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8 to Updated Final Safety Analysis Report, Chapter 6, Engineered Safeguards
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Site:	Oconee
Issue date:	06/29/2020
From:	; Duke Energy Carolinas
To:	; Office of Nuclear Reactor Regulation
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Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - i Table of Contents 6.0 Engineered Safeguards 6.1 Engineered Safeguards 6.1.1 General Systems Description 6.1.2 Equipment Operability 6.1.3 Leakage and Radiation Considerations 6.1.4 Quality Control Standards 6.1.5 Piping Design Conditions 6.1.6 Engineered Safeguards Materials 6.1.7 References 6.2 Containment Systems 6.2.1 Containment Functional Design 6.2.1.1 Containment Structure 6.2.1.1.1 Design Bases 6.2.1.1.2 Design Features 6.2.1.1.3 Design Evaluation 6.2.1.2 Containment Subcompartments 6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents 6.2.1.3.1 Short-Term Mass and Energy Release Data 6.2.1.3.2 Long-Term Mass and Energy Release Date 6.2.1.3.3 Energy Sources 6.2.1.3.4 Description of Analytical Models 6.2.1.3.5 Single Failure Analysis 6.2.1.3.6 Metal-Water Reaction 6.2.1.4 Mass and Energy Release Analyses for Postulated Secondary System Pipe Ruptures Inside Containment 6.2.1.4.1 Mass and Energy Release Data 6.2.1.4.2 Single Failure Analysis 6.2.1.4.3 Initial Conditions 6.2.1.4.4 Description of Blowdown Model 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies on Emergency Core Cooling System 6.2.1.6 Coating Materials 6.2.2 Containment Heat Removal Systems 6.2.2.1 Design Bases 6.2.2.2

System Design

6.2.2.2.1 Piping and Instrumentation Diagrams 6.2.2.2.2 Codes and Standards 6.2.2.2.3 Materials Compatibility 6.2.2.2.4 Component Design 6.2.2.2.5 Reliability Considerations 6.2.2.2.6 Missile Protection 6.2.2.2.7 System Actuation 6.2.2.2.8 Environmental Considerations 6.2.2.2.9 Quality Control 6.2.2.3 Design Evaluation 6.2.2.4 Tests and Inspection 6.2.3 Containment Isolation System 6.2.3.1 Design Bases 6.2.3.2

System Design

6.2.3.3 Periodic Operability Tests

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - ii 6.2.4 Containment Leakage Testing 6.2.4.1 Periodic Leakage Testing 6.2.4.2 Continuous Leakage Monitoring 6.2.5 References 6.3 Emergency Core Cooling System 6.3.1 Design Bases 6.3.2

System Design

6.3.2.1 Schematic Piping and Instrumentation Diagrams 6.3.2.2 ECCS Operation 6.3.2.2.1 High Pressure Injection System 6.3.2.2.2 Low Pressure Injection System 6.3.2.2.3 Core Flooding System 6.3.2.3 Equipment and Component Descriptions 6.3.2.3.1 Piping 6.3.2.3.2 Pumps 6.3.2.3.3 Heat Exchangers 6.3.2.3.4 Valves 6.3.2.3.5 Coolant Storage 6.3.2.3.6 Pump Characteristics 6.3.2.3.7 Heat Exchanger Characteristics 6.3.2.3.8 Relief Valve Settings 6.3.2.3.9 Component Data 6.3.2.3.10 Quality Control 6.3.2.4 Applicable Codes and Classifications 6.3.2.5 Material Specifications and Compatibility 6.3.2.6 System Reliability 6.3.2.6.1 High Pressure Injection Operability 6.3.2.6.2 Core Flood Tank Valve Operability 6.3.2.6.3 Active Valve Operability 6.3.2.7 Protection Provisions 6.3.2.7.1 Seismic Design 6.3.2.7.2 Missile Protection 6.3.2.8 Post-Accident Environmental Consideration 6.3.3 Performance Evaluation 6.3.3.1 High Pressure Injection System (HPI) 6.3.3.2 Low Pressure Injection and Core Flooding Systems 6.3.3.2.1 Boron Precipitation Evaluation 6.3.3.3 Loss of Normal Power Source 6.3.3.4 Single Failure Assumption 6.3.4 Tests and Inspections 6.3.4.1 ECCS Performance Tests 6.3.4.2 Reliability Tests and Inspections 6.3.4.3 Gas Management 6.3.5 Instrumentation Requirements 6.3.6 References 6.4 Habitability Systems 6.4.1 Design Bases 6.4.2

System Design

6.4.2.1 Definition of Control Room Envelope 6.4.2.2 Ventilation System 6.4.2.3 Leak Tightness 6.4.2.4 Interaction With Other Zones and Pressure-Containing Equipment 6.4.2.5 Toxic Gas Protection 6.4.3 Testing and Inspection

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - iii 6.4.4 Instrumentation Requirements 6.4.5 References 6.5 Fission Product Removal and Control Systems 6.5.1 Engineered Safeguards (ES) Filter Systems 6.5.1.1 Design Bases 6.5.1.2

System Design

6.5.1.3 Design Evaluation 6.5.1.4 Tests and Inspections 6.5.1.5 Instrumentation Requirements 6.5.1.6 Materials 6.5.2 Containment Spray Systems 6.5.3 Deleted per 2001 Update 6.5.4 References 6.6 Inservice Inspection of Class 2 and 3 Components 6.6.1 Components Subject to Examination 6.6.2 Accessibility 6.6.3 Examination and Procedures 6.6.4 Inspection Intervals 6.6.5 Examination Categories and Requirements 6.6.6 Evaluation of Examination Results 6.6.7 System Pressure Tests 6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures

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Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - v List of Tables Table 6-1. Deleted per 1995 Update Table 6-2. Deleted per 2000 Update Table 6-3. Quality Control Standards for Engineered Safeguards Systems Table 6-4. Engineered Safeguards Piping Design Conditions Table 6-5. Single Failure Analysis Reactor Building Spray System Table 6-6. Single Failure Analysis For Reactor Building Cooling System Table 6-7. Reactor Building Penetration Valve Information Table 6-8. High Pressure Injection System Component Data Table 6-9. Low Pressure Injection System Component Data Table 6-10. Core Flooding System Components Data Table 6-11. Single Failure Analysis - Emergency Core Cooling System Table 6-12. Oconee Nuclear Station Analysis of Valve Motors Which May Become Submerged Following A LOCA Table 6-13. Equipment Operational During An Accident and Located Outside Containment Table 6-14. Equipment Operational During an Accident and Located Within the Containment Table 6-15. Emergency Core Cooling Systems Performance Testing Table 6-16. Deleted Per 1999 Update Table 6-17. Deleted Per 1999 Update Table 6-18. Inventory of Iodine Isotopes in Reactor Building (at t = o)

