2.2.2. 1.

Delete 9.2.5 Control Room Ventilation Chilled Water System (WC)

The WC System is shown schematically on Figure 9-24.

9.2.5.1 Design Basis The WC System provides chilled water for the Control Room Ventilation System for all three units. The major equipment of the chilled water system is arranged in two parallel redundant trains with one supply and return line and each train capable of supplying the required cooling capacity. A temporary cooling train and piping may be installed in parallel with the permanent chilled water system equipment. The temporary cooling train and piping will connect to the system supply and return piping and be capable of supplying the required cooling capacity. The bases to one of the Technical Specification 3.7.16 addresses the use a temporarily installed full capacity control area cooling train as one of the Tchnical Specification 3.7.16 required WC trains.

9.2.5.2 System Description and Evaluation The WC permanently installed chillers are each made up of a compressor, an evaporator and a refrigerant condenser. The single stage compressor is driven by an open drip proof motor. Both the evaporator and condenser are horizontal shell and finned tube design with individually replaceable tubes. Two chillers are provided for this system, each with 100% capacity.

For the permanently installed WC cooling trains condenser water temperature is measured by a thermistor located upstream of the condenser. This sensor is used to generate a signal that modulates a three-way bypass valve located downstream of the condenser, to maintain proper condenser water temperature. (i.e.

as the temperature increases, the bypass port on the three-way valve is modulated towards closed, to maintain proper entering condenser water temperature for operating the chiller.)

A temporary cooling train with piping connected to the WC system chilled water return and supply piping may be used, providing 100% cooling capacity.

(31 DEC 2012) 9.2 -11

Oconee Nuclear Station UFSAR Chapter 9 In addition to bil food and bloed, Duke indicated otn a numaber of ocasionsgthat the Stationbwuiliart auxiciwa@r* system ans a athetrbnae meildans of decay hueat rdemoial Duke did not readit ths ration aumilipryser~c waterssystm in the mitigation of the turbine building flood. In addition, the NRC did not cridit thn station apuiliary ervic water-systm for the mitigation of the trbrine building flood @;or1 in the licening basisanalysis.

backfit or ELECTRICAL SEPARATION CRITERIA J Selected motor operated valves and selected pressurizer heaters are capable of being powered and controlled from either the normal station electrical systems or the SSF electrical system. Suitable electrical separation is provided in the following manner. Electrical distribution of the SSF is identified in Figure 9-40 and Figure 9-41 is provided by the SSF motor control centers (MCC's). Loads fed from MCC's I XSF, 2XSF, 3XSF, and XSF are capable of being powered from either an existing plant load center or the SSF load center through key interlocked breakers at the MCC's. These breakers provide separation of the power supplies to the SSF loads.

Loads fed from MCC PXSF are capable of being powered from either Unit 2 B2T or the SSF Diesel via switchgear OTS 1. Breakers feeding OTS I are electrically interlocked and provide separation of the power supplies to the SSF loads.

During normal operation, these loads are powered from a normal (non-S SF) load center via the SSF MCC's IXSF, 2XSF, 3XSF (Group B) or switchgear OTS 1 via SSF MCC PXSF (Group C).

During operation of the SSF, these loads are powered from the SSF diesel generator via the SSF load center/switchgear and SSF MCC's.

9.6.3 System Descriptions 9.6.3.1 Structure The Standby Shutdown Facility (SSF) is a reinforced concrete structure consisting of a diesel generator room, electrical equipment room, mechanical pump room, control room, central alarm station (CAS), and ventilation equipment room. The general arrangement of major equipment and structures is shown in Figure 9-30, Fi2ure 9-31, Figure 9-32, Figure 9-33 and Figure 9-34.

The SSF has a seismic classification of Category 1. The following load conditions are considered in the analysis and design:

1. Structure Dead Loads

2. Equipment Loads

3. Live Loads

4. Normal Wind Loads

5. Seismic Loads

6. Tornado Wind Loads

7. Tornado Missile Loads

8. High Pressure Pipe Break Loads (31 DEC 2012) 9.6-3

Oconee Nuclear Station UFSAR Chapter 9 9.7 Protected Service Water System 9.7.1 General Description The Protected Service Water (PSW) system is designed as a standby system for use under emergency conditions. The PSW system provides added "defense in-depth" protection by serving as a backup to existing safety systems and as such, the system is not required to comply with single failure criteria. The PSW system is provided as an alternate means to achieve and maintain safe shutdown conditions for one, two or three units following certain postulated scenarios. The PSW system reduces fire risk by providing a diverse power supply to power safe shutdown equipment in accordance with the National Fire Protection Association (NFPA) 805 safe shutdown analyses. The PSW system requires manual activation and can be activated if normal emergency systems are unavailable.

The function of the PSW system is to provide a diverse means to achieve and maintain safe shutdown by providing secondary side decay heat removal, RCS pump seal cooling, RCS primary inventory control, and RCS boration for reactivity management following plant scenarios that disable the 4160V essential electrical power distribution system. Following achieving safe shutdown, a plant cooldown is initiated within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of event initiation. The PSW system is not an Engineered Safety Feature Actuation System (ESFAS) and is not credited to mitigate design basis events as analyzed in UFSAR Chapters 6 and

15. No credit is taken in the safety analyses for PSW system operation following design basis events.

Based on its contribution to the reduction of overall plant risk, the PSW system satisfies Criterion 4 of 10 CFR 50.36 (c)(2)(ii) and is therefore included in the Station Technical Specifications.

Core decay heat removal is provided by feeding the steam generators from the PSW pumps (booster and high head pumps) via PSW flow control valves. Core reactivity is controlled in a safe manner by injecting borated water from the borated water storage tank (BWST) into the RCS to maintain adequate shutdown margin. RCS inventory control is provided by existing plant equipment that can be selectively powered from the PSW Electrical Distribution System. Specifically, one High Pressure Injection (HPI) pump (either "A"or "B"), the associated suction valve from the BWST (HP-24), the RCP seal injection flow control valves (HP-139 and HP-140), and the "A" HPI injection valve (HP-26) can be powered from PSW to provide RCS makeup. RCS letdown can be provided by repowering the Reactor Vessel (RV) Head Vents (RC-159 and RC-160) and the RCS Loop High Point Vent Valves (RC-155, -156, -157, -158) and repositioning the valves as needed to control RCS inventory. These valves are capable of being supplied with electrical power from the PSW switchgear. Manual power transfer control switches for these components are located in each respective unit's control room.

The PSW Electrical Distribution System can be used to repower a number of pressurizer heaters to establish and maintain a steam bubble in the pressurizer to aid in RCS pressure control. Selected pressurizer heaters with a nominal combined capacity of >400 KW are capable of being supplied with electrical power from the PSW switchgear. Manual power transfer switches for these components are located in each respective unit's East Penetration Room.

(Date) 9.7-1

Oconee Nuclear Station UFSAR Chapter 9 The PSW Electrical Distribution System also supplies power to the Vital Instrumentation and Control (I&C) Battery Chargers to maintain electrical power on the vital I&C buses. The PSW Electrical Distribution System can also be aligned to supply power to the Standby Shutdown Facility (SSF) Electrical Distribution System should the normal and emergency power sources to the SSF be lost.

The PSW system does not provide the primary success path for core decay heat removal following design basis events and transients. The Emergency Feedwater (EFW) System serves as the primary success path for design basis events and transients in which the normally operating main feedwater system is lost and the steam generators are relied upon for core decay heat removal. The PSW system serves as a backup to the EFW system and adds a layer of defense-in-depth to the SSF Auxiliary Service Water (ASW) System, which also serves as a backup to the EFW system.

The PSW system reduces fire risk by providing a diverse QA-1 power supply to power safe shutdown equipment thus enabling the use of plant equipment for mitigation of certain fires as defined by the ONS Fire Protection Program. For certain scenarios inside the Turbine Building (TB) resulting in loss of 4160V essential power, either the SSF or PSW system is used for reaching safe shutdown. The PSW system can achieve and maintain safe shutdown conditions for all three units for an extended period of operation during which time other plant systems required to cool down to MODE 5 conditions will be restored and brought into service as required. Similar to the SSF, the PSW system is equipped with a portable pumping system that may be utilized as necessary to replenish water to the Unit 2 embedded Condenser Circulating Water (CCW) piping. The water in the Unit 2 embedded CCW piping is used as a suction source for the PSW system. Electrical power is supplied from the PSW electrical system. The PSW portable pump is located in an onsite storage location. The portable pumping system is not expected to be necessary unless there is a prolonged use of the PSW system to feed the steam generators. Should there be a prolonged use of the PSW system, the portable pumping system would be used to replenish the water in the CCW piping since the PSW system takes suction off the CCW pipe at its low point in the Unit 2 Auxiliary Building.

The PSW system consists of the following:

1. PSW building and associated support systems.

2. Conduit duct bank from the Keowee Hydroelectric Station underground cable trench to the PSW building.

3. Conduit duct bank and raceway from PSW Building to Unit 3 Auxiliary Building (AB).

4. Conduit duct bank from PSW Building to SSF trench and from SSF trench to SSF.

5. Electrical power distribution system from breakers at Keowee Hydro Units (KHUs) and from breakers connecting the PSW building to the Central Tie Switchyard, and from there to the AB and SSF.

