Text

Updated Final Safety Analysis Report, Revision 24, Chapter 9 - Auxiliary Systems
	ML15197A316
Person / Time
Site:	Oconee
Issue date:	06/30/2015
From:	; Duke Energy Carolinas
To:	; Office of Nuclear Reactor Regulation
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ML15196A571	List: ML15196A558; ML15196A560; ML15196A570; ML15197A315; ONS-2015-072, Oconee, Units 1, 2, and 3, Submittal of Updated Final Safety Analysis Report, Revision 24; ONS-2015-072, Oconee, Units 1, 2, and 3, Updated Final Safety Analysis Report, Revision 24, Chapter 2 - Site Characteristics; ONS-2015-072, Oconee, Units 1, 2, and 3, Updated Final Safety Analysis Report, Revision 24, Chapter 2 - Site Characteristics, Tables - Appendix 2A. Tables; ONS-2015-072, Oconee, Units 1, 2, and 3, Updated Final Safety Analysis Report, Revision 24, Chapter 8 - Electric Power; ONS-2015-072, Updated Final Safety Analysis Report, Revision 24, Chapter 9 - Auxiliary Systems;
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Oconee Nuclear Station UFSAR Chapter 9 Table of Contents 9.0 Auxiliary Systems 9.1 Fuel Storage and Handling 9.1.1 New Fuel Storage 9.1.2 Spent Fuel Storage 9.1.2.1 Spent Fuel Storage - Oconee 1, 2 9.1.2.1.1 Design Bases 9.1.2.1.2 Design Description 9.1.2.2 Spent Fuel Storage - Oconee 3 9.1.2.2.1 Design Bases 9.1.2.2.2 Design Description 9.1.2.3 System Evaluation 9.1.2.3.1 Structural and Seismic Analysis 9.1.2.3.2 Criticality Analysis 9.1.2.3.3 Material, Construction, and Quality Control 9.1.2.3.4 Interface of High Capacity Fuel Storage Rack and Spent Fuel Storage Pool 9.1.2.4 Safety Evaluation 9.1.2.5 Boraflex 9.1.3 Spent Fuel Cooling System 9.1.3.1 Design Bases 9.1.3.1.1 Units 1 and 2 Spent Fuel Pool Cooling System 9.1.3.1.2 Unit 3 Spent Fuel Pool Cooling System 9.1.3.2 System Description 9.1.3.3 System Evaluation 9.1.3.3.1 Normal Operation 9.1.3.3.2 Failure Analysis 9.1.3.4 Safety Evaluation 9.1.4 Fuel Handling System 9.1.4.1 Design Bases 9.1.4.1.1 General System Function 9.1.4.1.2 New Fuel Storage 9.1.4.1.3 Spent Fuel Pool 9.1.4.1.4 Fuel Transfer Tubes 9.1.4.1.5 Fuel Transfer Canal 9.1.4.1.6 Fuel Handling Equipment 9.1.4.2 System Description and Evaluation 9.1.4.2.1 Receiving and Storing Fuel 9.1.4.2.2 Loading and Removing Fuel 9.1.4.2.3 Safety Provisions 9.1.5 Overhead Heavy - Load Handling Systems 9.1.5.1 Introduction and Licensing Background 9.1.5.2 Design Basis 9.1.5.3 Scope of Heavy Load Handling Systems 9.1.5.4 Control of Heavy Lifts Program 9.1.5.4.1 Oconee Commitments in Response to NUREG-0612, Phase 1 Elements 9.1.5.4.2 Oconee Response to NEI Initiative on Heavy Load Lifts 9.1.5.5 Safety Evaluation 9.1.6 References 9.2 Water Systems 9.2.1 Component Cooling System 9.2.1.1 Design Bases (31 DEC 2014) 9-1

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) 9.2.2.2.5 Essential Siphon Vacuum and Siphon Seal Water Systems 9.2.3 Auxiliary Service Water System (Deleted Per 2014 Update - Refer to UFSAR Section 9.7, Protected Service Water System)

Deleted Per 2014 Update 9.2.3.1 Deleted Per 2014 Update 9.2.3.2 9.2.4 Ultimate Heat Sink 9.2.5 Control Room Ventilation Chilled Water System (WC) 9.2.5.1 Design Basis 9.2.5.2 System Description and Evaluation 9.2.6 References 9.3 Process Auxiliaries 9.3.1 Chemical Addition and Sampling System 9.3.1.1 Design Bases 9.3.1.2 System Description and Evaluation 9.3.1.2.1 Mode of Operation 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 (31 DEC 2014) 9-2

Oconee Nuclear Station UFSAR Chapter 9 9.3.3.2.6 Operational Limits 9.3.3.2.7 Failure Considerations 9.3.4 Coolant Storage System 9.3.4.1 Design Bases 9.3.4.2 System Description and Evaluation 9.3.5 Coolant Treatment System 9.3.6 Post-Accident Sampling System 9.3.6.1 Post-Accident Liquid Sampling System 9.3.6.1.1 Design Bases 9.3.6.1.2 Deleted Per 2013 Update 9.3.6.1.3 Deleted Per 2005 Update 9.3.6.2 Post-Accident Containment Air Sampling System 9.3.6.2.1 Design Bases 9.3.6.2.2 Deleted Per 2005 Update 9.3.6.2.3 Deleted Per 2005 Update 9.3.7 Containment Hydrogen Monitoring System 9.3.7.1 Design Bases 9.3.7.2 System Description 9.3.7.3 Safety Evaluation 9.4 Air Conditioning, Heating, Cooling and Ventilation Systems 9.4.1 Control Room Ventilation 9.4.1.1 Design Bases 9.4.1.2 System Description 9.4.1.2.1 Control Room Oconee 1 and 2 9.4.1.2.2 Control Room Oconee 3 9.4.1.3 Safety Evaluation 9.4.1.4 Inspection and Testing Requirements 9.4.2 Spent Fuel Pool Area Ventilation System 9.4.2.1 Design Bases 9.4.2.2 System Description 9.4.2.3 Safety Evaluation 9.4.2.4 Inspection and Test Requirements 9.4.3 Auxiliary Building Ventilation System 9.4.3.1 Design Bases 9.4.3.2 System Description 9.4.3.3 Safety Evaluation 9.4.3.4 Inspection and Testing Requirements 9.4.4 Turbine Building Ventilation System 9.4.4.1 Design Bases 9.4.4.2 System Description 9.4.4.3 Safety Evaluation 9.4.4.4 Inspection and Testing Requirements 9.4.5 Reactor Building Purge System 9.4.5.1 Design Bases 9.4.5.2 System Description 9.4.5.3 Safety Evaluation 9.4.5.4 Inspection and Testing Requirements 9.4.6 Reactor Building Cooling System 9.4.6.1 Design Bases 9.4.6.2 System Description 9.4.6.3 Safety Evaluation 9.4.6.4 Inspection and Testing Requirements 9.4.7 Reactor Building Penetration Room Ventilation System 9.4.7.1 Design Bases 9.4.7.2 System Description (31 DEC 2014) 9-3