Table 6-19. Deleted Per 2015 Update Table 6-20. Parameters for Boron Precipitation Analysis Table 6-21. Summary of Calculated Containment Pressures and Temperatures for LOCA Cases Table 6-22. Containment Response Analyses Initial Conditions Table 6-23. Containment Structural Heat Sink Data Table 6-24. Accident Chronology for Limiting Break for Equipment Qualification Table 6-25. Minimum Acceptable Combinations of Containment Heat Removal Equipment Performance

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - vi Table 6-26. Engineered Safety Feature Assumptions in Containment Response Analyses Table 6-27. Summary of Calculated Containment Pressures and Temperatures for Secondary System Pipe Rupture Cases Table 6-28. Steam Generator Compartment Pressure Response Flowpath Discharge Coefficients Table 6-29. Peak Pressure Mass and Energy Release Data Table 6-30. Relap5 Long-Term Mass And Energy Release Data Table 6-31. BFLOW/FATHOMS Long-Term Mass and Energy Releases Table 6-32. Steam Line Break Mass and Energy Releases for Double-Ended Guillotine Break Table 6-33. NPSH Available and Required for LPI and BS Pumps (Limiting Flow Case)

Table 6-34. Deleted Per 2008 Update Table 6-35. Rotsg Peak Pressure Mass And Energy Release Data

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - vii List of Figures Figure 6-1. Flow Diagram of Emergency Core Cooling System Figure 6-2. Flow Diagram of Reactor Building Spray System Figure 6-3. Reactor Building Cooling Schematic Figure 6-4. Reactor Building Purge and Penetration Ventilation System Figure 6-5. Reactor Building Spray Pump Characteristics Figure 6-6. Reactor Building Cooler Heat Removal Capacity Figure 6-7. Reactor Building Cooler Heat Removal Capability as a Function of Air-Steam Mixture Flow Figure 6-8. Reactor Building Post-Accident Steam-Air Mixture Composition Figure 6-9. Reactor Building Isolation Valve Arrangements Figure 6-10. Deleted Per 1993 Update Figure 6-11. Deleted Per 1993 Update Figure 6-12. Deleted Per 1993 Update Figure 6-13. Deleted Per 1999 Update Figure 6-14. Deleted Per 1999 Update Figure 6-15. Deleted Per 1991 Update Figure 6-16. High Pressure Injection Pump Characteristics Figure 6-17. Low Pressure Injection Pump Characteristics Figure 6-18. Low Pressure Injection Cooler Capacity Figure 6-19. Control Rooms 1-2 And 3 Locations Figure 6-20. General Arrangement Control Room 1-2 Figure 6-21. General Arrangement Control Room 3 Figure 6-22. Penetration Room Ventilation Fan And System Characteristics Figure 6-23. Penetrations In Penetration Room 809'3" Floor And Wall Areas Figure 6-24. Penetrations In Penetration Room 838'0" Floor Figure 6-25. Penetration Rooms Details, Mechanical Openings Figure 6-26. Penetration Rooms Details, Electrical Openings

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - viii Figure 6-27. Penetration Rooms Details Construction Details Figure 6-28. ONS ROTSG Peak Pressure Analysis. 14.1 ft2 break - Rx Vessel Outlet Figure 6-29. ONS ROTSG Peak Pressure Analysis. 14.1 ft2 break - S/G Inlet Figure 6-30. ONS ROTSG Peak Pressure Analysis. 8.55 ft2 break - Cold Leg Pump Discharge Figure 6-31. ONS ROTSG Peak Pressure Analysis. 8.55 ft2 break - Cold Leg Pump Suction Figure 6-32. ONS ROTSG Peak Pressure Analysis. 14.1 ft2 break - Rx Vessel Outlet Figure 6-33. ONS ROTSG Peak Pressure Analysis. 14.1 ft2 break - S/G Inlet Figure 6-34. ONS ROTSG Peak Pressure Analysis. 8.55 ft2 break - Cold Leg Pump Discharge Figure 6-35. ONS ROTSG Peak Pressure Analysis. 8.55 ft2 break - Cold Leg Pump Suction Figure 6-36. Oconee Large Break LOCA Long-term Containment Response. Limiting Reactor Building Pressure Profile Figure 6-37. Oconee Large Break LOCA Long-term Containment Response. Limiting Vapor Temperature Profile Figure 6-38. Deleted Per 2003 Update Figure 6-39. Deleted Per 2003 Update Figure 6-40. Deleted Per 2003 Update Figure 6-41. Deleted Per 2003 Update Figure 6-42. Oconee Steam Line Break:Containment Pressure. With Automatic MFW Isolation Figure 6-43. Oconee Steam Line Break:Containment Temperature. With Automatic MFW Isolation Figure 6-44. LOCA-Mass Release for the Subcompartment Pressure Response Analysis Figure 6-45. LOCA-Energy Release Rate for the Subcompartment Pressure Response Analysis Figure 6-46. LOCA-Reactor Compartment Pressure Response Figure 6-47. LOCA-Steam Generator Compartment Vent Discharge Coefficient Figure 6-48. LOCA-Steam Generator Compartment Pressure Response Figure 6-49. Deleted Per 2003 Update Figure 6-50. LOCA-Mass Released to the Reactor Building. For the 8.55 ft2 Cold Leg Pump Discharge Break Figure 6-51. LOCA-Energy Released to the Reactor Building. For the 8.55 ft2 Cold Leg Pump Discharge Break

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6 - ix Figure 6-52. LOCA-Reactor Building Pressure. For the 8,55 ft2 Cold Leg Pump Discharge Break Figure 6-53. Deleted Per 1997 Update

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Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.0 - 1 6.0 Engineered Safeguards THIS IS THE LAST PAGE OF THE TEXT SECTION 6.0.

UFSAR Chapter 6 Oconee Nuclear Station 6.0 - 2 (31 DEC 2019)

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Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.1 - 1 6.1 Engineered Safeguards Engineered safeguards are those systems and components designed to function under accident conditions to prevent or minimize the severity of an accident or to mitigate the consequences of an accident. During accident conditions when reactor coolant is lost, or in the event of secondary system pipe breaks, the engineered safeguards act to provide emergency cooling to assure structural integrity of the core, to maintain the integrity of the Reactor Building, and to collect and filter Potential Reactor Building penetration leakage. Separate and independent engineered safeguards are provided for each of the three reactor units at Oconee. Special precautions are taken to assure high quality in the system design and components.

The engineered safeguards include provisions for:

a. High pressure injection.

b. Low pressure injection.

c. Core flooding.

d. Two types of Reactor Building cooling.

e. The collection and control of Reactor Building penetration leakage.

f. Reactor Building isolation.