6. PSW booster pump, PSW primary pump, and mechanical piping taking suction from Unit 2 embedded CCW System to the EFW headers supplying cooling water to the respective unit's SGs and HPI pump motor bearing coolers.

7. PSW portable pumping system.

8. PSW pump room exhaust fan (in AB).

(Date) 9.7- 2

Oconee Nuclear Station UFSAR Chapter 9 Portions of the PSW System are credited to meet the Extensive Damage Mitigation Strategies (B.5.b) commitments, which have been incorporated into the Oconee Nuclear Station operating license Section H

- Mitigation Strategy License Condition.

The PSW mechanical system is shown on Figure 9.44. The interface of the PSW system and the EFW system is shown on Figure 10.8. The PSW AC electrical distribution system is shown on Figure 9.45. The PSW DC electrical distribution system is shown on Figure 9.46.

In order to ensure PSW/HPI mitigating component design temperature limits will not be exceeded during PSW/HPI System operation, alternate cooling water and power to the existing ventilation systems is provided to recover from the potential loss of ventilation to the AB and RB (refer to Section 9.7.3.4.5).

9.7.2 Design Bases The design criteria for the PSW System are as follows:

• Major PSW components are Duke Energy Quality Assurance Condition 1 (QA-1). Components that receive backup power from PSW or systems that connect to PSW retain their existing seismic and quality classifications.

• Maintain a minimum water level above the reactor core and maintain Reactor Coolant Pump Seal cooling.

• Provide steam generator secondary side cooling water from Lake Keowee to promote natural circulation core cooling.

• Transfer decay heat from the RCS by steaming the steam generator(s) (SGs) to atmosphere.

• Maintain Keff < 0.99 after all normal sources of RCS makeup have become unavailable, by providing makeup via the HPI system which supplies makeup of a sufficient boron concentration from the BWSTs.

• Control of PSW primary and booster pumps, motor operated valves and solenoid valves, required to bring the system into service are controlled from the Main Control Rooms (MCRs).

9.7.3 System Description 9.7.3.1 Mechanical The mechanical portion of the PSW system is designed to provide decay heat removal by feeding Keowee Lake water to the secondary side of the steam generators. The system, consisting of one booster pump and one primary (high-head) pump, is designed to provide 375 gpm per unit at 1082 psig with SG pressure at the lowest relief valve lift set point. In addition, the system is designed to supply Lake Keowee water at 10 gpm per unit to the HPI pump motor bearing coolers. Refer to Figures 9-44, and 10-8 for more information.

(Date) 9.7-3

Oconee Nuclear Station UFSAR Chapter 9 The PSW system utilizes the inventory of lake water contained in the plant Unit 2 CCW embedded piping. The PSW pumps are located in the AB at Elevation 771'. The PSW booster pump takes suction from the Unit 2 CCW embedded piping and with the aid of the PSW primary pump, discharges into the SG(s) of each unit via separate lines into the emergency feedwater headers. The raw water is vaporized in the SG(s) removing residual heat and discharged to the atmosphere. For extended operation, a portable pump can be utilized via recovery actions to pump water directly from Lake Keowee to the Unit 2 CCW embedded piping.

During periods of very low decay heat the PSW system will be used to establish conditions that support the formation of subcooled natural circulation between the core and the SGs; however, natural circulation 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 minimum HPI flow through the RCS vent valves. 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.

The piping system has pump minimum flow lines that discharge back into the Unit 2 CCW embedded piping. For flow testing to the steam generators, the system is connected to a condensate water source located in the TB that is normally isolated using valves in the AB.

The PSW pumps are controlled from the Unit 2 main control room. Electrically operated valves, used to control flow to the SGs, are controlled from each unit's control rooms. PSW transfer switches for the HPI motor and motor operated valves, required to operate the system, are located in each unit's respective control room. Check valves and manual handwheel operated valves are used to prevent back-flow, accommodate testing, or are used for system isolation during system maintenance. Pumps and valves in the system are ASME Section III Class 3. Piping is designed to the 1967 Edition of USAS B31.1 (Reference 11). The PSW System piping is classified as Oconee Class F.

In service testing of pumps and valves are accomplished in accordance with the provisions of ASME Section Xl and ONS's In-Service Test (IST) program, except for the portable pump. The portable pump is tested periodically to verify flow capability. A recirculation flow path and instrumentation is available for testing of the PSW Booster and PSW Primary Pumps. Active motor operated valves are included in the ONS Generic Letter (GL) 89-10 monitoring program.

9.7.3.2 Electrical The PSW electrical system is designed to provide power to PSW mechanical and electrical components as well as other system components needed to establish and maintain a safe shutdown condition. The system is designed with adequate capacity and capability to supply the necessary loads and is electrically independent from the station electrical distribution system.

(Date) 9.7 -4

Oconee Nuclear Station UFSAR Chapter 9 A separate PSW electrical equipment structure (PSW building) is provided for major PSW electrical equipment. Normal power is provided from the Central Tie Switchyard via a 100 kV transmission line to a 100/13.8 kV substation located adjacent to Oconee Nuclear Station (ONS) and then via a 13.8 kV feeder that enters an underground ductbank leading to the PSW building. This power path from the Central Tie Switchyard to the PSW Switchgear is non QA-1. Alternate QA-1 power is provided from the KHUs via a tornado protected underground path. These external power sources provide power to transformers, switchgear, breakers, load centers, batteries, and battery chargers located in the PSW electrical equipment structure (PSW building). The PSW DC system consists of two (2) batteries, two (2) battery chargers, a distribution center and panelboards. Either battery can be aligned to either battery charger. Refer to Figures 9-45 and 9-46 for additional information.

The power system provides primary or backup power to the following:

• PSW booster pump

• PSW primary pump

• Required 125 VDC Vital I&C Normal Battery Chargers (CA & CB)

" One HPI pump (either "A" or "B") motor per unit

" HPI valves needed to align the HPI pumps to the BWSTs

" HPI valves and instruments that support RCP seal injection and RCS makeup

• RCS and Reactor Vessel Head high point vent valves

" Portable pump (if not self-powered)

• Select groups of pressurizer heaters (nominal capacity in excess of 400 KW)

" Standby Shutdown Facility (SSF)

The PSW Electrical Distribution System does not provide the primary success path for supplying electrical power to systems and components used to mitigate design basis events and transients. The two main feeder buses and the three Engineered Safeguards (ES) power strings are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ES systems so that the fuel, RCS, and containment design limits are not exceeded. The main feeder buses and the ES power strings are the primary success path, consistent with the initial assumptions of the accident analyses, and are credited to meet the design basis of the unit. The PSW Electrical Distribution System serves as a backup source of power for certain components normally powered from the three ES power strings.

9.7.3.2.1 Electrical Separation Criteria The PSW electrical power distribution system has only one train; however, the PSW Primary and Booster Pump circuits, and the associated valve circuits in the PSW system are separate to the SSF ASW pump and valve circuits with one exception. The PSW 4.16 kV switchgear has a circuit that can repower the SSF 4.16 kV switchgear in the event the SSF normal and emergency power sources are not available.

This circuit is normally electrically isolated from the SSF switchgear. Whenever the PSW 4.16 kV (Date) 9.7- 5

Oconee Nuclear Station UFSAR Chapter 9 switchgear is providing power to the SSF, there will no longer be electrical separation between the PSW and SSF electrical systems.

The KHU generator output breakers to the PSW circuits (KPF-9 and KPF-1O) are electrically interlocked such that both breakers cannot be closed simultaneously. This feature prevents inadvertent connection of the outputs of KHU-1 and KHU-2, maintaining train separation and preventing potential damage to the generators.

9.7.3.2.2 Electrical Testing Requirements The electrical power system components are tested consistent with ONS' testing philosophy as described in the fleet nuclear station directives.

9.7.3.3 Instrumentation and Control (I&C)

The PSW system has dedicated instrumentation and controls located in each main control room (MCR) as follows:

" Two (2) high flow controllers (one per SG)

" Two (2) low flow controllers (one per SG)

" Two (2) flow indicators (one per SG)

• One (1) SG header isolation valve

• Two (2) HPI pump power transfer switches

" Power transfer control switches to HPI valves needed to align the HPI System.

" Power transfer control switches for the Reactor Vessel Head and RCS High Point vent valves.

SG parameters and critical reactor coolant system parameters are monitored in the MCRs. The critical reactor parameters needed to support PSW operation are:

" Two (2) Hot Leg Temperature

• Two (2) Cold Leg Temperature

• Twelve (12) Core Exit Thermocouples

" RCS Pressure (Trains A & B)

• RCP Seal Injection Flow

• HPI Injection Flow (Train A)

" Pressurizer Level (Train A & B)

• PSW Flow (Date) 9.7-6

Oconee Nuclear Station UFSAR Chapter 9 9.7.3.4 Support Systems The PSW Building support systems are designed to provide:

1. Emergency Lighting

2. Fire Protection and Detection

3. Heating, Ventilation, and Air Conditioning

4. Duct Bank and Building Drainage

5. Battery power backup.

9.7.3.4.1 PSW Building Lighting System The PSW Building lighting system consists of exit/emergency signs, security lighting fixtures, indoor and exterior building lighting.

Emergency DC lighting for the PSW Building is provided by self-contained 12VDC battery pack lighting units. These units are located to provide adequate levels of lighting for control panel operation and for entering and leaving the structure.