Oconee Nuclear Station UFSAR Chapter 9 9.4.7.3 Safety Evaluation 9.4.7.4 Inspection and Test Requirements 9.4.8 Ventilation Systems in the Station Battery Rooms 9.4.9 References 9.5 Other Auxiliary Systems 9.5.1 Fire Protection 9.5.1.1 Design Bases 9.5.1.1.1 Defense-in Depth (DID) 9.5.1.1.2 Fundamental Fire Protection Program and Design Elements 9.5.1.1.3 Nuclear Safety Goals, Objectives and Performance Criteria 9.5.1.1.4 Radioactive Release Goal, Objectives and Criteria 9.5.1.2 System Description 9.5.1.2.1 Power Block 9.5.1.2.2 Nuclear Safety Capability Systems and Equipment 9.5.1.2.3 Fire Protection Systems and Features 9.5.1.3 Safety Evaluation 9.5.1.3.1 Fundamental Fire Protection Program and Design Elements 9.5.1.3.2 Nuclear Safety at Power 9.5.1.3.3 Nuclear Safety at Non-Power Operation (NPO) 9.5.1.3.4 Radioactive Release 9.5.1.3.5 Fire PRA 9.5.1.3.6 Summary of NFPA 805 Compliance 9.5.1.4 Inspection and Testing 9.5.1.4.1 Inspection, Testing and Surveillance Methodology 9.5.1.4.2 Monitoring Program Methodology 9.5.1.5 Personnel Qualification and Training 9.5.1.5.1 Fire Brigade 9.5.2 Instrument and Breathing Air Systems 9.5.2.1 Design Basis 9.5.2.2 System Description 9.5.2.3 Instrument Air (IA) System Tests and Inspections 9.6 Standby Shutdown Facility 9.6.1 General Description 9.6.2 Design Bases 9.6.3 System Descriptions 9.6.3.1 Structure 9.6.3.2 Reactor Coolant Makeup (RCM) System 9.6.3.3 Auxiliary Service Water (ASW) System 9.6.3.4 Electrical Power 9.6.3.4.1 General Description 9.6.3.4.2 Diesel Generator 9.6.3.5 Instrumentation 9.6.3.5.1 SSF Reactor Coolant Makeup System Instrumentation 9.6.3.5.2 SSF Auxiliary Service Water Instrumentation 9.6.3.6 Support Systems 9.6.3.6.1 SSF Lighting System Description 9.6.3.6.2 SSF Fire Protection and Detection 9.6.3.6.3 SSF Service Water 9.6.3.6.4 Heating Ventilation and Air Conditioning 9.6.3.6.5 SSF Sump System 9.6.4 System Evaluations 9.6.4.1 General 9.6.4.2 Structure Design 9.6.4.3 Seismic Subsystem Analysis (31 DEC 2014) 9-4

Oconee Nuclear Station UFSAR Chapter 9 9.6.4.4 Dynamic Testing and Analysis of Mechanical Components 9.6.4.5 ASME Code Class 1, 2, and 3 Components, Component Supports and Core Support Structures 9.6.4.6 Fire Protection 9.6.4.6.1 Safe Shutdown Systems 9.6.4.6.2 Performance Goals 9.6.4.6.3 Instrumentation Guidelines 9.6.4.6.4 Repairs for Hot Shutdown 9.6.4.6.5 Fire Protection Conclusion 9.6.4.7 Flooding Review 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 Bases 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 2014) 9-5

Oconee Nuclear Station UFSAR Chapter 9 List of Tables Table 9-1. Spent Fuel Cooling System Data, Units 1, 2 Table 9-2. Spent Fuel Cooling System Data, Oconee 3 Table 9-3. Component Cooling System Performance Data (For Normal Operation on a Per Oconee Basis)

Table 9-4. Cooling Water Systems Component Data (Component Data on a Per Unit Basis)

Table 9-5. Chemical Addition and Sampling System Component Data Table 9-6. High Pressure Injection System Performance Data Table 9-7. High Pressure Injection System Component Data Table 9-8. Low Pressure Injection System Performance Data Table 9-9. Low Pressure Injection System Component Data Table 9-10. Coolant Storage System Component Data (Component Quantities for Three Units)

Table 9-11. Ventilation System Major Component Data Table 9-12. Deleted Per 2002 Update.

Table 9-13. Component Cooling System Component Data (Component Data on a Per Unit Basis)

Table 9-14. SSF System Main Components Table 9-15. SSF Primary Valves Table 9-16. SSF Instrumentation Table 9-17. Design Basis Tornado Missiles And Their Impact Velocities Table 9-18. Design Basis Tornado Missiles Minimum Barrier Thicknesses Table 9-19. Codes and Specifications For Design of Category I Structures (31 DEC 2014) 9-6

Oconee Nuclear Station UFSAR Chapter 9 List of Figures Figure 9-1. Fuel Storage Rack (Module)

Figure 9-2. Fuel Storage Rack (Assembly)

Figure 9-3. Spent Fuel Pool Outline Oconee 1, 2 Figure 9-4. Spent Fuel Pool Outline Oconee 3 Figure 9-5. Spent Fuel Cooling System Figure 9-6. Deleted per 1990 Update Figure 9-7. Fuel Handling System (Units 1&2 Page 1 and Unit 3 Page 2)

Figure 9-8. Component Cooling System Figure 9-9. Condenser Circulating Water System Figure 9-10. High Pressure Service Water System Figure 9-11. Low Pressure Service Water System Figure 9-12. Low Pressure Service Water System Figure 9-13. Recirculated Cooling Water System Figure 9-14. Deleted Per 1997 Update Figure 9-15. Chemical Addition and Sampling System Figure 9-16. Chemical Addition and Sampling System Figure 9-17. High Pressure Injection System Figure 9-18. High Pressure Injection System Figure 9-19. Low Pressure Injection System Figure 9-20. Coolant Storage System Figure 9-21. Coolant Treatment System Figure 9-22. Post-Accident Liquid Sample System Figure 9-23. Post-Accident Containment Air Sample System Figure 9-24. Control Room Area Ventilation and Air Conditioning System Figure 9-25. Spent Fuel Pool Ventilation System Unit 1 and 2 Figure 9-26. Spent Fuel Pool Ventilation System Unit 3 (31 DEC 2014) 9-7

Oconee Nuclear Station UFSAR Chapter 9 Figure 9-27. Auxiliary Building Ventilation System Unit 1 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 9-31. 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 Figure 9-45. PSW AC Electrical Distribution Figure 9-46. PSW DC Electrical Distribution (31 DEC 2014) 9-8

Oconee Nuclear Station UFSAR Chapter 9 9.0 Auxiliary Systems The Auxiliary Systems required to support the reactor during normal operations and servicing of the Oconee Nuclear Station are described in this section. Some of these systems have also been described and discussed in Chapter 6, since they serve as engineered safeguards. The information in this section deals primarily with the functions served by these systems during normal operation.

The design of the Auxiliary Systems has included consideration of system sharing, where feasible, between the three Oconee Nuclear Station units. This section describes the equipment for each unit and states where equipment is shared.

The majority of the components in these systems are located within the Auxiliary Building. Those systems connected by piping between the Reactor Building and the Auxiliary Building are equipped with Reactor Building isolation valves as described in Chapter 6.