Figure 6-1 and Figure 6-4 depict the portion of the Engineered Safeguards System related to core and building protection (see (a) through (d) above). A general description of the engineered safeguards provisions is presented below and a more detailed description is presented in the latter portion of this section. Since each reactor unit has the same arrangement of Emergency Safeguard Systems, the performance of the systems is described on a unit basis.

6.1.1 General Systems Description The High and Low Pressure Injection Systems and the Core Flooding Tanks are designed to form collectively an overall Emergency Core Cooling System (ECCS), which is designed to prevent melting or physical disarrangement of the core over the entire spectrum of Reactor Coolant System break sizes. Figure 6-1 shows the Emergency Core Cooling Systems for one reactor unit. The High Pressure Injection System is arranged so that three pumps are available for emergency use. The Low Pressure Injection System is arranged to assure that two pumps are normally available and a third pump is installed but normally valved off. The Core Flooding System for each unit is composed of two separate pressurized tanks containing borated water at Reactor Building ambient temperature. These tanks automatically discharge their contents into the reactor vessel at a preset Reactor Coolant System pressure without reliance on any actuating signal, any electrical power or any external actuated component.

Reactor Building integrity is assured by two pressure reducing systems operating on different principles; the Reactor Building Spray System and the Reactor Building Emergency Cooling System. (Refer to Figure 6-2 and Figure 6-3). These systems have the redundancy required to meet the single failure criterion. These systems operate to lower Reactor Building pressure over the spectrum of Reactor Coolant System break sizes and to reduce the driving force for leakage of radioactive materials from the Reactor Building. They also serve to reduce Reactor Building pressure and temperature in the event of a main steam line break.

The Reactor Building Penetration Room Ventilation System shown on Figure 6-4 collects and filters air leakage to control and minimize the release of radioactive material from Reactor Building penetrations following an accident. Two full capacity filtering paths are provided. This

UFSAR Chapter 6 Oconee Nuclear Station 6.1 - 2 (31 DEC 2019) system is no longer required due to adoption of the alternate source term. The system is not credited for event mitigation and serves an ALARA function only.

6.1.2 Equipment Operability Operability of engineered safeguards equipment is assured in several ways. Much of the equipment in these systems serves a function during normal reactor operation. In those cases where equipment is used for emergency functions only such as the Reactor Building Spray System, systems have been designed to permit meaningful periodic tests. Operational reliability is achieved by using proven component design, and by conducting tests where either the component or its application was considered unique. In-house quality control procedures are imposed on the components of the Engineered Safeguards Systems. These procedures include use of accepted codes and standards as well as supplementary test and inspection requirements to assure that all components will perform their intended function under the design conditions following a loss-of-coolant accident. See also Section 6.3.4.3 for additional considerations related to Engineered Safeguards system operability.

The purpose of this section is to describe the physical arrangement, design, and operation of the Engineered Safeguards Systems as related to their safety function.

Reactor Building isolation is described in Section 6.2. Other sections of the report contain information which is pertinent to the Engineered Safeguards Systems. Chapter 7 describes the actuation instrumentation of these systems. Chapter 15 describes the analysis of the Engineered Safeguards Systems' capability to provide adequate protection during accident conditions. Chapter 9 discusses functions performed by these systems during normal operation and gives further design details and descriptive information concerning those systems.

6.1.3 Leakage and Radiation Considerations The use of normally operating equipment for engineered safeguards functions and location of some of this equipment outside the Reactor Building require that consideration be given to direct radiation levels after fission products have accumulated in these systems and leakage from these systems.

The shielding for components of the engineered safeguards is designed to meet the following objectives in the event of a maximum hypothetical accident:

a. To provide protection for personnel to perform all operations necessary for mitigation of the accident.

b. To provide sufficient accessibility in all areas around the station to permit safe continued operation of the unaffected nuclear units.

Summary of Post-Accident Recirculation Following a loss-of-coolant accident, flow is initiated in the Low Pressure Injection System from the borated water storage tank to the reactor vessel. Flow is also initiated by the Reactor Building Spray Systems to building spray headers. When most of the borated water storage tank inventory is exhausted, the operators initiate steps to transfer the pump suction to the Reactor Building emergency sump for both the reactor core cooling flow and the Reactor Building sprays. System resistance will maintain Reactor Building Spray pumps flow rates between 700 and 1200 gpm to ensure adequate NPSH during injection and recirculation modes, with no throttling required. The post-accident recirculation system includes all piping and

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.1 - 3 equipment both internal and external to the Reactor Building as shown on Figure 6-1, up to the stop and test line valves leading to the borated water storage tank.

The NPSH available to the Low Pressure Injection and Reactor Building Spray pumps during the post-LOCA recirculation phase has been calculated based on:

a. "As Built" piping drawings.

b. Pipe and fitting losses calculated using the information in Crane Technical Paper No.

410, Cameron Hydraulic Data, and Idelchik Handbook of Hydraulic Resistance.

c. Total flow in a single string (i.e., consisting of one Low Pressure Injection Pump and one Reactor Building Spray pump served by a single sump suction line) is determined through hydraulic analysis.

d. BS pump flow is allowed to vary while LPI flow is maximized. Resistances have been added to the BS system piping so that flow rates are limited to a range of 700-1200 gpm.

This provides improvements in NPSH available, and eliminates the requirements for throttling the flow to maintain pump NPSH. The system hydraulic analysis reflect those additional system resistances and the elimination of pump throttling.

e. Sump water temperatures and Reactor Building pressures were determined from analysis of hot leg break with conservative building cooling assumptions (LPSW temperature, RBCU capacity, etc.).

f. Reactor Building Spray pump shaft center line at elevation 760 ft. 1 in.

g. Low Pressure Injection pump shaft center line at elevation 761 ft. 1 in.

h. Water level in the Reactor Building sump is at elevation 781.28 ft. based on the following assumptions: (height above Reactor Building basement level is 4.4 ft.)

1) The Technical Specification minimum initial levels were used for the BWST and the CFT's, with six feet of level remaining in the BWST at completion of switchover to RBES.

2) Some water is maintained in the Reactor Building atmosphere as vapor. The quantity was determined using the results of a FATHOMS Computer Run for a 14.1 ft2 break.

3) The break is conservatively assumed to occur at the top of the hot leg, thereby keeping the Reactor Coolant System full.

i.

Unit-specific and train-specific models were constructed to maximize the friction losses in the LPI and BS pump suction piping.

Results of the NPSH analysis are presented in Table 6-33. Credit is taken for 0.44 psi of reactor building overpressure in the calculation of available NPSH for the RBS and LPI pumps from approximately 3,000 seconds to approximately 30,000 seconds post-accident.