9.7.3.4.2 PSW Building Fire Protection and Detection System Fire protection for the PSW Building is provided by two hose reel stations inside the building and adjacent fire hydrants outside of the building. The hose reels are located such that hose spray can reach any interior portion of the building. The High Pressure Service Water (HPSW) System at the north and south ends of the PSW building supplies the fire protection water. An outside isolation valve must be opened to operate the interior hose reel stations. This design prevents potential flooding in the PSW Building.

The PSW Building fire detection system consists of a local fire alarm control panel (FACP), a remote fire alarm annunciator panel, photoelectric smoke detectors, heat detectors, an outdoor horn/strobe, indoor horn/strobes and multiple manual pull stations. The system is connected to the Unit 3 FACP via two monitor modules, alarm and trouble. The Unit 3 FACP will alert operators when either module actuates.

9.7.3.4.3 PSW Building Heating Ventilation and Air Conditioning System The PSW building Heating, Ventilation, and Air Conditioning (HVAC) system consists of two subsystems, a ventilation system and an air conditioning system. The PSW HVAC System supports operation of systems and equipment located in the PSW building by maintaining temperature within design limits.

The air conditioning system is normally operating while the ventilation system is in standby. The ventilation system will actuate in the event the air conditioning is lost. Both systems are shut down in event of fire in the building.

(Date) 9.7-7

Oconee Nuclear Station UFSAR Chapter 9 The PSW Building HVAC is designed to maintain transformer and battery rooms within their design temperature range. The HVAC System consists of two (2) systems; a non QA-1/non credited system designed to maintain the PSW Transformer and Battery Rooms environmental profile and a QA-1/credited system designed to actuate whenever the non QA-1 system is not able to meet its design function.

VENTILATION SYSTEM The PSW Building Transformer Room and Battery Rooms have independent ventilation systems. These two systems contain exhaust fans, duct heaters, tornado dampers, backdraft dampers, motor-operated dampers, air inlet dampers, and associated ductwork. Ventilation for the Battery Rooms is designed to provide adequate air flow to prevent buildup of hydrogen emitted from charging batteries in accordance with IEEE-484 (Reference 21). Both ventilation systems are located within the PSW Building and protected from tornado loads. The purpose of the ventilation systems is to maintain the PSW Building at temperatures between 60°F and 120 0 F.

AIR CONDITIONING SYSTEM The PSW Building Transformer Room and Battery Rooms have independent air conditioning systems.

Both systems are similar in that the condensing units are located on concrete pads outside the PSW Building. The transformer space air handling units are mounted on platforms inside the PSW Building east wall. Cooling coils and fans for the Battery Rooms are integral with the Battery Room ventilation system. The purpose of the air conditioning systems is to maintain the PSW Building at approximately 751F. The air conditioning systems are designed in accordance with ASME AG-1-2003 (Reference 17).

9.7.3.4.4 PSW Building Underground Duct Bank Drainage System The underground duct banks and manholes associated with the PSW Building are designed and installed to preclude water entry. In the event of water entry, duct bank conduits are sloped to manholes to prevent standing water accumulation. Manholes and duct banks are provided with gravity drains that exit the ductbank or lead to existing yard drains, or in the case of Manhole 7 and the Technical Support Building (TSB) cable vault, to the Radwaste and Interim Radwaste Trenches.

Manhole inspection ports are provided to ensure that the manholes drains are working properly and there is no standing water in the manholes. The inspection ports are located such that the bottom of the manhole is visible and inspection of the manhole interior may be accomplished by video camera without removing the manhole cover. Manhole drain exit points are provided with animal screens.

Underground drain fields or dry wells are not used.

(Date) 9.7- 8

Oconee Nuclear Station UFSAR Chapter 9 9.7.3.4.5 Alternate Cooling for the Reactor and Auxiliary Buildings Alternate cooling water and power to the existing ventilation systems is provided to recover from the potential loss of normal AB and RB ventilation and to support extended PSW System operation to meet NFPA 805 requirements.

The alternate cooling equipment is included in the QA-5 program in accordance with the Duke Quality Assurance Topical Report as discussed in UFSAR Chapter 17. Existing repowered equipment retains its current quality classification. Cooling water to the RB equipment is supplied from Lake Keowee. Cooling water to the AB is supplied by portable chillers. The equipment is not protected from tornado or external flood damage and is not single failure proof. The equipment is not seismically designed; however, it is designed to preclude interactions with other seismically-designed SSCs during a seismic event.

9.7.3.5 Civil/Structural 9.7.3.5.1 Building Structures The PSW system is housed in four new structures, as follows:

1. PSW Building

2. Conduit duct banks and manholes connecting the Keowee Underground to the PSW Building

3. Conduit duct banks, Technical Support Building (TSB) cable vault, Elevated Raceway, and Manhole 7 connecting the PSW Building with the Unit 3 Auxiliary Building (AB)

4. Conduit duct banks connecting Manhole 7 to the SSF cable trench and the SSF trench to the SSF.

The PSW building houses the major electrical equipment. The building is a reinforced concrete structure consisting of a transformer room, a mezzanine, a cable spreading area, and two battery rooms. The building is seismically qualified to the Maximum Hypothetical Earthquake (MHE) and designed to withstand tornado missiles, wind and differential pressure in accordance with Regulatory Guide 1.76, Revision 1 (Reference 7). The PSW building has a seismic classification of Category 1. The following load combinations were considered in the analysis and design:

• Structure Dead Load

" Equipment Loads

" Live Loads

" Normal Wind Loads

• Seismic Loads

• Tornado Wind Loads

• Tornado Missile Loads (Date) 9.7 -9

Oconee Nuclear Station UFSAR Chapter 9 0 Tornado Differential Pressure Loads A reinforced concrete conduit duct bank connects the Keowee Underground power path to the PSW building. From the PSW building, a second reinforced concrete conduit duct bank/elevated raceway connects to the Unit 3 AB. A third conduit duct bank connects the PSW Building to the existing SSF cable trench. These structures were seismically qualified to the Maximum Hypothetical Earthquake (MHE) and designed to withstand tornado missiles, wind and differential pressure in accordance with Regulatory Guide 1.76, Revision 1 (Reference 7).

The PSW Building and the three duct banks were designed in accordance with the following codes and standards:

1. ACI 349-97 (Reference 3).

2. AISC Manual of Steel Construction, 13th edition, 2006 (Reference 4).

3. ANSI / AISC, N690-1984 (Reference 5).

4. ASCE 4-98 (Reference 19)

5. NUREG-0800, Chapter 3, Revision 3, March 2007 (Reference 20).

6. Regulatory Guide 1.122, Revision 1, February 1978 (Reference 15).

7. Regulatory Guide 1.142, Revision 2, November 2001 (Reference 6).

8. Regulatory Guide 1.76, Revision 1, March 2007 (Reference 7).

9. Topical Report BC-TP-9A, Revision 2, Bechtel Power Corporation, 1974 (Reference 8).

The existing sections of the Interim Radwaste Trench, which the conduit duct bank/elevated raceway from the PSW Building to the Unit 3 AB connects to, were designed in accordance with ACI 318-63 (Reference 9). The existing sections of the SSF trench, which the conduit duct bank from the PSW Building to the SSF connects to, were designed in accordance with ACI 318-71, "Building Code Requirements for Reinforced Concrete" (Reference 10).

The design response spectra for the new structures correspond to the May 1990 El Centro North-South earthquake normalized to a peak ground acceleration of 0.15g for structures founded on structural fill in accordance with the Oconee Nuclear Station current licensing basis. The PSW Building is founded on structural fill. The building design response spectra were developed in accordance with Regulatory Guide 1.122 (Reference 15). The dynamic analysis of the PSW Building is made using the STAAD-PRO computer program with amplified response spectra generated at elevations of significant nodal mass.

9.7.3.5.2 Subsystem Seismic Analysis The PSW mechanical piping system was seismically designed using dynamic modal analysis techniques.

The system was modeled using the lumped mass piping analysis program SUPERPIPE. An adequate number of lumped masses or degrees of freedom are included in the model to determine the response of significant modes. Rigid range acceleration effects are included in the modal analysis. The Oconee Nuclear Station (ONS) earthquake motion is two directional in accordance with UFSAR Section 3.7.2.5.

(Date) 9.7-10

Oconee Nuclear Station UFSAR Chapter 9 Therefore, the PSW structures, systems and components (SSCs) have been analyzed for maximum horizontal component (either X or Z) and the vertical component (Y)for seismic loads applied simultaneously. Pipe supported from multiple levels or structure is designed for an envelope of the response spectra for all supporting structures. Resulting analysis stresses were evaluated using the ASME USAS B31.1.0, 1967 edition (Reference 11).

The PSW mechanical piping was evaluated for potential effects from non-seismic piping and components that may be proximate to the system.

The PSW HVAC system was designed in accordance with ASME AG-1, 2003 (Reference 17).

PSW piping supports were designed in accordance with the AISC Manual of Steel Construction, 6th edition, 1963 (Reference 12) per UFSAR Section 3.9.3.4.2. Tube steel shapes were designed using AISC 7th Edition (Reference 18) with the equations used reconciled with the 6th Edition.

Cable trays located in the PSW Building, the ONS AB, and the Keowee Hydro Station, installed to support the PSW electrical distribution system, were evaluated by the Seismic Qualification Utility Group Generic Implementation Procedure (SQUG GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 3A (Reference 13).