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UFSAR Chapter 9 Oconee Nuclear Station THIS PAGE LEFT BLANK INTENTIONALLY.

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Oconee Nuclear Station UFSAR Chapter 9 9.1 Fuel Storage and Handling 9.1.1 New Fuel Storage New fuel will normally be stored in the spent fuel pool serving the respective unit. New or irradiated fuel assemblies with initial nominal enrichments up to 5.00 weight percent U-235 which do not meet the requirements for unrestricted storage must be placed in a restricted loading pattern.

Deleted paragraph(s) per 2002 update.

Reactivity analyses for these assemblies, stored in a checkerboard type configuration in the spent fuel pool, were performed using the methods discussed in Section 9.1.2.3.2.

New fuel may also be stored in the fuel transfer canal. The fuel assemblies are stored in five racks in a row having a nominal center-to-center distance of 2 ft 1-3/4 inches. One rack is oversized to receive a failed fuel assembly container. The other four racks are normal size and are capable of receiving new fuel assemblies.

New fuel may also be stored in shipping containers.

9.1.2 Spent Fuel Storage 9.1.2.1 Spent Fuel Storage - Oconee 1, 2 The Spent Fuel Pool common to Oconee Units l and 2 has been re-racked to increase the spent fuel storage capacity to 1312 fuel assemblies. This modification is pursuant to License Amendment Nos. 90, 90 and 87 for License Nos. DPR-38, DPR-47 and DPR-55 for the Oconee Nuclear Station.

9.1.2.1.1 Design Bases The Spent Fuel Pool designed for Oconee 1 and 2 is an integral part of the Oconee 1 and 2 Auxiliary Buildings and conforms to Safety Guide 13, Fuel Storage Design Basis. The fuel pools were designed for tornado wind and missiles, turbine generator missile, and seismic conditions as listed in Table 3-23.

The Spent Fuel Pools were analyzed for the postulated cask drop accident as described in Section 3.8.4.4.

The spent fuel pool is constructed of reinforced concrete lined with stainless steel plate. The fuel pool concrete, reinforcing steel, liner plate and welds connecting the liner plate to the fuel pool floor concrete embedments are analyzed based on consideration of the new racks and additional fuel. Design criteria including loading combinations and allowable stresses are in compliance with Oconee FSAR Section 3.8.4 for Class I structures. The determination of Ta (abnormal thermal load condition to be used in combination with E') is based on the failure of one pump or cooler during normal operating conditions.

The function of the spent fuel storage racks is to provide for storage of spent fuel assemblies in a flooded pool, while maintaining a coolable geometry, preventing criticality, and protecting the fuel assemblies from excess mechanical or thermal loadings.

A list of design criteria is given below:

1. The racks are designed in accordance with the NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, dated April 14, 1978 and revised January 18, 1979.

2. The racks are designed to meet the nuclear requirements of ANSI/ANS-57.2-1983. Oconee complies with the criticality accident requirements of 10CFR50.68(b) (Reference 27). The effective multiplication factor, Keff, in the spent fuel pool is less than or equal to 0.95, including all uncertainties and under all credible conditions with partial credit for soluble boron.

(31 DEC 2014) 9.1 - 1

UFSAR Chapter 9 Oconee Nuclear Station

3. The racks are designed to allow coolant flow such that boiling in the water channels between fuel assemblies does not occur.

4. The racks are designed to Seismic Category 1 requirements, and are classified as ANS Safety Class 3 and ASME Code Class 3 Component Support structures.

5. The racks are designed to withstand loads which may result from fuel handling accidents and from the maximum uplift force of the fuel handling crane.

6. Each storage position in the racks is designed to support and guide the fuel assembly in a manner that will minimize the possibility of application of excessive lateral, axial and bending loads to fuel assemblies during fuel assembly handling and storage.

7. The racks are designed to preclude the insertion of a fuel assembly in other than design locations.

8. The materials used in construction of the racks are compatible with the storage pool environment and do not contaminate the fuel assemblies 9.1.2.1.2 Design Description The Oconee fuel storage racks are composed of individual storage cells made of stainless steel interconnected by grid assemblies to form integral module structures as shown in Figure 9-1. Each cell has a lead-in opening which is symmetrical and is blended smooth to facilitate fuel insertion. The cells are open at the top and bottom to provide a flow path for convective cooling of spent fuel assemblies through natural circulation. The fuel assembly storage cells are structurally connected to form modules through the use of plates and box beams which limit structural deformations and maintain a nominal center-to-center spacing between adjacent storage cavities during design conditions including the Safe Shutdown Earthquake. The racks utilize a neutron absorber, Boraflex, which is attached to each cell.

However, due to degradation of the absorber material, no reactivity holddown credit is taken for any remaining Boraflex in the storage cells. The modules are neither anchored to the floor nor braced by the pool walls. The following information applies to the Oconee 1 and 2 fuel storage pool.

Number of Cells 1312 Number of Modules 4 - 8 x 11 10 - 8 x 12 Deleted row(s) per 2002 Update.

Center-to-Center Spacing 10.65 in.

Deleted row(s) per 2004 Update.

Approximate Rack Assembly Dimensions and 8 x 10 - 85.5 x 107 x 172 - 18, 060 lbs.

Maximum Weights 8 x 12 - 85.5 x 128 x 172 - 21,800 lbs.

The pool outline and rack arrangements are shown in Figure 9-3 and Figure 9-4.

9.1.2.2 Spent Fuel Storage - Oconee 3 The Spent Fuel Pool serving Oconee Unit 3 has been re-racked to increase the spent fuel storage capacity to 822 fuel assemblies, plus 3 additional storage spaces for failed fuel containers. This modification is pursuant to License Amendment Nos. 123, 123, and 120 for License Nos. DPR-38, DPR-47 and DPR-55 for the Oconee Nuclear Station.

9.1 - 2 (31 DEC 2014)

Oconee Nuclear Station UFSAR Chapter 9 9.1.2.2.1 Design Bases The Oconee 3 Spent Fuel Pool has the same Design Bases as the Oconee 1 and 2 pool described in Section 9.1.2.1.1.

9.1.2.2.2 Design Description The Oconee 3 Spent Fuel Pool storage racks are similar to the Oconee 1 and 2 racks described in Section 9.1.2.1.2. The following information applies to Oconee Unit 3 spent fuel storage racks.

Number of Cells 822 plus storage locations for 3 failed fuel containers Number Rack Arrays 7 - 8 x 11 2 - 8 x 12 1 - 8 x 10 x/3 container locations Deleted row(s) per 2002 Update.

Center-to-Center Spacing 10.60 in.

Deleted row(s) per 2004 Update.

Approximate Rack Assembly Dimensions and 8 x 10 - 85.5 x 107 x 172 - 18, 060 lbs.

Maximum Weights 8 x 12 - 85.5 x 128 x 172 - 21,800 lbs.

The pool outline and rack arrangements are shown in Figure 9-3 and Figure 9-4.