Available NPSH has been determined to meet or exceed the required NPSH for worst case accident conditions with conservative inputs as identified above. Curves of total dynamic head and NPSH versus flow are shown in Figure 6-5 for the Reactor Building Spray Pumps and in Figure 6-17 for the Low Pressure Injection Pumps. These curves are representative in nature and are provided for information only. They are not intended to constitute design commitments or performance requirements for the pumps. Refer to the Inservice Test Program for actual performance requirements for BS and LPI pumps.

UFSAR Chapter 6 Oconee Nuclear Station 6.1 - 4 (31 DEC 2019)

The NRC issued Generic Letter 97-04, "Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps," on October 7, 1997, requesting that licensees submit information necessary to confirm the adequacy of the net positive suction head (NPSH) available for emergency core cooling (including core spray and decay heat removal) and containment heat removal pumps. A review of the current design-basis analyses used to determine the available NPSH for the applicable pumps was performed and report submitted by the letter from M. S. Tuckman to the NRC, dated January 5, 1998 (Reference 2 ). This report concluded that the available NPSH for these pumps is adequate under all design-basis accident scenarios. As a result of a subsequent review by Duke, the NPSH calculations were determined to be outside the design basis as discussed in a letter from W. R. McCollum to the NRC, dated Septemer 17, 1998 (Reference 3). A License Admendment was submitted to revise the UFSAR to be consistent with the design basis calculations. License Amendments 305/305/305 approving incorporations of the UFSAR changes to the NPSH bases were issued in a letter from D. E. LeBarge (NRC) to W. R. McCollum (Duke), dated July 19, 1999 (Reference 4).

Bases of Leakage Estimates While the reactor auxiliary systems involved in the recirculation complex are closed to the Auxiliary Building atmosphere, leakage is possible through component flanges, seals, instrumentation, and valves.

The leakage sources considered are:

a. Valves

1) Disc leakage when valve is on recirculation system boundary.

2) Stem leakage.

3) Bonnet flange leakage.

b. Flanges

c. Pump shaft seals While leakage rates have been assumed for these sources, maintenance and periodic testing of these systems will preclude all but a small percentage of the assumed amounts. With the exception of the boundary valve discs, all of the potential leakage paths may be examined during periodic tests or normal operation. Boundary valves which have been identified to have leakage paths are tested periodically. All other valve disc leakage is retained in the other closed systems and, therefore, will not be released to the Auxiliary Building.

While valve stem leakage has been assumed for all valves, a number of manual valves in the recirculation complex are backseating and do not rely on packing alone to prevent stem leakage.

Leakage Assumptions Source Quantities

a.

Valves - Process

1.

Disc Leakage 10 cc/hr./in. of nominal disc diameter

2.

Stem leakage 1 drop/min.

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.1 - 5 Source Quantities

3.

Bonnet flange 10 drops/min.

b.

Valves - Instrumentation Bonnet flange and stem 1 drop/min.

c.

Flanges 10 drops/min.

d.

Pump seals 50 drops/min.

For the analysis, it was assumed that the water leaving the Reactor Building was at 252 °F. This assumption is conservative as this peak temperature would only exist for a short period during the post-accident condition. Water downstream of the coolers was assumed to be 115 °F. The Auxiliary Building was assumed to be at 70 °F and 30 percent relative humidity. Under these conditions, approximately 22 percent of the leakage upstream of the coolers and 4 percent of the leakage downstream of the coolers would flash into vapor. For the analysis, however, it was assumed that 50 percent of the leakage upstream of the coolers would become vapor because of additional heat transfer from the hot metal.

Design Basis Leakage The design basis leakage for the LPI, HPI, and BS systems is 12 gallons per hour.

Leakage Analysis Conclusions It was concluded from analysis of the 12 gph limit (in conjunction with the discussion and analysis in Section 15.15.4) that leakage from Engineered Safeguards Systems outside the Reactor Building does not pose a public safety problem.

6.1.4 Quality Control Standards Quality Control Standards for the Engineered Safeguards Systems are listed in Table 6-3.

6.1.5 Piping Design Conditions Piping Design Conditions for the Engineered Safeguards Systems are listed on Table 6-4.

6.1.6 Engineered Safeguards Materials Materials used in Engineered Safeguards components are addressed in applicable sections where appropriate.

6.1.7 References

1. Nuclear Regulatory Commission, Letter to Holders of Operating Licenses for Nuclear Power Plants, Except Those Who Have Permanently Ceased Operations and Have Certified That Fuel Has Been Permanently Removed for the Reactor Vessel, from Jack W. Roe, October 7, 1997, "Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps (Generic Letter 97-04)."

UFSAR Chapter 6 Oconee Nuclear Station 6.1 - 6 (31 DEC 2019)

2. Duke Power Company, Letter from M. S. Tuckman to the NRC, January 5, 1998, re:

Response to Generic Letter 97-04, "Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps."

3. Duke Power Company, Letter from W. R. McCollum to the NRC, dated September 17, 1998. Re: Response to Request for Additional Information Related to Generic Letter 97-04, dated August 11, 1998.

4. D. E. LeBarge (NRC) to W. R. McCollum (Duke), dated July 19, 1999, re: Issuance of Amendments (License Amendments 305/305/305 for NPSH changes related to Generic Letter 97-04).

THIS IS THE LAST PAGE OF THE TEXT SECTION 6.1.

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.2 - 1 6.2 Containment Systems 6.2.1 Containment Functional Design 6.2.1.1 Containment Structure 6.2.1.1.1 Design Bases The Reactor Building completely encloses the Reactor Coolant System to minimize release of radioactive material to the environment should a serious failure of the Reactor Coolant System occur. The structure provides adequate biological shielding for both normal operation and accident situations. The Reactor Building is designed for an internal pressure of 59 psig. The leakage rate will not exceed 0.2 percent by volume in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months under the conditions of the maximum hypothetical accident as described below.

The Reactor Building is designed for an external pressure 3.0 psi greater than the internal pressure. The design external pressure of 3.0 psi corresponds to a margin of 0.5 psi above the differential pressure that could be developed if the building is sealed with an internal temperature of 120°F with a barometric pressure of 29.0 inches of Hg and the building is subsequently cooled to an internal temperature of 80°F with a concurrent rise in barometric pressure to 31.0 inches of Hg. The weather conditions assumed here are conservative since an evaluation of National Weather Service records for this area indicates that from 1918 to 1970 the lowest barometric pressure recorded is 29.05 inches of Hg and the highest of 30.85 inches of Hg.