The structural attachment of equipment within the PSW Building was designed in accordance with the following codes and standards:

1. AISC Manual of Steel Construction, 13th edition, 2006 (Reference 4).

2. Regulatory Guide 1.142, Revision 2, November 2001 (Reference 6).

3. AISI North American Specification for the Design of Cold-Formed Steel Structural Members, 2001 Edition (Reference 14).

4. ANSI / AISC N690-1984 (Reference 5).

5. Regulatory Guide 1.122, Revision 1, February 1978 (Reference 15).

6. Regulatory Guide 1.199, November 2003 (Reference 16).

The anchorage of PSW related equipment in the ONS AB was designed in accordance with ACI 318-63 (Reference 9) and the AISC Manual of Steel Construction, 13th edition, 2006 (Reference 4).

9.7.3.5.3 Dynamic Testing and Analysis of Mechanical Components As part of the PSW System implementation process, procedures were established for the startup testing of the PSW mechanical system to verify the following information:

1. An "as-built" verification process is used to verify that the piping, components, and piping support/restraints have been erected within the design tolerance.

(Date) 9.7 - 11

Oconee Nuclear Station UFSAR Chapter 9

2. Vibration monitoring was completed to verify that vibration levels for system components during PSW Booster and PSW Primary Pump operations are within acceptable limits.

9.7.4 Safety Evaluation To verify PSW system performance criteria, thermal-hydraulic (T/H) analysis was performed to demonstrate that the PSW system could achieve and maintain safe shutdown following postulated fires that disable the 4160 V essential power distribution system, without reliance on equipment located in the turbine building. The analysis evaluates RCS subcooling margin using inputs that are representative of plant conditions as defined by Oconee's NFPA 805 fire protection program. 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 and 4160 VAC power 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 PSW system is capable of meeting the relevant NFPA 805 nuclear safety performance criteria.

During periods of very low decay heat the PSW system will be used to establish conditions that support the formation of subcooled natural circulation between the core and the SGs; however, natural circulation 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 minimum HPI flow through the Reactor Coolant (RC) vent valves. 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 and 2 other than operation at nominal full power, the duration of operation in these conditions is insufficient to result in an appreciable contribution to overall plant risk.

As a result, T/H analysis was performed assuming full power initial conditions, as described above and in the Oconee Fire Protection Program, Nuclear Safety Capability Assessment (Reference 2). The plant configuration examined in the T/H analysis is representative of risk significant operating conditions and provides reasonable assurance that a fire mitigated by PSW during these MODES will not prevent the plant from achieving and maintaining fuel in a safe and stable condition.

9.7.5 References

1. Not used (reserved for Nuclear Station Report ONDS-351, "Analysis of Postulated High Energy Line Breaks (HELBs) Outside of Containment," (Rev. 2)).

2. NFPA 805 SER for the Oconee Nuclear Station dated December 29, 2010.

(Date) 9.7 - 12

Oconee Nuclear Station UFSAR Chapter 9

3. American Concrete Institute (ACI) 349-97, "Code Requirements for Nuclear Safety Related Concrete Structures" (and its supplements, except Appendix B).

4. American Institute of Steel Construction (AISC), Manual of Steel Construction, 13th edition, 2006.

5. American National Standards Institute (ANSI) / AISC, N690-1984, "Specification for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities."

6. Regulatory Guide 1.142, "Safety Related Concrete Structures for Nuclear Power Plants,"

Revision 2, November 2001.

7. Regulatory Guide 1.76, "Design Basis Tornado and Tornado Missiles for Nuclear Power Plants,"

Revision 1, March 2007.

8. Topical Report BC-TP-9A, "Design of Structures for Missile Impact," Revision 2, Bechtel Power Corporation, 1974.

9. American Concrete Institute (ACI) 318-63, "Building Code Requirements for Reinforced Concrete."

10. American Concrete Institute (ACI) 318-71, "Building Code Requirements for Reinforced Concrete."

11. American Society of Mechanical Engineers (ASME), United States of America Standard (USAS)

B31.1.0-1967, "Power Piping."

12. American Institute of Steel Construction (AISC) Manual of Steel Construction, 6th edition, 1963.

13. Seismic Qualification Utility Group (SQUG), Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 3A.

14. American Iron and Steel Institute (AISI), North American Specification for the Design of Cold-Formed Steel Structural Members, 2001 Edition.

15. Regulatory Guide 1.122, "Development of Floor Design Response Spectra for Seismic Design of Floor Supported Equipment or Components," Revision 1, February 1978.

16. Regulatory Guide 1.199, "Anchoring Components and Structural Supports in Concrete," November 2003.

17. American Society of Mechanical Engineers (ASME) AG-1-2003, "Code on Nuclear Air and Gas Treatment."

18. American Institute of Steel Construction (AISC) Manual of Steel Construction, 7th Edition.

19. American Society of Civil Engineers (ASCE) 4-98, "Seismic Analysis of Safety-Related Nuclear Structures and Commentary."

20. NUREG-0800, Chapter 3, USNRC Standard Review Plan 3.7.2, Seismic System Analysis, Revision 3, March 2007.

THIS ISTHE LAST PAGE OF THE TEXT SECTION (Date) 9.7 - 13

Oconee Nuclear Station UFSAR Figure 9-12 (Page 1 of 1)

Figure 9-12. Low Pressure Service Water System (31 DEC 2011)

Oconee Nuclear Station UFSAR Chapter 10 10.4.6.5.1 Turbine Trips 10.4.6.5.2 Automatic Actions 10.4.6.5.3 Principal Alarms 10.4.6.6 Interactions with Reactor Coolant System 10.4.7 Emergency Feedwater System 10.4.7.1 Design Bases 10.4.7.1.1 Deleted Per 2002 Update 10.4.7.1.2 Deleted Per 2002 Update 10.4.7.1.3 Deleted Per 2002 Update 10.4.7.1.4 Deleted Per 2002 Update 10.4.7.1.5 Deleted Per 2002 Update 10.4.7.1.6 Deleted per 1996 Revision 10.4.7.1.7 Deleted Per 2002 Update 10.4.7.1.8 Deleted Per 2002 Update 10.4.7.1.9 Deleted Per 2002 Update 10.4.7.1.10 Deleted Per 2002 Update 10.4.7.2 System Description 10.4.7.2.1 Motor Driven EFW Pumps (MDEFWPs) 10.4.7.2.2 Turbine Driven EFW Pump (TDEFWP) 10.4.7.2.3 EFW Pump Suction Source 10.4.7.2.4 EFW Pump Minimum Recirculation 10.4.7.2.5 EFW Discharge Flow Control Valves 10.4.7.2.6 Instrumentation and Controls 10.4.7.2.7 Alternate Flow Path 10.4.7.2.8 Alarms 10.4.7.3 Safety Evaluation 10.4.7.3.1 EFW Reponse Following a Loss of Main Feedwater 10.4.7.3.2 EFW Response Following a HELB 10.4.7.3.3 EFW Response Following a SBLOCA 10.4.7.3.4 EFW Response Following a SGTR 1/.la.: "StationASW'with"PSW" 10.4.7.3.5 EFW Response Following a MHE 10.4.7.3.6 EFW Response Following a Tornado 10.4.7.3.7 EFW Response FollqM"1 Wa 10.4.7.3.8 Initiation of SSF A , Station ASW, d HPI Forced Cooling 10.4.7.4 Inspection and Test 10.4.7.5 Instrumentation Requirements 10.4.7.5.1 Turbine Driven Emergency Feedwater Pump 10.4.7.5.2 Motor Driven Emergency Feedwater Pumps 10.4.7.5.3 EFW Flow Indication to the Steam Generators 10.4.7.5.4 UST Level Indication 10.4.8 OTSG Condenser Recirculation System 10.4.8.1 Design Bases 10.4.8.2 System Description 10.4.9 References (31 DEC 2012) 10 -2

Oconee Nuclear Station UFSAR Chapter 10 The ansemlnsadtemi he mai steam lnes and he main and..

and emergen /mergeRplace:

System is "5. The Protected capable Service of supplying both Water ,tea and SGs fteam n Power Conversion System which penetrate the all three units at full secondary system d by the turbine stop valves and the normal and emergePnf pressure."

Feedwater supply to the steam generators f owing a reactor shutdown is assured by one of the following methods:

1. Either of the two main feed ter pumps is capable of supplying both steam generators at full secondary system pressure.

2. The hotwell and cond ate booster pump combination has discharge shutoff head of approximately 620 psia. Three se of half-size pumps are provided. If required, the Turbine Bypass System can be used to reduce condary system pressure to the point where one of the hotwell and condensate booster puwm ombinations can supply feedwater to both steam generators.

3. A sep e Emergency Feedwater System for each unit will supply feedwater at full system pressure (see ction 10.4.7).

4. ternate auxiliary feedwater supplies are available from the Emergency Feedwater System of each of

6. The SSF Auxiliary Service Water System is capable of supplying both steam generators of all three units at full secondary system pressure.

10.4.6.3 Safety Evaluation The design, material, and details of construction of the feedwater heaters are in accordance with the ASME Code,Section VIII, Unfired Pressure Vessels.