9.1.2.3 System Evaluation 9.1.2.3.1 Structural and Seismic Analysis Fuel assembly storage rack and associated structures are designed to withstand the maximum forces generated during normal operation combined with the Safe Shutdown Earthquake according to the requirements of a Seismic Class l structure. For these conditions, the storage rack design is such that all stresses fall within the allowable stress limits specified in the AISC Specifications for Design, Fabrication and Erection of Structural Steel.

Normal operating loads include dead weight (in air) and thermal expansion loads. Lateral and vertical seismic loads along with the fluid forces generated by seismically generated pool water sloshing are considered to be acting simultaneously.

The seismic input spectra conform to the requirements of Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants.

Reference is made to Project 81 PSAR, Docket Nos. STN50-488 through -493, Section 3.7. The smoothed response spectra shown on Figure 2E-2A were normalized to 10 percent g for Safe Shutdown Earthquake (SSE). An earthquake acceleration-time history compatible with these spectra, as shown in Figures 2E-2B through 2E-2E, was used as a base motion on the model of the Auxiliary Building.

The seismic response of the Auxiliary Building to the base excitation is determined by a dynamic analysis. The dynamic analysis is made by idealizing the structure as a series of lumped masses with weightless elastic columns acting as spring restraints. The base of the structure is considered fixed. The choice of the location of the mass-joints depends on the distribution of masses in the real structure.

The seismic analysis of the racks was performed in two phases:

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UFSAR Chapter 9 Oconee Nuclear Station First a seismic time history analysis of a simplified non-linear 2-dimensional model was conducted. The model consisted of spring, mass, damping, friction, and gap elements to simulate a fuel bundle in a simplified model of a rack. The fuel assembly-to-cell impact loads, support pad lift-off values, rack sliding, and overall rack response were determined from the non-linear analysis. Coefficients of friction were varied between minimum and maximum possible values in order to determine worst case conditions of sliding and tipping respectively. Rack-to-rack impacts were precluded by spacing the racks beyond maximum possible excursion. The gap spaces are large enough to accomodate lateral module motion due to earthquake forces. In order to account for 3-dimensional effects, the results of independent orthogonal loadings were combined by the SRSS method.

Next, a seismic response spectrum analysis of a 3-dimensional finite element model of the racks, using inputs from the results of the non-linear analysis, and superimposed with other applicable loads, was conducted. Design stresses and safety margins for appropriate components in the racks were tabulated and found to be acceptable.

The structural damping values used are 4 percent for an SSE and 2 percent for an OBE.

The maximum uplift load available from the fuel handling crane on the storage rack is limited to 3000 lbs or less by the hoist interlock. A separate fuel assembly drop analysis was performed. A 3000 pound object was postulated to impact the top of the rack from a height of 6 feet. Calculations show that the resulting stresses are within acceptable stress limits.

Structural design precludes placing a fuel assembly between cells, and the rack will withstand the loadings imposed by a postulated dropped fuel assembly.

9.1.2.3.2 Criticality Analysis The design methodology which ensures the criticality safety of the fuel assemblies in the spent fuel storage rack is discussed in Section 9.1.2.3.2.3 and in Reference 8.

9.1.2.3.2.1 Neutron Multiplication Factor Criticality of fuel assemblies in the spent fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies and inserting neutron poisons between assemblies.

The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (keff) of the fuel assembly array will be less than 1.0 in unborated spent fuel pool water, and less than 0.95 with partial credit for soluble boron, in accordance with Reference 26. Oconee complies with the criticality accident requirements of 10CFR50.68(b) (Reference 27). The acceptance criteria for criticality is further discussed in Section 9.1.2.3.2.5.

9.1.2.3.2.2 Normal Storage Under normal storage conditions, the following assumptions were used in the criticality analysis.

1. Credit is taken for the decrease in reactivity associated with the fuel assembly burnup.

2. The fuel assembly is the most reactive fuel assembly to be stored based on a minimum burnup. The fuel designs analyzed include the fuel assembly designs described in Chapter 4, and earlier designs.

Additionally, a small number of alternate fuel configurations are analyzed (e.g. lead test asssemblies, failed rod canisters, and rod consolidation canisters).

3. The moderator is at the temperature within the design limits of the pool which yields the largest reactivity, and contains at least 430 ppm boron (pursvant to License Amendment Nos. 323, 323, 324 9.1 - 4 (31 DEC 2014)

Oconee Nuclear Station UFSAR Chapter 9 for License Nos. DPR-38, DPR-47, and DPR-55), to maintain Keff 0.95 for normal storage conditions. Full credit for soluble boron is taken for postulated accident conditions and during fuel movement. For accident conditions the double contingency principle of ANSI N16.1-1975 is applied.

This principle states that it shall require at least two unlikely, independent, and concurrent events to produce a criticality accident. During fuel movement the presence of dissolved boron in the spent fuel pool water is assumed since this is only a temporary condition and only a single assembly is handled at a time.

4. The array is either infinite in the lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the design. The nominal case calculation is infinite in the lateral extent.

However, poison plates are not necessary on the periphery of the modular array and between widely spaced modules because calculations show that this finite array is less reactive than the nominal case infinite array. The assemblies are also infinite in the axial extent. A reactivity bias is included in all burned-fuel criticality calculations to conservatively account for reactivity differences between a detailed 3-D axial burnup model and the 2-D average burnup model employed for nominal calculations.

5. Mechanical uncertainties and biases due to mechanical tolerances during construction are treated by either using "worst case" conditions or by performing sensitivity studies and obtaining appropriate values. The items included in the analysis are:

a. Deleted row per 2002 Update.

b. Deleted row per 2002 Update.

c. Can ID

d. Stainless steel thickness

e. Center-to-center spacing

f. Fuel enrichment

g. Fuel pellet density

h. Fuel pellet OD Other applicable uncertainties and biases are discussed in Section 9.1.2.3.2.3.

6. No credit is taken for the assembly spacer grids.

7. No credit is taken for fuel assembly control components which can be removed (e.g. burnable poisons and control rods).

8. Credit is taken for the inherent neutron absorbing effect of some of the rack structure materials in accordance with ANSI/ANS-57.2-1983 and Reference 26.

9.1.2.3.2.3 Criticality Analysis Methodology Criticality of fuel assemblies outside the reactor is precluded by adequate design of fuel transfer, shipping and storage facilities and by administrative control procedures. The two principal methods of preventing criticality are limiting the fuel assembly array size and limiting assembly interaction by fixing the minimum separation between assemblies and/or inserting neutron poisons between assemblies.

The design basis for preventing criticality outside the reactor is that, considering possible variations, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (keff) of the fuel assembly array will be less than or equal to 0.95, with partial credit for soluble boron. Oconee complies with the criticality accident requirements of 10CFR50.68(b) (Reference 27). The conditions that are assumed in meeting this design basis are outlined in Section 9.1.2.3.2.2.