The principal design basis for the structure is that it be capable of withstanding the internal pressure resulting from a loss-of-coolant accident or a secondary line rupture with no loss of integrity. In a LOCA, the total energy contained in the water of the Reactor Coolant System is assumed to be released into the Reactor Building through a break in the reactor coolant piping.

In a secondary line break event the energy contained in the water in the secondary coolant system, as well as energy transferred across the steam generator tubes from the Reactor Coolant System is assumed to be released into the Reactor Building through a break in the steam line piping. However, in the case of a secondary line break, the release of energy essentially stops when the faulted steam generator empties and is no longer being supplied with feedwater. In either case, subsequent pressure behavior is determined by the building volume, engineered safeguards, and the combined influence of energy source and heat sinks.

Energy is available for release into the containment structure from the following sources:

LOCA Secondary Line Break Reactor Coolant Stored Energy Secondary Coolant Stored Energy Reactor Stored Energy Secondary System Stored Energy Reactor Decay Heat Reactor Coolant Stored Energy Metal-Water Reactions Reactor Stored Energy Secondary Coolant Stored Energy Reactor Decay Heat Secondary System Stored Energy

UFSAR Chapter 6 Oconee Nuclear Station 6.2 - 2 (31 DEC 2019) 6.2.1.1.2 Design Features Since the design of the Engineered Safeguards Systems and their operation is discussed more fully in Section 6.3, only their relation to the basis of Reactor Building design is discussed below.

The Engineered Safeguards Systems are provided to limit the consequences of an accident.

Their energy removal capabilities limit the internal pressure after the initial peak so that Reactor Building design limits are not exceeded and the potential for release of fission products is minimized.

Following a LOCA, the Emergency Core Cooling Systems inject borated water into the Reactor Coolant System to remove core decay heat and to minimize metal-water reactions and the associated release of heat and fission products. Flashed primary coolant, Reactor Coolant System sensible heat, and core decay heat transferred to Reactor Building are removed by two engineered safeguards systems: the Reactor Building Spray and/or the Reactor Building Cooling Systems.

Following a secondary line break at power, main feedwater and turbine-driven emergency feedwater flow to the faulted steam generator are isolated by the Automatic Feedwater Isolation System (AFIS) on low steam line pressure. AFIS will also isolate motor-driven emergency feedwater to the affected steam generator on a high rate of steam line depressurization concurrent with low steam line pressure. For break sizes that do not exceed the AFIS rate of depressurization setpoint, manual operator action is credited at 10 minutes to isolate motor-driven emergency feedwater flow to the affected steam generator. Section 6.2.1.4.4 discusses AFIS actions during secondary line blowdown into containment. Section 7.9 provides a detailed description of AFIS operation.

The Reactor Building Spray System removes heat directly from the Reactor Building atmosphere by cold water quenching of the Reactor Building steam.

The air recirculation and cooling systems remove heat directly from the Reactor Building atmosphere to the Service Water System with recirculating fans and cooling coils.

The low pressure injection coolers remove heat from the containment sump liquid to the Service Water System with heat exchange through tubes.

Section 3.8 provides a detailed description of the Reactor Building design.

6.2.1.1.3 Design Evaluation 6.2.1.1.3.1 LOCA Short Term Containment Pressure Response This section provides analyses of the short-term (3 minutes) pressure response of the containment to a spectrum of postulated Reactor Coolant System pipe ruptures. The analyses results provide the bounding post-LOCA containment responses with respect to the containment design pressure, with all input assumptions and boundary conditions chosen to provide a conservatively high containment pressure per the methodology presented in Reference 1. The break size and location of each postulated loss of coolant accident is given in Table 6-21. The pressure and temperature response of the four break location sensitivity studies, Cases 1A through 1D, are given in the following figures:

Figure 6-28 Containment pressure for a 14.1 ft2 break at the reactor vessel outlet (1A)

Figure 6-29 Containment pressure for a 14.1 ft2 break at the steam generator inlet (1B)

Figure 6-30 Containment pressure for a 8.55 ft2 break at the RCP discharge (1C)

Figure 6-31 Containment pressure for a 8.55 ft2 break at the RCP suction (1D)

Figure 6-32 Containment temperature for a 14.1 ft2 break at the reactor vessel outlet (1A)

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.2 - 3 Figure 6-33 Containment temperature for a 14.1 ft2 break at the steam generator inlet (1B)

Figure 6-34 Containment temperature for a 8.55 ft2 break at the RCP discharge (1C)

Figure 6-35 Containment temperature for a 8.55 ft2 break at the RCP suction (1D)

Analysis Method and Computer Codes The analysis method used in this section is described in Reference 1. The computer codes used in this section are RELAP5/MOD2-B&W (Reference 2) for calculating the mass and energy releases and FATHOMS (Reference 3) for calculating the containment pressure and temperature response.

Mass and Energy Release Rate Data The mass and energy release rate data used for the LOCA analyses described in this section are given in Section 6.2.1.3.

Initial Condition Assumption Conservatisms Initial condition assumptions in the LOCA containment peak pressure response analyses are adjusted to give a conservative answer:

1. The initial pressure assumption is equal to the upper Technical Specification limit.

Instrument uncertainty is taken into account through operation of the Reactor Building purge, before this upper limit is reached.

2. The initial temperature assumption is conservatively low for full power operation. This maximizes the initial containment air mass, which maximizes the air partial pressure contribution to the pressure peak.

3. The nominal containment free volume is reduced by 2% from the Table 6-22 value.

4. A low initial relative humidity is used to maximize the initial air mass.

The initial conditions used are tabulated in Table 6-22.

Containment Heat Removal Systems No credit is taken in the LOCA peak pressure analysis for either the Reactor Building cooling units or the Reactor Building Spray System. The peak pressure occurs within the first 20 seconds after the postulated break, prior to the assumed actuation of either of these heat removal systems.

Emergency Core Cooling Systems The emergency core cooling systems are not explicitly modeled in FATHOMS for the LOCA peak pressure analysis, but are considered in the mass and energy releases discussed in Section 6.2.1.3.

Single Failure A component single failure generally has little impact on the peak pressure analysis. This is because peak pressures usually occur before the engineered safeguards equipment has time to activate and become effective.

Structural Heat Sinks The structural heat sinks within containment are divided into nine groups for the purposes of containment pressure and temperature response modeling. These nine structures are tabulated in Table 6-23. The concrete and steel portions of the building cylinder, the building dome, and the building base are combined in three structures of two materials each.