The Feedwater System has been reviewed to determine the potential for "water hammer" during anticipated operational occurrences. It has been concluded that the existing Oconee Feedwater System is adequate to prevent flow instabilities. Because design features of the feedwater system preclude the probability of destructive "water hammer" forcing functions resulting from uncovering feedwater lines, no analyses have been performed nor test program conducted regarding this occurrence. The following considerations support this conclusion:

1. Neither the Main nor Emergency Feedwater Systems has horizontal or downward-sloping pipe runs adjacent to the steam generator. The auxiliary piping remains below the level of its junction with the steam generator. The main feedwater line rises above its steam generator connection only after downward and horizontal runs which effectively form a loop seal. Only in the unlikely event of steam generator shell pressure near the vapor pressure of the water in this pipe could a steam void occur.

2. The main and emergency feedwater distribution heads on the steam generator are designed to remain flooded regardless of steam generator water level, and would in any event be self-venting if steam were introduced. The main ring header is fed from the bottom, external to the steam generator, and empties upward through the vertical inlet lines. The auxiliary ring headers are similar in design to the main header. None of the feedwater headers can spontaneously drain into the steam generator.

3. Each steam generator has its auxiliary header separate from the main header. Therefore, there is no need to deliver the relatively cool auxiliary feedwater through the normal path for main feedwater. In addition, the QA- 1 portions of Main FDW have been analyzed for pressure transient forces due to control valve closure and pump trip resulting from actuation of the Automatic Feedwater Isolation System circuitry.

(31 DEC 2012) 10.4- 5

UFSAR Chapter 10 Oconee Nuclear Station EFW system capacity is sufficient to support a 50'F per hour cooldown rate, this rate is not achievable during certain events, such as a natural circulation cooldown. The EFW System design basis includes the ability to perform its function in the event of a single active failure. However, in some instances, as addressed in Section 10.4.7.3, alternate capability and operator actions are credited for performing the EFW function to compensate for specific single failures and system vulnerabilities that have been identified. The EFW System is not considered to be an Engineered Safeguard System and therefore was not designed to meet all of the design criteria applicable to Engineered Safeguard Systems. The EFW System is shown in Figure 10-8.

For diversity, the EFW System includes two AC motor-driven pumps and one turbine-drive pump that is independent of AC power. Sources of steam for driving the turbine-driven EFW pump (TDEFWP) are available from both steam generators. Following a loss of all AC power, the turbine-driven EFW pump will automatically actuate and is capable of operating for at least two hours completely independent of AC power. The water inventory that is immediately available to the turbine-driven EFW pump is sufficient to supply feedwater to the steam generators for at least 40 minutes assuming automatic steam generator level control and no reliance on operator action.

The EFW System is designed to start automatically in the event of loss of both main feedwater pumps.

The automatic start on loss of both main feedwater pumps meets the single failure criterion. All automatic initiation logic and control functions associated with the EFW pumps and control valves FDW-315 and FDW-316 are independent from the Integrated Control System (ICS). Each OTSG is provided with a level control system (see UFSAR Section 7.4.3.2) that, on demand, enables the EFW System to supply sufficient initial and subsequent flow to the necessary SG to assure adequate decay heat removal.

The seismic qualification of the EFW System and Quality Group Classification is described in UFSAR Section 3.2.2. Only those components listed in UFSAR Section 3.2.2 are seismically qualified. The TDEFWP supporting equipment is not fully seismically qualified and therefore is not credited for Maximum Hypothetical Earthquakes (MHEs). However, it has been evaluated against Seismic Qualification Utility Group (SQUG) criteria and is expected to be available following a seismic event.

Although redundancy is provided by two full-capacity seismically qualified Motor Driven Emergency Feedwater Pumps (MDEFWPs), they are also susceptible to failure in a seismic event due to flooding induced by the event. However, alternative seismically qualified means of decay heat removal are provided by the Standby Shutdown Facility (SSF) Auxiliary Service Water (ASW) System and the High Pressure Injection (HPI) System.

The EFW System is seismically qualified to the MHE level out through the first isolation valves, consistent with the design criteria given in Section 3.7.3.9. Piping beyond these boundary points is not seismically qualified. The primary suction to the EFW pumps is from the UST. The Upper Surge Tank (UST) is seismically qualified. Operator action is relied upon to shift the suction of the EFW pumps from the UST to the non-safety condenser hotwell before the UST is completely depleted. The condenser hotwell is seismically qualified with a nominal capacity of 120,000 gallons (References 12, 13, and 14).

However, not all piping from the condenser hotwell, such as the suction supply to the TDEFWP and to the hotwell pumps, is seismically qualified. The piping from the hotwell to the TDEFWP; however, is designed and supported such that it would be expected to withstand the design basis earthquake. The piping from the hotwell to the MDEFWPs is seismically qualified.

Portions of the EFW System are vulnerable to tornado missiles. Thus, the plant relies upon diverse means to provide fdý,fdVI event of a tornado. These diverse means include the SSF ASW System and t station ASW System. Repiao: "station ASW' with "PSW" The Emergen s not designed to withstand the effects of internally generated missiles. If such an event were to occur and if main feedwater were unavailable, the single train SSF ASW System would provide an assured means of providing heat removal from the SGs. A detailed 10.4- 8 (31 DEC 2012)

Oconee Nuclear Station UFSAR Chapter 10

19. TDEFWP tripped

20. FDW-315 controller Bypassed

21. FDW-316 controller Bypassed

22. Loss of Primary control power for FDW-315

23. Loss of Primary control power for FDW-316

24. FDW-315 Hand/Auto Station Failure

25. FDW-316 Hand/Auto Station Failure

26. FDW-315 Hand/Auto Station in Manual Mode

27. FDW-316 Hand/Auto Station in Manual Mode

28. FDW-315 Automatic Control on Primary Control

29. FDW-316 Automatic Control on Primary Control

30. FDW-315 Nitrogen Pressure A Low

31. FDW-316 Nitrogen Pressure A Low

32. FDW-315 Nitrogen Pressure B Low ,-.Replace: "7. The Protected

33. FDW-316 Nitrogen Pressure B Low Service Water System is capable of supplying both

_SGs of all three units at full

..... _E.alua ...Safety ...

10.4.7.3 .... / ,. ,secondary system/

Feedwater inventory is maintained in the SGs following reac r shutdipressure."

methods listed: r

1. Either of the two main feedwater pumps in combinatio with a hotwell pump and a condensate booster pump are capable of supplying both SGs at full s ondary system pressure.

2. The two MDEFWPs are capable of supplying their ass iated SG at full secondary system pressure.

3. The single TDEFWP is capable of supplying both S at full secondary system pressure.

4. An alternate EFW supply available from the EF System of one of the other units, capable of supplying both SGs at full secondary system pres e.

5. The hotwell and condensate booster pump corn nation has a discharge shutoff head of approximately 620 psia. There are three hotwell pumps and t ee condensate booster pumps. If required, the Turbine Bypass System or the Atmospheric Dump alves (ADVs) can be used to reduce secondary system pressure to the point where one hotwell and condensate booster pump combination can supply feedwater to both SGs.

6. The SSF Auxiliary Service Water Sys is capable of supplying both SGs of all three units at full

7. The station Auxiliary Service Water System may be used to maintain SG water inventory following.

SG depressurization to remove decay heat in the long term.

of the diverse methods listed above. Although redundancy and diversity is provided as listed above, the EFW System has been designed with special considerations to enable it to function when conventional means of feedwater makeup may be unavailable.

(31 DEC 2012) 10.4- 15

Oconee Nuclear Station UFSAR Chapter 10 10.4.7.3.4 EFW Response Following a SGTR This event does not assume a loss of offsite power has occurred. With offsite power available, main feedwater should continue to operate and provide inventory to the SGs. In addition, the condenser should remain available as a means of removing heat from the SGs via the Turbine Bypass System to the Condenser Circulating Water (CCW) System. However, should the Main Feedwater System be unavailable, the EFW System would be required to provide secondary side cooling. All three EFW pumps would be available to provide inventory to the SGs. Prior to isolation of the ruptured SG, EFW inventory requirements are diminished to a certain degree due to primary system leakage boiloff in the ruptured SG.

If the EFW flow control valve for the unaffected SG failed to open, the flow path can be realigned to bypass the failed valve and reach the SG through the main feedwater startup flow path. This alternate path through the main feedwater startup control valve relies on non-safety equipment and non-safety support systems (electrical power and instrument air). With offsite power being available, the main feedwater startup path should remain available. However, if this path were unavailable, the SSF ASW System provides an alternate means of establishing feedwater flow to the unaffected SG. Prior to cooling the unit down to DHR conditions, one RCP per loop is tripped, further reducing the demand for EFW. The flowrate and inventory demands for EFW following a SGTR event is bounded by the demand for EFW following a loss of main feedwater with offsite power available.

If the EFW flow control valve on the unaffected SG fails open (on a loss of compressed air and nitrogen),

this could result in the SG overcooling. The safety analyses assume action outside the Control Room for local manual control of the EFW control valve if the valve failed open. The EFW flow control valves are located in the penetration room adjacent to the Control Room.

10.4.7.3.5 EFW Response Following a MHE The EFW pumps are located in the basement of the Turbine Building and are therefore, subject to complete failure as a result of flooding caused by a rupture of the non-seismic portion of the condenser circulating water line. In such an event, the SSF ASW System would be relied upon for shutdown decay heat removal. The SSF ASW System is not single failure proof. Penetration seals and waterproof doors have been installed between the Turbine Building and Auxiliary Building in each unit to provide waterproofing up to a height of twenty feet above the Turbine Building basement floor. Thus the High Pressure Injection (HPI) System, located in the Auxiliary Building, would be available as an alternative to the EFW System and the SSF ASW System for shutdown decay heat removal (Reference 6).