(31 DEC 2014) 9.1 - 5

UFSAR Chapter 9 Oconee Nuclear Station In order to justify storage of fuel up to 5.0 w/o, the burnup credit approach was utilized in the spent fuel pools. The burnup credit approach to fuel rack criticality analysis requires calculation and comparison of reactivity values over a range of burnup and initial enrichment conditions. In order to accurately model characteristics of irradiated fuel which impact reactivity, a criticality analysis method capable of evaluating arrays of these irradiated assemblies is needed. The advanced nodal methodology combining CASMO-3/TABLES-3/SIMULATE-3 is used for this purpose. CASMO-3 (Reference 4) is an integral transport theory code, SIMULATE-3 (Reference 6) is a nodal diffusion theory code, and TABLES-3 (Reference 5) is a linking code which reformats CASMO-3 data for use in SIMULATE-3. This methodology permits direct coupling of incore depletion calculations and resulting fuel isotopics with out-of-core storage array criticality analyses. The variable effects of fission product poisoning, fissile material production and utilization and other related effects are accurately modeled with the CASMO-3/TABLES-3/SIMULATE-3 methodology. Applicable biases and uncertainties are developed and become inputs to the methodology.

The results for the criticality methodology are validated by comparison to measured results of fuel storage critical experiments. The criticality experiments used to benchmark the methodology were the Babcock and Wilcox close proximity storage critical experiments performed at the CX-10 facility (Reference 7).

The B&W critical experiments used are specifically designed for benchmarking reactivity calculation techniques. The experiments are analyzed, and the statistical accuracy of the calculated reactivity results are assessed.

The bias associated with the benchmarks is -0.00142 k with a standard deviation of 0.00412 k. The 95/95 one-sided tolerance limit factor for 10 values is 2.911. Therefore, there is a 95 percent probability at a 95 percent confidence level that the uncertainty in reactivity due to the method is not greater than 0.01199 k.

For burned fuel, the maximum reactivity occurs approximately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after shutdown due to the decay of Xe135. Therefore, all fuel assemblies in the spent fuel pool are modeled at no xenon conditions.

An additional bias and uncertainty are required to quantify the reactivity of burned nuclear fuel assemblies. Two burnup uncertainties associated with this methodology are accounted for in the criticality analysis. The first penalty accounts for uncertainties in the reactivity due to uncertainties in the burnup of the assembly, while the second penalty accounts for the reactivity holddown effect of lumped burnable absorbers.

The exposure reactivity uncertainty accounts for the uncertainty on the assembly burnup. Since the final burnup qualification curves are based on a code calculated burnup, the uncertainty in that calculated burnup must be considered. Rather than determining the uncertainty on the actual burnup, the uncertainty on reactivity due to burnup was applied to account for the burnup uncertainty. A 95/95 one-sided tolerance was determined to account for the maximum reactivity error associated with the burnup of the fuel.

As required by the standards, no removable poisons are accounted for in the criticality analyses. Thus, all assemblies are modeled with no burnable poisons (BPs). However, this can be slightly non-conservative due to the increase in reactivity associated with the removal of the BP. Thus a burnable poison removal (BP-Pull) penalty is developed to account for this effect. BPs are used in the core design to hold down reactivity, and hence peaking of fresh assemblies. Thus, the reactivity of the BPd assembly is less than the non-BPd assembly. However, once the BP is removed (from the previously BPd assembly), a reactivity increase is seen due to the shadowing effect the BPs had on the assembly. This difference in reactivity is applied as an additional bias on reactivity.

The basic approach in the burnup credit methodology is to use reactivity equivalencing techniques to construct burnup versus enrichment curves which represent equivalent and acceptable reactivity 9.1 - 6 (31 DEC 2014)

Oconee Nuclear Station UFSAR Chapter 9 conditions over an applicable range of burnups and initial enrichments. These burnup versus enrichment curves are established for each type of storage, e.g. unrestricted and restricted storage.

Generation of the applicable burnup credit curves requires a two part calculation process. The first part is to create two types of reactivity versus burnup curves. The first type of curve defines the maximum reactivity for the spent fuel pool such that the appropriate design criteria are met including allowances for both calculational uncertainties and manufacturing tolerances. The second type of curve represents the reactivity versus burnup for a particular enrichment, and is generated for the range of enrichments. The intersection of the maximum design reactivity curve with the multiple enrichment curves provides data points for the second part of the process.

The second part of the process generates the burnup versus initial enrichment curves by plotting the burnup where the maximum design reactivity equals the reactivity of a particular enrichment for each enrichment. Two curves are generated which represent the qualification criteria for a particular storage configuration. Each burnup versus enrichment curve shows the minimum amount of burnup required to qualify fuel for storage in the applicable loading pattern as a function of the fuel's initial enrichment.

Additional details of the methods used can be found in Reference 8.

The SCALE-4 system of computer codes (Reference 10) was used to analyze the boundary restrictions between Checkerboard, Restricted, and Unrestricted storage configurations to assure that the storage configurations at the boundary do not cause an increase in the nominal keff for the individual regions.

This analysis is performed to determine if there is a need for new administrative restrictions at the boundaries.

This methodology utilizes two dimensional Monte Carlo theory. Specifically, this analysis method used the CSAS25 sequence contained in Criticality Analysis Sequence No. 4 (CSAS4). CSAS4 is a control module contained in the SCALE-4.2 system of codes. The CSAS25 sequence utilizes two cross section processing codes (NITAWL and BONAMI) and a 3-D Monte Carlo code (KENO Va) for calculating the effective multiplication factor for the system. The 27 Group NDF4 cross section library was used exclusively for this analysis.

Acceptable interface boundary conditions between storage configurations were determined by varying the boundaries between various storage regions to determine the worst case configurations for coupling between assemblies in different regions. The boundaries were then reflected to simulate an infinite array.

The keff of these infinite boundary arrays were compared to the base keff of infinite arrays of either fuel storage region creating the boundary. If the infinite boundary array keff did not represent an increase in the keff of the regions making the boundary, then no storage restrictions were imposed at the interface. When the worst case did represent an increase, conservative storage restrictions were applied.

These methods conform with ANSI N18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System; ANSI/ANS-57.2-1983, Design Requirements for LWR Spent Fuel Storage Facilities at Nuclear Power Stations, Section 6.4.2.2; ANSI N16.9-1975, "NRC Standard Review Plan," Section 9.1.2 and the NRC guidance contained in Reference 26.

9.1.2.3.2.4 Postulated Accidents As part of the criticality analysis for the Oconee spent fuel pools, abnormal and accident conditions are considered to verify that acceptable criticality margin is maintained for all conditions. Most accident conditions will not result in an increase in keff of the rack. For an assembly dropped on top of the storage rack, the section of the rack structure which is essential for preventing criticality is not excessively deformed. Furthermore, the dropped assembly has more than eight inches of water separating it from the active fuel height of stored assemblies which precludes any interaction between the dropped assembly and the stored assemblies. Although the dropped assembly is more reactive outside of the storage cell rather (31 DEC 2014) 9.1 - 7

UFSAR Chapter 9 Oconee Nuclear Station than inside, the assembly is no more reactive dropped on top of the storage rack than located anywhere else in the pool outside the storage rack.

However, accidents can be postulated which would increase reactivity. Misloading of an assembly would increase reactivity; in particular, misloading the highest reactive assembly in place of the lowest reactive assembly. This is either the misplacement of a fresh assembly in an empty cell in the checkerboard pattern or in a filler cell in the restricted pattern.

For loss of spent fuel pool cooling scenarios, the reactivity increases with decreasing water density for the Oconee fuel storage racks and the current analyzed fuel designs. Two accident scenarios are postulated:

heat load due to the loss of one cooling train and cold water emergency makeup. The emergency makeup event encompasses a dilution event, since one source of makeup is Lake Keowee.

For accident conditions, the double contingency principle is employed. The double contingency principle of ANSI/ANS-57.2-1983 states that it is not required to assume two unlikely, independent concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition, since not assuming its presence would be a second unlikely event.

The acceptance criteria for criticality are further discussed in 9.1.2.3.2.5.

9.1.2.3.2.5 Acceptance Criteria for Criticality The acceptance criteria for the spent fuel pools will be keff 0.95. This assumes full credit may be taken for soluble boron under accident conditions as allowed by the double contingency principle in ANSI/ANS-57.2-1983, and that only partial credit is taken for soluble boron under normal conditions, per Reference 26. Oconee complies with the criticality accident requirements of 10CFR50.68(b) (Reference 27).

Deleted paragraph(s) per 2002 update.

9.1.2.3.2.6 Cask Drop Accident Cask drop accidents are analyzed for criticality consequences in Section 15.11.2.5.1.

9.1.2.3.2.7 Criticality Analyses for Loading NUHOMS Dry Storage Canisters (DSC)

The criticality analysis of the NUHOMS-24P/24PHB DSC, for loading and unloading operations in the Oconee spent fuel pools, has been performed in accordance with the requirements of 10CFR50.68(b).

The evaluation takes partial credit for soluble boron in the spent fuel pools. Minimum burnup requirements were developed for fuel to be placed without location restrictions in the NUHOMS-24P/24PHB DSC. These burnup requirements, applicable for eligible fuel assemblies with a minimum 5 years post-irradiation cooling time, are a function of initial U-235 enrichment.

The criticality analysis demonstrated that the current minimum boron concentration required in the Oconee spent fuel pools is adequate to maintain the maximum 95/95 Keff below 0.95 for all normal conditions and credible accident scenarios associated with loading fuel assemblies into the NUHOMS-24P/24PHB DSCs.

Consistency was maintained between the spent fuel pool rack and DSC normal and accident analyses.

Accidents analyzed included assembly dropped on top of the storage rack, misloading of an assembly, and loss of spent fuel pool cooling scenarios.

9.1 - 8 (31 DEC 2014)

Oconee Nuclear Station UFSAR Chapter 9 9.1.2.3.3 Material, Construction, and Quality Control The entire fuel assembly storage rack is constructed of type 304 stainless steel, with Boraflex panels attached to each cell. All welded construction is used in the fabrication of the fuel assembly storage rack.

The all-welded construction ensures the structural integrity of the storage modules and provides assurance of smooth, snag-free passage in the storage cavities so that it is highly improbable that a fuel assembly could become stuck in the rack.

The material, construction and quality control procedures are in accordance with the quality assurance requirements of Duke Power Company, as described in Duke Power Company Topical Report, DUKE-1.

9.1.2.3.4 Interface of High Capacity Fuel Storage Rack and Spent Fuel Storage Pool The pool floor will support the high capacity storage rack as a free-standing structure during all design conditions. During installation, no racks are moved over spent fuel assemblies in the pool. All spent fuel assemblies in Unit 3 are removed prior to removing existing racks.

For the free-standing rack structure, conservative analysis shows that under simultaneous forces from vertical and lateral seismic excitation, the residual displacement of the rack relative to the pool floor is less than 1 inch for full-loaded condition (i.e., much less than minimum clearance of 2.75 inches to pool walls and installed equipment.)

The maximum sliding distance of the Westinghouse free-standing fuel rack is obtained by equating the kinetic energy developed in the fuel rack, in response to the SSE seismic event, to the energy dissipated by friction between the fuel rack supports and the pool floor, during sliding. The maximum kinetic energy in the fuel rack, produced by the SSE seismic event, is calculated from the spectral response to the SSE response spectrum. The horizontal displacement of the rack is 1.414 times the sum of the deflecton of the top of the rack (0.245 in) and the maximum sliding distance (0.432). The coefficient of friction is assumed to be 0.20.

The rack/pool floor normal force on which the lateral friction forces used in the analysis are based includes the effect of vertical seismic acceleration.

The maximum lateral seismic force exerted by any rack module on pool floor is 189000 pounds and results in a stress of 2440 psi in the floor liner and 3296 psi in the weld connecting the floor liner to embedments in the concrete. The maximum combined seismic and thermal stress in the floor liner is 21640 psi and 30610 psi in the weld between liner and embedments. The maximum stresses are below the design allowable stress of 27,000 psi in the liner and 32000 psi in the welds.

9.1.2.4 Safety Evaluation The storage rack is designed and constructed to retain the integrity of the structure under all anticipated loads, including the Safe Shutdown Earthquake, with the maximum number of fuel assemblies occupying the storage locations.

The rack design provides protection against damage to the fuel and precludes the possibility of a fuel assembly being placed between cells. Although not required for safe storage of spent fuel assemblies, the spent fuel pool water is normally borated to a concentration of at least 2220 ppm, or higher as specified by the Core Operating Limits Report (COLR). The rack design also assures a Keff of less than 1.0 even when the entire array of fuel assemblies, assumed to be in their most reactive condition and within the limits specified in the Technical Specifications, are immersed in unborated water at room temperature.

Furthermore, if the pools were filled with the most reactive fuel allowed, which is clearly in violation of the Technical Specifications, Keff would be ~ 0.85 with full credit for soluble boron. Under these conditions a criticality accident during refueling or storage is not considered credible.

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UFSAR Chapter 9 Oconee Nuclear Station 9.1.2.5 Boraflex The spent fuel storage racks contain Boraflex, which is the trade name for a silicon polymer that contains a specified amount of Boron 10 that was originally used as the neutron absorber to assure that the design basis for criticality control was met through the service life of the racks. The Boraflex is affixed to each of the four exterior sides of the fuel storage cell by means of stainless steel wrappers. Boraflex was originally used in spent fuel storage racks for the nonproductive absorption of neutrons such that the NRC established acceptance criterion of keff no greater than 0.95 was maintained. However, due to degradation of the absorber material, no reactivity holddown credit is taken any longer for the remaining Boraflex in the storage cells.

Since reactivity hold-down credit is no longer being taken for Boraflex in the Spent Fuel Pool storage cells, the License Renewal commitment to inspect the Boraflex panels is no longer required, and the inspection program has been discontinued.

Deleted paragraph(s) per 2002 update.

9.1.3 Spent Fuel Cooling System 9.1.3.1 Design Bases 9.1.3.1.1 Units 1 and 2 Spent Fuel Pool Cooling System The primary function of Spent Fuel Pool Cooling System for Units 1 and 2 is to provide decay heat removal for the spent fuel stored in the Units 1 and 2 spent fuel pool. The cooling system design requirements are the criteria imposed by the 1980 re-racking (References 11, 12). Other system functions are to maintain the pool inventory, clarity and chemistry at acceptable levels.

Revised criteria have been imposed during the 1980 re-racking modification, pursuant to Amendments 90, 90, and 87 for License Nos. DPR-38, DPR-47 and DPR-55 for the Oconee Nuclear Station. The thermal-hydraulic analyses associated with the spent fuel pool racks assumes that the bulk spent fuel pool temperature remain at or below 150°F, for normal heat loads (Reference 11). The Units 1 and 2 Spent Fuel Cooling System is designed to keep the pool bulk water temperature:

1. Below 150°F for normal heat loads and two or three pump-cooler configurations in operation (Reference 11)

2. Below 150°F for abnormal heat loads and three pump-cooler configurations in operation (Reference 11)

3. Below 205°F for abnormal heat loads and any two pump-cooler configurations in operation (Reference 11).

For the Units 1 and 2 spent fuel cooling system, the design basis normal heat load assumes that Units 1 and 2 are refueled consecutively, and the rack positions are filled with previous discharges, except for 118 spaces reserved for a full core discharge (Reference 11). The design basis abnormal heat load assumes that Units 1 and 2 are refueled consecutively, followed by a full core discharge after a short period of operation. In this case, all rack positions contain spent fuel (References 11 and 12). The licensing basis decay heat predictions were performed with the methodology outlined in Reference 13. Various operational evolutions may utilize decay heat predictions based on the ORIGEN methodology (e.g.,

ORIGEN-ARP or SAS2H/ORIGEN-S) presented in References 24 and 25.

It should be noted that, while all temperature conditions above represent design criteria associated with specific analytical assumptions, only the higher temperature of 205°F represents an actual operating limit.

Analyses have been performed to ensure that seismic and structural integrity of the pool liner, supporting concrete, and fuel racks are not compromised at this temperature limit. Thermal - hydraulic analysis of the 9.1 - 10 (31 DEC 2014)

Oconee Nuclear Station UFSAR Chapter 9 racks has also shown that boiling within the fuel cells does not occur with pool temperatures maintained at or below this limit, provided normal operating pool level is maintained.

In addition to the primary function of decay heat removal, the system provides for purification of the spent fuel pool water, the fuel transfer canal water, and the contents of the borated water storage tank, in order to remove fission and corrosion products and to maintain water clarity for fuel handling operations.

The system also provides inventory makeup for the fuel transfer canal and the incore instrument handling tank.

The system is designed to withstand the effects of a seismic event and meet the requirements of Duke piping Class C for Oconee.

Portions of the Spent Fuel Cooling 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.

9.1.3.1.2 Unit 3 Spent Fuel Pool Cooling System The Unit 3 Spent Fuel Pool Cooling System duplicates the equipment used for the Units 1 and 2 system.

The Unit 3 system is designed to remove the decay heat from the stored fuel in the Unit 3 spent fuel pool.

The cooling system heat removal requirements are as set forth in NRC Standard Review Plan Section SRP-9.1.3 (References 14, 15). Other system functions are to maintain the pool inventory, clarity and chemistry at acceptable levels.

The Unit 3 system heat removal design requirements, as stipulated by Standard Review Plan 9.1.3, are:

1. For the maximum normal heat load with the normal cooling systems in operation, and assuming a single active failure, the temperature of the pool water shall be maintained at or below 140°F and the liquid level in the pool should be maintained.

2. For the abnormal maximum heat load with the normal cooling systems in operation, the pool water temperature should be kept below boiling and the liquid level in the pool should be maintained. A single active failure need not be considered.

The design basis maximum normal and abnormal decay heat loads are as defined in SRP 9.1.3 (Reference 15), for fuel racks with greater than 1 1/3 core storage capacity. The licensing basis decay heat predictions were performed with the methodology outlined in Reference 13. Various operational evolutions may utilize decay heat predictions based on the ORIGEN methodology (e.g., ORIGEN-ARP or SAS2H/ORIGEN-S) presented in References 24 and 25.

It should be noted that, while both temperature conditions above represent design criteria associated with specific analytical assumptions, only the boiling criterion represents an actual design limit. An operating limit of 205°F is imposed for conservatism. Analyses have been performed to ensure that seismic and structural integrity of the pool liner, supporting concrete, and fuel racks are not compromised at this temperature limit. Thermal - hydraulic analysis of the racks also has shown that boiling within the fuel cells does not occur with pool temperatures maintained at or below this limit, provided normal operating pool level is maintained.

In addition to the primary function of decay heat removal, the system provides for purification of the spent fuel pool water, the fuel transfer canal water, and the contents of the borated water storage tank, in order to remove fission and corrosion products and to maintain water clarity for fuel handling operations.

The system also provides inventory makeup for the fuel transfer canal and the incore instrument handling tank.

The system is designed to withstand the effects of a seismic event and meet the requirements of Duke piping Class C for Oconee.

(31 DEC 2014) 9.1 - 11

UFSAR Chapter 9 Oconee Nuclear Station Portions of the Spent Fuel Cooling 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.

9.1.3.2 System Description The Spent Fuel Cooling System (Figure 9-5, Units 1, 2 and 3) provides cooling for the spent fuel pool to remove fission product decay heat energy. System performance data are shown in Table 9-1 (Units 1 and

2) and Table 9-2 (Unit 3). Major components of the system are briefly described below.

Spent Fuel Coolers The spent fuel coolers are designed to maintain the temperature of the spent fuel pool as noted in Section 9.1.3.1. There are three coolers for Oconee 1 and 2, and three coolers for Unit 3, arranged in parallel.

Spent Fuel Coolant Pumps The spent fuel coolant pumps take suction from the spent fuel pool and recirculate the fluid back to the pool after passing through the coolers. A portion of the flow is demineralized and filtered depending on conditions. There are three pumps for Oconee Units l and 2, and three pumps for Oconee 3. The spent fuel coolant pumps are also used for filling the fuel transfer canal or incore instrumentation handling tank with borated water from the borated water storage tank.

Spent Fuel Coolant Demineralizers One spent fuel coolant demineralizer will process approximately one-half of the spent fuel pool volume in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . There is one demineralizer for Units 1 and 2, and one for Unit 3.

Spent Fuel Coolant Filters The spent fuel coolant filters are designed to remove particulate matter from the spent fuel pool water.

They are sized for the same flow rate as the demineralizers (180 gpm). There are two filters for Units l and 2, and two for Unit 3.

Borated Water Recirculation Pump This pump removes water from the borated water storage tank for demineralization and filtering. The pump may also be used while demineralizing and filtering the water in the fuel transfer canal during a transfer of fuel. It may also be used for emptying the fuel transfer canal if spent fuel coolant pumps are unavailable for use. There is one pump for Units 1 and 2, and one for Unit 3.

9.1.3.3 System Evaluation 9.1.3.3.1 Normal Operation The normal operation of the Spent Fuel Cooling System provides several functions. The most safety significant of these functions is to maintain pool inventory so that stored fuel is always covered with water. In order to protect against loss of inventory by boil-off, the system maintains the pool temperature below the design bases limits specified in Section 9.1.3.1. The system also maintains the pool clarity and chemistry at acceptable levels.

Spent fuel pool heat removal is accomplished by recirculating spent fuel coolant water through heat exchangers and then back to the pool. The spent fuel pumps take suction from the spent fuel pool and transport the flow through the coolers, which are arranged in parallel. The waste heat is removed from the shell side of the coolers by the Recirculated Cooling Water System. The cooled spent fuel pool water is then directed back to the spent fuel pool.

9.1 - 12 (31 DEC 2014)

Oconee Nuclear Station UFSAR Chapter 9 The spent fuel pool water temperature is a direct function of the decay heat load produced by the fuel in the racks, in conjunction with the heat removal capability of the spent fuel cooling system. The total heat removal capacities are the same for the Units 1 and 2 and the Unit 3 spent fuel pool coolant systems.

Both systems use the same numbers of pumps and coolers, with the same design specifications and overall equipment configurations. The expected decay heat loads vary with the number of fuel assemblies present in the pool, the burnups of the various fuel assemblies, and the post-irradiation decay times.

At the time that the Units 1 and 2 spent fuel pool was re-racked, its spent fuel cooling system was upgraded to handle the higher total heat load expected from the increased number of stored fuel assemblies. The heat removal capability of the upgraded spent fuel cooling system has been sized to meet the design limits specified in Section 9.1.3.1. A specific analysis of expected maximum normal and abnormal heat loads was performed, as described in Reference 58. The Spent Fuel Cooling System was analyzed to predict the pool temperatures which would result from these heat loads. Temperatures meet the design requirements as specified in Section 9.1.3.1. Core offloads are controlled such that the ultimate heat load from these analyses are not exceeded.

At the time that the Unit 3 spent fuel pool was re-racked, its spent fuel cooling system was upgraded to handle the higher total heat load expected from the increased number of stored fuel assemblies. The heat removal capability of the upgraded spent fuel cooling system has been sized to meet the design limits specified in Section 9.1.3.1. A specific analysis of expected maximum normal and abnormal heat loads was performed, as described in Reference 59. Again, the Spent Fuel Cooling System was analyzed to predict the pool temperatures resulting from these heat loads. These temperatures meet the design requirements as specified in Section 9.1.3.1. Core offloads are controlled such that the ultimate heat load from these analyses are not exceeded.

During an actual refueling outage for any unit at ONS, it is now common practice to offload a full core (177 fuel assemblies) into the pool. The resulting heat load under this condition will be less than the abnormal heat load cases evaluated in Sections 9.1.3.1.1 and 9.1.3.1.2 for the Units 1 and 2 fuel pool and Unit 3 fuel pool respectively. In addition, the resulting temperature will be less than 205°F in the fuel pools in the abnormal heat load case, assuming a single active failure. Normal practice at Oconee during the abnormal heat load case is to limit the maximum pool temperature to 150°F. This is accomplished via plant procedures. The seismic structural integrity of the storage racks, pools, and supporting structures has been evaluated at or above this temperature, and found to be adequate. Also, the thermal-hydraulic analysis of the storage racks indicates that localized boiling will not occur if water entering the storage cells reaches this temperature, as long as normal pool level is maintained.

A bypass purification loop is provided to maintain the purity of the water in the spent fuel pool. This loop is also utilized to purify the water in the borated water storage tank following refueling, and to maintain clarity in the fuel transfer canal during refueling. Water from the borated water storage tank or fuel transfer canal can be purified by using the borated water recirculation pump.

9.1.3.3.2 Failure Analysis An analysis of the maximum fuel cladding temperature has been performed for the postulated case of complete loss of coolant circulation to the pool. The analysis assumes maximum anticipated heat load in the pool, with the hottest assembly located in the least cooled storage area. The maximum cladding temperature will occur at the location of maximum heat flux. For a fuel assembly having the maximum value for decay heat power of 80 kw, and for an axial peak to average power density ratio of 1.2, the maximum local fuel rod heat flux is 1200 BTU/hr-ft2. Natural circulation flow rates within the storage tubes have been calculated which give confidence that convection film coefficients in excess of 50 BTU/hr-ft2 °F can be expected. Assuming this low value for conservatism, the clad surface temperature is 24°F above the coolant temperature. Because the heat flux is small, very large uncertainties in the film coefficient are acceptable without causing prohibitively high clad temperatures. For example, a reduction (31 DEC 2014) 9.1 - 13

UFSAR Chapter 9 Oconee Nuclear Station by a factor of five in the film coefficient would result in a clad surface temperature of 120°F above the coolant temperature. A reduction by a factor of ten, from 50 BTU/hr-ft2 °F to 5 BTU/hr-ft2 °F would result in a clad surface temperature of 240°F above the coolant temperature. These temperatures are below 650°F, which is the normal operating temperature of the fuel clad in the core.

9.1.3.4 Safety Evaluation The Spent Fuel Cooling System provides adequate capacity and component redundancy to assure the cooling of stored spent fuel, even when large quantities of fuel are in storage. Multiple component failures or complete cooling failures permit ample time to assure that protective actions are taken. The system is arranged so that loss of fuel pool water by piping or component failure is highly improbable.

The system performs no emergency functions. Alarms are provided to alert operator of abnormal pool level and temperature.

The Spent Fuel Cooling System has one process line connecting to the Reactor Coolant System through the SSF RC Makeup line. Its major penetration to the Reactor Building is through the fuel transfer tube.

The fuel transfer tube is isolated inside the Reactor Building by a blind flange connection in the fuel transfer canal.

9.1.4 Fuel Handling System 9.1.4.1 Design Bases 9.1.4.1.1 General System Function The fuel handling system shown on Figure 9-7 (sheets 1 & 2) is designed to provide a safe, effective means of transporting and handling fuel from the time it reaches the station in an unirradiated condition until it leaves the station after postirradiation cooling. The system is designed to minimize the possibility of mishandling or maloperations that could cause fuel assembly damage and/or potential fission product release.

Separate fuel handling equipment is provided for each reactor. A common fuel storage area serves Oconee 1 and 2, while a separate fuel storage area is provided for Oconee 3.

The reactors are refueled with equipment designed to handle the spent fuel assemblies underwater from the time they leave the reactor vessels until they are placed in a cask for shipment from the spent fuel pools. Underwater transfer of spent fuel assemblies provides an effective, economic, and transparent radiation shield, as well as a reliable cooling medium for removal of decay heat. Use of borated water assures reactor subcriticality during refueling.

9.1.4.1.2 New Fuel Storage New Fuel Storage is described in Section 9.1.1.

9.1.4.1.3 Spent Fuel Pool Each spent fuel pool is a reinforced concrete pool located in its respective Auxiliary Building. The Oconee 1, 2 pool is lined with stainless clad plate. The Oconee 3 pool is lined with stainless steel plate.

The unit 1 and 2 spent fuel pool will hold 1312 fuel assemblies. The unit 3 spent fuel pool will hold 822 assemblies plus 3 spaces for failed fuel canisters. Fuel components (such as control rods, BP's, or APSR's) requiring removal from the reactors are stored in the spent fuel assemblies or in brackets suspended from the top of the fuel racks.

9.1 - 14 (31 DEC 2014)