UFSAR Chapter 6 Oconee Nuclear Station 6.2 - 4 (31 DEC 2019) 6.2.1.1.3.2 LOCA Long-Term Containment Temperature Response This section provides analyses of the long-term (>1 day) temperature response of the containment to a spectrum of postulated Reactor Coolant System pipe ruptures. The analyses results provides the bounding long-term containment temperature response for use in Equipment Qualification (EQ) evaluation of equipment within the Reactor Building. The bounding post-LOCA containment pressure responses are discussed in Section 6.2.1.1.3.1, while the bounding short-term containment temperature results are discussed in Section 6.2.1.1.3.3. The long-term large break containment analysis considers only a single break size and location: a double-ended guillotine break located at the A1 cold leg pump discharge. There is no need to analyze a spectrum of large break locations since a suitably bounding site can be chosen by inspection. The qualitative bases for this position is explained in the following paragraphs. Reference 1 also extensively analyzed the mass and energy releases from and containment response to small break LOCAs. The conclusion of these analyses is that small break LOCAs are not more limiting than large break LOCAs with respect to challenging the containment equipment qualification acceptance criteria.

The basis for choosing a cold leg break as opposed to a hot leg break is obvious once the characteristics of each break are considered. Although it is true that an identical quantity of decay heat will be generated regardless of the break location, the manner in which this energy is partitioned between the vapor and liquid break flow streams is the dominant consideration.

Because the long-term containment response is concerned with temperature in containment as a function of time, it is expected that an energy release profile which is dominated by steam relief will generate a more severe containment response. This is because steam relief to the atmosphere will have a greater impact on containment temperature than if the energy is released primarily in the liquid phase, which has only a slight interaction with the containment atmosphere (convection at the pool surface). Indeed, this observation has been validated with the FATHOMS computer code in numerous analyses. It might appear that the Reactor Building Spray System acts to homogenize the containment atmosphere such that the phase in which the energy is released is insignificant. However, when the complicated interactions between the equipment used to cool the containment atmosphere (Reactor Building coolers and sprays) and the equipment used to cool the containment sump (LPI coolers) are examined by analysis, it is apparent that containment will never become completely homogenized. Therefore, the partitioning of energy released to containment between the vapor and liquid phases is the dominant factor in the long-term containment response.

Due to the geometry of a B&W reactor system, steaming from a cold leg break location will never become completely suppressed. This means that steam will always exit the break no matter how much the decay heat power drops. In contrast, it is possible to completely suppress steaming from a hot leg break site as decay power decreases. Decay heat will eventually be absorbed as sensible heat by the injection fluid and thus steaming from the break will cease.

Naturally, when this occurs, decay heat will be transferred to containment in the liquid phase, resulting in a less severe containment response.

The cold leg pump discharge break location is selected rather than the pump suction location.

For a pump suction break the cold HPI fluid injected into the broken cold leg pump discharge piping will interact with steam exiting the core through the vent valves and condense a large portion of this steam before it reaches the break. Thus, the steam release will be less for a pump suction break. Consequently, the pump discharge break location is limiting.

An accident chronology is presented in Table 6-24 for the most limiting LOCA, an 8.55 ft2 double-ended guillotine cold leg break at the reactor coolant pump outlet. Table 6-25 presents

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.2 - 5 results for various combinations of initial Reactor Building temperature, LPSW temperature and RBCU heat removal rate. The results of the limiting case are presented in the following figures:

Figure 6-36 Containment pressure Figure 6-37 Containment atmosphere temperature Deleted row(s) per 2003 update These cases represent the combinations of minimum and maximum LPSW temperature and RBCU heat removal rate. All cases assume LPI cooler heat removal performance which conservatively bounds that for the coolers at all three Oconee units. The equipment qualification criteria for all equipment required for LOCA mitigation are documented in the Oconee Environmental Qualification Criteria Manual, referred to UFSAR Section 3.11. These criteria are met for all cases analyzed.

Analysis Method and Computer Codes The analysis method used in this section is described in Reference 1. The computer codes used in this section are RELAP5/MOD2-B&W (Reference 2) for calculating the mass and energy releases during the first 30 minutes, BFLOW for calculating the longer term LOCA mass and energy releases, and GOTHIC (Reference 17) for calculating the containment pressure and temperature response.

Mass and Energy Release Rate Data The mass and energy release rate data used for the LOCA analyses described in this section are given in Section 6.2.1.3.

Initial Condition Assumption Conservatisms Initial condition assumptions in the containment response analyses are adjusted to give a conservative answer:

1. A nominal initial pressure is used, although this parameter has very little effect due to the long duration of this analysis.

2. The initial temperature assumption is conservatively high for full power operation.

3. The nominal containment free volume is reduced by 2% from the Table 6-22 value.

4. A high initial relative humidity is used, although this parameter has very little effect due to the long duration of this analysis.

The initial conditions used are tabulated in Table 6-22 Containment Heat Removal Systems The Reactor Building Cooling Units (RBCUs) are modeled based on performance at design conditions. This reference is based on the heat removal rate associated with:

1. Low Pressure Service Water (LPSW) temperature of 75°F

2. Containment air temperature of 286°F

3. Containment air/stream mixture characteristic of 70°F and 14.7 psia initial condition

4. Entering air flow rate of 54,000 acfm The performance of the two operating coolers (refer to single failure discussion below) is adjusted for the safety analysis to reflect different air/stream mixtures based on the higher containment initial temperatures and higher LPSW temperatures assumed in the safety analysis. The RBCU performance requirements are adjusted back to the reference performance conditions above to compare with periodic cooler performance test results.

UFSAR Chapter 6 Oconee Nuclear Station 6.2 - 6 (31 DEC 2019)

Low Pressure Injection (LPI) Cooler test data at various flow rates are used to determine the relationships between cooler degradation (number of plugged tubes and amount of tube surface fouling) and thermal performance parameters such as fluid flow rates and temperatures. These relationships are then modeled to determine an LPI Cooler overall heat transfer coefficient as a function of LPI inlet temperature. Since assumed cooler degradation, fluid flow rates, and LPSW temperature are constant during a simulation, and since LPI inlet temperature changes during the accident, this change determines the LPI cooler heat removal rate for a particular case. No credit is taken for heat removal by the LPI coolers during the injection phase.

Calculations using hydraulic models of the Oconee Reactor Building Spray (RBS) System result in a minimum flow rate of 750 gpm during the injection phase, taking suction from the Borated Water Storage Tank (BWST). Likewise, minimum flow rates of 900 gpm during the sump recirculation phase (taking suction from the Reactor Building Emergency Sump (RBES)) are demonstrated. In the FATHOMS containment response analyses, the RBS flow rate during injection mode is conservatively assumed to be 700 gpm. During sump recirculation mode, the RBS flow rate is assumed to be 900 gpm. The injection phase temperature used is a conservatively high for the borated water storage tank, the source of RBS water during the injection phase. The recirculation phase RBS temperature is the sump temperature calculated by FATHOMS. No credit is taken for aligning the RBS pumps to take suction from the outlet of the LPI coolers.