As defined in Reference 6, Oconee was deemed to meet the criteria of Generic Letter 81-14 regarding adequate post-seismic event decay heat removal capability by:

1. requiring portions of the EFW System (defined in UFSAR Section L.2.2) to be capable of withstanding a MHE, and

2. providing alternative seismically qualified means of decay heat removal with the SSF ASW System and the anHPIthe---Sstm.Add System. new sentence: "Subsequently,[

/ PSW replaced station ASW relative 10.4.7.3.6 EFW Response Following a Tor do to this function."

Reference 7 concludes that the Standard Revie Plan probabilistic criterion is met based upon the probability of failure of the EFW and statio Systems combined with the protection against tornado missiles afforded the SSF ASW System.

10.4.7.3.7 EFW Response Following a SBO This event is similar to the LMFW with LOOP analysis with the additional assumption that the onsite emergency AC power sources have been lost. This results in the loss of the MDEFWPs. The TDEFWP should be available for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during this event because of its AC power independence. The SSF ASW (31 DEC 2012) 10.4- 19

UFSAR Chapter 10 Replace: "station ASW' with "PSW." Oconee Nuclear Station "Station ASW ' !with "PSW '

System; however, is credited to removtthe decay heat in event.

ts The SBO event, which is not a design basis event, is described in UFSAR S tion 8.3 .2.2.4.

10.4.7.3.8 InitIat Stationd HPI Forced Cooling The SSF ASW Syste , station ASW ys em, and orced cooling serve as alternate means of decay heat removal for some gn events described in Section 10.4.7.3.

Once the control room decides to use the SSF ASW system, the system can be aligned within 14 minutes, consistent with the assumptions in the safety analyses. 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.

The SSF contains adequate instrumentation to maintain the plant in a safe shutdown condition. The SSF ASW System is described in Section 9.6 of the UFSAR.

at *rmes hsi taton ystm i a ow eadsystem that requir-es depr-essurization of the steam generators. The The dump SW Voatl 4- +-Ato o*

system is plac in se',d by'cgeoialeoal operatorA S1pum

÷ r~.+U station action

÷r~ . r to depr..s r ize

, a ndh openng

.. the stam genr. tors using the anal,, valwa Lqt+*t*;h e. pensm;.t iep*

If feedwater is unavailable, oper or action is taken on hi-h RC* nregu1re nr nreagr1i7er level to initiate HPI forced cooling. These actions e from the control roo Replace entire paragraph with the following: I PORV, and throttling HPI flow a snecessary. HPI for The Protected Service Water (PSW) system is designed exceeding the initiation criteria. Thee- System is describ as a standby system for use under emergency conditions. The PSW System is powered from either Th the Central Tie Switchyard via a 100 kV transmission 10.4.7.4 Inspection and Testing Re irements line to a 100/13.8 kV substation or the Keowee A comprehensive test program is followed fo he EFW S Hydroelectric Station. The PSW system provides of the activation logic and mechanical compon ts to ass additional "defense in-depth" protection by serving as a unit. backup to existing safety systems and as such, the system is not required to comply with single failure During unit operation, the EFW System is tested by tlizi criteria. The PSW system is provided as an alternate tank dome. Pump head and flow is verified utilizing thi means to achieve and maintain safe shutdown for one, two, or three units.

10.4.7.5 Instrumentation Requirements The PSW System is capable of cooling each unit's RCS to approximately 250 OF and maintaining this condition Sufficient instrumentation and controls are provided to ad' for an extended period. Failures in the PSW system will The safety related instrumentation and controls that monitc not cause failures or inadvertent operations in existing EFW pumps, and automatically align the supply, meet the plant systems. The PSW system is operated from the and separation Main Control Rooms (MCRs) when existing diverse emergency systems are not available. The power to 10.4.7.5.1 Turbine Driven Emergency Feedwat aligned IPSW controlled outside the MCR. heaters must be_manually pressurizer tr outsidethe_

Instrum entation used in the autom atic initiation circui TDEFWP is safety grade, as listed in Section 3.1.1.1, but not all of the equipment required to provide auto start capability is safety grade. This non-safety grade equipment includes: the TDEFWP Auxiliary Oil Pump, the 250VDC Load Center DP (which supplies power to the TDEFWP Auxiliary Oil Pump), the limit switch for MS-93 and and the pressure switch (FDWPS0300) which senses hydraulic pressure for the TDEFWP. Instrumentation used in the automatic initiation of the pump following an ATWS event is not required to be safety grade. A failure in the automatic initiation circuitry will not prevent manual start capability from the Control Room.

10.4 -20 (31 DEC 2012)

Oconee Nuclear Station UFSAR Figure 10-8 (Page 1 of 1)

Figure 10-8. Emergency Feedwater System (31 DEC 2000)

Oconee Nuclear Station UFSAR Chapter 18 18.2 One-Time Inspections for License Renewal 18.2.1 Cast Iron Selective Leaching Inspection Purpose - The purpose of the Cast Iron Selective Leaching Inspection will be to characterize loss of material due to selective leaching of cast iron components in Oconee raw water, treated water, and underground environments.

Scope - The results of this inspection will be applicable to the cast iron components falling within the scope of license renewal. These components include pump casings in several systems along with piping, valves and other components. The Oconee raw and treated water systems containing cast iron components potentially susceptible to loss of material due to selective leaching are the Auxiliary Service Water System,le Low Pressure Service Water System, the Condenser Circulating Water System, the Service Water S tem (Keowee), the Chilled Water System, the Condensate System, and the High Pressure Service Wa System.

Aging Effects - The ins tion will determine the existence of loss of material due to selective leaching, a form of galvanic corrosion d assess the likelihood of the impact of this aging effect on the component intended function. Selective le hing is the dissolution of iron at the metal surface that leaves a weakened network of graphite and iron corr 'on products.

Method - The Cast Iron Selective Le ing Inspection will inspect a select set of cast iron pump casings to determine whether selective leaching f the iron has been occurring at Oconee and whether loss of material due to selective leaching will be an ing effect of concern for the period of extended operation.

A Brinell Hardness check will be performed o e inside surface of a select set of cast iron pump casings to determine if this phenomenon is occurring. Th results of the Cast Iron Selective Leaching Inspection will be applicable to all cast iron components with license renewal scope and installed in applicable environments.

Sample Size - A representative sample of six pump casin will be ir Add Eootnote 1: "The Auxiliary Service leaching, one from each of the following systems on-site: Water System has been replaced by the Protected Service Water System. The

1. Auxiliary Service Water System Auxiliary Service Water pump was one of

2. Chilled Water System selected pumps inspected; hence, reference to Auxiliary Service Water System is correct

3. Low Pressure Service Water System Ifor this program.

4. High Pressure Service Water System

5. Service Water System (Keowee) Footnote 1)"

6. Condensate System (one inspection location on any of the three Oconee Units.)

Industry Codes or Standards - No specific codes or standards exist to address this inspection.

Frequency - The Cast Iron Selective Leaching Inspection is a one-time inspection.

Acceptance Criteria or Standard - No unacceptable indication of loss of material due to selective leaching as determined by engineering analysis. Component wall thickness acceptability will be judged in accordance with the Oconee component design code of record.

Corrective Action - Any unacceptable loss of material due to selective leaching requires an engineering analysis be performed to determine potential impact on component intended function. Specific corrective actions will be implemented in accordance with the Problem Investigation Program. The Problem Investigation Program will apply to all structures and components within the scope of the Cast Iron Selective Leaching Inspection.

(31 DEC 2012) 18.2- 1

UFSAR Chapter 18 Oconee Nuclear Station Timing of New Program or Activity - Following issuance of renewed operating licenses for Oconee Nuclear Station, this inspection will be completed by February 6, 2013 (the end of the initial license of Oconee Unit 1).

Regulatory Basis - Application [Reference 1] and Final SER [Reference 2].

18.2.2 Galvanic Susceptibility Inspection Purpose - The purpose of the Galvanic Susceptibility Inspection will be to characterize the loss of material by galvanic corrosion in carbon steel - stainless steel couples in the Oconee raw water systems.

Scope - The results of this inspection will be applicable to all galvanic couples with the focus on the carbon steel - stainless steel couples in the Oconee raw water systems falling within the scope of license renewal.

Aging Effects - The inspection will determine the existence of loss of material due to galvanic corrosion and assess the likelihood of the impact of this aging effect on the component intended function.

Method - A volumetric or destructive examination at the junction of the carbon steel - stainless steel components will be performed to determine material loss from the more anodic carbon steel. The most susceptible locations will be identified. The exact method of examination will be determined at the time of the inspection.

Sample Size - A sentinel population of the more susceptible locations on all three Oconee units, Keowee, and Standby Shutdown Facility will be selected for this inspection from the following raw water systems within the scope of license renewal.

1. Auxiliary Service Water System

2. Chilled Water System (raw water porti of the chillers)

3. Component Cooling System (raw water porti f the Component Cooler)

4. Condensate System (raw water portions of the Co nsate Cooler and Main Condenser within the scope of license renewal)

5. C r t rAdidn "The inspection "2:

5. Condenser Circulating Water System of the Auxiliary Service Water

6. Diesel Jacket Water Cooling System (raw water portion of the j ac System piping was completed prior to upgrading the system to the

7. High Pressure Service Water System Protected Service Water System."

8. Low Pressure Injection (raw water portion of the Decay Heat Removal Cooler)

9. Low Pressure Service Water System

10. Service Water System (Keowee)

11. Standby Shutdown Facility Auxiliary Service Water System

12. Turbine Generator Cooling Water System (Keowee)

13. Turbine Sump Pump System (Keowee)

Areas of low flow to stagnant conditions in Oconee raw water systems which contain carbon steel -

stainless steel couples are the most susceptible locations. Engineering practice at Duke has been to use stainless steel as a replacement material in raw water systems for several years. Since engineering practice will continue to use stainless steel as an acceptable substitute material, the size of the sentinel population will be dependent on the number of susceptible locations at the time of the inspection.

18.2- 2 (31 DEC 2012)

Oconee Nuclear Station UFSAR Chapter 18 Acceptance Criteria or Standard - For the items comprised of plates, forgings, and welds that will be visually inspected, critical crack size will be determined by analysis. Acceptance criteria for all aging effects will be developed prior to the inspection.

For baffle bolts, any detectable crack indication is unacceptable for a particular baffle bolt. The number of baffle bolts needed to be intact and their locations will be determined by analysis. Acceptance criteria for dimensional changes due to void swelling will be developed prior to the inspection.

For core barrel bolts, and thermal shield bolts any detectable crack is unacceptable. Acceptance criteria for all aging effects will be developed prior to the inspection.

For items fabricated from CASS and martensitic steel, critical crack size will be determined by analysis.

Acceptance criteria for all aging effects will be developed prior to the inspection.

Corrective Action - If the results of the inspection are not acceptable, then actions will be taken to repair or replace the affected items or to determine by analysis the acceptability of the items. Specific corrective actions will be implemented in accordance with the Duke Quality Assurance Program.

Timing of New Program or Activity - The inspections among the three sets of reactor vessel internals will be spaced out over the twenty-year period of extended operation. The first inspection will occur early in the period. The second will occur near the middle of the period, and the third will occur in the latter third of the twenty-year period. (The third inspection will be scheduled prior to the last year of the twenty-year period of extended operation for the unit inspected.)

Regulatory Basis - Renewal Applicant Action Item 4.1 (Items 5, 6, 7, 8, and 9) in the Safety Evaluation Report for BAW-2248A. Duke letter dated December 17, 1999 [Reference 39], and Final SER [Reference 2].

18.3.21 Service Water Piping Corrosion Program Purpose - The purpose of the Service Water Piping Corrosion Program is to assess and manage loss of material due to corrosion for the various component materials in Oconee, Keowee and Standby Shutdown Facility raw water systems and selected Keowee air and gas systems that may challenge the component intended function of pressure boundary. The following raw A= systems within the scope of license repawpj h*in the scope of the Service Water Piping Corrosion Program:

I- Auxliary rvice Water System, ater tem (raw water portion of the coolers),

3. Component Cooling em (raw water side of the component coolers),

4. Condenser Circulating Wa stem,w

5. Diesel Jacket Water Cooling Syste raw water side of the heat exchangers),

6. Essential Siphon Vacuum System, Replace: "Auxiliary" with

7. High Pressure .. . Water Service . It System, "Protected."

o orions)" afteAdd: "(carbon z r "S yste m _ steel

8. Keowee Service Water System, i ) a

9. Keowee Turbine Generator Cooling Water System,

10. Keowee Turbine Sump Pump System,

11. Keowee Vacuum Break System,

12. Low Pressure Injection System (for the raw water side of the Decay Heat Cooler),

13. Low Pressure Service Water System, (31 DEC 2012) 18.3 -27

Oconee Nuclear Station UFSAR Chapter IS Keowee Governor Air systems, the aging effect is loss of material due to general corrosion of carbon steel components.

Method - Inspection methods for susceptible component locations include use of volumetric examinations using ultrasonic testing. Also, visual examination is used as a general characterization tool in conjunction with ultrasonic testing when access to interior surfaces is allowed such as during plant modifications.

Industry Codes and Standards - No code or standard exists to guide or govern this inspection.

Component wall thickness acceptability is judged in accordance with the component design code of record.

Frequency - Because the corrosion phenomena is slow-acting, inspection frequency varies for each location with a periodicity on the order of five to ten years. The frequency of re-inspection depends on previous inspection results, calculated rate of material loss, piping analysis review, pertinent industry events and plant operating experiences.

Acceptance Criteria - No inspection locations falling below the minimum pipe wall thickness values for the inspection locations as defined in the program. These minimum values have been determined based on design pressure or structural loading using the piping design code of record and then applying additional conservatism.

Corrective Action - Inspection locations that fall below the acceptance criteria are repaired or replaced prior to the system returning to service unless an engineering analysis allows further operation. In the cases where a component may be allowed to continue in service, a re-inspection interval is established in the program.

Specific corrective actions are implemented in accordance with the Problem Investigation Program. The Problem Investigation Program applies to all structures and components within the scope of the Service Water Piping Corrosion Program.

Regulatory Basis - The Service Water Piping Corrosion Program is a formalization of a portion of the commitments made in response to GL 89-13, primarily those associated with component pressure boundary maintenance [References 19, 20, 21, 22, and 23]; Application [Reference 1] and Final SER

[Reference 2].

18.3.22 System Performance Testing Activities The following raw water systems have been identified as containing smaller diameter piping that could be aff 4ting and will be managed by System Performance Testing Activities:

1I,.Auxiliary 4rvice Water System,

2. ýeowre bi e Cooling Water System,

3. Keowee Turbine Sump Pump System,

4. Low Pressure Service Water System, Proctd.,Auxiliarbon wisth

5. Siphon Seal Water System, and portions)" after "System"

6. SSF Auxiliary Service Water System.

7. Essential Siphon Vacuum System.

Performance testing for these systems will provide assurance that the components are capable of delivering adequate flow at a sufficient pressure as required to meet system and accident load demands.

Performance testing includes other alternate techniques, for example, periodic monitoring of system operating parameters, for those systems whose design or operation renders conventional testing (31 DEC 2012) 18.3 -29

UFSAR Chapter 18 Oconee Nuclear Station techniques unfeasible. For the Keowee Turbine Generator Cooling Water system, monitoring or bearing temperatures is acceptable.

Periodic operation, inspections and testing are completed for the above systems at a range of frequencies.

The Turbine Generator Cooling Water System is operated at design conditions every time the Keowee operate with unitsS*;eWater bea piping system. s areevluted cintf bet*-:'ee01 b om C-CWI-!00 gring operations.

CCWT 101 ts of

-+-A theI**_'e_.at tests And ^.inpciaons fiy:'@

4 OVO;e:e' Y7es.

For other systems, periodic testing frequencies range from quarterly o every third refueling outage, depending on the system. Fouling is not a concern in the Essential SiPhon Vacuum System since the system is primarily an air system, and any raw water intrusion is insuffici t o allow for fouling.

System performance is determined and iia compared to test acceptance crit established by engineering.

The results of visual inspections are evaluated by engineering. If the resut s of the tests and inspections do not meet acceptance criteria, then corrective actions, which could re uire piping replacement, are undertaken, Specific corrective actions are implemented in accordance w h the Problem Investigation Process. The Problem Investigation Process applies to all structures and co ponents within the scope of the System Performance Testing Activities.

The activities credited here for license renewal are consistent with the Oc ee commitments made in response to Generic Letter 89-13 [References 19, 20, 21, 22 and 23].

The continued implementation of the System Performance Testing Activ ties provides reasonable assurance that the aging effects will be managed such that mechanical com onents will continue to perform their intended functions consistent with the current licensing basis fo the period of extended operation.\ oprtin .

IDelete sentence Regulatory Basis - Application [Reference 1] and Final SER [Reference 2]. beginning "Visual

/inspection ..... "

18.3.23 Tendon - Secondary Shield Wall - Surveillance Program Purpose - The purpose of the Tendon - Secondary Shield Wall - Surveillance Program is to inspect the Secondary Shield Wall Post-Tension Tendon System to ensure that the quality and structural performance of the secondary shield wall is consistent with the licensing basis.

Scope - The scope of this program includes the tendon wires and tendon anchorage hardware, including bearing plates, anchorheads, bushing, buttonheads, and shims of the Units 1, 2, and 3 Secondary Shield Wall Tendons.

Aging Effects - The applicable aging effects include loss of material due to corrosion and cracking of tendon anchorage; wire force relaxation; loss of material due to corrosion and breakage of wires; loss of material due to corrosion and cracking of bearing plate; cracked, split, and broken buttonheads; cracking and loss of material due to corrosion of shims.

Method - Lift-off tests and visual inspections are performed on three randomly selected horizontal tendons.

Industry Code or Standard - No code or standard exists to guide or govern this program.

Frequency - Lift-off tests and visual inspections are performed on three randomly selected horizontal tendons every other refueling outage or every 48 months.

Acceptance Criteria or Standard - No unacceptable visual indication of moisture, discoloration, foreign matter, rust, corrosion, splits or cracks in the buttonheads, broken or missing wires, and other obvious damage as identified by the accountable engineer. Lift-off forces are measured and compared to established acceptance criteria. The minimum required forces for the tendon groups range from 390 kips to 560 kips depending on the location of the group.

18.3 -30 (31 DEC 2012)

Attachment 3 UFSAR 9.7 (PSW System) Figures 9-44 Protected Service Water 9-45 PSW AC Electrical Distribution 9-46 PSW DC Electrical Distribution 10-8 Emergency Feedwater System

t 0) 2'~~

u-t C) 0)

01

-D 0)

U 0) 0 a

AL VALVES

'PS ASNOTED EXCEPT

FIGURE 9-45 PSW AC ELECTRICAL DISTRIBUTION KE *WEE FANT KEOWEE FANT HYORO LINE HYORD LINE I)  :) I) ',)

Oconee Nuclea Station Appendix 10. Chapter 10 TWes arw Figures Ocoi~ee Nuclear Stotioii Appendix 10. Chapter 10 TobIem and Figures Figure 10-8 System Emergency Feedeoter

Attachment 4 UFSAR Markups in Response to RAI Item No. 109(b)

Oconee Nuclear Station UFSAR Chapter 3 3.2 Classification of Structures, Components, and Systems 3.2.1 Seismic Classification 3.2.1.1 Structures The design bases for normal operating conditions are governed by the applicable building design codes.

The basic design criterion for the worst loss-of-coolant accident and seismic conditions is that there shall be no loss of function if that function is related to public safety.

AEC publication TID 7024, "Nuclear Reactors and Earthquake," as amplified in Chapter 3 is used as the basic design guide for seismic analysis.

The design basis earthquake ground acceleration at the site is 0.05g. The maximum hypothetical earthquake ground acceleration is 0.10g. and 0.15g for Class 1 structures founded on bedrock and overburden respectively.

The plant structures are classified as one of three classes according to their function and the degree of integrity required to protect the public.

3.2.1.1.1 Class 1 Class 1 structures are those which prevent uncontrolled release of radioactivity and are designed to withstand all loadings without loss of function. Class 1 structures include the following:

Portions of the Auxiliary Building that house engineered safeguards systems, control room, fuel storage facilities and radioactive materials.

Reactor Building and its penetrations.

CT4 Transformer and 4KV Switchgear Enclosures (Blockhouses) (Reference Section 8.3.1.4.1.)

Unit Vent.

Standby Shutdown Facility (SSF) (Reference Section 9.6.3.4.1.)

o - From the license renewal review, it was determined that Class I civil structures are included in the sco or license renewal.

3.2.1.1.2 Clas' Class 2 structures are thos hose limited damage would not result in a release of radioactivity and would permit a controlled plant shut but could interrupt power generation. Class 2 structures include the following:

Oconee Intake Structure Oconee Turbine and Auxiliary Buildings, cept as included in Class 1 Oconee Intake Canal Dike Oconee Intake Underwater Weir Add new line item: "Protected Service Water (F Keowee Powerhouse Building (Reference Section 9.7.3.5)."

Keowee Spillway Keowee Service Bay Substructure Keowee Breaker Vault Keowee Intake Structure Keowee Power and Penstock Tunnels Keowee Dam (31 DEC 2012) 3.2- 1

UFSAR Chapter 3 Information relating: t~~l¶ seismic design ofplace: SSF"ansystems auxiliaryand components service water pum" is contained

`[ .. Rein Section with place  :" m2 the and9.64.3Re 9.64.1

4. Tornado 9 ."6.*4.1 and 9.64..A with "Protected Service Water pumps" PsW pumps."

The Reactor Coolant Syste ill not be damagedm by a tornado. A loss of Reacto Coolant Pump (RCP) seal integrity all was t units.

t postulated as part of the tornado design basis. Capabili is provided to shutdown safely The Reactor Coo t System, by virtue of its location within the Reactor Building, protected from of the sx. s e oly of secondary sr ieof secondarfo eoesafe .. .

. shutdr .

is assured by an auxilir rvice water pump d~

cateddiverse Irt in the sources Auxiliary Building and taking suctior from Oconee 2 tornado mitigation. These include: 1) the other units' EFW Syste, 2) the ASW "torna ll tump iays after trip of the three reactors.

unitoe uneapproximately 37 I

iplace entire sentence with: "The sources of power to the PSW pumps athe Keowee Hydro Station and the Central Tie Switchyard via a 100 transmission line to a 100/13.8 kV substation."

a. The SSF Reactor Coolant Makeup Pump can take suction from the Spent Fuel Pool. The pump can be supplied power from the SSF Diesel.

sueI ur b.esAIngjee y

listed . The statement, "Capabilityis provided to shutdown powered fromt evPW ilableu

/c ory alernate prehensivel systerPms b Swi e avalbet lowsudw fth ln.Ia "Either..the."A".or."B".High e ctinrgans Prtch hratospecific stemsi sholdte trna iiadprotof. As par ofth pror gina nor were F theytA prwessre Intection Pump can bet to prove this judgement, ow normal qualitative were not performed confirmed that the original tornado is analyses studies have a specitic accident probabilistic safe shutdown after development, able to achieve the NRC. Subsequent risk of not being requested by correct. The assessments were sufficiently small that additional protection is not required.

In addition, there was considerable correspondence between Duke and NRC in the years post-TMI discussing Oconee's ability to survive tornado generated missiles. Based upon the probability of failure of the EFW and Station ASW systems combined with the protection against tornado missiles afforded by the SSF ASW system, the NRC concluded that the secondary side heat removal function complied with the criterion for protection against tornadoes. -Ad 3.2- 4 (31 DEC 2012)

Oconee Nuclear Station UFSAR Chapter 3 3.2.2.4 System Component Classification In the absence of definitive codes, the design criteria applied to pressure retaining system components are generally consistent with the intent of Sections III and VIII of the ASME Boiler and Pressure Vessel Code, the piping system classification applicable to that portion of the system which includes the component, and the required function of the component. Atmospheric water storage tanks important to safety conform to American Waterworks Association Standard for Steel Tanks, Standpipes, Reservoirs and Elevated Tanks for Water Storage, D 100, or equivalent.

Components are listed by system in Table 3-2. This tabulation shows the code to which the component was designed, whether the component was designed to withstand the seismic load imposed by the maximum hypothetical earthquake, and the analytical technique employed in seismic analysis.

Code Applicability: Due to the numerous code references located throughout the UFSAR, no attempt is made to revise these references as codes are amended, superseded, or substituted. Consequently, the station specifications applicable to a given component should be relied upon to determine appplicable codes.

3.2.3 Reference

1. Applicationfor Renewed OperatingLicensesfor Oconee Nuclear Station, Units 1, 2, and 3, submitted by M. S. Tuckman (Duke) letter dated July 6, 1998 to Document Control Desk (NRC), Docket Nos.

50-269, -270, and -287.

2. NUREG-1723, Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2, and 3, Docket Nos. 50-269, 50-270, and 50-287.

3. License Amendment No. 338, 339, and 339 (date of issuance - June 1, 2004); Adoption of Alternate Source Term.

THIS IS THE LAST PAGE OF THE TEXT SECTION 3.2.

Add: "4. NRC Protected Service Water Safety Evaluation Report, dated xx-xx-2013."

(31 DEC 2012) 3.2- 7

UFSAR Chapter 3 Oconee Nuclear Station 3.6.1.3 Safety Evaluation e sis ffects resulting from postulated piping breaks outside containment is contained in Duke Power MVDS Repo o. OS-73.2, dated April 25, 1973 including revisions through supplement 2.

An evaluation of potent non-safety grade control Add new section 3.6.'1.3: "3.6.1.3 'Prote ed Service Water line break accidents is contai in the Duke Power/ (PSW) System An exception to report OS-73.2, e ding the time The PSW System is designed as a standby system for use breaks from 30 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, has evaluate under emergency conditions. With the exception of testing of the system, the system is not normally pressurized.

Testing of the system is infrequent, typically every quarter.

3.6.2 References In addition, the duration of the test configuration is short,

1. Duke Power MDS Report No. OS-73.2, dated A compared to the total plant (unit) operating time. Due to the

2. combination of the infrequent testing and the short duration of the test, pipe ruptures are not postulated or evaluated for

2. Duke Power/B&W Report, Oconee Nucle the PSW System."

Environmental Effects on Non-Safety Grade Coi

3. USNRC Standard Review Plan (NUREG 0800) Section 3.6.2 Branch Technical Position MEB 3-1.

4. NRC Safety Evaluation of B&W Owners Group Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops, dated December 12, 1985.

5. NRC Safety Evaluation Relating To Elimination of Dynamic Effects of Postulated Primary Loop Pipe Ruptures from Design Basis in Regard to TMI- 1, dated November 5, 1987.

6. B&W Topical Report BAW-1847, Revision 1, "Leak-Before-Break Evaluation of Margin Against Full Break for RCS Primary Piping of B&W Designed NSS," September 1985.

7. B&W Topical Report BAW-1889P, "Piping Material Properties for Leak-Before-Break Analysis,"

October 1985.

8. B&W Topical Report BAW-1999, "TMI-1 Nuclear Power Plant Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping", April 1987.

Repac: "3.6.1.3 with 3.6.1.4" THIS IS THE LAST PAGE OF THE TEXT SECTION 3.6.

3.6- 2 (31 DEC 2012)