Assumed values for containment heat removal equipment performance parameters are given in Table 6-26.

Emergency Core Cooling Systems The single operating Low Pressure Injection (LPI) pump (refer to single failure discussion below) is assumed in the FATHOMS computer code to be supplying a conservatively low flow rate to the reactor vessel. The injection phase temperature used is a conservatively high for the borated water storage tank, the source of LPI water during the injection phase. The recirculation phase temperature is calculated by FATHOMS based on the heat removal from the LPI coolers.

The two operating High Pressure Injection (HPI) pumps (refer to single failure discussion below) are assumed to be supplying a conservatively low flow rate to the cold legs. The HPI water injected into the broken cold leg is added directly to the containment sump. The injection phase temperature used is a conservatively high for the borated water storage tank, the source of HPI water during the injection phase. No credit is taken for HPI flow during the recirculation phase.

Liquid injection from the core flood tanks is not explicitly modeled in FATHOMS but is considered in the mass and energy releases discussed in Section 6.2.1.3. The nitrogen cover gas from these tanks is assumed to be injected to increase the containment pressure calculated by FATHOMS. The amount of injected nitrogen is based on the mass which would be present at the pressure and temperature initial conditions of the mass and energy release calculation.

Assumed values for ECCS equipment performance parameters are given in Table 6-26.

Single Failure While a component single failure generally has little impact on the peak pressure analysis described in the previous section, it has a much greater impact on the long-term containment response. The most limiting single failure is therefore chosen to yield a conservative long-term containment response. The most restrictive single failure is chosen as the one which disables the greatest number of containment heat removal components.

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.2 - 7 An evaluation was performed to determine the most limiting single failure with respect to containment cooling. This evaluation indicated that the failure of a 4160V switchgear represents the most limiting single failure. Electrical switchgear power a myriad of safety related equipment including injection systems and containment cooling systems. The failure of one of the three available switchgear will result in the loss of the following components:

1. One HPI pump

2. One LPI pump

3. One RBS pump

4. One RBCU All other ECCS equipment is available following a nominal, transient-specific actuation delay.

The switchgear failure is more limiting than a loss of offsite power (LOOP) and failure of one Keowee hydroelectric unit because the second hydroelectric unit is available to power the standby busses through CT-4 (underground) or through the switchyard (overhead). Therefore, all ECCS equipment would be available after a small time delay.

Structural Heat Sinks The structural heat sinks within containment are those described in Section 6.2.1.1.3.1 and tabulated in Table 6-23. For the LOCA long-term containment response calculation the surface areas of these heat structures are reduced by 1% for conservatism.

6.2.1.1.3.3 Steam Line Break Containment Pressure and Temperature Response This section provides analyses of the pressure and temperature response of the containment to postulated secondary system pipe ruptures. A spectrum of break sizes is analyzed to determine the limiting break size for peak containment pressure and temperature. For peak containment pressure, the response depends mainly on the steam mass flow rate. The limiting break size for peak containment pressure and temperature is the double-ended guillotine break (12.6 ft2). This break size results in the highest initial rate of mass and energy release to the containment and thus maximizes the increase in containment pressure and temperature during the steam line break transient.

The results of the limiting case are given in Table 6-27. The pressure and temperature response of these limiting cases are given in Figure 6-42 (containment pressure) and Figure 6-43 (containment temperature). The period of time during which the calculated temperature exceeds the equipment qualification limit is very short compared to the time that the equipment is exposed to high temperatures during its qualification testing. This short duration of calculated temperatures above the equipment qualification limit is not long enough to cause the equipment internal temperatures to reach values as high as those reached during the qualification testing.

Analysis Method and Computer Codes The analysis method used in this section is described in Reference 1. The computer codes used in this section are RETRAN-3D (Reference 9) for calculating the steam line break mass and energy releases and FATHOMS (Reference 3) for calculating the containment pressure and temperature response.

Mass and Energy Release Rate Data The mass and energy release rate data used for the steam line break analyses described in this section are given in Section 6.2.1.4.

UFSAR Chapter 6 Oconee Nuclear Station 6.2 - 8 (31 DEC 2019)

Initial Condition Assumption Conservatisms Initial condition assumptions in the containment response analyses are adjusted to give a conservative answer:

1. The initial pressure assumption is equal to the upper Technical Specification limit for cases in which high initial pressure is conservative.

2. The initial temperature assumption is conservatively high for full power operation. It is known from the LOCA analyses described in the previous section that a lower initial temperature maximizes the containment peak pressures due to a higher initial air mass.

However, a higher initial temperature reduces the cooling capacity of the structural heat sinks, outweighing the impact of a higher initial air mass.

3. The nominal containment free volume is reduced by 2% from the Table 6-22 value.

4. A low initial relative humidity is used to maximize the initial air mass.

The initial conditions used are tabulated in Table 6-22.

Containment Heat Removal Systems The Reactor Building cooling units (RBCUs) are modeled as described in Section 6.2.1.1.3.2.

The steam line break peak pressure is reached long before the borated water storage tank empties. Therefore the recirculation phase is not simulated, and no credit is taken for heat removal by the LPI coolers during the injection phase. The RBS is modeled as described in Section 6.2.1.1.3.2 for the injection phase.

Single Failure The assumed single failure is the same as discussed above for LOCA, the failure of a 4160 V switchgear, resulting in the loss of one HPI pump, one LPI pump, one RBS pump and one RBCU.

Structural Heat Sinks The structural heat sinks within containment are those described in Section 6.2.1.1.3.1 and tabulated in Table 6-23. For the steam line break containment response calculation the surface areas of these heat structures are reduced by 1% for conservatism.

6.2.1.1.3.4 Functional Capability of Normal Containment Ventilation Systems Normal containment ventilation is provided by four Reactor Building auxiliary cooling units (RBACUs) and two of the three RBCUs. The function of these units during normal operation is described in Section 9.4.6. Upper and lower limits on containment pressure during normal operation are maintained by complying with the Technical Specifications.

6.2.1.1.3.5 Post-Accident Monitoring of Containment Conditions Post-accident monitoring instrumentation is provided for the following containment parameters:

Reactor Building pressure Reactor Building air temperature Reactor Building normal sump level Reactor Building emergency sump level Reactor Building wide range sump level Reactor Building hydrogen concentration Section 7.5 discusses the range, accuracy, and response of this instrumentation and the tests conducted to qualify the instruments for use in the post-accident containment environment.

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.2 - 9 6.2.1.2 Containment Subcompartments The pressure response of the Reactor Building subcompartments following the design basis LOCA has been evaluated using mass and energy release rates calculated by the CRAFT code (Reference 4) using the system model in Reference 5, with the pressure response calculated by the COPRA code (Reference 6). The Reactor Building subcompartments include the reactor compartment and the east and west steam generator compartments. For each compartment the worst case LOCA break size and location is identified, including the effect of piping restraints on the maximum break size. The flow through the subcompartment vents is calculated using a sonic choking model for a homogeneous steam-water-air mixture, with a vent discharge coefficient of 0.6. A discharge coefficient of 1.0 is used for the system blowdown calculation.

The reactor compartment has a volume of 5520 ft3, one 6 ft2 vent flowpath, and concrete shield plugs with a total flow area of 69 ft2. Only the vent flowpath is assumed to be available for pressure relief. Although the maximum break area within the compartment has been determined to be 3.0 ft2, hot leg breaks of 8.0, 5.0, and 3.0 ft2 were analyzed, as well as the maximum cold leg break of 8.55 ft2. The CRAFT mass and energy release rates are given in Figure 6-44 and Figure 6-45. The resulting pressure differential across the compartment walls are shown in Figure 6-46. The peak pressure of 160 psi, which occurs for the 8.0 ft2 hot leg break, is only 78 percent of the design differential pressure of 205 psi.

The west steam generator compartment has a volume of 61,700 ft3 and a total vent flow area of 1333 ft2. The east compartment has a volume of 60,400 ft3 and a flow area of 1222 ft2. The discharge coefficients for each of the flowpaths and the effective discharge coefficient calculated to result in the correct choked flow are given in Table 6-28 and Figure 6-47. The maximum hot leg break of 14.1 ft2 was analyzed using the CRAFT mass and energy release rates in Figure 6-44 and Figure 6-45. The resulting pressure differentials across the compartment walls are shown in Figure 6-48. The structural integrity of the compartments is sufficient to withstand 130 percent of the peak differential pressure of 15 psi.

6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents 6.2.1.3.1 Short-Term Mass and Energy Release Data From hot leg and cold leg break studies, the limiting break location for peak containment pressure is the hot leg at the steam generator inlet. Studies have shown that reduced reactor coolant system average temperature and differential cold leg temperature cases do not result in the limiting peak containment pressure response. The short-term LOCA peak pressure mass and energy releases are given in Table 6-29. The short-term LOCA peak pressure mass and energy releases using the replacement steam generators are given in Table 6-35.

6.2.1.3.2 Long-Term Mass and Energy Release Date The long-term LOCA mass and energy releases calculated by RELAP5/MOD2-B&W for the cold leg injection phase of the transient are given in Table 6-30. The mass and energy releases used by the FATHOMS computer code for periods beyond this point in time are given in Table 6-31.

6.2.1.3.3 Energy Sources The generated energy sources considered in the LOCA mass and energy release calculations are fission power, fission product and actinide decay energy, and metal-water reaction. The analyzed (actual) core power level is conservatively assumed to be 2% above the licensed

UFSAR Chapter 6 Oconee Nuclear Station 6.2 - 10 (31 DEC 2019)

(indicated) power. The assumed core axial power distribution is chosen to maximize the amount of steam exiting the break. Fission product and actinide decay power is calculated as a function of time based on the methodology from Reference 7. For conservatism an upper bound of two standard deviations above the mean value is used. The modeling of metal-water reaction energy is discussed in Section 6.2.1.3.6.

The stored energy sources considered in the LOCA mass and energy release calculations are fluid stored energy in the initial primary and secondary system inventories, stored energy in the primary and secondary structural metal components, stored energy in the fuel rods, and the energy content of the fluid added to the primary and secondary systems during the accident.

6.2.1.3.3.1 Short-Term Energy Sources The short-term LOCA mass and energy release calculation conservatively determines fission power by modeling moderator density, fuel, temperature, and boron feedbacks as described in Reference 1.

The specific conservatisms used in modeling the stored energy sources considered in the short-term LOCA mass and energy release calculations are:

1. The nominal volume of the Reactor Coolant System calculated based on cold dimensions is increased by 1% to account for the increase in volume due to thermal expansion to operating temperatures.

2. Initial pressurizer liquid mass is increased by assuming an initial level corresponding to the maximum Technical Specification allowable value of 285 inches plus an indication uncertainty of 25 inches.

3. The assumed reactor vessel average temperature is the nominal at power value, 579°F, plus a 2°F uncertainty allowance. This maximizes the stored energy in the Reactor Coolant System.

4. Reactor Coolant System pressure is at the nominal at power value, 2155 psig, plus a 30 psi uncertainty allowance. This maximizes the saturation temperature and therefore the energy content of the pressurizer fluid.

5. The core flood tank (CFT) temperature is assumed to be a conservatively high value of 130°F. This minimizes available sensible heat capacity of the CFT liquid and therefore maximizes the break steaming rate.

6. The CFT initial pressure used is the upper Technical Specification limit plus a 30 psi instrument uncertainty. This maximizes the amount of noncondensible gas released to containment and therefore the containment pressure.

7. The CFT liquid volume is the lower Technical Specification limit less an instrument uncertainty of 38 ft3. This minimizes available sensible heat capacity of the CFT liquid and maximizes the amount of noncondensible gas released to containment, both of which, as explained above, tend to increase containment pressure and temperature.

8. The main steam safety valve lift setpoints incorporate 3% drift and 4% blowdown to minimize the potential for steam relief. This maximizes the amount of hot secondary side fluid remaining in the steam generators and able to transfer its energy via the primary side to containment.

9. The BWST temperature is assumed to be a conservatively high value of 115°F, which minimizes available sensible heat capacity of the BWST liquid and therefore maximizes the break steaming rate.

Oconee Nuclear Station UFSAR Chapter 6 (31 DEC 2019) 6.2 - 11

10. Main feedwater temperature is at the nominal at power value, 453°F, plus a 7°F uncertainty allowance. This maximizes the potential energy release to containment. For further conservatism this assumed main feedwater temperature is maintained during the analysis although actual temperature would decrease as bleed steam was lost due to the break.

11. Main feedwater flow is maximized by controlling flow to the higher natural circulation setpoint even before the reactor coolant pumps are tripped. The nominal level setpoint is increased by a 12.5% operating range allowance for instrument uncertainty. This is conservative since it maximizes the amount of higher energy secondary side inventory available to transfer heat to the containment via the primary side.

12. Emergency feedwater temperature is at a conservatively high value of 130°F. This maximizes the potential energy release to containment.

6.2.1.3.3.2 Long-Term Energy Sources The long-term LOCA mass and energy release calculation conservatively determines fission power by modeling moderator density, fuel, temperature, and boron feedbacks as described in Reference 1.

The specific conservatisms used in modeling the stored energy sources considered in the long-term LOCA mass and energy release calculations are: