Text

Amendment 63 to Final Safety Analysis Report, Chapter 6, Engineered Safety Features. (Redacted)
	ML20147A023
Person / Time
Site:	Harris
Issue date:	05/15/2020
From:	; Duke Energy Progress
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML20147A016	List: ML20147A017; ML20147A018; ML20147A021; ML20147A022; ML20147A023; ML20147A024; ML20147A025; ML20147A026; ML20147A027; ML20147A028; ML20147A029; ML20147A030; ML20147A031; ML20147A032; ML20147A033; ML20147A034; ML20147A035; ML20147A036; ML20147A037; ML20147A038; ML20147A039; ML20147A040; ML20147A041; ML20147A042; RA-20-0134, Amendment 63 to Updated Final Safety Analysis Report, Drawings (Redacted) Part 2 of 2;
References
	RA-20-0134
	Download: ML20147A023 (739)
	v • d • e

{{#Wiki_filter:Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.0 ENGINEERED SAFETY FEATURES ........................................................................... 1 6.1 ENGINEERED SAFETY FEATURES MATERIALS ..................................................... 1 6.1.1 METALLIC MATERIALS ......................................................................................... 1 6.1.1.1 Materials Selection and Fabrication ................................................................. 1 6.1.1.2 Composition, Compatibility and Stability of Containment and Core Spray Coolants .......................................................................................................... 2 6.1.2 ORGANIC MATERIALS .......................................................................................... 3 6.1.2.1 Balance of Plant Organic Materials .................................................................. 3 6.1.2.2 NSSS Organic Materials .................................................................................. 4

REFERENCES:

SECTION 6.1 ................................................................................................. 5 6.2 CONTAINMENT SYSTEMS ......................................................................................... 5 6.2.1 CONTAINMENT FUNCTIONAL DESIGN ............................................................... 5 6.2.1.1 Containment Structure ..................................................................................... 5 6.2.1.2 Containment Subcompartments ..................................................................... 19 6.2.1.2a Evaluation of SGR/PUR ................................................................................. 24 6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents ....................................................................................................... 26 6.2.1.4 Mass and Energy Release Analysis For Postulated Secondary System and Pipe Ruptures ......................................................................................... 37 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies of Emergency Core Cooling System ................................................ 41 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEM ....................................................... 43 6.2.2.1 Design Bases ................................................................................................. 44 6.2.2.2 System Design ............................................................................................... 45 6.2.2.3 System Design Evaluation ............................................................................. 52 6.2.2.4 Testing and Inspection ................................................................................... 56 6.2.2.5 Instrumentation Requirements ....................................................................... 57 6.2.3 SECONDARY CONTAINMENT FUNCTIONAL DESIGN...................................... 58 6.2.4 CONTAINMENT ISOLATION SYSTEM ................................................................ 58 6.2.4.1 Design Bases ................................................................................................. 58 6.2.4.2 System Design ............................................................................................... 60 Amendment 63 Page i of v

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.4.3 Design Evaluation .......................................................................................... 68 6.2.4.4 Tests and Inspections .................................................................................... 69 6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT ......................................... 69 6.2.5.1 Design Bases ................................................................................................. 70 6.2.5.2 System Design ............................................................................................... 73 6.2.5.4 Test and Inspections ...................................................................................... 76 6.2.5.5 Instrumentation Requirements ....................................................................... 76 6.2.5.6 Materials ......................................................................................................... 76 6.2.6 CONTAINMENT LEAKAGE TESTING .................................................................. 76 6.2.6.1 Containment Integrated Leakage Rate Test (Type A Test) ........................... 76 6.2.6.2 Containment Penetration Leakage Rate Tests (Type B Tests) ...................... 80 6.2.6.3 Containment Isolation Valve Leakage Rate Tests (Type C Tests)................. 81 6.2.6.4 Scheduling and Reporting of Periodic Tests .................................................. 83

REFERENCES:

SECTION 6.2 ............................................................................................... 84 APPENDIX 6.2A ...................................................................................................................... 86

REFERENCES:

APPENDIX 6.2A ........................................................................................... 86 6.3 EMERGENCY CORE COOLING SYSTEM ................................................................ 87 6.3.1 DESIGN BASES .................................................................................................... 87 6.3.2 SYSTEM DESIGN ................................................................................................. 90 6.3.2.1 Schematic Piping and Instrumentation Diagrams .......................................... 90 6.3.2.2 Equipment and Component Descriptions ....................................................... 91 6.3.2.3 Applicable Codes and Classifications .......................................................... 100 6.3.2.4 Material Specifications and Compatibility ..................................................... 100 6.3.2.5 System Reliability ......................................................................................... 100 6.3.2.6 Protection Provisions ................................................................................... 105 6.3.2.7 Provisions for Performance Testing ............................................................. 105 6.3.2.8 Manual Actions ............................................................................................. 106 6.3.3 PERFORMANCE EVALUATION ......................................................................... 113 6.3.3.1 Inadvertent Opening of a Steam Generator Relief or Safety Valve.............. 114 6.3.3.2 Small Break LOCA ....................................................................................... 115 6.3.3.3 Large Break LOCA ....................................................................................... 115 Amendment 63 Page ii of v

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.3.3.4 Major Secondary System Pipe Failure ......................................................... 116 6.3.3.5 Steam Generator Tube Failure .................................................................... 117 6.3.3.6 Existing Criteria Used to Judge the Adequacy Of the ECCS. Criteria from 10CFR50.46 ................................................................................................ 118 6.3.3.7 Use of Dual Function Components .............................................................. 119 6.3.3.8 Limits on System Parameters ...................................................................... 120 6.3.3.9 Time Sequence for the Operation of the ECCS Components ...................... 120 6.3.4 TEST AND INSPECTIONS ................................................................................. 121 6.3.4.1 ECCS Performance Tests ............................................................................ 121 6.3.4.2 Reliability Tests and Inspections .................................................................. 122 6.3.5 INSTRUMENTATION REQUIREMENTS ............................................................ 124 6.3.5.1 Temperature Indication ................................................................................ 124 6.3.5.2 Pressure Instrumentation ............................................................................. 124 6.3.5.3 Flow Indication ............................................................................................. 125 6.3.5.4 Level Indication ............................................................................................ 125 6.3.5.5 Valve Position Indication .............................................................................. 126

REFERENCES:

SECTION 6.3 ............................................................................................. 127 6.4 HABITABILITY SYSTEMS ........................................................................................ 127 6.4.1 DESIGN BASIS ................................................................................................... 127 6.4.2 SYSTEM DESIGN ............................................................................................... 130 6.4.2.1 Control Room Envelope ............................................................................... 130 6.4.2.2 Ventilation System Design ........................................................................... 130 6.4.2.3 Leak Tightness ............................................................................................. 131 6.4.2.4 Interaction with Other Zones and Pressure-Containing Equipment ............. 132 6.4.2.5 Shielding Design .......................................................................................... 132 6.4.3 SYSTEM OPERATIONAL PROCEDURES ......................................................... 132 6.4.4 DESIGN EVALUATION ....................................................................................... 133 6.4.4.1 Radiological Protection ................................................................................ 133 6.4.4.2 Toxic Gas Protection .................................................................................... 133 6.4.5 TESTING AND INSPECTIONS ........................................................................... 134 6.4.5.1 Emergency HEPA/Charcoal Filter Trains ..................................................... 134 Amendment 63 Page iii of v

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.4.5.2 Water Chillers ............................................................................................... 135 6.4.5.3 Fan or Fan Coil Units ................................................................................... 135 6.4.5.4 Pumps .......................................................................................................... 135 6.4.5.5 Considerations Leading to the Selected Test Frequency............................. 135 6.4.6 INSTRUMENTATION REQUIREMENT .............................................................. 135 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS ................................. 137 6.5.1 ENGINEERED SAFETY FEATURE (ESF) FILTER SYSTEMS .......................... 137 6.5.1.1 Design Bases ............................................................................................... 137 6.5.1.2 System Design ............................................................................................. 139 6.5.1.3 Design Evaluation ........................................................................................ 144 6.5.1.4 Test and Inspection ...................................................................................... 145 6.5.1.6 Materials ....................................................................................................... 147 6.5.2 CONTAINMENT SPRAY SYSTEM ..................................................................... 147 6.5.2.1 Design Bases ............................................................................................... 147 6.5.2.2 System Design ............................................................................................. 148 6.5.2.3 Design Evaluation ........................................................................................ 149 6.5.2.4 Testing and Inspection ................................................................................. 153 6.5.2.5 Instrumentation Requirement ....................................................................... 153 6.5.2.6 Materials ....................................................................................................... 153 6.5.3 FISSION PRODUCT CONTROL SYSTEMS....................................................... 154 6.5.3.1 Primary Containment ................................................................................... 154 6.5.3.2 Secondary Containment ............................................................................... 154 6.5.4 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM ................. 154

REFERENCES:

SECTION 6.5 ............................................................................................. 155 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS ........................... 155 6.6.1 COMPONENTS SUBJECT TO EXAMINATION.................................................. 155 6.6.2 ACCESSIBILITY .................................................................................................. 155 6.6.3 EXAMINATION TECHNIQUES AND PROCEDURES ........................................ 156 6.6.4 INSPECTION INTERVALS ................................................................................. 156 Amendment 63 Page iv of v

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.6.5 EXAMINATION CATEGORIES ........................................................................... 156 6.6.6 EVALUATION OF EXAMINATION RESULTS .................................................... 156 6.6.7 SYSTEM PRESSURE TESTS ............................................................................ 157 6.6.8 AUGMENTED INSERVICE INSPECTION TO PROTECT AGAINST POSTULATED PIPING FAILURES ..................................................................... 157 6.7 MAIN STEAM LINE ISOLATION VALVE LEAKAGE CONTROL SYSTEM.............. 157 Amendment 63 Page v of v

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.0 ENGINEERED SAFETY FEATURES 6.1 ENGINEERED SAFETY FEATURES MATERIALS 6.1.1 METALLIC MATERIALS 6.1.1.1 Materials Selection and Fabrication Typical materials specifications used for components in the engineered safety features (ESF) are listed in Table 6.1.1-1. For NSSS supplied equipment, this list of materials may not be totally inclusive; however, the listed specifications are representative of those materials used. Identification of the actual materials used in Class 2 and 3 components is available in the SHNPP Site QA records. The utilized materials conform to the applicable requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. Table 6.1.1-1 lists materials utilized by ESF components within the Containment that would be exposed to core cooling water and containment sprays in the unlikely event of a loss-of-coolant accident (LOCA). These components are manufactured primarily of stainless steel or other corrosion-resistant material. The integrity of the materials of construction for ESF equipment when exposed to post-design basis accident (DBA) conditions has been evaluated. Post-DBA conditions were conservatively represented by test conditions. The test program performed by Westinghouse considered spray and core cooling solutions of the design chemical compositions, as well as the design chemical compositions contaminated with corrosion and deterioration products which may be transferred to the solution during recirculation. The effects of sodium (free caustic), chlorine (chloride), and fluorine (fluoride) on austenitic stainless steels were considered. Based on the results of this investigation, as well as testing by Oak Ridge National Laboratory and others, the behavior of austenitic stainless steels in the post-DBA environment will be acceptable. The inhibitive properties of alkalinity (hydroxyl ion) with respect to chloride cracking have been demonstrated. All parts of components in contact with borated water are fabricated of, or clad with, austenitic stainless steel or equivalent corrosion-resistant material. The integrity of the safety related components of the ESF is maintained during all stages of component manufacture. Austenitic stainless steel is utilized in the final heat-treated condition as required by the respective ASME Code, Section II, material specification for the particular type or grade of alloy. Furthermore, it is required that austenitic stainless steel materials used in the ESF components be handled, protected, stored, and cleaned according to recognized and accepted methods that are designed to minimize contamination which could lead to stress corrosion cracking. The rules covering these controls are discussed in Section 5.2.3. Additional information concerning austenitic stainless steel, including the avoidance of sensitization and the prevention of intragranular attack, can be found in Section 5.2.3. The welding materials used for joining the ferritic base materials of the ESF conform to or are equivalent to ASME Code Section II, Part C, Material Specifications SFA 5.1, 5.5, 5.17, 5.18 and 5.20. The welding materials used for joining nickel-chromium-iron alloy in similar base material combination and for joining dissimilar ferritic or austenitic base material combination conform to ASME Code, Section II, Part C, Material Specifications SFA 5.11 and 5.14. The welding materials used for joining the austenitic stainless steel base material conform to ASME Code, Section II, Part C, Material Specifications SFA 5.4 and 5.9. These materials are tested and qualified to the requirements of the ASME Code and are used in procedures which have Amendment 63 Page 1 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 been qualified to these same rules. The methods utilized to control delta ferrite content and to avoid hot cracking (fissuring) in austenitic stainless steel weldments are discussed in Section 5.2.3 for Westinghouse supplied equipment, and in Sections 10.3 and 1.8 (Regulatory Guide 1.31) for piping and Ebasco specified equipment. Materials for Class 2 and 3 components are selected for their compatibility with core and containment spray solutions, as described in ASME Code, Section III, Articles NB-2160 and NB-3120; the materials are selected from those which are included in Appendix I to Section III. The mechanical properties of materials specified for use in Class 2 and 3 components are as indicated in ASME Code, Section III, Appendix I or ASME Code, Section II, Parts A, B or C All materials for ESF components which are in contact with core cooling and/or containment spray water are considered compatible with the cooling solutions as described below: a) Austenitic stainless steels and nickel base alloys are not subject to significant corrosion in borated water or borated water with sodium hydroxide additives. b) Any carbon steel components, requiring protective coatings will be coated to meet the intent of Regulatory Guide 1.54. The integrity of ESF components is maintained during all stages of component manufacture and reactor construction. Specific assurance of integrity is based on compliance with Regulatory Guides 1.31, 1.36, 1.37 and 1.44 as described in Section 1.8. Additionally, all austenitic stainless steels are provided in the solution annealed condition. Yield strengths for these materials are of the order of 30,000 to 50,000 psi. No cold-worked austenitic stainless steels having yield strengths greater than 90,000 psi are used for components of the ESF. Any cold bent piping is re-solution annealed after cold bending, except where the piping is bent to a radius of at least 20 times the pipe radius, in which case the resulting strain in the outer pipe fibers is under 2.5 percent, which causes no significant increase in yield strength. Information regarding the selection, procurement, testing, storage, and installation of nonmetallic thermal insulation, and demonstrating that the leachable concentrations of chloride, fluoride, sodium, and silicate are comparable to the recommendations of the Regulatory Guide 1.36, is contained in Section 5.2.3 for Westinghouse supplied insulation and in Section 1.8 for all other insulation. Use of aluminum and zinc will be minimized in the Containment. An aluminum and zinc inventory in the Containment is given in Table 6.1.1-2. 6.1.1.2 Composition, Compatibility and Stability of Containment and Core Spray Coolants The pH of the containment spray will be adjusted during the injection mode by the addition of a 27-29 weight-percent sodium hydroxide (NaOH) solution to provide a minimum pH of 7.0. A discussion of the NaOH addition design basis is provided in Section 6.5.2.3.3. In no case will the solution pH fall outside the range of 7.0 to 11.0. The refueling water storage tank is the source of borated cooling water during injection. The boron concentration, as boric acid, is 2,400 - 2,600 ppm. The tank is maintained above 40F, thus ensuring that the boric acid remains soluble. Amendment 63 Page 2 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 In order to ensure materials compatibility during storage, the sodium hydroxide chemical additive is contained in a stainless steel tank. The spray additive solution is not corrosive to the stainless steel components of the system with which it comes into contact. The spray and sump solutions will tend to severely corrode aluminum alloys, but will not attack stainless steel or copper-nickel alloys. Hydrogen release within the Containment due to corrosion of materials by the sprays and cooling water in the event of a LOCA is controlled as described in Section 6.2.5. The use of aluminum within the Containment is minimized to the greatest extent practical, thereby precluding concern over excessive hydrogen generation due to the corrosion of aluminum. The vessels used for storing engineered safety features coolant include the accumulators, the boron injection tank, and the refueling water storage tank (RWST). The accumulators are carbon steel clad with austenitic stainless steel and the boron injection tank is austenitic stainless steel. Because of the corrosion resistance of these materials, significant corrosive attack of the storage vessels is not expected. The accumulators are vessels filled with borated water and pressurized with nitrogen gas. The boron concentration, as boric acid, is 2400 - 2600 ppm. Samples of the solution in the accumulators are taken periodically for checks of boron concentration. Principal design parameters of the accumulators are listed in Table 6.3.2-1. Principal design parameters of the boron injection tank are listed in Table 6.3.2-1. 6.1.2 ORGANIC MATERIALS 6.1.2.1 Balance of Plant Organic Materials Significant quantities of organic materials that exist within the primary containment consist of lubricants and protective coatings for containment surfaces, equipment and pipe. Protective coatings applied to major equipment, piping, steel surfaces and concrete surfaces have been applied in accordance with the applicable guidelines included in ANSI N101.4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," and Regulatory Guide 1.54," Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," with the exception of areas which are not accessible for the required preparation, application, or inspection. Such areas will be prepared and coated as best as possible with approved coatings using manufacturers recommendations and industry standards as guidelines. In addition, the coatings used to meet the requirements of ANSI N101.2-1972 "Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities" for the design basis accident (DBA) are resistant to an integrated radiation exposure of 4.2 x 106 rads. over a period of 25 hours (5.5 x 105 rad./hr. initial dose rate). In the event coatings repair work (touch up) is required, (for both steel and concrete surfaces), the damaged coatings will be replaced and recoated in accordance with the manufacturer's recommendations. Film thickness is checked with a nondestructive film thickness gauge, where applicable. Amendment 63 Page 3 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Based on tests conducted in accordance with the requirements of ANSI N101.2 and ANSI N512 "Protective Coatings (Paints) for the Nuclear Industry," no materials (gases or others) were reported to be released and no decomposition by radiation or chemical reaction was reported when exposed to 1 x 109 rads. Test panels were also inspected for any breakdown on the coating system, i.e., flaking, peeling, delamination and blistering (allowable blisters - few intact, Size No. 4 per ASTM D714). None of the foregoing occurred. Compliance with Regulatory Guide 1.54 is discussed in Section 1.8. Any equipment, excluding small valves, pumps, motors and other small miscellaneous items, not coated in accordance with Regulatory Guide 1.54 will be recoated at the site with an acceptable system. All thermal insulation jacketing material will be stainless steel. This is applicable to all insulated equipment including pressurizers and steam generators. The design life of all applied thermal insulation is 40 years. The construction is such that it will not sag, settle, corrode or disintegrate during its design life. The aging management reviews for insulation within the scope of License Renewal determined that the insulation has no aging effects requiring management. Therefore, the insulation is capable of performing its intended function through the period of extended operation. Quantities of miscellaneous organic materials such as diaphragms, valve packing and O-rings for mechanical nuclear equipment are not considered significant. The total weight of electrical cable insulation materials and their chemical compositions, along with a breakdown of cable diameters and associated conductor cross sections is given in Table 6.1.2-2. The RCPs are not required following a DBA, therefore, the lubricating oil need not perform its function under DBA conditions. Likewise, steam generator snubbers are not required under post DBA conditions and therefore the snubber oil need not perform its function. During a DBA, the snubber and oil will perform satisfactorily. 6.1.2.2 NSSS Organic Materials Quantification of significant amounts of protective coatings on Westinghouse supplied components located inside the Containment Building is given in Table 6.1.2-1; the painted surfaces of Westinghouse supplied equipment comprise a small percentage of the total painted surfaces inside Containment. For large equipment requiring protective coatings (specifically itemized in Table 6.1.2-1), Westinghouse specifies or approves the type of coating systems utilized; requirements with which the coating system must comply are stipulated in Westinghouse specifications. For these components, the generic types of coatings used are zinc rich silicate or epoxy based primer with or without chemically-cured epoxy and epoxy modified phenolic top coat. Amendment 63 Page 4 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The remaining equipment requires protective coatings on much smaller surface areas and is procured from numerous vendors; for this equipment, Westinghouse specifications require that high quality coatings be applied using good commercial practices. Table 6.1.2-1 includes identification of this equipment and total quantities of protective coatings on such equipment. Protective coatings for use in the Containment have been evaluated as to their suitability in post-design basis accident conditions. Tests have shown that certain epoxy and modified phenolic systems are satisfactory for in containment use. This evaluation (Reference 6.1.2-1) considered resistance to high temperature and chemical conditions anticipated during a LOCA as well as high radiation resistance. Information regarding compliance with Regulatory Guide 1.54 is discussed in Section 1.8. Further compliance information has been submitted to the NRC for review via Reference 6.1.2-2 and accepted via Reference 6.1.2-3.

REFERENCES:

SECTION 6.1 6.1.2-1 Picone, L. F., "Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198-L (Proprietary), April, 1968 and WCAP-7825 (Non-Proprietary), December, 1971. 6.1.2-2 Letter NS-CE-1352, dated February 1, 1977, C. Eicheldinger (Westinghouse) to C. J. Heltemes, Jr. (NRC). 6.1.2-3 Letter dated April 27, 1977, C. J. Heltemes, Jr. (NRC) to C. Eicheldinger (Westinghouse). 6.1.2-4 Whyte, D. D. and Picone, L. F., "Behavior of Austenitic Stainless Steel in Post Hypothetical Loss-Of-Coolant Accident Environment," WCAP-7798-L (Proprietary), November, 1971 and WCAP-7803 (Non-Proprietary), December, 1971. 6.2 CONTAINMENT SYSTEMS 6.2.1 CONTAINMENT FUNCTIONAL DESIGN 6.2.1.1 Containment Structure 6.2.1.1.1 Design Bases The containment systems protect the public from the consequences of any postulated break in the Reactor Coolant System. The containment systems consist of the steel lined concrete Containment Building, and the Engineered Safety Feature Systems which include the Containment Heat Removal System (Containment Spray System and Containment Cooling System), the Containment Isolation System, and the Containment Hydrogen Control System. The containment structure provides biological shielding and missile protection for the Nuclear Steam Supply System. A physical description of the Containment and the design criteria relating to construction techniques, static loads, and seismic loads are provided in Section 3.8. This section pertains to those aspects of containment design, testing, and evaluation that relate to the accident mitigation function. Amendment 63 Page 5 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The containment structure is designed to withstand the pressure and temperature transient calculated to exist after a design basis accident (DBA). Post-accident conditions are determined by evaluating the combined influence of the energy sources, heat sinks, and engineered safety features (ESF) operation. The capability of the containment structure to maintain design leaktight integrity and to provide a predictable environment for the operation of ESF systems is ensured by a comprehensive design, analysis, and testing program. This program considers the results of both the peak containment pressures and temperatures resulting from a LOCA or a main steam line break (MSLB) and the maximum containment external (differential) pressure resulting from inadvertent containment heat removal system operation that reduces containment internal pressure below outside atmospheric pressure. The containment systems are designed to provide protection to the public from the consequences of a loss-of-coolant accident (LOCA) up to and including a double-ended rupture of the largest reactor coolant pipe assuming unobstructed discharge from both ends coincident with the safe shutdown earthquake (SSE), loss of normal offsite power, and any single active component failure. The containment structure and the engineered safety features ensure that the radiological exposure to the public resulting from such an occurrence is below the guidelines established in 10CFR 50.67. The spectrum of postulated accidents considered in determining the design containment peak pressure and temperature, the subcompartment peak pressure, the containment external (differential) pressure, and the ECCS minimum containment pressure analysis are summarized in Table 6.2.1-1. The spectrum of break sizes was chosen to establish the upper bounds of containment pressure and temperature following a design basis accident (DBA). For postulated subcompartment pipe break accidents, a discussion of the criteria for selecting break locations is given in Section 6.2.1.2. The accident controlling design for each of the categories of containment peak pressure, containment peak temperature, subcompartment peak pressure, containment external (differential) pressure, and containment minimum pressure is defined as a design basis accident (DBA), and is that case which produces the most severe loadings for the spectrum of accidents postulated. The margin between values calculated for a DBA and design values is established by comparing Tables 6.2.1-2 and 6.2.1-3. Table 6.2.1-2 defines the DBA for each design category and Table 6.2.1-3 gives the margin and the formula used as the bases for calculating margin. For the containment structure peak pressure analysis, subcompartment peak pressure analyses, containment peak temperature analysis, and the ECCS minimum containment pressure analysis, it is assumed that each accident is concurrent with the most limiting single active failure. No two accidents are assumed to occur simultaneously or consecutively. For the LOCA maximum injection case, one containment spray system train and four containment fan coolers were assumed to operate in conjunction with both trains of safety injection (i.e., one containment spray pump failure). For the LOCA minimum injection case, one containment spray system train and two containment fan coolers were assumed to operate in conjunction with one train of safety injection (i.e., one diesel generator failure). Amendment 63 Page 6 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The time dependent LOCA mass and energy release for the postulated accidents under the categories of containment peak pressure and temperature analyses are referenced in Table 6.2.1-1. The computer codes and assumptions used for deriving each of the mass and the energy release tables are discussed in Section 6.2.1.3. Energy released to the containment atmosphere as a result of the postulated pipe break accidents is transferred to the containment sump by the Containment Heat Removal System discussed in Section 6.2.2. During recirculation, energy is removed from the containment sump and atmosphere by the Residual Heat Removal System (RHRS) used in conjunction with the Containment Cooling System (i.e., fan coolers). For the purpose of the LOCA containment peak pressure analysis, the most restrictive single active failure is the failure of one onsite diesel generator (and therefore, one containment spray train and two fan coolers) resulting in minimum containment heat removal capability. Assuming this most restrictive single active failure, the Containment Heat Removal System is capable of reducing post-LOCA pressures to less than 50 percent of the containment peak calculated pressure within 24 hours following the postulated accident. This capability, demonstrated by containment pressure response curves, is consistent with the offsite radiological consequences discussed in Chapter 15. The most severe containment peak pressure results from a LOCA while the most severe temperature results from a main steam line break (MSLB). The most limiting single active failure for the MSLB's are discussed in Sections 6.2.1.1.3 and 6.2.1.4. The time dependent MSLB mass and energy release for the postulated accidents under the categories of containment peak pressure and temperature analyses are referenced in Table 6.2.1-1 and discussed in Section 6.2.1.4. The analysis of containment minimum pressure is based on confirming the ECCS core reflood capability under the conservative set of assumptions that maximize the heat removal effectiveness of ESF systems, structural heat sinks, and other potential heat removal processes. These assumptions are discussed in Section 6.2.1.5. 6.2.1.1.2 Design Features The design bases and design measures taken to assure that the containment structure is adequately protected against the dynamic effects of postulated accidents are discussed in Sections 3.5 and 3.6. The codes, standards, and guides applied in the design of the containment structure and internal structures are described in Section 3.8. Redundant containment vacuum breakers have been provided for protection against loss of containment integrity under external loading conditions. Calculations of containment pressure following an inadvertent operation of the Containment Spray System have resulted in pressures within the containment design external (differential) allowable pressure. Details of this evaluation are provided in Section 6.2.1.1.3. The margin between calculated and design pressure differentials is shown in Table 6.2.1-3. The Containment Cooling System, discussed in Section 6.2.2, maintains the containment and subcompartment atmospheres within required pressure and temperature limits during normal plant operation. This system recirculates air in the upper Containment through fan coolers which are located above the operating floor. The Containment Cooling System and the Amendment 63 Page 7 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 containment ventilation system during normal plant operation is functionally capable of maintaining the pressure and temperature within the limits used for equipment design and assumed for DBA analyses. The systems used for normal containment ventilation include the Containment Purge System and Containment Cooling System. The limiting containment conditions for normal plant operation are contained in the Technical Specifications. During the injection phase, water entering the reactor cavity is trapped from returning to the containment recirculation sump. The volume of trapped liquid has been determined to be 53,600 gals. This quantity of water has been accounted for in determining the available NPSH for the recirculation pumps. Water entering the refueling cavity will be directed to the recirculation sump via a locked open floor drain. This drain will be manually closed during refueling operations and a separate normally closed drain will be provided which will direct decontamination washdown to the equipment drain. The arrangement is shown on Figure 9.1.3-3. The range, accuracy, and response time of instrumentation provided which is capable of operating in the post-accident environment for monitoring containment atmosphere and containment sump water temperature, is listed in Table 6.2.1-65. Continuous indication and display of containment wide range [(-5) to 135 psig] pressure will be provided in the Control Room. This recorded range will be three (3) times the design pressure of the Shearon Harris concrete containment of 45 psig. Containment wide range pressure monitoring instrumentation will consist of two (2) redundant Class 1E pressure transmitters. The transmitters will be physically located inside the containment building. The accuracy of the transmitter to be used is +/- 0.5 percent (normal) and +/- 10 percent (accident) of calibrated span. The pressure transmitter output signal will be processed by a process instrumentation control system (PIC) which in turn will furnish signals for the Class 1E indicator and the Safety Parameters Display System CRT in the control room. The operator has the capability for continuous recording if desired for trending. The containment pressure monitoring channels shall meet the design and qualification criteria of Reg. Guide 1.97, Revision 3 Appendix A. Additionally, for a narrower range, redundant Class 1E indicators and non-1E (seismic only) recorders whose inputs are derived from signals used in the Engineered Safety Features System are provided in the Control Room for a maximum available pressure range of 0 - 55 psig. Another set of redundant indicators which monitor the effectiveness of the Containment Vacuum Relief System are provided on the main control board panel having a range of +/- 5 inches of water column. Continuous Class 1E indication of containment water level is provided in the Main Control Room as follows: Containment Recirculation Sumps (Narrow Range Level Monitoring) - The recirculation sumps are provided with one Level Transmitter each (LT-7160A/B). The 219'4" and 224'4" Elevations correspond to 0% and 100% indicated level, respectively. The corresponding level indicators Amendment 63 Page 8 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 are LI-7160A/B. A low level alarm occurs at 43% indicated level (i.e., 221.5') if a recirculation sump isolation valve is open. These low level alarms are annunciated on monitor light boxes in the Main Control Room. Containment Wide Range Sump Level Monitoring - The containment wide range level instruments are provided for post-accident monitoring and consist of Level Transmitters LIT-7162A/B. The 211'9-3/4" and 230'3-3/4" Elevations correspond to 0% and 100% indicated levels, respectively. The corresponding level indicators are LI-7162A/B. A high level alarm occurs at 88% indication level (i.e., 228'1"). These alarms are annunciated via the plant computer. Non-1E indication of containment sump level is also available from the same level instrumentation via isolated inputs to the plant computer. Qualification is in accordance with the criteria for Class IE transmitters located inside Containment. The narrow range monitors will meet the requirement of Regulatory Guide 1.89. 6.2.1.1.3 Design Evaluation 6.2.1.1.3.1 Containment Pressure - Temperature Analysis In the event of a postulated loss-of-coolant accident (LOCA), or main steam line break (MSLB), the release of coolant from the rupture area causes the high pressure fluid to flash to steam. This release of mass and energy raises the temperature and pressure of the containment atmosphere. The severity of the resulting temperature and pressure peaks developed depends upon the nature, location, and size of the postulated rupture. In order to establish the controlling rupture for containment design, the spectrum of primary and secondary breaks described in Table 6.2.1-1 were analyzed to determine their significance in selecting the containment design basis accidents. Table 6.2.1-4 presents the results (the calculated pressure, temperature, time of peak pressure) of these analyses and the containment design basis accidents are noted in Table 6.2.1-2. Additional information for the selection of break size, location, etc. is provided in 6.2.1.3 (LOCA) and 6.2.1.4 (MSLB). The calculated transients following a postulated accident are a direct consequence of the energy balance within the Containment. Of particular importance are the initial conditions postulated at the start of the accident, the ability of the heat sinks within the Containment to absorb energy during the accident, and the capability of the Containment Heat Removal System to reduce the total energy within the Containment, thus bringing the containment heat sinks, sump water, and atmosphere into thermal equilibrium. The containment pressure analysis input data are based upon plant design features. A conservative prediction of consequences was assured by determining upper and lower bounding values of containment initial conditions, geometric parameters, and thermodynamic properties, and by applying these values in the manner producing maximum pressure and temperature results. Amendment 63 Page 9 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.1.1.3.2 LOCA Analysis LOCA analyses were re-performed for SGR/PUR. The initial conditions within the Containment and Reactor Coolant Systems prior to accident initiation are given in Table 6.2.1-5. The minimum containment volume is assumed to be at the highest value of operating pressure for both LOCA peak pressure and temperature cases since a sensitivity study indicates that the initial Containment pressure had negligible impact on the peak temperature case, but does result in higher peak pressures with a high initial Containment pressure. Sensitivity studies are performed to determine the conservative initial humidity and temperature for each analysis. The maximum operating temperature was assumed to exist in all heat sinks. The heat sink temperatures were chosen to reflect the initial containment air temperature for each case. For the LOCA analysis, the assumed reactor coolant system inventory is based on design overpower of 102 percent with normal liquid levels. The surface area of the liquid pool formed in the bottom of the containment following a LOCA is calculated with GOTHICs default calculations. The following additional assumptions are made in performing the containment LOCA analysis: a) No leakage into or out of the Containment occurs; b) The mass diffusion calculation of the GOTHIC computer code is used for the heat transfer coefficient between the containment vapor and the sump liquid region; and c) Hot metal surfaces in the NSSS not cooled by safety injection water, such as the reactor vessel above the nozzles, are simulated as hot walls in contact with the containment steam-air mixture. A sensitivity study was performed varying free volume and the size of the heat sinks within the Containment (Reference 6.2.1-1). The results show that the containment free volume is the principal factor responsible for large changes in the peak containment pressure and temperature responses. In the sensitivity study, the surface area of the heat sinks in the Containment was varied over a range of +/- 20 percent. Two sets of analyses were done, one in which only the surface area of the internal heat sinks was varied and the other in which all heat sinks, including the size of the containment height were varied with a proportional change in the free volume. For the change of +/- 20 percent in the surface area of the internal heat sinks the peak pressure was found to vary by +/- 5 percent. For the change of all the heat sinks and a proportional variation of the free volume of about 25 percent, the peak pressure varied by about +/- 25 percent. For the purpose of the LOCA analyses, the ECCS and the Containment Heat Removal Systems were assumed to operate maximizing the containment pressure response. The operating assumptions are discussed below. For the Containment Heat Removal System, minimum system capacity shown in Table 6.2.1-6 is conservative for calculating the containment peak pressure response. Therefore, the Containment Heat Removal System was assumed to be affected by the most restrictive single active failure which has been determined to be a loss of a diesel generator (one containment spray train and two fan coolers). For LOCA analyses, the following describes the conservative assumptions made with respect to ESF system operations and parameters: Amendment 63 Page 10 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

1. The contents of all 3 accumulators including nitrogen gas discharge into the reactor coolant system when reactor coolant system pressure drops below the tank pressure setpoint.

2. All ECCS pumps are assumed to operate for the maximum injection case, and only one train of ECCS pumps are assumed to operate for the minimum injection case.

3. One containment spray pump operates and sprays 1730 gpm of water at 125°F into the Containment until the start of recirculation. This assumes the limiting single active failure of one containment spray train a maximum refueling water storage tank temperature, minimum refueling water storage tank level, and peak containment pressure equal to design pressure of containment (45 psig.)

For the maximum safety injection case, four containment fan coolers are assumed to operate. For the minimum safety injection case, two containment fan coolers are assumed to operate.

4. For the maximum injection case, both residual heat removal pumps circulate water through their associated RHR heat exchangers during recirculation. For the minimum injection case, one RHR pump circulates water through its associated RHR heat exchanger during recirculation.

The faulted overall heat transfer coefficient (UA) of the heat exchanger is calculated internally by the GOTHIC code, and the heat exchanger is assumed to be supplied with cooling water flowing with the maximum recirculation component cooling water temperature. Refer to Table 6.2.1-6.

5. The time until initiation of the recirculation mode is calculated on the basis of a minimum refueling water storage tank volume and ESF pumps operating as specified in b) and c) above. This volume is assumed to be injected into Containment before a recirculation actuation signal is generated. At this time, the containment spray water is drawn from the containment sump.

6. The Containment Spray System was assumed to commence spray at 58.4 seconds following a LOCA. This time delay takes into account signal process time, diesel starting time, sequencer delay time, breaker closing time, pump start up time and spray line fill up time. No credit was applied for partial containment spray heat removal during fill up of the spray headers.

The sizing of the Containment Spray System was based on the heat removal rate necessary to keep the peak pressure reached during a LOCA less than the design pressure of the Containment. The peak pressure occurs during the DBA LOCA blowdown or reflood phase. This peak is larger than the rise in containment pressure that occurs during the recirculation period, when the containment energy balance is coming into equilibrium.

7. The Containment Isolation Activation Signal (CIAS), (T), H1-1 (3-0 psig) setpoint is reached within 1 second after the postulated ruptures for the most severe temperature and pressure cases. The fan coolers are assumed in full operation 70 seconds (Reference ESR9400546) after the CIAS setpoint. Note: Although the service water Amendment 63 Page 11 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 valves have stroked sufficiently at 70 seconds to allow full flow to the fan coolers, the occurrence of two-phase flow during a LOCA coincident with a LOOP delays full flow to the coolers until 110 seconds. This input has been included in the latest containment accident analysis for both MSLB and LOCA. The containment heat sink data used in accident analyses are described in Tables 6.2.1-7 and 6.2.1-8. Table 6.2.1-7 is a detailed list of the geometry of each heat sink and Table 6.2.1-8 describes the resulting simplified heat sink models used for computer input. Node spacing used for concrete, steel and steel-lined concrete heat sinks is fine enough to ensure an accurate representation of the thermal gradient in each slab. The given values for surface area and thickness reflect the total areas and surface area weighted thicknesses for all steel exposed to the containment atmosphere from all sources. These sources include structural steel, polar crane and moving platform structures, instrumentation and control equipment (cabinet, tubes), HVAC equipment (duct, fan coolers, valves), refueling machine, miscellaneous piping, and containment penetration nozzles. All steel is coated with a layer of paint and finisher. All steel except the primary containment liner is assumed to be insulated on one side and in contact with the containment atmosphere by a condensing heat transfer coefficient on the other. The initial temperature is 135°F. The given values of the concrete surface area and thickness reflect total areas and surface-area weighted thicknesses of all concrete within the Containment. This concrete includes the unlined reactor cavity wall, primary and secondary shield, walls, pressurizer room, regenerative heat exchanger room, valve room, pipe tunnel, reactor sump pump wall, and the steam generator foundation. All concrete is coated with paint. The concrete is assumed to have a zero temperature gradient at the center. The concrete is exposed to the containment atmosphere with a condensing heat transfer coefficient or to the sump water with a free convection heat transfer coefficient. Table 6.2.1-8 also lists values for surface area and thickness of the remaining heat sinks which represent the total surface area and mass-weighted mean thickness of the similar heat sinks in the Containment. The initial temperatures given likewise reflect a mass-weight average. Initially, a free convection heat transfer correlation was used for the heat sink with initial thermal gradients. A complete list of the thermophysical properties used in the analysis is also given in Table 6.2.1-

8. No credit was taken for heat transfer to reinforcing steel in the internal concrete structures and a low value of thermal conductivity was used for these structures.

The node spacing within a heat sink is dependent upon the gradient of temperature within the sink. To ensure an accurate representation of the temperature gradient, a maximum number of node points are placed where the temperature gradient is at a maximum. Too few node points simulate a more gradual slope implying excessive energy stored within the heat sink. An accurate energy balance ensures the adequacy of the node spacing. The node spacing for all conductors is such that the calculated Biot number for each node is less than 0.1. Amendment 63 Page 12 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The high thermal conductivity and relative thinness of the steel heat sinks results in a rapidly uniform temperature distribution throughout the sink. It is, therefore, only necessary to provide for sufficient nodes to adequately define its relative thickness (in relation to other heat sinks). A complete tight contact between the steel liner and the concrete wall has been assumed in the analysis since no steel liner buckling has been calculated to occur (see Section 3.8.1). Blowdown mass and energy release rates for LOCA are discussed in Section 6.2.1.3. The containment accident analyses are performed using the GOTHIC computer code (Reference 6.2.1-23). A description of the computer code is contained in Appendix 6.2A. The containment pressure and temperature response and containment sump water temperature response versus time are given on Figures 6.2.1-1 through 6.2.1-6b for the most severe LOCA's. Pipe break locations, peak pressures and temperatures, and times of peak pressure, are summarized in Table 6.2.1-4 for each LOCA analyzed for SGR/PUR conditions. The DBA's are identified in Table 6.2.1-2. Similar passive heat sinks listed in Table 6.2.1-7 were combined into larger heat sinks. The resultant heat sink model would maintain the sum of the combined heat sinks surface area, with a volume-weighted thickness. Heat Sinks No. 4 through 11 in Table 6.2.1-8 are resultant heat sink models used in the containment functional analyses. Figures 6.2.1-11 and 6.2.1-12 are plots of the containment condensing heat transfer coefficient versus time for the most severe LOCA. The Direct heat transfer option with the DLM (Diffusion Layer Model) condensation option as described in Appendix 6.2A is applied for all LOCA analyses. For the primary system breaks (LOCA), the containment pressure reaches a peak near the end of the blowdown period. Continued heat removal by the concrete and steel heat sinks and the Containment Spray System after initiation results in a decrease of this pressure peak. For the double ended pump suction guillotine (DEPSLG) break case, further mass and energy release from the break during the core reflood period causes the pressure to rise once more until a new balance between energy release and energy removal is reached. Continued heat absorption by the steel and concrete and the Containment Spray System results in a decrease of this second pressure peak. The long-term results for the peak pressure DBA were evaluated to verify the ability of the Containment Heat Removal System (CHRS) to maintain the Containment below the design conditions. These evaluations were based upon the conservatively assumed performance of the engineered safety features as discussed above. The mechanism by which heat from the Containment is assumed to be rejected to the outside environment during the accident is the following:

1. The heat from the Containment is rejected to the Component Cooling Water System by the water/water heat transfer in the RHR heat exchanger.

2. For the maximum injection case, four containment fan coolers are assumed to be supplied with 95°F service water. For the minimum injection case, two containment fan Amendment 63 Page 13 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 coolers are assumed to be supplied with 95°F service water. This is conservative-based on a maximum operational service water inlet temperature of 94°F. For the maximum injection case, each RHR heat exchanger was conservatively assumed to be supplied with 120°F component cooling water. For the minimum injection case, one RHR heat exchanger was also conservatively assumed to be supplied with 120°F component cooling water.

3. The CCWS heat exchanger serves as the mechanism by which heat is rejected to the outside environment. As part of the accident heat removal system the CCWS heat exchanger performance is included in the determination of the RHR heat exchanger outlet temperature.

4. A maximum coolant inlet temperature and minimum coolant inlet flow are assumed so as to minimize the heat being removed during the recirculation phase.

At the start of the recirculation mode, water is drawn from the containment sump by the RHR pumps and returned to the core after passing through the RHR heat exchangers. Simultaneously the containment spray pump takes suction from the sump and sprays the water back into Containment. As a result of the higher safety injection system (charging) pump inlet temperature at the start of recirculation, steam continues to be generated in the reactor core at a high rate, due primarily to the release of decay heat and stored energy in the system internals. This causes the containment pressure to rise. The higher temperature of the recirculation containment spray further contributes to this pressure rise by reducing the ability of the sprays to remove heat from the containment atmosphere. This rise in containment pressure and temperature occurs until a heat balance is reached. For the containment peak pressure LOCA, the maximum heat load on the RHR heat exchanger occurs when the containment sump water temperature is at maximum and, hence, a maximum temperature difference exists in the RHR heat exchanger. The Component Cooling Water System is designed to accept this peak post-LOCA heat load from the RHR heat exchanger and the heat generated by station emergency auxiliaries. This rise in pressure is reversed when the combined RHR heat exchanger and structural heat removal rate becomes greater than the net heat addition to the Containment. The containment pressure and temperature responses out to 10 million seconds are calculated for the LOCA-DBA, identified in Table 6.2.1-4, with the ESF performance mode in Table 6.2.1-6. The same initial conditions are used in the analysis of the pump suction leg, and hot leg breaks. Figures 6.2.1-1 through 6.2.1-6b show the calculated transient containment temperature, containment sump temperature, and containment pressure for the most severe hot leg break. In contrast to the pump suction leg breaks, for the hot leg break there is no physical mechanism to rapidly remove the residual steam generator secondary energy either during or after reflood. Accident chronologies for the most severe reactor coolant system breaks are provided in Table 6.2.1-9. It is assumed that time equals zero at the start of each accident. Figure 6.2.1-13 provides a typical rate of energy distribution inside Containment for the LOCA containment pressure DBA (does not represent the latest analysis). The long-term performance Amendment 63 Page 14 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 is essentially the same for all the primary system break cases. All mechanisms of energy storage within the Containment are addressed. Included are the vapor energy (steam plus air), containment sump (liquid) energy, and energy contained in heat sinks. 6.2.1.1.3.3 Main steam line breaks The following breaks were postulated for the SGR/Uprate:

   - Double-ended ruptures (1.4 ft2) at 100.34%, 68.6%, 29.4%, and 0% power
   - Split rupture (0.687 ft2) at 100.34% power
   - Split rupture (0.675 ft2) at 68.6% power
   - Split rupture (0.666 ft2) at 29.4% power
   - Split rupture (0.558 ft2) at 0% power However, previous studies indicate that a full double-ended break at a given power is more severe than a corresponding split break. Consequently, only the double-ended breaks at the four power levels were analyzed for the SGR/Uprate. Small double-ended ruptures were not postulated since they result in Containment pressure and temperature responses that are less severe than those associated with full double-ended and split breaks.

Mass and Energy release data used in the analysis for each of the four postulated double-ended breaks reflects the failure of the faulted-loop main steam isolation valve (MSIV). In addition to the consequences of this initial MSLB and MSIV failure assumption, the analyses include the following additional single failures:

   - An active failure of a main feedwater isolation valve (MFIV); or
   - An active failure of a feedwater flow control valve (MFCV or MFBCV at 0% power); or
   - A single failure of one cooling train for heat removal.

The peak containment pressure and temperature are calculated to occur following the DBA MSLB indicated in Table 6.2.1-2. The containment pressure and temperature transients for the most severe MSLB cases are shown on Figures 6.2.1-9 through 6.2.1-10b. Figure 6.2.1-14 shows a typical transient containment liner surface temperature for the maximum MSLB containment temperature DBA. Pipe break areas, peak pressures and temperatures, times of peak pressure, initial power level, and single active failure assumed are summarized in Table 6.2.1-4 for each MSLB analyzed. Figure 6.2.1-12 is a plot of the condensing heat transfer coefficient versus time for the containment temperature DBA. The Uchida heat transfer coefficient contained in the CONTEMPT computer code has been used for the analysis of all secondary system breaks (MSLBs). Amendment 63 Page 15 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The containment analyses for the MSLBs have been performed using all the containment initial conditions, heat sinks and methodology assumed for the LOCA analyses except for the following:

1. For the MFIV failure case, both containment heat removal trains (four fan coolers and two spray pumps) are assumed to operate. For the cases of one heat removal system train failure, two fan coolers and one spray pump are assumed to operate.

2. The mass and energy release rates for the MSLB Containment transient are calculated with the assumption of the availability of the offsite power, as further described in Section 6.2.1.4.8. The Containment Isolation Actuation Signal (CIAS) (T), HI-1 (3.0 psig) setpoint or low steamline press (LSP), is reached within 1 second after the postulated ruptures for both the most severe temperature and the most severe pressure cases.

3. The CONTEMPT-LT/28 computer code was used in the analyses since CONTEMPT-LT/26 code has excess conservatism.

4. The mass diffusion calculation of the CONTEMPT-LT computer code for the containment vapor-sump heat and mass transfer was conservatively omitted for all MSLB analyses.

5. The most severe MSLB containment transients were evaluated using a conservative in fan cooler capacity, fan cooler start up delay time, the containment spray flow rates and fill up time, feedwater control valve and feedwater isolation valve closure times, and FW and AFW flow rates.

6. A sensitivity study indicates that the MSLB peak pressure case is more severe with a high containment initial pressure and a low initial relative humidity (same as for LOCA analyses). However, the study also indicates that for the MSLB temperature case, the peak temperature is obtained assuming a low initial containment pressure and a high initial relative humidity.

As discussed in 6.2.1.4, feedwater addition to the faulted steam generator includes unisolable feedwater piping volumes and pumped feedwater addition until isolation (FW and AFW). The amounts of feedwater added to the faulted steam generator for the 100.34%, 68.6%, and 29.4% power levels have been calculated using the RELAP5 computer code. Feedwater addition at 0% power was determined based upon a conservative hand calculation. Since the dryout times for the MFCV failure case are significantly smaller than those for the MFIV failure case, the consequences of the MFCV failure cases are enveloped by the MFIV failure cases and are not analyzed. The most limiting MSLB cases are the full double-ended break at 30% power for maximum pressure and the full double-ended break at 102% power for the maximum temperature. For all MSLBs analyzed following blowdown of the ruptured steam generator unit, the RCS decay heat is transferred to the intact units which, in turn, vent to the atmosphere when their safety relief valves open. Therefore, there is no physical mechanism for the release of significant amounts of mass or energy to the Containment after the end of blowdown. Main feedwater line breaks (MFWLB) are not analyzed since such breaks result in a blowdown less Amendment 63 Page 16 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 limiting than the MSLB because the pipe break mass flow for the MFWLB is limited by the steam generator internals design. Fluid enthalpy for the MFWLB is also less than the enthalpy of the fluid in the MSLB. A discussion of the computer codes, and the assumptions, including all assumed single active failures, used in deriving the MSLB mass and energy releases are discussed in Section 6.2.1.4. Accident chronologies for the most severe secondary system break are provided in Table 6.2.1-

9. It is assumed that time equals zero at the start of the accident.

The instrumentation provided to monitor and record the post-accident containment pressure and temperature is discussed in Section 7.5. This instrumentation is designed and qualified for the SSE and the environmental conditions discussed in Section 3.11. 6.2.1.1.3.4 Containment external (differential) pressure analysis An analysis was made of the transient for a positive external containment differential pressure which results from actuation of the Containment Spray System during normal plant operation. The analysis was performed using the Ebasco modified version of the CONTEMPT computer code described in Appendix 6.2A. The assumptions used in the analysis of an inadvertent containment spray system actuation are listed in Table 6.2.1-11. The calculated external (differential) pressure transient, is shown as a function of time on Figure 6.2.1-15. The containment external (differential) pressure design provides substantial margin over this conservatively calculated value as shown in Table 6.2.1-3. There is no single failure which could result in the operation of both containment spray trains as was assumed for the purposes of this analysis. To evaluate the adequacy of the containment design and the containment vacuum relief system, sensitivity analyses were performed with different parameters, such as varying initial temperature and humidity from the minimum value to the maximum value. The worst case of combining the most severe parameters resulted in a negative pressure differential of 1.814 psid. This case considered the simultaneous application of the worst summer and winter conditions which would not occur in a real situation. Since there are additional conservatisms in the calculation model, such as ignoring the heat-sink effect and keeping the RAB at the worst initial conditions etc., the containment design margin for the external pressure differential should be more than 0.07 psid. The transient for the vacuum relief system is the accidental initiation of the containment spray system (both pumps) while the containment is at its calculated bulk average temperature of 135°F. The containment spray pumps are assumed to reach full runout flow (4293 gpm total for both) instantaneously, the initial humidity is assumed to be 65 percent, with an initial pressure of negative 4 inches w.g. and one (1) vacuum relief subsystem is assumed not operating. This is the worst combination for negative pressure. The outside air is taken as 105°F and 100 percent humidity. The temperature of the spray water is taken as 40°F and the temperature of the service water to the fans is taken as 33°F; both are the lowest and the most conservative temperatures. For other assumptions and data see Table 6.2.1-11. An analysis was performed to verify the sizing of the vacuum relief system. Calculations were performed with the computer code CONTEMPT-LT which considers conditions in the Amendment 63 Page 17 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Containment and allows only leakage from atmosphere to Reactor Auxiliary Building and from Reactor Auxiliary Building to the Containment. Refer to Sections 6.2.1.1 and 6.2.1.2 for a discussion of the Containment subcompartment analysis. As a result of the analysis, the use of a 24-inch nominal vacuum relief valve was verified. Protection of the containment vessel against excessive external pressure is provided by two independent vacuum relief lines, each sized to prevent the differential pressure between the containment and the outside atmosphere from exceeding the design value of negative 2.0 psid. The vacuum system conforms to the requirements of Paragraph NE-7116 of ASME Section III. The containment vacuum relief system is shown on Figure 6.2.2-3. The system consists of a check valve and an automatic air operated butterfly valve outside the containment building. The check valve is provided with a short pipe spool permanently attached to the valve, and a removable test flange. Actuation of the butterfly valves are controlled by differential pressure between the outside atmosphere and the containment. Safety grade differential pressure transmitters, as described in FSAR Section 7.3.1.5.12, are provided, two for monitoring and two for control. One set of transmitters provide a signal for control action to open the butterfly valves when the differential pressure between the containment and outside reaches (-) 2.5 inches water gauge (w.g.). The second set of transmitters, which are of a different manufacturer, provide a continuous signal to the MCB for indication and will alarm Hi Containment Vacuum when the differential pressure between the containment and the outside atmosphere reaches (-) 1.0 inches w.g. The vacuum relief check valve is set to open at a differential pressure of 1.5 in. w.g. and the butterfly valve is set to open at a differential pressure of 2.5 in. w.g. The total loss coefficients for the Containment Vacuum Relief System are shown on Table 6.2.1-11 for the components illustrated in Figure 6.2.1 306. Both the vacuum relief check valves inside the Containment and the butterfly valves outside the Containment perform the dual safety functions of providing an open flow path for relieving negative containment pressure and providing containment pressure integrity for positive containment pressures. These valves are designed to satisfy Safety Class 2 and Seismic Category I requirements. Each valve is designed to take the full containment design pressure. Since the containment vacuum relief check valves also perform as containment isolation valves in the event of a LOCA, the pneumatically operated butterfly valves are designed to fail closed. A Seismic Category I air accumulator is provided for each butterfly valve to ensure a reliable energy source for operation of each valve. Each air accumulator is sized to allow three cycles of operation of its associated air operated valve. The Seismic Class I air supply is isolated from the normal Non-Seismic Class I air supply system by a set of check valves which will prevent the loss of air from the accumulator in the event of failure of the Non-Seismic Category I air supply system. Refer to Table 6.2.1-64 for a single failure analysis of the Containment Vacuum Relief System. Each vacuum relief assembly is provided with independent instrumentation and controls in accordance with IEEE-279 requirements. The electrical supply for the control operations of each valve is from a separate emergency 125V DC bus. No single failure of system component can prevent operation of the Containment Vacuum Relief System. Amendment 63 Page 18 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The Containment Vacuum Relief System pre-operational tests are described in Section 14.2.12.1.67. Periodic tests as required by the Technical Specifications in Section 16.2 will be performed. In-service inspection will be performed in accordance with Section 6.6 and valve testing requirements in Section 3.9.6 will be followed. 6.2.1.2 Containment Subcompartments 6.2.1.2.1 Design bases The containment subcompartments are subject to pressure transients and jet impingement forces caused by the mass and energy releases from postulated high energy pipe ruptures within their boundaries. Subcompartments within which high energy ruptures are postulated include the reactor cavity, the pressurizer subcompartment, and the three steam generator subcompartments. The original HNP design and license bases did not apply leak-before-break (LBB) methodology. As a result, dynamic effects of large RCS pipe breaks (DECLG, DEPSG, DEHLG, and 150 sq. in. hot/cold leg) were considered in the structural design basis for containment subcompartment analysis. Since LBB has subsequently been approved for application at HNP, the large RCS breaks are eliminated from consideration (Reference 6.2.1-15) Instead, for SGR/PUR, evaluation of postulated breaks in the pressurizer surge and spray lines, RHR lines, and accumulator nozzles were performed to demonstrate that the associated dynamic effects are bounded by the original design bases. Discussions and referenced tables and figures in Section 6.2.1.2 reflect the original design basis subcompartment analysis. Section 6.2.1.2a discusses results of SGR/PUR evaluations. Analyses were made to determine the peak pressure that could be produced by a line break discharging into the subcompartments. Venting of these chambers is employed to keep the differential pressures within structural limits. In addition, restraints on the reactor coolant pipes, reactor vessel, steam generators, and other pressurized equipment are designed so that neither pipe whip nor other forces transmitted through component supports threaten the integrity of the containment structures (see Section 3.6). Break locations for the pressurization analyses were chosen to maximize the pressures. The inherent stiffness of the systems, together with the pipe whip restraints, limits the break openings to no more than the break sizes considered. The spectrum of pipe breaks analyzed for each subcompartment are listed in Table 6.2.1-1. The location and characteristics of the reactor coolant pipe ruptures were determined mechanistically in accordance with the methods and criteria of Reference 6.2.1-3. The accident that resulted in the maximum differential pressure across the walls of a respective compartment is designated as the subcompartment design basis accident (DBA).

Section 6.2.1.2a presents discussion of a subsequent evaluation to assess the effects of plant operation with Steam Generator Replacement and Power Uprate. Amendment 63 Page 19 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Calculated DBA differential pressures are compared to the design differential pressure values used in the structural design of subcompartment walls and equipment to ensure that calculated values are less than design values. 6.2.1.2.2 Design features Plan and elevation drawings for each subcompartment showing detailed design, nodes, and component and equipment locations are shown on Figures 6.2.1-18 through 6.2.1-20. The reactor cavity is a heavily reinforced concrete structure that performs the dual function of providing reactor vessel support and radiation shielding. Figures 6.2.1-21 and 6.2.1-22 show the reactor cavity model. The walls of the steam generator compartments are constructed of reinforced concrete that serves to support the equipment enclosed and provides radiation shielding. Figures 6.2.1-23 and 6.2.1-27 present the steam generator subcompartment model. There are three steam generator compartments, each having different geometry. Since the pressurizer subcompartment is immediately adjacent to the Loop 2 SG compartment and the opening between the compartments is fairly large, the pressurizer subcompartment model has been incorporated into the SG subcompartment model as shown on Figure 6.2.1-27. Due to the small mass and energy release rates associated with the pressurizer line breaks and due to the routing of these lines and the location of the connection to the pressurizer vessel, the pressurization inside the pressurizer compartment was found to be less severe than that of the Double Ended Hot Leg Guillotine Break inside the Loop 2 SG compartment. As can be seen from Table 6.2.1-27, the peak pressure differential across the pressurizer compartment wall for the case of DEHLG Break inside the Loop 2 SG compartment is higher than the other two pressurizer line break cases. Figures 6.2.1-249 through 6.2.1-301 are the results of DEHLG in Loop 2 SG compartment, therefore, the pressures in some SG subcompartments can be higher than the peak pressures inside the pressurizer subcompartment. The bounding peak pressurizer compartment pressure was found to be resulted from the DEHLG break side of the Loop 2 SG compartment. The calculated DBA differential pressure for the Loop 2 SG compartment is, therefore, considered as the calculated DBA differential pressure for the pressurizer compartment as shown in Tables 6.2.1-2 and 6.2.1-3. 6.2.1.2.3 Design evaluation

1. Computer Codes - The analytical model used to calculate mass and energy release rates is fully described in Reference 6.2.1-3. Tables 6.2.1-12 through 6.2.1-18 provide a tabulation of mass and energy release rates versus time.

See Section 6.2.1.2a for subsequent evaluation, with Steam Generator Replacement and Power Uprate. Amendment 63 Page 20 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Analysis of the pressure transients in the reactor cavity, the steam generator subcompartment, and the pressurizer subcompartment were performed using the RELAP-4 Mod 6 computer code, Reference 6.2.1-4. The options used in running the code are:

a. The RELAP-4 CONTAINMENT option.

b. The compressible single-stream form of the momentum equation with momentum flux, except where flow oscillations are present. For this case the RELAP-4 manual recommends use of the incompressible single-stream form of the momentum equation.

c. The thermal homogeneous equilibrium critical chocked flow correlation for air-steam-water mixtures.

Two sub-compartment-analyses which yielded the maximum p between two compartments (1. Steam Generator Loop 3; double ended cold leg guillotine break and 2. Reactor Cavity; 150 in2 cold leg guillotine break) were chosen for sensitivity studies using minimum humidity. The maximum pressure differences were found to be the same. Similar sensitivity studies with an initial pressure of 14.841 psia (maximum normal operating pressure inside the containment) in the volumes were carried out; the pmax was found to be smaller by a negligible amount. Therefore, the assumed initial pressures and humidities do not affect the calculated pmax. An initial pressure of 14.7 psia, an initial temperature of 120 F and a 100 percent initial humidity have been used in all the subcompartment pressure analyses. The junction effective inertia (1/A) was calculated in a manner consistent with the methods included in RELAP-4. For a pair of volumes vi and vk, with cross-sectional areas, Ai and Ak, and lengths in the direction of flow, li and lk and a junction with area Aj and length lj, where Aj Ai, Ak and lj < li, lk, the inertia coefficient 1/A, was computed as:

                        =     +     +                                                            (1)

Flow coefficients for the subcompartment analysis were computed in a manner consistent with the calculations done by the RELAP-4 code. The junction friction coefficient utilized in the analyses is a combination of the wall friction losses (Kf), and any irreversible friction losses (KT) such as area changes, flow obstructions due to turns and gratings. The wall friction loss is computed as: KFi = (2) KFj = (3) KFk = (4) where DHk are the hydraulic diameters of the system and f is conservatively assumed to be 0.02. KT is drawn from References 6.2.1-5 and 6.2.1-6 and is chosen to account for all friction loss within the associated volumes as well as loss within the junction itself. Amendment 63 Page 21 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The total friction loss coefficient at a junction (KRELAP) is then represented as: KRELAP = KFi + KFj + KFk + KT where AT represents the reference area to which KT applies.

2. Subcompartment Modeling - Subcompartment nodalization models are principally determined by physical flow restrictions within each compartment. These flow restrictions consider the presence of steel and concrete obstructions, doorways, vent shafts, grating, reactor coolant pumps, piping, the steam generator, the pressurizer, the reactor vessel, and the reactor cavity missile and neutron shields. By choosing node boundaries at the various physical flow restrictions in a manner consistent with the flow model used by RELAP-4, calculated differential pressures and consequent support loads are realistically maximized. The nodalization sensitivity study performed in the SHNPP PSAR showed that the peak calculated differential pressure is very sensitive to an increasing number of nodes until that number equals the number of critical physical restrictions to flow. Increasing the number of nodes beyond the number of critical physical restrictions does not result in increased pressure differentials. It is therefore concluded that further arbitrary increase in the number of subcompartment nodes modeled is neither sensible nor realistic unless additional physical flow obstructions exist. The subcompartment models, discussed below, take into account all critical physical flow obstructions present.

For all analyses, insulation was assumed to remain in place and was included in the volume and vent area calculations. No displaced objects are assumed to exist.

1. Reactor Vessel Cavity - For the analyses of the pressure transient in the reactor cavity following a line break, the flow models are illustrated on Figures 6.2.1-21 and 6.2.1-22. The control volume and vent path descriptions are given in Tables 6.2.1-19 and 6.2.1-20. The mass and energy release for the affected piping system, location and size for each break is given in Tables 6.2.1-12 and 6.2.1-13. Vertical and horizontal forces on the reactor vessel and the moment which arises are provided in Reference 6.2.1-12. Detailed information concerning the computer codes used are also contained in Reference 6.2.1-12.

2. Steam Generator Compartments* - For the analysis of the pressure transient in the three steam generator compartments following a LOCA, the flow models used are shown on Figures 6.2.1-23 through 6.2.1-26. Each control volume and vent path is shown in Tables 6.2.1-21 through 6.2.1-24. The mass and energy release for the affected piping system, location, and size for each break is given in Tables 6.2.1-14 through 6.2.1-16.

3. Pressurizer Compartment* - For analysis of the pressurizer subcompartment pressure transient and uplift force, the pressurizer compartment was modeled as depicted on Figure 6.2.1-27. Control volume and vent path descriptions are given in Tables 6.2.1-25 and 6.2.1-26.

The mass and energy release for the affected piping system, location, and size for each break is given in Tables 6.2.1-17 and 6.2.1-18. The breaks considered for the analyses were a double ended pressurizer surge line guillotine break within the pressurizer skirt area, a pressurizer spray line system breaks in the pressurizer subcompartment.

See Section 6.2.1.2a for subsequent evaluation, with Steam Generator Replacement and Power Uprate. Amendment 63 Page 22 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 In the subcompartment pressurization analysis, the insulation for Reactor Vessel and piping system was assumed to remain in place and the volume occupied by the insulation was deducted from the free volume of each subcompartment. The reflecting mirror type insulation used inside the containment cannot sustain any pressure buildup. During the subcompartment pressure transient, the insulation near the pipe break will be most likely crushed to near zero occupied volume or torn loose and carried to adjacent volumes. In either case, the net free volume of the subcompartment where the pipe break is located will be substantially increased and the junction area to the adjacent volume will also be enlarged accordingly. Both will cause the reduction in pressure inside the break subcompartment. Since the peak pressure inside the break subcompartment has been used as the representing pressure for the comparison with the design pressure, the assumptions used in the current subcompartment analysis are believed to be conservative and justified. The neutron streaming shield is located below the postulated nozzle rupture elevation for a reactor cavity pressurization transient. For this reason it is considered to remain in place during the transient. The effects of the shielding blockage and occupied volumes were considered in the reactor cavity subcompartment modeling. A Permanent Cavity Seal Ring (PCSR) has been installed within the cavity annulus at the refueling seal ledge. This seal ring has eight open hatches for venting during normal operation. With credit for "Leak Before Break" approach, (see Section 6.2.1-2a), no large RCS loop piping break is postulated within the primary shield wall. The primary shield wall piping penetrations provide a vent path for a break of smaller attached lines to the reactor coolant loop in the steam generator/pressurizer cubicle subcompartments. This results in some flow through the PCSR. These effects have been evaluated and have been determined to be acceptable. Due to the small mass and energy release associated with the pressurizer spray line guillotine, (see Table 6.2.1-18), and due to the routing of these lines and location of their connections to the pressurizer vessel, this break is not capable of producing significant lateral pressure differentials across the pressurizer. Results - The design value and peak calculated values for pressure in the subcompartments is shown in Tables 6.2.1-2, 6.2.1-3, and 6.2.1-27. Graphs of the subcompartment pressure response versus time for the limiting break for the reactor cavity, steam generator loop 1, steam generator loop 3, steam generator loop 2, and pressurizer subcompartments are given in Figures 6.2.1-28 through 6.2.1-75, Figures 6.2.1-76 through 6.2.1-132, Figures 6.2.1-133 through 6.2.1-195, Figures 6.2.1-196 through 6.2.1-248, and Figures 6.2.1-249 through 6.2.1-301, respectively. Peak pressure differentials for all cases analyzed are given in Table 6.2.1-27. The peak calculated differential pressure is limited to a small portion of the total wall area and is less than the design pressures. The external asymmetric loadings to the reactor pressure vessel are the result of the pressure differentials inside the reactor cavity throughout the cavity pressurization transient. The worst loadings would be caused by the 150 in.2 break at the Second-loop Inlet (Cold Leg) Nozzle. The component pressure forces acting on the reactor vessel are calculated by multiplying the pressure in each volume with the appropriate projected areas. Since the vessel insulation is considered to remain in place throughout the transient, the projected areas are conservatively

 *See Section 6.2.1.2a for subsequent evaluation with Steam Generator Replacement and Power Uprate.

Amendment 63 Page 23 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 calculated by using the insulation surface area rather than the vessel surface area. The values of the projected areas used in the component force calculation are listed in Table 6.2.1-20A. The forces are assumed to act through the midpoint of the vessel insulation located in each volume. The coordinate system used in the force and moment calculation is identified in Figure 6.2.1-21. The component forces in each volume were then summed to form the resultant forces in X, Y, Z directions. These resultant forces are shown in Figure 6.2.1-307. They should be applied to the vessel through the origin of the coordinate system. They also include the pressure differential forces across the nozzles and the primary pipings inside the cavity. The resultant moments about X and Y axes (shown in Figure 6.2.1-308) were calculated by summing up the product of the component forces and the appropriate lever arms. The lever arms for each volume is determined by the vertical distance between the nozzle center line elevation (Elevation 253.75 ft.) and the elevation of the insulation midpoint in each volume. The values of the lever arms used in the moment calculation are presented in Table 6.2.1-20B. The magnitude of Mz is relatively small and it has been neglected in the evaluation. Westinghouse assumed a 150 in.2 rupture area when analyzing for asymmetric loads in the reactor cavity. Using this area as the maximum allowable break area, Ebasco designed the reactor vessel hot and cold leg restraints. Using geometric parameters from the restraints, Westinghouse then calculated the actual rupture areas of approximately 32 in.2 for the hot leg and 90 in.2 for the cold leg. Since these areas are enveloped by the assumed 150 in.2 break area the reactor cavity subcompartment analysis is conservative. 6.2.1.2a Evaluation of SGR/PUR The short-term LOCA-related Mass and Energy releases are used as input to the subcompartment analyses, which are performed to ensure that the walls of a subcompartment can maintain their structural integrity during the short pressure pulse (generally less than 3 seconds) accompanying a high-energy line pipe rupture within that subcompartment. The subcompartments evaluated include the steam generator (SG) compartment, the reactor cavity region, and the pressurizer compartment. For the SG compartment and the reactor cavity region, the fact that the HNP is approved for leak-before-break (LBB) was used to qualitatively demonstrate that any changes associated with the SGR/Uprating program are offset by the LBB benefit of using the smaller RCS nozzle breaks. This demonstrates that the current licensing bases for these subcompartments remain bounding. For the pressurizer compartment, the Reference 6.2.1-3 methodology was applied to calculate pressurizer spray line and surge line Mass and Energy (M&E) releases. The results of this reanalysis are discussed below. A reanalysis was conducted to determine the effect of the SGR/Uprating on the short-term LOCA-related M&E releases that support the pressurizer subcompartment analyses for HNP FSAR, 6.2.1.2a. Since HNP was licensed for LBB by Reference 6.2.1-15, only breaks in the largest branch lines are analyzed (the pressurizer surge line and spray line break). The RCL breaks have been eliminated by LBB and therefore, the original design bases (pre LBB) M&E releases associated with these breaks would bound any RCS primary break considered under the LBB exemption. This evaluation addresses the impact of the SGR/Uprating and other relevant issues on the current licensing basis for HNP. The magnitude of the pressure differential across the walls is a function of several parameters, which include the blowdown M&E release rates, the subcompartment volume, vent areas, and vent flow behavior. The blowdown M&E release rates are affected by the initial RCS Amendment 63 Page 24 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 temperature conditions. Since short-term releases are linked directly to the critical mass flux, which increases with decreasing temperatures, the short-term LOCA releases would be expected to increase due to any reductions in RCS coolant temperature conditions. Short-term blowdown transients are characterized by a peak M&E release rate that occurs during a subcooled condition. Therefore, using lower temperatures, which maximizes the short-term LOCA M&E releases, data representative of the lowest inlet and outlet temperatures (with uncertainty subtracted) were used for the HNP SGR/Uprating analysis. The evaluation considered a temperature operating range of 572°F to 588.2°F for the RCS average temperature. For this evaluation, an RCS pressure of 2301 psia (2250 + 51 psi uncertainty), a vessel outlet temperature of 598.2°F, and a vessel/core inlet temperature of 530.6°F were considered for the uprating, which includes consideration of the lower end of the operating range with the temperature uncertainty of 6°F. Additionally, due to the short time period (0-3 seconds) that these events are analyzed for, the ECCS system is not modeled. Since the ECCS will not start in this short time period, single failures in the ECCS and Engineering Safeguards are not of a concern and are not considered. The M&E data for the pressurizer surge line and spray line break analyses are given in Tables 6.2.1-17a and 6.2.1-18a. The methodology described in Reference 6.2.1-3 was used. Per Reference 6.2.1-15, HNP is approved for LBB. LBB eliminates the dynamic effects of postulated primary loop pipe ruptures from the design basis. This means that the current breaks (a double-ended circumferential rupture of the reactor coolant cold leg, hot leg, and the steam generator inlet nozzle, used for the SG compartments, and a 150 in2 RV inlet break for the reactor cavity region) no longer have to be considered for the short-term effects. Since the RCL piping has been eliminated from consideration, the large branch nozzles must then be considered. This includes the surge line, accumulator line, and the RHR line. These smaller breaks, which are outside the cavity region, would result in minimal asymmetric pressurization in the reactor cavity region. Additionally, compared to the large RCL double-ended ruptures, the differential loadings are significantly reduced. For example, peak compartment pressure can be reduced by a factor of greater than 2, and the peak differential across an adjacent wall can be reduced by a factor of greater than 3, if only the nozzle breaks are considered. Therefore, since the HNP is approved for LBB, the decrease in M&E releases associated with the smaller RCL nozzle breaks, as compared to the larger RCL pipe breaks, more than offsets any increased releases associated with the lower RCS temperatures as a result of the SGR/Uprating. The current licensing basis subcompartment analyses that consider breaks in the RCL remain bounding, as discussed below: Reactor Cavity The design basis for the Reactor cavity is a 150 in2 break in the RCS piping. The break sizes associated with the postulated Surge line, the RHR lines, Pressurizer Spray line, accumulator nozzle are outside the cavity and are significantly smaller than 150 in2. Therefore, the lower Mass and Energy releases from these smaller RCS breaks would offset any changes associated with SGR/Power Uprate. Consequently, the existing design basis subcompartment differential pressures and associated forces and moments envelope those due to smaller breaks outside the cavity. Amendment 63 Page 25 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Steam Generator Compartment The original design basis analysis considered a double-ended rupture in hot-leg, cold-leg, and pump suction of the RCS piping. The Mass and Energy release rates for these breaks are significantly larger than those due to postulated breaks in RHR, Pressurizer Spray and Pressurizer Surge lines. Although modifications have been made in the geometry (such as installing different type of insulation on the SG and rerouting feedwater pipe inside the cubicle), the subcompartment pressurization and associated forces and moments obtained in the original design basis analysis remain bounding. Pressurizer Compartment The original design basis for the Pressurizer compartment is the double-ended pump suction break which enveloped the postulated breaks in the Surge and the Spray lines. FSAR Tables 6.2.1-17a and 6.2.1-18a, provide Mass and Energy releases for the Surge line and Spray line breaks for Power Uprate/SGR conditions. A review of the new Mass and Energy release rates reveals the new rates are actually lower than those used in the original design basis calculation up to 0.1 second into transient. Since the peak pressure in the Pressurizer compartment occurs at 0.0175 second after the break, the new differential peak pressure would be lower than 7 psid calculated previously (see Table 6.2.1-27) The blowdown data for the Spray line has increased due to Power uprate by about 15%. This increase is expected to increase the peak differential pressure from 0.9 psid to about 1.2 psid. However, this peak differential pressure is much lower than those due to Pump Suction and Surge line breaks. 6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents The analysis for the SGR/Uprating program, used Westinghouse generated mass and energy (M&E) releases using the March 1979 model, described in Reference 6.2.1-10, which includes the NRC review and approval letter. This methodology has previously been applied to the HNP (Reference 6.2.1-18). Input Parameters and Assumptions The mass and energy release analysis is sensitive to the assumed characteristics of various plant systems, in addition to other key modeling assumptions. Where appropriate, bounding inputs are utilized and instrumentation uncertainties are included. For example, the Reactor Coolant System (RCS) operating temperatures are chosen to bound the highest average coolant temperature range of all operating cases, and a temperature uncertainty allowance of (+6.0°F) is then added. Nominal parameters are used in certain instances. For example, the RCS pressure in this analysis is based on a nominal value of 2250 psia plus an uncertainty allowance (+51 psi). All input parameters are chosen consistent with accepted analysis methodology. Some of the most-critical items are the RCS initial conditions, core decay heat, safety injection flow, and primary and secondary metal mass and steam generator heat release modeling. Specific assumptions concerning each of these items are discussed next. The core rated power of 2958 MWt which includes calorimetric error was used in the analysis. As previously noted, the use of RCS operating temperatures to bound the highest average coolant temperature range were used as bounding analysis conditions. The use of higher Amendment 63 Page 26 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 temperatures is conservative because the initial fluid energy is based on coolant temperatures that are at the maximum levels attained in steady state operation. Additionally, an allowance to account for instrument error and deadband is reflected in the initial RCS temperatures. The selection of 2250 psi, plus an uncertainty allowance, as the limiting pressure is considered to affect the blowdown phase results only, since this represents the initial pressure of the RCS. The RCS rapidly depressurizes from this value until the point at which it equilibrates with containment pressure. The rate at which the RCS blows down is initially more severe at the higher RCS pressure. Additionally the RCS has a higher fluid density at the higher pressure (assuming a constant temperature) and subsequently has a higher RCS mass available for releases. Thus, 2250 psia plus uncertainty was selected for the initial pressure as the limiting case for the long-term M&E release calculations. The selection of the fuel design features for the long-term M&E release calculation is based on the need to conservatively maximize the energy stored in the fuel at the beginning of the postulated accident (i.e., to maximize the core stored energy). The margin in core-stored energy was chosen to be +15 percent. Thus, the analysis very conservatively accounts for the stored energy in the core. Margin in RCS volume of 3 percent (1.6 percent allowance for thermal expansion and 1.4 percent for uncertainty) is modeled. The LOCA transient is typically divided into four phases: (a) Blowdown - which includes the period from accident occurrences (when the reactor is at steady state operation) to the time when the total break flow stops. (b) Refill - the period of time when the lower plenum is being filled by accumulator and safety injection water. (This phase is conservatively neglected in computing mass and energy releases for containment evaluations). (c) Reflood - begins when the water from the lower plenum enters the core and when the core is completely quenched. (d) Post-Reflood - begins immediately after the core is quenched and continues until primary and secondary energy has been removed to 212°F. A uniform steam generator tube plugging (SGTP) level of 0 percent is modeled. This assumption maximizes the reactor coolant volume and fluid release by considering the RCS fluid in all SG tubes. During the post-blowdown period the steam generators are active heat sources, as significant energy remains in the secondary metal and secondary mass that has the potential to be transferred to the primary side. The 0-percent SGTP assumption maximizes heat transfer area and therefore, the transfer of secondary head across the SG tubes. Additionally, this assumption reduces the reactor coolant loop resistance, which reduces the pressure drop upstream of the break for the pump suction breaks and increases break flow. Thus, the analysis very conservatively accounts for the level of SGTP. Amendment 63 Page 27 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The following assumptions were employed to ensure that the M&E releases are conservatively calculated for the limiting hot leg break cased and DEPS maximum SI case thereby maximizing energy release to containment:

1. Maximum expected operating temperature of the RCS (100-percent full-power conditions)

2. Allowance for RCS temperature uncertainty (+6.0°F)

3. Margin in RCS volume of 3 percent (which is composed of 1.6-percent allowance for thermal expansion, and 1.4 percent for uncertainty)

4. Core rated power of 2958 MWt including calorimetric error.

5. Deleted by Amendment No. 58.

6. Conservative heat transfer coefficients (i.e., steam generator primary/secondary heat transfer and reactor coolant system metal heat transfer)

7. Allowance in core-stored energy for effect of fuel densification

8. A margin in core-stored energy

9. An allowance for RCS initial pressure uncertainty (+51 psi)

10. A maximum containment backpressure equal to design pressure (45 psig)

11. Allowance for RCS flow uncertainty (-2.1 percent)

12. SGTP leveling (0-percent uniform)

 -     Maximizes reactor coolant volume and fluid release
 -     Maximizes heat transfer area across the SG tubes
 -     Reduces coolant loop resistance, which reduces the P upstream of the break for the pump suction breaks and increases break flow Later analyses considering measurement uncertainty recapture demonstrated that the hot leg break described above remains the limiting break relative to peak LOCA containment pressure.

The pump suction break with minimum safeguards peak pressure, while still less than the hot leg break pressure, was not bounded. This analysis contained the following assumptions to ensure that M&E releases were conservatively calculated:

1. Maximum expected operating temperature of the RCS (100 percent full-power conditions)

2. Allowance for RCS temperature uncertainty (+3.8°F)

3. Margin in RCS volume of 3 percent (which is composed of 1.6 percent allowance for thermal expansion and 1.4 percent for uncertainty)

Amendment 63 Page 28 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

4. Bounding upper core power of 2958 MWt

5. Conservative heat transfer coefficients (i.e., steam generator primary/secondary heat transfer and reactor coolant system metal heat transfer)

6. Allowance in core-stored energy for effect of fuel densification

7. A margin in core-stored energy

8. An allowance for RCS initial pressure uncertainty (+51 psi)

9. A maximum containment backpressure equal to design pressure (45 psig) during blowdown, and 42 psig during post-blowdown

10. Allowance for RCS flow uncertainty (-2.1 percent)

11. SGTP leveling (0 percent uniform)

     - Maximizes reactor coolant volume and fluid release
     - Maximizes heat transfer area across the SG tubes
     - Reduces coolant loop resistance, which reduces the P upstream of the break for the pump suction breaks and increase break flow

12. The steam generator secondary metal mass was modeled to include only the portion of the steam generators which is in contact with the fluid on the secondary side. Portions of the steam generators such as the elliptical head, upper shell and miscellaneous internals have poor heat transfer due to location. The heat stored in these areas available for release to containment will not be able to effectively transfer energy to the RCS, thus the energy will be removed at a much slower rate and time period (>10000 seconds).

Thus based on the previously discussed conditions and assumptions, a bounding analysis for the HNP was made for the release of M&E from the RCS in the event of a LOCA at 2958 MWt. LOCA mass and Energy Release Phases The containment system receives mass and energy releases following a postulated rupture in the RCS. These releases continue over the time period, which, for the LOCA M&E analysis, is typically divided into four phases.

1. Blowdown - the period of time from accident initiation (when the reactor is at steady state operation) to the time that the RCS and containment reach an equilibrium state.

2. Refill - the period of time when the lower plenum is being filled by accumulator and Emergency Core Cooling System (ECCS) water. At the end of blowdown, a large amount of water remains in the cold legs, downcomer, and lower plenum. To conservatively consider the refill period for the purpose of containment M&E releases, it is assumed that this water is instantaneously transferred to the lower plenum along with sufficient accumulator water to completely fill the lower plenum. This allows an uninterrupted release Amendment 63 Page 29 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 of M&E to containment. Thus, the refill period is conservatively neglected in the M&E release calculation.

3. Reflood - begins when the water from the lower plenum enters the core and ends when the core is completely quenched.

4. Post-reflood (Froth) - describes the period following the reflood phase. For the pump suction break, a two-phase mixture exits the core, passes through the hot legs, and is superheated in the steam generators prior to exiting the break as steam. After the broken loop steam generator cools, the break flow becomes two-phase.

Computer Codes The Reference 6.2.1-10 mass and energy release evaluation model is comprised of M&E release versions of the following codes: SATAN VI, WREFLOOD, FROTH, and EPITOME. These codes were used to calculate the long-term LOCA M&E releases for HNP SGR/Uprating program. SATAN VI calculates blowdown, the first portion of the thermal-hydraulic transient following break initiation, including pressure, enthalpy, density, M&E flowrates, and energy transfer between primary and secondary systems as a function of time. The WREFLOOD code addresses the portion of the LOCA transient where the core reflooding phase occurs after the primary coolant system has depressurized (blowdown) due to loss of water through the break and when water supplied by the ECCS refills the reactor vessel and provides cooling to the core. The most important feature of WREFLOOD is the steam/water mixing model. FROTH models the post-reflood portion of the transient. The FROTH code is used for the steam generator heat addition calculation from the broken and intact loop steam generators. EPITOME continues the FROTH post-reflood portion of the transient from the time at which the secondary equilibrates to containment design pressure to the end of the transient. It also compiles a summary of data on the entire transient, including formal instantaneous M&E release tables and M&E balance tables with data at critical times. Break Size and Location Generic studies (Reference 6.2.1-10, Section 3) have been performed with respect to the effect of postulated break size on the LOCA M&E releases. The double-ended guillotine break has been found to be limiting due to larger mass flow rates during the blowdown phase of the transient. During the reflood and froth phases, the break size has little effect on the releases. Three distinct locations in the reactor coolant system loop can be postulated for pipe rupture for any release purposes:

1. Hot leg (between vessel and steam generator)

2. Cold leg (between pump and vessel)

Amendment 63 Page 30 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

3. Pump suction (between steam generator and pump)

The break locations analyzed for this program are the DEPS rupture (10.48 ft2) and the DEHL rupture (9.18 ft2). Break M&E releases have been calculated for the blowdown, reflood, and post-reflood phases of the LOCA for the DEPS cases. For the DEHL case, the releases were calculated only for the blowdown. The following information provides a discussion of each break location. - The DEHL rupture has been shown in previous studies (Reference 6.2.1-10, Section 3.1) to result in the highest blowdown M&E release rates. Although the core flooding rate would be the highest for this break location, the amount of energy released from the steam generator secondary is minimal because the majority of fluid that exits the core vents directly to containment, bypassing the steam generators. As a result, the reflood M&E releases are reduced significantly as compared to either the pump suction, or cold-leg break locations where the core exit mixture must pass through the steam generators before venting through the break. For the hot-leg break, generic studies have confirmed that there is no reflood peak (i.e., from the end of the blowdown period the containment pressure would continually decrease). Therefore, only the M&E releases for the hot-leg break blowdown phase are calculated. - The cold-leg break location has also been found in previous studies (Reference 6.2.1-10, Section 3.1) to be much less limiting in terms of the overall containment energy releases. The cold-leg blowdown is faster than that of the pump suction break, and more mass is released into the containment. However, the core heat transfer is greatly reduced, and this results in a considerably lower energy release into containment. Studies have determined that the blowdown transient for the cold leg is, in general, less limiting than that for the pump suction break. During reflood, the flooding rate is greatly reduced, and the energy release rate into the containment is reduced. Therefore, the cold-leg break is bounded by other breaks and no further evaluation is necessary. - The pump suction break combines the effects of the relatively high core-flooding rate, as in the hot-leg break, and the additional stored energy in the steam generators. As a result, the pump suction break yields the highest energy flow rates during the post-blowdown period by including all of the available energy of the RCS in calculating the releases to containment. Application of Single-Failure Criterion An inherent assumption in the generation of the mass and energy release is that offsite power is lost. This results in the actuation of the emergency diesel generators, required to power the safety injection system. This is not an issue for the blowdown period, which is limited by the DEHL break, since the combination of signal delay, plus diesel delay and additional delays in starting the ECCS pumps result in an SI delivery time after the end of blowdown. Generally, two cases are analyzed to assess the effects of a single failure. The first case assumes minimum safeguards SI flow based on the postulated single failure of an emergency diesel generator. This results in the loss of one train of safeguards equipment. The other case assumes maximum safeguards SI flow based on no postulated failures that would impact the amount of ECCS flow. Amendment 63 Page 31 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Acceptance Criteria A large LOCA is classified as an ANS Condition IV event, an infrequent fault. To satisfy the NRC acceptance criteria presented in the Standard Review Plan Section 6.2.1.3, the relevant requirements are as follows: - 10 CFR 50, Appendix A - 10 CFR 50, Appendix K, paragraph I.A In order to meet these requirements, the following must be addressed: - Sources of energy - Break size and location - Calculation of each phase of the accident 6.2.1.3.1 Mass and Energy Release Data The SATAN-VI code is used for computing the blowdown transient. The code utilizes the control volume (element) approach with the capability for modeling a large variety of thermal fluid system configurations. The fluid properties are considered uniform, and thermodynamic equilibrium is assumed in each element. A point kinetics model is used with weighted feedback effects. The major feedback effects include moderator density, moderator temperature, and Doppler broadening. A critical flow calculation for subcooled (modified Zaloudek), two-phase (Moody), or superheated break flow is incorporated into the analysis. The methodology for the use of this model is described in Reference 6.2.1-10. Table 6.2.1-33 presents the calculated mass and energy release for the blowdown phase of the DEHL break. For the hot-leg break M&E release tables, break path 1 refers to the M&E exiting from the reactor vessel side of the break; and break path 2 refers to the M&E exiting from the steam generator side of the break. Table 6.2.1-29a presents the calculated M&E releases for the blowdown phase of the DEPS break with maximum ECCS flows. Table 6.2.1-29b presents the calculated M&E releases for the blowdown phase of the DEPS break with minimum ECCS flows. For the pump suction breaks, break path 1 in the M&E release tables refers to the M&E exiting from the steam generator side of the break; break path 2 refers to the M&E exiting from the pump side of the break. The WREFLOOD code is used for computing the reflood transient. The WREFLOOD code consists of two basic hydraulic models--one for the contents of the reactor vessel and one for the coolant loops. The two models are coupled through the interchange of the boundary conditions applied at the vessel outlet nozzles and at the top of the downcomer. Additional transient phenomena, such as pumped safety injection and accumulators, reactor coolant pump performance, and steam generator releases are included as auxiliary equations that interact with the basic models are required. The WREFLOOD code permits the capability to calculate variations during the core reflooding transient of basic parameters such as core flooding rate, core and downcomer water levels, fluid thermodynamic conditions (pressure, enthalpy, density) throughout the primary system, and mass flowrates through the primary system. The code Amendment 63 Page 32 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 permits hydraulic modeling of the two flow paths available for discharging steam and entrained water from the core to the break, the path through the broken loop and the path through the unbroken loops. A complete thermal equilibrium mixing condition for the steam and ECCS injection water during the reflood phase has been assumed for each loop receiving ECCS water. This is consistent with the usage and application of the Reference 6.2.1-10 M&E release evaluation model in recent analyses, for example, D. C. Cook Docket (Reference 6.2.1-19). Even though the Reference 6.2.1-10 model credits steam/water mixing only in the intact loop and not in the broken loop, the justification, applicability, and NRC approval for using the mixing model in the broken loop has been documented (Reference 6.2.1-19). Moreover, this assumption is supported by test data and is further discussed below. The model assumes a complete mixing condition (i.e., thermal equilibrium) for the steam/water interaction. The complete mixing process, however, is made up of two distinct physical processes. The first is a two-phase interaction with condensation of steam by cold ECCS water. The second is a single-phase mixing of condensate and ECCS water. Since the steam release is the most important influence to the containment pressure transient, the steam condensation part of the mixing process is the only part that needs to be considered. (Any spillage directly heats only the sump.) The most applicable steam/water mixing test data has been reviewed for validation of the containment integrity reflood steam/water mixing model. This data, generated in 1/3-scale tests (Reference 6.1.1-20), are the largest scale data available and thus, most clearly simulate the flow regimes and gravitational effects that would occur in a Pressurized Water Reactor (PWR). These tests were designed specifically to study the steam/water interaction for PWR reflood conditions. A group of 1/3-scale tests corresponds directly to containment integrity reflood conditions. The injection flowrates for this group cover all phases and mixing conditions calculated during the reflood transient. The data from these tests were reviewed and discussed in detail in Reference 6.2.1-10. For all of these tests, the data clearly indicate the occurrence of very effective mixing with rapid steam condensation. The mixing model used in the containment integrity reflood calculation is therefore wholly supported by the 1/3-scale steam/water mixing data. Additionally, the following justification is also noted. The double ended pump suction break results in the highest containment pressure post-blowdown. For this break, there are two flowpaths available in the RCS by which mass and energy may be released to containment. One is through the outlet of the steam generator, the other via reverse flow through the reactor coolant pump. Steam that is not condensed by ECCS injection in the intact RCS loops passes around the downcomer and through the broken loop cold leg and pump in venting to containment. This steam also encounters ECCS injection water as it passes through the broken loop cold leg, complete mixing occurs and a portion of it is condensed. It is this portion of steam that is condensed that is taken credit for in this analysis. This assumption is justified based upon the postulated break location, and the actual physical presence of the ECCS injection nozzle. Descriptions of the test and test results are contained in References 6.2.1-10 and 6.2.1-19. Table 6.2.1-36 presents the calculated M&E release for the reflood phase of the pump suction double-ended rupture with minimum safeguards. Amendment 63 Page 33 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.2.1-35 presents the calculated M&E release for the reflood phase of the pump suction double-ended rupture with maximum safeguards. The transient responses of the principal parameters during reflood are given in Table 6.2.1-49 for the DEPS maximum safeguards case. The FROTH code (Reference 6.2.1-3) is used for computing the post-reflood transient. The FROTH code calculates the heat release rates resulting from a two-phase mixture present in the steam generator tubes. The M&E releases that occur during this phase are typically superheated due to the depressurization and equilibration of the broken-loop and intact-loop steam generators. During this phase of the transient, the RCS has equilibrated with the containment pressure, but the steam generators contain a secondary inventory at an enthalpy that is much higher than the primary side. Therefore, there is a significant amount of reverse heat transfer that occurs. Steam is produced in the core due to core decay heat. For a pump suction break, a two-phase fluid exits the core, flows through the hot legs, and becomes superheated as it passes through the steam generator. Once the broken loop cools, the break flow becomes two-phase. During the FROTH calculation, ECCS injection is addressed for both the injection phase and the recirculation phase. The FROTH code calculation stops when the secondary side equilibrates to the saturation temperature (Tsat) at the containment design pressure. After this point, the EPITOME code completes the SG depressurization. The methodology for the use of this model is described in Reference 6.2.1-10. The M&E release rates are calculated by FROTH and EPITOME until the time of containment depressurization. After containment depressurization (14.7 psia), the M&E release available to containment is generated directly from core boil-off/decay heat. Table 6.2.1-41 presents the two-phase post-reflood M&E release data for the pump suction double-ended case minimum safeguards case. Table 6.2.1-40 presents the two-phase post-reflood M&E release data for the pump suction double-ended case maximum safeguards case. The maximum safeguards mass & energy release data was subsequently reevaluated using a higher cold leg re-circulation flowrate than that assumed in the above analysis as documented in Reference 6.2.1-22. The evaluation concluded that the impact on Containment pressure and temperature, due to a higher cold leg re-circulation flowrate, remains non-limiting with respect to the pressure & temperature of the minimum safeguards case. Decay Heat Model On November 2, 1978, the Nuclear Power Plant Standards Committee (NUPPSCO) of the American Nuclear Society approved ANS Standard 5.1 (Reference 6.2.1-21) for the determination of decay heat. This standard was used in the M&E release. Table 6.2.1-66 lists the decay heat curve used in the M&E release analysis, post blowdown, for the HNP SGR/Uprating program. Significant assumptions in the generation of the decay heat curve for use in the LOCA M&E releases analysis include the following:

1. Decay heat sources considered are fission product decay and heavy element decay of U-239 and Np-239.

Amendment 63 Page 34 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

2. Decay heat power from the following fissioning isotopes are included: U-238, U-235 and Pu-239.

3. Fission rate is constant over the operating history of maximum power level.

4. The factor accounting for neutron capture in fission products has been taken from Equation 11 of Reference 6.2.1-21, up to 10,000 seconds and from Table 10 of Reference 6.2.1-21, beyond 10,000 seconds.

5. The fuel has been assumed to be at full power for 1096 days.

6. The number of atoms of U-239 produced per second has been assumed to be equal to 70 percent of the fission rate.

7. The total recoverable energy associated with one fission has been assumed to be 200 MeV/fission.

8. Two-sigma uncertainty (two times the standard deviation) has been applied to the fission product decay.

Based upon NRC staff review, Safety Evaluation Report (SER) of the March 1979 evaluation model (Reference 6.2.1-10), use of the ANS Standard-5.1, November 1979 decay heat model was approved for the calculation of M&E releases to the containment following a LOCA. A plant specific decay heat curve was developed for Shearon Harris in support of the rework necessary for the Shearon Harris measurement uncertainty recapture (MUR). The decay heat fraction as a function of time was calculated using ANS 1979 decay heat curve with plant specific parameters. Bounding values used to generate the decay heat fractions include the following:

1. A core average burnup of 50,000 MDW/MTU

2. A minimum average core enrichment of 3.0%

3. A maximum core fuel loading of 74 MTU

4. Standard 17 x 17 Westinghouse fuel, which has a nearly equal pellet diameter to the AREVA fuel used at Shearon Harris

5. A two sigma uncertainty has been applied Steam Generator Equilibration and Depressurization Steam generator equilibration and depressurization is the process by which secondary side energy is removed from the steam generators in stages. The FROTH computer code calculates the heat removal from the secondary mass until the secondary temperature is the saturation temperature (Tsat) at the containment design pressure. After the FROTH calculations, the EPITOME code continues the FROTH calculation for SG cooldown removing steam generator secondary energy at different rates (i.e., first and second stage rates). The first stage rate is applied until the steam generator reaches Tsat at the user specified intermediate equilibration pressure, when the secondary pressure is assumed to reach the actual containment pressure.

Then the second stage rate is used until the final depressurization, when the secondary reaches Amendment 63 Page 35 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 the reference temperature of Tsat at 14.7 psia, or 212°F. The heat removal of the broken-loop and intact-loop steam generators are calculated separately. During the FROTH calculations, steam generator heat removal rates are calculated using the secondary side temperature, primary side temperature, and a secondary side heat transfer coefficient determined using a modified McAdam's correlation. Steam generator energy is removed during the FROTH transient until the secondary side temperature reaches saturation temperature at the containment design pressure. The constant heat removal rate used during the first heat removal stage is based on the final heat removal rate calculated by FROTH. The SG energy available to be released during the first stage interval is determined by calculating the difference in secondary energy available at the containment design pressure and that at the (lower) user specified intermediate equilibration pressure, assuming saturated conditions. The intermediate equilibrium pressures are selected as discussed in Reference 6.2-10, Section 2.2. This energy is then divided by the first stage energy removal rate, resulting in an intermediate equilibrium time. At this time, the rate of energy release drops substantially to the second stage rate. The second stage rate is determined as the fraction of difference in secondary energy available between the intermediate equilibration and final depressurization at 212°F, and the time difference from the time of the intermediate equilibration to the user-specified time of the final depressurization at 212°F. With current methodology (Reference 6.2.1-10), all of the secondary energy remaining after the intermediate equilibration is conservatively assumed to be released by imposing a mandatory cooldown and subsequent depressurization down to atmospheric pressure at 3600 seconds, i.e., 14.7 psia and 212°F. Sources of Mass and Energy The sources of mass consideration in the LOCA M&E release analysis are given in Tables 6.2.1-47, 6.2.1-44 and 6.2.1-43. These sources are the reactor coolant system, accumulators, and pumped safety injection. The energy inventories considered in the LOCA M&E release analysis are given in Tables 6.2.1-55, 6.2.1-52, and 6.2.1-51. The energy sources are listed below. - RCS water - Accumulator water (all three inject) - Pumped SI water - Decay heat - Core stored energy - RCS metal (includes SG tubes) - SG metal (includes transition cone, shell, wrapper, and other internals) - SG secondary energy (includes fluid mass and steam mass) - Secondary transfer of energy (feedwater into, and steam out of, the SG secondary) The energy reference points are as follows. - Available energy: 212°F; 14.7 psia - Total energy content: 32°F; 14.7 psia Amendment 63 Page 36 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The mass and energy inventories are presented at the following times, as appropriate: - Time zero (initial conditions) - End of blowdown time - End of refill time - End of reflood time - Time of broken loop steam generator equilibration to pressure setpoint - Time of full depressurization (3600 seconds) In the M&E release data presented, no Zirc-water reaction heat was considered because the clad temperature is assumed not to rise high enough for the rate of the Zirc-water reaction heat to be of any significance. 6.2.1.4 Mass and Energy Release Analysis For Postulated Secondary System and Pipe Ruptures 6.2.1.4.1 Mass and energy data A complete analysis of main steam line breaks inside Containment has been performed using the methods described in WCAP-8822, including Supplement 1 and Supplement 2 (Reference 6.2.1-17). A total of 12 cases covering four power levels, two break types, and two single failures have been analyzed. However, as discussed in 6.2.1.1.33 previous studies have indicated that a full double-ended break at a given power is more severe than a corresponding split break. Consequently, only the double-ended breaks at the four power levels were analyzed for the SGR/Uprate. (A confirmatory split break case at 29.4% power and a cooling train failure was evaluated to ensure it was bounded by full DER cases at 29.4% power levels.) Mass and energy release data used in the analysis of SGR/Uprate conditions for each of the four postulated double-ended breaks reflects the failure of the faulted-loop main steam isolation valve (MSIV). In addition to the MSLB and MSIV failure assumption, the SGR/Uprate analyses include the following additional single failures: - An active failure of a main feedwater isolation valve (MFIV); or - An active failure of feedwater flow control valve (MFCV)(or MFBCV at 0% power) or; - A single failure of one cooling train for heat removal. Tables 6.2.1-58A and 6.2.1-58B present the blowdown data for the mass and energy release rates for the most limiting MSLB cases are the full double-ended break at 29.4% power for maximum pressure and the full double-ended break at 100.34% power for the maximum temperature respectively. The actual integrated mass & energy releases for these two cases are different, primarily due to the differences in the initial steam/water mass in the faulted steam generator and the pumped feedwater addition until isolation. All the blowdown used in the analysis was conservatively assumed to consist of dry steam although entrainment can be expected on the double-ended rupture. The significant parameters Amendment 63 Page 37 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 affecting the mass and energy releases to containment following a steam line break are discussed below. 6.2.1.4.2 Plant power level Steam line breaks can be postulated to occur with the plant in any operating condition ranging from zero to full power. Since steam generator mass decreases with increasing power levels, breaks occurring at lower power generally result in a greater total mass release to the Containment. However, because of increased energy storage in the primary plant, increased heat transfer in the steam generators, and the additional energy generation in the nuclear fuel, the energy release to the Containment from breaks postulated to occur during power operation may be greater than for breaks occurring with the plant at lower power levels. Additionally, pressure in the steam generators changes with increasing power and has a significant influence on the rate of blowdown. Because of the opposing effects of changing power level on steam line break mass and energy releases, no single power level can be singled out as a worst case initial condition for a steam line break. Therefore, a spectrum of power levels spanning the operating range (100.34%, 68.6%, and 29.4%), as well as zero power, has been considered. 6.2.1.4.3 Break type, area, and location

1. Break Type - There are two possible types of pipe ruptures which must be considered in evaluating steam line breaks.

The first is a split rupture in which a hole opens at some point on the side of the steam pipe, but does not result in a complete severance of the pipe. A single, distinct break area is fed uniformly by all steam generators until steam line isolation occurs. The blowdown from the individual steam generators is not independent since fluid coupling exists among all steam lines. Because of the flow-limiting orifices in each steam generator, the largest possible split rupture can have an effective area, prior to isolation, that is not greater than the throat area of the flow restrictor times the number of reactor coolant loops. Following isolation, the effective break area for the steam generator with the broken line can be no greater than the flow restrictor throat area. However, split ruptures have been evaluated to be non-limiting cases. The second break type is the double-ended guillotine rupture in which the steam pipe is completely severed and the ends of the break displace from each other. Guillotine ruptures are characterized by two distinct break locations, each of equal area, but are fed by different steam generators. The largest possible guillotine rupture can have an effective area no greater than the throat area of one steam line flow restrictor for each steam generator.

2. Break Area - Two break areas (one full double-ended, and one split rupture) have been analyzed at each of the four initial power levels, as follows:

a. A full double-ended pipe rupture downstream of the steam line flow restrictor. For this case, the actual break area equals the cross sectional area of the steam line, but the blowdown from the steam generator with the broken line is controlled by the flow restrictor throat area (1.4 ft.2). The reverse flow from the intact steam generators is Amendment 63 Page 38 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 controlled by the smaller of the pipe cross section, or the total flow restrictor throat area for both the intact loops.

b. A split break that represents the largest break which will not generate a steam line isolation signal from the primary protection equipment. Steam and feedwater line isolation signals will be generated by high containment pressure signals for these cases. However, split ruptures have been determined to be non-limiting cases.

3. Break Location - Break location affects steam line blowdown by virtue of the pressure losses which would occur in the length of piping between the steam generator and the break. The effect of the pressure loss is to reduce the effective break area seen by the steam generator. Although this would reduce the rate of blowdown, it would not significantly change the total release of energy to the Containment. Therefore, piping loss effects have been conservatively ignored in all blowdown results.

6.2.1.4.4 Main feedwater addition prior to feedwater line isolation All of the double-ended ruptures generate main steam and feedwater isolation signals very quickly following the break. Isolation of these lines is assumed to be complete following a time delay sufficiently long to allow for instrument response time and signal processing delay (2 seconds) and valve closing time. The total delay to complete isolation of the steam lines is 7 seconds including the instrument response and signal processing delay. The total delay to complete isolation of the feedwater lines is 10 seconds including the instrument response and signal processing delay. (For steam line breaks initiated at zero power, the total feedwater isolation delay is 12 seconds.) For the split ruptures, the feedwater isolation signal and the main stream line isolation signal result from high containment pressure protective trips. The containment pressure setpoints for feedwater line and steamline isolation signals is assumed to be 3.0 psig. The isolation is assumed to be complete 7 seconds (instrument/signal delay and valve closure time) after the setpoint is reached for the main steam lines and 10 seconds after the setpoint is reached for the feedwater lines. (For the steam line breaks initiated at zero power, the total feedwater isolation delay is 12 seconds.) Prior to complete isolation, the depressurization of the steam generator results in significant amount of feedwater being added to the broken loop steam generator through the Feedwater System. The quantity of feedwater added is conservatively evaluated using the following assumptions:

1. Two main feedwater pumps operating and feedwater control valve position is the same as that expected for normal operation at a given power level. At zero power, two pumps are assumed to be operating, however flow is controlled by the feedwater control bypass valve and the main control valve is closed. An alternate flow path at zero power using the AFW pump and AFW valves to control flow was evaluated and found to be bounded by the mass addition using the MFW pumps and MFCBV flow alignment.

2. The feedwater control valves in the intact loops maintain their initial flow until feedwater isolation signal is received. Immediate closure of the feedwater isolation valves and control valves in the intact loops upon receipt of the isolation signal.

Amendment 63 Page 39 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

3. In the faulted steam generator loop, three failure scenarios are postulated:

For the cooling train failure case, both MFIV and MFCV (or MFBCV) are expected to function and isolate feedwater upon receipt of a MFIS. Flow reduction through the valves is not credited as they stroke closed. For the MFIV failure case, the MFCV (or MFBCV) is expected to close upon receipt of a MFIS. Flow reduction through the valve is not credited as it strokes closed. For the MFCV failure case (or MFBCV failure at 0% power), the MFCV (MFBCV) is assumed to ramp open immediately upon a MSLB and feedwater flow to the faulted steam generator increases. The MFIV is assumed to close upon receipt of a MFIS. Flow reduction through the valve is not credited as it strokes closed.

4. The pressure in the intact loop steam generators remains at the level existing prior to a double-ended guillotine rupture, while the broken loop steam generator depressurizes. The pressure in the intact loop steam generators decays at the same rate as the broken loop steam generator pressure subsequent to a split rupture.

These assumptions were used along with the feedwater system hydraulic resistances and pump performance curves to determine the amount of feedwater added to the steam generator with the broken loop. The amounts of main feedwater added to the faulted steam generator for the 100.34%, 68.6%, and 29.4% power levels have been calculated using the RELAP5 computer code. Feedwater addition at 0% power was determined based upon a conservative calculation. 6.2.1.4.5 Auxiliary feedwater system design Generally within the first minute following a steam line break, the Auxiliary Feedwater System is initiated on any one of several protection system signals. Addition of auxiliary feedwater to the steam generators increases the secondary mass available for release to the Containment as well as increasing the heat transferred to the secondary fluid. A conservative bounding AFW flowrate of 3000 gpm was assumed to enter the faulted steam generator from all 3 AFW pumps up until the time of isolation. After isolation, the AFW isolation valves were assumed to leak, and a leak flow of 20 gpm was assumed to enter the faulted steam generator. Auxiliary feedwater flow is assumed up until the time automatic auxiliary feedwater isolation takes place. For a description of the automatic auxiliary feedwater isolation logic see Section 10.4.9. 6.2.1.4.6 Fluid stored in the feedwater piping prior to isolation The unisolated feedwater line volume between the steam generator and the isolation valve is a source of additional high energy fluid to be discharged through the break. This volume was assumed to be 245 cubic feet. For the MFIV failure case, an additional 455 cubic feet of fluid stored in the piping between the MFIV and the MFCV was also assumed to be discharged through the break. In addition a purge volume of 50 cubic feet of AFW piping between the steam generator and AFW isolation valve was also assumed to be discharged into the faulted steam generator and out of the break. Amendment 63 Page 40 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.1.4.7 Limiting feedwater valve failure As a result of the pumped feedwater and unisolable feedwater piping volume, the analysis addresses the maximum amounts of feedwater addition to the faulted steam generator in calculating the dry out time. Since the dryout times for the MFCV failure case are significantly smaller than those for the MFIV failure case, the consequences of the MFCV failure cases are enveloped by the MFIV failure cases and were not analyzed. 6.2.1.4.8 Fluid stored in the steam piping prior to isolation For the double-ended ruptures, all the steam in the steam lines up to the turbine stop valve (9415 ft3) is assumed to be released to the containment following the break. The split ruptures that do not assume a failure of the MSIV use the steam between the steam generator and the MSIV (1025 ft3) as the unisolable volume. 6.2.1.4.9 Availability of offsite power Loss of offsite power following a steam line rupture would result in tripping of the reactor coolant pumps, motor-driven main feedwater pumps, and a possible delay of auxiliary feed initiation due to standby diesel generator starting delays. Each of these occurrences aids in mitigating the effects of the steam line break releases by either reducing the fluid inventory available to feed the blowdown or reducing the energy transferred from the Reactor Coolant System to the steam generators. Thus, blowdowns occurring in conjunction with a loss of station power are less severe than cases where offsite power is available; these cases are not presented. 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies of Emergency Core Cooling System The containment backpressure for the limiting case for the ECCS analysis is calculated using the methods and assumptions described in Section 15.6.5. Input parameters including the containment initial conditions, net free containment volume, passive heat sink materials, thicknesses, surface areas, starting time, and number of containment heat removal systems used in the analysis are described below. The large break LOCA ECCS performance analysis was performed with a loss of offsite power as the most limiting condition with respect to margin to the 10CFR50.46 acceptance criteria. That is, a more challenging PCT results from assuming a loss of offsite power (reactor coolant pumps trip) rather than offsite power being available (reactor coolant pumps running). This results from core thermal hydraulics behavior during blowdown and is true even though the calculated containment pressure may be lower when offsite power is available due to faster actuation of the engineered safeguards. For the ECCS performance analysis, a dominant effect during the blowdown phase is the time to critical heat flux (CHF). The time to CHF significantly affects the amount of stored energy released to the coolant prior to entering the subsequent core heatup and reflood phase. The remaining stored energy at the end of blowdown significantly affects the peak clad temperature (PCT). If offsite power is lost at event initiation, an immediate flow reversal occurs as reactor coolant system mass exits the cold leg break. The flow reversal results in flow stagnation in the core decreasing the time to CHF and reducing clad-to-coolant heat transfer. The shorter time to Amendment 63 Page 41 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 CHF minimizes the stored energy released from the fuel rods during blowdown and presents a greater challenge to the PCT acceptance criterion. For the offsite power available scenario, faster actuation of the engineered safeguards can result in a small decrease in containment pressure which leads to a small decrease in core reflood rate and a small increase in PCT. However, the time to CHF is significantly delayed if the reactor coolant pumps remain running since flow reversal and stagnation do not occur. The more dominant effect on PCT of delayed time to CHF more than offsets the secondary effect of slightly reduced containment pressure. The PCT is less challenging to the 10CFR50.46 acceptance criterion when offsite power is available. Thus, the overall effect of assuming offsite power is available during a large break LOCA event is to obtain a more favorable result. The ECCS performance analysis assumption of loss of offsite power is limiting and the results presented in Section 15.6.5 demonstrate compliance with 10CFR50.46 for this limiting case. 6.2.1.5.1 Mass and energy release data The mathematical models which calculate the mass and energy releases to the Containment are described in Section 15.6.5. Since the requirements of Appendix K of 10 CFR 50 are very specific in regard to the modeling of the RCS during blowdown and the models used are in conformance with Appendix K, no alterations to those models have been made in regard to the mass and energy releases. A break spectrum analysis is performed (see references in Section 15.6.5) that analyzes various break sizes, break locations, and Moody discharge coefficients for the double ended cold leg guillotines which do affect the mass and energy released to the Containment. This effect is considered for each case analyzed. During reflood, the effect of steam-water mixing between the safety injection water and the steam flowing through the RCS intact loops reduces the available energy released to the containment vapor space and therefore tends to minimize containment pressure. 6.2.1.5.2 Initial containment internal conditions The following initial values were used in the analysis: Containment pressure 14.0 psia Containment temperature 80°F RWST temperature (ECCS) 82.5°F RWST temperature (sprays) 40°F Outside temperature 60°F Initial Relative Humidity 100 % The combination of containment initial conditions used in the analysis are conservative relative to the values anticipated during normal full power operation. 6.2.1.5.3 Containment volume The volume used in the analysis is 2.344 x 106 ft.3. Amendment 63 Page 42 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.1.5.4 Active heat sinks The Containment Spray System and the containment fan coolers operate to remove heat from the Containment. Pertinent data for these systems which were used in the analysis are presented in Table 6.2.1-

62. The heat removal capability of each fan cooler is presented in Figure 6.2.1-303.

The containment sump temperature was not used in the analysis because the maximum peak cladding temperature occurs prior to initiation of the recirculation mode for Containment Spray System. In addition, heat transfer between the sump water and the containment vapor space was not considered in the analysis. 6.2.1.5.5 Steam-water mixing Water spillage rates from the broken loop accumulator are determined as part of the core reflooding calculation and are included in the containment code calculation model. 6.2.1.5.6 Passive heat sinks The passive heat sinks used in the analysis, with their thermophysical properties, are given in Table 6.2.1-63. Concrete thermophysical properties utilized were taken directly from BTP CSB 6 1. A carbon steel thermal conductivity value of 26Btu/hr-ft-F is specified for the temperature range of interest for Shearon Harris from Reference 6.2.5-5; likewise, a volumetric heat capacity value is obtained from that reference. The values shown in Table 6.2.1-63 were used in the analysis. 6.2.1.5.7 Heat transfer to passive heat sinks The condensing heat transfer coefficients used for heat transfer to the steel containment structures were calculated in accordance with NRC Branch Technical Position CSB6-1. 6.2.1.5.8 Containment purging during a LOCA The containment purge system consists of two 8-inch diameter lines and associated isolation valves. During the time between event initiation and complete closure of the isolation valves, containment purging occurs which can adversely affect the core reflood rate and PCT for a large break LOCA by reducing containment backpressure. Over the short period of time that the isolation valves are open, the pressure decrease resulting from containment purging is small and will have an insignificant effect on the core reflood rate and PCT. 6.2.1.5.9 Other parameters No other parameters have a substantial effect on the minimum containment pressure analysis. 6.2.2 CONTAINMENT HEAT REMOVAL SYSTEM The purpose of the Containment Heat Removal System (CHRS), is to rapidly reduce the containment pressure and temperature following a reactor or steam generator energy release Amendment 63 Page 43 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 and to maintain them at acceptably low levels. The CHRS also serves to limit offsite radiation levels by reducing the pressure differential between the containment atmosphere and the external environment, thereby decreasing the driving force for fission product leakage across the Containment. In addition, airborne iodine following a loss-of-coolant accident (LOCA) is removed by spraying borated sodium hydroxide solution into the Containment. The fission product removal function is carried out by the Iodine Removal System (IRS) as discussed in Section 6.5.2. The systems provided for containment heat removal include the Containment Cooling System (CCS) and Containment Spray System (CSS). The Containment Cooling System is designed to operate during both normal plant operations and under LOCA or main steam line break (MSLB) conditions. The operations of the CCS are discussed in Section 6.2.2.2.1. The CSS is designed to operate during accident conditions only. The operation of the CSS is discussed in Section 6.2.2.2.2. 6.2.2.1 Design Bases The CCS and the CSS are designed to remove heat from the containment atmosphere following a LOCA accident or a secondary system rupture inside Containment, as required by General Design Criteria 38. The CCS also provides a supply of cooling air to the annular clearance between the reactor vessel and primary shield wall, the reactor vessel supports and the annular space between the reactor coolant legs and the concrete wall.

1) The sources and amounts of energy released to the Containment as a function of time which were used as the basis for sizing the Containment Heat Removal System are given in Sections 6.2.1.3 and 6.2.1.4. The CCS is designed to remove its heat load while the essential portions of the Service Water System (SWS) is providing cooling water to the CCS at 95°F. This is conservative-based on a maximum operational Service Water Inlet temperature of 94°F. A description of the SWS is presented in Section 9.2.1.

2) The heat removal capacity of either train of the CCS and CSS is sufficient to keep the containment temperature and pressure below design conditions for any size break up to and including a double ended break of the largest reactor coolant pipe. The system is also designed to mitigate the consequences of any size break in the secondary systems, up to and including a double ended break of the largest main steam line inside Containment.

3) The CCS and CSS each consist of two redundant loops and are designed such that failure of any single active or passive component will not prevent adequate post-accident cooling of the containment atmosphere.

4) The safety related portions of the CCS and CSS are designed to Safety Class 2, Seismic Category I requirements.

5) The CCS and CSS are protected against the dynamic effects associated with postulated fluid system piping failures as described in Section 3.6.

6) Protection of the CCS and CSS from the effects of missiles is described in Section 3.5.

7) Protection of the CCS and CSS from the effects of wind/tornado and flooding is described in Sections 3.4 and 3.5.

Amendment 63 Page 44 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

8) Both the CCS and CSS are designed to permit periodic inspection and testing as described in Section 6.2.2.4.

9) The essential portions of the CCS and CSS located inside the Containment are designed to withstand the containment environment resulting from a LOCA or MSLB. The environmental conditions resulting from a LOCA or MSLB are described in Section 3.11.

10) The Primary Shield Cooling system and the Reactor Supports Cooling System are designed to supply cooling air to the annular clearance between the reactor vessel and primary shield wall, the reactor vessel supports and the annular space between the reactor coolant legs and the concrete wall. The systems are designed to limit the temperature of the shielding concrete, instrumentation and concrete base at the reactor vessel supports to a maximum of 150°F. The systems are designed to Safety Class 3 and Seismic Category I requirements.

11) The CSS is capable of withstanding the dynamic effects associated with hydraulic instabilities occurring during any mode of operation.

6.2.2.2 System Design 6.2.2.2.1 Containment Cooling System (CCS) 6.2.2.2.1.1 Functional description The CCS has the following functions:

1. In the event of a design basis accident, LOCA or MSLB, containment fan coolers are designed to remove heat in the following manner:

a. Four containment fan coolers will operate with one of the two fans in each cooler running at half speed (the other fans are idle). Heat removal capacity per containment fan cooler is stated in FSAR Table 6.2.2 1.

b. In the case of single train failure, two containment fan coolers will operate with one of the two fans in each cooler running at half speed (the other fans are idle).

2. During normal operation, the CCS is designed to maintain the indicated containment temperature below 120°F.

3. Mixing the containment atmosphere following an accident.

6.2.2.2.1.2 Design description The CCS consists of four safety related fan cooler units and three non-safety fan coil units. Following a design basis accident only the safety related fan cooler units are required to operate. During normal power operation, safety related units operate in conjunction with the non-safety units to maintain required containment temperature. See Table 6.2.2-1 for major system components. Figure 6.2.2-3 describes the extent of essential portions of the ductwork and equipment for the CCS. Air is supplied to the steam generator and pressurizer subcompartments, the operating floor, the ground floor and the mezzanine floor. Figures 6.2.2-10 through 6.2.2-16 describe the plan and elevation drawings of the Containment showing the Amendment 63 Page 45 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 routing of air distribution ductwork. A portion of supply air is tapped to serve the Reactor Support Cooling System and Primary Shield Cooling System described in Section 6.2.2.2.3. Two of the four safety related fan cooler units are located at Elevation 236', the remaining two safety related units are located at Elevation 286'. Two separate trains are provided, each consisting of two fan cooler units with each unit supplying air to an independent, vertical concrete air shaft. Train A Components Train B Components Fan Cooler AH-2 Fan Cooler AH-1 Fan Cooler AH-3 Fan Cooler AH-4 Service Water Loop A Service Water Loop B Emergency Power Diesel A Emergency Power Diesel B Train selection of each fan cooler with its respective water supply is under administrative control. Each fan cooler is served by water from the Service Water System. A detailed description of the Service Water System is given in Section 9.2.1. Each safety related fan cooler consists of cooling coil sections and two direct driven vane axial flow fans. Unit performance data is shown in Table 6.2.2 1. Each fan is equipped with a two speed motor enabling half speed operation at DBA conditions and integrated leak rate test conditions. A gravity damper is provided at the discharge side of each fan to prevent air flow in the reverse direction when only one fan per unit is required to operate. Both fans of the unit discharge into a common duct which is connected to a concrete air shaft through a locked open damper. A branch duct connection is provided to serve as a post-accident discharge nozzle and is normally isolated by means of a separate pneumatically operated, fail open damper. The three non-nuclear safety fan-coil units are all located at the same elevation. These units are required to operate during normal plant operating conditions only; their air is directed to Reactor Coolant Pump and Steam Generator Compartments. The fan-coil units are served by the Service Water System. A detailed description of Service Water System is given in Section 9.2.1. Each unit has cooling coil section and two one hundred percent capacity, direct driven, vane axial fans. Unit performance is shown in Table 6.2.2-1. 6.2.2.2.1.2.1 Post-accident operation During post-accident operation, four fan cooler units operate with one fan per unit running at half speed. The system can operate in this mode as long as both emergency diesel generators and both service water system trains are available. In the event of failure of one of the emergency diesel generators or one service water system train only two fan cooler units will operate. The damper in the post-accident discharge branch duct will be opened. The post-accident discharge duct is provided with high velocity nozzles to diffuse air to accelerate the temperature mixing inside containment. These nozzles are directed to selected areas of heat release, to achieve thorough mixing of containment atmosphere. The Amendment 63 Page 46 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 high velocity nozzles direct turbulent air jets from discharge points at two levels inside containment where two separate trains of containment fan coolers are located. Two sets of nozzles are located at Elevation 286 ft. as shown on Figure 6.2.2-14, Sections C-14-1 and C-12-1, and the other two nozzles are shown on Figure 6.2.2-10 (plan at Elevation 221.00 ft.) as post-accident discharge nozzles. Seismic Category I ductwork is used from the fan coolers to the discharge outlets. As the post-accident containment atmosphere steam-air mixture passes through the system cooling coils, it is cooled and a portion of the steam is condensed. In the event of a single active failure in one train, one containment spray pump and two containment fan coolers will provide the adequate cooling capacity. The fan cooler units receive electric power from the diesel generators approximately 15 seconds after SIAS generation through a timer-sequencer. However, due to a time delay relay a fan running in high speed will be allowed to coast down for 15 seconds to allow for low speed synchronization. Approximately 8 additional seconds are required to bring the fans to the operational speed. The containment fan cooler performance data, showing the energy removal rate is shown on Figure 6.2.2-4 and Table 6.2.2-3. 6.2.2.2.1.2.2 Normal operation During normal power operation, three non-safety fan coil units are in continuous operation along with two of or all four of safety-related fan cooler units. The following describes their operation: a) When containment average temperature is 118°F or below: Normally two fan cooler units will operate with both fans of the unit running at full speed. Each of the two vertical concrete air shafts is served by an operating fan cooler unit. In this mode of operation, the idle train is serving as standby. Each shaft supply air damper is locked open and each nozzle damper is closed. b) When the containment average temperature is above 118°F or if additional cooling is desired, additional coolers will be operated. Fan cooler units located at floor Elevation 236 ft. will operate with one of the two fans of the units running at full speed and the other fan is on standby. Each shaft supply damper is locked open and each nozzle damper is closed. The other two fan cooler units located at Elevation 286 ft. will operate with both fans per unit operating at full speed. Each shaft supply damper is locked open and each nozzle damper is open. If containment average temperature continues to rise or if additional cooling is desired, the two standby fans of the fan coolers at Elevation 236 ft. will be manually energized to operate at full speed and the nozzle dampers will remain closed. c) With (2) safety related fan cooler units and (3) non-safety related fan coil units operating at a service water temperature of 50°F, their total heat removal capacity is approximately 11.1x106 Btu/hr. These capacities are based on air entering the units at 80°F DB and between 48°F and 67°F WB. The containment heat gain is approximately 13.8x106 Btu/hr. This includes heat contributed from equipment, lighting, piping, motors as well as fan motors. Amendment 63 Page 47 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Since heat gain is greater than the heat removal rate the temperature in the Containment cannot fall below 80°F. 6.2.2.2.2 Containment Spray System (CSS) 6.2.2.2.2.1 Functional description The purpose of the CSS is to spray borated sodium hydroxide solution into the Containment to cool the atmosphere and to remove the fission products that may be released into the containment atmosphere following a LOCA or MSLB. A summary of the design and performance data for the CSS is presented in Section 6.2.1. The fission product removal effectiveness and the pH control of the containment sump water of the CSS is described in Section 6.5.2. 6.2.2.2.2.2 Design description The CSS consists of two independent and redundant loops each containing a spray pump, piping, valves, spray headers, and spray valves. Figure 6.2.2-1 provides the process flow and instrumentation details of the system. The operation of the CSS is automatically initiated by the containment spray actuation signal (CSAS) which occurs when a containment pressure HI-3 signal is reached. Section 7.3 describes the design bases criteria for the CSAS. Upon receipt of a CSAS, the containment spray pumps start operation and the containment spray isolation valves open. The CSS has two principal modes of operation which are: a) The initial injection mode, during which time the system sprays borated water which is taken from the refueling water storage tank (RWST). Section 6.2.2.3.2.3 describes the criteria used for sizing the RWST. b) The recirculation mode, which is initiated when low-low level is reached in the RWST. Pump suction is transferred from the RWST to the containment sump by opening the recirculation line valves and closing the valves at the outlet of the refueling water storage tank. This switch over is accomplished automatically. See Section 7.3 for further details. Upon receipt of the CSAS the containment spray pumps are started and borated water from the RWST is discharged into the Containment through the containment spray headers. The CSAS starts the two containment spray pumps and opens the motor operated containment spray isolation valves. Upon reaching full speed of the containment spray pumps, water will reach the nozzles and start spraying within approximately 33 seconds. The spray headers are located to maximize heat removal. Each train at the CSS has two headers which conform to the shape of the Containment and contain a total of 106 spray nozzles per train. The number of spray nozzles in the system provides 100 percent redundancy for effective heat removal and iodine removal. Figure 6.2.2-2 provides the location of spray piping and nozzles and the resulting spray pattern. Refer to Section 6.5.2 for a discussion of Containment sprayed and unsprayed volumes. Amendment 63 Page 48 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 A flow element is installed in each containment spray pump's discharge line to monitor the system operation. The spray nozzles, which are of open throat design, without any moving parts (minimum inside diameter of approximately 0.375 in.), break the flow into small droplets, which increases the cooling effectiveness on the containment atmosphere. As these droplets fall through the containment atmosphere they absorb heat until they reach the temperature of the containment air-steam mixture. The spray nozzles are protected from clogging by the following means: There are two independent sumps which serve as reservoirs and provide suction to the ECCS and Containment Spray (CT) system pumps during the recirculation mode of operation. The recirculation sumps are located inside the containment building outside the secondary shield wall at elevation 221'-0" and at azimuths 2250° and 3150°. The sumps are covered with checker plate steel covers. Before water enters the fine strainer assemblies, it passes through coarse trash racks which are vertical. The vertical trash racks have approximately 2" x 2" openings except that the bottom 12" of these racks have been removed to assure that water is always able to flow under them even if the openings become plugged (Figure 6.2.2-19). The fine strainer assemblies behind the trash racks consist of a total of one hundred thirty-six (136)(68 per sump) high-performance top hat style assemblies and four (4) top hat inspection port assemblies (2 per sump) which will provide a total net effective surface area of approximately 6,000 ft2 (3,000 ft2 per sump) (Figure 6.2.2-20). A concrete wall is located inside each recirculation sump separating the Residual Heat Removal (RHR) pump intake from the containment spray (CT) pump intake. Thirty five vertical top hats are located on each side of the concrete wall for a total of 70 top hats per sump. Since RHR flows exceeds CT flow, there is a 4" x 18" opening in the concrete wall connecting the two sides of the sump to allow water to flow from the CT side to the RHR side of the sump. The top hats are 66 inches long with a 13 1/4" x 14 1/2" flange (baseplate) on one end. The high-performance top hat assemblies consist of four tubes (12-inch, 10-inch, 7-inch and 5-inch diameter) fabricated from perforated stainless steel plate with 3/32" perforations. The top hat inspection port assemblies consist of three tubes (12-inch, 10-inch, and 7-inch) fabricated from perforated stainless steel plate plus a non-perforated tube (5-inch) with a blank flange on top that can be removed to look through the top hat. A top hat support frame is anchored to the sump walls with vertical supports going to the recirculation sump floor at elevation 216'-4 1/2". In this design, water enters through the perforated plate surfaces of the strainers and travels through the annuli created between the two outer tubes and the two inner tubes (note that the top hat inspection port assemblies do not contain a 5-inch inner perforated tube). The flow then travels underneath the support frame to the RHR and CT suction intakes. A vortex suppressor made from standard floor grating is installed above the vertical top hat modules in each recirculation sump to prevent air from being drawn into the top hat modules. A curb approximately 18 in. high and located 2 ft. in front of the screen structure is provided to prevent heavy or sunken debris from impingement upon the screens. The floor outside of the curb slopes away from the sump to minimize debris from entering the sump. The containment recirculation sumps have been designed and constructed to ensure the functional capability of the sumps to provide an adequate supply of water during the recirculation mode of operation for the Containment Spray System and the Residual Heat Amendment 63 Page 49 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Removal System. In addition, the containment recirculation sumps have been evaluated against the guidelines provided in NUREG 0869, Revision 1 "Unresolved Safety Issue A-43 Regulatory Analysis". This evaluation concluded that post-LOCA insulation debris will not degrade either the performance of the containment sumps or that of the RHR pumps and the CS pumps. The evaluation demonstrates that based upon sump location, containment building layout and the jet impingement effects associated with a postulated LOCA, SHNPP insulation cannot be transported to the sump screens either in the short-term as a direct result of blowdown forces, or during long-term recirculation since the 0.1 ft/sec velocity of the water as it approaches the sumps is less than that required to transport insulation debris to the screens. The sump structures and screens are designed to withstand the effects of a Safe Shutdown Earthquake (SSE) without loss of structural integrity. Thus, the sump screens are designed to the Seismic Category I structural criteria. Figures 6.2.2-7, 6.2.2-8, and 6.2.2-9 show the plan and section views of the containment sump. Piping and equipment insulation is considered to be the primary source of post-accident debris inside Containment which could potentially clog the sump screening. The possibility of paint chips peeling off has been minimized by requiring proper surface preparation and by painting larger surface components with coatings which have been qualified under design basis accident condition. Non-NSSS-supplied thermal insulation inside Containment consists primarily of metallic reflective insulation. The insulation is constructed of stainless steel interior and exterior sheets. All insulation assemblies are designed to be self-supporting from the associated piping and equipment or from adjacent removable or permanent covering. Permanent insulation assemblies are attached by stainless steel straps and fasteners of the expansion type which prevent overstressing of the bands or damage to the coverings due to thermal expansion of the equipment surface. Removable assemblies are attached by means of stainless steel buckles or other fasteners of the quick release type which vary depending upon installation requirements. Each insulation assembly is jacketed in heavy gage stainless steel or stainless steel wire mesh for the RSG primary side channel heads and designed to withstand vibration and seismic shock associated with postulated accident conditions inside Containment. With the exception of local failure in the vicinity of postulated pipe ruptures, insulation assemblies are expected to remain intact during and after an accident. Westinghouse-supplied insulation for inside containment equipment application consists mainly of stainless steel reflective panels of various sizes ranging from 12 by 18 inches to 24 by 48 inches. The thickness ranges from 3 to 3 1/2 inches. This type of insulation may be found on the pressurizer, reactor coolant pump casings, and the primary piping consisting of the hot, cold, and crossover legs and the pressurizer surge line. As a result of the steam generator replacement, the steam generators are insulated with fiberglass blanket insulation with stainless steel jacketing (or wiremesh for the primary side channel heads). The reactor vessel is also covered with this type of insulation with the exception of the beltline region, from the nozzles down approximately 48 inches. In this region, the vessel is covered with a heavier sandwich design consisting of Microtherm thermal insulation and Ricorad neutron Amendment 63 Page 50 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 shielding encapsulated in stainless steel. The Microtherm insulation is closest to the reactor vessel and is approximately 1 to 1 1/2 inches thick. Surrounding this is 1 to 2 inches of Ricorad shielding. The stainless steel exists both around and between this combination. This heavier insulation paneling is supported from the reactor vessel nozzles through inter-fastening of sheet metal screws to adjacent panels and by vertical support straps. The sizes for these panels fall within the sizes given for the reflective panels. Cutouts are provided in the insulation for equipment, seismic supports, and tie downs. The containment sump has screens with 3/32in. openings. This is adequate because there are no openings in the ECCS or containment spray system that are more restrictive than 3/32in. Adequate means are provided for convenient access to the sump for inspection and maintenance purposes. The containment recirculation sumps are periodically inspected as delineated in the Technical Specifications. 6.2.2.2.3 Primary shield and reactor supports cooling system The Primary Shield Cooling System and the Reactor Supports Cooling System are shown on Figure 6.2.2-3. 6.2.2.2.3.1 Primary shield cooling system The Primary Shield Cooling System consists of two Safety Class 3, 100 percent capacity, direct driven supply fans. Each fan serves as a standby for the other fan and is served by a separate power channel. Fan design data are provided in Table 6.2.2-4. Each fan is provided with a locked open inlet damper and a gravity type discharge damper to prevent air recirculation through the standby fan. Each axial supply fan draws 18,000 cfm cool air from the vertical concrete air shaft and supplies it to the annular clearance between the reactor vessel and primary shield wall through connecting ductwork. The cooling provided by the Primary Shield Cooling System minimizes the possibility of concrete dehydration and subsequent faulting. 6.2.2.2.3.2 Reactor Supports Cooling System The Reactor Supports Cooling System consists of two Safety Class 3, 100 percent capacity direct driven vane axial fans. Each fan serves as a standby for the other fan. Fan design data are presented in Table 6.2.2-5. Each fan is provided with a locked open inlet damper and a gravity type discharge damper to prevent air recirculation to the idle fan. The system draws 27,600 cfm of cooling air from the vertical concrete air shaft and supplies 21,600 cfm of air to the reactor vessel supports and 1000 cfm each to the annular space between reactor coolant legs (nozzle) and sleeves. Cool air is forced through these spaces uniformly by means of a ductwork distribution system. The cooling provided by the Reactor Supports Cooling System limits thermal expansion of the reactor vessel supporting steelwork. Amendment 63 Page 51 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.2.3 System Design Evaluation 6.2.2.3.1 Containment Cooling System (CCS) Cooling units, with associated piping, valves, and instrumentation, are located outside the primary shield and above the maximum possible post-accident water height to provide protection against flooding. Although the ECCS is designed to rapidly cool the water in the core below saturation temperature following a LOCA, the CCS design is based on the assumption that all core residual heat appears as steam in the Containment. The CCS cooling coil design provides for rapid drainage of large quantities of condensed steam, preventing loss of capacity and maintaining cooling water temperatures below the boiling point. A relief valve is provided to prevent excess tube pressure. Since the cooling coils are in constant use, tube clogging during an accident is highly unlikely. Surface fouling on the secondary side of the fan cooler heat exchanger by the cooling water is minimized by the use of Cu Ni 90/10 tubes. Performance of the cooling unit was predicted assuming a fouling factor of .001. Service water flow to the cooling unit coils is unregulated to eliminate the possibility of a failure due to a modulating valve or controller malfunction. Each containment fan cooling unit has a separate branch supply and return run through the containment wall, with an isolation valve located outside the Containment. High reliability is maintained through careful quality control and assurance procedures and by general arrangement of equipment and piping to provide access for inspection and maintenance. Safety-related components are designed to operate in, and to withstand, post-accident environment, resulting from postulated design basis accidents. See Section 3.11 for a description of the design basis for environmental considerations. All safety-related dampers are pneumatically operated. Dampers will fail in the safe (either closed or open) position in the event that electrical power or air is lost to the damper operator. The heat sink for the containment cooling units is the Service Water System. Failure of an inlet or outlet valve to a containment cooling coil will be detected due to flow reduction since water side flowrates are monitored via appropriate instrumentation. During the post-accident period, most of the containment cooling ductwork system is not required. Cooling air is reapportioned by means of Safety Class 2 dampers to discharge nozzles adjacent to the fan discharge. With the exception of a small amount of ductwork between the fan outlet and concrete air shaft, the major portion of the ductwork which could collapse and damage other safety-related systems is provided with Seismic Category I duct supports. The essential portions of the CCS ductwork and equipment housings are designed for a two psid pressure differential to prevent overpressurization. No single failure in the CCS would render the containment heat removal system incapable of performing its post-accident cooling function. See Table 6.2.2-6. Amendment 63 Page 52 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.2.3.2 Containment Spray System CSS The single failure characteristics of the CSS have been evaluated to show that failure of any single active component will not prevent adequate post-accident cooling of the containment atmosphere during the injection phase. No single active or passive failure (not in addition to a single active failure in the injection phase) during the recirculation phase will render the Containment Heat Removal System incapable of performing its required safety function. See Table 6.2.2-7. One containment spray pump and two of the containment cooling units will provide at least 100 percent cooling capacity. One of two spray additive eductors will supply adequate sodium hydroxide solution to provide minimum required iodine removal. See Section 6.5.2 for further details. The containment spray pumps take suction from the refueling water storage tank during the injection phase. The pumps take suction from the containment sumps during the recirculation phase. Each pump has a separate suction line from its associated sump. Class 1E level instruments LE-7160 SA & SB are provided in containment sumps 1A & 1B, respectively. These level instruments provide indication in the main control room of the water level in the respective sumps upstream of the strainer screens. The containment recirculation sumps are located at the outer perimeter of containment floor Elevation 221.00 feet. Any water from pipe breaks, drain flow or spray flow will be intercepted at higher elevations and directed to the reactor cavity sump by means of the floor drains system. Water must then flow radially out to the recirculation sump location which guarantees uniform flow approach. Figures 1.2.2-3 and 6.2.2-7, 6.2.2-8 and 6.2.2-9 provide additional details on sump layout and location. The containment sumps will be inspected following extended shutdowns for any materials which have the potential for becoming debris capable of blocking the recirculation of coolant following a LOCA. There will also be a periodic inspection of sump components such as screens and intake structures in accordance with Regulatory Guide 1.82. Figure 6.2.2-1 indicates the containment isolation valves provided for each of the independent lines leading from the containment sump to the suction of the containment spray pumps. There is one motor operated isolation valve located on each line external to the Containment. A secondary containment boundary, which incorporates an airtight protective valve chamber is provided. This secondary boundary completely encloses the sump line and the isolation valve and is not open to the containment atmosphere. Figure 6.2.2-1 indicates the containment isolation valves provided for each of the independent lines leading from the containment sump to the suction of the containment spray pumps. There is one motor operated isolation valve located on each line external to the Containment. A secondary containment boundary, which incorporates an airtight protective valve chamber is provided. This secondary boundary completely encloses the sump line and the isolation valve and is not open to the containment atmosphere. Amendment 63 Page 53 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The design basis fabrication requirements and quality control procedures for the containment secondary boundaries and valve chambers are identical to those used for the containment liner and the other containment penetrations (see Section 3.8). No single failure in the CSS sump lines or isolation valves during the recirculation phase will result in a loss of containment integrity. Reliability of the containment spray actuation signal is discussed in Section 7.3. Accidental initiation of the spray system will not affect the safety of the plant since all engineered safety feature instruments will be designed to operate in the resulting environment. All piping or equipment insulation which may come in contact with sprays will be covered with lagging to prevent large quantities of water from penetrating the insulation. Small amounts of seepage will not cause thermal shock to hot equipment. Receipt of the containment spray actuation signal will be alarmed. If the operator determines that initiation was inadvertent he may terminate spray flow, thus minimizing the amount of water entering the Containment. The procedures for terminating inadvertent containment spray are based on criteria, which require at least two operator errors to effect incorrect termination of the CCS. No automatic corrective systems to account for operator error are provided. The decision to terminate containment spray will be made only if a) the containment pressure, as indicated at least by three channels of the containment pressure instrument is less than the containment HI 3 pressure setpoint, or b) the containment pressure on two channels of the containment pressure instrumentation is less than the HI-3 pressure setpoint and a high pressure alarm is not activated. 6.2.2.3.2.1 CSS NPSH Requirements The NPSH requirements of the containment spray pumps have been evaluated for both the injection and recirculation phase following a loss-of-coolant accident. The minimum NPSH requirements and the available NPSH, (based upon final design) and the maximum expected flow through the pumps are listed in Table 6.2.2-8. As indicated in Table 6.2.2-8, recirculation operation gives the limiting NPSH conditions. The formulae and parameters used in the evaluation of the NPSH during both the injection phase and recirculation phase are the same as in the case of the low head injection pumps. No reliance is placed on the containment pressure for meeting the NPSH requirements for the containment spray pumps (however, credit is taken for the pressure necessary to maintain the fluid in its liquid phase, i.e., liquid vapor pressure). a) Injection phase: NPSH available = hrwst + hstatic - hfriction - hvapor pressure

                           = 34.3 + 70.6 - 8.4 - (4.5)
                           = 92.0 ft.

b) Recirculation phase: NPSH available = hcontainment + hstatic - hfriction - hvapor pressure Amendment 63 Page 54 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

                            = hcontainment + 28.1 - 2.6 - hvapor pressure Where:

hcontainment = hvapor pressure NPSHavailable = 25.5 ft. The minimum NPSH requirements are 13.0 ft. and 12.4 ft. for the injection phase and recirculation phase respectively. Positive net positive suction head margin is maintained with a postulated debris bed on the recirculation sump screens. 6.2.2.3.2.2 CSS spray coverage Two sets of spray nozzles are provided, each set oriented for effective coverage of the containment volume. Each spray header is located inside the containment dome. Figure 6.2.2-2 indicates the location of the spray nozzles within the Containment and indicates the expected spray pattern. The average height above the operating deck for containment spray trains A and B is 133 ft. and 140 ft. respectively. The average fall height of the spray droplets is conservatively taken as 125 ft. for determination of the iodine removal coefficient. The Spray Engineering Company spray nozzle, model number 1713A, is used for the CSS. Each spray nozzle is designed for a flow rate of 15.2 gpm with a 40 psi pressure drop across the nozzles. The nozzles are designed to produce droplets of approximately 700 microns mean diameter at the rated system conditions. Figure 6.2.2-6 is a sample spray nozzle drop size histogram. Reference 6.2.2-1 describes and presents the results of the spray nozzle test program performed by Spray Engineering Company which predicts the performance of the nozzle and the analytical methods employed to determine the mean spray drop size. 6.2.2.3.2.3 Refueling Water Storage Tank (RWST) The RWST capacity was determined on the basis of the following requirements: a) The tank must provide a minimum inventory to assure adequate containment sump level for proper recirculation phase operation. The tank will also provide that quantity of water required for at least 20 minutes of operation during the injection phase, with two high-head safety injection pumps, two low-head safety injection pumps and two containment spray pumps in operation. b) The tank must provide a quantity of water required to fill the Refueling Cavity, the Fuel Transfer Tube and the Fuel Transfer Canal, during refueling. c) The tank must provide an alternate source of boration for plant shutdown. d) The tank must provide a minimum inventory to assure a post-LOCA containment sump boron concentration sufficient to meet core subcriticality requirements for long-term cooling. The RWST is designed for a 469,260 gallon capacity with a minimum water inventory of 434,302 gallons maintained during all normal modes. This minimum inventory will only be Amendment 63 Page 55 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 removed from the RWST during unit refueling after shutdown and will always be maintained for post-accident recirculation mode operation and system testing. The two RWST vent lines are protected from freezing by redundant ambient sensing heat tracing on each vent line. The power supply for the heat tracing on each vent line is supplied by separate trains. The water in the RWST will be maintained at a temperature of not less than 40 F utilizing heaters, the minimum temperature for injection of borated water during emergency core cooling as indicated in Section 6.3. The freezing point of 2400 ppm boron solution, 1.37 weight percent boric acid, is below the normal freezing point of water, therefore a 40 F minimum temperature precludes freezing. In addition, the solubility temperature for a 2400-2600 ppm boron, 1.37-1.49 weight percent boric acid solution is below 40 F. The refueling water storage tank is a Seismic Category I field-fabricated tank of stainless steel construction. It is designed, fabricated, erected, and tested in accordance with the requirements of ASME Boiler and Pressure Vessel Code Section III, Winter Addenda 1971, Class 2. The RWST is designed for the horizontal and vertical seismic loads for both the Design and Operating Basis Earthquakes. The RWST would not be required for plant shutdown following a tornado. The tank is therefore, not designed for tornado winds or pressure drops. The major design parameters for the refueling water storage tank are indicated in Table 6.2.2-9. 6.2.2.3.2.4 Primary shield and reactor supports cooling system The Primary Shield and Reactor Support Cooling Systems are safety related and designed to Safety Class 3 and Seismic Category I requirements. Each system is provided with redundant fans to assure continuity and reliability of operation. Each fan is supplied with onsite emergency power from the diesel generators, in the event of loss of offsite power. 6.2.2.4 Testing and Inspection 6.2.2.4.1 Containment cooling system CCS The CCS undergoes preoperational startup tests as described in Section 14.2.12. Periodic tests are required as described in the Technical Specifications. Inservice inspection requirements are described in Section 6.6 and pump and valve testing requirements of Section 3.9.6 will apply. Factory tests verify cooling coil and motor performance. 6.2.2.4.2 Containment Spray System CSS See Section 6.5.2.4 for the testing and inspection requirements of the CSS. 6.2.2.4.3 Primary Shield and Reactor Supports Cooling System Refer to Section 14.2 for a discussion of testing provisions as they apply to the Primary Shield and Reactor Supports Cooling System. Amendment 63 Page 56 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.2.5 Instrumentation Requirements 6.2.2.5.1 Containment Cooling System CCS The instrumentation details and design requirements of the CCS are discussed in Section 7.3. 6.2.2.5.2 Containment Spray System CSS The following control room indications, utilizing the four containment pressure channels, aid the operator in determining pressure status. a) The four containment pressure channels activate the CSS (see Section 7.3) on the HI-3 pressure. The output of these four channels are shown on four indicators located on the control board.

1) These outputs activate one common annunciator alarm.

2) Each channel has individual trip status lights.

b) Three of these channels are utilized for SIS generation on high pressure.

1) Any of the three channels activates one common annunciator alarm.

2) Each channel has individual trip status lights.

c) Three channels (the same as in b above) are also utilized for main steam line isolation (2 out of 3 operation) on "HI-2" pressure.

1) These outputs activate one common annunciator alarm.

2) Each channel has individual trip status lights.

Flow measurement devices are provided, one on each of the two independent and redundant CSS loops. The containment spray flow is indicated by the ERFIS. The control room instrumentation which indicates RCS pressure is as follows: a) Three protection channels which supply signals to three pressurizer pressure indicators. b) Two control channels which supply signals to two pressurizer pressure indicators. c) The two control channels supply signals to low and high pressurizer pressure annunciation. The control room operator may utilize the following Control Room instrumentation to determine whether the "HI-3" containment pressure is a result of a steam line break or a primary system break (LOCA). The control room instrumentation which indicates steam generator pressure is as follows: a) Three channels and three pressure indicators. Amendment 63 Page 57 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 b) Three high differential pressure alarms. c) A low pressure alarm for each of the three steam generators. The control room instrumentation which indicates containment radiation is as follows: a) Containment radioactive air particulate indication. b) Containment area radiation monitoring. c) Containment room alarm for high radiation from above instrumentation. 6.2.2.5.3 Primary Shield and Reactor Supports Cooling System Indicator lights are provided to show blower status. 6.2.3 SECONDARY CONTAINMENT FUNCTIONAL DESIGN This section is not applicable to the Shearon Harris Nuclear Power Plant. 6.2.4 CONTAINMENT ISOLATION SYSTEM 6.2.4.1 Design Bases The Containment Isolation System consists of the valves and actuators required to isolate the Containment following a loss-of-coolant accident, steam line rupture, or fuel handling accident inside the Containment. The Containment Isolation System is designed to the following bases: a) The Containment Isolation System provides isolation of lines penetrating Containment, which are not required to be open for operation of the Engineered Safety Features Systems, to limit the release of radioactive materials to the atmosphere during a loss-of-coolant accident (LOCA). b) Upon failure of a main steam line, the Main Steam Line Isolation System, described in Section 7.3, isolates the faulted steam generator to prevent excessive cooldown of the Reactor Coolant System or overpressurization of the Containment, and as described in Section 7.3, the Containment Isolation System isolates the Containment. c) Upon failure of a main feedwater line, the Main Feedwater Isolation System, described in Section 7.3, isolates the faulted steam generator, and as described in Section 7.3, the Containment Isolation System isolates the Containment. d) Upon detection of high containment atmosphere radioactivity, isolation valves in the Containment Atmosphere Purge Exhaust System, discussed in Section 9.4.7, are shut to control release of radioactivity to the environment. The Containment Purge Isolation Actuation System is discussed in Section 7.3. Airborne radioactivity monitoring is discussed in Section 12.3.4.

Further information is contained in the TMI Appendix. Amendment 63 Page 58 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 All containment purge and vent isolation valves with the exception of those serving the Hydrogen Purge System as discussed in Section 6.2.5.1 close automatically on a high radiation signal generated as a result of inputs from containment airborne radiation sensors. All the automatically actuated valves have status indication lights in the Main Control Room. e) The Containment Isolation System is designed in accordance with 10 CFR 50, Appendix A, General Design Criterion 54 and Westinghouse Systems Standard Design Criteria, Number 1.14, Rev. 2. f) There are no lines that are part of the reactor coolant pressure boundary (RCPB) that penetrate the Containment (i.e., no safety class 1 lines), therefore GDC 55 is not applicable to SHNPP. However, for lines such as charging, safety injection, and letdown there is not an applicable GDC because these lines are connected to the RCPB but not part of the RCPB. Each line that is connected to the reactor coolant pressure boundary, and instrument lines as discussed in FSAR Section 6.2.4.2.4 is provided with containment isolation valves in accordance with 10 CFR 50, Appendix A, General Design Criterion 55, with the exception of the RHR hot leg suction lines as described below. g) Each line that connects directly to the containment atmosphere and penetrates Containment, with the exception of the residual heat removal and containment spray recirculation sump lines as discussed below and instrument lines as discussed in FSAR Section 6.2.4.2.4, is provided with containment isolation valves in accordance with 10 CFR 50, Appendix A, General Design Criterion 56. h) Each line that forms a closed system inside Containment, with the exception of the containment pressure sensing lines as described below, is provided with containment isolation valves in accordance with 10 CFR 50, Appendix A, General Design Criterion 57. i) Emergency power from the diesel generators is provided to ensure system operation in the event of a loss of offsite power. j) All air/spring-actuated valves are designed to fail to their required position to perform their safety function upon loss of the instrument air supply and/or electrical power. k) The containment isolation system design is such that the containment design leakage rate is not exceeded during a design basis accident. l) The Containment Isolation System is designed to remain functional during and following the safe shutdown earthquake. m) Closure times for containment isolation valves are established on the basis to minimize the release of containment atmosphere to the environment, to mitigate the offsite radiological consequences, and to assure that emergency core cooling system effectiveness is not degraded by a reduction in the containment back-pressure. n) Relief valves which are located between containment isolation valves are designed to meet the requirements for containment isolation valves. o) The steam generator shell and lines connected to the secondary side of the steam generator are considered to be an extension of the Containment and therefore, need no containment isolation valves located inside the Containment. Amendment 63 Page 59 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 p) The welding and qualification requirements for all welds associated with the spare penetration sleeve assemblies listed in Table 6.2.4-1 are in accordance with the appropriate requirements of Section III of the ASME B & PV Code. Provisions are made for leak testing the weld between the closure plate/cap and the embedded wall sleeve. The design requirements for spare penetration sleeves including their closure plates/caps listed in FSAR Table 6.2.4-1 are further described in Sections 3.8.2.2 through 3.8.2.7 inclusive, for Type II penetrations. q) The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions. A conservative value of 3.0 psig was established based on inputs to the Shearon Harris containment accident analysis. This value was selected to optimize: a) ability of safety injection systems to maintain containment within maximum allowable pressure and b) provide sufficient response time for instruments. The pressure setpoint is above the maximum expected pressure inside containment during normal operation so that inadvertent containment isolation will not occur during normal operation as a result of instrument drift, pressure fluctuations and instrument errors. The containment isolation setpoint pressure is established along with the plant Technical Specifications because of its association with other parameters. The basis for this setpoint has been established. 6.2.4.2 System Design The Containment Isolation System, in general, closes fluid penetrations that support those systems not required for emergency operation. Fluid penetrations supporting Engineered Safety Features (ESF) Systems have remote manual isolation valves which may be closed from the Control Room, if necessary. Automatic isolation valves close upon receipt of an isolation signal from a sensor. All power operated isolation valves have position indication in the Control Room. Design information regarding the containment isolation provisions for fluid system lines and fluid instrument lines penetrating the Containment is presented in Table 6.2.4-1. 6.2.4.2.1 Codes and standards The portions of the Containment Isolation System which are a part of the reactor coolant pressure boundary are designed and constructed in accordance with Quality Group A recommendations of Regulatory Guide 1.26. The remainder of the Containment Isolation System is designed and constructed in accordance with Quality Group B recommendations of Regulatory Guide 1.26. The Containment Isolation System is designed in accordance with Seismic Category I requirements as discussed in Section 3.2.1. 6.2.4.2.2 System integrity All containment isolation valves are located inside either the Containment, the Reactor Auxiliary Building, or the Fuel Handling Building. These structures are of Seismic Category I design and Amendment 63 Page 60 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 are protected against damage from missiles. The reinforced concrete containment provides a major mechanical barrier for protection against missiles which may be generated external to the Containment. Protection against damage from missiles is provided for the penetrations and associated piping, tubing, and isolation valves, actuators, and controls. Refer to Section 3.5 for a discussion of missile protection. Section 3.6 contains a discussion of protection provided against dynamic effects of pipe-whip, while Section 3.7 contains a discussion of the seismic design analysis performed on containment penetration piping. Screens are provided on the open-ended containment atmosphere purge exhaust system lines inside Containment to minimize the debris entering the lines and, in turn, entering the purge isolation valves. 6.2.4.2.3 Valve Operability Each containment isolation valve is designed to ensure its performance under all anticipated environmental conditions including maximum differential pressure, seismic occurrences, steam-laden atmosphere, high temperature, and high humidity. Section 3.11 presents a discussion of the environmental conditions, both normal and accident, for which the Containment Isolation System is designed. Dynamic analysis procedures, used in the design of Seismic Category I mechanical equipment, are discussed in Section 3.9.1. The analytic and empirical methods used for design of valves are discussed in Section 3.9.3. A discussion of the vibration operational test program to verify that the piping and piping restraints have been designed to withstand dynamic effects for valve closures is included in Section 3.9.2 A discussion of the inservice testing program for valves to assure their operability is included in Section 3.9.6. The valve types utilized for containment isolation service are designs which provide rapid closure and near zero leakage. Therefore, essentially no leakage is anticipated through the containment isolation valves when in closed position. Verification that actual leakage rates from the Containment are within design limits is provided by periodic leakage rate testing in accordance with 10 CFR 50, Appendix J as described in Section 6.2.6. Plant conditions and loads which the valves are expected to withstand are delineated in Sections 3.10 and 3.11, and will be described in the Equipment Qualification Report. 6.2.4.2.4 Isolation Barriers As stated in Section 6.2.4.1, the design of isolation valving for lines penetrating the Containment follows the intent of GDC 54 through 57, and Westinghouse Systems Standard Design Criteria Number 1.14, Rev. 2. Isolation valving for instrument lines which penetrate the Containment follows the guidance of Regulatory Guide 1.11. Those cases where literal interpretation of GDC 54 through 57 have not been followed are included in the following discussions. 6.2.4.2.4.1 General Design Criterion 54 All piping penetrations meet the intent of GDC 55, 56, or 57. In doing so, they also conform to the intent of GDC 54 to the extent that all piping systems penetrating the Containment are provided with containment isolation capabilities which reflect the importance to safety isolating Amendment 63 Page 61 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 these piping systems. In addition, Table 6.2.4-1 lists each piping penetration to be tested periodically in accordance with 10 CFR 50, Appendix J. In some penetrations, sealed closed barriers are used. Sealed closed barriers include blind flanges and locked closed isolation valves, which may be closed manual valves, closed remote-manual valves, and closed automatic valves which remain closed after a loss-of-coolant accident. Locked closed isolation valves are under administrative control to assure that they cannot be inadvertently opened. 6.2.4.2.4.2 General Design Criterion 55 Lines which are connected to the reactor coolant pressure boundary are shown in Table 6.2.4-1. Each penetration is provided with one of the following valve arrangements conforming to the requirements of 10 CFR 50, Appendix A, General Design Criterion 55, as follows: a) One locked-closed-isolation valve inside and one locked-closed-isolation valve outside Containment; or b) One automatic-isolation valve inside and one locked-closed-isolation valve outside Containment; or c) One locked-closed-isolation valve inside and one automatic-isolation valve outside Containment; a simple check valve is not used as the automatic isolation valve outside Containment; or d) One automatic-isolation valve inside and one automatic-isolation valve outside Containment; a simple check valve is not used as the automatic isolation valve outside Containment. Isolation valves are located as close to the Containment as practical and, upon loss of actuating power, solenoid and air-operated automatic-isolation valves fail closed. An exception of GDC 55 is taken for the RHR suction lines. The lines from the RCS hot legs to the RHR pump suctions each contain two remote manual (motor operated) valves, which are locked closed during normal plant power operation and are under administrative control to assure that they cannot be inadvertently opened, in accordance with SRP Section 6.2.4 Item II.f. The valves are interlocked such that they cannot be opened when the RCS pressure is greater than the design pressure of the RHR system. This valve arrangement is provided in accordance with Westinghouse Systems Standard Design Criteria, Number 1.14, Revision 2 and Appendix B of ANSI Standard N271-1976. An exception to Criterion 55 is taken for several isolation valves in lines which penetrate Containment and are required to perform safeguards functions following an accident. Lines which fall into this category include the RHR and safety injection lines, and RCP seal injection lines. Since these valves must remain open or be opened, a trip signal cannot be used. Instead, each of these motor operated valves is capable of remote manual operation. Upon completion of the safeguards function of the line, the operator can close the isolation valve from the Control Room. Leak detection capabilities for these lines is discussed in Section 5.2.5. An exception to GDC 55 is taken for the Reactor Vessel Level Instrumentation System (RVLIS) sensing lines. The six sensing lines penetrate the containment and are required to remain Amendment 63 Page 62 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 functional following a LOCA or steam break. These lines sense reactor vessel level and reactor coolant pressure, and are connected to pressure and level transmitters outside Containment. Although the RVLIS instrumentation does not prevent or mitigate the consequences of an accident, it provides an important post-accident function of providing indication of reactor vessel level and approach to inadequate core cooling. In view of this function, it is essential that the lines remain open and not be isolated following an accident. Based on this requirement, sealed sensing lines as described below are used: Each of the two sets of three sensing lines has a separate penetration, with pressure and level transmitters located immediately outside the containment wall in Seismic Category I instrument racks. The transmitters are connected to a sealed bellows located inside Containment by means of a hydraulic isolator and a sealed fluid filled tube. This arrangement provides a double barrier (one inside and one outside) between the Containment and the outside atmosphere should a leak occur outside Containment. The sealed bellows inside Containment, which is designed to withstand full reactor coolant design pressure, will prevent the escape of reactor coolant. Should a leak occur inside Containment, the diaphragm in the hydraulic isolator, which is designed to withstand full reactor coolant design pressure, will prevent any escape from Containment. This arrangement provides automatic double barrier isolation without operator action and without sacrificing any reliability with respect to its function (i.e., no valves to be inadvertently closed or to close spuriously). Both the bellows and tubing inside Containment and the transmitters and hydraulic isolators outside Containment are protected against missiles and pipe whip. Because of this sealed fluid filled system, a postulated severance of the line during either normal operation or accident conditions will not result in any release from Containment. If the fluid in the tubing is heated during the accident, the flexible bellows will allow for expansion of the fluid without overpressurizing the system. Temperature sensors have been placed in critical vertical sensing line runs to compensate for temperature induced effects on system accuracy. The RVLIS instrument lines are capillaries, not pipes, and as such are not subject to ASME code requirements. They are the same as Westinghouse has historically supplied for this application. The Westinghouse qualification groups follow the ANS definitions. The capillaries are made of Type 304 stainless steel and are procured to ASME SA-213. Although these capillaries do not fit ASME Safety Class 2 definition, they are seismically designed, and thus, it is appropriate to designate them as safety related. 6.2.4.2.4.3 General Design Criterion 56 The lines that penetrate the Containment and communicate directly with both the atmosphere inside and outside of the Containment are of two types. The first type communicates directly with the atmospheres inside and outside of Containment, i.e., the atmosphere purge line. The second type encompasses those penetrations for non-nuclear safety class lines penetrating the Containment, i.e., service air, fire protection, etc. As stated in GDC 56, two isolation valves, one inside and one outside Containment, are required in lines which penetrate the Containment and connect directly to the containment Amendment 63 Page 63 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 atmosphere. However, GDC 56 allows for alternatives to these explicit isolation requirements where the acceptable basis for each alternative is defined. The following are alternatives to explicit conformance with GDC 56. An exception is taken to Criterion 56 for the lines from the containment recirculation sumps to the suctions of the residual heat removal (RHR) pumps and containment spray pumps. Each line is provided with motor operated gate valves. These valves are enclosed in valve chambers that are leaktight at containment design pressure. Each line from the containment sump to the valve is enclosed in a separate concentric guard pipe which is also leaktight. A seal is provided so that neither the chamber nor the guard pipe is connected directly to the containment sump or to the containment atmosphere. This design arrangement is provided in accordance with Westinghouse Systems Standard Design Criteria Number 1.14, Revision 2 and Appendix B of ANSI Standard N271-1976. The vacuum relief lines to the Containment are essential for containment integrity. Isolation is provided through a power-to-open, spring-to-close butterfly valve and a check valve inside Containment. Power from divisional electrical buses is applied to the butterfly valves at all times to keep the valves closed, except when air is required to relieve a vacuum inside the Containment. The four containment vacuum relief sensing lines associated with the containment vacuum relief lines and the containment purge sensing lines associated with normal containment pressure control utilize an alternative arrangement to those described in GDC-56. These lines, although they do not prevent or mitigate the consequences of an accident, provide important functions. The vacuum relief sensing lines support the function of providing vacuum relief in the event of an inadvertent containment spray actuation while the purge sensing lines support the function of maintaining the containment pressure within design limits during normal operation. Commensurate with this function the sensing lines meet Quality Group B standards. The piping, tubing, isolation valves, actuators, and controls associated with these lines are Seismic Category I, Class 2, as applicable, and are protected against missiles and pipe whip. The lines are open to the containment atmosphere and their containment isolation arrangement is detailed on FSAR Table 6.2.4-1. A locked open manual shut-off valve in series with a manual reset type excess flow check valve is provided outside the Containment. The excess flow check valve closes on excess flow. Open/Closed Status indication is provided in the control room for the excess flow check valve. The lines utilize a Type II mechanical penetration. The Class 1E differential pressure transmitters include an isolation diaphragm which is qualified to assure post-accident operability and structural integrity. The transmitters are designated safety class 2 components as defined in ANS-51.8/ANSI N18.2a-75 and ANSI 18.2-73. The transmitters, sensing lines, and isolation valves associated with containment vacuum relief sensing are designated seismic Category 1 and are protected from missiles and pipe break effects. When relief valves are provided in fluid system penetrations as overpressure protection devices, the relief set point is greater than 1.5 times the containment design pressure. Because of the orientation required, each of these relief valves are isolation valves for the applicable penetration. The piping and valve designs are Quality Group B, Seismic Category I, and will withstand temperatures and pressures at least equal to the containment design pressure and temperature. Should the postulated loss-of-coolant accident occur, containment pressure would Amendment 63 Page 64 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 be felt on the downstream side of a relief valve inside the Containment and would act in conjunction with the spring pressure setting of the relief valve to further enhance seating. 6.2.4.2.4.4 General Design Criterion 57 Closed systems used as an isolation barrier, inside the Containment, meet the following requirements:

1. The systems are protected against postulated missiles and pipe-whip.

2. The systems are designed to Seismic Category I.

3. The systems meet Safety Class 2 standards and are inservice inspected as described in Section 6.6.

4. The systems are designed to at least the maximum temperature and pressure of the Containment.

5. The systems will be leak tested in accordance with Section 6.6.

In addition, closed systems inside Containment meet the following requirements:

1. They are designed to withstand external pressure from the Containment structural integrity test.

2. They are designed to withstand the design basis accident and accompanying environment.

3. They do not communicate with either the Reactor Coolant System or the containment atmosphere.

The steam generator shell, and all connected lines are designed as Seismic Category I, Quality Group B, and are missile protected. This design allows these components to be considered as an extension of the Containment. Isolation valves are provided outside Containment on all lines emanating from the steam generator. These valves are either normally closed or close automatically to effect steam generator isolation, except for steam supply lines to auxiliary feed pump turbine and safety valve lines which may operate intermittently. During a LOCA, the secondary side of the steam generator will be pressurized to a greater pressure than the containment atmosphere by the Auxiliary Feedwater System. This pressure within the steam generator constitutes an additional barrier to the release of the containment atmosphere. All feedwater lines including all associated branch lines are provided with positive isolation valves (gate or globe type valves) which are either automatically or remote-manually operated, and located outside the containment as close as possible to the containment. This design complies with GDC 57 criteria for containment isolation provisions. Further details are shown on Table 6.2.4-1 and Figure 10.1.0-3. There are four instrument lines which penetrate the Containment and are required to remain functional following a LOCA or steam break. These lines sense the pressure of containment atmosphere and are connected to pressure transmitters outside Containment. Signals from Amendment 63 Page 65 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 these transmitters can initiate safety injection and containment isolation on high containment pressure, HI-1. They also, upon HI-3 containment pressure, produce the signal to initiate containment spray. In view of this function, it is essential that the line remain open and not be isolated following an accident. Based on this requirement, a sealed sensing line as described below is used. Each of the four channels has a separate penetration and each pressure transmitter is located immediately outside the containment wall. The transmitter is connected to a sealed bellows located immediately adjacent to the inside containment wall by means of a sealed fluid filled tube. This arrangement provides a double barrier (one inside and one outside) between the Containment and the outside atmosphere. Should a leak occur outside Containment, the sealed bellows inside Containment, which is designed to withstand full containment design pressure, will prevent the escape of containment atmosphere. Should a leak occur inside Containment, the diaphragm in the transmitter, which is designed to withstand full containment design pressure, will prevent any escape from Containment. This arrangement provides automatic double barrier isolation without operator action and without sacrificing any reliability with regard to its safeguards functions (i.e. no valves to be inadvertently closed or to close spuriously). Both the bellows and tubing inside Containment and the transmitter and tubing outside Containment are enclosed by protective shielding. This shielding (box, channel or guard pipe, etc.) prevents mechanical damage to the components from missiles, water jets, dropped tools, etc. Because of this sealed fluid filled system, a postulated severance of the line during either normal operation or accident conditions will not result in any release from the Containment. If the fluid in the tubing is heated during the accident, the flexible bellows will allow for expansion of the fluid without overpressurizing the system and without significant detriment to the accuracy of the transmitter. The RHR, Containment Spray, and Safety Injection are closed loop systems, outside Containment. The systems are designed to Seismic Category I standards, classified as Quality Group B and C, and will maintain their integrity should the Containment experience its design temperature and pressure transient. All portions of the CSS which are subject to containment pressure meet the requirements of SRP Section 6.2.4 and ANSI-N271 to qualify as a closed system. Due to the use of eductors, the NaOH suction flow is drawn into the recirculation piping of the CSS pump and thus this portion of the system does not provide a leakage path for release of containment atmosphere. The NaOH system has been designed to Safety Class 3 criteria and is capable of withstanding a design basis earthquake. In addition, the entire CSS is subject to inservice inspection. The containment pressure instrument lines are capillaries, not pipes, and as such are not subject to ASME Code requirements. They are the same as Westinghouse has historically supplied for this application. The Westinghouse qualification groups follow the ANS definitions. The capillaries are made of SA-316 stainless steel and are procured to ASTM A-269. Although these capillaries do not fit the ASME Safety Class 2 definition, they are seismically designed, and thus it is appropriate to designate them as safety-related. Amendment 63 Page 66 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Provisions to detect possible leakage from these systems include instruments to measure flow rate, containment sump water level, temperature, pressure, and radiation level. The systems will be periodically leak tested as described in Section 6.6. 6.2.4.2.5 Valve Closure Times The containment isolation valve closure times have been selected to assure rapid isolation of the Containment following postulated accidents. A closure time of 3.5 seconds has been established for the normal containment purge make-up and exhaust lines which provide an open path from the Containment to the environment. The Pre-Entry Purge make-up and exhaust lines have a closure time of 15 seconds but are normally locked closed per NUREG-0737. Isolation valve closing times are verified during the functional performance tests prior to reactor startup. Upon receipt of the actuating signals, automatic valves will close within the times indicated in Table 6.2.4-1. The historical basis evaluation of radiological consequences of a LOCA during purge for compliance with Branch Technical Position CSB 6-4 were evaluated and are reported in Section 6.2.4.2.7. 6.2.4.2.6 Valve redundancy and actuation The Containment Isolation System is automatically actuated by signals developed by the Engineered Safety Features Actuation System, described in Section 7.3. The sequence of events and diversity in the parameters sensed which culminates in the initiation of containment isolation is discussed fully in Section 6.2.1 and 7.3. Redundancy and physical separation are provided in the electrical and mechanical design to ensure that no single failure in the Containment Isolation System prevents the system from performing its intended functions. Where a penetration is part of a redundant train in an ESF system, isolation valves for that train may receive power from a single electrical division. This is desirable so that a single failure of an electrical division cannot disable both trains of the ESF system. In these cases, a redundant mechanical barrier (that is closed systems beyond the isolation valves) exists so that containment isolation is not lost as a result of a single electrical failure. Emergency power is supplied from the diesel generators in the event of loss of offsite power as discussed in Section 8.3.1. When an automatic phase A containment isolation signal is actuated, the standby diesels are started concurrently as described in Section 7.3. The power train assignment for each isolation valve is shown on Table 6.2.4-1. Diesel generator 1 supplies power to Train A and diesel generator 2 supplies power to Train B. Automatic actuation causes required containment isolation valves to function. In addition, containment isolation valves equipped with power operators may be controlled individually by positioning hand switches in the Control Room. Also, in the case of certain valves with actuators, a manual override is installed to permit manual control of the associated valve. The override control function can be performed as described in Section 7.3. Containment isolation valves with power operators are provided with open/closed indication which is displayed in the Control Room. The valve mechanism also provides a local, mechanical indication of valve Amendment 63 Page 67 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 position. Air/spring operated isolation valves are driven to the closed position on loss of actuating power by a self-contained spring actuator. 6.2.4.2.7 Evaluation of containment purge system design Based on guidance given in Branch Technical Position CSB 6-4, the following historical basis analysis was performed to justify the containment purge system design:

1. An analysis of the radiological consequences of a loss-of-coolant accident. The analysis was performed using the following assumptions and parameters:

a. The Containment was assumed to be filled with a steam at a fission product concentration of 60 Ci/gm of I-131 equivalent.

b. The temperature and pressure inside the Containment were given in Figures 6.2.1-1 and 6.2.1-2 for the most severe hot leg break.

c. Steam was released unfiltered through the normal containment purge and purge makeup 8 in. lines prior to isolation of the Containment.

d. The purge line isolation valves were assumed to remain fully open for 2 sec.

following a containment isolation signal (CIS) and were to be fully closed with 5.5 sec. following a CIS.

e. The steam was assumed to be discharged by adiabatic flow through an abrupt inlet with a frictional resistance of 0.5 velocity heads. Conservatively, all other frictional loses were neglected. The amount of steam released to the environment was calculated by the method described on Pages 380 and 381 of chemical Engineers' Handbook, J. H. Perry, Editor, Third Edition, McGraw-Hill Book Company, Inc., 1950.

The amount of steam released was calculated to be 98 lbs. Using the atmospheric dilution factors in Table 2.3.4-5 and the dose calculation method described in Appendix 15.0A of the original version of the HNP FSAR, this release was calculated to result in offsite inhalation thyroid doses of 0.85 rem at the exclusion area boundary and 0.2 rem at the boundary of the low population zone. The eight (8) inch butterfly valves used for continuous purge were evaluated against the operability criteria set forth in BTP CSB 6-4. This was a one-time only evaluation showing overall acceptability of the containment purge system design, and has therefore not been repeated or updated for changes to dose analysis methods. When analytical methods are used in fulfillment of the provisions of the component operability assurance program, they will meet the requirements of NUREG-0737. 6.2.4.3 Design Evaluation The purpose of the Containment Isolation System is to provide a minimum of one protective barrier between the Containment and the environment. To fulfill its role as a barrier, the Containment is designed to remain intact before, during, and subsequent to any failure involving fluid systems, either inside or outside the Containment. Where fluid lines penetrate the Containment, the penetration has the same integrity as the containment structure itself. In Amendment 63 Page 68 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 addition, the fluid line isolation valves perform the containment isolation function for leakage through the fluid lines. Since a rupture of a large line connected to the Reactor Coolant System may be postulated, isolation valves for lines of this type are required to be located within the Containment. These isolation valves are required to close automatically on various indications of reactor coolant loss or high energy line break. Additional reliability is added when a second valve, located outside and as close as practical to the Containment, is included. This second valve also closes automatically. A single active failure can be accommodated since a second valve is available to perform the containment isolation function. By physically separating the two valves, there is little likelihood that a failure of one valve would cause failure of the second. Series valves of this type are provided with independent power sources. 6.2.4.4 Tests and Inspections Components of the Containment Isolation System are tested for correct functional performance during the preoperational test program described in Section 14.2. A capability is provided to operate the isolation valves in order to verify continued availability in accordance with the requirements of the inservice test program described in Section 3.9.6 and the Technical Specifications. The capability and test procedures used to verify that the leaktightness of containment isolation valves is in accordance with 10 CFR 50, Appendix J, as described in Sections 6.2.6 and the Technical Specifications. Provisions are made to allow ASME Code Class 2 and 3 inservice inspection per ASME B&PV Code Section XI as described in Section 6.6. 6.2.5 COMBUSTIBLE GAS CONTROL IN CONTAINMENT Following a beyond design-basis accident, hydrogen gas may be generated inside Containment by reactions such as Zirconium metal with water, corrosion of materials of construction, and radiolysis of aqueous solution in the core and containment sump. This subsection describes the systems that are provided in accordance with General Design Criteria 41 to control the buildup of hydrogen within the Containment. Three mechanisms for monitoring and controlling hydrogen inside the Containment are considered in the SHNPP design:

1. Containment hydrogen purge.

2. Containment hydrogen mixing and,

3. Containment hydrogen monitoring.

The design basis for each of these mechanisms is described in the following subsections.

Further information is contained in the TMI Appendix. Amendment 63 Page 69 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.5.1 Design Bases 6.2.5.1.1 Deleted By Amendment 62 6.2.5.1.2 Containment Hydrogen purge system The following design bases apply to the containment hydrogen purge system.

1. The system up to the first isolation valve outside Containment is Safety Class 2, Seismic Category I, designed to retain its integrity and operability under all conditions following a design basis loss-of-coolant accident. The remainder of the system is non-safety-related.

2. The system is designed to exhaust the air and hydrogen from the Containment and replace it with air from the outside.

3. Functional and operational redundancy of the system is not provided. The system may be used to control hydrogen inside Containment following a beyond design-basis accident.

4. Since the control power is disconnected by a remotely operated key locked switch, the air-operated containment hydrogen purge isolation valve inside containment is "sealed closed" during normal plant operation. The keylock switch will restore the power and allow the remote manual opening of the valve from the main control room. Valve status and keylock switch position indication is provided in the main control room. The outside containment isolation valves are normally locked closed and are manually operated locally.

5. All materials and equipment required by this system inside Containment are compatible with the environmental conditions anticipated during normal operation and accident conditions and are suitable for a lifetime consistent with that of the plant.

6.2.5.1.3 Hydrogen monitoring system The following design bases apply to the hydrogen monitoring system.

1. The Hydrogen Analyzer is non-safety-related but qualified using the methodology of IEEE-323-1974. It is designed to be functional, reliable, and capable of continuously measuring the concentration of hydrogen in the containment atmosphere following a significant beyond design-basis accident for combustible gas control and accident management, including emergency planning. The Hydrogen Analyzer is powered through associated circuits from a 1E source.

2. The hydrogen analyzer systems lines between and including the containment isolation valves for the sample feed header and sample return line are ASME Section III, Class 2, Seismic Cat I and are designed to retain their integrity and operability under all conditions following a design basis accident. Portions downstream of both the inboard and outboard containment isolation valves, beyond the safety-related boundary, up to and including the next analyzed point meet the requirements of RG 1.29 C.3 (Quality Class A-12)

Amendment 63 Page 70 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

3. All materials and equipment required by this system are selected to be compatible with the environmental conditions anticipated during accident operation and are suitable for a lifetime consistent with that of the plant.

4. The system samples containment air, providing the means to measure the containment air hydrogen concentration and to alert the operator in the event that a high hydrogen concentration is detected, in accordance with the requirements of Regulatory Guide 1.7.

5. Containment isolation valves for the A Train hydrogen analyzer are normally open and fail closed on loss of electrical power. The containment isolation valves for the B Train analyzer are normally shut. They will fail shut on a loss of electrical power when open.

Means are provided to reopen valves, when required, after power is restored. In the event of a containment isolation signal, valves 2SP-V301 SA-1 and 2SP-V349SA-1 close and isolate containment penetration 73B. Valves 2SP-V300 SA-1 and 2SP-V348 SA-1 close to isolate penetration 73A. On power failure, all valves fail closed, insuring isolation. The hydrogen analyzer cabinet, tag number AT-1SP-7438A, is qualified for beyond design-basis accident operation. The sample line coming from and going to penetrations 73A and 73B respectively, contain only train A associated valves. Likewise, containment penetrations 86A and 86B use only train B associated valves on the hydrogen analyzer sample lines. As a result, if one train fails then the redundancy for hydrogen sampling is still provided. If the associated valves fail to close when they should close, safety is not compromised since the hydrogen analyzer is qualified for beyond design-basis accident operation.

6. The Hydrogen Analyzer System consists of two identical units which are completely independent of each other and are powered from independent onsite sources to assure process capability is available to monitor the hydrogen concentration in the Containment.

See Table 6.2.5-7 which provides a failure modes and effects analysis.

7. The system is designed for remote-manual sampling capability with an intermittent cycle of Hydrogen indication for six (6) different sample points. The Hydrogen Analyzer will have a continuous sampling and indicating capability for a single sample point. The sample point locations are as follows:

a. Dome

b. Reactor Coolant Pump and Steam Generator 1A

c. Reactor Coolant Pump and Steam Generator 1B

d. Reactor Coolant Pump and Steam Generator 1C

e. Pressurizer

f. Area Below Flux Mapping Room Floor Amendment 63 Page 71 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 These points are located on various elevations providing a broad coverage of the Containment for monitoring of Hydrogen Concentration in a beyond design-basis accident.

8. Remote control, readout, alarm, and recording will be from the Main Control Room. An alarm will be activated for the Hydrogen Analyzer malfunction, loss of power, and high H2 Concentration.

9. The H2 Analyzers will be capable of measuring in the 0-10 percent H2 range by volume, with an accuracy of +/- 2.0 percent of full scale and a sensitivity of 0.1 percent H2 by volume.

10. Provisions will be made for Containment air grab sample via Remote Sample Dilution Panel to be diluted, cooled, and transported to the laboratory for analyses.

The Remote Sample Dilution Panel was designed in accordance with the criteria stated in Regulatory Guide 1.97 Rev. 3 and NUREG-0737, Section II.B.3 to meet the following requirements: a) To provide, with sufficient rapidity, a sample of containment atmosphere, so that analysis can be completed within 3 hours from the time of decision to take a sample, without requiring the use of an isolated auxiliary system. b) To obtain samples suitable for analysis for hydrogen, and for gamma spectrum analysis for noble gases and iodines. c) To obtain, and permit analysis of, a sample without a dose to any person exceeding the criteria of GDC-19 of Appendix A to 10 CFR 50 (i.e., 5 rem whole body, 75 rem extremities) assuming a fission product release per Regulatory Guide 1.4, Rev. 2. d) To provide samples such that background radiation will be low enough to permit sample analysis with an error of approximately a factor of two. e) To be capable of providing at least one sample per day for seven days and at least one sample per week for the duration of the accident condition. f) To give design consideration:

1) provisions for purging, reducing plateout, and preventing blockage in sample lines.

2) samples that are representative of the containment atmosphere following a transient or accident.

3) minimizing the volume of gas taken from containment and returning residues to containment.

4) providing ventilation exhaust from the panel filtered with charcoal and HEPA filters.

g) The RSDP is classified in Regulatory Guide 1.97 Rev. 3 as Category 3 which specifies "high quality commercial grade" construction "selected to withstand the specific service environment." This equipment is, therefore, classified as Non-Nuclear Safety and is Amendment 63 Page 72 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 non-seismic Category I. The valves isolating this system from the Containment Hydrogen Analyzer System are Class 1E and operated from a Class 1E power source.

11. Capability will be provided for obtaining samples under both positive and negative containment pressure condition.

12. Proper shielding and other provisions will be incorporated into the design to assure that personnel exposure does not exceed the limits of GDC 19, and that the required radiological analysis can be performed on the containment air sample.

13. A hydrogen monitoring system capable of diagnosing beyond design-basis accidents is installed at Harris Nuclear PLant (HNP). HNP committed to maintain a containment spray hydrogen monitoring system as part of the justification for the removal of the requirements for these monitors from the Technical Specifications, which was approved in License Amendment No. 131. HNP's containment hydrogen monitoring system will comply with the Category 3 criteria of Regulatory Guide 1.97, as categorized by the Commission and published in the Model Safety Evaluation for TSTF-447, Revision 1 (Federal Register 55418).

6.2.5.1.4 Containment hydrogen mixing The following design basis applies to mechanisms or systems for mixing of hydrogen bearing gases inside the reactor containment.

1. Local hydrogen concentrations inside the reactor containment shall be maintained at less than 4 percent by volume.

2. The Containment Cooling System which provides heat removal and active mixing of containment air meets the redundancy, environmental, seismic, and quality requirements described in Section 6.2.2.1.

6.2.5.2 System Design 6.2.5.2.1 Deleted By Amendment 62 6.2.5.2.2 Containment hydrogen purge system The Containment Hydrogen Purge System is provided as a means of purging the hydrogen from the Containment. The system consists of a purge make-up penetration line, an exhaust penetration line and a filtered exhaust system; it is shown on Figure 6.2.2-3. Design data for principal system components are presented in Table 6.2.5-2. The filtered exhaust system includes in the direction of air flow, a demister, electrical heating coil, a medium efficiency filter, HEPA pre-filter, a charcoal adsorber, a HEPA after-filter, a motorized isolation valve and a centrifugal fan. Amendment 63 Page 73 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The filtered system draws 100 cfm from the Containment, mixes with 400 cfm dilution air from Reactor Auxiliary Building and discharges to the vent stack. A motorized isolation valve, and a check valve are provided in the dilution air line from the Reactor Auxiliary Building. The hydrogen purge filtered exhaust unit is located in the Reactor Auxiliary Building and the hydrogen purge intake point is located inside the containment building. The intake is fastened to the inside of the Containment and routed through its containment penetration. The system is actuated by opening the inboard containment isolation valve in the exhaust line by a remote keylock manual action from the Control Room, manually opening the locked closed outboard containment isolation valves in both the exhaust and make-up lines and then starting the exhaust fan. The operator would make the decision to use the purge system based on readings from the containment hydrogen analyzers and the containment pressure indicators. The inboard isolation valve is a normally closed remote keylock manually air operated valve which is defined as a sealed closed barrier per SRP Section 6.2.4 Item II.3.f. Administrative control is provided on the outboard containment isolation valve in the form of a locked closed manual valve. Both of these valves will be used only post-beyond design-basis accidents. The only portions of the system which would be exposed to the post LOCA or a beyond design-basis accident environment in the Containment are the system isolation valves and associated piping. The isolation valves and associated piping are safety-related and seismically supported. The following items will be locked closed and will be verified that they are closed at least every 31 days as required by NUREG-0737, Item II.E.4.2: a) The local air supply isolation valves to the 42-inch pre-entry purge and makeup valves 2CP-B4SB-1 and 2CP-B8SB-1 (outside containment). b) The manual remote keylock switches for the 42-inch pre-entry purge and makeup valves 2CP-B3SA-1 and 2CP-B7SA-1 (inside containment). c) The manual operated hydrogen purge exhaust and makeup valves 2CM-B4SA-1 and 2CM-B6SA-1 (outside containment). d) The manual remote keylock switch for the hydrogen purge air operated exhaust valve 2CM-B5SA-1 (inside containment). 6.2.5.2.3 Containment Hydrogen Monitoring System The Hydrogen Monitoring System consists of containment sampling valve manifolds, containment isolation valves, Hydrogen Analyzers, remote control panel, sample dilution panel, and sample return line. The hydrogen monitor system will be placed in service upon direction by the plant Emergency Operating Procedures (EOPs) following a beyond design-basis accident upon diagnosis of inadequate core cooling and prior to venting noncondensibles from the reactor vessel head. When placed in service, the system will provide continuous indication and recording of containment hydrogen concentration. Samples will be taken from six (6) various Containment locations to monitor H2 concentration or provide a sample for laboratory analysis. The hydrogen analyzer systems lines between and including the containment isolation valves for the sample feed header and sample return line are ASME Section III, Class Amendment 63 Page 74 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 2, Seismic Cat 1 and are designed to retain their integrity and operability under all conditions following a design basis accident. Portions downstream of both the inboard and outboard containment isolation valves, beyond the safety-related boundary, up to and including the next analyzed point meet the requirements of RG 1.29 C.3 (Quality Class A-12). The sample point is selected from the Main Control Room on the Remote Control Panel automatic sequencing and manually available by opening the appropriate valve. The system has provisions for purging and for returning the residue of the sample to the containment. The H2 concentration is then measured by an in line Analyzer with the result displayed on the local panel. Hydrogen concentration is recorded and displayed on the remote control panel located within the main control room envelope. The sampling is repeated at specified intervals for each location to establish a trend in hydrogen generation and subsequent control. A high hydrogen concentration (3 volume percent) at any sample point will activate an alarm in the Main Control Room. A recorder will be provided to record the H2 concentration at each sample point. A remote sample dilution panel is provided to cool and dilute the sample as required to obtain a containment air sample for laboratory analyses. The sample dilution panel is connected to one of the hydrogen analyzers. The sample for analysis is collected via syringe and a sample septum. The sample will then be immediately injected into a preevacuated vial for transport to the laboratory. Because the sample is diluted, shielding for transporting to the laboratory and for analysis is minimized, as is exposure to personnel collecting the sample. Equipment specifically designated for hydrogen and radioisotopic analysis of radioactive samples in the laboratory will be provided. The Post-Accident Hydrogen Monitoring System is schematically shown on Figure 6.2.5 7. 6.2.5.2.4 Containment hydrogen mixing As described in Section 6.2.5.3.3, thorough mixing of hydrogen generated by metal-water reactions, radiolysis and corrosion of metals in the Containment does not rely on any active systems. Mass diffusion of hydrogen from the source of generation within the Containment Building is sufficient to ensure thorough and uniform mixing of hydrogen to ensure that local concentrations do not exceed four volume percent. The internal structures of the Containment were designed to provide vertical compartments around each of the steam generators and the reactor vessel, which project upward from the basemat. Following beyond design-basis accidents, the lower portions of the Containment will be flooded. The surface of the water is assumed to be the main source of Hydrogen Gas. The use of grating in applicable areas promotes the circulation of air. The design of the containment is such that there are no rooms where hydrogen could accumulate in concentrations in excess of four volume percent. During the accident mitigation period mass diffusion is sufficient for thorough and uniform mixing; however, the Safety Class 2, Seismic Category I Containment Cooling System, described in Section 6.2.2, provides heat removal and active containment air mixing. Natural circulation when coupled with the active mixing provided by the containment fan coolers and the containment sprays assure proper uniform mixing of hydrogen with steam and air inside the containment throughout the beyond design-basis accident. Therefore, these local concentrations will never exceed the bulk containment concentration, which remains well below the four percent limit. Amendment 63 Page 75 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.5.4 Test and Inspections 6.2.5.4.1 Deleted By Amendment 62 6.2.5.4.2 Hydrogen Purge System All safety-related equipment is qualified by the vendor to meet the codes and standards required by the system classification. Functional testing is performed after installation, but prior to plant startup to verify the system performance capability. Preoperational tests are described in Section 14.2.12.1.68. Periodic testing of the system components will be performed in accordance with manufacturer's recommendations. 6.2.5.4.3 Hydrogen Monitoring System All equipment for this system is vendor qualified to meet the codes and standards required by the system classification. Functional and preoperational testing is performed after installation and prior to plant startup to verify the system performance capability. Preoperational tests are described in Section 14.2.12.1.68. 6.2.5.5 Instrumentation Requirements 6.2.5.5.1 Deleted By Amendment 62 6.2.5.5.2 Hydrogen Purge System All instrumentation and controls for this system are located outside of the Containment in the Reactor Auxiliary Building or in the Control Room. Control switches and status indication for the fans, isolation valves, and control valves are provided in the Control Room. 6.2.5.6 Deleted By Amendment 62 6.2.6 CONTAINMENT LEAKAGE TESTING The Containment and containment penetrations are designed to permit periodic leakage rate testing in accordance with General Design Criteria (GDC) 52 and 53 and Appendix J to 10 CFR 50. Testing requirements for piping penetration isolation barriers and valves have been established by using the intent of GDC 54, as interpreted in Appendix J to 10 CFR 50. Exceptions taken to Appendix J for Type A, B, or C tests are described and justified in Subsections 6.2.6.1, 6.2.6.2, and 6.2.6.3, respectively. 6.2.6.1 Containment Integrated Leakage Rate Test (Type A Test) The design leakage rate for the Containment is 0.1 weight percent per day. The actual leakage rate is tested and verified using the methods and requirements of Appendix J to 10 CFR 50 for Type A tests. In accordance with Appendix J, a margin for possible deterioration of the Containment integrity during the service intervals between integrated leakage rate test (ILRT) is provided. The Amendment 63 Page 76 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 measured leak rate (Lam at peak test pressure) shall not exceed 0.75 of the maximum allowable value. The structural integrity test (SIT) is conducted during the same test program as the preoperational peak pressure integrated leakage rate test. The SIT is conducted in conformance with the descriptions contained in Section 3.8.1 and with the exceptions taken to Regulatory Guide 1.18 as specified in Section 1.8. After the SIT peak pressure requirements and the containment stabilization at required pressurization are completed, the initial peak calculated pressure (Pa) ILRT and SIT depressurization phase of the test are conducted. This sequence of testing is chosen to satisfy paragraph II.F of Appendix J to 10 CFR 50, which specifies that the initial ILRT shall be conducted after the Containment is completed and is ready for operation. Subsequent peak calculated pressure tests are conducted as specified in Section 6.2.6.4. Reduced pressure ILRT's (as described in paragraphs III.A.4 and III.A.5 of Appendix J to 10 CFR 50) are not performed during pre-operational testing or during periodic ILRT's. Industry experience has shown that extrapolation factors used to correlate the reduced and full pressure tests are not reliable and may be erroneous in some cases. 6.2.6.1.1 Pretest requirements The performance of Local Leak Rate Tests (LLRT) is not a prerequisite to the ILRT. If a Containment Boundary (isolation valve, airlock seal, etc.) is repaired prior to the ILRT and during the same outage as the ILRT, then the difference between the measured local leak rates before and after the repair are used to adjust the subsequent ILRT measured Type A Leakage Rate to determine the "As-Found" Leakage Rate. The calculated difference is based upon minimum pathway leakage for the affected containment barriers. Minimum pathway leakage is the smaller leakage rate of in-series barriers tested individually, one-half the leakage rate of in-series barriers tested simultaneously by pressurizing between them, and the combined leakage rate for barriers tested in parallel. The primary prerequisite for conducting an ILRT is a general inspection of the accessible interior and exterior surfaces of the containment structures and components to uncover any evidence of structural deterioration which may affect either the structural integrity or leaktightness of the Containment. If there is evidence of structural deterioration, Type A tests shall not be performed until corrective action is taken in accordance with repair procedures, nondestructive examinations, and tests as specified in the applicable code specified in 10 CFR 50.55a. 6.2.6.1.2 Valve positioning for the ILRT The containment isolation valves are positioned to their post-accident position by the normal method with no accompanying adjustments. Normal, LOCA, and ILRT positions for each isolation valve are shown on Table 6.2.4-1. 6.2.6.1.3 System preparation for Type A tests Systems are properly isolated, drained, or vented to reflect their worst potential status following a LOCA to assure that the Type A test results accurately reflect the most restricting LOCA conditions. Systems required to maintain the Unit in a cold shutdown condition are operable in Amendment 63 Page 77 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 their normal mode and are not vented or drained. However, any of these system penetrations that require Type C local leakage tests as defined in Section 6.2.4 have the results of the local leakage tests added to the result of the Type A test. Per ANSI/ANS-56.8-1994, Systems that are not vented or drained during the Type A test which could become exposed to the containment atmosphere during a LDBA shall be Type C tested and the Type C test leakage rate for the penetration path shall be added to UCL. The leakage shall be based on minimum pathway leakage and shall include instrumentation system error. Systems used during the Type A test for sensing the leakage are not lined up in the post-accident positions. Any leakage from the isolation valves in these systems is determined by local methods and the results are added to the Type A test. Systems that operate in post-accident conditions filled with fluid as defined in Section 6.2.4 need not be vented or drained for the Type A test. Systems which form closed Seismic Category I systems inside Containment (as defined by GDC 57) are not vented to the containment atmosphere. Leakage testing of instrumentation lines that penetrate Containment is done in conjunction with the Type A test. These lines will be open to the containment atmosphere. Liner plate weld leak chase channels will not be vented during the Type A test. All systems which are provided with isolation capabilities to satisfy GDC 55 or 56 are either normally open to the containment atmosphere or are vented to the containment atmosphere during the Type A tests, except those systems required to maintain the unit in a cold shutdown condition and those penetrations that are water sealed. Table 6.2.4-1 contains the applicable GDC or other defined criteria for the isolation valve arrangements provided. The electrical penetration pressurization system, supplied by dry pressurized nitrogen, serves to exclude moisture-laden air from each containment electrical penetration. During the Type A test, the nitrogen pressure in each electrical penetration will be locked in by shutting each penetration's nitrogen supply valve. Nitrogen supply to the penetration pressurization system will be isolated and the system headers vented to the outside atmosphere. During a type A test, the steam generator secondary side is to be vented outside the containment atmosphere. The systems connected to the secondary side of the steam generator are identified in Table 6.2.4-1. The service water lines to the emergency containment air coolers are neither vented or drained, as these lines are designed to GDC 57. The coolers may be required to cool the containment atmosphere during the Type A test. Pressurized gas and water systems are isolated downstream of the outside isolation valve for the system and vented outside of the Containment. This is done to preclude inleakage into the Containment and to expose the outside isolation valve to an atmospheric back pressure to obtain accurate leakage characteristics. The reactor coolant drain tank, pressurizer relief tank, and the accumulator tanks are vented to the containment atmosphere. This is done to protect the tanks from the external pressure of the test and to preclude leakage to or from the tanks to help assure the accuracy of the test results. The following systems are considered closed systems inside containment that need not be vented and drained for a Type A test: Amendment 63 Page 78 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 a) Main Feedwater System b) Auxiliary Feedwater System c) Steam Generator Blowdown System d) Safety-Related Portion of SW. System to and from emergency fan coolers AH 1 through AH-4 e) Portion of component Cooling Water System (to and from Reactor Coolant Drain Tank HX and Excess Letdown HX) f) Portion of the Steam Generator Sampling System Inside Containment Out to the Containment Isolation Valve The system design meets the following requirements of SRP 6.2.4.II.0 for a closed system inside containment: a) The system does not communicate with either the reactor coolant system or the containment atmosphere. b) The system is protected against missiles and pipe whip. c) The system is designated seismic category I. d) The system is classified Safety Class 2. e) The system is designed to withstand temperature at least equal to the containment design temperature. f) The system is designed to withstand the external pressure from the containment structural acceptance test, g) The system is designed to withstand the loss-of-coolant-accident transient and environment. 6.2.6.1.4 ILRT test method The air used to pressurize the Containment is conditioned for temperature and water vapor to prevent moisture condensation in the Containment at the test pressure. The air used to pressurize the Containment is essentially oil-free to prevent coating of the containment wall with oil or interfering with the test instrumentation. Sensing devices are located at different locations in the Containment to measure average temperature and humidity. Location of the temperature and humidity sensors are made with consideration to their respective patterns in the Containment. These patterns are employed in determination of the mean representative temperature and humidity for the absolute method of leakage rate testing. These data are periodically monitored during the test and analyzed as they are taken so that the leakage rate and its statistical significance is known as the test progresses. Amendment 63 Page 79 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The leakage rate test period extends to 24 hours of sustained internal pressure. If it is demonstrated to the satisfaction of the NRC that the leakage rate can be accurately determined during a shorter test period, the agreed upon shorter period may be used. At the conclusion of the leakage rate test, the accuracy of the Type A test is verified by either of the supplemental test methods described in ANSI/ANS 56.8-1994, Appendix C. The supplemental test bleeds from the Containment an accurately measured amount of air. The supplemental test method selected is conducted for a sufficient duration to establish accurately the change in leakage rate between Type A test and the supplemental test. The difference between the supplemental test data and the Type A test data shall agree within 0.25 La. Except as noted below, the following aspects of Type A testing follow 10 CFR 50, Appendix J guidelines are adhered to: a) Pretest requirements including a general inspection b) Conduct of tests c) Acceptance criterion d) Periodic retest schedule e) Inspection and reporting of test Corrective actions and test frequencies for Type A tests will be determined as specified in Technical Specifications. 6.2.6.2 Containment Penetration Leakage Rate Tests (Type B Tests) Each of the following containment penetrations are tested with a Type B test. a) Personnel air locks b) Emergency air locks c) Equipment hatch d) Fuel transfer tube e) Residual heat and containment spray valve chambers f) Electrical penetrations g) Refueling access sleeve (M-66) h) Refueling access sleeve (M-102) All other mechanical penetrations do not incorporate any expansion joints or resilient seals. They consist of sleeve embedded in the containment wall and welded to the liner with the process pipe passing through the sleeve and sealed by welding to the sleeve as described in Amendment 63 Page 80 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 3.8.1. These penetrations are tested by a Type C test performed on the isolation valves as described in Section 6.2.6.3. The test pressure for Type B tests is the calculated peak pressure for the containment, Pa. The combined leakage rate for all Type B and C tests will be less than 0.6 La (maximum allowable leakage rate). The individual leakage rate testing performed and the acceptance criteria on the personnel air lock and the emergency air lock is as described in Technical Specifications. The test equipment utilized to perform the Type B tests is the same equipment used for Type C tests. The test equipment is described in Section 6.2.6.3. The test procedure is the same as the one used for Type C tests. Type B tests are performed in accordance with Appendix J to 10 CFR 50, with the following addition and exception: a) An additional test method may be used. This method measures the air flow rate to maintain the test volume at a constant pressure. b) Air locks subject to Type B testing, in accordance with Section III.B.1, as required by Section III.D.2, may use the method for testable seals described in Section III.D.2(b)(iii) to fulfill the Type B test requirement, subject to all time interval restrictions contained therein. c) Periodic leakage testing of containment penetrations (except air locks) need not be done during a refueling outage, but may be scheduled at any time during an operating cycle. However, the test interval for any penetration shall not exceed 2 years. 6.2.6.3 Containment Isolation Valve Leakage Rate Tests (Type C Tests) Table 6.2.4-1 lists all valves which are associated with the penetrating piping systems. Table 6.2.4-1 also indicates for all valves listed which are considered to be containment isolation valves and which of the containment isolation valves are to be subjected to Type C test. The valves associated with Penetrations 15 and 16 are not Type C leak tested. The lines passing through Penetrations 15 and 16 terminate at the suction to the RHR pumps. At this elevation (RHR pump suction), the Refueling Water Storage Tank (RWST) provides a water seal during the Engineered Safety Features (ESF) injection phase and the containment sumps provide a water seal during the ESF recirculation phase. A single active failure of any valve or pump in the RHR system will not affect the existence of this water seal. The RHR system is a closed system outside containment in accordance with FSAR Section 6.2.4.2.4.4. The valves associated with Penetrations 47, 48, 49, and 50 are not Type C leak tested. The lines passing through Penetrations 47, 48 (residual heat removal (RHR)/low pressure safety injection (LPSI) recirculation) 49, and 50 (containment spray recirculation) are connected to the containment sump. During and after a LOCA, the sump will provide a water seal to the associated isolation valves. A single active failure of any component will not affect the existence of this water seal nor will activation of the recirculation modes in either system. The RWST also provides a water seal to the piping system outside containment for each of these penetrations during the ESF injection phase. This seal is applied directly to the containment Amendment 63 Page 81 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 isolation valve on Penetrations 49 and 50. It is applied to the valves immediately downstream of the containment isolation valve on Penetrations 47 and 48. The containment isolation valves for the high head safety injection lines (Penetrations 17, 20, 21, and 22), the low head safety injection penetrations (13, 14, and 18), and the RCP seal injection penetrations (9, 10, and 11) are not Type C leak tested. These penetrations are provided with a pressurized water seal at a pressure greater than 1.10 Pa for a minimum of 30 days following an accident. This water seal is provided by the ECCS LHSI pumps via the piping to these penetrations and with the post-accident lineup specified in FSAR Section 6.3.1. The water supply to these penetrations is virtually unlimited because the LHSI pumps are supplied initially from the RWST and then from the containment recirculation sumps after transfer to the recirculation mode. No single active failure can prevent penetration pressurization via this pressurized water seal. The containment isolation valves on the LHSI and HHSI injection lines are gate valves with a single piece wedge. Upon closure and pressurization, the wedge seals the downstream seat (toward containment). The upstream seat is not seated and this allows the packing and body/bonnet gasket to be pressurized above 1.10 Pa. Thus, the containment atmosphere does not enter the valves nor is it released to the outside environment through the packing or gasket. Service Water to and from the fan coolers is a closed ASME Class 2 system inside containment in accordance with FSAR paragraph 6.2.4.2.4.4. No single active failure of any component could provide a potential leakage path for post-LOCA containment atmosphere. This system is described in FSAR Section 9.2.1. General Design Criterion 57 is applicable to this system as discussed in FSAR Section 6.2.4. The containment pressure transmitters are designed to meet the requirements of Regulatory Guide 1.11 and are described in Section 7.3. These lines have no isolation valves and rely on closed systems both inside and outside of the Containment to preclude the release of the containment atmosphere. The integrity of these closed systems is verified during the periodic Type A tests. These lines penetrate the Containment at penetrations 69, 70, 71, and 72. The Containment Isolation Valves on the Component Cooling Water lines which provide cooling water to the Reactor Coolant Drain Tank and CVCS Excess Letdown Heat Exchangers via penetrations M-37 and M-38 are not leak tested. The components inside Containment provide a closed system designed in accordance with General Design Criteria 57. Two simultaneous passive failures of Class 2, or better, systems is not considered a credible event. A LOCA is a passive failure of the Reactor Coolant System, and therefore, it is not credible to assume a simultaneous passive failure of the CCW closed system. Therefore, the closed system inside Containment is sufficient to insure that the containment atmosphere is not released to the environment following an accident. As noted in Subsection 6.2.4.2.4.4, all portions of the secondary side of the steam generators are considered an extension of the Containment. These systems penetrate the containment shell at penetrations numbered 1 through 6, 51 through 56, and 108 through 110. These systems provide a closed system designed in accordance with General Design Criteria 57. They are not leak tested because the closed system inside Containment is sufficient to insure that the containment atmosphere is not released to the environment following an accident. Amendment 63 Page 82 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The test, vent and drain (TVD) connections that are used to facilitate local leak testing are under administrative control, and subject to periodic surveillance, to assure their integrity and verify the effectiveness of administrative controls. These procedures meet the requirements of SRP Section 6.2.6 Item II. The test equipment to be used during the Type C tests consists of a connection to an air supply source, pressure regulator, pressure gauge, flow indicator, and associated valving or equivalent test setup. Isolation valves are positioned to their post-accident position by the normal method with no accompanying adjustments. Fluid systems are properly drained and vented with the valves aligned to provide a test volume and atmospheric air back pressure on the isolation valve(s) being tested. The test volume is pressurized to the test pressure Pa. The pressure regulator(s) maintain the test volume at a minimum of the calculated peak pressure for the containment, Pa. The air flow rate into the test volume is recorded, as is the pressure reading, at the intervals specified on the data form. These records are utilized to determine the leakage rate in cubic centimeters per minute. For larger test volumes, a pressure decay method may be utilized to determine the leakage rate. The total leakage rate for Type B and C tests will be less than 0.6 La. The individual testing performed on valves requiring a Type C test is described in Technical Specifications. In accordance with 10 CFR 50 Appendix J III.C.1, valves may be tested in the non-accident pressure direction when it can be determined that the results from the tests for the pressure applied in the non-accident direction will provide equivalent or more conservative results. The packing leakage for any valve tested in the non-accident direction shall be included in the reported leak rate for that valve if the packing provides a leakage path from the containment atmosphere to the outside environment (i.e., packing is part of containment isolation boundary). The criteria for determining the direction in which the test pressure is applied to the isolation valves is as follows: a) Check, ball, plug, and non-wedge disc gate valves are tested in the accident pressure direction. b) Wedge disc gate, butterfly, and diaphragm valves are tested in either direction since seat leakage is the same in either direction. c) Globe valves may be tested in the non-accident pressure direction if the test pressure would tend to unseat the valve and the accident pressure would tend to seat the valve. 6.2.6.4 Scheduling and Reporting of Periodic Tests Types A, B, and C tests will be conducted at the intervals specified in Technical Specifications. These intervals are in accordance with Appendix J to 10 CFR 50, with the exception of the testing of the air locks as described in Section 6.2.6.2. Amendment 63 Page 83 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Periodic leak testing of the containment isolation valves need not be done during a refueling outage but may be scheduled at any time during an operating cycle. However, the test interval for any valve shall not exceed two years. The test results will be the subject of a summary report filed on site approximately three months after each Type A test. The preoperational test report will contain a schematic of the leak measuring system, instrumentation used, supplemental test method, test program, and analysis and interpretation of the leakage test data for the Type A test.

REFERENCES:

SECTION 6.2 6.2.1-1 Waterford Steam Electric Station Unit 3 PSAR, Docket 5382, letter LPL 2656. 6.2.1-2 Whent, L. L., et.al, CONTEMPT-LT-A Computer Code for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident, Acrojet Nuclear Company, ANCR-1219, June, 1975 6.2.1-3 Shepard, R. M., Massie, H. W., Mark R. H. and Docherty, P. J., "Westinghouse Mass and Energy Release Data for Containment Design," WCAP 8264 P A (Proprietary) and WCAP 8312 A (Non-Proprietary), Revision 2, August, 1975. 6.2.1-4 RELAP-4/Mode 6 - A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems - Users Manual, EG&G Idaho, Inc., CDAP TR003, January, 1978. 6.2.1-5 Crane Technical Paper No. 410 - Flow of Fluids Through Valves, Fittings and Pipe, 1978. 6.2.1-6 Idel Chik, I.E. Handbook of Hydraulic Resistance and of Friction, Translation, Israel Program for Scientific Translations, Jerusalem, 1966 AEC-TR-6630. 6.2.1-7 Westinghouse ECCS Evaluation Model, February 1978 Version WCAP-9220, February 1978 6.2.1-8 F. M. Bordelon, et al., SATAN-VI Program: Comprehensive Space Time Dependent Analysis of Loss-of-Coolant, WCAP-6174, June 1974. 6.2.1-9 G. Collier, et al, Calculation Model for Core Reflooding After a Loss-of-coolant accident (WREFLOOD Code), WCAP-8170, June 1974. 6.2.1-10 WCAP -10325-P-A, (Proprietary), WCAP-10326-A (Nonproprietary), "Westinghouse LOCA Mass & Energy Release Model for Containment Design-March 1979 Version," May 1983. 6.2.1-11 Deleted By Amendment No. 51 6.2.1-12 "Dynamic Analysis of RPV for Postulated Loss-of-coolant accidents," WCAP-9192. Amendment 63 Page 84 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2.1-13 Shapiro, A.H., "The Dynamics and Thermodynamics of Compressible Fluid Flow, Volume 1, p. 85. 6.2.1-14 1967 ASME Steam Tables, p. 301. 6.2.1-15 NRC Docket 50-400, Letter from George W. Knighton of NRC to Mr. E. E. Utley of CP&L, "Request for Exemption from a Portion of General Design Criteria 4 of Appendix A to 10 CFR 50 Regarding the Need to Analyze Large Primary Loop Pipe Ruptures as Structural Design Basis for Shearon Harris Nuclear Power Plant, Unit 1", June 5, 1985. 6.2.1-16 Deleted By Amendment No. 51 6.2.1-17 Land, R.E., "Mass and Energy Releases Following a Steam Line Rupture," WCAP-8822 (Proprietary) and WCAP-8860 (Nonproprietary), September 1976; Osborne, M.P. and Love, D.S., "Supplement 1 - Calculations of Steam Superheat in Mass/Energy Releases Following a Steam Line Rupture," WCAP-8822-S1-P-A (Proprietary) and WCAP-8860-S1-A (Non-proprietary), September 1986; Butler, J.C. and Linn, P.A., "Supplement 2 - Impact of Steam Superheat in Mass/Energy Releases Following a Steam Line Rupture for Dry and Subatmospheric Containment Designs," WCAP-8822-S2-P-A (Proprietary) and WCAP-8860-S2-A (Nonproprietary), September 1986. 6.2.1-18 NUREG-1038, "Safety Evaluation Report Related to the Operation of Shearon Harris Nuclear Plant, Units 1 and 2," Section 6.2.1.3, November 1983. 6.2.1-19 Docket No. 50-315, "Amendment No. 126, Facility Operating License No. DPR-58 (TAC No. 7106), for D.C. Cook Nuclear Plant Unit 1," June 9. 1989 6.2.1-20 EPRI 294-2, "Mixing of Emergency Core Cooling Water with Steam; 1/3-Scale Test and Summary," (WCAP-8423), Final Report, June 1975. 6.2.1-21 ANSI/ANS-5.1 1979, "American National Standard for Decay Heat Power in Light Water Reactors," August 1979. 6.2.1-22 Westinghouse letter, CQL-01-130, dated November 21, 2001, "Double Ended Pump Suction LOCA with Revised Maximum ECCS Recirculation Flow. 6.2.1-23 GOTHIC Thermal Hydraulic Package, Version 8.2 (QA), Electric Power Research Institute, October 2016. 6.2.2-1 Spray Nozzle Performance Test Results - SPRACO Model 1713A Nozzle, by Spray Engineering Company, dated January 1, 1973. Amendment 63 Page 85 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 APPENDIX 6.2A GOTHIC COMPUTER CODE The containment pressure and temperature transient analyses are performed with the GOTHIC computer code. In this computer code, the containment volume is divided into two regions, the atmosphere region (water vapor and air mixture) and the sump region (liquid water). Each region is assumed to be completely mixed and in thermal equilibrium. The temperature of each region may be different. Mass and energy additions are made to the appropriate region to simulate the mass and energy release from the Reactor Coolant System or Secondary System during and after blowdown with the contribution of the Safety Injection System (SIS) and Containment Spray System (CSS) water, and decay energy from the core. Account is taken of boiling in the liquid region and condensing in the vapor region, and mass and energy transfers between regions are considered. The model represents the heat conducting and absorbing materials in the Containment by dividing them into segments with appropriate heat transfer coefficients and heat capacities. Thermal behavior is described by the one-dimensional, multi-region, transient heat conduction equation. The heat conducting segments are used to describe materials and surfaces in the Containment which act as heat sinks. The model mechanistically simulates cooling of the containment atmosphere by fan coolers and/or by water sprays, and cooling of the containment sump water being recirculated to the SIS by the RHR heat exchangers. The GOTHIC computer code has been qualified for the design of the containment liner and concrete structure by simulated design basis accident tests such as the Carolinas Virginia Tube Reactor (CVTR) (see References 6.2A-1 through 6.2A-3) blowdown experiment. Calculations are begun by computing initial steady state containment atmosphere conditions. Subsequent calculations are performed at incremental time steps. Following the pipe rupture, the mass and energy addition to the atmosphere or liquid region is determined for each time interval. Heat losses or gains due to heat-conducting segments are calculated. Then the mass and energy balance equations are solved to determine containment pressure, temperature of the liquid and vapor region, and heat and mass transfer between regions. The Direct heat transfer option with the DLM (Diffusion Later Model) condensation option is used for all containment passive heat sinks except the sump floor. With the Direct option, all condensate goes directly to the liquid pool at the bottom of the volume. The effects of the condensate film on the heat and mass transfer are incorporated in the formulation of the DLM option. Under the DLM option, the condensation rate is calculated using a heat and mass transfer analogy to account for the presence of noncondensing gases. It compares well with Nusselt's theory for the condensation of pure steam where the rate is controlled by the heat transfer through the condensate film. The DLM option generally under predicts the condensation rate.

REFERENCES:

APPENDIX 6.2A 6.2A 1 Norberg, J. A., Bingham, G. E., Schmitt, R. C., Waddoups, D. A., Simulated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment Preliminary Results, Idaho Nuclear Corporation, IN 1325, October, 1969. Amendment 63 Page 86 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.2A 2 Schmitt, R. C., Bingham, G. E., Norberg, J. A., Simulated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment Final Report, Idaho Nuclear Corporation, IN-1403, December, 1970. 6.2A 3 Krotiuk, W. J., Rubin, M. B., "Condensing Heat Transfer Following a LOCA", Nuclear Technology, February, 1978. 6.2A 4 through 6.2A 12 deleted by Amendment No. 63. 6.3 EMERGENCY CORE COOLING SYSTEM 6.3.1 DESIGN BASES The Emergency Core Cooling System (ECCS) is designed to cool the reactor core and provide shutdown capability following initiation of the following accident conditions: a) Loss-of-coolant accident (LOCA) including a pipe break or a spurious relief or safety valve opening in the Reactor Coolant System (RCS) which would result in a discharge larger than that which could be made up by the normal makeup system. b) Rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident. c) Steam or feedwater system break accident including a pipe break or a spurious power operated relief or safety valve opening in the secondary steam system which would result in an uncontrolled steam release or a loss of feedwater. d) A steam generator tube failure. The primary function of the ECCS is to remove the stored and fission product decay heat from the reactor core during accident conditions. The ECCS provides shutdown capability for the accidents above by means of boron injection. The system is designed to tolerate a single active failure (injection phase), or a single active or passive failure (recirculation phase). Table 6.3.1-1 provides a failure modes and effects analysis which demonstrates the capability of the ECCS to perform following a single active failure. This analysis also shows that single failures occurring during ECCS operation do not compromise the ability to prevent or mitigate accidents. The capabilities are accomplished by a combination of suitable redundancy, instrumentation for indication and/or alarm of abnormal conditions, and relief valves to protect piping and components against malfunctions. The ECCS can meet its minimum required performance level with onsite or offsite electrical power. The ECCS consists of the centrifugal charging pumps, residual heat removal pumps, accumulators, a boron injection tank, residual heat removal heat exchangers, a refueling water storage tank, along with associated piping, valves, instrumentation and other related equipment. See Section 1.3.1 for comparison of the SHNPP ECCS with similar facility designs. The design bases for selecting the functional requirements of the ECCS are derived from data which is consistent with 10CFR50 Appendix K limits following any of the above accidents as Amendment 63 Page 87 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 delineated in 10CFR50.46. The subsystem functional parameters are integrated such that the Appendix K requirements are met over the range of anticipated accidents and single failure assumptions. Redundant components are provided where the loss of one component would impair reliability. Valves are provided in series where isolation is desired and in parallel when redundant flow paths are to be established for ECCS performance. Redundant sources of the ECCS actuation signal are available so that the proper and timely operation of the ECCS will not be inhibited. Sufficient instrumentation is available so that a failure of an instrument will not impair readiness of the system. The active components of the ECCS are powered from separate buses which are energized from offsite power supplies. In addition, redundant sources of auxiliary onsite power are available through the use of the standby diesel generators to assure adequate power for all ECCS requirements. Each diesel generator is capable of driving all pumps, valves and necessary instruments associated with one train of the ECCS. Spurious movement of a motor operated valve due to the actuation of its positioning device coincident with a loss-of-coolant has been analyzed and found not to be credible for consideration in design. Since there are two valves in each RHR sump line, the opening of one of them would have no impact. There is not a credible failure that would open both valves at once. In compliance with BTP ICSB-18(PSB), power is locked out of the ECCS valves noted below. Further information regarding this BTP and its application to other manually operated valves may be found in Section 8.3.1.2.38. Information regarding valve monitoring may be found in Sections 6.3.5.5 and 7.5.1.10. Valve Position During Normal Operation 8808 A, B, and C Open 8886 Closed 8885 Closed 8884 Closed 8889 Closed 8888 A and B Open To prevent their spurious operation in the event of certain postulated fires, power is normally removed from Charging Pump Suction and Discharge Header Crossover valves 8130 A/B, 8131 A/B, 8132 A/B and 8133 A/B. Power to these manually-controlled, electrically-operated valves has been interrupted by locking the valve motor operator supply breaker in the OFF position at Motor Control Centers 1A35-SA and 1B35-SB. The intent is that the operator must restore power to the valves to enable closure as required, thus preventing spurious operation. These valves are provided with diverse valve position monitoring capability in the control room. A 125 VDC powered valve position limit switch provides out-of-position annunciation on ALB-3 for each valve. In addition, a separate valve position limit switch provides a white monitor light for each valve that indicates when the valve is not open. For a description of these monitor lights, refer to Section 7.5.1.10.3. Inadvertent opening of motor operated valving to the containment spray pumps could not allow the draining of the RWST into the containment because whenever a signal is generated, either automatically or manually, to open the sump valves a close signal is simultaneously sent to Amendment 63 Page 88 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Valves 2CT-V2SA and 2CT-V3SB (see FSAR Figure 6.2.2-1) which isolate the RWST. Also, Class 1E level monitors in the sump provide level indication and alarm in the Control Room such that the postulated event would be known to the operators at an early stage. The spurious opening of the containment sump motor operated valving to the containment spray pumps is not considered possible due to the design of the control system servicing the operators. The valves in question, 2CT-V6SA and 2CT-V7SB are automatically opened on a two out of four "low-low" signal in the RWST provided that the Containment Spray pumps are running. Remote manual opening of the sump valves from the Control Room is possible at any time regardless of whether the CS pumps are running or not, but the operator must refer to sump level indication beforehand. For these valves to open, power must be applied to the MCC output relays servicing the motor operators. There is no way for this to occur other than by the normal means. It is not possible for a loss of power to the logic cabinets associated with the RWST level detectors to cause a false two out of four signal to be generated to open the valves because the output bistables and output relays must be energized to operate. All the equipment associated with the RWST and Containment Spray System is designed to safety grade standards as described in Chapters 3 and 6 of the FSAR to assure maximum reliability. The elevated temperature of the sump solution during recirculation is well within the design temperature of all ECCS components. In addition, consideration has been given to the potential for corrosion of various types of metals exposed to the fluid conditions prevalent immediately after the accident or during long-term recirculation operations. ECCS equipment, which is located inside the Containment and which is required to operate following a LOCA, is environmentally qualified as discussed in Section 3.11. To prevent hot leg injection during the ECCS cold leg injection phase as well as SI initiation following Safeguards Actuation, power lockout in accordance with Branch Technical Position ICSB-18 will ensure that motor-operated hot leg recirculation isolation valves 8886 and 8884 are, or will remain, in the correct position, which is closed; and will not be subject to inadvertent operator mispositioning. Further information regarding compliance with BTP ICSB-18 may be found in Section 8.3.1.2.38. Additionally, these valves have monitor panel position indication and alarms should they be mispositioned during normal operation. The monitor panel position indication is in addition to the normal red-green position indication, and is provided via stem mounted limit switches which are independent of the normal limitorque position switches. There are no instruments, valves, or valve motors required for ECCS/RHR operation which will be flooded following a postulated LOCA. The following instruments, which have been provided for operator information only (per Regulatory Guide 1.97), will be below flood level and have been designed for submerged conditions. TE-7133ASA Containment sump water temperature TE-7133BSB LE-7160ASA Containment sump water level LE-7160BSB Amendment 63 Page 89 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The RHR pumps and the safety injection system piping provide a pressurized water seal to containment penetrations M-9, M-10, M-11, M-13, M-14, M-17, M-18, M-20, M-21, and M-22 for a minimum period of 30 days following a design basis accident. This seal is maintained following any single active failure. This water seal ensures that the containment atmosphere cannot leak to the environment following a design basis accident (see Section 6.2.6). The requirement to maintain this seal imposes the following restrictions on valve positions during the specified period. a) The charging pump suction header crossover valves must remain open during a post-accident injection and recirculation modes. An additional benefit of these valves being open is that a failure of an RHR pump in the recirculation modes will not result in loss of a charging pump because one RHR pump can provide sufficient flow and NPSH for two charging pumps. b) At least one of the boron injection tank inlet isolation valves must remain open during the post-accident injection and recirculation modes. c) The RHR system crossover valves at the connection to the line supplying flow to the RCS hot legs for hot leg recirculation must remain open during the post-accident injection and recirculation modes. d) A motor-operated Containment Isolation Valve on one of the low head flow paths to the RCS cold legs must be closed during the post-accident cold leg recirculation mode to prevent RHR pump runout should a single active failure of an RHR pump occur. 6.3.2 SYSTEM DESIGN The ECCS is designed to tolerate a single active failure (injection phase) or a single active or passive failure (recirculation phase). The redundant onsite standby diesel generators assure adequate emergency power to at least one train of electrically operated components in the event that a loss of offsite power occurs simultaneously with a LOCA. 6.3.2.1 Schematic Piping and Instrumentation Diagrams Flow diagrams of the ECCS are shown in Figures 6.3.2-1 through 6.3.2-3. Pertinent design and operating parameters for the components of the ECCS are given in Table 6.3.2-1. The codes and standards to which the individual components of the ECCS are designed are listed in Table 3.2.1-1. The component interlocks used in different modes of system operation are listed below: a) The safety injection signal, "S", is interlocked with the following components and initiates the indicated action:

1) Centrifugal charging pumps start on "S" signal.

2) Refueling water storage tank suction valves to charging pumps open on "S" signal.

3) Boron injection tank discharge parallel isolation valves open on "S" signal.

Amendment 63 Page 90 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

4) Normal charging path valves close on "S" signal.

5) Normal charging pump miniflow valves close on "S" signal.

6) Alternate miniflow path valves open/close on "S" signal coincident with RCS pressure.

7) Residual heat removal pumps start on "S" signal.

8) Any closed accumulator isolation valves open on "S" signal.

9) Volume control tank outlet isolation valves close on "S" signal.

b) Switchover from injection mode to recirculation involves the following interlocks:

1) The suction valves from the containment sump open when two out of four low level transmitters indicate a low-low level in the RWST in conjunction with an "S" signal.

2) The charging pump suction (recirculation line) isolation valve from the RHR pump discharge line cannot be opened unless one of the RHR suction isolation valves from the RCS hot legs is closed.

3) The recirculation flow paths from the RHR pumps discharge to the charging pump suction is interlocked such that the isolation valves in these lines cannot be opened unless the alternate miniflow paths are isolated.

6.3.2.2 Equipment and Component Descriptions The component design and operating conditions are specified as the most severe conditions to which each respective component is exposed during either normal plant operation or during operation of the ECCS. For each component, these conditions are considered in relation to the code to which it is designed. By designing the components in accordance with applicable codes, and with due consideration for the design and operating conditions, the fundamental assurance of structural integrity of the ECCS components is maintained. Components of the ECCS are designed to withstand the appropriate seismic loadings in accordance with their safety class as given in Table 3.2.1-1. The discussion of each major mechanical component of the ECCS follows below: 6.3.2.2.1 Accumulators The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. During normal operation each accumulator is isolated from the RCS by two check valves in series. Should the reactor coolant system pressure fall below the accumulator pressure, the check valves open and borated water is forced into the RCS. One accumulator is attached to each of the cold legs of the RCS. Mechanical operation of the swing disc check valves is the only action required to open the injection path from the accumulators to the core via the cold leg. Connections are provided for remotely adjusting the level and boron concentration of the borated water in each accumulator during normal plant operation as required. Accumulator Amendment 63 Page 91 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 water level may be adjusted either by draining to the refueling water storage tank or by pumping borated water from the refueling water storage tank to the accumulator. Samples of the solution in the accumulators are taken periodically for checks of boron concentration. Accumulator pressure is provided by a supply of nitrogen gas, and can be adjusted as required during normal plant operation; however, the accumulators are normally isolated from this nitrogen supply. Gas relief valves on the accumulators protect them from pressures in excess of design pressure. The accumulators are located within the Containment, but outside of the secondary shield wall which protects them from missiles. Accumulator gas pressure is monitored by indicators and alarms. The operator can take action as required to maintain plant operation within the requirements of the Technical Specification covering accumulator operability. 6.3.2.2.2 Boron Injection Tank The boron injection tank is connected to the discharge of the centrifugal charging pumps. Upon actuation of the safety injection signal, the charging pumps deliver boric acid solution from the refueling water storage tank into the RCS by way of the boron injection tank. In the original design of the ECCS, the boron injection tank contained a high concentration boric acid solution (12% wt.). This highly concentrated boric acid solution was determined to be unnecessary and has been eliminated. The boron injection tank has been left in place but it serves no function other than being part of the safety injection flow path. In the steam line break accident analysis (Section 15.1.5) the system was analyzed assuming the BIT boron concentration was 0 ppmB. This assumption provides the most limiting case for this analysis; however, the BIT may contain a boron concentration within the range of 0 - 2600 ppmB. 6.3.2.2.3 Deleted by Amendment No. 27 6.3.2.2.4 Residual heat removal pumps In the event of a LOCA the residual heat removal pumps are started automatically on receipt of an "S" signal. The residual heat removal pumps take suction from the refueling water storage tank during the injection phase and from the containment sump during the recirculation phase. Each residual heat removal pump is a single stage vertical position centrifugal pump. A minimum flow bypass line is provided for the pumps to recirculate and return the pump discharge fluid to the pump suction should these pumps be started with the reactor coolant system pressure above their shutoff head. Once flow satisfies an established setpoint, the bypass line is automatically closed. This line prevents deadheading of the pumps and permits pump testing during normal operation. The safety intent of Regulatory Guide 1.1 is met by the design of the ECCS such that adequate net positive suction head is provided to system pumps. The most limiting condition with respect to net positive suction head exists when the residual heat removal pumps are switched to the Amendment 63 Page 92 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 recirculation mode of operation. In addition to considering the static head and suction line pressure drop, the calculation of available net positive suction head in the recirculation mode assumes that the vapor pressure of the liquid in the sump is equal to the containment ambient pressure. This ensures that the actual available net positive suction head is always greater than the calculated net positive suction head. Available and required net positive suction head for the residual heat removal pumps is indicated on Table 6.3.2-1. ECCS pump specifications include a specified maximum required NPSH which the pump is required to meet. Pump vendors have verified that the required NPSH for the Shearon Harris pumps is less than the maximum required NPSH through testing in accordance with the criteria established by the Hydraulic Institute Standards. Ample experience with the same vendors and similar ECCS pumps has shown the variability in their NPSH requirements to be minimal. Pumps are deemed acceptable based on their vendor certified NPSH requirements being less than the maximum allowable specified by the ECCS designers. Although one specific pump may vary slightly from the certified curve, the curve is representative of all the pumps supplied and is always lower than the maximum available specified by the system designers. Furthermore, this number specified to the vendor is conservative compared to the ECCS layout criteria. The vendor supplied curve, which is used to confirm that the actual system piping provides adequate NPSH, is derived from repeated testing of the same type of pump. In addition to random testing to demonstrate that variation in pump performance is insignificant, each impeller casting is inspected to ensure that dissimilarity from one pump to the next is minimized. For the RHR pump NPSH calculation, when taking suction from the containment sump, in equilibrium with containment ambient pressure (i.e., no credit is taken for subcooling of the sump fluid), the equation is: NPSHavailable = hstatic head - hline losses For other system pumps, or for RHR pump NPSH when operating in other modes, this equation becomes: NPSHavailable = hambient pressure + hstatic head -hline losses - hvapor pressure The net positive suction head of the residual heat removal pumps is evaluated for normal plant cooldown operation, and for both the injection and recirculation modes of operation for the design basis accident. Recirculation operation gives the limiting net positive suction head requirement and the net positive suction head available is determined from the containment water level relative to the pump elevation and the pressure drop in the suction piping from the sump to the pumps. Positive net positive suction head margin is maintained with a postulated debris bed on the recirculation sump screens. The residual heat removal pumps are discussed further in Section 5.4.7. A pump design performance curve is given in Figure 6.3.2-8. 6.3.2.2.5 Centrifugal charging pumps In the event of an accident, the charging pumps are started automatically on receipt of an "S" signal and are automatically aligned to take suction from the refueling water storage tank during Amendment 63 Page 93 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 injection. During recirculation, suction is provided from the residual heat removal pump discharge. These high head pumps deliver flow through the boron injection tank to the RCS at the prevailing reactor coolant system pressure. Each centrifugal charging pump is a multistage diffuser design, barrel-type casing with vertical suction and discharge nozzles. A minimum flow bypass line is provided on each pump discharge to recirculate flow to the pump suction after cooling via the seal water heat exchanger during normal plant operation. Each charging pump has double valve isolation on the minimum flow bypass line. Safety injection signal closes the valves to isolate the normal charging line and volume control tank and opens the charging pump-refueling water storage tank suction valves to align the high head portion of the ECCS for injection. The charging pumps may be tested during power operation via the minimum flow bypass line or the normal charging line. The two operable charging pumps are each provided with an alternate miniflow path which is automatically aligned on receipt of an "S" signal plus an RCS pressure permissive signal. Simultaneously, the normal miniflow paths are isolated. Each of the alternate miniflow motor operated isolation valves is powered from the same train as the pump it is protecting. This control logic is intended to ensure maximum safety injection flow while providing pump protection against a dead head condition. The orifice in the alternate miniflow path prevents the charging pump from reaching a runout condition. Net positive suction head design considerations for the charging pumps are similar to those for the RHR pumps discussed in Section 6.3.2.2.4. The net positive suction head for the centrifugal charging pumps is evaluated for both the injection and recirculation modes of operation for the design basis accident. The end of the injection mode of operation gives the limiting net positive suction head available (minimum static head). The net positive suction head available is determined from the elevation head and vapor pressure of the water in the refueling water storage tank, which is at atmospheric pressure, and the pressure drop in the suction piping from the tank to the pumps. A pump design performance curve for the centrifugal charging pumps is presented in Figure 6.3.2-9. 6.3.2.2.6 Positive displacement hydrostatic test pump The positive displacement hydrostatic test pump is provided to accomplish two non-safety related functions. It is designed primarily for use in hydrotesting the RCS. This pump is also used to initially fill and maintain level in the three SIS accumulators. The suction of this pump is permanently connected to a branch line from the RWST discharge header and its discharge is permanently connected to the accumulator fill line header. The discharge pressure of the pump is regulated by an air operated control valve located in a return line to the pump suction. 6.3.2.2.7 Deleted by Amendment No. 27 6.3.2.2.8 Residual heat exchangers The residual heat exchangers are conventional shell and U-tube type units. During normal cooldown operation, the residual heat removal pumps recirculate reactor coolant through the Amendment 63 Page 94 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 tube side while component cooling water flows through the shell side. During emergency core cooling recirculation operation, water from the containment sump flows through the tube side. The tubes are seal welded to the tube sheet. A further discussion of the residual heat exchangers is found in Section 5.4.7. 6.3.2.2.9 Valves Design parameters for all types of valves used in the ECCS are given in Table 6.3.2 1. Design features employed to minimize valve leakage include:

1) Where possible, packless valves are used.

2) Other valves which are normally open, except check valves and those which perform a control function, are provided with backseats to limit stem leakage.

3) Normally closed globe valves are installed with recirculation fluid pressure under the seat to prevent stem leakage of recirculated (radioactive) water.

4) Relief valves are enclosed, i.e., they are provided with a closed bonnet.

6.3.2.2.9.1 Motor operated valves The seating design of motor operated gate valves is of the flexible wedge design. This design releases the mechanical holding force during the first increment of travel so that the motor operator works only against the frictional component of the hydraulic unbalance on the disc and the packing box friction. The disc is guided throughout the full disc travel to prevent chattering and to provide ease of gate movement. The seating surfaces are hard faced to prevent galling and to reduce wear. Motor operators may also be installed on globe valves. Where a gasket is employed for the body to bonnet joint, it is either a fully trapped, controlled compression, spiral wound asbestos gasket with provisions for seal welding, or it is of the pressure seal design with provisions for seal welding. Generally, the motor operator incorporates a "hammer blow" feature that assists with opening gate valves. This feature allows the motor to attain its operational speed prior to exerting a force on the stem to unseat the valve disc. The NRC issued Generic Letter (GL) 89-10 which impacted select safety-related motor operated valves (MOV). The GL recommended development and implementation of a program which ensures that MOV switch settings are set and maintained such that they will operate under design-basis conditions for the life of the plant. As part of the program, valves were tested under static and dynamic conditions, where practicable, to determine performance characteristics. This test information was used to assist in establishing appropriate MOV switch settings. The GL (89 10 recommendations have been completed in accordance with plant-specific commitments (see References 6.3.2-3, 6.3.2-4 and 6.3.2-5. NRC Generic Letter 96-05 requests the establishment of a program to verify on a periodic basis that safety-related MOVs continue to be capable of performing their safety functions within the Amendment 63 Page 95 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 current licensing basis. The Harris Nuclear Plant has committed to implement the Periodic Verification (PV) program developed by the Joint Owners Group (JOG) in response to Generic Letter 96-05. 6.3.2.2.9.2 Manual globes, gates, and check valves Gate valves employ a wedge design and are straight through. The wedge is either split or solid. Gate valves have backseats and outside screw and yoke construction. Globe valves, "T" and "Y" style are full ported with outside screw and yoke construction. Check valves are spring loaded lift piston types for sizes 2 in. and smaller, swing type for sizes 2-1/2 in. to 4 in., and tilting disc type for sizes 4 in. and larger. Stainless steel check valves have no penetration welds other than the inlet, outlet, and bonnet. The check hinge is serviced through the bonnet. The stem packing and gasket of the stainless steel manual globe and gate valves are similar to those described above for motor operated valves. Carbon steel manual valves are employed to pass nonradioactive fluids only and therefore do not contain the double packing and seal weld provisions. 6.3.2.2.9.3 Accumulator check valves (Swing-Disc) The accumulator check valve is designed with a low pressure drop configuration with all operating parts contained within the body. Design considerations and analyses which assure that leakage across the check valves located in each accumulator injection line will not impair accumulator availability are as follows:

1. During normal operation the check valves are in the closed position with a nominal differential pressure across the disc of approximately 1650 psi. Since the valves remain in this position except for testing or when called upon to open following an accident and are therefore not subject to the abuse of flowing operation or impact loads caused by sudden flow reversal and seating, they do not experience significant wear of the moving parts, and are expected to function with minimal backleakage. This backleakage can be checked via the test connection as described in Section 6.3.4.

2. When the RCS is being pressurized during the normal plant heatup operation, the check valves are tested for leakage in accordance with Technical Specifications prior to exceeding 1000 psi RCS pressure. This test confirms the seating of the disc and whether or not there has been an increase in the leakage since the last test. When this test is completed, the accumulator discharge line motor operated isolation valves are opened and the RCS pressure increase is continued. There should be no increase in leakage from this point on since increasing reactor coolant pressure increases the seating force and decreases the probability of leakage.

3. The experience derived from the check valves employed in the emergency injection systems indicate that the system is reliable and workable; check valve leakage has not been a problem. This is substantiated by the satisfactory experience obtained from operation of the Robert Emmett Ginna and subsequent plants where the usage of check valves is identical to SHNPP.

Amendment 63 Page 96 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

4. The accumulators can accept some in-leakage from the RCS without affecting availability. Continuous in-leakage would require, however, that the accumulator water volume be adjusted periodically to Technical Specification requirements.

6.3.2.2.9.4 Relief valves Relief valves are installed in various sections of the ECCS to protect lines which have a lower design pressure than the RCS. The valve stem and spring adjustment assembly are isolated from the system fluids by a bellows seal between the valve disc and spindle. The closed bonnet provides an additional barrier for enclosure of the relief valves. Table 6.3.2-2 lists the system's relief valves with their capacities and setpoints. 6.3.2.2.9.4.1 Accumulator relief valves Accumulator relief valves are procured to specifications requiring certain gas relieving capacities at certain temperatures (60 - 120 F). Water relief is not a design basis for the accumulator relief valves. The maximum fill rate for the accumulator at the relief valve setpoint (700 psi) is 35 gpm or 4.7 scfm (liquid). This additional water volume increase during an event which requires relief capacity can be assumed to be negligible compared to the relief valve capacity of 1500 scfm. Since the design transient is the case of maximum nitrogen make-up to the accumulator, a coincident water fill operation has a very small effect on the relief valve capability. If these valves had to relieve water, it is expected that they would do so at rates of from 50 to 150 gpm at temperatures of 60° to 120° F. 6.3.2.2.9.5 Butterfly valves Each main residual heat removal line has an air operated butterfly valve which is normally open and is designed to fail to the open position. The actuator is arranged such that air pressure on the diaphragm overcomes the spring force, causing the linkage to move the butterfly to the closed position. Upon loss of air pressure, the spring returns the butterfly to the open position. These valves are left in the full open position during normal operation to maximize flow from this system to the RCS during the injection mode of the ECCS operation. These valves are used during normal residual heat removal system (RHRS) operation to control cooldown flowrate. Each residual heat removal heat exchanger bypass line has an air operated butterfly valve which is normally closed and is designed to fail closed. Those valves are used during normal cooldown to avoid thermal shock to the residual heat exchanger. 6.3.2.2.9.6 Accumulator motor operated valve controls As part of the plant shutdown administrative procedures, the operator is required to close these valves. This prevents a loss of accumulator water inventory to the RCS and is done shortly after the RCS has been depressurized below 1000 psig. The redundant pressure and level alarms on each accumulator would remind the operator to close these valves, if any were inadvertently left open. Power is disconnected after the valves are closed. During plant startup, the operator is instructed via procedures to energize and open these valves before the RCS pressure reaches 1000 psig. Monitor lights in conjunction with an Amendment 63 Page 97 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 audible alarm will alert the operator should any of these valves be left inadvertently closed once the RCS pressure increases beyond the safety injection unblock setpoint. The accumulator isolation valves are not required to move during power operation or in a post-accident situation except for valve testing. For a discussion of limiting conditions for operation and surveillance requirements of these valves, refer to the Technical Specifications. For further discussions of the instrumentation associated with these valves refer to Sections 6.3.5 and 7.6.1.2. 6.3.2.2.9.7 Motor operated valves and controls Remotely operated valves for the injection mode which are under manual control (i.e., valves which normally are in their ready position and do not require a safety injection signal) have their positions indicated on a common portion of the main control board. If a component is out of its proper position, its monitor light will indicate this on the control panel. At any time during operation when one of these valves is not in the ready position for injection, this condition is shown visually on the board, and an audible alarm is sounded in the Control Room. The ECCS delivery lag times are given in Chapter 15.0. The accumulator injection time varies as the size of the assumed break varies since the RCS pressure drop will vary proportionately to the break size. Inadvertent mispositioning of a motor operated valve due to a malfunction in the control circuitry in conjunction with an accident has been analyzed and found not to be a credible event for use in design. Table 6.3.2-3 is a listing of motor operated isolation valves in the ECCS showing interlocks, automatic features, and position indications. 6.3.2.2.10 Auxiliary Systems Required for Operation and Support of the ECCS 6.3.2.2.10.1 Primary Auxiliary Systems The primary auxiliary systems required to support the ECCS are as follows: a) The engineered safety features (ESF) electrical buses; to provide electric power to the ECCS pumps and motor operated valves. If offsite power is available, loading of the emergency diesel generators onto the ESF buses is not required (see Section 8.3.1). b) The component cooling water system; to provide cooling to the RHR pumps and RHR heat exchangers (in recirculation mode only). The standby component cooling water pump is started by the "S" signal. Flow to the RHR heat exchangers is initiated by the operator prior to the switch to recirculation. c) The chilled water system (see Section 9.2.8) to provide cooling water to the ECCS pump room cooling system air handling units. The standby chiller and standby chilled water pump are started by the "S" signal. Amendment 63 Page 98 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 d) The service water system (see Section 9.2.1) to provide bearing and gear oil cooling for the charging pumps. 6.3.2.2.10.2 Secondary Auxiliary Systems Secondary auxiliary systems required to directly support the primary auxiliary systems listed above: a) The emergency diesel generators (see Section 8.3.1); to provide electric power to the ESF buses in the event of loss of offsite power. The emergency diesel generators are started upon receipt of the "S" signal. Supporting systems for operation of the emergency diesel generators and methods for actuation of these systems are as follows:

1) Diesel generator fuel oil storage and transfer system; started by a low level signal from the day tank. (See Section 9.5.4).

2) Diesel generator cooling water system; cooling water is supplied by operation of the associated service water system cooling loop. (See Section 9.5.5).

3) Diesel generator starting system; started by the "S" signal. (See Section 9.5.6).

4) Diesel generator lubrication system; components are engine driven. (See Section 9.5.7).

5) Diesel generator combustion air intake and exhaust system; system is passive and includes no operating components. (See Section 9.5.8).

6) Diesel generator building ventilation system; fans start when diesel generators start.

b) The service water system; to supply cooling water to the following:

1) Component cooling heat exchangers.

2) Diesel generator starting system air compressor aftercoolers.

3) Diesel generator lubrication system oil coolers.

4) Chilled water system water chiller condensers.

The emergency service water pump is started by the "S" signal. c) ECCS pump room air handling unit fans for the charging pump rooms and residual heat removal/reactor building spray pump rooms to provide ventilation and cooling for the ECCS pumps. These fans are operated during normal plant conditions and are started by the "S" signal. d) Ventilation systems for the control room, relay room and ESF switchgear rooms (see Sections 9.4.1.2.1, 9.4.1.2.2 and 9.4.5.2.2); to provide ventilation for controls associated with ECCS equipment. Amendment 63 Page 99 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.3.2-10 is a list of pumps and valves required for ECCS operation along with their safety classification. Each component is Seismic Category I. 6.3.2.3 Applicable Codes and Classifications Applicable industry codes and classifications for the ECCS are discussed in Section 3.2. 6.3.2.4 Material Specifications and Compatibility Materials employed for components of the ECCS are given in Table 6.3.2-4. Materials are selected to meet the applicable material requirements of the codes in Table 3.2.1-1 and the following additional requirements: a) All parts of components in contact with borated water are fabricated of or clad with austenitic stainless steel or equivalent corrosion resistant material. See Table 6.1.1-1. b) All parts of components in contact (internal) with sump solution during recirculation are fabricated of austenitic stainless steel or equivalent corrosion resistant material. See Table 6.1.1-1. c) Valve seating surfaces are hard faced to prevent galling and to reduce wear. d) Valve stem materials are selected for their corrosion resistance, high tensile properties, and resistance to surface scoring by the packing. 6.3.2.5 System Reliability a) General - Reliability of the ECCS is considered in all aspects of the system from initial design to periodic testing of the components during plant operation. The ECCS is a two train, fully redundant standby safeguard feature. The system has been designed and proven by analysis to withstand any single credible active failure during injection or active or passive failure during recirculation and maintain the performance objectives desired in Section 6.3.1. Two trains of pumps, heat exchangers, and flow paths are provided for redundancy. Only one train is required to satisfy the performance requirements. Due to this concept, either of the two subsystems can be isolated and removed from service if maintenance is required on any ECCS component. The initiating signals for the ECCS are derived from independent sources measured from process variables (e.g., low pressurizer pressure) or environmental variables (e.g., containment pressure). Redundant as well as functionally independent variables are measured to initiate the safeguards signals. Each train is physically separated and protected where necessary so that a single event cannot initiate a common failure. Power sources for the ECCS are divided into two independent trains supplied from offsite power via the emergency buses. Sufficient diesel generating capacity is also maintained onsite to provide required power to each train. The diesel generators and their auxiliary systems are completely independent and each supplies power to one of the two ECCS trains. Each compartment is provided with adequate radiation shielding such that access to any compartment for required maintenance is permissible during the recirculation phase. To obtain access to a given compartment, pumps in that compartment would be stopped and the lines flushed with water from the RWST. Provisions for washing down the floors and walls are provided to reduce contamination in the event that a leak occurs in a compartment during the Amendment 63 Page 100 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 recirculation phase. Adequate ventilation is provided to permit access for maintenance. The piping and valves associated with the pumps (refer to FSAR Figure 5.4.7-1) are arranged so that the system can be drained and flushed prior to maintenance. To meet this requirement, manual valves are provided with extended reach rods so that they can be operated from a position external to the pump compartments. The quality assurance program, as approved by the NRC during the construction permit review, assures receipt of components only after manufacture and test to the applicable codes and standards. The reliability program extends to the procurement of the ECCS components such that only designs which have been proven by past use in similar applications are acceptable. For example, the equipment specification for the ECCS pumps (safety injection, centrifugal charging, and residual heat removal pumps) require them to be capable of performing their long-term cooling function for one year. The same type of pump has been used extensively in other operating plants. Their function during recurrent normal power and cooldown operations in such plants as Zion, D. C. Cook, Trojan, and Farley has successfully demonstrated their performance capability. Reliability tests and inspections (see Subsection 6.3.4.1) further confirm their long-term operability. Nevertheless, design provisions are included that would allow maintenance on ECCS pumps, if necessary, during long-term operation. All of the Westinghouse active pump applications have gathered extensive operating time. These pumps are seismically qualified by a combination of analysis and tests which includes structural and operability analysis. Each pump is tested in the vendor's shop to verify hydraulic and mechanical performance. Performance is again checked at the plant site during preoperational system checks and quarterly per ASME Section XI. Pump design is specified, with strong consideration given to shaft critical speed, bearing, and seal design. Thermal transient and 100-hour endurance tests have been completed on the centrifugal charging and the safety injection pumps. Additional rotor dynamics tests have been performed on the centrifugal charging pumps which are the highest speed applications. A thermal transient analysis has been performed on the RHR pump; this analysis is supported by the vendor's test on a similar design. Endurance and leak determination testing has been completed on the mechanical seals by the seal supplier. This testing included various temperature, pressure, radiation, and boric acid concentration levels. These conditions were all substantially elevated over those expected during normal or post-accident conditions. The preoperational testing program assures that the systems as designed and constructed meet the functional requirements as calculated in design. The ECCS is designed with the ability for on-line testing of most components so the availability and operational status can be readily determined. All ECCS equipment has been designed to perform its system operating function at least one year without any periodic maintenance. The two independent ECCS subsystems/or trains allow maintenance to be performed on any pump, if it is necessary, during long-term operation. In addition to the above, the integrity of the ECCS is assured through examination of critical components during the routine inservice inspection. Amendment 63 Page 101 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.3.2.5.1 Active failure criteria The ECCS is designed to accept a single failure following an incident without loss of its protective function. The system design will tolerate the failure of any single active component in the ECCS itself or in the necessary associated service systems at any time during the period of required system operations following an incident. A failure modes and effects analysis for a single active failure is presented in Table 6.3.1-1, and demonstrates that the ECCS can sustain the failure of any single active component in either the injection or recirculation phase and still meet the required level of performance for core cooling. Since the operation of the active components of the ECCS following a steam line rupture is identical to that following a LOCA, the same analysis is applicable and the ECCS can sustain the failure of any single active component and still meet the required level of performance for the addition of shutdown reactivity. 6.3.2.5.2 Passive failure criteria As discussed in the following, sufficient redundancy is provided in ECCS component and system arrangement to meet the intent of the General Design Criteria on single failure as it specifically applies to failure of passive components. Thus, for the recirculation phase, the system design is based on accepting either a passive or an active failure. The non-safety section of the SI pumps miniflow header is seismically designed and analyzed pipe; failure is thus not expected. 6.3.2.5.2.1 Redundancy of Flow Paths and Components for Long-term Emergency Core Cooling In design of the ECCS, the following criteria are utilized:

1) During the long-term cooling period following a loss of coolant (recirculation phase), the emergency core cooling flow paths shall be separable into two subsystems, either of which can provide minimum core cooling functions and return spilled water from the floor of the Containment back to the RCS.

2) Either of the two subsystems can be isolated and removed from service in the event of a leak outside the Containment.

3) Adequate redundancy of check valves is provided to tolerate passive failure of a check valve during recirculation.

4) Should one of these two subsystems be isolated in this long-term period, the other subsystem remains operable.

5) Provisions are also made in the design to detect leakage from components outside the Containment, collect this leakage and to provide for maintenance of the affected equipment.

Amendment 63 Page 102 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Thus, for the long-term emergency core cooling function, adequate core cooling capacity exists with one flow path removed from service. 6.3.2.5.2.2 Subsequent Leakage from Components in Safeguards Systems With respect to piping and mechanical equipment outside the Containment, considering the provisions for visual inspection and leak detection, leaks will be detected before they propagate to major proportions. A review of the equipment in the system indicates that the largest sudden leak potential would be the sudden failure of a pump shaft seal. Evaluation of leak rate assuming only the presence of a seal retention ring around the pump shaft showed flows less than 50 gpm would result. Piping leaks, valve packing leaks, or flange gasket leaks usually build up slowly with time and are considered less severe than the pump seal failure. Larger leaks in the ECCS are prevented by the following:

1) The piping is classified in accordance with ANS Safety Class 2 and receives the ASME Class 2 quality assurance program associated with this safety class.

2) The piping, equipment and supports are designed to ANS Safety Class 2, Seismic Category I permitting no loss of function for the design basis earthquake.

3) The system piping is located within a controlled area on the plant site.

4) The piping system receives periodic pressure tests and is accessible for periodic visual inspection.

5) The piping is austenitic stainless steel which, due to its ductility, can withstand severe distortion without failure.

Based on this review, the design of the Reactor Auxiliary Building and related equipment is based upon handling of leaks up to a maximum of 50 gpm. Means are also provided to detect and isolate such leaks in the emergency core cooling flow path within four hours. A single passive failure analysis is presented in Table 6.3.2-5. It demonstrates that the ECCS can sustain a single passive failure during the recirculation phase and still retain an intact flow path to the core to supply sufficient flow to maintain the core covered and affect the removal of decay heat. The procedure followed to establish the alternate flow path also isolates the component which failed. Figures 6.3.2-4 through 6.3.2-7 are simplified illustrations of the ECCS. The notes provided with Figures 6.3.2-4 through 6.3.2-7 contain information relative to the operation of the ECCS in its various modes. The modes of operation illustrated are full operation of all ECCS components, cold leg recirculation with residual heat removal pump number 2 operating, and hot leg recirculation with residual heat removal pump number 1 operating. These are representative of the operation of the ECCS during accident conditions. Lag times for initiation and operation of the ECCS are limited by pump startup time and consequential loading sequence of these motors onto the emergency buses. Most valves are normally in the position conducive to safety, therefore valve opening time is not considered for these valves. If the normal offsite power supply is available, all pump motors and valve motors Amendment 63 Page 103 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 are started immediately upon receipt of the "S" signal. Without offsite power, a 10 second delay is assumed for diesel generator startup, then pumps and valves are loaded according to the sequencer. In any case, full injection flow is achieved within 27 seconds of reaching the safety parameter setpoint. In both the large and small break LOCA analyses, full injection flow was conservatively assumed to occur within 29 seconds. ECCS piping is designed such that normal system operation and testing assures that the systems remain water-filled to preclude the effects of water hammer. Interfaces with normally pressurized non-ECCS systems preclude a loss of water from ECCS systems. Leakage from ECCS systems through valve packing, pump seals, etc., will be detected by any number of methods including: 1) normal operator rounds, 2) performance during testing, 3) the plant leak reduction inspection program, 4) various sump level alarms, 5) decreasing water levels in various tanks. Should significant leakage be discovered, where an introduction of air into the system could have occurred, provisions have been made in the system design to permit refilling and venting of the affected components or piping following repair to the source of leakage. 6.3.2.5.2.3 Potential boron precipitation Boric acid buildup considerations during long-term cooling have been addressed in the letter from C. Caso of Westinghouse to T. Novak of NRC dated April 1, 1975. This letter presents the method, assumptions, and results of analysis for a typical plant. During cold leg recirculation for a cold leg pipe break the analysis shows that boric acid concentrations within the reactor vessel and core regions remain at acceptable levels up to the time of the initiation of hot leg recirculation. An analysis has been performed for Shearon Harris to determine the maximum boron concentration in the reactor vessel following a hypothetical LOCA. The analysis considers the increase in boric acid concentration in the reactor vessel during the long-term cooling phase of a LOCA assuming a conservatively small effective vessel volume including only the free volumes of the reactor core and the upper plenum below the bottom of the hot leg nozzles. This assumption conservatively neglects the mixing of boric acid solution with directly connected volumes, such as the reactor vessel lower plenum. The calculation of boric acid concentration in the reactor vessel considers a cold leg break of the reactor coolant system in which steam is generated in the core from decay heat while the boron associated with the boric acid solution is completely separated from the steam and remains in the effective vessel volume. The results of the analysis show that the maximum allowable boric acid concentration established by the NRC, which is the boric acid solubility limit minus 4 w/o, will not be exceeded in the vessel if hot leg recirculation is initiated when the following criteria are met:

1. The Safety Injection System has previously been aligned for cold leg recirculation (meaning that the Refueling Water Storage Tank level has been depleted), and

2. Approximately 6.5 hours have passed since the beginning of the event, and

3. Safety Injection has not been terminated such that a single Charging Safety Injection Pump has been realigned to the charging header (meaning that Reactor Coolant System subcooling and Pressurizer level have been established).

Amendment 63 Page 104 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 (Reference 6.3.1-1) (See Sections 6.3.2.8 and 15.6.5.2) The SHNPP will utilize alternating hot and cold leg recirculation to prevent excessive concentration in the reactor vessel during long-term operation following a LOCA. The switch between hot leg recirculation and cold leg recirculation should occur approximately every 6.5 hours after the initiation of hot leg recirculation. This method of preventing boron concentration complies with the requirements of the NRC staff position concerning boron dilutions. The amount of flow which must be maintained through the core at the time of hot leg switchover is greater for a hot leg break than for a cold leg break. If at least one RHR pump and one CSIP are successfully aligned to the hot leg, sufficient flow is maintained through the core for either the hot or cold leg break (Reference 6.3.1-1). Sufficient flow cannot be maintained through the core with only one CSIP aligned to the hot legs if the break is at the hot leg. In order to establish Low Head Safety Injection to the hot legs, a single motor operated valve must be remotely opened by the operator. If this valve fails to open, the operator is directed by the Emergency Operating Procedures to re-establish flow from the RHR pump to the cold leg. This action ensures that sufficient flow is maintained for a hot leg break. The operator is also directed to complete the alignment of the CSIP to the hot leg to ensure sufficient flow is maintained for a cold leg break (Reference 6.3.1-1). Discussions of hot leg and cold leg recirculation modes of the ECCS are presented in Section 6.3. Since the ECCS is designed to meet the single failure criterion, no back up means to prevent the buildup of boron concentration is provided. All components of the ECCS are ANS Safety Class 2. ECCS testing is discussed in Section 6.3.4. 6.3.2.6 Protection Provisions The provisions taken to protect the system from damage that might result from dynamic effects of pipe rupture are discussed in Section 3.6. The provisions taken to protect the system from missiles are discussed in Section 3.5. The provisions to protect the system from seismic damage are discussed in Sections 3.7, 3.9 and 3.10. The provisions to protect the system from flooding are discussed in Section 3.4. Thermal stresses on the RCS are discussed in Section 5.2. 6.3.2.7 Provisions for Performance Testing Test lines are provided for performance testing of the ECCS as well as individual components. These test lines and instrumentation are shown in Figures 6.3.2-1 through 6.3.2-3. All pumps have miniflow lines for use in testing operability. Additional information on testing can be found in Section 6.3.4.2. Amendment 63 Page 105 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.3.2.8 Manual Actions The plant emergency operating procedures include instructions and verification steps to ensure proper manual realignment of the ECCS for recirculation by the operator. The failure to perform one step or the performance of one step out of order, as a single failure, should not reduce ECCS recirculation capability below minimum safeguards. Should the operator fail to take any action following automatic ECCS switchover initiation, the consequences will be the loss of the safety injection (charging) pumps. The residual heat removal pumps will be protected from damage by automatic ECCS switchover initiation. In the unlikely event of losing all high head injection capability, the situation could lead from a small break LOCA to core uncovering and inadequate core cooling. Analyses have been performed and are addressed in WCAP-9691 as the loss of the Emergency Coolant Recirculation (ECR) function following a small break LOCA. Inadequate Core Cooling guidelines instruct the operator on the appropriate actions to be taken for this event. No manual actions are required of the operator for proper operation of the ECCS during the injection mode of operation. Only limited manual actions are required by the operator to realign the system for the cold leg recirculation mode of operation, and for the hot leg recirculation mode of operation. These actions are delineated in Table 6.3.2 6. Subsequent to establishing recirculation mode of operation, re-injection from the RWST may be necessary to compensate for ECCS leakage outside of containment in order to maintain adequate recirculation sump inventory. Re-injection can be accomplished by manually aligning a CT pump (preferred) for injection from the RWST. The changeover from the injection mode to recirculation mode is initiated automatically and completed manually by operator action from the Control Room. The design of the Refueling Water Storage Tank (RWST) at Shearon Harris Nuclear Plant includes allowances to account for working and transfer water allowance, instrument error, single failure and the unusable volume of water present in the bottom of the tank. Consideration has been given to the amount of water required for core reflood and cooling and the pH requirements for water entrained in the containment sump. Additionally, the positioning of the instrument (alarm) levels permits sufficient time for appropriate operator action required for ECCS switchover to recirculation. The shortest times available for ECCS injection and switchover are as follows: a) Injection Mode Allowance - The safety injection mode of ECCS operation consists of the ECCS pumps (charging pumps and residual heat removal pumps) and the containment spray pumps taking suction from the refueling water storage tank (RWST) and delivering to the reactor coolant system (RCS) and containment, respectively. In order to analyze the shortest time available for injection mode operation, the following conservative bases are established:

1) The minimum RWST volume available for injection mode operation is 266,625 gallons.

2) To maximize flow out of the RWST, the following conservative assumptions are utilized:

Amendment 63 Page 106 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

a. Containment spray flowrate is based on a containment pressure equal to the Containment Spray Actuation Signal Pressure which occurs when a containment pressure HI-3 signal is reached.

b. Charging and RHR pumps are operating in their minimum resistance configuration, with 0.0 psig Containment and RCS backpressure.

3a) Minimum safeguards response-Flow out of the RWST during the injection mode includes conservative allowances for one pump of each type operating at the following flow rates: Charging pump - 685 gpm per pump RHR pump - 4410 gpm per pump Spray pump - 2127 gpm per pump Total RWST outflow during injection mode operation is 7222 gpm. Based on the above minimum available RWST volume for injection mode operation and the maximum total flow rate out of the RWST, the shortest injection mode operation time for single train ECCS operation is approximately 2215 sec. or 36.92 minutes 3b) Maximum safeguards response -Flow out of the RWST during the injection mode includes conservative allowance for pumps operating at the following flow rates: Charging Pump - 426 gpm per pump (two pumps) RHR Pump - 3096 gpm per pump (two pumps) Containment Spray Pump 4417 gpm for both pumps Total outflow during injection mode operation is 11,461 gpm. The Containment Spray flow rates are based on discharge to a backpressure of 0.0 psig. Based on the above minimum available RWST volume for injection mode operation and the maximum total flow rate out of the RWST, the shortest injection mode operation time for maximum safeguards operation is approximately 23 minutes. For conservatism this was reduced to 20 minutes in the containment pressure - temperature transient analysis. b) Transfer Allowance - During the safety injection mode of ECCS operation, the operator monitors the RWST level and containment recirculation sump level in anticipation of switchover. During this time, the operator normally opens the component cooling water inlet isolation valves to the residual heat removal (RHR) heat exchanger. The ECCS switchover from safety injection to cold leg recirculation is initiated automatically upon receipt of the RWST low-low level signal in conjunction with the safety injection signal and is completed via timely operator action at the main control board. Switchover is initiated via automatic opening of the containment recirculation sump isolation Valves (8811 A/B and 8812 A/B). This automatic action aligns the suction of the RHR pumps to the containment recirculation sump to ensure continued availability of a suction source. Amendment 63 Page 107 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The low-low RWST level signal, which initiated the automatic opening of the containment sump valves, also provides an alarm to inform the operator that he must initiate the manual actions required to complete switchover. Manual actions of Table 6.3.2-6 must be performed following switchover initiation prior to loss of the RWST transfer allowance to ensure that all ECCS pumps are protected with suction flow available from the containment sump. The ECCS switchover procedure is structured so that the operator simultaneously switches both trains of the ECCS from injection to recirculation, repositioning functionally similar switches as part of the same procedural steps. The time available for switchover is dependent on the flow rate out of the RWST as the switchover manual actions are performed. As ECCS valves are repositioned, the flow rate out of the RWST is reduced in magnitude. In order to analyze the shortest time available for switchover, the following conservative bases are established:

1) Valve stroke times used are conservative design values.

2) The RWST water volume "required" for switchover is approximately 64,688 gallons.

3) To maximize the RWST outflow and thus minimize the switchover duration, the RCS is assumed to be at 0 psig, and the containment pressure equal to the Containment Spray Actuation Signal pressure. Thus, no credit is taken for the reduction in RWST outflow that will result with the higher containment and RCS pressure following a large break.

The same conservative assumption is made for the small break conditions (except that RCS pressure is assumed to be greater than RHR pump shutoff head resulting in no RHR pump flow to the RCS for small break conditions).

4) Flow out of the RWST during switchover includes allowances for both pumped flow to the RCS and containment and backflow to the containment sump based on the 0 psig containment pressure assumption. Average flow rates are assumed during switchover and include the following conservative flow rate allowances assuming two pumps of each type are operating:

Charging pump - 426 gpm per pump RHR pump - 3096 gpm per pump Spray pump - 2137 gpm per pump Backflow to the containment sump may occur during ECCS switchover based on the 0 psig containment pressure assumption and ECCS operating conditions. Backflow, if it occurs, will vary as the switchover proceeds depending on ECCS alignment.

5) Flow rate out of the RWST for the worst ECCS single failure condition is determined assuming one of the RWST/RHR isolation valves (8809A or 8809B) fails to close on demand. This single failure maximizes RWST outflow during switchover. Flow rates out of the RWST assume no operator corrective action to mitigate the single failure (i.e., stop the affected RHR pump and close the appropriate sump isolation valves).

Amendment 63 Page 108 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Based on the criteria, the calculated flow rates out of the RWST as a function of switchover manual action are itemized in Table 6.3.2-9 for large breaks. The large break with single failure constitutes the condition where RWST outflow is the greatest. Flow rate data for small breaks is less than for large breaks and is not included in Table 6.3.2-9. Table 6.3.2-9 also identifies the operator action time assumed per switchover step and shows the change in RWST volume per switchover step. Analyzing the flow rate out of the RWST for large LOCA with single failure indicates that approximately 64,688 gallons are consumed in switchover. The volume of water available (transfer allowance) between the nominal RWST "Lo-Lo" level setpoint and the nominal "Empty" level setpoint is approximately 84,000 gallons. This shows that the switchover steps necessary to protect all ECCS pumps can be accomplished before the transfer allowance is depleted. Protection logic is provided to automatically open the ECCS recirculation containment sump isolation valves when two out of four refueling water storage tank level channels indicate a refueling water storage tank level less than a low-low level setpoint in conjunction with the initiation of the safety injection actuation signal ("S" signal). This automatic action aligns the two RHR pumps to take suction directly from the containment sump. The charging pumps will continue to take suction from the RWST, following the above automatic action, until manual operator action is taken to align these pumps in series with the RHR pumps. The low-low RWST level signal, which initiated the automatic opening of the containment sump valves, also provides an alarm to inform the operator that he must initiate the manual actions required to realign the RHR and charging pumps for the recirculation phase. The manual switchover sequence is delineated in Tables 6.3.2-6, 6.3.2 9, and 6.3.2-10. The RHR pumps would continue to operate during this changeover from injection mode to recirculation mode. Following the automatic and manual switchover sequence, the two RHR pumps would take suction from the containment sump and deliver borated water directly to the RCS cold legs. The RHR pump discharge flow would be used to provide suction to the two charging pumps which would also deliver directly to the RCS cold legs. The hot leg recirculation phase will be initiated when the following criteria are met:

1. The safety injection system has previously been aligned for cold leg recirculation (meaning that the Refueling Water Storage Tank level has been depleted), and

2. Approximately 6.5 hours have passed since the beginning of the event, and

3. Safety Injection has not been terminated such that a single Charging Safety Injection Pump has been realigned to the charging header (meaning that Reactor Coolant System subcooling and Pressurizer level have been established). (Reference 6.3.1-1) (See Sections 6.3.2.8 and 15.6.5.2)

The refueling water storage tank level protection logic consists of four level channels with each level channel assigned to a separate process control protection set. Four refueling water storage tank level transmitters provide level signals to corresponding normally deenergized level channel bistables. Each level channel bistable would be energized on receipt of a refueling water storage tank level signal less than the low-low level setpoint. Amendment 63 Page 109 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 A two out of four coincident logic is utilized in both protection cabinets A and B to ensure a trip signal in the event that two out of the four level channel bistables are energized. This trip signal, in conjunction with the "S" signal, provides the actuation signal to automatically open the corresponding containment sump isolation valves. As part of the manual switchover procedure, the discharge of the residual heat removal pumps are aligned to the suctions of the charging pumps. Charging pump discharge header cross connect valves are closed in order to establish two separate and redundant high head recirculation systems. The suction header cross connect valves are not closed to ensure that subsequent RHR pump failure does not cause an immediate charging pump failure. During startup and shutdown operation, the normal alignment of Emergency Core Cooling System (ECCS) equipment is changed from that which is available during power operation. Only shutdown is discussed here since shutdown conditions would be more limiting than startup due to the higher decay heat level following reactor shutdown. During the Reactor Coolant System (RCS) cooldown and depressurization following reactor shutdown, the low pressurizer pressure and low steamline pressure safety injection (SI) actuation logic is manually blocked when below the P 11 setpoint of 2000 psig. This action disarms the SI signal from the pressurizer and steamline pressure transmitters to prevent automatic SI initiation during the subsequent RCS cooldown and depressurization. The containment high pressure SI signal will actuate SI if the setpoint is exceeded. Manual SI actuation is also available for RCS temperatures above 200°F. At 1000 psig, the ECCS accumulator isolation valves are locked out to prevent accumulator injection when the RCS pressure is reduced below the accumulator pressure. For temperatures above 350°F, the Technical Specifications require that both ECCS subsystems (each subsystem consists of one centrifugal charging pump, one RHR pump, and one RHR heat exchanger) be operable, whereas only one ECCS subsystem is required to be operable for RCS temperatures between 200°F and 350°F. Also below 325°F only one centrifugal charging pump is allowed to be operable by the Technical Specifications to reduce the possibility of overpressurizing the RCS at low temperature conditions. The residual heat removal (RHR) pumps may also be in the RHR cooling mode where suction is drawn from the RCS hot legs when the temperature is less than 350°F. If a steamline rupture occurs while both the low pressurizer pressure and low steamline pressure SI actuation signals are blocked, steamline isolation will occur on high negative steam pressure rate. An alarm for steamline isolation will alert the operator of the accident. Although a steamline rupture may result in a significant cooldown of the RCS, there is no danger of uncovering the core, and thus the ECCS is not required for core cooling. The ECCS is also not required for post-accident reactivity control for this case since procedural requirements provide for boration of the RCS to cold shutdown conditions prior to blocking the SI actuation signals. Thus, if a steamline rupture occurs with the SI actuation signals blocked, there would not be any return to criticality, and the core would be protected. With regards to a LOCA, it has been determined that shutdown operating conditions are so far below the conditions for which the RCS has been designed that a large LOCA is not credible and for all practical purposes can be assumed not to occur. With the equipment status described above, it is concluded that operator actions can be taken for a credible LOCA to avoid exceeding the ECCS performance criteria. The supporting information for these statements is presented below. Amendment 63 Page 110 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 A rupture in the RCS pressure boundary piping greater than 6 inches in nominal diameter is considered to be highly unlikely even at normal operating pressure. Engineering studies and operating experience have shown that through wall cracks in the RCS Class 1 pressure boundary piping greater than 6 inches in nominal pipe diameter are highly unlikely. A leak-before-break analysis has been performed for the Shearon Harris plant and approved by the NRC. In addition, leak-before-break analyses have been performed for the RCS pressure boundary piping down to 10 inches in nominal pipe diameter for the McGuire and Catawba plants. It is expected that a similar analysis for the RCS pressure boundary piping down to 12 inches in diameter for the Shearon Harris plant would yield results comparable to those for the McGuire and Catawba plants. The results of the leak-before-break analyses demonstrated that even if a through wall crack is postulated at normal operating pressure, RCS pressure boundary leakage would be detected with existing leak detection systems and the crack will remain stable (i.e., not propagate to a pipe rupture). The maximum size leak which could occur in the piping down to 12 inches in diameter without being detected would be very small (i.e., less than 1 inch in equivalent diameter). Since there is no 10 inch or 8 inch nominal diameter piping in the Shearon Harris RCS pressure boundary, the next smaller piping size to be considered is 6 inches in nominal diameter (5.187 inches inside diameter). Thus, based on the available leak-before-break analyses, the maximum size pipe which could be assumed to rupture is the 6-inch piping which would result in a 5.187 inch diameter LOCA. Below the RCS normal operating pressure, a rupture in the RCS pressure boundary piping greater than 6 inches in nominal diameter is considered even more unlikely. Normal operation at 2000-2250 psig serves as a more severe condition which demonstrates that pipe ruptures below normal operating pressures are highly unlikely since additional margins of safety exist at the lower pressures. The condition which could lead to a pipe rupture, a large through wall crack, would be identified during operation. However, even with the presence of such a crack, the piping system would remain stable and a piping rupture would be unlikely at the reduced RCS pressure. Therefore, based on the above information, it is concluded that the maximum credible LOCA to be considered for the RCS pressure boundary during shutdown operations is 5.187 inches in diameter, corresponding to the rupture of a 6-inch pipe. The above information is not applicable to the Class 2 portions of the RHR system piping since it is not operated at the higher system pressures and has not been subjected to a leak before break analysis. The design pressure of the RHR system is approximately 600 psig and due to the nature of the operation of the system is considered a moderate energy system. Large ruptures of moderate energy system piping have not been considered as a part of the design basis of Westinghouse supplied PWRs. Breaks of relatively small size have been considered. Any leakage of the RHR system piping would be expected to occur when the system is initially pressurized at less than 400 psig. The RCS conditions are under manual control by the reactor operator and the operator will be monitoring the pressurizer level and the RCS loop pressure so that any significant leakage would be immediately detected. If a break is detected, the operator would isolate the RHR system from the RCS, terminating the loss of coolant, and initiate safety injection, if necessary. Based on the results from small LOCA studies provided in various plant license applications, the operator will have ample time to take these actions. Further discussions on shutdown LOCA are considered to be applicable to only breaks in the RCS inside containment which are not isolable. For a credible LOCA, the RCS break flow rate and depressurization rate is significantly less than for a design basis large break LOCA. For shutdown conditions, the break flow and Amendment 63 Page 111 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 depressurization rates would be further reduced due to the lower initial RCS pressure and temperature. In addition, the initial fuel rod temperatures and decay heat level would be significantly less than for full power since the reactor would have been shutdown for a period of time. Automatic SI actuation may not occur since the pressurizer pressure SI signal is blocked and the high containment pressure SI signal may not be operable or may not be reached for lower initial RCS temperatures. Operator action would be relied upon to initiate sufficient SI flow to maintain sufficient reactor vessel inventory for adequate core cooling. The indications available to the operator that there is a small or medium LOCA in progress would be the following:

1. Loss of pressurizer level

2. Decrease of RCS pressure

3. Loss of RCS subcooling

4. Radiation alarms inside containment

5. RVLIS Other potential indications include an increase in containment pressure and sump water level increasing. However, the reduced break flow rate and reduced energy in the break flow for small breaks at low initial RCS temperatures may not noticeably increase the containment pressure or increase the sump water level at a rate which would be readily detected.

If a credible LOCA should occur during shutdown conditions when the RCS temperature is above 350°F, the operators would only have to manually initiate SI since both ECCS subsystems would be available and the suction flow path would be automatically aligned to the RWST upon the SI signal. The accumulators would also be available for injection for initial RCS conditions above 1000 psig. Adequate ECCS performance is also expected below 1000 psig without the accumulators because of the lower decay heat levels which would exist due to the longer cooldown and depressurization time required to achieve this condition. Below 350°F, the operators would have to manually align the suction of the available SI pumps to the RWST and manually align required SI flow. Actuating SI is not desired due to isolating instrument air and nitrogen to the Pressurize PORVs, which may be needed to mitigate a cold repressurization event. Follow-up action would also be required to restore the remaining SI pumps or the accumulators to service. The RHR pumps will be aligned in the RHR cooling mode during part of the operating time below 350°F. Since the RHR pumps may be damaged by operating with highly voided flow, they will be tripped as soon as possible following a loss of RCS subcooling. This will ensure that the RHR pumps are available for long-term core cooling during the recirculation phase of operation. Thus, if the RHR pumps are operating in the RHR cooling mode, the operator would first have to stop the RHR pumps and then perform the actions indicated above to provide SI flow. The indications noted above will alert the operators to a LOCA so that they can perform the required manual actions. It is expected that the operators can complete the manual actions to align SI, align the flow path from the RWST, and restore the SI equipment to operable status as required during shutdown conditions such that the level of protection for a LOCA during shutdown conditions will be equivalent to that during power operation. However, the cooldown Amendment 63 Page 112 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 period necessary to assure safe operation in ECCS mode is not consistent with this design, therefore operation is restricted to the highest indicated RHR hot leg temperature of 201°F or 200°F if ERFIS is not available. At temperatures between 201°F and 350°F, one train of RHR is maintained in the ECCS mode to assure its capability to quickly respond in an ECCS injection function. To provide instruction to the operators on the above, procedures and training are provided. However, instrumentation is available to aid in detecting problems including recirculation sump level, RHR and Containment Spray Pumps discharge pressure and flow indications (all of which are on the main control board) and local indications of RHR and Containment Spray Pumps' suction pressure. An unexplained decrease in the discharge pressure and flow of any RHR or Containment Spray pump coupled with abnormal recirculation sump level at its corresponding intake might be indicative of vortex formation or screen blockage. The operator in this situation would closely monitor the recirculation flow and discharge pressure of the affected pump to ensure that it is stopped before damage occurs. The facility is designed for recirculation using only one RHR or CS train; therefore, this single train operation could be performed while the situation is diagnosed and appropriate corrective actions taken. Maintaining one train of RHR in ECCS mode above an indicated RCS temperature of 201°F (200°F without ERFIS) ensures that this train is not subject to fluid flashing and can perform its ECCS injection-mode function following a large-break LOCA and an associated rapid depressurization of the RCS and connected RHR piping. On plant start-up, placing an RHR train in ECCS mode prior to exceeding an indicated RCS temperature of 249°F (245°F without ERFIS), together with forced cooling of that train of specified minimum duration, ensures that excessive fluid flashing will not occur in the RHR suction piping of that train once it is realigned to the comparatively low-pressure RWST. See Section 7.5 for process information available to the operator in the Control Room following an accident. 6.3.3 PERFORMANCE EVALUATION Accidents which require ECCS operation are as follows:

1. Inadvertent opening of a steam generator power operated relief or safety valve (see Section 15.1.4).

2. Small break LOCA (see Section 15.6.5).

3. Large break LOCA (see Section 15.6.5).

4. Major secondary system pipe failure (see Section 15.1.5).

5. Steam generator tube failure (see Section 15.6.3).

Amendment 63 Page 113 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.3.3.1 Inadvertent Opening of a Steam Generator Relief or Safety Valve The most severe core conditions resulting from an accidental depressurization of the Main Steam Supply System are associated with an inadvertent opening of a single steam dump, power operated relief or safety valve. A safety injection system actuation can occur from any of the following:

1. Low pressurizer pressure signal.

2. Low steam line pressure.

3. Hi-l containment pressure.

4. Manual actuation.

A safety injection signal will rapidly trip the main turbine, close all feedwater control valves, trip the main feedwater pumps, and close the feedwater isolation discharge valves. Following the actuation signal, the isolation valves between the RWST and charging pump suction open and the suction of the centrifugal charging pumps is diverted from the volume control tank to the refueling water storage tank. Simultaneously, the valves isolating the boron injection tank from the injection header automatically open. After the isolation valves between the RWST and charging pump suction are opened, the charging pumps then force boric acid solution from the refueling water storage tank through the header and injection line (including the boron injection tank) and into the cold legs of each loop. The passive injection system (accumulators) and the low head system provide no flow during these events since reactor coolant system pressure remains relatively high. This event is described in further detail in FSAR Section 15.1.4. The steam dump control circuitry is designed on a de-energize to close principle, so that the preferred failure mode on loss of energy source is to close. Although the single failure criteria is not a design basis for control grade circuitry such as the steam dump controller, a review of credible single failures will show that the probability of failure (open) of more than one steam dump valve is low. A malfunction in the steam dump controller that causes the steam dump open initiating signal to be present when either a turbine load decreased or a turbine trip has not occurred will not cause steam dump to fail open. This is because the steam dump is blocked by the loss of load interlock unless a large turbine load decrease has occurred. Likewise a failure of the loss of load interlock will not cause the steam dump to fail open when the control signal is not introduced. In the unlikely event that control system failure opens the steam dump, the Protection System provides diverse protection grade actuation signals, that is, low-low TAVE Block of Steam Dump and Main Steamline Isolation (MSLI) that prevent a sustained cooldown. The low probability of failure of more than one steam dump valve recognizes that there is a distinction between potential non-design basis systems interaction and a random single failure of a component. An unanticipated systems interaction is not ruled out. The review shows that if more than one steam dump valve were to open, it would not be the effect of a single random failure of a component. Amendment 63 Page 114 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.3.3.2 Small Break LOCA A LOCA is defined as a rupture of the RCS piping or of any line connected to the system. Ruptures of a small cross section will cause expulsion of the coolant at a rate which can be accommodated by the charging pumps which would maintain an operational water level in the pressurizer permitting the operator to execute an orderly shutdown. The maximum break size for which the reactor coolant makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the RCS through the postulated break against the charging pump makeup flow at normal RCS pressure, i.e., 2250 psia. A makeup flow rate from one centrifugal charging pump is adequate to sustain pressurizer level at 2250 psia for a break through a 0.375 in. diameter hole. This break results in a loss of approximately 17.5 lb./sec. (127 gpm at 130°F and 2250 psia). Although automatic makeup to the VCT is set less than or equal to 120 gpm, the charging pumps are automatically realigned to the RWST upon receipt of a low VCT level signal. The makeup capability of the CSIPs when taking suction from the RWST is in excess of 127 gpm. The safety injection signal stops normal feedwater flow by closing the main feedwater line isolation valves and initiates emergency feedwater flow by starting auxiliary feedwater pumps. The small break analyses (Section 15.6.5) deals with breaks ranging from a 0.75-inch pipe size up to a 9.0-inch pipe size, where the charging pumps play an important role in the initial core recovery because of the slower depressurization of the RCS. The analysis of this accident has shown that the high head portion of the ECCS, together with accumulators, provide sufficient core flooding to keep the calculated peak clad temperature below required limits of 10CFR50.46. Hence, adequate protection is afforded by the ECCS in the event of a small break LOCA. The SBLOCA spectrum includes break sizes where RHR injection occurs, however, the limiting breaks have only high head injection and SI accumulators. 6.3.3.3 Large Break LOCA A major LOCA is defined as a rupture of the RCS piping including the double ended rupture of the largest pipe in the RCS or of any line connected to that system. The boundary considered for LOCA as related to connecting piping is defined in Section 3.6. Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. Reactor trip occurs and the safety injection system is actuated when the pressurizer low pressure trip or Hi l containment pressure setpoints are reached. These countermeasures will limit the consequences of the accident in three ways:

1. Reactor trip and borated water injection provide additional negative reactivity insertion to supplement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

2. Injection of borated water ensures sufficient flooding of the core to prevent excessive clad temperatures.

Amendment 63 Page 115 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

3. During long-term recirculation and cooling, borated water serves to maintain the core sub-critical.

When the pressure falls below approximately 600 psi the accumulators begin to inject borated water. The conservative assumption is made that accumulator water injected bypasses the core and goes out through the break until the termination of the blowdown phase. This conservatism is again consistent with 10CFR50 Appendix K. The pressure transient in the Containment during a LOCA affects ECCS performance in the following ways. The time at which end of blowdown occurs is determined by zero break flow which is a result of achieving pressure equilibrium between the RCS and the Containment. In this way the amount of accumulator water bypass is also affected by the containment pressure, since the amount of accumulator water discharged during blowdown is dependent on the length of the blowdown phase and RCS pressure at end of blowdown. During the reflood phase of the transient, the density of the steam generated in the core is dependent on the existing containment pressure. The density of this steam affects the amount of steam which can be vented from the core to the break for a given downcomer head, the core reflooding process, and, thus, the ECCS performance. It is through these effects that containment pressure affects ECCS performance. For breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS will limit the clad temperature to well below the melting point and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved. For breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS will meet the acceptance criteria as presented in 10CFR50 Appendix K. That is:

1. The calculated peak fuel element clad temperature provides margin to the requirement of 2200°F.

2. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17 percent are not exceeded during or after quenching.

4. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

6.3.3.4 Major Secondary System Pipe Failure The steam release arising from a rupture of a main steam pipe would result in energy removal from the RCS causing a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin. There is an increased possibility that the core will become critical and return to power. A return to power following a steam pipe rupture is a potential problem. The core is ultimately shut down by the boric acid injection delivered by the Safety Injection System. Amendment 63 Page 116 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 For the cases where offsite power is assumed, the sequence of events in the safety injection system is the following. After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the high head safety injection (charging) pumps start. In 12 seconds after initiation of a safety injection signal, the high head safety injection discharge valves are assumed to be in their final position and the pumps are assumed to be at full speed. The RWST to charging pump suction valves are fully open 17 seconds after initiation of a safety injection signal. This delay, described above, is inherently included in the modeling. The VCT is isolated from the charging pump suction in 10 additional seconds after the RWST to CSIP suction valves are fully open. This delay and the transport delay for boron from the RWST to the core are included in the modeling. In cases where offsite power is not available, an additional 10 second delay is assumed to start the diesel generators and to load the necessary safety injection equipment. The analysis has shown that even assuming a stuck rod cluster control assembly with or without offsite power, and assuming a single failure in the engineered safeguards the core remains in place and intact. Radiation doses will not exceed 10 CFR 50.67 guidelines. Departure from nucleate boiling and possible clad perforation following a steam rupture are not necessarily unacceptable and not precluded in the criterion. The detailed analysis of whether departure from nucleate boiling may be expected to occur is presented in Section 15.1.5. 6.3.3.5 Steam Generator Tube Failure The accident examined is the complete severance of a single steam generator tube at power. Assuming normal operation of the various plant control systems, the following sequence of events is initiated by a tube failure:

1. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side there is a steam flow/feedwater flow mismatch before trip as feedwater flow to the affected steam generator is reduced due to the break flow which is now being supplied to that steam generator from the primary side.

2. The condenser vacuum pump effluent radiation monitor, steam generator blowdown line radiation monitor, and/or main steamline radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system.

3. Continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure or by overtemperature T. Resultant plant cooldown following reactor trip leads to a rapid decrease in RCS pressure and pressurizer level. A safety injection (SI) signal, initiated by low pressurizer pressure, follows soon after the reactor trip. The SI signal automatically terminates steam generator blowdown, normal feedwater supply and initiates auxiliary feedwater (AFW) addition via the motor-driven AFW pumps. If the steam generator level decreases below the low-low level setpoint in two of the three steam generators or a loss of off-site power occurs, the turbine-driven AFW pump will also be started.

Amendment 63 Page 117 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

4. The reactor trip automatically trips the turbine and if off site power is available, the steam dump valves open permitting steam dump to the condenser. In the event of a coincident loss of off site power, the steam dump valves would automatically close to protect the condenser. The steam generator pressure would rapidly increase resulting in steam discharge to the atmosphere through the steam generator power operated relief valves (and safety valves if their setpoint is reached).

5. Following reactor trip and SI actuation, the continued action of the AFW supply and borated SI flow (supplied from the refueling water storage tank) provide a heat sink which absorbs some of the decay heat. This reduces the amount of steam bypass to the condenser or in the case of loss of off site power, steam relief to the atmosphere.

6. SI flow results in stabilization of the RCS pressure and pressurizer water level, and the RCS pressure trends toward an equilibrium value where the SI flow rate equals the break flow rate.

In the event of an SGTR, the plant operators must diagnose the SGTR and perform the required recovery actions to stabilize the plant and terminate the primary to secondary leakage. The operator actions for SGTR recovery are provided in the plant Emergency Operating Procedures. The major operator actions include identification and isolation of the ruptured steam generator, cooldown and depressurization of the RCS to restore inventory, and termination of SI to stop primary to secondary leakage. A steam generator tube rupture will cause no subsequent damage to the RCS or the reactor core. An orderly recovery from the accident can be completed even assuming simultaneous loss of offsite power. 6.3.3.6 Existing Criteria Used to Judge the Adequacy Of the ECCS. Criteria from 10CFR50.46

1. Peak clad temperature calculated shall not exceed 2200°F.

2. The calculated total oxidation of the clad shall nowhere exceed 0.17 times the total clad thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the clad with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the clad cylinders surrounding the fuel, excluding the clad around the plenum volume, were to react.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptable low value and decay heat shall be removed for the extended period of time required by long lived radioactivity remaining in the core.

In addition to and as an extension of the final acceptance criteria of 10 CFR 50 Appendix K, two accidents have more specific criteria as shown below. Amendment 63 Page 118 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 In the case of the inadvertent opening of a steam generator power operated relief or safety valve, an additional criteria for adequacy of the ECCS is: Assuming a stuck rod clustered control assembly, offsite power available, and a single failure in the engineered safety features, there will be no return to criticality after reactor trip for a steam release equivalent to the spurious opening with failure to close, of the larger of a single steam dump power operated relief, or safety valve. For a major secondary system pipe failure, the added criteria is: Assuming a stuck RCCA with or without offsite power, and assuming a single failure in the engineered safeguards, the core remains inplace and intact. 6.3.3.7 Use of Dual Function Components The ECCS contains components which have no other operating function as well as components which are shared with other systems. Components in each category are as follows:

1. Components of the ECCS which perform no other function are:

a. One accumulator for each loop which discharges borated water into its respective cold leg of the reactor coolant loop piping.

b. One boron injection tank.

c. Associated piping, valves and instrumentation.

2. Components which also have a normal operating function are as follows:

a. Residual heat removal pumps and the residual heat exchangers These components are normally used during the latter stages of normal reactor cooldown and when the reactor is held at cold shutdown for core decay heat removal or for flooding the refueling cavity. However, during all other plant operating periods, they are aligned to perform the low head injection function.

b. Centrifugal charging pumps - These pumps are normally aligned for charging service. As a part of the Chemical and Volume Control System, the normal operation of these pumps is discussed in Section 9.3.4. During safety injection conditions, however, they are aligned with the RWST to perform the high head injection function.

c. Refueling water storage tank (RWST) - This tank is used to fill the refueling cavity for refueling operations and to provide borated makeup to the spent fuel pools.

However, during all other plant operating periods it is aligned to the suction of the residual heat removal pumps. The charging pumps are automatically aligned to the suction of the refueling water storage tank upon receipt of the safety injection signals or volume control tank low level alarm. During normal operation they take suction from the volume control tank.

3. Positive Displacement Hydrostatic Test Pump - Normally this pump takes suctions from the RWST. It serves three functions, none of which are safety related. By temporary connections at the discharge of the pump to the Chemical and Volume Control System, Amendment 63 Page 119 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 this pump is used for hydrotesting the high pressure parts of the Reactor Coolant System. Permanent connections to the accumulators provide for using the hydrostatic test pump in supplying borated water to the accumulators. Also through the permanent connection, this pump provides a high pressure source for leak testing ECCS pressure isolation valves. A locked closed manual boundary isolation valve, 2CT V144SAB-1, separates the safety related RWST and the non-safety, non-seismically qualified hydrostatic test pump. Strict administrative controls are invoked by plant procedures whenever this valve is opened in modes when the RWST is required operable. An evaluation of all components required for operation of the ECCS demonstrates that either:

1. The component is not shared with other systems, or

2. If the component is shared with other systems, it is either aligned during normal plant operation to perform its accident function or if not aligned to its accident function, two valves in parallel are provided to align the system for injection. These valves are automatically actuated by the safety injection signal.

Table 6.3.2-7 indicates the alignment of components during normal operation, and the realignment required to perform the accident function. In all cases of component operation, safety injection has the priority usage such that an "S" signal will override all other signals and start or align systems for injection. 6.3.3.8 Limits on System Parameters The analyses show that the design basis performance characteristic of the ECCS is adequate to meet the requirements for core cooling following a LOCA with the minimum engineered safety features equipment operating. In order to ensure this capability in the event of the simultaneous failure of any single active component to operate, Technical Specifications are established for reactor operation. Normal operating status of ECCS components is given in Table 6.3.2-8. The ECCS components are available whenever the coolant energy is high and the reactor is critical. During low temperature physics tests, there is a negligible amount of stored energy in the coolant and low decay heat; therefore, an accident comparable in severity to accidents occurring at operating conditions is not possible and ECCS components are not required. The principal system parameters and the number of components which may be out of operation in test, quantities and concentrations of coolant available, and allowable time in a degraded status are specified in the Technical Specifications. If efforts to repair the faulty component are not successful the plant is placed into a lower operational status. 6.3.3.9 Time Sequence for the Operation of the ECCS Components The ECCS response times supported in the Chapter 15 non-LOCA safety analysis are 12 seconds if offsite power is assumed available and 27 seconds if offsite power is assumed to be lost. The respective response times account for trip logic delays, valve alignment and the time Amendment 63 Page 120 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 of the SI pumps to reach full speed. In addition, if offsite power is lost, a delay time for starting the diesel generators and loading the SI pumps is included. These times are verified by the procedures in the plant Technical Specifications. For the steam generator tube rupture (SGTR), immediate actuation of the ECCS on an SI signal is assumed in the analysis and is conservative for a SGTR. In large and small break LOCA analyses it is conservatively assumed that offsite power is lost. Following a loss of offsite power the diesel generators must activate automatically and then be loaded with the ECCS components sequentially. The current Small Break and Large Break LOCA analyses are conservatively based on a 29 second SI response time. Refer to section 6.3.3.4 for discussion of additional time delay for injection of borated water from RWST to suction of charging pumps. 6.3.4 TEST AND INSPECTIONS 6.3.4.1 ECCS Performance Tests 6.3.4.1.1 Preoperational Test Program at Ambient Conditions Preliminary operational testing of the ECCS is conducted during the hot functional testing of the RCS following flushing and hydrostatic testing, with the system cold and the reactor vessel head removed. Provision will be made for excess water to drain into the reactor cavity. The ECCS must be aligned for normal power operation with the boron injection tank filled with refueling water. Simultaneously, the safety injection block switch is reset and the breakers on the lines supplying offsite power are tripped manually so that operation of the standby diesel generators is tested in conjunction with the safety injection system. This test will provide information including the following facets: a) Satisfactory safety injection signal generation and transmission. b) Proper operation of the standby diesel generators, including sequential load pickup. c) Valve operating times. d) Pump starting times. e) Pump delivery rates at ECCS design flows (one point on the operating curve). Recirculation tests of the ECCS are performed under the requirements of Reg. Guide 1.79 with exceptions/clarifications as noted in Section 14.2.7(g). Testing of the containment recirculation sumps, to demonstrate vortex control and acceptable pressure drops across screening and suction lines and valves, is provided in Section 14.2.12.1.66. 6.3.4.1.2 Components a) Pumps - Separate flow tests of the pumps in the ECCS are conducted during the operational startup testing (with the reactor vessel head off) to check capability for Amendment 63 Page 121 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 sustained operation. The centrifugal charging, and residual heat removal pumps will discharge into the reactor vessel through the injection lines, the overflow from the reactor vessel passing into the reactor cavity. Each pump will be tested separately with water drawn from the refueling water storage tank. Data will be taken to determine pump head and flow at this time. Pumps will then be run on miniflow circuits and data taken to determine a second point on the head flow characteristic curve. b) Accumulators - Each accumulator is filled with water from the refueling water storage tank and pressurized with the motor operated valve on the discharge line closed. Then the valve is opened and the accumulator allowed to discharge into the reactor vessel as part of the operational startup testing with the reactor cold and the vessel head off. 6.3.4.2 Reliability Tests and Inspections 6.3.4.2.1 Description of tests planned Routine periodic testing of the ECCS components and all necessary support systems at power is planned. Valves which operate after a LOCA are operated through a complete cycle, and pumps are operated individually in this test on their miniflow lines except the charging pumps which are tested by their normal charging function. If such testing indicates a need for corrective maintenance, the redundancy of equipment in these systems permits such maintenance to be performed without shutting down or reducing load under certain conditions. These conditions include considerations such as: the period within which the component should be restored to service, and the capability of the remaining equipment to provide the minimum required level of performance during such a period. The operation of the remote stop valve and the check valve in each accumulator tank discharge line, may be tested by opening the remote test line valves just downstream of the stop valve and check valve respectively. Flow through the test line can be observed on instruments and the opening and closing of the discharge line stop valve could be sensed on this instrumentation if other methods of position indication were suspect. Where series pairs of check valves form the high pressure to low pressure isolation barrier between the RCS and safety injection system piping outside the Containment, periodic testing of these check valves must be performed to provide assurance that certain postulated failure modes will not result in a loss of coolant from the low pressure system outside Containment with a simultaneous loss of safety injection pumping capacity. The tests performed verify that each of the series check valves can independently sustain differential pressure across its disc, and also verify that the valve is in its closed position. The required periodic tests are to be performed after each refueling just prior to plant startup, after the RCS has been pressurized. To implement the periodic component testing requirements, Technical Specifications have been established. During periodic system testing, a visual inspection of pump seals, valve packings, flanged connections, and relief valves is made to detect leakage. Inservice inspection provides further confirmation that no significant deterioration is occurring in the ECCS fluid boundary. Design measures have been taken to assure that the following testing can be performed: a) Active components may be tested periodically for operability (e.g., pumps on miniflow, certain valves, etc.). Amendment 63 Page 122 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 b) An integrated system actuation test can be performed when the plant is cooled down and the RHRS is in operation. The ECCS will be arranged so that no flow will be introduced into the RCS for this test. Details of the testing of the sensors and logic circuits associated with the generation of a safety injection signal together with the application of this signal to the operation of each active component are given in Section 7.2. c) A coordinated full flow test of the ECCS operational sequence can be performed at refuelings. However, normally only the high and low head safety injection lines will be tested for delivery of coolant to the vessel. The design features which further assure this test capability are specifically: a) Power sources are provided to permit individual actuation of each active safety related component of the ECCS. b) The residual heat removal pumps are used every time the RHRS is put into operation. They can also be tested periodically when the plant is at power either by using the miniflow recirculation lines or a full flow recirculation path. c) The centrifugal charging pumps are either normally in use for charging service or can be tested periodically on miniflow to ensure operability. d) Remote operated valves can be exercised during routine plant maintenance and normal operation. e) Redundant level and pressure instrumentation is provided for each accumulator tank, for continuous monitoring of these parameters during plant operation. f) Flow from each accumulator tank can be directed through a test line in order to determine valve operability. The test line can be used, when the RCS is pressurized, to ascertain backleakage through each of the accumulator check valves individually. g) A flow indicator is provided in the common charging pump, the SI BIT flowpath, the alternate SI flowpath, and each residual heat removal pump headers. Pressure instrumentation is also provided in these lines. h) An integrated system test can be performed when the plant is cooled down and the RHRS is in operation. This test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry including diesel generator starting and the automatic loading of ECCS components on the diesel generators (by simultaneously simulating a loss of offsite power to the emergency electrical buses). A closeout inspection procedure is established to perform an inspection of the containment, in particular the containment sump area, to identify any materials having the potential to become debris capable of blocking the containment sump when required for recirculation. This inspection will be performed at the end of each shutdown as soon as practical before containment isolation. A procedure for inspection of the structural components of the containment recirculation sump will be established in accordance with Regulatory Guide 1.82. Amendment 63 Page 123 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 See the Technical Specifications, and Section 3.9.6 for the selection of test frequency, acceptability of testing, and measured parameters of pumps and valves. The inservice inspection program described in Section 6.6 is also included in the Technical Specifications. ECCS components and systems are designed to meet the intent of the ASME Code, Section XI for inservice inspection. 6.3.5 INSTRUMENTATION REQUIREMENTS Instrumentation and associated analog and logic channels employed for initiation of ECCS operation is discussed in Section 7.3. This section describes the instrumentation employed for monitoring ECCS components during normal plant operation and also ECCS post-accident operation. All alarms are annunciated in the Control Room. 6.3.5.1 Temperature Indication 6.3.5.1.1 Deleted by Amendment No. 26. The fluid temperature at both the inlet and the outlet of each residual heat exchanger is recorded in the Control Room. The outlet temperature of each residual heat exchanger is also indicated locally. 6.3.5.2 Pressure Instrumentation 6.3.5.2.1 Deleted by Amendment No. 26. 6.3.5.2.2 Charging Pump Inlet, Discharge Pressure There is local pressure indication at the suction and discharge of each centrifugal charging pump. 6.3.5.2.3 Accumulator Pressure Duplicate pressure channels are installed on each accumulator. Pressure indication in the Control Room and high and low pressure alarms are provided by each channel. 6.3.5.2.4 Test Line Pressure A local pressure indicator used to check for proper seating of the accumulator check valves between the injection lines and the RCS is installed on the leakage test line. 6.3.5.2.5 Residual Heat Removal Pump Suction Pressure Local pressure indication is provided at the inlet to each residual heat removal pump. 6.3.5.2.6 Residual Heat Removal Pump Discharge Pressure Residual heat removal discharge pressure for each pump is indicated locally and remotely in the Control Room. A high pressure alarm is actuated by each channel. Amendment 63 Page 124 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.3.5.3 Flow Indication 6.3.5.3.1 Deleted by Amendment No. 26. 6.3.5.3.2 Charging Pump Injection Flow Injection flow to the reactor cold legs is indicated in the Control Room. These flow indicators are non-safety and are operable with and without offsite power. They are powered by the uninterruptable power supply and are backed up by a DC battery, which is connectable to the diesel generator. 6.3.5.3.3 Test Line Flow Local indication of the leakage test line flow is provided to check for proper seating of the accumulator check valves between the injection lines and the RCS. 6.3.5.3.4 Residual Heat Removal Pump Hot Leg Injection Flow Indication of the flow recirculated to the RCS hot legs by the residual heat removal pumps is provided on the main control board. 6.3.5.3.5 Residual Heat Removal Pump Cold Leg Injection Flow The flow from each residual heat removal subsystem to the RCS cold legs is indicated in the Control Room. These instruments also control the residual heat removal bypass valves, maintaining constant return flow to the RCS during normal cooldown. 6.3.5.3.6 Residual Heat Removal Pump Minimum Flow The flowmeter installed in each residual heat removal pump discharge header provides control for the valve located in the pump minimum flow line. These flow indicators are non-safety and are operable with and without off site power. They are powered by the uninterruptable power supply and are backed up by a DC battery, which is connectable to the diesel generator. 6.3.5.4 Level Indication 6.3.5.4.1 Refueling water storage tank level There are four safety related locally mounted level transmitters provided for the RWST. These four level transmitters are used to provide inputs to the RWST low level protection logic. The RWST low level protection logic produces an actuation signal to automatically open the containment recirculation sump isolation valves when two of four RWST level channel bistables receive an RWST level signal lower than a predetermined low-low level setpoint in conjunction with an "S" signal. There are level alarms for high, low, low-low, 2 out of 4 low-low, and empty levels. The high level alarm is provided to warn of possible overflow of the RWST. The low level alarm is provided to assure that a sufficient volume of water is always available in the RWST in conformance with the Technical Specifications. The low-low level alarms alerts the operator to realign the ECCS from the injection to the recirculation mode following an accident Amendment 63 Page 125 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 and automatically opens the containment sump isolation valves. The empty alarm indicates that the useable volume of the RWST has been exhausted. Each channel also provides level indication in the Control Room. A local level indication is also provided. Two of the four channels provide input signals to the recorder in the Control Room. 6.3.5.4.2 Accumulator water level Duplicate water level channels are provided for each accumulator. Both channels provide indication in the Control Room and actuate high and low water level alarms. 6.3.5.4.3 Deleted by Amendment No. 26 6.3.5.5 Valve Position Indication Valve position for those valves provided with engineering safety features monitoring lights (see Section 7.5) are indicated on the main control board by an on/off system, i.e., should the valve be out of position, the associated light will be different (on or off) from the other valves with which it is grouped and, thus, provide a highly visible indication to the operator. Valve position for remote manual ECCS valves is also indicated on the main control board by red and green indicator lights. Certain "critical" valves also have an annunciator to indicate and alarm in the Control Room a change to the wrong position. 6.3.5.5.1 Accumulator isolation valve position indication The accumulator motor operated isolation valves are provided with red (open) and green (closed) position indicating lights located at the control module for each valve. These lights are powered by separate Class IE, 120 VAC supply and actuated via valve motor operator limit switches, in order to maintain valve position indication during normal operation when valve power is locked out. A white indicating light is also provided at the control module, powered by the valve control power, to indicate a thermal overload condition at the MCC breaker. Redundant red and green position indicating lights for these valves are also provided at the MCB, via separate Class IE stem mounted limit switches and powered by separate 125 volt DC power. In addition, a white monitor light is provided for each valve to indicate that valve is not in fully open position. These lights are combined to indicate the proper valve positions for the safeguard operation. The total array of lights is powered from a separate Class IE, 120 Volt AC source and actuated via valve motor operated limit switches. For description of these monitor lights, refer to FSAR subsection 7.5.1.10.3. An alarm annunciator point is activated by both a valve motor-operator limit switch and by a valve position limit switch activated by stem travel whenever an accumulator valve is not fully open for any reason with the system at pressure (the pressure at which the safety injection block is unblocked is approximately 1900 psig). A separate annunciator point is used for each accumulator valve. This alarm will be recycled at approximately 1 hour intervals to continuously remind the operator of the improper valve lineup, until corrective action is taken. Amendment 63 Page 126 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

REFERENCES:

SECTION 6.3 6.3.1-1 Carolina Power and Light Harris Nuclear Plant Steam Generator Replacement/Uprating Analysis and Licensing Project NSSS Engineering Report, WCAP-14778, Revision 1, September, 2000. 6.3.2-1 Geets, J. M., "MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop PWR System," WCAP-7909, June, 1972. 6.3.2-2 Deleted by Amendment No. 55 6.3.2-3 Letter from CP&L's W. R. Robinson to the NRC, dated July 15, 1994, NLS-94-055. 6.3.2-4 Letter from CP&L's W. R. Robinson to the NRC, dated February 28, 1995, HNP 027. 6.3.2-5 Letter from CP&L's W. R. Robinson to the NRC, dated December 7, 1995, HNP 077. 6.3.3-1 "VCT to RWST Alignment for Steam Line Break," FCQL-465, April 13, 1987. 6.4 HABITABILITY SYSTEMS The Control Room Habitability Systems include equipment, supplies and procedures which give assurance that the control room operators can remain in the Control Room and take effective actions to operate the nuclear power plant safely under normal conditions and maintain the facility in a safe condition following a postulated accident as required by the General Design Criterion 19 contained in Appendix A to 10 CFR 50. The habitability systems and provisions include: a) Control Room Air Conditioning System (which includes the Emergency Filtration System). b) Radiation protection c) Food and water storage d) Kitchen and sanitary facilities e) Breathing apparatus The above systems and provisions provide adequate operator protection under normal and emergency operating conditions (including the design basis loss-of-coolant accident) and postulated release of toxic gases and smoke. 6.4.1 DESIGN BASIS The habitability systems for the control room include shielding, air handling and filtration systems, temperature control, dehumidifiers, instrumentation to protect against airborne Amendment 63 Page 127 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 radioactivity, air breathing apparatus, sufficient storage for food and water, and other provisions for extended occupancy by control room personnel, including kitchen and sanitary facilities. The bases upon which the functional design of these systems and provisions are designed include the following: Control Room Envelope: The control room envelope includes, in addition to the Control Room, the following auxiliary spaces: a) Office areas b) Relay and termination cabinet rooms c) Kitchen and sanitary facilities d) Component cooling water surge tank room Period of Habitability The period of habitability for control room operators is based on the habitability systems' capability to provide protection from the introduction into the control room envelope of airborne contaminants that present an immediate danger to life or health. The most severe hazards are posed by airborne radioactivity. After the detection of airborne radioactivity the control room envelope will be pressurized and all air will be filtered through charcoal adsorbers. This system will ensure that the control room operators will not receive doses of radiation in excess of the limits specified in GDC 19 of Appendix A to 10 CFR 50 during the time required for the safe shutdown of the plant. Capacity The Control Room has been designed (1) to allow continuous occupancy of five persons for a seven-day period following a design basis accident and (2) for replacement of the crews following the seven days. This includes sufficient food, water, medical supplies and sanitary facilities. Food, Water, Medical Supplies, and Sanitary Facilities For habitability of the Control Room during certain emergencies, a seven day supply of food and potable water is provided within the control room area. Basic medical supplies, kitchen and sanitary facilities are provided within the control room envelope. Radiation Protection The Control Room envelope has been designed to ensure continuous occupancy during normal operation and extended occupancy throughout the duration of any one of the following postulated design basis accidents: Amendment 63 Page 128 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 a) Loss-of-coolant accident (LOCA) b) Fuel handling accident c) Radioactive releases due to radwaste system failure The radiation exposures shall not exceed 5 rem TEDE for the duration of any of the above accidents. As documented in the SHNPP SER (NUREG-1038 Supplement 2), the postulated design basis LOCA event has been established as the most limiting event for demonstrating compliance with the Control Room Habitability Dose Criteria. Dose to the Control Room personnel resulting from a LOCA is discussed in Section 15.6.5.4.4. Respiratory, Eye, and Skin Protection for Emergencies An adequate number of respirators is provided in the Control Room for emergency use. Habitability System Operation During Emergencies The Control Room Air Conditioning System is safety related and designated as Safety Class 3 and Seismic Category I. The system is capable of performing its functions assuming an active component single failure. The air conditioning system will not promote the propagation of smoke and fire from other areas in the Reactor Auxiliary Building to the control room envelope. Refer to Section 9.5.1 for a discussion of fire protection criteria for the Control Room. Provisions have been made for control room smoke purge operation, as described in Section 9.4.1.2.3. The system has been designed to maintain the ambient temperature in the Control Room at 75 F DB and 50 percent (max.) relative humidity during normal conditions and a design basis accident. During a postulated LOCA, the Control Room is pressurized to 1/8 in. wg. by the capability of introducing a maximum of 400 cfm outside air into the Control Room which will keep the carbon dioxide and oxygen concentrations within safe levels. Calculations of CO2 and O2 concentrations within the Control Room consider that the concentrations of these gases are homogenous within the control room envelope, excluding the air above the hung ceiling. Design maximum concentration of carbon dioxide is taken as 1.0 percent. Design minimum concentration of oxygen will be taken as 17 percent. The Control Room has been designed to protect the control room operators from all design basis natural phenomena and design basis accidents. Emergency Monitors and Control Equipment Provisions have been made to detect radioactivity and smoke in the Control Room air intake. Following detection, the control room envelope is automatically isolated. Sensitivities of the detectors and isolation time including delays in the control circuits are designed to meet the requirements of GDC 19. Amendment 63 Page 129 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.4.2 SYSTEM DESIGN 6.4.2.1 Control Room Envelope The control room envelope includes those areas listed in Section 6.4.1. During an emergency, the areas which the control room operator could require access to are the Control Room, office areas, kitchen and sanitary facilities and control room emergency air intake valves located in the relay and termination and cabinet rooms. 6.4.2.2 Ventilation System Design The control room envelope air conditioning process includes an environmental control operation and an emergency air cleanup operation. The environmental control operation is the primary function of the air conditioning system and it is accomplished by the use of redundant air conditioning trains. The Control Room will be isolated upon receipt of a Safety Injection Signal, following a detection of radioactivity or smoke at the Normal Outside Air Intake (OAI), or following a detection of radioactivity at the Emergency Outside Air Intakes. A loss of power to any of the OAI Radiation Monitors will also result in a Control Room Isolation. Redundant, motorized butterfly valves are provided in the control room envelope outside air intake and exhaust ducts for automatic isolation of the system from the surrounding atmosphere. Redundant trains of the Control Room Air Conditioning System are provided for the system to fulfill its essential functions. The redundant filtration train is located in a separate equipment room. The system is located within the Reactor Auxiliary Building which is designed to withstand effects of the safe-shutdown earthquake and other design basis natural phenomena. To assure continued operation following a design basis accident, the Control Room Air Conditioning System is designed to Seismic Category I requirements. This includes equipment and ductwork up to and including the connection into the Control Room (except portions of the normal exhaust and smoke purge fans). The air intakes and exhaust of the Control Room Areas Ventilation System are tornado and missile protected. Active system components meet the single failure criteria as described in IEEE 279-71. Refer to Table 9.4.1-4 for a single failure analysis of the Control Room Air Conditioning System. The redundant air conditioning units are served by separate Essential Services Chilled Water Systems so that loss of one train of the chilled water systems will not affect the ability of the system to control the thermal environment in the control room envelope. The Control Room Area Ventilation System including equipment, ductwork, valves, and air flows for both normal and emergency modes is discussed in detail in Section 9.4.1. Design data for principal components of the Control Room Area Ventilation System are presented in Table 9.4.1-1. The airflow diagram for the Control Room Area Ventilation System is shown on Figure 9.4.1-1. The Emergency Filtration System is discussed in Section 9.4.1.2. The operational status of valves, fans and corresponding airflow rates for the Control Room Air Conditioning System and Emergency Filtration System are presented in Table 9.4.1-2. The design data is presented in Table 9.4.1-1. Amendment 63 Page 130 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The degree of compliance of the Emergency Filtration System with the requirements of Regulatory Guide 1.52 is discussed in Section 6.5.1. The layout drawings of the Control Room showing doors, corridors, stairwells, shield walls, and the placement and type of equipment within the Control Room are shown on Figure 1.2.2-35. Elevation and plan views showing building dimensions and the location of control room air intakes are also presented on Figure 1.2.2-35. Under a completely isolated Control Room, occupied with up to ten people, the CO2 concentration would build up from zero to one percent in 71 hours. This buildup time is based upon a net control room envelope of 0.71 x 105 ft3) which includes space above the egg crate hung ceiling and a breathing rate of 30 ft3)/hr to generate 1 ft3)/hr CO2 per person. Considering there are no postulated design conditions which would require that the control room envelope be isolated for an extended period of time, 71 hours provides more than adequate time of the operator actions required to reestablish control room ventilation. A ventilation rate of 3.4 cfm fresh air per person will maintain the carbon monoxide level in the control room below 0.5 percent. Since the emergency pressurization mode of the Control Room Ventilation System permits the continuous introduction of up to 400 cfm (outside air from the uncontaminated air intake) through the control room emergency air cleaning unit, there will be no excessive buildup of CO2 in the control room. The actual pressurization flow rate will be determined by testing to maintain a positive pressure differential of 1/8 inch of water gauge. A ventilation rate of 0.5 cfm fresh air per person will maintain the oxygen level in the Control Room at 17 percent, min. The design ventilation rate capability of up to 400 cfm is therefore adequate. Smoke purge fans are provided to expedite firefighting efforts. Refer to Section 9.4.1.2.3 for a detailed discussion of the smoke purge operation. Adequate bottled air capacity (of at least six hours) is readily available onsite for the five Control Room occupants to assure that sufficient time is available to locate and transport bottled air from offsite locations. This offsite supply is capable of delivering several hundred hours of bottled air to the members of the emergency crew. 6.4.2.3 Leak Tightness The control room envelope is pressurized to 1/8 in. of water gauge differential pressure relative to the adjacent areas at all times during normal plant operation and outside air is continuously introduced to the control room envelope. During a postulated LOCA, a maximum rate of 400 cfm may be required in order to maintain 1/8 inch of water gauge. The control room is automatically isolated following a design basis radionuclide accident. In case of a radionuclide accident, the operator will re-pressurize the control room by drawing in filtered outside air through one of two emergency air intakes. The 400 cfm pressurization flow rate is approximately 0.34 volume change per hour. All openings to the Control Room have a low leakage design. This includes doors, valves, penetrations and walls. The control room leakage rate estimate through valves, doors, penetrations and walls is shown in Tables 6.4.2-1 and 6.4.2-2. The estimate is based on AEC R&D Report NAA-SR-101000. Amendment 63 Page 131 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 A maximum of 400 cfm makeup air will not make the overall doses to the control room operator exceed the radiation dose limit of General Design Criterion 19 of Appendix A to 10 CFR 50 under design basis accidents. An acceptance test is performed at startup to verify that the control room leakage rate is less than the value assumed in the analysis. 6.4.2.4 Interaction with Other Zones and Pressure-Containing Equipment The following provisions are taken into consideration in the Control Room Area Ventilation System design to assure that there are no toxic or radioactive gases and other hazardous material that would transfer into the Control Room: a) The control room envelope is pressurized to 1/8 in. w.g. relative to the adjacent areas. b) The Control Room Area Ventilation system is independent and completely separated from other adjacent ventilation zones. c) There is no other HVAC equipment within the Control Room envelope that serves other ventilation zones. d) All doors, duct and cable penetrations are of low leakage design. e) On a Control Room Isolation Signal, initiated on either a Safety Injection Signal or following a detection of radioactivity or smoke at the Outside Air Intakes, the RAB Normal Ventilation System is secured, and the RAB Emergency Exhaust System (RABEES) is started. The RAB Normal Ventilation System must be secured to preclude the possibility of postulated system failures from impacting the ability of the Control Room Envelope (CRE) to maintain a positive pressure of 1/8 INWG relative to adjacent areas. When the RAB Normal Ventilation System is secured, the RAB Emergency Exhaust System is initiated to maintain the potentially contaminated areas of the RAB at sub-atmospheric pressure in an effort to limit outleakage and to remove radon gas from the RAB. 6.4.2.5 Shielding Design The Control Room envelope is shielded against direct sources of radiation which are present during normal operating conditions and following a postulated accident. There are no significant sources of direct or streaming radiation near the control room envelope during normal operating conditions. The shielding walls and floor provided for the accident conditions are more than sufficient to limit the dose rates to less than 0.25 mr/hr. in the Control Room during normal operation. Refer to Section 12.3.2.14 for a discussion of the control room shielding design. 6.4.3 SYSTEM OPERATIONAL PROCEDURES The normal operation of the Control Room Areas Ventilation System is discussed in detail in Section 9.4.1.2.1; the post-accident operation and smoke purge operation are discussed in detail in Sections 9.4.1.2.2 and 9.4.1.2.3. Amendment 63 Page 132 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Upon failure of the normal power supply, all electrically operated safety related components of the system will be automatically switched to their respective emergency power source. Upon receipt of a Safety Injection Actuation Signal (SIAS) or a high radiation signal from the radiation monitor located within each air intake (one Normal OAI and two Emergency OAIs), all outside air intakes and exhausts will be automatically isolated, and the Emergency Filtration System will be put into operation. In addition, the RAB Normal Ventilation System will be secured, and the RAB Emergency Exhaust System (RABEES) will be started. The RAB Normal Ventilation System must be secured to preclude the possibility of postulated system failures from impacting the ability of the Control Room Envelope (CRE) to maintain a positive pressure of 1/8 INWG relative to adjacent areas. When the RAB Normal Ventilation System is secured, the RAB Emergency Exhaust System is initiated to maintain the potentially contaminated areas of the RAB at sub-atmospheric pressure in an effort to limit outleakage and to remove radon gas from the RAB. After a high radiation signal has automatically isolated the Control Room Air Conditioning System (CRACS) the operator will monitor the CRACS air intake radiation detectors and select the emergency air intake from which to draw the least radioactive make-up air. This selection will be based on the readings of the radiation detectors located in the redundant air intakes on either side of the Reactor Auxiliary Building. The control room operator will manually open the selected closed air intake allowing up to a maximum of 400 cfm of the outside air into the control room envelope. To maintain a positive pressure of 1/8 inch water gauge, a make-up air rate within the range of 71 to 132 cfm is required. The actual control room boundary leakage shall be determined by testing and is expected to be well below the 400 cfm make-up air assumed in the radiological analysis. 6.4.4 DESIGN EVALUATION 6.4.4.1 Radiological Protection The evaluation of the radiological exposure to the control room operators is presented in the control room accident dose analysis given in Chapter 15. Section 15.6.5.4.4 shows the doses following the design basis accident (LOCA) and demonstrates compliance with GDC 19. 6.4.4.2 Toxic Gas Protection Accidents involving off-site hazardous chemical releases are discussed in Section 2.2.3. A summary analysis of off-site and on-site toxic chemical hazards that may impact control room habitability, performed in accordance with Regulatory Guide 1.78, is contained in Calculation 9-CRH. The analysis found no impact on control room habitability from toxic chemical sources. The leakage rate of the control room HVAC valves are given in Tables 6.4.2-1 and 6.4.2-2. The valves that isolate the control room outside air intakes and exhausts are designed with a 15 second closure time. The Control Room Area Ventilation System is discussed in detail in Section 9.4.1. Toxic chemicals stored onsite are listed in Table 6.4.4 1. Sulfuric acid and sodium hydroxide do not present dangers to control room habitability because they are non-volatile. Hydrogen and nitrogen are simple asphyxiants and would pose a threat to control habitability only if they were to appreciably reduce the oxygen concentration in the control room, while carbon dioxide levels Amendment 63 Page 133 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 of up to 1% can be tolerated for a limited period of time. Since an analysis based on Regulatory Guide 1.78 shows that the concentration of these gases in the Control Room will be below one percent under the condition of accidental release, these gases have no potential for adversely affecting control room habitability (Calculation CPL-X-5). Refer to Section 1.8 for the SHNPP position on Regulatory Guide 1.78. Chlorine Detection System Since Shearon Harris no longer stores chlorine in large quantities on site, the chlorine detection system is no longer required at the Shearon Harris Nuclear Power Plant. The chlorine leak detectors, both local and remote, have been deactivated and the equipment is abandoned in place. 6.4.5 TESTING AND INSPECTIONS The major items of equipment required to maintain the habitability of the Control Room are the emergency HEPA/charcoal filter trains, mechanical refrigeration water chillers, fans and fan coil units, and chilled water pumps. These units are thoroughly tested in a program consisting of the following: a) Shop component qualification test. b) Field preoperational tests. These systems and their components, which maintain Control Room habitability, are subjected to documented preoperational testing and in-service surveillance to ensure continued integrity. Testing and inspection is also discussed in Sections 6.6, 9.4.1.4, and 14.2.12. Pump and valve testing is delineated in Section 3.9.6. Tests are conducted to verify the following for both normal and emergency conditions. a) System integrity and leaktightness. b) Inplace testing of emergency filter trains to establish leaktightness and removal efficiency of the high-efficiency particulate air and charcoal filters. c) Proper functioning of system components and control devices. d) Proper electrical and control wiring. e) System balance for design airflow, water flow and operational pressures. 6.4.5.1 Emergency HEPA/Charcoal Filter Trains Initial performance verification and periodic surveillance tests are conducted to ensure operability and performance of both emergency HEPA/charcoal filter systems. Components in these filter systems have been designed to, and are tested in accordance with, the codes and requirements cited within Regulatory Guide 1.52 (see Section 1.8), with the exceptions listed in Table 6.5.1-2. Amendment 63 Page 134 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.4.5.2 Water Chillers During shop testing the water chiller impellers are subjected to an overspeed test and dynamic balancing. This overspeed test is in excess of 125 percent of the impeller operating speed. The rotor part of the compressor drive motor is dynamically balanced. Preoperational testing in the field is discussed in Section 14.2. Inservice inspection on the safety Class 3 components of the chillers will be performed in accordance with Section 6.6. 6.4.5.3 Fan or Fan Coil Units Cooling coils are hydrostatically pressurized and leak tested. A performance test or manufacturer's certified rating in accordance with Air Moving and Conditioning Association (AMCA) or Air Conditioning and Refrigeration Institute (ARI) standards is required. Preoperational testing is delineated in Section 14.2.12. Operating fan or fan coil units will be checked periodically for unusual vibration. 6.4.5.4 Pumps Each chilled water pump is tested to verify the pump performance characteristics. Preoperational testing shall be delineated in Section 14.2.12. Operating pumps will be observed for leaks, suction and discharge pressures, and flowrates. The pumps will be rotated periodically. 6.4.5.5 Considerations Leading to the Selected Test Frequency The frequency of performing the surveillance tests is determined by the following considerations: a) Preoperational test data. b) Normal control room area ventilation system performance data. c) Continuous monitoring of the Control Room Area Ventilation System, which gives an indication of building tightness and system performance. 6.4.6 INSTRUMENTATION REQUIREMENT The control room air conditioning system instrumentation is designed to assist the operator to monitor habitability conditions in the Control Room. System instrumentation, control switches and alarms on the Main Control Board provide the operator with the information concerning the status of the system and enables the operator to take the proper course of action. System instrumentation and control switches, with the exception of those for the emergency filtration trains and emergency intake valves, are located on the auxiliary control panel/auxiliary transfer panels for use when the Control Room is evacuated. The radiation monitors are provided with adjustable setpoints and associated alarms such that the operator is notified if any predetermined increase in radiation levels occurs at the air intakes. In the event that the high radiation setpoint is reached, the normal outside air and emergency air intakes and exhausts are automatically isolated. The operator will override the isolation signal Amendment 63 Page 135 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 and open the emergency air intake that has the least radioactive level to pressurize the control room envelope. The radiation detectors and displays meet the requirements for the post-accident monitoring systems including IEEE-279, as discussed in Sections 7.3, 11.5 and 12.5. Redundant flow indicators are provided for the emergency air intake flow to show the operator the make-up air flowrate required for pressurization. Smoke detectors are provided at the normal outside air intake and throughout the control room area. In the event of a smoke alarm in the control room area, the operator manually initiates the smoke purge fans which convert the Control Room HVAC System to a "once through" system. If smoke is detected at the normal outside air intake, the Control Room isolation signal is activated as described in Section 7.3.1. The following Control Room Air Conditioning System parameters are monitored and alarmed when abnormal conditions exist: a) Normal outside and emergency air intake radiation level b) Normal outside air intake smoke concentration c) Control room area smoke concentration d) Control room air handling unit prefilter differential pressure e) Control room air handling unit inlet temperature f) Control room air handling unit entering heating and leaving cooling coil temperature g) Control room air handling unit fan failure (low flow) h) Control room exhaust fan failure (low flow) i) Control room purge fan failure (low flow) j) Control room pressure (relative to the adjacent area) k) Emergency air intake fan failure (low flow) l) Emergency air filtration train status (diff. press.) m) Emergency air filtration train humidity n) Emergency air filtration train inlet temperature o) Emergency air filtration train charcoal filter status p) Control room isolation train actuation status Amendment 63 Page 136 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Refer to Sections 7.3 and 7.5 for a more detailed discussion of the control room air conditioning system instrumentation and controls. 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5.1 ENGINEERED SAFETY FEATURE (ESF) FILTER SYSTEMS All filters that are required to perform a safety related function following a DBA are discussed in this section. The Engineered Safety Filter Systems include the following:

1. FHB Emergency Exhaust System which is discussed in Sections 6.5.1.1.1 and 6.5.1.2.1.

2. RAB Emergency Exhaust System which is discussed in Sections 6.5.1.1.2 and 6.5.1.2.2.

3. Control Room Emergency Filtration System which is discussed in Section 9.4.1.

6.5.1.1 Design Bases 6.5.1.1.1 FHB emergency exhaust system The FHB Emergency Exhaust System is designed to the following bases:

1. The system is designed to mitigate the consequences of the fuel handling accident by removing the airborne radioactivity from the FHB exhaust air prior to releasing to the atmosphere.

2. The system is designed to maintain the site boundary dose within the guidelines of 10 CFR 50.67 following a fuel handling accident. The fuel handling accident analysis, in accordance with the guidance given in Regulatory Guide 1.183, is presented in Section 15.7.

3. The components of the system are designed and sized in accordance with Regulatory Guide 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2.

4. The fission product removal capacity of the filters is based on the requirements of Regulatory Guide 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2.

5. The system is designed to satisfy all applicable requirements of GDC 61 of 10CFR50, Appendix A.

6. The system establishes and maintains the operating floor of the FHB under negative pressure following a fuel handling accident to prevent unfiltered outleakage of airborne radioactive materials. The FHB will only be held under negative pressure following an event involving the release of radioactivity in the FHB atmosphere.

7. The system is designed to withstand the SSE without loss of function.

8. Any single active failure in the FHB Emergency Exhaust System will not impair the ability of the system to comply with design bases 1 to 7 above.

Amendment 63 Page 137 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

9. Components and piping or ducting have sufficient physical separation or barriers to protect essential portions of the system from missiles and pipe whip.

10. Failures of non-Seismic Category I equipment or components will not affect the FHB Emergency Exhaust System.

6.5.1.1.2 RAB emergency exhaust system The RAB Emergency Exhaust System serves to limit the post-accident radiological releases from selected potentially contaminated portions of the RAB. These areas include the charging pump, RHR heat exchanger, containment spray and RHR pump room, mechanical, electrical and H&V rooms and mechanical, electrical and H&V penetration areas. Since leakage in these areas following a SIAS is a potential source of additional offsite dose, the RAB Emergency Exhaust System is provided to ensure that such airborne leakage is filtered prior to release to the environment. Portions of the Post-accident ECCS Recirculation flow path are outside of the RAB Emergency Exhaust System boundary. These areas include the mezzanine above the CSIP rooms and the CVCS filter area and valve galleries. Other areas affected or potentially affected by the ECCS recirculation flow path pressure boundary include various heat exchanger rooms and valve galleries. Postulated leakage from components in these areas (valves, strainers, filters) is not filtered and will be limited to 2 gallons per hour. Radiological consequences of leakage from ECCS is discussed in 15.6.5.4. The RAB Emergency Exhaust System will meet the following design bases:

1. The system is designed to maintain the post-accident radiological releases within the guidelines of 10CFR50.67, if a postulated leak occurs in the containment sump water recirculation system. The guidance provided in SRP 15.6.5, Appendix B, has been followed in assessing the offsite doses.

2. The system is designed to satisfy all applicable requirements of GDC 61 of 10CFR50, Appendix A.

3. The fission product removal capacity of the filters is based on the requirements of Regulatory Guide 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2.

4. The system establishes and maintains selected potentially contaminated areas of the RAB below atmospheric pressure following a SIAS, minimizing unfiltered outleakage of airborne radioactive materials.

5. The system is designed with sufficient redundancy to meet single active failure criteria.

6. The system is designed to withstand the SSE without loss of function.

7. The components of the system are designed and sized in accordance with Regulatory Guide 1.52, Revision 2 with the exceptions listed in Table 6.5.1-2.

Amendment 63 Page 138 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.5.1.1.3 Control room emergency filtration system Refer to Section 9.4.1 for a discussion of the Control Room Emergency Filtration System. 6.5.1.2 System Design 6.5.1.2.1 FHB emergency exhaust system The FHB Emergency Exhaust System is shown on Figure 9.4.2-1 and consists of two 100 percent capacity redundant fan and filter subsystems. Each of the two subsystem filter trains includes a manual locked open inlet butterfly valve, demister, electric heating coil, medium efficiency pre-filter, HEPA pre-filter, charcoal adsorber, HEPA after-filter, and decay heat cooling air connection. System component design data are shown in Table 6.5.1-1. Connected to each subsystem outlet is a centrifugal fan with a motor operated butterfly valve on its inlet and a back draft damper on its outlet to prevent reverse airflow through the inactive fan. The fan is furnished with variable inlet vanes and an air flow monitor to control and measure air flow. Interconnecting duct between the two emergency air cleaning units was originally provided, as shown in the system flow diagram, to allow one air cleaning unit to draw a small quantity of bleed air through the second inactive filtration train for decay heat cooling. However, the actual temperature increase of the carbon in the shutdown unit is calculated to be well below minimum auto-ignition or desorption temperatures. Therefore, the interconnecting duct is not needed and is blanked off at each unit. Following a fuel handling accident radioactivity released from fuel rods will be detected by the radiation monitors located around the fuel pools. These radiation monitors will then signal the switchover from the normal to the emergency ventilation and filtration system. The switchover time is 30 seconds for the emergency ventilation and filtration system to become fully operational. The isolation of the normal ventilation system is accomplished in 10 seconds. Either train may then be manually de-energized from the Control Room and placed on standby. Negative pressure is established at 1/8 in. wg. by continuously exhausting air from the operating floor. Pressure is then controlled by the Airflow Control System which adjusts the variable inlet vanes of the exhaust fans. Operating procedures ensure that no irradiated fuel (outside of sealed casks) will be handled or transported inside the FHB unless the operating floor hatch to the unloading area is in place. See Section 9.1.4.2. System design compliance with Regulatory Guide 1.52, Revision 2, is discussed in Table 6.5.1-2. The total travel time for radioactive gases to travel from the spent fuel pool surface to the isolation damper was conservatively calculated to be 29.95 seconds; however, the closure time of the normal ventilation isolation damper is 10 seconds. Thus no radioactive gases are released through the normal ventilation pathway. Amendment 63 Page 139 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 To initiate operation of the emergency ventilation system and terminate normal ventilation system operation, radiation monitors are provided at appropriate locations as shown in Figure 12.3.2-9 (four sets of three for safety function). The radiation detectors are located on the FHB walls and are extended low range GM tube detectors as described in Section 11.5.2.5.2, monitoring the air volume over the fuel pools. In the event of a fuel handling accident the gaseous radioactive material that is assumed to be released into the fuel pool will rise to the pool surface and be swept up into the FHB ventilation system. The radiation (gamma) that would emanate from the gaseous cloud of radioactive material in the airflow would cause alarms when the exposure rate exceeded preset limits. The detectors high radiation alarm will actuate switchover from normal FHB ventilation to emergency ventilation system operation. The radiation detectors are described in FSAR Section 11.5.2.5.2 and 12.3.4.1.8.3. The monitor's range is 1 x 10-2 through 1 x 103 mr/hr with the high alarm set-point at 1 x 102 mr/hr. The capability of the GM tube detectors was based on assuming the activity releases from the accident discussed in Section 15.7.4. This assumed source provides an exposure rate for the GM tube detectors to monitor. The time required for the accident released activity to provide an exposure rate to exceed the high alarm set-point for the monitors and initiate switchover is such that any doses will be within required limits as discussed in Section 15.7.4. The FHB operating floor, spent and new fuel pool areas are provided with two ventilation systems each of which have Particulate, Iodine, Gas (PIG) airborne effluent monitors monitoring the ventilation exhausts as described in Sections 11.5.2.7.2.2 and 11.5.2.7.2.3. The FHB normal exhaust is provided with effluent airborne monitoring for indication of airborne activity to operations personnel. Operations personnel have the capability to initiate the FHB Emergency Exhaust System from the Control Room as described in Section 7.3.1.3.4. The FHB Emergency Exhaust is provided with a PIG monitor for monitoring effluent exhaust downstream of the emergency exhaust systems HEPA-Charcoal filter units. The particulate and iodine channels of the PIG have been abandoned in place and are not used. Only the gas channel is used. This airborne effluent monitor measures effluent releases during and after a fuel handling accident. Any airborne activity release by the FHB normal ventilation system prior to switchover to the emergency exhaust system will be monitored by the FHB normal exhaust monitors. After switchover the FHB ventilation exhaust will also be monitored. The analyses performed to determine the adequacy of the 30 second switchover time for the FHB ventilation system is described in Section 6.5.1.2.1.1. The two analyses determine the following: a) The time of travel of radioactive gases from the spent fuel pool surface to the normal ventilation intake vents isolation damper; (calculated conservatively assuming these dampers remain open). b) The maximum allowable bypass period following a fuel handling accident. The ventilation system for the Fuel Handling Building (FHB) shown on Figures 9.4.2-1 and 9.4.2-2 shows applicable areas covered by the FHB ventilation system. The control drawing for the ventilation system switchover from normal ventilation to emergency exhaust is shown on Figures 7.3.1-13 and 7.3.1-14. The FHB operating floor area has GM tube area monitors at appropriate locations on the FHB walls, monitoring the building volume by the spent and new fuel pools. As described in Section 12.3.4.1.8.3 these GM tube area monitors will detect gamma radiation emanating from airborne material being drawn up into the FHB ventilation system from a fuel handling accident. When preset levels are reached, a high alarm signal will initiate switchover from normal ventilation to emergency ventilation. The FHB normal and emergency ventilation exhausts are monitored by Amendment 63 Page 140 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 airborne effluent Particulate, Iodine, Gas (PIG) monitors in the exhaust ducts as described in Sections 11.5.2.7.2.2 and 11.5.2.7.2.3. The capability to initiate the emergency ventilation system is provided in the Control Room as described in Section 7.3.1.3.4. The FHB emergency ventilation system switchover time and the FHB normal ventilation dampers isolation time of 10 seconds is within an acceptable duration that limits offsite doses to less than 10 CFR 50.67 limits as described in Section 15.7.4. 6.5.1.2.1.1 Time of Travel of Radioactive Gases Time for radioactive gases to travel from the spent fuel pool surface to the isolation damper of the normal ventilation system consists of travel time from the pool surface to the intake header and from there to the isolation damper through the length of ventilation duct. These times are evaluated as follows: a) Travel Time from Refueling Pool Surface to Exhaust Duct The equation of flow for round hoods is obtained from reference No. 5 of NUREG-0800 "Industrial Ventilation," 8th edition, by the American Conference of Governmental Industrial Hygienists. The velocity profile is given by: V= (1) x

where, V= Centerline velocity at distance X from hood, ft/min.

X= Distance outward along axis, ft (equation is accurate only for limited distance of X, where X is within 1.5D, where D is duct diameter or side of rectangular register). Q= Air flow rate, cfm. Q= Area of hood opening, ft2. D= Diameter of round hoods or side of essentially square hood, ft. Using Equation (1) above, the average velocity between the hood and any distance X can be obtained as follows: Vavg = ( / ) = ' Vavg = tan

        '(    ) /                    =0 The distance between pool surface and intake header is 44 ft.

Amendment 63 Page 141 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 X = 1.5D = 1.5 ft. (X is evaluated using the smaller side of the intake header.) Q = 2,300 cfm. A = 2 ft2. Vavg. in the first 1.5 ft. of the distance from the intake header equals: Vavg. = 439.2 ft/min Travel time for the first 1.5 ft. of the vertical distance from the exhaust register to the fuel pool surface becomes: t1 = x 60 = 0.2 sec. Air velocity at 1.5 ft. given by Equation (1) is: V1.5 ft. = 94 ft/min Conservatively assuming that velocity beyond 1.5 ft. does not decrease, then travel time required for balance of the distance can be calculated as follows: t2 = x 60 = 27.1 sec. The travel time, t3, from the intake header to the isolation damper is calculated to be 2.65 sec. The total travel time is then: ttotal = t1 + t2 + t3 = 29.95 sec. The exposure rate from a FHA will be almost instantaneous at the fuel pool surface, and therefore, the gaseous puff of activity will have just breached the fuel pool water surface and not have ascended toward the exhaust duct any significant distance before the monitors will have sensed the activity and initiated ventilation isolation and de-energization. It was conservatively assumed that half the distance to the exhaust register (22 feet) was required to provide an exposure rate to the radiation detectors to initiate damper closure. However, calculations show a point source at the pool surface provides exposure rates from Xe-133 and I-131 to be well in excess of the monitor setpoint. At this exposure rate and a detector sensitivity of 103 cpm/mr/hr, the monitor response time will be 0.6 seconds. Assuming the maximum distance traveled by the FHA radioactive gas is half the total distance to the exhaust register, the resulting travel time is 13.65 seconds. This is the longest time considered possible before the radiation detectors initiate an alarm signal to isolate. The time for the normal ventilation dampers to fully close once receiving an isolation signal is 10 seconds. Therefore, the total time for isolation is 24.25 seconds. b) Conclusions of Analysis of the Switchover Period Following Fuel Handling Accident. Amendment 63 Page 142 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The analysis of the bypass time was performed using the guidance given in Regulatory Guide 1.25. Bypass time is defined as the time period during which gaseous radioactivity is released unfiltered, therefore bypassing the Emergency Exhaust Filtration System. The time for FHA radioactive gas to actuate the radiation detectors and initiate ventilation damper isolation until the damper closes (24.75 seconds) is less than the total time calculated for the FHA radioactive gas to travel from the fuel pool surface to the first isolation damper (29.95 seconds). Therefore, there is zero bypass time, assuring that no unfiltered bypass leakage of radioactive gases can be released through the normal ventilation system. 6.5.1.2.2 RAB Emergency Exhaust System The RAB Emergency Exhaust System is shown on Figure 9.4.3-2 and consists of redundant 100 percent capacity fan and filter subsystems. Design data for principal system components are presented in Table 6.5.1-3. Each of the two subsystem filter trains includes a motor operated butterfly valve, decay heat cooling air connection, demister, electric heating coil, medium efficiency filter, HEPA pre-filter, charcoal adsorber and HEPA after filter. Connected to each filter train outlet is a centrifugal fan with a motor operated butterfly valve on its inlet and a backdraft damper on its outlet to prevent reverse airflow through the inactive fan. Upon receipt of a Safety Injection Actuation Signal (SIAS) or a Control Room Isolation Signal (CRIS), air operated valves on the normal ventilation penetrations into the areas containing equipment essential for safe shutdown close and both RAB Emergency Exhaust Systems are automatically energized. Either unit may then be manually de-energized from the Control Room, and placed on standby. Access into the areas in the RAB Emergency Exhaust System pressure seal boundary from other parts of the RAB is through leaktight doors under administrative controls. All penetrations into the enclosed area are provided with proper seals which limit the amount of inleakage. The seals permit differential movement between the penetration and the wall due to thermally or seismically induced motion. Negative pressure is established at 1/8 in. wg. by continuously exhausting air. Pressure is then controlled by the Airflow Control System which adjusts the variable inlet vane of the exhaust system. The system is provided with a locked open cross connection line that, in the original system design, allowed for room air to be drawn into either filter train after it had been shut down in order to provide for decay heat removal. It has since been shown that forced air flow is not required for decay heat removal and the room air lines have been permanently isolated. However, the cross connection line remains in place. Cooling for all areas exhausted by RAB Emergency Exhaust System is provided by the RAB ESF Equipment Cooling System. Refer to Section 9.4.5 for a detailed discussion. System design compliance with Regulatory Guide 1.52, Revision 2, is discussed in Table 6.5.1-2. Amendment 63 Page 143 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.5.1.2.3 Control Room Emergency Filtration System Refer to Section 9.4.1 for a detailed discussion of Control Room Emergency Filtration System. System design compliance with Regulatory Guide 1.52, Revision 2, is discussed in Table 6.5.1-2. 6.5.1.3 Design Evaluation 6.5.1.3.1 FHB Emergency Exhaust System Two 100 percent capacity subsystems are provided for the FHB Emergency Exhaust System, either of which is capable of meeting the design bases. The subsystems are located within the FHB which protects them from the effect of natural phenomena and missiles. The system components are designed to meet the applicable environmental conditions specified in Section 3.11. All components, ductwork and piping of each subsystem are physically separated from one another so that a single active failure in any component will not impair the system's ability to meet the design bases. A single failure analysis for the FHB Emergency Exhaust System is presented in Table 6.5.1-4. Instruments and controls and power to the redundant subsystems are electrically separated and powered from separate onsite power sources. The subsystems are actuated through separate channels of high radiation signals or FHB operating floor isolation signals. The FHB Emergency Exhaust System is designed to meet Safety Class 3 and Seismic Category I requirements. The temperature of air leaving each charcoal adsorber assembly is monitored. If the temperature rises above a pre*high or high level, an alarm on a local detection panel and in the Control Room will be annunciated. Interconnecting duct between the two emergency air cleaning units was originally provided, as shown in the system flow diagram, to allow one air cleaning unit to draw a small quantity of bleed air through the second inactive filtration train for decay heat cooling. However, the actual temperature increase of the carbon in the shutdown unit is calculated to be well below minimum auto-ignition or desorption temperatures. Therefore, the interconnecting duct is not needed and is blanked off at each unit. The HEPA filters meet ASME AG-1 or military specification MIL-F-51068 and MIL-F-51079 and are of fire and water resistant construction in accordance with UL-586, Class 1. They are individually factory tested and certified to have an efficiency not less than 99.97 percent when tested with 0.3 micron dioctylphthalate smoke in accordance with military Standard MIL-STD-282 and USAEC Health and Safety Bulletin, Issue No. 120.306. Charcoal adsorbers are filled with activated coconut shell charcoal. Laboratory tests of representative samples of charcoal are conducted to demonstrate their capability to attain the decontamination efficiency as indicated in Table 2 of Regulatory Guide 1.52 Revision 2, with the exceptions listed in Table 6.5.1 2. Each air cleaning unit is designed to be tested in place to verify that the unit meets the particulate filtration, iodine adsorption and leaktightness requirements. Amendment 63 Page 144 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.5.1.3.2 RAB Emergency Exhaust System Two 100 percent capacity subsystems are provided for the RAB Emergency Exhaust System, either of which is capable of meeting the design bases. The subsystems are located in separate compartments within the RAB which protects them from the effects of natural phenomena and missiles. All components are qualified to meet the applicable environmental conditions specified in Section 3.11. Instruments and controls and power to the redundant subsystems are electrically separated and powered from separate onsite power sources. The subsystems are actuated through separate channels of the Safety Injection Logic or the Control Room Isolation Logic. A single active failure in any component of the RAB Emergency Exhaust System will not impair the system's ability to fulfill the objectives given in the design bases. A single failure analysis is presented in Tables 6.5.1-4, and 6.5.1-5. The RAB Emergency Exhaust System is designed to meet Seismic Category I and Safety Class 3 requirements. The temperature of air leaving each charcoal adsorber assembly is monitored. If temperature rises above a pre-high or high level, an alarm on a local detection panel and in the Control Room will be annunciated. The maximum decay heat load in the RABEES charcoal filters has been shown to remain low enough that forced air cooling is not required for these units. The HEPA filters meet ASME AG-1 or military Specification MIL-F-51068 and MIL-F-51079 and are of fire and water resistant construction in accordance with UL-586, Class 1. They are individually factory tested and certified to have an efficiency not less than 99.97 percent when tested with 0.3 micron dioctylphthalate smoke in accordance with military standard MIL-STD-282 and USAEC Health and Safety Bulletin, Issue No. 120.306. Charcoal adsorbers are filled with activated coconut shell charcoal. Laboratory tests of representative samples of charcoal are conducted to demonstrate their capability to attain the decontamination efficiency as indicated in Table 2 of Regulatory Guide 1.52, Revision 2 (see Table 6.5.1 2). Each air cleaning unit is designed to be tested in place to verify the unit meets the particulate filtration, iodine adsorption and leaktightness requirements. 6.5.1.3.3 Control Room Emergency Filtration System Refer to Section 9.4.1 for a detailed discussion of the Control Room Emergency Filtration System. Safety evaluation is described in Section 9.4.1.3. 6.5.1.4 Test and Inspection Testing and maintenance are primary factors in assuring the reliability and the post-accident fission product removal capability of the Emergency Exhaust Systems. Amendment 63 Page 145 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The qualification tests of filtration system components comply to the requirements of Regulatory Guide 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2. The in-place airflow distribution test for HEPA filters and charcoal adsorbers, DOP test for HEPA filters, leak test for charcoal adsorber section and laboratory test for activated carbon are in accordance with Sections 5 and 6 of Regulatory Guide 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2. The system will undergo preoperational and start-up tests as described in Section 14.2.12.1.58. Periodic tests as required by the Technical Specifications will be performed. Inservice inspection will be performed in accordance with Section 6.6 and the valve testing requirements of Section 3.9.6 will apply. 6.5.1.5 Instrumentation Requirements for the RAB and FHB Emergency Exhaust Systems Indication is provided in the Control Room for the normal flow and low flow conditions for each filtration train. A low flow signal from the operating train initiates an alarm on the main control board. A Fire Detection Control System is provided for the adsorber section. The temperature of air leaving each charcoal adsorber assembly is monitored by a thermister wire traced over each charcoal adsorber outlet. On temperature rising above a pre-high or high level, an alarm on the detection panel and in the Control Room is annunciated. This will permit initiation, if necessary, of procedures that will prevent high temperature iodine desorption. Thermometers are provided for the filtration unit inlet downstream of charcoal adsorber. Indicators and recorders are provided in the Control Room for temperature of air entering and leaving the electric heating coil. If the leaving air temperature from the electric heating coil reaches a dangerous level, the temperature alarm for the charcoal adsorber will alert the Control Room Operator to survey the appropriate temperature indicator and manually de-energize the fan serving the subsystem with the high temperature. A relative humidity controller, Hydrocon-1, with a Chemical Research Corp. PCR-55 relative humidity sensor is provided for the charcoal adsorber section. High relative humidity annunciation is provided in the Control Room. This ensures that the relative humidity of the air stream is maintained within a range of 0-70 percent in order that an acceptable methyl iodide trapping efficiency is maintained. Pressure differential indicators and alarms in the Control Room are provided for the following components of the filtration unit. a) Prefilter - Alarm b) HEPA Prefilter - Indicator, Alarm c) Entire Filtration Unit - Indicator, Alarm Refer to Chapter 7 for further discussion of Emergency Exhaust System instrumentation. Amendment 63 Page 146 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.5.1.6 Materials The ESF Filter Systems are located outside the Containment Building. The system operates at a relatively low working temperature and the radiolytic or pyrolytic decomposition of the system material does not therefore pose any significant problem. Filtration components excluding filter media meet the material requirements described in ANSI/ASME N509-1976. The ESF filter, housing, frames and floor are made of stainless steel. Refer to Tables 6.5.1-1 and 6.5.1-3 for component material description. 6.5.2 CONTAINMENT SPRAY SYSTEM 6.5.2.1 Design Bases The Containment Spray System (CSS) performs the dual functions of removing heat and fission products from a post-accident containment atmosphere (fission products are discussed in Section 15.6). The heat removal capability of this system is discussed in Section 6.2.2 (Containment Heat Removal). The fission product removal function is carried out by the Iodine Removal System (IRS) in conjunction with containment heat removal. The IRS removes radioiodines from the containment atmosphere following a loss-of-coolant accident (LOCA) by adding controlled amounts of sodium hydroxide (NaOH) to the containment spray water. The design bases for the Containment Spray System as a fission product removal system are as follows:

1. To provide adequate capability for the fission product scrubbing of the containment atmosphere following a LOCA so that offsite doses and doses to operators in the Control Room are within the guidelines of 10 CFR 50.67. The radioactive material release assumptions of Regulatory Guide 1.183 (see Section 1.8 for compliance) are used in determining system capability. The fission product inventories in the containment are discussed in Section 15.6.

2. To blend Sodium Hydroxide (NaOH) into the spray stream to enhance absorption and retention of iodine by chemical reaction by maintaining a pH value of not less than 7.0 and not more than 11.0 during the long-term recirculation period (the pH range is discussed in greater detail in Section 6.5.2.3.3). The fission product iodine removed from the containment atmosphere remains mixed in the spray solution and will not evolve back into the containment atmosphere.

3. To remove elemental iodines and particulates with the minimum first order removal coefficients in accordance with WASH 1329 as follows:

Iodine Form First Order Removal Coefficient Elemental 20 hours-1 Particulate 3.938 hours-1

Filter media meets the material requirements described in ANSI/ASME N509-1980. Except that, ANSI requires HEPA filters to be in accordance with MIL-F-51068, and MIL-F-51068 has been canceled and replaced by ASME AG-1. Therefore, HEPA filters will be allowed to either specification. Amendment 63 Page 147 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

4. To meet all removal requirements based on an effective spray coverage of 85.9 percent of the containment free volume. This includes volumes beneath areas of grating in the operating floor (Elevation 286 ft.). The specified grating has 80 percent free area.

5. To perform its function following a LOCA, assuming a single active component failure coincident with loss of offsite power.

6. To perform its function following a safe shutdown earthquake.

7. To perform its function under the post-accident environmental conditions specified in Section 3.11.

8. To provide system materials which are compatible with fluid chemistry and applied codes and standards. System component design data parameters are given in Table 6.5.2-1.

6.5.2.2 System Design The system flow diagram is shown on Figure 6.2.2-1. System component design data parameters are given in Table 6.5.2-1. A discussion of the spray header design including a description of the number of nozzles per header, nozzle spacing, and nozzle is contained in Section 6.2.2. System operation is automatically initiated by a HI 3 signal. The signal starts the two spray pumps and the motor operated spray isolation valves. Within approximately 33 seconds after the spray pumps reach full speed, water will reach the nozzles and start spraying (see Section 6.2.2). The motor operated NaOH isolation valves will be opened automatically by the HI-3 signal. After the opening of the NaOH Isolation valve, the kinetic energy in the eductor will create a negative pressure to draw the Sodium Hydroxide solution (NaOH) from the containment spray additive tank, NaOH solution will be injected into the Containment Spray System (CSS) lines just upstream of the CS pump suction at a rate sufficient to provide the required range of pH for the containment spray. Turbulence in the fluid passing through the pump is sufficient to assure complete and uniform mixing of the fluid. The NaOH isolation valves will automatically close when the containment spray additive tank is empty. Additional NaOH can be added to the tank or through an emergency NaOH addition line outside the Tank Building. If necessary, the operator may reopen these NaOH isolation valves at any later time. The containment spray pumps initially take suction from the refueling water storage tank (RWST). The minimum operating capacity of the RWST (see Section 6.2.2) is more than adequate to supply enough water for the injection mode of operation. When low-low level tank water level is reached in the RWST, pump suction is transferred to containment recirculating sump automatically by opening the recirculation line valves and closing the valves at the outlet of the RWST. The Containment Spray System can provide one year of operation if required. The layout of the containment spray system headers and nozzle orientation (see Section 6.2.2) provides a minimum spray coverage of 92.6 percent of the containment free volume and 95 percent of the surface area of the operating floor (Elevation 286 ft.) with only one spray train in Amendment 63 Page 148 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 operation. This includes the volume beneath the grating in the operating floor. The specified grating has 80 percent free area. The drop size spectrum is discussed in Section 6.2.2. Forced air ventilation is provided to avoid stagnant air regions (see Section 6.2.2). The small amount of aluminum in Containment reacting with the spray solution will not form a colloidal suspension or a precipitate which could subject the nozzles to clogging. 6.5.2.3 Design Evaluation 6.5.2.3.1 Theory of iodine removal by containment spray Using the models described in WASH 1329, an evaluation of the effectiveness of the IRS in removing radioiodines from the containment atmosphere post LOCA has been performed. The removal of radioiodine is considered to be a first order rate phenomenon and is mathematically described below:

which integrates to: C = Coe-t where C = airborne concentration

= removal rate constant Co = initial concentration t = time of spray operation (injection phase)

The iodine removal rate constant () is a function of the iodine absorption efficiency of the spray droplets (E), the iodine partition coefficient (H), the flow rate of the Containment Spray System (F), and the containment volume (V). The iodine absorption efficiency of the spray droplets (E) takes into account the mass mixture transport process of iodine from the containment air stream to the spray drops and within the drops themselves, and the hydrodynamic and aerodynamic behavior of the drops as they fall through the Containment. The mass transfer model used to calculate the transfer of iodine considers both the interface gas film resistance and the liquid phase resistance of the drops. The effect of drop saturation inhibiting mass transfer rates is not considered since calculations show that saturation does not occur in the time interval of drop transit through the containment atmosphere. Amendment 63 Page 149 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The partition coefficient, H, is defined as the equilibrium ratio of the concentration of iodine in the liquid phase to concentration in the gas phase. It is a function of temperature, pH, and iodine concentration. 6.5.2.3.2 Effectiveness of Fission Product Cleanup by the Spray System a) Introduction This section discusses the removal of airborne iodine by the Containment Spray System. In the event of a design basis accident, large amounts of steam and possibly a substantial amount of radioactive iodine (typically, I-131), particulates, and noble gases will be released to the containment atmosphere. The CSS is actuated by Hi-3 containment atmosphere pressure signal. A basic borate solution will be pumped through spray nozzles located near the top of the Containment Building at one or more intermediate levels in the building. In falling through the Containment to the floor below, the spray droplets will cool the containment atmosphere and remove from the atmosphere the inorganic (molecular) iodine (I2), particulate iodines, and other particulates released in the accident. The model used to compute the rate of Elemental iodine removal by the spray solution is based on guidance in SRP 6.5.2, Section III.4. Elemental Iodine Spray Removal Coefficient The following formula, used in the FSAR at the time of this calculation, is the functional equivalent of that provided in SRP 6.5.2, Section III.4.c(1): S=1470 1470 Constant for conversions to yield consistent units 0.0235 = VT/ut 1730 = F = Spray Flow (gpm) 125 = h = fall height (ft) [Average fall height to operating floor] 2,013,730 = Vc = Sprayed Volume of Containment (ft3) 0.1 = d = Droplet diameter (cm) 37.1 = s = calculated spray removal coefficient (1/hr) 20 = s = used in accident analysis (1/hr), max allowed per SRP 6.5.2, Section III.4.c(1) Particulate Spray Removal Coefficient Amendment 63 Page 150 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 The model used to compute the rate of Particulate removal by the spray solution is based on guidance in SRP 6.5.2, Section III.4.c(4) The following formula determines the Particulate Spray Removal Coefficient in units of 1/hr: p=

Where, E/D is a dimensionless collection efficiency E / average spray drop diameter D:

10 Initial Value (1/meter) 1 Value after particulate depletion factor of 50 obtained 13,876 = F = spray flow (cu. ft./hr)@ 1730 gpm

60 min/hr / 7.481 gal/cu.ft.

38.1 = h = fall height (meters) [from lowest spray ring to operating deck] = 125 ft. 2,013,730 = V = Sprayed Volume of Containment (ft3) 3.938 = initial p (1/hr) used in accident analysis 0.3938 = p (1/hr) after particulate DF of 50 obtained used in accident analysis The spray flow rate used in the calculations is 1730 gpm, the flow with only one of the two containment spray system pumps operating at maximum containment internal pressure. The majority of the spray droplets will fall a distance of 125 ft., which is the average distance from the spray headers located in the hemispherical containment dome to the operating floor. Some of the droplets will fall through the open areas in the operating floor and some of the droplets will be stopped above the operating floor. The spray system is designed to deliver droplets with an average diameter of 1000 at rated flow with a minimum available nozzle pressure of 40 psi above the actual containment pressure. The average drop size assumed in the spray calculation is 0.10 cm or 1000. 6.5.2.3.3 Injection of spray solution The Containment Spray System (CSS) is designed to deliver a spray to the Containment during the short-term injection phase and the initial period of recirculation with a minimum pH of approximately 7.0 to enhance the absorption of iodine and to prevent stress-corrosion cracking of austenitic stainless steel. To assure long-term retention of iodine, the CSS is designed to assure a minimum pH of 7.0 in the sump solution at the onset of the recirculation mode and at the completion of NaOH addition from the Spray Additive Tank (SAT). The maximum spray or sump pH will not exceed 11.0. This pH range is maintained by the controlled addition of sodium hydroxide to the spray solution. The containment spray eductors are sized to deliver sodium hydroxide into each of the two containment spray loops as discussed in Section 6.5.2.2. Figures 6.5.2-2 and 6.5.2-3 show the Amendment 63 Page 151 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 pH time history of the water both in the containment spray and in the containment sump for the following cases:

1. Case 1 - Minimum pH, longest time

2. Case 2 - Maximum pH, shortest time The time history curves assume total mixing in the sump for the minimum and maximum RWST volumes.

The figures indicate that the containment spray in all cases achieves a long-term recirculation phase pH within the range of 8.5 - 11.0 (see Figures 6.5.2-2 and 6.5.2-3). Fission products and sump pH may be monitored, via the sumps connected to the Safety Injection System, from samples taken at the discharge of the RHR heat exchangers (see Sections 5.4.7, 6.3, and 9.3.2). The pH of the water in the sump during recirculation and after the contents of the spray additive tank have been introduced into the Containment can be determined by using known volumes and chemical compositions of the Reactor Coolant System, refueling water storage tank, accumulators, and spray additive tank. If necessary, additional sodium hydroxide may be added to the Containment and recirculated sump fluid. If additional sodium hydroxide must be added to the recirculated water in excess of the amount injected from the spray additive tank, this can be accomplished by utilizing the emergency sodium hydroxide addition connections (see Figure 6.2.2-1). The emergency addition connections are fitted with a hose connector. If it is determined (by sampling and pH measurement, in conjunction with trend monitoring of the recirculation fluid) that additional sodium hydroxide will be required, a tank truck of sodium hydroxide will be obtained from a local supplier. A hose will be run from the truck unloading area to an emergency addition connection. After the hose connection is made, the sodium hydroxide will be added to the system utilizing the tank truck pump. A 30 weight percent NaOH solution has a freezing point of approximately 32°F, and a boiling point of 240°F at atmospheric pressure. The NaOH is stored in the containment spray additive tank in the RAB; thus, no special provisions for temperature control are installed on the tank. An N2 blanket is maintained in the tank to assure solution stability and to prevent degradation during long-term storage. The 30 weight percent sodium hydroxide solution used has no fire or flash point, and thus does not pose a fire hazard. The system components are fabricated of corrosion-resistant materials. They are designed to operate in the environment to which they will be exposed following the worst postulated design basis accident as discussed in Section 3.11. A cavitating venturi is installed downstream of each CS pump to ensure that both the motor and pump will not exceed the limit of operation during the short term initial injection mode. Both the containment spray pump and cavitating venturi characteristics are given in Section 6.2.2. Amendment 63 Page 152 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.5.2.3.4 Single failure analysis The single failure characteristics of the Containment Spray System have been evaluated in Section 6.2.2. One of two spray additive eductors will supply adequate sodium hydroxide solution to provide minimum required iodine removal. 6.5.2.4 Testing and Inspection The Containment Spray System will undergo preoperational and startup tests as described in Section 14.2.12. Periodic tests as required by the Technical Specifications, Section 16.2, will be performed. Inservice inspection will be performed in accordance with Section 6.6 and the pump and valves testing requirements of Section 3.9.6 will apply. 6.5.2.5 Instrumentation Requirement Instrumentation is provided for monitoring the actuation and performance of the Iodine Removal System. The instrumentation is as follows: Instrumentation Function

1. Sodium Hydroxide (NaOH) Tank Pressure Indicates pressure *** and alarms** on high and low N2 cover gas pressure

2. NaOH Tank Level Indicates level** and alarms** on low level and empty

3. NaOH Flow Indicates flow rate**

Local and Control Room

** Control Room

- - Local Instrumentation is provided to monitor NaOH Tank N2 pressure and tank level. Refer to Section 7.5, Safety Related Display Instrumentation, for the detailed descriptions of these monitors.

Also refer to Section 7.3 for the interface between the system instrumentation and operation. The following abnormal operating conditions will be alarmed in the Control Room: high or low N2 NaOH tank pressure, and low or empty NaOH tank level. 6.5.2.6 Materials The materials used in the Iodine Removal System are compatible with the NaOH solution and the environment for the following reasons: a) The specifications restrict metals to austenitic stainless steel, Type 316, 304 or an acceptable alternative material. b) None of the materials used are subject to decomposition by the radiation or thermal environment. The specifications require that the materials be unaffected when exposed to the equipment design temperature, the total integrated radiation dose, and the boric acid and NaOH solution. A listing of all the materials utilized in the Iodine Removal System is provided on Table 6.5.2-1. Amendment 63 Page 153 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 A complete discussion of the materials utilized in the engineered safety features systems is provided in Section 6.1. 6.5.3 FISSION PRODUCT CONTROL SYSTEMS 6.5.3.1 Primary Containment For a discussion of the primary containment structural and functional design and other containment systems, refer to the following sections: Concrete Containment 3.8.1 Containment Functional Design 6.2.1 Containment Heat Removal System 6.2.2 Containment Isolation System 6.2.4 Combustible Gas Control in Containment 6.2.5 Containment Leakage Testing 6.2.6 Containment Ventilation System 9.4.7 A summary of the Containment's capacity to control fission product releases following a design basis accident is shown in Table 6.5.3-1. Refer to Sections 6.2.2 and 6.5.2 for a discussion of Containment Spray System. Credit is taken for Containment Spray System as a safety related fission product removal system. A non-nuclear safety airborne radioactivity removal system is provided for the Containment to maintain the fission product activity at low level for safe personnel entry during normal operation. The system is discussed in Section 9.4.7. Another non-nuclear safety Containment Atmosphere Purge Exhaust System (CAPES) is provided for the Containment to dilute the noble gases and other airborne containment concentrations by continuously venting the Containment to the vent stack. The Containment Atmosphere Purge Exhaust System consists of two subsystems. One is used for low flow during normal operation and the other is used for high flow during reactor shutdown. The former is referred to as the Normal Containment Purge System (NCP) and the latter as the Containment Pre-Entry Purge System (CPP). Both are discussed in Section 9.4.7. 6.5.3.2 Secondary Containment The Shearon Harris Nuclear Power Plant does not utilize a secondary containment system. Refer to Figures 1.2.2-3 through 1.2.2-18 for the general arrangements of the Containment. 6.5.4 ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM This section is not applicable to the Shearon Harris Nuclear Power Plant. Amendment 63 Page 154 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6

REFERENCES:

SECTION 6.5 6.5.2-1 J.F. Croft, et al, "Experiments on the Deposition of Airborne Iodine of High Concentration," AEEW-R265, June, 1963. 6.5.2-2 J. D. McCormack, R. K. Hilliard, "Natural Removal of Fission Products Released from UO2 Fuel in Condensing Steam Environments," International Symposium on Fission Product and Transport Under Accident Conditions," CONF-650407, April, 1965. 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS This section discusses the inservice inspection program for ASME Class 2 and 3 components. Preservice inspection will be conducted in accordance with ASME Code Section XI, 1980 edition through Winter '81 addenda, except where specific relief is requested. Inservice inspection will be conducted in accordance with the ASME Section XI Edition required by 10CFR50.55a(g) as detailed in the Technical Specifications. 6.6.1 COMPONENTS SUBJECT TO EXAMINATION The scope of the program encompasses those ASME Class 2 and 3 components which are within the boundaries of safety-related systems. Safety-related systems and components are those required to (1) permit safe reactor shutdown, (2) mitigate the consequences of an accident, and (3) limit the radiation dose at the site boundary to the limits of 10CFR50.67. ASME Class 2 components are exempted from inservice examination requirements in accordance with IWC-1220 and the Examination Plan. ASME Class 3 components are exempted from inservice examination requirements in accordance with IWD-1220 and the Examination Plan. Detailed inservice examinations will be performed on ASME Class 2 non-exempt components in accordance with Table IWC-2500-1 except that RHRS, ECCS, and CHRS category C-F welds will be examined in accordance with the requirements of this section and the Examination Plan. Detailed inservice examinations will be performed in ASME Class 3 components in accordance with Table IWD-2500-1. ASME Code Category C-F has been subsumed by the adoption of EPRI Technical Report TR-I 12657, Rev. 8-A methodology, which is supplemented by Code Case N-578-1, for implementing risk-informed inservice inspections. This approach replaces the categorization, selection, examination volume requirements for portions of ASME Section XI Examination Categories B-F, B-J, C-F-I, and C-F-2 applicable to HNP with Examination Category R-A as defined in Code Case N-578-1. Implementation of the RISI program is in accordance with Relief Request 13R-

02. Detailed Examination Plan, including information on areas subject to examination, method of examination, and extent and frequency of examination, will be provided.

6.6.2 ACCESSIBILITY The design and arrangement of ASME Class 2 components provide adequate clearance in accordance with IWA-1500 to conduct the required examination and inspections of Subsection Amendment 63 Page 155 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 IWC. Where volumetric or surface examinations are performed, direct access to the component has been provided. The design and arrangement of ASME Class 3 components also provide adequate clearances in accordance with IWA-1500 to conduct the required examinations and inspections of Subsection IWD. A continuing program of radiation surveys during the refueling programs will be performed to ensure that any possible future problem areas are detected at an early stage. Should additional experience in the maintenance and inspection of operating plants indicate that areas exist where access will be either limited or impossible, requests for relief from Section XI requirements will be made. 6.6.3 EXAMINATION TECHNIQUES AND PROCEDURES Examination techniques and procedures for ASME Class 2 and 3 components will be in accordance with IWA-2200, IWC-2500 and IWD-2500, and are identical with those of ASME Class 1 as discussed in Section 5.2.4.3. 6.6.4 INSPECTION INTERVALS An inspection program for ASME Class 2 components will be developed in accordance with the guidance of IWC-2400 and Table IWC-2500-1. The inspection program will be based upon "Inspection Program B" given in Table IWC-2412-1. An inspection program for ASME Class 3 components will be developed in accordance with the guidance of IWD-2400 and Table IWD-2500-1. 6.6.5 EXAMINATION CATEGORIES The inservice inspection categories for ASME Class 2 components are in agreement with Table IWC-2500-1 and as noted above. The inservice inspection categories for ASME Class 3 components are in agreement with IWD-2500 and Table IWD 2500-1. The examination categories and inspection requirements are described in the Examination Plan. 6.6.6 EVALUATION OF EXAMINATION RESULTS Articles IWC-3000 and IWD-3000, concerning evaluation of examination results on ASME Class 3 components, have not yet been prepared. After their publication, this article will be reviewed and incorporated in this section as applicable. In the meantime, the evaluation of examination results for ASME Class 2 and 3 components will be performed in accordance with IWA-3000, IWB-3000, and IWC-3000. The repair procedures for ASME Class 2 components will comply with the requirements of IWA-4000. The repair procedures for ASME Class 3 components will comply with the requirements of IWA-4000. Amendment 63 Page 156 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 6.6.7 SYSTEM PRESSURE TESTS System leakage and hydrostatic tests of ASME Class 2 and 3 components are conducted per the requirements of ASME Section XI Articles IWA-5000, IWC-5000 and IWD-5000. 6.6.8 AUGMENTED INSERVICE INSPECTION TO PROTECT AGAINST POSTULATED PIPING FAILURES High-energy fluid system piping between containment isolation valves or from the Containment to the first valve outside Containment for piping without an isolation valve inside Containment and piping in the break exclusion region receive an augmented ISI as follows: a) Protective measures or structures do not prevent the access necessary for conducting the inservice inspection in accordance with Section XI. b) HNP has adopted EPRI Topical Report TR-1006937, Rev. 0-A methodology for additional guidance for adoption of the RISI evaluation process to Break Exclusion Region(BER) piping, also referred to as the High Energy Link Break (HELB) region. This change to the BER program was made under 10CFR50.59 evaluation criteria. The RISI evaluation for BER piping is in effect for the entire third ten-year 1st interval. One hundred percent of all pipe welds that are greater than 4 in. nps in the break exclusion region of main steam and feedwater shown in Figure 3.6.2 1 will be 100 percent volumetrically inspected within each inspection interval. c) One hundred percent of all welds that are less than or equal to 4 in. nps and greater than 1 in. nps in the break exclusion regions of Main Steam and Feedwater shown in Figure 3.6.2-1 and in the break exclusion regions of Steam Generator Blowdown and CVCS Letdown (at the containment wall in the mechanical penetration area) and all socket welds will receive a surface inspection within each inspection interval. d) Welded attachments, if any, in the inspection area will receive surface examination. e) Guard pipe is not used on the SHNPP piping system. f) The areas subject to examination are defined in accordance with Examination Categories C-F for Class 2 welds in Table IWC-2500-1 of Section XI. 6.7 MAIN STEAM LINE ISOLATION VALVE LEAKAGE CONTROL SYSTEM This section is not applicable to the Shearon Harris Nuclear Power Plant. Amendment 63 Page 157 of 157

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE TITLE 6.1.1-1 ENGINEERED SAFETY FEATURES MATERIALS 6.1.1-2 ALUMINUM AND ZINC INVENTORY INSIDE CONTAINMENT 6.1.2-1 PROTECTIVE COATINGS ON WESTINGHOUSE SUPPLIED EQUIPMENT INSIDE CONTAINMENT 6.1.2-2 ELECTRICAL CABLE INSULATION MATERIALS INVENTORY INSIDE CONTAINMENT 6.2.1-1 POSTULATED ACCIDENTS FOR CONTAINMENT DESIGN 6.2.1-2 CALCULATED VALUES FOR CONTAINMENT PARAMETERS 6.2.1-3 PRINCIPAL CONTAINMENT DESIGN PARAMETERS 6.2.1-4 CALCULATED CONTAINMENT PRESSURE & TEMPERATURE 6.2.1-5 INITIAL CONDITIONS FOR CONTAINMENT PEAK PRESSURE-TEMPERATURE ANALYSIS 6.2.1-6 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS 6.2.1-7 CONTAINMENT PASSIVE HEAT SINKS 6.2.1-8

SUMMARY

OF PASSIVE HEAT SINKS USED IN THE CONTAINMENT ANALYSES 6.2.1-9 ACCIDENT CHRONOLOGIES 6.2.1-10 DELETED BY AMENDMENT NO. 51 6.2.1-11 ASSUMPTIONS USED IN ANALYSIS OF INADVERTENT CONTAINMENT SPRAY SYSTEM ACTUATION 6.2.1-12 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL COLD LEG NOZZLE 150 IN2 BREAK 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK 6.2.1-14 STEAM GENERATOR SUBCOMPARTMENT DOUBLE ENDED HOT LEG GUILLOTINE BREAK MASS AND ENERGY RELEASE DATA 6.2.1-15 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED PUMP SUCTION LEG GUILLOTINE BREAK MASS AND ENERGY RELEASE DATA 6.2.1-16 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED COLD LEG GUILLOTINE BREAK MASS AND ENERGY RELEASE DATA 6.2.1-17 PRESSURIZER SUBCOMPARTMENT DOUBLE-ENDED SURGE LINE GUILLOTINE BREAK MASS AND ENERGY RELEASE RATES FOR ORIGINAL DESIGN BASES 6.2.1-17a PRESSURIZER SUBCOMPARTMENT DOUBLE-ENDED SURGE LINE GUILLOTINE BREAK MASS AND ENERGY RELEASES FOR SGR/PUR 6.2.1-18 PRESSURIZER SUBCOMPARTMENT SPRAY LINE DOUBLE ENDED PRESSURIZER BREAK MASS Amendment 63 Page 1 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE TITLE AND ENERGY RELEASE RATES FOR ORIGINAL DESIGN BASES 6.2.1-18a PRESSURIZER SUBCOMPARTMENT SPRAY LINE DOUBLE ENDED PRESSURIZER BREAK MASS AND ENERGY RELEASE RATES FOR SGR/PUR 6.2.1-19 REACTOR CAVITY SUBCOMPARTMENT PRESSURIZATION MODEL RELAP-4 VOLUME INPUT DATA 6.2.1-20 REACTOR CAVITY SUBCOMPARTMENT PRESSURIZATION MODEL RELAY-4 JUNCTION INPUT DATA 6.2.1-20A LIST OF PROJECTED AREAS 6.2.1-20B LIST OF LEVER ARMS 6.2.1-21 STEAM GENERATOR - LOOP 1 SUBCOMPARTMENT ANALYSIS-VOLUME INPUT DATA 6.2.1-22 STEAM GENERATOR LOOP 1 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA 6.2.1-23 STEAM GENERATOR - LOOP 3 SUBCOMPARTMENT ANALYSIS VOLUME DATA 6.2.1-24 STEAM GENERATOR LOOP 3 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA 6.2.1-25 STEAM GENERATOR AND PRESSURIZER LOOP - 2 SUBCOMPARTMENT ANALYSIS VOLUME DATA 6.2.1-26 STEAM GENERATOR AND PRESSURIZER LOOP 2 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA 6.2.1-27

SUMMARY

OF CALCULATED SUBCOMPARTMENT PEAK PRESSURES FOR ORIGINAL DESIGN BASES 6.2.1-28 CASE ANALYZED AND RESULTS 6.2.1-29a DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN M&E RELEASES (SAME FOR ALL DEPS RUNS-MAX. AND MIN. S.I.) 6.2.1-29b DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN M&E RELEASES (SAME FOR ALL DEPS RUNS-MAX. AND MIN. S.I.) 6.2.1-30 DELETED BY AMENDMENT NO. 51 6.2.1-31 DELETED BY AMENDMENT NO. 51 6.2.1-32 DELETED BY AMENDMENT NO. 51 6.2.1-33 DOUBLE ENDED HOT-LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES 6.2.1-34 DELETED BY AMENDMENT NO. 51 6.2.1-35 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES 6.2.1-36 DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND Amendment 63 Page 2 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE TITLE ENERGY RELEASES 6.2.1-37 DELETED BY AMENDMENT NO. 51 6.2.1-38 DELETED BY AMENDMENT NO. 51 6.2.1-39 DELETED BY AMENDMENT NO. 51 6.2.1.40 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES 6.2.1-41 DOUBLE-ENDED PUMP SUCTION BREAK MIN SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES 6.2.1-42 DELETED BY AMENDMENT NO. 51 6.2.1-43 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MAXIMUM SAFEGUARDS 6.2.1-44 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MINIMUM SAFEGUARDS 6.2.1-45 DELETED BY AMENDMENT NO. 51 6.2.1-46 DELETED BY AMENDMENT NO. 51 6.2.1-47 DOUBLE-ENDED HOT LEG BREAK MASS BALANCE 6.2.1-48 DELETED BY AMENDMENT NO. 51 6.2.1-49 DOUBLE-ENDED PUMP SUCTION BREAK, MAXIMUM SAFEGUARDS PRINCIPAL PARAMETERS DURING REFLOOD 6.2.1-50 DELETED BY AMENDMENT NO. 63 6.2.1-51 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE -MAXIMUM SAFEGUARDS 6.2.1-52 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE -MINIMUM SAFEGUARDS 6.2.1-53 DELETED BY AMENDMENT NO. 51 6.2.1-54 DELETED BY AMENDMENT NO. 51 6.2.1-55 DOUBLE-ENDED HOT LEG BREAK ENERGY BALANCE 6.2.1-56 DELETED BY AMENDMENT NO. 51 6.2.1-57 DELETED BY AMENDMENT NO. 51 6.2.1-58A MSLB FULL DOUBLE-ENDED RUPTURE (1.4 FT2) AT 102% POWER (WITH MFIV FAILURE) 6.2.1-58B MSLB FULL DOUBLE-ENDED RUPTURE (1.4 FT2) AT 30% POWER (WITH MFIV FAILURE) 6.2.1-59 DELETED BY AMENDMENT NO. 46 6.2.1-60 DELETED BY AMENDMENT NO. 46 Amendment 63 Page 3 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE TITLE 6.2.1-61 DELETED BY AMENDMENT NO. 46 6.2.1-62 ACTIVE HEAT SINK DATA FOR MINIMUM POST-LOCA CONTAINMENT PRESSURE 6.2.1-63 PASSIVE HEAT SINK DATA FOR MINIMUM POST-LOCA CONTAINMENT PRESSURE 6.2.1-64 SINGLE FAILURE ANALYSIS - CONTAINMENT VACUUM RELIEF SYSTEM 6.2.1-65 POST-ACCIDENT MONITORING CONTAINMENT ATMOSPHERE TEMPERATURE AND CONTAINMENT SUMP WATER TEMPERATURE 6.2.1-66 LOCA M&E RELEASE ANALYSIS CORE DECAY HEAT FRACTION 6.2.2-1 CONTAINMENT COOLING SYSTEM COMPONENTS 6.2.2-2 DELETED BY AMENDMENT NO. 48 6.2.2-3 CONTAINMENT FAN COOLER PERFORMANCE DATA 6.2.2-4 PRIMARY SHIELD COOLING SYSTEM COMPONENTS SAFETY CLASS - 3 UNITS 6.2.2-5 REACTOR SUPPORTS COOLING SYSTEM COMPONENTS SAFETY CLASS 3 UNITS 6.2.2-6 SINGLE FAILURE ANALYSIS CONTAINMENT COOLING SYSTEM 6.2.2-7 SINGLE FAILURE ANALYSIS CONTAINMENT SPRAY SYSTEM 6.2.2-8 CSS PUMP NPSH EVALUATION 6.2.2-9 CONTAINMENT SPRAY SYSTEM COMPONENT PARAMETERS 6.2.2-10 DELETED BY AMENDMENT NO. 43 6.2.2-11 DELETED BY AMENDMENT NO. 43 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA 6.2.4-2 CONTAINMENT ISOLATION VALVE POSITION FOLLOWING AN ACCIDENT 6.2.5-1 DELETED BY AMENDMENT NO. 62 6.2.5-2 CONTAINMENT HYDROGEN PURGE SYSTEM COMPONENTS NON NUCLEAR SAFETY UNITS 6.2.5-3a POST-LOCA CONTAINMENT TEMPERATURES 6.2.5-4 ALUMINUM INVENTORY IN CONTAINMENT 6.2.5-5 GALVANIZED ZINC INVENTORY IN CONTAINMENT 6.2.5-6 ZINC-BASE PAINT INVENTORY IN CONTAINMENT 6.2.5-7 FAILURE MODE AND EFFECTS ANALYSIS HYDROGEN MONITORING SYSTEM Amendment 63 Page 4 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE TITLE 6.2B-1 DELETED BY AMENDMENT NO. 51 6.2B-2 DELETED BY AMENDMENT NO. 51 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS 6.3.2-1 EMERGENCY CORE COOLING SYSTEM COMPONENT PARAMETERS 6.3.2-2 EMERGENCY CORE COOLING SYSTEM RELIEF VALVE DATA 6.3.2-3 MOTOR OPERATED ISOLATION VALVES IN THE EMERGENCY CORE COOLING SYSTEM 6.3.2-4 MATERIALS EMPLOYED FOR EMERGENCY CORE COOLING SYSTEM COMPONENTS 6.3.2-5 EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS 6.3.2-6 SEQUENCE OF SWITCHOVER OPERATION FROM INJECTION TO RECIRCULATION 6.3.2-7 EMERGENCY CORE COOLING SYSTEM SHARED FUNCTIONS EVALUATION 6.3.2-8 NORMAL OPERATING STATUS OF EMERGENCY CORE COOLING SYSTEM COMPONENTS FOR CORE COOLING 6.3.2-9 RWST OUTFLOW LARGE BREAK - NO FAILURES 6.3.2-10 PUMPS AND VALVES REQUIRED FOR ECCS OPERATION 6.4.2-1 CONTROL ROOM BUTTERFLY VALVES LEAKAGE RATE ESTIMATE 6.4.2-2

SUMMARY

OF MAIN CONTROL ROOM LEAK RATE CALCULATION 6.4.4-1 TOXIC CHEMICALS STORED ONSITE 6.5.1-1 DESIGN DATA FOR FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM 6.5.1-2 COMPARISON OF FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM, REACTOR AUXILIARY EMERGENCY EXHAUST SYSTEM AND CONTROL ROOM EMERGENCY FILTRATION SYSTEM WITH REGULATORY POSITIONS OF R.G. 1.52, REVISION 2 6.5.1-3 DESIGN DATA FOR REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM 6.5.1-4 FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM SINGLE FAILURE ANALYSIS 6.5.1-5 REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM SINGLE FAILURE ANALYSIS 6.5.2-1 IODINE REMOVAL SYSTEM COMPONENTS 6.5.3-1 PRIMARY CONTAINMENT OPERATION FOLLOWING A DESIGN BASIS ACCIDENT Amendment 63 Page 5 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.1.1-1 ENGINEERED SAFETY FEATURES MATERIALS Materials Employed for Safety Injection and Containment Spray Systems Components Component Material Pumps Containment spray SA-351, Grade CF8 or CF8M; SA182, Grade F304 or F316 NSSS Vendor Supplied Pumps Pump casing and heads SA-351, Grade CF8 or CF8M; SA-182, Grade F304 or F316 Flanges and nozzles SA-182, Grade F304 or F316; SA-403, Grade WP316L Seamless Piping SA-312, Grade TP304 or TP316 Seamless Stuffing or packing box cover SA-351, Grade CF8 or CF8M: SA-240, Type 304 or Type 316; SA-182, Grade F304 Pipe fittings SA-403, Grade WP316L Seamless; SA-213, Grade TP304, TP304L, TP316 or TP316L Closing bolting and nuts SA-193, Grade B6, B7 or B8M; SA-104, Grade 2H or 8M; SA-453, Grade 660 and Nuts SA-194, Grade 2H, 6, 7, or 8M NSSS Vendor Supplied Heat Exchangers Heads SA-240, Type 304 Nozzle necks SA-182, Grade F304; SA-312, Grade TP 304; SA-240, Type 304 Tubes SA-213, Grade TP304; SA-249, Grade TP304 Tube sheets SA-182, Grade F304; SA-240, Type 304; SA-516, Grade 70 with Stainless Steel Cladding A-7 Analysis Shells SA-240 and SA-312, Grade TP304 and SA-351, Grade CF8 Pressure retaining bolting SA-193, Grade B7 Valves Containing radioactive fluids: Pressure-containing parts Type 316 and 304 Seating surfaces Stellite No. 6 or equivalent Stems Type 410 or 17-4PH stainless (Type 630) Containing nonradioactive, boron-free fluids: Pressure-retaining parts SA-216 Grade WCB NSSS Vendor Supplied Valves Bodies SA-182, Grade F316 or SA-351, Grade CF8 or CF8M Bonnets SA-182, Grade F316 or SA-351, Grade CF8 or CF8M Discs SA-182, Grade F316 or SA-564, Grade 630 or SA-351, Grade CF8 or CF8M Amendment 62 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.1.1-1 ENGINEERED SAFETY FEATURES MATERIALS Materials Employed for Safety Injection and Containment Spray Systems Components Pressure retaining bolting SA-453, Grade 660 Pressure retaining nuts SA-453, Grade 660 or SA-194, Grade 6 Component Material Relief Valves Bodies SA-182, Type F316 or SA-351, Grade CF8 or CF8M All nozzles, discs, spindles, and guides SA-182, Type F316 or SA-564, Grade 630 Bonnets for stainless steel valves without a SA-182, Type F316 or SA-351, Grade CF8 or CF8M balancing bellows Piping All piping in contact with borated water SA-312, SA-376 or 358 Class 1, Type 304 or 316 Deleted in Amendment 62 Materials Employed for Containment System Reinforcing Steel A-615, Grade 60 Containment liner SA-516 Grade 70 NSSS Vendor Supplied Pressure Vessels, Tanks, Filters, etc. Shells and heads SA-351, Grade CF8A; SA-240, Type 304; SA-264 Clad Plate of SA-537, Class 1 with SA-240, Type 304 Clad and Stainless Steel Weld Overlay A-8 Analysis Flanges and nozzles SA-182, Grade F304; SA-350, Grade LF2 or LF3 with SA-240, Type 304 and Stainless Steel Weld Overlay A-8 Analysis Piping SA-312 and SA-240, Grade TP304 or TP316 Seamless Pipe fittings, Closure bolting, and nuts SA-403, Grade WP304 Seamless SA-193, Grade B7 and SA-194, Grade 2H Amendment 62 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.1.1-2 ALUMINUM AND ZINC INVENTORY INSIDE CONTAINMENT* Components Surface ft.2 Weight (lbs.) Material Flux Mapping Drive System 75 171 Al Source Intermediate and Power Range 83 244 Al Detectors Control Rod Drive Mech Conn 65 191 Al Rod Position Indicators 79 178 Al Miscellaneous Valves 86 230 Al Contingency 75 200 Al Polar Crane 37.25 71.5 Al Jib Crane - Removed 0 0 Hoist (Estimate) 26 50 Al Elevator (Estimate) 26 10 Al TOTAL WEIGHT OF ALUMINUM = 1345.5 pounds Components Surface ft.2 Weight (lbs.) Material Cable Trays 8776 4 Mils Zn Conduits 20531 1.5 Mils Zn Pull and Junction Boxes 1166 2 Mils Zn Pull and Junction Boxes 2180 2 Mils Zn Pull and Junction Boxes 530 2 Mils Zn Pull and Junction Boxes 103 5 Mils Zn Ductwork over Elevation 228.14 ft. 8960 .8-.9 oz/ft.2 Zn from Elevation 236 ft. to 261 ft. 7226 .8-.9 oz/ft.2 Zn 1728 .8-.9 oz/ft.2 Zn 672 .8-.9 oz/ft.2 Zn 614 .8-.9 oz/ft.2 Zn 2106 .8-.9 oz/ft.2 Zn from Elevation 261 ft. to 286 ft. 1656 .8-.9 oz/ft.2 Zn 1058 .8-.9 oz/ft.2 Zn 4514 .8-.9 oz/ft.2 Zn 540 .8-.9 oz/ft.2 Zn from Elevation 286 ft. - Up 1128 .8-.9 oz/ft.2 Zn 5366 .8-.9 oz/ft.2 Zn 1876 .8-.9 oz/ft.2 Zn Ductwork Miscellaneous 2648 .8-.9 oz/ft.2 Zn Grating & Stair Treads 51400 1.7 Mils Zn

Refer to Tables 6.2.5-5 and 6.2.5-6 for inventories used in post-LOCA hydrogen generation analysis.

Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.1.2-1 PROTECTIVE COATINGS ON WESTINGHOUSE SUPPLIED EQUIPMENT INSIDE CONTAINMENT Component Painted Surface Area (ft.2) RCS component supports 7600 Reactor coolant pump motors/motor supports 2550 Accumulator tanks 4050 Manipulator crane 2600 Other refueling equipment 2125 Remaining equipment (such as valves, auxiliary tanks and heat <1300 exchanger supports, transmitters, alarm horns, small instruments) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.1.2-2 ELECTRICAL CABLE INSULATION MATERIALS INVENTORY INSIDE CONTAINMENT WEIGHT (LBS) MAX OD INSULATION JACKET CABLE B/M NO. DESCRIPTION (in) TRAY COND TRAY CONDUIT NOTES D10-01 1/C 750 MCM 1.62 796 153 655 124 1. Insulation is EPR material

2. Jacket is CPE material D25-02 3/C # 10 0.65 31 69 69 155 1. HTK Insulation (N-98 material (Kerite)

D25-03 1/C # 6 0.47 55 9 130 22 2. Jacket is Vulcanized Chlorinated Rubber Material (Kerite) D25-04 3/C # 6 0.87 49 40 74 61 D25-06 3/C # 2 1.15 360 542 616 929 D25-07 1/C - 4/0 0.85 422 59 669 93 D25-08 3/C - 4/0 1.84 40 131 63 209 D25-09 1/C - 350 MCM 1.05 595 249 745 311 D50-01 2/C # 10 0.58 46 94 92 190 1. HTK Insulation (N-98) material (Kerite) D50-02 4/C # 10 0.70 4 21 4 32 2. Jacket is Vulcanized Chlorinated Rubber Material (Kerite) D50-06 2/C # 12 0.52 111 168 240 362 D50-07 4/C # 12 0.63 148 208 251 354 D50-08 7/C # 12 0.75 332 477 431 619 D50-09 10/C # 12 1.00 139 321 209 483 D50-10 12/C # 12 1.03 26 45 29 51 1. HTK Insulation (N-98) material (Kerite) D50-11 2/C # 16 0.435 90 99 311 342 2. Jacket is Vulcanized Chlorinated Rubber material (Kerite) D50-12 4/C # 16 0.495 43 70 85 140 D50-13 7/C # 16 0.54 112 128 127 146 D50-14 10/C # 16 0.74 13 9 18 14 D50-15 12/C # 16 0.76 126 55 153 67 D61-01 1 Pr # 16 0.38 69 87 318 400 1. Insulation material is EPR D61-02 2 Pr # 16 0.69 35 85 129 305 2. Jacket is CPE material D61-04 4 Pr # 16 0.77 188 194 544 561 D61-05 3/C # 16 0.39 107 157 382 562 D61-07 16/C # 16 0.76 5 11 7 16 D61-08 32/C # 16 1.06 87 22 91 23 D65-01 1 Quad # 16 0.45 126 181 241 346 1. Insulation material is EPR

2. Jacket is CPE Material D86-01 22C # 20+ 0.96 508 4 484 4 1. Insulation is EPR material 2C # 12 D86-02 11C # 20+ 0.70 269 - 349 - 2. Jacket is CPE 2C #12 D88-02 1 Pr # 16 0.34 4 1 15 2 1. Insulation material is FR-EP

2. Jacket is CPE material Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.1.2-2 ELECTRICAL CABLE INSULATION MATERIALS INVENTORY INSIDE CONTAINMENT WEIGHT (LBS) MAX OD INSULATION JACKET CABLE B/M NO. DESCRIPTION (in) TRAY COND TRAY CONDUIT NOTES D84-01 Triaxial 0.469 218 393 126 227 1. Insulation is tefzel

2. Jacket is chlorosulphonated polyethylene D80-01 4/C # 10+ 0.801 239 - 311 - 1. Insulation is EPR material 2/C # 8 D80-04 4/C # 4+ 1.303 343 282 446 367 2. Jacket is CPE.

2/C #8 D82-02 1/C # 2 0.505 87 13 21 3 1. Jacket is glass braid

2. Insulation is silicon rubber D70-01 1 Pr # 16 0.354 94 94 291 263 1. Jacket is hypalon

2. Insulation is FR-EPDM D97-01 1/C # 14 0.15 6 1 7 1 1. Insulation material is HTK (Kerite)

D97-02 12/C # 12 1.20 9 9 10 10 2. Jacket material is Vulcanized Chlorinated Rubber D98-01 1/C # 14 0.15 7 1 23 3 1. Insulation material is EPR

   -03             16/C # 16               0.73         25        6      23         6      2. Jacket material is CPE
   -04             26/C # 16               0.89         49       12      43        11 D59-02             1/C # 10                0.37          9        2      30         7      1. Insulation material is HTK D59-04             1/C # 2                 0.59         10        6      22        12      2. Jacket material is Vulcanized Chlorinated Rubber D99-01             15/C # 18               0.52          6        7       7         8      1. Insulation material is FR-EP D99-33             STQ                     0.80         38        1      42         2      2. Jacket material is CPE D99-61             3 PR # 16 & 1/C # 16    0.45          2       11       2        12 D99-67             19/30 # 18              0.63         11       19      13        21 D99-90             35/C # 20               0.68         18       34      20        38 D87-04             17/C # 16               0.79         36       12      38        13      1. Insulation material is FR-EP

2. Jacket material is CPE D91-02 RG-59U (Coax) 0.24 22 39 24 44 1. Insulation material is FR-EP

2. Jacket material is CPE D90-01 32/C # 16 0.98 1 22 1 20 1. Insulation material is FR-EP

2. Jacket material is CPE LIGHTNING CABLE CABLE B/M NO. DESCRIPTION MAX OD (in) WEIGHT (lbs) JACKET WEIGHT (lbs) INSULATION NOTES D25-20 thru D-25-23 1/C # 4 0.43 155 233 1. Jacket is Hypalon.

D25-24 thru D25-31 1/C # 6 0.38 169 510 2. Insulation is EPR. D25-32 thru D25-40 & D25-49 1/C # 10 0.24 104 208 3. All B/Ms in conduit only. D25-41 thru D25-48 1/C # 12 0.22 24 46 Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-1 POSTULATED ACCIDENTS FOR CONTAINMENT DESIGN Mass and Energy Release Containment Design Parameter Postulated Accidents Reference A. Containment Peak Pressure Loss-of Coolant Accidents (LOCA) Temperature Double-ended pump suction leg guillotine Section 6.2.1.3 (DESLG) with maximum safety injection (SI) Double-end pump suction leg guillotine Section 6.2.1.3 (DESLG) with minimum safety injection Double-ended hot leg guillotine(DEHLG) with Section 6.2.1.3 minimum safety injection Main Stem Line Breaks (MSLB) Full double-ended rupture, 100.34% power Section 6.2.1.4 2 0.687 ft. split rupture, 100.34% power Section 6.2.1.4 Full double-ended rupture, 68.6% power Section 6.2.1.4 2 0.675 ft. split rupture 68.6 % power Section 6.2.1.4 Full double-ended rupture 29.4% power Section 6.2.1.4 2 0.666 ft. split rupture 29.4% power Section 6.2.1.4 Full double-ended rupture zero power Section 6.2.1.4 2 0.558 ft. split rupture zero power Section 6.2.1.4 B. Subcompartment Peak Pressure Reactor Cavity 2 150 in. hot leg break Section 6.2.1.2 2 150 in. hot leg break Section 6.2.1.2 Steam Generator Compartment Double-ended hot leg guillotine (DEHLG) Section 6.2.1.2 Double-ended suction leg guillotine (DESLG) Section 6.2.1.2 Double-ended cold leg guillotine (DECLG) Section 6.2.1.2 Pressurizer Subcompartment Pressurizer surge line guillotine Section 6.2.1.2 Pressurizer spray line break Section 6.2.1.2 C. External (Differential) Pressure Inadvertent Operation of the Containment N/A Heat Removal System (Both trains) D. Containment ECCS Minimum Refer to Section 6.2.1.5 N/A Pressure Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-2 CALCULATED VALUES FOR CONTAINMENT PARAMETERS Parameter Design Basis Accident Calculated Value Peak Containment Atmosphere DEHLG 41.8 psig Pressure (LOCA) Peak Pressure (MSLB) 30% Power full DE Rupture MSIV 41.3 psig and MFIV Failure Peak Containment Atmosphere DEHLG 258.9°F Temperature (LOCA) Peak Temperature (MSLB) 102% Power Full DE Rupture MSIV 364.4°F and MFIV Failure Peak Subcompartment Differential Pressure Reactor Cavity 150 in.2 CLG 29.8 psid Steam Generator (Loop 1) DEHLG 22.2 psid Steam Generator (Loop 2) and DESLG 22.4 psid Pressurizer Steam Generator (Loop 3) DECLG 29.7 psid External Differential Pressure Containment Inadvertent Operation of the 1.814 psid Containment Heat Removal System Minimum Pressure See Section 6.2.1.5 Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-3 PRINCIPAL CONTAINMENT DESIGN PARAMETERS Parameter Design Margin(1) Containment 45.0 7.1% Internal Design pressure, psig (LOCA) 45.0 8.2% (MSLB) External design pressure differential, psid 2.0 9.3% Net free volume, 106 ft.3 2.266 Not Applicable Design leak rate, percent free volume per day at 45.0 psig 0.1 Not Applicable Subcompartments Reactor cavity design wall loading, psid 64.0 53.4% Steam generator compartment design wall loading, psid Loop 1 38 41.6% Loop 2 (Including pressurizer compartment) 38 41.1% Loop 3 38 21.8% Notes: (1) Margin (%) =100 Actual margin, i.e., the margin between design values and peak calculated* values when using realistic or mediam parameter values would be much larger.

From Table 6.2.1-2.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-4 CALCULATED CONTAINMENT PRESSURE AND TEMPERATURE Full D/E MSLB - Cooling Train Failure Percent Power 102 70 30 0 Peak Pressure (psia) 49.8 50.8 53.0 52.6 Peak Temperature (F) 364.4 361.3 359.3 355.7 Time of Peak Pressure (sec) 108.2 124.7 149.2 291.5 Full D/E MSLB - Main Feedwater Isolation Valve Failure Power Level (%) 102 70 30 0 Peak Pressure (psia) 52.5 53.9 56.0 54.5 Peak Temperature (F) 364.4 361.3 359.3 355.7 Time of Peak Pressure (sec) 132.7 150.2 176.2 365.0 Full DEHLG LOCA Power Level (%) 102 Peak Pressure (psia) 56.5 Peak Temperature (°F) 258.9 Time of Peak Pressure (sec) 18.5 Full DEPSLG LOCA(1) Minimum ECCS Power Level (%) 102(2) Peak Pressure (psia) 53.6 Peak Temperature (°F) 257.9 Time of Peak Pressure (sec) 18.5 Notes:

1) The DEPSLG minimum ECCS case peak pressure is most limiting post blowdown but still bounded by the DEHLG case. See Figure 6.2.1-4.

2) Corresponds to a bounding core power of 2958 MWt.

Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-5 INITIAL CONDITIONS FOR CONTAINMENT PEAK PRESSURE-TEMPERATURE ANALYSIS Parameter Value Reactor Coolant System and Secondary System* Reactor power level, Mwt 2958 Core Inlet Temperature, F 560.4 Steam Pressure, psia 1011 Containment Pressure,*** max. (psig) 1.6 min. (in. W.G.) -1.0 Temperature***, F 138** Relative humidity, % max. 75 min. 20 Component cooling water temperature, F 120 Refueling water storage tank temperature, F 125 Net free volume (minimum), x 106 ft.3 2.266 Stored Water Minimum RWST volume available for Safety Injection, gal. 266,625

*NOTE: Values include uncertainties, where applicable.

- 138°F represents the bulk average atmosphere temperature when the indicated temperature is at the Technical Specification limit of 120°F, plus uncertainty.
  - Values used in the analyses depends on DBA (MSLB or LOCA) and peak temperature or pressure case.

Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-6 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS Value Used for Peak Pressure System/Item Full Capacity Analyses Containment Spray System Number of lines 2 1 Number of pumps 2 1 Number of headers 2 1 Spray Flow rate, gpm/pump Pump A - 1740 1730 Pump B - 1730 Containment Fan Coolers Number of units 4 2 Air side flow rate, acfm/unit 31,250 31,250 Heat removal rate, 106 Btu/hr. See Figure 6.2.1 16* See Figure 6.2.1 16* Fouling factor 0.001 0.001 Cooling water flow rates, gpm/unit 1,300 1,300 Source of cooling water Service Water Service Water Cooling water temperature, F 95 95

The cooling water flow rate assumed for Figure 6.2.1-16 is 1360 gpm. For SGR/PUR a more conservative value of 1300 gpm is assumed with one tube bundle plugged per safety train. Consequently a slightly lower heat removal rate is assumed in the analysis compared to that provided in Figure 6.2.1-16 and Table 6.2.2-3.

System/Item Full Capacity(3) Value Used for LOCA Analyses Heat Exchangers Shutdown heat exchangers (shell and U-tube) Number 1* Overall heat transfer coefficient, Btu/hr.-F (UA) per Calculated in GOTHIC(1)(2) heat exchanger: Flow Rates per heat exchanger: Recirculation side, gpm 3650(1) Exterior side, gpm 4850(1) Temperature: Exterior side, F 120 (max) Source of cooling water Component Cooling Water CCWS flow begins, sec. (Loss of offsite power) 28 NOTES: (1) These numbers are design minimum values. Actual values are greater than or equal to the design values. CCW flow rates do not reflect CCW flow to SFPHX after re-alignment. (2) The overall heat transfer coefficient used in the analysis varies due to recirculation side fluid (containment sump water) temperature and CCW temperature varying throughout the transient. (3) The full heat removal capacity of the RHR heat exchangers was modeled in the analysis of the Safety Injection temperature during the re-circulation phase of a LOCA used for determining the long-term mass and energy release data. The full capacity (maximum safeguards case) used the single train containment sump temperature-time history and the single train case heat exchanger removal rates, but utilized the two-train SI flowrate. The temperature of the SI flow for the full Amendment 63 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 capacity case would thus be conservatively high and therefore the containment atmospheric and sump temperatures would also be conservatively high.

This is based on single train operation.

Amendment 63 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-7 CONTAINMENT PASSIVE HEAT SINKS (1) Total Thermal Conductivity Volumetric Heat Capacity 2 2 3 Component Surface Area(ft. ) Thickness (in.) Btu/hr.-ft. -F Btu/ft. -F) A. Steel Containment Cylinder Liner 64038.20 0.375 25.9 53.5 Containment Dome Cylinder Liner 26015.08 0.50 25.9 53.5 Exposed Steel Liner in Refueling Cavity and Primary Shield Wall 6621.0 0.1875 25.9 53.5 Grating 52493.0 0.088 25.9 53.5 0.001 64.0 40.6 Structural 5716.0 0.03125 25.9 53.5 Steel Columns 12687.0 0.0625 25.9 53.5 Equipment Supports, 21211.0 0.125 25.9 53.5 Platform Framing, 18267.0 0.1875 25.9 53.5 Elevator Shafts 23097.0 0.25 25.9 53.5 Equipment Hatch 13689.3 0.3125 25.9 53.5 13230.0 0.375 25.9 53.5 2528.0 0.432 25.9 53.5 3696.0 0.4375 25.9 53.5 17070.0 0.50 25.9 53.5 2090.3 0.5625 25.9 53.5 3289 0.625 25.9 53.5 1656.0 0.6875 25.9 53.5 6106.0 0.75 25.9 53.5 1232.0 0.8125 25.9 53.5 134.0 0.875 25.9 53.5 1793.00 0.9375 25.9 53.5 6161.0 1.0 25.9 53.5 2765.0 1.125 25.9 53.5 348.0 1.1875 25.9 53.5 3330.0 1.25 25.9 53.5 2821.0 1.50 25.9 53.5 154.0 1.782 25.9 53.5 885.0 0.40116 8.6 54 HVAC (Duct and Equipment) 17039.1 0.05704 25.9 53.5 0.0013 64.0 40.6 Plumbing Piping 8573.1 0.1248 25.9 53.5 2240.0 0.1046 8.6 54 1210.5 0.0617 8.6 54 Amendment 62 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-7 CONTAINMENT PASSIVE HEAT SINKS (1) Total Thermal Conductivity Volumetric Heat Capacity 2 2 3 Component Surface Area(ft. ) Thickness (in.) Btu/hr.-ft. -F Btu/ft. -F) A. Steel (continued) Instrument and Control Equipment (Including Racks) 3240.0 0.125 25.9 53.5 Electrical Conduits, Cable Trays and Equipment 11062.0 0.4277 25.9 53.5 Fuel Trans. Sys. Control Panel 46.1 7.18 25.9 53.5 Manipulator Crane 1641.0 13.545 25.9 53.5 React. Lower Core Int.Stand 175.0 2.835 25.9 53.5 React. Upper Core Int.Stand 350.0 1.155 25.9 53.5 Fire Hose Racks 118.13 1.995 25.9 53.5 Internals Lifting Rig 603.75 1.575 25.9 53.5 RC Pump Motor and Drive 541.63 15.23 25.9 53.5 Flux Mapping Room Equip. and Thimbles 933.63 1.05 25.9 53.5 Neutron Detect. Pos. Device 262.5 0.525 25.9 53.5 RCCA Changing Fixture 12.25 1.365 25.9 53.5 RC Pump Handling Fixture 502.25 16.17 25.9 53.5 Control Rod Drive Shafts (6 Spare) 68.25 0.42 25.9 53.5 Reactor Vessel Hd. Guide Stud (3) 80.50 1.47 25.9 53.5 RCCA Thimble Plug Handling Tool 21.88 1.47 25.9 53.5 Full-Length CRD Shaft Unlatching Tool 12.25 2.73 25.9 53.5 Irradiation Sample Handling Tool 23.63 0.945 25.9 53.5 Head and Internals Lifting Rig Load Cell Linkage Assembly 46.38 3.15 25.9 53.5 Stud Tensioners (3) 215.25 4.095 25.9 53.5 Stud Tensioner Hydraulic Unit 35.0 4.2 25.9 53.5 Primary Loop Hot Leg Restr. 209.13 0.945 25.9 53.5 RC Pump Tie-Rods and Brackets 126.88 1.365 25.9 53.5 Pri Loop Restr. X-Over Leg Vert.Leg 61.25 1.26 25.9 53.5 Prim. Loop Restr. SG; RC Pump 35.88 0.945 25.9 53.5 Prim. Loop Restr. SG; Side Elbow 42.88 0.945 25.9 53.5 SG Upper Lat. Support 538.13 1.05 25.9 53.5 SG Lower Lat. Support 357.88 2.73 25.9 53.5 SG MWY Cover Supports 14.88 2.205 25.9 53.5 Reactor Vessel Supports 286.13 3.465 25.9 53.5 Triaxial Accelerograph 57.75 4.725 25.9 53.5 SG Vertical Column, Supports and Adaptors 898.63 1.68 25.9 53.5 RC Pump Vertical Column, Supports and Adaptors 636.13 1.365 25.9 53.5 Pressurizer Supports 52.5 1.785 25.9 53.5 Pipes and Equipment below LOCA flood line 5894.2 0.322 25.9 53.5 Misc. Stainless Piping 451.0 0.204 8.6 54.0 Piping with Refractory Insulation 17325.0 0.2562 8.6 54.0 Amendment 62 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-7 CONTAINMENT PASSIVE HEAT SINKS (1) Total Thermal Conductivity Volumetric Heat Capacity 2 2 3 Component Surface Area(ft. ) Thickness (in.) Btu/hr.-ft. -F Btu/ft. -F) A. Steel (continued) Misc. Piping 5210.0 0.2829 25.9 53.5 Seismic Restraints and Hangers 70322.0 0.21 25.9 53.5 B. Concrete Above Water Level Total Thickness (ft.) CRDM and Flux Mapping RM Wall 4413.92 1.5 1.0 31.9 Slab 1688.54 1.5 1.0 31.9 SG Shield Wall 8745.52 2.0 1.0 31.9 Operation Floor 7793.94 2.0 1.0 31.9 Pressurizer Room Wall 2984.1 1.25 1.0 31.9 Slab 560.56 2.0 1.0 31.9 Air Duct Shaft Wall 1373.96 1.5 1.0 31.9 1275.96 0.75 RCP Pedestal 2078.58 5.5 1.0 31.9 SG Pedestal 2511.74 6.0 1.0 31.9 Heat Exchanger Room Wall 1858.08 1.0 1.0 31.9 Slab 1208.34 0.75 1.0 31.9 RC Drain Tank Room Wall 587.0 0.875 1.0 31.9 Slab 433.16 1.0 1.0 31.9 Personal Shield 422.0 1.25 1.0 31.9 Wall 117.6 0.75 1.0 31.9 Secondary Shield Wall 25071.34 2.0 1.0 31.9 Primary Shield Wall (Unlined Portion) 3012.52 4.625 1.0 31.9 2837.1 2.25 1.0 31.9 Refueling Cavity Wall 1568.0 2.0 1.0 31.9 3670.1 2.5 1.0 31.9 1205.4 3.0 1.0 31.9 666.4 3.75 1.0 31.9 Refueling Cavity Slab 754.6 2.5 1.0 31.9 C. Concrete Below Water Level Air Duct Shaft Wall 176.4 1.5 1.0 31.9 186.2 0.75 1.0 31.9 RCP Pedestal 934.92 5.5 1.0 31.9 Amendment 62 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-7 CONTAINMENT PASSIVE HEAT SINKS (1) Total Thermal Conductivity Volumetric Heat Capacity 2 2 3 Component Surface Area(ft. ) Thickness (in.) Btu/hr.-ft. -F Btu/ft. -F) C. Concrete Below Water Level (continued) SG Pedestal 1420.02 6.0 1.0 31.9 Internal Mat 10844.68 5.0 1.0 31.9 Pit Wall and Slab 141.12 1.5 1.0 31.9 Personal Shield Wall 3735.76 1.25 1.0 31.9 823.2 0.75 1.0 31.9 Secondary Shield Wall 3483.9 2.0 1.0 31.9 Primary Shield Wall 1244.6 4.625 1.0 31.9 Refueling Cavity Wall 1229.9 2.5 1.0 31.9 443.0 3.0 1.0 31.9 Amendment 62 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-8

SUMMARY

OF PASSIVE HEAT SINKS USED IN THE CONTAINMENT ANALYSES Surface Area Exposed to Thermal Conductivity Volumetric Heat Capacity (1) 2 3 Structure Thickness Containment Interior(ft. ) (Btu/hr.-ft.-F) (Btu/ft. -F) A. Heat Sinks

1. Containment Primary Dome Paint film - 10 mils. 26,546 1.379 43.75 Steel liner - 0.5 in. 25.9 53.5 Concrete - 2.5 ft. 1.0 31.9

2. Containment Primary Cylinder Paint film - 10 mils. 61,220 1.379 43.75 Steel liner - .375 in. 25.9 53.5 Concrete - 4.5 ft. 1.0 31.9

3. Concrete Mat (Floor slab) Paint film - 27 mils. 12,256 1.379 43.75 Concrete - 4.6 ft.

4. Concrete exposed to Paint film - 27 mils. 13,678 0.16 14.93 Containment Sump water for Concrete - 2.67 ft. 1.0 31.9 LOCA or Containment atmosphere for MSLB (shield walls and foundations under flood line)

5. Concrete exposed to Paint film - 27 mils. 76,858 0.16 14.93 Containment Atmosphere (shield Concrete - 2.17 ft. 1.0 31.9 walls and concrete pads above flood line)

6. Stainless Steel (Refueling Pool Stainless Steel 0.016963 ft. 2,546 8.6 54.0 and Piping)

7. Steel Lined Concrete Paint film - 13 mils. 6,621 0.13833 10.55 Steel - 0.015625 ft. 25.9 53.5 Concrete - 2.2933 ft. 1.0 31.9

8. Galvanized Steel (conduit cable Zinc - 1.35 mils 17,039 64.0 40.6 trays)

Steel - 0.00453 ft 25.9 53.5

8. Galvanized Steel (conduit cable Zinc - 1.35 mils 52,493 64.0 40.6 trays) (continued)

Amendment 62 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-8

SUMMARY

OF PASSIVE HEAT SINKS USED IN THE CONTAINMENT ANALYSES Surface Area Exposed to Thermal Conductivity Volumetric Heat Capacity (1) 2 3 Structure Thickness Containment Interior(ft. ) (Btu/hr.-ft.-F) (Btu/ft. -F) Steel - 0.00733 ft 25.9 53.5 Zinc - 1.5 mils 11,062 64.0 40.6 Steel - 0.0354 ft 25.9 53.5

9. Structured + Miscellaneous Exposed Steel Paint film 10 mils 80,978 1.379 43.75 Steel (A&S) 0.0132 ft 25.9 53.5 Paint film 10 mils 70,322 0.078 28.8 Steel (MN-Hangers, etc) 0.0175 ft 25.9 53.5 Paint film 10 mils 5,210 0.078 28.2 Steel (MN-Equipment) 0.023574 ft 25.9 53.5 Paint film 10 mils 50,213 1.379 43.75 Steel (A&S) 0.03384 ft 25.9 53.5 Paint film 10 mils 22,461 1.379 43.75 Steel (A&S) 0.06644 ft 25.9 53.5 Paint film 10 mils 9,418 1.379 43.75 Steel (A&S) 0.11792 ft 25.9 53.5 Paint film 12 mils 11,813 0.2333 38.4 Steel (HVAC Ductwork) 0.0104 ft 25.9 53.5

10. Steel exposed to Containment Paint film - 13 mils 8,134 0.13833 10.55 Sump Water for LOCA or Steel - 0.0268 ft 25.9 53.5 Containment atmosphere for MSLB (below flood line)

(2)

11. MN Piping with refractory Insulation - 0.16887 ft 17,325 0.01 0.00127 (2) insulation (steel initial at Stainless steel - 0.021352 ft 8.6 54.0 422.5 F)

NOTES: (1) One side insulated (Coating thickness at the maximum allowed by installation specification has a negligible effect) (2) Total thickness Amendment 62 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-9 ACCIDENT CHRONOLOGIES A. Worst Case Hot Leg Break (DEHLG) Time (Seconds) Event 0.0 Break occurs 18.5 Peak containment pressure (blowdown) 20.40 End of blowdown start of pumped injection B. Worst Case Suction Leg Break (DESLG) Assuming Max. SI Time Assuming Min. SI (Seconds) Time (Seconds) Event 0.0 0.0 Break occurs 18.0 18.5 Peak containment pressure during blowdown 21.4 20.2 End of blowdown 32.4 32.3 Start of pumped injection 58.4 58.4 Start containment spray injection 58.4 58.4 Containment spray reaches full flow 110.0 110.0 Start of containment fan coolers 223.8 206.6 End of core reflood 1000.0 1000.0 Peak containment pressure subsequent to end of blowdown (2nd peak) 1283.5 1003.28 The time when the broken loop steam generator reaches thermal reaches 1st intermediate pressure during EPITOME period. 1200.00 2210.0 Start ECCS recirculation 1200.00 2210.0 End containment spray injection 1200.00 2210.0 Start containment spray recirculation 9500 (approx.) 24500 (approx.) 50 percent of containment design pressure (29.85 psia) reached 14500 (approx.) 40000 (approx.) 50 percent of containment peak calculated pressure (26.8 psia) reached Amendment 63 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.2.1-9(Continued) D. Worst Pressure Case MSLB (Full DEB, MFIV and MSIV Failure, 30% Power) Time(Seconds) Event 0.0 Break occurs 1.0 Main Steam Isolation Signal 1.0 Main Feedwater Isolation Signal 1.0 Containment Isolation Actuation Signal (CIAS) 3.0 MFIV's start to close 3.0 MSIV's start to close 8.0 MSIV's closed 11.0 MFIV's closed 58.4 Start Containment spray injection 58.4 Containment spray reaches full flow 110.0 Start of Containment fan coolers 176.2 Peak Containment Pressure 176.2 End of blowdown E. Worst Temperature Case MSLB (Full DEB, MSIV and MFIV Failure, 102% Power) Time(Seconds) Event 0.0 Break occurs 1.00 Main Steam Isolation Signal 1.00 Main Feedwater Isolation Signal 1.00 CIAS Signal 3.0 MFIV's start to close 8.0 MSIV's closed 11.0 MFIV's closed 58.40 Start of containment spray injection 58.40 Containment spray reaches full flow 58.40 Peak containment temperature 132.7 End of blowdown Amendment 63 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-11 ASSUMPTIONS USED IN ANALYSIS OF INADVERTENT CONTAINMENT SPRAY SYSTEM ACTUATION Item Assumed Value Containment Initial temperature, F 135 Initial pressure, inches w.g. -4 Relative humidity, % 65 Net free volume (minimum) ft.3 2.266 x 106 Passive heat sinks ignored for conservatism Containment Spray System Number of trains in operation 2 Flow rate per train, gpm 2146.5 Refueling water temperature, F 40 Spray efficiency, % 100 Containment Vacuum Breaker System Connecting to the RAB No. of vacuum breaker systems 2 No. assumed failed 1 Setpoint differential pressure to start opening vacuum 2.75 in. w.g. breaker system. Delay time to start opening vacuum breaker system. 5.5 sec. Vacuum breaker system exit flow area 425 in.2 Fully open loss coefficient referenced to exit flow area 3.912 Reactor Auxiliary Building (RAB) Initial temperature, F 104 Initial pressure, psia 14.7 Relative humidity, % 100 Net free volume (minimum), ft.3 65033 Passive heat sinks ignored for conservatism Setpoint differential pressure to open HVAC damper. 3.75 in. w.g. Delay time to open HVAC damper. 10.5 sec. Vent Connecting the RAB to the Outside Environment Exit Flow area, in.2 2304 Loss coefficient referenced to exit area. 15.017 Outside Environment Temperature, F 105 Pressure, psia 14.7 Relative humidity, % 100 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-12 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL COLD LEG NOZZLE 150 IN2-BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

         .00000                          0.                 0.00
         .00101                    8.8319843E+03           557.61
         .00200                    1.2459125E+04           557.40
         .00300                    1.2857556E+04           557.52
         .00400                    1.4251487E+04           557.41
         .00502                    1.5723959E+04           557.18
         .00600                    1.6310674E+04           556.84
         .00700                    1.7785577E+04           556.94
         .00802                    1.9498763E+04           556.90
         .00902                    2.0097690E+04           556.60
         .01004                    2.0531009E+04           556.35
         .01102                    2.1053367E+04           556.14
         .01202                    2.1309527E+04           555.82
         .01303                    2.1338598E+04           555.52
         .01402                    2.1975745E+04           555.51
         .01502                    2.3248402E+04           555.72
         .01603                    2.4329433E+04           555.76
         .01702                    2.4700207E+04           555.57
         .01803                    2.4666827E+04           555.27
         .01901                    2.4485169E+04           554.95
         .02001                    2.4259530E+04           554.64
         .02104                    2.3992691E+04           554.33
         .02203                    2.3822256E+04           554.11
         .02303                    2.3846373E+04           553.99
         .02401                    2.4001944E+04           553.92
         .02508                    2.4185665E+04           553.86
         .02607                    2.4312700E+04           553.79
         .02701                    2.4441169E+04           553.74
         .02801                    2.4672517E+04           553.75
         .02904                    2.5047176E+04           553.82
         .03002                    2.5467242E+04           553.92
         .03106                    2.5879331E+04           554.01
         .03202                    2.6173354E+04           554.06
         .03307                    2.6426705E+04           554.08
         .03404                    2.6599390E+04           554.08
         .03504                    2.6702468E+04           554.04
         .03604                    2.6719592E+04           553.97
         .03708                    2.6663322E+04           553.87
         .03802                    2.6568524E+04           553.76
         .03902                    2.6449194E+04           553.65
         .04010                    2.6322343E+04           553.53
         .04104                    2.6241842E+04           553.46
         .04207                    2.6226095E+04           553.43
         .04306                    2.6286025E+04           553.43
         .04403                    2.6400784E+04           553.47
         .04503                    2.6551812E+04           553.52
         .04604                    2.6705544E+04           553.57
         .04710                    2.6849091E+04           553.61
         .04805                    2.6940198E+04           553.62
         .04902                    2.6986217E+04           553.61
         .05004                    2.6978871E+04           553.58
         .05101                    2.6919492E+04           553.52
         .05204                    2.6804340E+04           553.44
         .05307                    2.6643852E+04           553.34 Amendment 61                                                       Page 1 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-12 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL COLD LEG NOZZLE 150 IN2-BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

         .05405                    2.6459392E+04           553.24
         .05501                    2.6261653E+04           553.13
         .05607                    2.6037337E+04           553.02
         .05701                    2.5854529E+04           552.93
         .05802                    2.5687588E+04           552.84
         .05909                    2.5526689E+04           552.77
         .06003                    2.5418755E+04           552.72
         .06106                    2.5323801E+04           552.68
         .06200                    2.5253101E+04           552.65
         .06306                    2.5187103E+04           552.62
         .06403                    2.5138770E+04           552.60
         .06506                    2.5100388E+04           552.59
         .06603                    2.5076526E+04           552.59
         .06710                    2.5062696E+04           552.59
         .06808                    2.5061362E+04           552.60
         .06901                    2.5066571E+04           552.61
         .07011                    2.5076735E+04           552.62
         .07105                    2.5084422E+04           552.63
         .07211                    2.5086728E+04           552.63
         .07300                    2.5079173E+04           552.63
         .07402                    2.5055283E+04           552.62
         .07503                    2.5010718E+04           552.59
         .07607                    2.4939836E+04           552.56
         .07706                    2.4846454E+04           552.52
         .07802                    2.4731583E+04           552.47
         .07907                    2.4583651E+04           552.40
         .08001                    2.4430026E+04           552.34
         .08105                    2.4248123E+04           552.26
         .08205                    2.4061514E+04           552.18
         .08302                    2.3886047E+04           552.11
         .08400                    2.3719146E+04           552.04
         .08502                    2.3557007E+04           551.98
         .08611                    2.3395041E+04           551.92
         .08706                    2.3255793E+04           551.87
         .08802                    2.3139040E+04           551.84
         .08904                    2.3032509E+04           551.80
         .09007                    2.2940099E+04           551.78
         .09106                    2.2862772E+04           551.75
         .09204                    2.2798691E+04           551.74
         .09305                    2.2740750E+04           551.72
         .09409                    2.2686062E+04           551.71
         .09501                    2.2641867E+04           551.70
         .09603                    2.2591886E+04           551.69
         .09710                    2.2534477E+04           551.68
         .09804                    2.2483379E+04           551.66
         .09912                    2.2425274E+04           551.65
         .10010                    2.2373914E+04           551.64
         .10205                    2.2291706E+04           551.62
         .10402                    2.2247664E+04           551.62
         .10600                    2.2255536E+04           551.65
         .10809                    2.2307643E+04           551.69
         .11007                    2.2363727E+04           551.73
         .11206                    2.2399816E+04           551.76
         .11415                    2.2413579E+04           551.77
         .11606                    2.2426812E+04           551.78 Amendment 61                                                       Page 2 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-12 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL COLD LEG NOZZLE 150 IN2-BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

         .11806                    2.2471239E+04           551.81
         .12001                    2.2559629E+04           551.86
         .12208                    2.2706561E+04           551.93
         .12406                    2.2879532E+04           552.00
         .12607                    2.3046557E+04           552.07
         .12810                    2.3187022E+04           552.12
         .13001                    2.3283942E+04           552.16
         .13204                    2.3360162E+04           552.18
         .13404                    2.3426168E+04           552.20
         .13604                    2.3495058E+04           552.22
         .13803                    2.3566056E+04           552.24
         .14011                    2.3637109E+04           552.26
         .14214                    2.3698584E+04           552.28
         .14409                    2.3738841E+04           552.29
         .14609                    2.3753420E+04           552.28
         .14801                    2.3735735E+04           552.26
         .15009                    2.3679773E+04           552.22
         .15201                    2.3595450E+04           552.18
         .15401                    2.3487359E+04           552.12
         .15601                    2.3372949E+04           552.07
         .15807                    2.3256413E+04           552.02
         .16008                    2.3156955E+04           551.98
         .16203                    2.3082081E+04           551.95
         .16412                    2.3027569E+04           551.94
         .16601                    2.3000076E+04           551.93
         .16813                    2.2995843E+04           551.94
         .17010                    2.3013489E+04           551.96
         .17213                    2.3048841E+04           551.99
         .17418                    2.3094130E+04           552.01
         .17604                    2.3136716E+04           552.04
         .17801                    2.3170910E+04           552.06
         .18012                    2.3189743E+04           552.06
         .18203                    2.3189483E+04           552.06
         .18405                    2.3173520E+04           552.05
         .18604                    2.3146530E+04           552.03
         .18803                    2.3115602E+04           552.02
         .19005                    2.3085559E+04           552.00
         .19209                    2.3059161E+04           551.99
         .19419                    2.3036158E+04           551.98
         .19607                    2.3018022E+04           551.97
         .19811                    2.2999305E+04           551.97
         .20008                    2.2980931E+04           551.96
         .20507                    2.2938673E+04           551.94
         .21010                    2.2933443E+04           551.95
         .21512                    2.2954971E+04           551.97
         .22014                    2.2950058E+04           551.96
         .22502                    2.2928353E+04           551.95
         .23010                    2.2952581E+04           551.97
         .23506                    2.3072520E+04           552.02
         .24003                    2.3281157E+04           552.11
         .24504                    2.3515535E+04           552.20
         .25013                    2.3643062E+04           552.23
         .25502                    2.3592732E+04           552.18
         .26003                    2.3390515E+04           552.06
         .26509                    2.3118090E+04           551.94 Amendment 61                                                       Page 3 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-12 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL COLD LEG NOZZLE 150 IN2-BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

         .27008                    2.2877506E+04           551.84
         .27502                    2.2720679E+04           551.79
         .28017                    2.2678034E+04           551.79
         .28510                    2.2747790E+04           551.84
         .29003                    2.2876473E+04           551.91
         .29507                    2.3010287E+04           551.97
         .30007                    2.3139240E+04           552.02
         .30509                    2.3310172E+04           552.08
         .31005                    2.3521898E+04           552.16
         .31506                    2.3682144E+04           552.21
         .32016                    2.3762989E+04           552.22
         .32503                    2.3747359E+04           552.19
         .33004                    2.3622728E+04           552.12
         .33508                    2.3384402E+04           551.99
         .34006                    2.3046295E+04           551.85
         .34503                    2.2720192E+04           551.72
         .35010                    2.2511082E+04           551.66
         .35510                    2.2501211E+04           551.69
         .36005                    2.2675792E+04           551.79
         .36503                    2.2924189E+04           551.91
         .37017                    2.3141922E+04           551.99
         .37511                    2.3264067E+04           552.03
         .38011                    2.3285253E+04           552.02
         .38500                    2.3215051E+04           551.97
         .39001                    2.3074833E+04           551.89
         .39502                    2.2918005E+04           551.83
         .40012                    2.2837686E+04           551.80
         .40508                    2.2886282E+04           551.84
         .41007                    2.3014952E+04           551.91
         .41500                    2.3146622E+04           551.97
         .42013                    2.3253525E+04           552.01
         .42505                    2.3319116E+04           552.03
         .43010                    2.3342656E+04           552.02
         .43502                    2.3315102E+04           552.00
         .44007                    2.3242087E+04           551.95
         .44511                    2.3148182E+04           551.91
         .45013                    2.3064584E+04           551.87
         .45508                    2.3005936E+04           551.85
         .46005                    2.2971344E+04           551.85
         .46507                    2.2960760E+04           551.85
         .47001                    2.2975247E+04           551.86
         .47506                    2.3020589E+04           551.89
         .48001                    2.3091317E+04           551.92
         .48524                    2.3171808E+04           551.96
         .49015                    2.3235774E+04           551.98
         .49507                    2.3274173E+04           551.99
         .50012                    2.3307522E+04           552.00
         .51012                    2.3457884E+04           552.06
         .52006                    2.3552873E+04           552.08
         .53004                    2.3491859E+04           552.03
         .54000                    2.3282689E+04           551.92
         .55007                    2.3105522E+04           551.86
         .56004                    2.3135759E+04           551.90
         .57008                    2.3257830E+04           551.96
         .58007                    2.3258786E+04           551.95 Amendment 61                                                       Page 4 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-12 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL COLD LEG NOZZLE 150 IN2-BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

          .59001                   2.3180063E+04           551.90
          .60006                   2.3111914E+04           551.88
          .61000                   2.3174064E+04           551.92
          .62012                   2.3324087E+04           551.99
          .63011                   2.3422696E+04           552.01
          .64000                   2.3433127E+04           552.00
          .65010                   2.3390772E+04           551.97
          .66005                   2.3355244E+04           551.96
          .67002                   2.3367725E+04           551.96
          .68009                   2.3374377E+04           551.96
          .69006                   2.3368667E+04           551.96
          .70013                   2.3344088E+04           551.94
          .71003                   2.3315262E+04           551.93
          .72011                   2.3307913E+04           551.93
          .73004                   2.3319851E+04           551.94
          .74004                   2.3350128E+04           551.95
          .75011                   2.3378937E+04           551.96
          .76004                   2.3437804E+04           551.98
          .77000                   2.3475805E+04           551.99
          .78015                   2.3461859E+04           551.97
          .79002                   2.3411475E+04           551.95
          .80017                   2.3363249E+04           551.93
          .81000                   2.3358638E+04           551.93
          .82002                   2.3375038E+04           551.94
          .83013                   2.3397300E+04           551.95
          .84016                   2.3414859E+04           551.95
          .85001                   2.3412072E+04           551.95
          .86004                   2.3408072E+04           551.94
          .87019                   2.3410564E+04           551.95
          .88002                   2.3427685E+04           551.95
          .89005                   2.3437173E+04           551.95
          .90001                   2.3430893E+04           551.95
          .91013                   2.3421783E+04           551.95
          .92016                   2.3419358E+04           551.95
          .93017                   2.3421140E+04           551.95
          .94008                   2.3418732E+04           551.95
          .95006                   2.3411082E+04           551.95
          .96005                   2.3405128E+04           551.94
          .97005                   2.3399867E+04           551.94
          .98004                   2.3397928E+04           551.94
          .99010                   2.3399184E+04           551.95 1.00005                   2.3404374E+04           551.95 Amendment 61                                                       Page 5 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                         .00000          0.                 0.00
                         .00100    1.0105250E+04           655.29
                         .00202    1.0610067E+04           655.26
                         .00301    1.0582770E+04           655.22
                         .00401    1.0558857E+04           655.19
                         .00502    1.0555990E+04           655.23
                         .00601    1.0586959E+04           655.30
                         .00701    1.0638869E+04           655.38
                         .00800    1.3771204E+04           655.92
                         .00901    1.2249198E+04           656.04
                         .01001    1.4499344E+04           656.38
                         .01100    1.4998327E+04           656.44
                         .01200    1.5540535E+04           656.51
                         .01302    1.5704632E+04           656.54
                         .01402    1.6151012E+04           656.73
                         .01502    1.7039374E+04           657.04
                         .01601    1.7868204E+04           657.22
                         .01701    1.8317086E+04           657.26
                         .01800    1.8197483E+04           657.10
                         .01901    1.7730485E+04           656.88
                         .02004    1.7230573E+04           656.70
                         .02103    1.6835673E+04           656.56
                         .02200    1.6508443E+04           656.44
                         .02303    1.6043866E+04           656.25
                         .02402    1.5469954E+04           656.06
                         .02502    1.5071526E+04           656.00
                         .02600    1.5113559E+04           656.09
                         .02700    1.5457384E+04           656.23
                         .02806    1.5892300E+04           656.39
                         .02902    1.6253449E+04           656.51
                         .03006    1.6557081E+04           656.59
                         .03103    1.6774392E+04           656.65
                         .03207    1.6925314E+04           656.68
                         .03306    1.6968386E+04           656.66
                         .03406    1.6903047E+04           656.60
                         .03502    1.6761249E+04           656.52
                         .03602    1.6565467E+04           656.43
                         .03708    1.6333301E+04           656.32
                         .03805    1.6119994E+04           656.24
                         .03903    1.5935794E+04           656.18
                         .04001    1.5819048E+04           656.15
                         .04108    1.5787209E+04           656.15
                         .04201    1.5817878E+04           656.17
                         .04301    1.5877223E+04           656.20
                         .04406    1.5950923E+04           656.23
                         .04508    1.6018533E+04           656.25
                         .04600    1.6061413E+04           656.26
                         .04709    1.6065018E+04           656.24
                         .04801    1.6026494E+04           656.22
                         .04903    1.5951212E+04           656.17
                         .05004    1.5854454E+04           656.13 Amendment 61                                                                 Page 1 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                         .05106    1.5742200E+04           656.08
                         .05200    1.5626358E+04           656.03
                         .05304    1.5498947E+04           655.98
                         .05404    1.5392633E+04           655.94
                         .05505    1.5316719E+04           655.92
                         .05602    1.5281903E+04           655.91
                         .05710    1.5286845E+04           655.93
                         .05809    1.5317933E+04           655.95
                         .05902    1.5359169E+04           655.97
                         .06005    1.5408450E+04           655.99
                         .06105    1.5446950E+04           656.01
                         .06203    1.5465310E+04           656.01
                         .06307    1.5458102E+04           656.01
                         .06404    1.5428972E+04           655.99
                         .06509    1.5377479E+04           655.97
                         .06605    1.5317925E+04           655.94
                         .06703    1.5249735E+04           655.92
                         .06807    1.5176995E+04           655.89
                         .06909    1.5111902E+04           655.87
                         .07002    1.5066013E+04           655.86
                         .07109    1.5028739E+04           655.85
                         .07213    1.5010312E+04           655.85
                         .07303    1.5005336E+04           655.85
                         .07402    1.5005746E+04           655.86
                         .07508    1.5003829E+04           655.86
                         .07610    1.4990368E+04           655.86
                         .07707    1.4960781E+04           655.84
                         .07803    1.4909122E+04           655.82
                         .07905    1.4829243E+04           655.79
                         .08005    1.4729065E+04           655.75
                         .08102    1.4611445E+04           655.71
                         .08201    1.4477009E+04           655.66
                         .08302    1.4331003E+04           655.61
                         .08405    1.4180635E+04           655.56
                         .08502    1.4043072E+04           655.52
                         .08607    1.3905251E+04           655.48
                         .08705    1.3799552E+04           655.44
                         .08804    1.3698738E+04           655.42
                         .08901    1.3610579E+04           655.40
                         .09011    1.3542166E+04           655.38
                         .09107    1.3486022E+04           655.37
                         .09208    1.3428712E+04           655.36
                         .09303    1.3390308E+04           655.35
                         .09404    1.3335577E+04           655.34
                         .09503    1.3278101E+04           655.33
                         .09604    1.3225822E+04           655.32
                         .09702    1.3165525E+04           655.30
                         .09802    1.3095734E+04           655.29
                         .09910    1.3031588E+04           655.27
                         .10009    1.2969180E+04           655.26
                         .10503    1.2768948E+04           655.22
                         .11001    1.2697740E+04           655.22
                         .11510    1.2479383E+04           655.14
                         .12002    1.2108701E+04           655.02
                         .12501    1.1892632E+04           654.98 Amendment 61                                                                 Page 2 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                         .13015    1.1780754E+04           654.95
                         .13507    1.1709595E+04           654.95
                         .14000    1.1690491E+04           654.95
                         .14500    1.1614640E+04           654.93
                         .15007    1.1440950E+04           654.89
                         .15504    1.1287310E+04           654.86
                         .16005    1.1158492E+04           654.83
                         .16501    1.1022980E+04           654.81
                         .17000    1.0927623E+04           654.81
                         .17502    1.0867090E+04           654.81
                         .18000    1.0795204E+04           654.80
                         .18505    1.0645821E+04           654.77
                         .19011    1.0643389E+04           654.79
                         .19501    1.0635996E+04           654.81
                         .20004    1.0625862E+04           654.82
                         .21001    1.0605421E+04           654.83
                         .22002    1.0592801E+04           654.84
                         .23004    1.0588173E+04           654.86
                         .24017    1.0587078E+04           654.88
                         .25015    1.0603374E+04           654.92
                         .26001    1.1284612E+04           655.18
                         .27003    1.0949213E+04           654.98
                         .28000    1.0754729E+04           654.91
                         .29004    1.0643705E+04           654.88
                         .30004    1.0629268E+04           654.89
                         .31008    1.0620766E+04           654.89
                         .32014    1.0621174E+04           654.91
                         .33010    1.0616396E+04           654.90
                         .34005    1.0606236E+04           654.89
                         .35024    1.0589862E+04           654.88
                         .36018    1.0577181E+04           654.87
                         .37013    1.0583748E+04           654.89
                         .38018    1.0603201E+04           654.93
                         .39001    1.0623727E+04           654.96
                         .40000    1.0849992E+04           654.99
                         .41001    1.0886893E+04           654.96
                         .42003    1.0878688E+04           654.93
                         .43015    1.0852306E+04           654.92
                         .44001    1.0763823E+04           654.78
                         .45013    1.0736766E+04           654.69
                         .46001    1.0722672E+04           654.57
                         .47002    1.0695395E+04           654.26
                         .48009    1.0717358E+04           654.18
                         .49003    1.0818851E+04           654.08
                         .50001    1.0970746E+04           654.22
                         .51000    1.1159034E+04           654.52
                         .52003    1.1054592E+04           654.65
                         .53004    1.1033647E+04           654.75
                         .54020    1.1018892E+04           654.82
                         .55001    1.1001423E+04           654.87
                         .56018    1.0986136E+04           654.92
                         .57017    1.0983624E+04           654.97
                         .58015    1.0952771E+04           655.01
                         .59011    1.0946959E+04           655.07
                         .60011    1.0974916E+04           655.14 Amendment 61                                                                 Page 3 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                          .61006   1.1000097E+04           655.22
                          .62008   1.1018301E+04           655.30
                          .63008   1.1035912E+04           655.38
                          .64004   1.1040225E+04           655.46
                          .65026   1.1035395E+04           655.53
                          .66006   1.1037813E+04           655.59
                          .67008   1.1097994E+04           655.64
                          .68000   1.1091621E+04           655.64
                          .69000   1.1089756E+04           655.61
                          .70001   1.1108036E+04           655.58
                          .71012   1.1142844E+04           655.54
                          .72009   1.1183058E+04           655.48
                          .73014   1.1223983E+04           655.41
                          .74019   1.1256362E+04           655.32
                          .75005   1.1282482E+04           655.22
                          .76015   1.1319898E+04           655.15
                          .77007   1.1376847E+04           655.12
                          .78012   1.1431539E+04           655.10
                          .79007   1.1471538E+04           655.09
                          .80008   1.1505581E+04           655.08
                          .81017   1.1540554E+04           655.08
                          .82004   1.1577494E+04           655.09
                          .83001   1.1618296E+04           655.09
                          .84012   1.1640193E+04           655.10
                          .85015   1.1659614E+04           655.10
                          .86018   1.1684006E+04           655.11
                          .87011   1.1719357E+04           655.12
                          .88005   1.1758180E+04           655.14
                          .89006   1.1790857E+04           655.15
                          .90001   1.1816577E+04           655.17
                          .91012   1.1841744E+04           655.18
                          .92013   1.1869260E+04           655.20
                          .93008   1.1897816E+04           655.22
                          .94012   1.1927211E+04           655.24
                          .95012   1.1946374E+04           655.28
                          .96002   1.1967142E+04           655.34
                          .97000   1.1990573E+04           655.40
                          .98005   1.2011992E+04           655.47
                          .99006   1.2027742E+04           655.54 1.00011   1.2041494E+04           655.61 1.01012   1.2055686E+04           655.70 1.02023   1.2058832E+04           655.78 1.03000   1.2065277E+04           655.86 1.04009   1.2068659E+04           655.94 1.05000   1.2071642E+04           656.02 1.06004   1.2080042E+04           656.10 1.07020   1.2094119E+04           656.17 1.08022   1.2108985E+04           656.24 1.09006   1.2122129E+04           656.29 1.10008   1.2140506E+04           656.34 1.11008   1.2160923E+04           656.37 1.12001   1.2180334E+04           656.40 1.13006   1.2201204E+04           656.41 1.14004   1.2222397E+04           656.41 1.15001   1.2244078E+04           656.41 Amendment 61                                                                 Page 4 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.) 1.16015 1.2268080E+04 656.40 1.17008 1.2294679E+04 656.38 1.18011 1.2321863E+04 656.36 1.19006 1.2348034E+04 656.34 1.20012 1.2373848E+04 656.33 1.21004 1.2395820E+04 656.35 1.22001 1.2415994E+04 656.38 1.23028 1.2435956E+04 656.44 1.24020 1.2455882E+04 656.50 1.25006 1.2475352E+04 656.57 1.26014 1.2495104E+04 656.64 1.27000 1.2514710E+04 656.71 1.28004 1.2535229E+04 656.80 1.29021 1.2553077E+04 656.88 1.30004 1.2565603E+04 656.97 1.31008 1.2574253E+04 657.05 1.32005 1.2594828E+04 657.14 1.33003 1.2600002E+04 657.23 1.34010 1.2602730E+04 657.33 1.35003 1.2608563E+04 657.42 1.36002 1.2616603E+04 657.52 1.37003 1.2626860E+04 657.63 1.38004 1.2638778E+04 657.74 1.39020 1.2649592E+04 657.85 1.40010 1.2652172E+04 657.98 1.41006 1.2656618E+04 658.11 1.42004 1.2663442E+04 658.26 1.43002 1.2666051E+04 658.40 1.44009 1.2665171E+04 658.55 1.45008 1.2666734E+04 658.71 1.46009 1.2664616E+04 658.86 1.47009 1.2664805E+04 659.01 1.48011 1.2664564E+04 659.16 1.49003 1.2660979E+04 659.31 1.50012 1.2656865E+04 659.46 1.51010 1.2655918E+04 659.61 1.52015 1.2657492E+04 659.75 1.53011 1.2659966E+04 659.89 1.54011 1.2663688E+04 660.03 1.55003 1.2668902E+04 660.16 1.56010 1.2675711E+04 660.29 1.57005 1.2683652E+04 660.42 1.58012 1.2691736E+04 660.55 1.59006 1.2698101E+04 660.67 1.60011 1.2703562E+04 660.79 1.61001 1.2708449E+04 660.92 1.62029 1.2713135E+04 661.05 1.63006 1.2715022E+04 661.19 1.64003 1.2714929E+04 661.35 1.65006 1.2713582E+04 661.52 1.66014 1.2711390E+04 661.70 1.67010 1.2708673E+04 661.89 1.68004 1.2705441E+04 662.08 1.69003 1.2701664E+04 662.27 1.70000 1.2698115E+04 662.47 Amendment 61 Page 5 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.) 1.71030 1.2692309E+04 662.68 1.72007 1.2684034E+04 662.88 1.73015 1.2675491E+04 663.09 1.74012 1.2666637E+04 663.30 1.75001 1.2653944E+04 663.51 1.76017 1.2637858E+04 663.72 1.77020 1.2622373E+04 663.94 1.78011 1.2606850E+04 664.16 1.79005 1.2589699E+04 664.38 1.80005 1.2572196E+04 664.60 1.81011 1.2554261E+04 664.83 1.82011 1.2536563E+04 665.06 1.83009 1.2521042E+04 665.29 1.84002 1.2504470E+04 665.53 1.85011 1.2496217E+04 665.77 1.86013 1.2492896E+04 666.01 1.87009 1.2478238E+04 666.25 1.88012 1.2459728E+04 666.49 1.89008 1.2444107E+04 666.73 1.90005 1.2427676E+04 666.98 1.91008 1.2411993E+04 667.23 1.92011 1.2394744E+04 667.48 1.93009 1.2377802E+04 667.73 1.94012 1.2360450E+04 667.98 1.95014 1.2346708E+04 668.24 1.96007 1.2333988E+04 668.50 1.97006 1.2318036E+04 668.77 1.98012 1.2299479E+04 669.03 1.99011 1.2278152E+04 669.30 2.00012 1.2257038E+04 669.58 2.01015 1.2234090E+04 669.85 2.02001 1.2210620E+04 670.13 2.03015 1.2186273E+04 670.42 2.04009 1.2161208E+04 670.70 2.05003 1.2138423E+04 670.99 2.06004 1.2117643E+04 671.29 2.07011 1.2096290E+04 671.59 2.08018 1.2074929E+04 671.90 2.09015 1.2051989E+04 672.21 2.10012 1.2032288E+04 672.52 2.11002 1.2008044E+04 672.83 2.12011 1.1981719E+04 673.15 2.13007 1.1958063E+04 673.48 2.14005 1.1938307E+04 673.81 2.15000 1.1902313E+04 674.13 2.16000 1.1871376E+04 674.46 2.17018 1.1840380E+04 674.80 2.18028 1.1813980E+04 675.14 2.19002 1.1779106E+04 675.48 2.20008 1.1746437E+04 675.82 2.21011 1.1717815E+04 676.17 2.22007 1.1682224E+04 676.52 2.23009 1.1656145E+04 676.88 2.24010 1.1635537E+04 677.24 2.25000 1.1606759E+04 677.59 Amendment 61 Page 6 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.) 2.26017 1.1566030E+04 677.96 2.27013 1.1529150E+04 678.32 2.28008 1.1506706E+04 678.69 2.29015 1.1470585E+04 679.05 2.30003 1.1426224E+04 679.42 2.31006 1.1196024E+04 679.69 2.32001 1.1138844E+04 679.92 2.33009 1.1308279E+04 680.21 2.34005 1.1273368E+04 680.55 2.35005 1.1197207E+04 680.88 2.36005 1.1153269E+04 681.42 2.37007 1.1157151E+04 681.97 2.38021 1.1151873E+04 682.50 2.39024 1.1144881E+04 682.99 2.40010 1.1136913E+04 683.44 2.41008 1.1133399E+04 683.92 2.42006 1.1133273E+04 684.39 2.43011 1.1131634E+04 684.88 2.44003 1.1127026E+04 685.36 2.45003 1.1122502E+04 685.85 2.46002 1.1121341E+04 686.37 2.47011 1.1124320E+04 686.91 2.48001 1.1128375E+04 687.44 2.49004 1.1129679E+04 687.99 2.50010 1.1126662E+04 688.52 2.51011 1.1120319E+04 689.08 2.52007 1.1115835E+04 689.65 2.53004 1.1111082E+04 690.22 2.54005 1.1107080E+04 690.80 2.55011 1.1100815E+04 691.40 2.56005 1.1093406E+04 691.98 2.57004 1.1088308E+04 692.61 2.58007 1.1091970E+04 693.24 2.59021 1.1091308E+04 693.88 2.60011 1.1085872E+04 694.50 2.61002 1.1077456E+04 695.14 2.62006 1.1071304E+04 695.78 2.63011 1.1064348E+04 696.42 2.64013 1.1055587E+04 697.08 2.65002 1.1046906E+04 697.72 2.66009 1.1039075E+04 698.38 2.67009 1.1032599E+04 699.08 2.68009 1.1030332E+04 699.77 2.69003 1.1027265E+04 700.45 2.70014 1.1020499E+04 701.14 2.71010 1.1009593E+04 701.81 2.72005 1.0994752E+04 702.50 2.73010 1.0986299E+04 703.20 2.74015 1.0981456E+04 703.91 2.75009 1.0976227E+04 704.62 2.76018 1.0969251E+04 705.34 2.77022 1.0959540E+04 706.06 2.78007 1.0946830E+04 706.77 2.79002 1.0933509E+04 707.50 2.80001 1.0923801E+04 708.24 Amendment 61 Page 7 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-13 REACTOR CAVITY SUBCOMPARTMENT MASS AND ENERGY RELEASE FROM REACTOR VESSEL HOT LEG NOZZLE 150 IN2 BREAK TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.) 2.81001 1.0915605E+04 708.99 2.82004 1.0911554E+04 709.75 2.83006 1.0916234E+04 710.52 2.84005 1.0911481E+04 711.27 2.85002 1.0899235E+04 712.00 2.86100 1.0881568E+04 712.72 2.87025 1.0864099E+04 713.45 2.88021 1.0846173E+04 714.19 2.89017 1.0835129E+04 714.95 2.90016 1.0927215E+04 715.73 2.91016 1.0815027E+04 716.50 2.92019 1.0799955E+04 717.29 2.93008 1.0782872E+04 718.08 2.94006 1.0770177E+04 718.92 2.95017 1.0754933E+04 719.76 2.96001 1.0744940E+04 710.61 2.97003 1.0736345E+04 711.50 2.98006 1.0730031E+04 722.38 2.99001 1.0724229E+04 723.27 3.00001 1.0716026E+04 724.15 Amendment 61 Page 8 of 8

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-14 STEAM GENERATOR SUBCOMPARTMENT DOUBLE ENDED HOT LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .00000     1.1024200E+04          647.51
                     .00100     6.0345458E+04          645.59
                     .00200     8.4532187E+04          645.51
                     .00301     8.8837001E+04          645.01
                     .00401     8.5213433E+04          644.35
                     .00501     7.8995087E+04          644.16
                     .00601     7.3606154E+04          645.02
                     .00701     7.0689097E+04          646.42
                     .00800     7.0271191E+04          647.45
                     .00902     7.1149568E+04          647.90
                     .01002     7.2017922E+04          648.08
                     .01100     7.2586755E+04          648.30
                     .01200     7.2894144E+04          648.64
                     .01302     7.3063533E+04          649.06
                     .01402     7.3250654E+04          649.49
                     .01502     7.3566696E+04          649.87
                     .01602     7.4025054E+04          650.17
                     .01702     7.4541059E+04          650.40
                     .01801     7.5021506E+04          650.59
                     .01900     7.5446979E+04          650.75
                     .02000     7.5828421E+04          650.91
                     .02101     7.6175152E+04          651.07
                     .02202     7.6485820E+04          651.22
                     .02301     7.6762416E+04          651.38
                     .02400     7.7013151E+04          651.53
                     .02500     7.7239288E+04          651.69
                     .02603     7.7440885E+04          651.84
                     .02703     7.7612024E+04          651.98
                     .02801     7.7760440E+04          652.11
                     .02902     7.7892833E+04          652.24
                     .03003     7.8011961E+04          652.36
                     .03101     7.8118668E+04          652.47
                     .03202     7.8220542E+04          652.57
                     .03302     7.8311089E+04          652.67
                     .03400     7.8389499E+04          652.76
                     .03502     7.8455142E+04          652.85
                     .03602     7.8503090E+04          652.95
                     .03701     7.8533679E+04          653.04
                     .03803     7.8548369E+04          653.14
                     .03902     7.8549141E+04          653.24
                     .04002     7.8540922E+04          653.35
                     .04101     7.8528741E+04          653.47
                     .04201     7.8517443E+04          653.60
                     .04300     7.8513601E+04          653.75
                     .04404     7.8522438E+04          653.92
                     .04500     7.8547709E+04          654.10
                     .04602     7.8598019E+04          654.31
                     .04705     7.8678656E+04          654.53
                     .04802     7.8787056E+04          654.75
                     .04902     7.8940381E+04          654.99
                     .05002     7.9143466E+04          655.22 Amendment 61                                                              Page 1 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-14 STEAM GENERATOR SUBCOMPARTMENT DOUBLE ENDED HOT LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .05103     7.9402317E+04          655.45
                     .05204     7.9713095E+04          655.65
                     .05303     8.0060956E+04          655.83
                     .05402     8.0451400E+04          655.98
                     .05501     8.0884873E+04          656.10
                     .05603     8.1380615E+04          656.18
                     .05705     8.1908338E+04          656.23
                     .05800     8.2428466E+04          656.24
                     .05903     8.3013314E+04          656.22
                     .06001     8.3582930E+04          656.17
                     .06101     8.4173855E+04          656.10
                     .06205     8.4785291E+04          656.00
                     .06306     8.5372076E+04          655.90
                     .06404     8.5932593E+04          655.78
                     .06501     8.64659313+04          655.66
                     .06601     8.7009247E+04          655.52
                     .06703     8.7537158E+04          655.37
                     .06803     8.8021412E+04          655.22
                     .06903     8.8474746E+04          655.08
                     .07004     8.8844812E+04          654.92
                     .07102     8.9199095E+04          654.78
                     .07203     8.9563003E+04          654.65
                     .07302     8.9902776E+04          654.53
                     .07401     9.0213915E+04          654.41
                     .07506     9.0517151E+04          654.28
                     .07600     9.0762604E+04          654.18
                     .07707     9.1011445E+04          654.06
                     .07802     9.1211308E+04          653.96
                     .07903     9.1398073E+04          653.87
                     .08000     9.1559811E+04          653.78
                     .08102     9.1709583E+04          653.69
                     .08202     9.1838578E+04          653.61
                     .08300     9.1949386E+04          653.53
                     .08404     9.2048213E+04          653.45
                     .08500     9.2122263E+04          653.39
                     .08602     9.2181252E+04          653.32
                     .08704     9.2220281E+04          653.26
                     .08803     9.2240201E+04          653.20
                     .08901     9.2244722E+04          653.15
                     .09005     9.2235370E+04          653.09
                     .09109     9.2214176E+04          653.04
                     .09202     9.2186144E+04          653.00
                     .09309     9.2146638E+04          652.96
                     .09405     9.2104726E+04          652.92
                     .09503     9.2056998E+04          652.88
                     .09609     9.2001425E+04          652.84
                     .09708     9.1945643E+04          652.81
                     .09804     9.1889123E+04          652.78
                     .09900     9.1828949E+04          652.75
                     .10002     9.1763167E+04          652.72
                     .10509     9.1418960E+04          652.56
                     .11008     9.1082248E+04          652.36
                     .11509     9.0685692E+04          652.07
                     .12011     9.0047466E+04          651.73
                     .12500     8.9088521E+04          651.45 Amendment 61                                                              Page 2 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-14 STEAM GENERATOR SUBCOMPARTMENT DOUBLE ENDED HOT LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .13002     8.7917136E+04          651.25
                     .13506     8.6813094E+04          651.12
                     .14003     8.5889536E+04          651.01
                     .14504     8.5084409E+04          650.89
                     .15017     8.4363565E+04          650.73
                     .15511     8.3757206E+04          650.55
                     .16002     8.3313163E+04          650.39
                     .16507     8.2828675E+04          650.15
                     .17002     8.2435660E+04          649.91
                     .17510     8.2127364E+04          649.64
                     .18010     8.1907487E+04          649.35
                     .18516     8.1751221E+04          649.01
                     .19017     8.1616559E+04          648.64
                     .19514     8.1470640E+04          648.25
                     .20017     8.1296682E+04          647.85
                     .21007     8.0875772E+04          647.10
                     .22018     8.0352270E+04          646.38
                     .23003     7.9766161E+04          645.71
                     .24006     7.9126450E+04          645.06
                     .25004     7.8531830E+04          644.44
                     .26017     7.7993138E+04          643.78
                     .27002     7.7517288E+04          643.14
                     .28019     7.7043431E+04          642.49
                     .29021     7.6575871E+04          641.86
                     .30011     7.6116015E+04          641.27
                     .31021     7.5666476E+04          640.69
                     .32021     7.5251120E+04          640.13
                     .33015     7.4876385E+04          639.59
                     .34022     7.4523285E504          639.05
                     .35009     7.4191459E+04          638.54
                     .36023     7.3861019E+04          638.03
                     .37025     7.3549326E+04          637.56
                     .38001     7.3267132E+04          637.10
                     .39021     7.3000331E+04          636.63
                     .40012     7.2766750E+04          636.18
                     .41005     7.2549020E+04          635.75
                     .42009     7.2343573E+04          635.31
                     .43019     7.2134293E+04          634.89
                     .44013     7.1916544E+04          634.51
                     .45021     7.1684609E+04          634.16
                     .46022     7.1448775E+04          633.83
                     .47008     7.1219459E+04          633.53
                     .48022     7.0992367E+04          633.23
                     .49005     7.0779993E+04          632.95
                     .50023     7.0567669E+04          632.66
                     .51020     7.0358933E+04          632.39
                     .52028     7.0141691E+04          632.12
                     .53019     6.9916584E+04          631.88
                     .54022     6.9673933E+04          631.66
                     .55010     6.9419968E+04          631.45
                     .56013     6.9149595E+04          631.27
                     .57005     6.8874145E+04          631.10
                     .58013     6.8590390E+04          630.95
                     .59014     6.8309118E+04          630.81
                     .60009     6.8032346E+04          630.68 Amendment 61                                                              Page 3 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-14 STEAM GENERATOR SUBCOMPARTMENT DOUBLE ENDED HOT LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .61001     6.7760825E+04          630.56
                     .62016     6.7487651E+04          630.45
                     .63006     6.7226833E+04          630.35
                     .64001     6.6971944E+04          630.27
                     .65008     6.6724623E+04          630.21
                     .66012     6.6424666E+04          630.15
                     .67002     6.6204897E+04          630.13
                     .68015     6.6019069E+04          630.12
                     .69031     6.5852324E+04          630.15
                     .70023     6.5713585E+04          630.19
                     .71010     6.5593430E+04          630.27
                     .72031     6.5487015E+04          630.36
                     .73028     6.5411988E+04          630.49
                     .74005     6.5361694E+04          630.63
                     .75037     6.5320448E+04          630.80
                     .76009     6.5286330E+04          630.98
                     .77010     6.5250534E+04          631.17
                     .78008     6.5207849E+04          631.38
                     .79033     6.5150591E+04          631.60
                     .80001     6.5080698E+04          631.82
                     .81023     6.4988834E+04          632.06
                     .82010     6.4881089E+04          632.30
                     .83006     6.4752780E+04          632.54
                     .84017     6.4602522E+04          632.78
                     .85020     6.4433998E+04          633.01
                     .86021     6.4247611E+04          733.24
                     .87024     6.4044061E+04          633.46
                     .88008     6.3829954E+04          633.67
                     .89003     6.3601424E+04          633.88
                     .90012     6.3360257E+04          634.08
                     .91010     6.3115084E+04          634.28
                     .92002     6.2868417E+04          634.47
                     .93015     6.2616753E+04          634.66
                     .94004     6.2374563E+04          634.84
                     .95021     6.3132634E+04          635.03
                     .96022     6.1904266E+04          635.23
                     .97008     6.1690561E+04          635.42
                     .98006     6.1485413E+04          635.62
                     .99012     6.1289450E+04          635.83 1.00000     6.1106371E+04          636.03 1.10005     5.9398440E+04          638.07 1.20014     5.7440815E+04          640.04 1.30003     5.5876625E+04          643.14 1.40024     5.4101666E+04          646.93 1.50008     5.1897498E+04          650.14 1.60013     5.0012016E+04          651.58 1.70014     4.8545063E+04          652.06 1.80016     4.7316810E+04          652.68 1.90007     4.6196784E+04          653.17 2.00021     4.5167348E+04          653.27 2.10034     4.4196706E+04          652.95 2.20004     4.3278882E+04          652.46 2.30003     4.2403800E+04          652.07 2.40007     4.1594908E+04          651.90 2.50020     4.0877363E+04          651.91 Amendment 61                                                              Page 4 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-14 STEAM GENERATOR SUBCOMPARTMENT DOUBLE ENDED HOT LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.) 2.60005 4.0298576E+04 651.90 2.70029 3.9805653E+04 651.73 2.80021 3.9341807E+04 651.61 2.90022 3.8837692E+04 651.67 3.00001 3.8391596E+04 651.66 Note (1)Tabulated mass flow rates include a 10% margin, added by Westinghouse. For all subcompartment analyses, using the Double Ended Hot Leg Guillotine Break Mass Release Data, are reduced by a factor of 0.90910 to remove the 10% margin. Amendment 61 Page 5 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-15 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED PUMP SUCTION LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .00000     9.3240000E+03          558.10
                     .00101     4.0524818E+04          553.41
                     .00201     5.8641831E+04          553.56
                     .00301     6.9038021E+04          553.67
                     .00402     7.4940058E+04          553.71
                     .00501     7.8151491E+04          553.69
                     .00600     7.9777886E+04          553.62
                     .00702     8.0390754E+04          553.54
                     .00800     8.0396615E+04          553.44
                     .00903     8.0003216E+04          553.33
                     .01004     7.9368163E+04          553.22
                     .01101     7.8594360E+04          553.11
                     .01200     7.7704920E+04          553.01
                     .01302     7.6755870E+04          552.91
                     .01403     7.5810000E+04          552.83
                     .01503     7.4915768E+04          552.76
                     .01602     7.4093202E+04          552.73
                     .01707     7.3364284E+04          552.70
                     .01802     7.2784237E+04          552.70
                     .01904     7.2342149E+04          552.72
                     .02005     7.2028741E+04          552.75
                     .02102     7.1844018E+04          552.80
                     .02202     7.1756271E+04          552.84
                     .02300     7.1726549E+04          552.87
                     .02404     7.1790534E+04          552.90
                     .02505     7.1911886E+04          552.91
                     .02606     7.2047660E+04          552.91
                     .02703     7.2163902E+04          552.89
                     .02807     7.2265460E+04          552.88
                     .02902     7.2328098E+04          552.86
                     .03001     7.2363317E+04          552.83
                     .03104     7.2371342E+04          552.81
                     .03205     7.2356728E+04          552.80
                     .03301     7.2327175E+04          552.79
                     .03405     7.2282264E+04          552.78
                     .03500     7.2234161E+04          552.78
                     .03605     7.2177208E+04          552.79
                     .03704     7.2124971E+04          552.80
                     .03802     7.2077306E+04          552.81
                     .03904     7.2036558E+04          552.82
                     .04001     7.2007572E+04          552.83
                     .04102     7.1990994E+04          552.84
                     .04202     7.1987955E+04          552.85
                     .04302     7.1998413E+04          552.86
                     .04402     7.2020808E+04          552.87
                     .04501     7.2053373E+04          552.87
                     .04601     7.2094186E+04          552.87
                     .04702     7.2141437E+04          552.86
                     .04802     7.2189397E+04          552.86
                     .04902     7.2236163E+04          552.85 Amendment 61                                                              Page 1 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-15 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED PUMP SUCTION LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .05001     7.2279684E+04          552.85
                     .05102     7.2320339E+04          552.84
                     .05202     7.2356686E+04          552.83
                     .05301     7.2387088E+04          552.82
                     .05401     7.2410978E+04          552.82
                     .05501     7.2428213E+04          552.81
                     .05602     7.2437962E+04          552.80
                     .05701     7.2439388E+04          552.79
                     .05803     7.2432507E+04          552.79
                     .05902     7.2417619E+04          552.78
                     .06001     7.2395333E+04          552.77
                     .06102     7.2365952E+04          552.77
                     .06202     7.2330702E+04          552.76
                     .06303     7.2289539E+04          552.76
                     .06402     7.2244552E+04          552.76
                     .06503     7.2197072E+04          552.75
                     .06604     7.2149289E+04          552.75
                     .06703     7.2098784E+04          552.75
                     .06804     7.2048644E+04          552.75
                     .06904     7.2002443E+04          552.75
                     .07002     7.1957255E+04          552.75
                     .07102     7.1912108E+04          552.75
                     .07201     7.1869418E+04          552.75
                     .07307     7.1827168E+04          552.76
                     .07402     7.1791578E+04          552.76
                     .07502     7.1756338E+04          552.76
                     .07600     7.1723584E+04          552.77
                     .07705     7.1690533E+04          552.78
                     .07804     7.1661002E+04          552.79
                     .07905     7.1632229E+04          552.80
                     .08000     7.1605788E+04          552.81
                     .08100     7.1579246E+04          552.82
                     .08205     7.1552302E+04          552.83
                     .08308     7.1527092E+04          552.85
                     .08401     7.1505759E+04          552.86
                     .08501     7.1484586E+04          552.88
                     .08602     7.1464931E+04          552.90
                     .08708     7.1448041E+04          552.93
                     .08801     7.1436754E+04          552.95
                     .08907     7.1428702E+04          552.98
                     .09003     7.1426647E+04          553.00
                     .09105     7.1430726E+04          553.03
                     .09206     7.1441798E+04          553.06
                     .09302     7.1459175E+04          553.09
                     .09406     7.1486298E+04          553.12
                     .09502     7.1519647E+04          553.15
                     .09602     7.1561712E+04          553.18
                     .09707     7.1616307E+04          553.22
                     .09802     7.1675605E+04          553.24
                     .09909     7.1746181E+04          553.27
                     .10003     7.1816384E+04          553.30
                     .10207     7.1987044E+04          553.35
                     .10402     7.2159440E+04          553.39
                     .10604     7.2338291E+04          553.43 Amendment 61                                                              Page 2 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-15 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED PUMP SUCTION LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .10802     7.2510237E+04          553.46
                     .11000     7.2677439E+04          553.49
                     .11203     7.2842062E+04          553.51
                     .11400     7.2997323E+04          553.54
                     .11606     7.3153938E+04          553.55
                     .11803     7.3297562E+04          553.57
                     .12008     7.3440044E+04          553.59
                     .12207     7.3572665E+04          553.60
                     .12400     7.3695376E+04          553.61
                     .12603     7.3818278E+04          553.63
                     .12807     7.3936712E+04          553.64
                     .13007     7.4045274E+04          553.66
                     .13208     7.4156446E+04          553.67
                     .13407     7.4258593E+04          553.69
                     .13603     7.4355862E+04          553.71
                     .13807     7.4453117E+04          553.73
                     .14009     7.4545568E+04          553.75
                     .14204     7.4630838E+04          553.77
                     .14412     7.4715983E+04          553.80
                     .14609     7.4791570E+04          553.82
                     .14805     7.4858228E+04          553.85
                     .15009     7.4918108E+04          553.87
                     .15209     7.4967134E+04          553.90
                     .15409     7.5007709E+04          553.93
                     .15602     7.5039961E+04          553.96
                     .15803     7.5067434E+04          554.00
                     .16005     7.5090314E+04          554.03
                     .16211     7.5110508E+04          554.07
                     .16401     7.5127519E+04          554.11
                     .16604     7.5145772E+04          554.15
                     .16804     7.5164912E+04          554.20
                     .17003     7.5187669E+04          554.24
                     .17202     7.5210518E+04          554.29
                     .17401     7.5239158E+04          554.34
                     .17601     7.5272690E+04          554.39
                     .17801     7.5313914E+04          554.45
                     .18000     7.5369749E+04          554.54
                     .18200     7.5469826E+04          554.63
                     .18401     7.5643771E+04          554.74
                     .18608     7.5886838E+04          554.83
                     .18802     7.6139351E+04          554.90
                     .19001     8.1490935E+04          555.04
                     .19202     7.7895991E+04          555.02
                     .19401     8.0581493E+04          555.13
                     .19602     8.1949739E+04          555.40
                     .19800     8.1739509E+04          555.55
                     .20002     8.1667397E+04          555.57
                     .20502     8.2119006E+04          555.65
                     .21003     8.1677496E+04          555.72
                     .21502     8.0771243E+04          555.84
                     .22001     8.0233868E+04          556.02
                     .22504     8.0339564E+04          556.17
                     .23001     8.0505049E+04          556.31
                     .23502     8.0933790E+04          556.48 Amendment 61                                                              Page 3 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-15 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED PUMP SUCTION LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .24007     8.1724803E+04          556.67
                     .24508     8.1981692E+04          556.78
                     .25006     8.1622762E+04          556.87
                     .25504     8.1432305E+04          556.97
                     .26000     8.1444665E+04          557.07
                     .26501     8.1601549E+04          557.17
                     .27001     8.1796453E+04          557.26
                     .27506     8.1997200E+04          557.35
                     .28008     8.2109578E+04          557.41
                     .28501     8.1936018E+04          557.45
                     .29000     8.1499004E+04          557.49
                     .29501     8.1082489E+04          557.55
                     .30010     8.0692508E+04          557.66
                     .30503     8.0211034E+04          557.79
                     .31010     7.9514336E+04          557.94
                     .31508     7.8733516E+04          558.11
                     .32009     7.7920017E+04          558.31
                     .32510     7.7155883E+04          558.52
                     .33006     7.6529340E+04          558.74
                     .33503     7.6092806E+04          558.97
                     .34003     7.5667555E+04          559.20
                     .34507     7.5164458E+04          559.43
                     .35001     7.4733004E+04          559.67
                     .35513     7.4381709E+04          559.93
                     .36010     7.4077393E+04          560.18
                     .36500     7.3799692E+04          560.43
                     .37004     7.3520636E+04          560.70
                     .37509     7.3234390E+04          560.95
                     .38012     7.2959480E+04          561.20
                     .38509     7.2725426E+04          561.45
                     .39000     7.2521810E+04          561.68
                     .39505     7.2329933E+04          561.90
                     .40001     7.2166141E+04          562.12
                     .40515     7.2060184E+04          562.35
                     .41001     7.1991945E+04          562.57
                     .41504     7.1859041E+04          562.79
                     .42000     7.1695083E+04          563.01
                     .42506     7.1516928E+04          563.23
                     .43003     7.1341627E+04          563.44
                     .43505     7.1184213E+04          563.66
                     .44010     7.1020662E+04          563.88
                     .44506     7.0864483E+04          564.09
                     .45011     7.0730762E+04          564.32
                     .45507     7.0631767E+04          564.54
                     .46002     7.0602771E+04          564.78
                     .46509     7.0546227E+04          565.00
                     .47002     7.0444621E+04          565.21
                     .47512     7.0363249E+04          565.41
                     .48003     7.0294940E+04          565.60
                     .48501     7.0235997E+04          565.78
                     .49005     7.0145287E+04          565.96
                     .49501     7.0029850E+04          566.14
                     .50008     6.9884963E+04          566.33
                     .51003     7.0039203E+04          566.77 Amendment 61                                                              Page 4 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-15 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED PUMP SUCTION LEG GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                     .52003     6.9731868E+04          567.18
                     .53012     6.9423078E+04          567.63
                     .54001     6.9140583E+04          568.12
                     .55007     6.8889484E+04          568.63
                     .56019     6.8664647E+04          569.12
                     .57017     6.8528725E+04          569.62
                     .58003     6.8507295E+04          570.13
                     .59009     6.8463834E+04          570.62
                     .60006     6.8378898E+04          571.10
                     .61000     6.8311163E+04          571.58
                     .62011     6.8272669E+04          572.06
                     .63012     6.8246816E+04          572.54
                     .64021     6.8221713E+04          573.00
                     .65009     6.8181963E+04          573.46
                     .66002     6.8132677E+04          573.91
                     .67012     6.8079209E+04          574.36
                     .68001     6.8025255E+04          574.78
                     .69008     6.7967724E+04          575.21
                     .70008     6.7905900E+04          575.63
                     .71002     6.7837950E+04          576.04
                     .72012     6.7760362E+04          576.45
                     .73012     6.7674460E+04          576.85
                     .74012     6.7581245E+04          577.24
                     .75009     6.7482820E+04          577.63
                     .76005     6.7383600E+04          578.02
                     .77010     6.7284210E+04          578.40
                     .78010     6.7185273E+04          578.78
                     .79008     6.7084243E+04          579.14
                     .80000     6.6983330E+04          579.49
                     .81003     6.6873004E+04          579.85
                     .82006     6.6759893E+04          580.20
                     .83007     6.6641933E+04          580.53
                     .84008     6.6518861E+04          580.87
                     .85012     6.6390083E+04          581.19
                     .86008     6.6258298E+04          581.52
                     .87005     6.6125310E+04          581.84
                     .88010     6.5990875E+04          582.15
                     .89006     6.5857848E+04          582.46
                     .90013     6.5723411E+04          582.76
                     .91009     6.5588268E+04          583.06
                     .92009     6.5451003E+04          583.35
                     .93004     6.5312934E+04          583.63
                     .94013     6.5170572E+04          583.91
                     .95000     6.5028819E+04          584.19
                     .96004     6.4882992E+04          584.46
                     .97007     6.4736852E+04          584.73
                     .98011     6.4590890E+04          584.99
                     .99066     6.4446435E+04          585.25 1.00012     6.4302055E+04          585.50 Amendment 61                                                              Page 5 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-16 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED COLD LET GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                         .00000    1.1324200E+04           553.81
                         .00100    4.4570762E+04           549.06
                         .00201    5.8805967E+04           549.08
                         .00301    6.4643873E+04           549.02
                         .00401    6.6481865E+04           548.88
                         .00501    6.6210787E+04           548.67
                         .00602    6.4790242E+04           548.40
                         .00700    6.2800952E+04           548.11
                         .00801    6.0591591E+04           547.90
                         .00901    5.8565061E+04           547.81
                         .01001    5.6928264E+04           547.88
                         .01101    5.5882027E+04           548.05
                         .01201    5.5469769E+04           548.28
                         .01301    5.5592852E+04           548.47
                         .01401    5.6036132E+04           548.59
                         .01512    5.6607215E+04           548.64
                         .01602    5.7144545E+04           548.66
                         .01703    5.7621577E+04           548.66
                         .01801    5.8010549E+04           548.67
                         .01905    5.8371809E+04           548.69
                         .02001    5.8677514E+04           548.72
                         .02102    5.8981775E+04           548.75
                         .02204    5.9293260E+04           548.81
                         .02301    5.9600001E+04           548.87
                         .02401    5.9927980E+04           548.92
                         .02504    6.0272597E+04           548.97
                         .02602    6.0612532E+04           549.01
                         .02701    6.0946329E+04           549.04
                         .02802    6.1263000E+04           549.05
                         .02901    6.1590646E+04           549.07
                         .03000    6.1885777E+04           549.08
                         .03104    6.2167943E+04           549.10
                         .03200    6.2431527E+04           549.11
                         .03301    6.2676997E+04           549.11
                         .03401    6.2902121E+04           549.13
                         .03502    6.3110234E+04           549.14
                         .03600    6.3308269E+04           549.16
                         .03703    6.3499393E+04           549.19
                         .03801    6.3672417E+04           549.21
                         .03903    6.3841553E+04           549.23
                         .04004    6.4101957E+04           549.26
                         .04101    6.4153357E+04           549.29
                         .04205    6.4312203E+04           549.33
                         .04304    6.4460158E+04           549.36
                         .04406    6.4610597E+04           549.39
                         .04507    6.4756613E+04           549.42
                         .04601    6.4894032E+04           549.46
                         .04701    6.5051944E+04           549.49
                         .04800    6.5200557E+04           549.53
                         .04900    6.5355294E+04           549.62
                         .05000    6.5525905E+04           549.76
                         .05100    6.5903704E+04           550.45
                         .05201    6.6627876E+04           550.55
                         .05300    6.7402684E+04           550.59
                         .05401    6.8107118E+04           550.60
                         .05501    7.6646458E+04           554.68 Amendment 61                                                                 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-16 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED COLD LET GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                         .05601    8.8017305E+04           550.87
                         .05701    8.4391009E+04           551.90
                         .05801    8.7244787E+04           550.76
                         .05900    8.7601522E+04           551.41
                         .06000    8.9796330E+04           551.21
                         .06100    8.9630540E+04           550.88
                         .06201    8.8424914E+04           550.82
                         .06302    8.8446761E+04           550.82
                         .06401    8.8933971E+04           550.84
                         .06500    8.9486174E+04           550.81
                         .06600    8.9042705E+04           550.61
                         .06702    8.9536751E+04           550.81
                         .06802    9.0402921E+04           550.88
                         .06901    9.1428447E+04           550.85
                         .07003    9.1832417E+04           550.83
                         .07105    9.2055693E+04           550.82
                         .07200    9.2489197E+04           550.84
                         .07302    9.2906966E+04           550.83
                         .07401    9.3094290E+04           550.85
                         .07505    9.3367771E+04           550.82
                         .07605    9.3858842E+04           550.86
                         .07705    9.4461599E+04           550.90
                         .07803    9.5062428E+04           550.92
                         .07903    9.5553439E+04           550.96
                         .08007    9.6195924E+04           551.00
                         .08105    9.6695119E+04           551.00
                         .08206    9.6963151E+04           550.97
                         .08304    9.7089364E+04           550.95
                         .08403    9.7243229E+04           550.94
                         .08507    9.7467380E+04           550.95
                         .08601    9.7696605E+04           550.95
                         .08702    9.7969430E+04           550.96
                         .08806    9.8279560E+04           550.98
                         .08908    9.8557361E+04           550.98
                         .09007    9.8727314E+04           550.96
                         .09107    9.8790135E+04           550.93
                         .09202    9.8782186E+04           550.90
                         .09303    9.8735463E+04           550.87
                         .09407    9.8678267E+04           550.85
                         .09501    9.8664334E+04           550.83
                         .09605    9.8735706E+04           550.84
                         .09706    9.8887470E+04           550.85
                         .09806    9.9066325E+04           550.86
                         .09905    9.9233805E+04           550.86
                         .10000    9.9350001E+04           550.85
                         .10507    9.9499448E+04           550.78
                         .11011    1.0010712E+05           550.78
                         .11506    1.0031991E+05           550.74
                         .12008    1.0095196E+05           550.79
                         .12507    1.0166041E+05           550.87
                         .13006    1.0251943E+05           550.99
                         .13501    1.0306787E+05           551.08
                         .14009    1.0340576E+05           551.15
                         .14503    1.0363687E+05           551.19
                         .15011    1.0397515E+05           551.23
                         .15509    1.0424490E+05           551.23 Amendment 61                                                                 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-16 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED COLD LET GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                         .16002    1.0438370E+05           551.20
                         .16508    1.0440389E+05           551.17
                         .17002    1.0449549E+05           551.16
                         .17515    1.0471349E+05           551.17
                         .18006    1.0497644E+05           551.18
                         .18513    1.0517078E+05           551.18
                         .19002    1.0528063E+05           551.17
                         .19511    1.0536931E+05           551.16
                         .20016    1.0546367E+05           551.15
                         .21008    1.0549102E+05           551.11
                         .22011    1.0533798E+05           551.07
                         .23014    1.0541258E+05           551.09
                         .24018    1.0588407E+05           551.19
                         .25004    1.0624301E+05           551.24
                         .26017    1.0605490E+05           551.22
                         .27012    1.0573770E+05           551.18
                         .28016    1.0564751E+05           551.18
                         .29016    1.0579764E+05           551.19
                         .30001    1.0582758E+05           551.18
                         .31002    1.0567966E+05           551.16
                         .32006    1.0539162E+05           551.13
                         .33004    1.0506220E+05           551.13
                         .34007    1.0475672E+05           551.19
                         .35002    1.0444833E+05           551.09
                         .36009    1.0403821E+05           551.07
                         .37009    1.0357630E+05           551.05
                         .38001    1.0329455E+05           551.05
                         .39006    1.0328083E+05           551.13
                         .40001    1.0334067E+05           551.14
                         .41002    1.0324684E+05           551.15
                         .42013    1.0302244E+05           551.14
                         .43008    1.0279151E+05           551.14
                         .44011    1.0252723E+05           551.14
                         .45004    1.0216415E+05           551.14
                         .46013    1.0173280E+05           551.14
                         .47009    1.0133983E+05           551.15
                         .48014    1.0103725E+05           551.17
                         .49001    1.0078626E+05           551.19
                         .50015    1.0050443E+05           551.21
                         .51007    1.0015614E+05           551.22
                         .52006    9.9740042E+04           551.24
                         .53015    9.9340565E+04           551.26
                         .54006    9.8996919E+04           551.29
                         .55014    9.8648668E+04           551.32
                         .56011    9.8275012E+04           551.36
                         .57009    9.7897646E+04           551.40
                         .58020    9.7795723E+04           551.48
                         .59014    9.7935754E+04           551.57
                         .60010    9.7993358E+04           551.63
                         .61001    9.7897358E+04           551.67
                         .62017    9.7802629E+04           551.73
                         .63013    9.7801473E+04           551.80
                         .64000    9.7822131E+04           551.87
                         .65005    9.7783336E+04           551.94
                         .66013    9.7622208E+04           552.00
                         .67009    9.7482789E+04           552.06 Amendment 61                                                                 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-16 STEAM GENERATOR SUBCOMPARTMENT DOUBLE-ENDED COLD LET GUILLOTINE BREAK - MASS AND ENERGY RELEASE DATA TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                                  .68001      9.7261452E+04                 552.13
                                  .69013      9.7237632E+04                 552.21
                                  .70015      9.7153080E+04                 552.28
                                  .71011      9.6924792E+04                 552.34
                                  .72004      9.6704687E+04                 552.42
                                  .73000      9.6622454E+04                 552.52
                                  .74006      9.6587028E+04                 552.62
                                  .75006      9.6489190E+04                 552.71
                                  .76009      9.6368978E+04                 552.81
                                  .77009      9.6303788E+04                 552.91
                                  .78001      9.6265856E+04                 553.01
                                  .79012      9.6178043E+04                 553.10
                                  .80000      9.6015706E+04                 553.21
                                  .81012      9.5810031E+04                 553.30
                                  .82000      9.5630221E+04                 553.42
                                  .83012      9.5499675E+04                 553.54
                                  .84004      9.5363709E+04                 553.66
                                  .85005      9.5164832E+04                 553.77
                                  .86011      9.4933621E+04                 553.89
                                  .87009      9.4735902E+04                 554.02
                                  .88006      9.4551854E+04                 554.15
                                  .89004      9.4326286E+04                 554.28
                                  .90008      9.4062638E+04                 554.42
                                  .91009      9.3815232E+04                 554.57
                                  .92006      9.3668819E+04                 554.72
                                  .93005      9.3418315E+04                 554.86
                                  .94004      9.3112967E+04                 555.01
                                  .95013      9.2726414E+04                 555.16
                                  .96005      9.2348072E+04                 555.31
                                  .97001      9.2008405E+04                 555.47
                                  .98008      9.1668053E+04                 555.63
                                  .99002      9.1301622E+04                 555.80 1.00008      9.0909494E+04                 555.98 1.10004      8.8562339E+04                 557.85 1.20017      8.5142881E+04                 559.79 1.30011      8.4053473E+04                 561.78 1.40004      8.1079123E+04                 563.86 1.50002      7.8671323E+04                 565.77 1.60002      7.5824157E+04                 567.54 1.70003      7.0812834E+04                 569.50 1.80013      6.8119105E+04                 571.19 1.90001      6.5842281E+04                 572.74 2.00002      6.1686321E+04                 574.53 2.10003      6.0056342E+04                 575.88 2.20006      5.8469684E+04                 577.33 2.30000      5.6960079E+04                 579.33 2.40013      5.5545962E+04                 581.72 2.50018      5.4351353E+04                 584.79 2.60023      5.2420102E+04                 588.92 2.63024      5.1916010E+04                 590.25 Notes:

(1)Tabulated mass flow rates include a 10% margin, added by Westinghouse. For all subcompartment analyses using the Double Ended Cold Leg Guillotine Break blowdown data, mass flow rates are reduced by a factor of 0.9091 to remove the 10% margin. Amendment 61 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-17 PRESSURIZER SUBCOMPARTMENT DOUBLE-ENDED SURGE LINE GUILLOTINE BREAK MASS AND ENERGY RELEASE RATES FOR ORIGINAL DESIGN BASES TIMES(S) MASS FLOW (lb./S) AVG ENTHALPY (Btu/lb.)

                       .00000            0.                 0.00
                       .00501      1.5865294E+04           679.56
                       .01102      2.0984553E+04           672.55
                       .01502      2.0322874E+04           672.22
                       .02003      1.9080382E+04           672.68
                       .02501      1.9588253E+04           671.76
                       .03005      2.0057482E+04           671.02
                       .03501      2.0522881E+04           670.40
                       .04110      2.0931338E+04           669.88
                       .05003      2.083487 E+04           669.74
                       .06007      2.0274783E+04           670.06
                       .07009      2.0007795E+04           670.26
                       .08001      1.9772454E+04           670.45
                       .09011      1.9213386E+04           670.97
                       .10007      1.9518224E+04           670.68
                       .15004      1.7950590E+04           672.12
                       .20001      1.6006467E+04           674.98
                       .30032      1.5635467E+04           675.35
                       .40050      1.5562509E+04           675.11
                       .50022      1.5499026E+04           674.85
                       .60041      1.5421848E+04           674.56
                       .70011      1.5366345E+04           674.22
                       .80033      1.5312382E+04           673.88
                       .90031      1.5242365E+04           673.55 1.00009      1.5175989E+04           673.22 2.00007      1.4471182E+04           669.87 Notes:

(1)Tabulated mass flow rates include 10% margin, added by Westinghouse. For the pressurizer subcompartment analysis using the double ended surge line guillotine break, the mass flow rates reduced by a factor of 0.90910 to remove the 10% margin. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-17a Pressurizer Subcompartment Double-Ended Surge Line Guillotine Break Mass and Energy Releases for SGR/PUR Time (sec) Break Flow (lbm/sec) Break Energy (Btu/sec) Enthalpy (Btu/lbm) 0.00000 0.0000000E+00 0.0000000E+00 0.00 0.00100 2.1103676E+03 1.3534502E+06 641.33 0.00904 9.4237946E+03 6.0692558E+06 644.04 0.01004 9.8820135E+03 6.3686254E+06 644.47 0.02004 1.3838759E+04 8.9630591E+06 647.68 0.03003 1.4863396E+04 9.6634563E+06 650.15 0.04003 1.6093160E+04 1.0502445E+07 652.60 0.05001 1.6159154E+04 1.0577799E+07 654.60 0.06002 1.6553060E+04 1.0864070E+07 656.32 0.07001 1.6655118E+04 1.0953752E+07 657.68 0.08006 1.6433660E+04 1.0835927E+07 659.37 0.09004 1.6416382E+04 1.0846517E+07 660.71 0.10002 1.6907343E+04 1.1180615E+07 661.29 0.20005 1.4938162E+04 1.0006279E+07 669.85 0.30001 1.4111884E+04 9.4848110E+06 672.12 0.40005 1.4341020E+04 9.6094731E+06 670.07 0.50001 1.4291815E+04 9.5529874E+06 668.42 0.60001 1.4319763E+04 9.5477666E+06 666.75 0.70001 1.4293723E+04 9.5143628E+06 665.63 0.80003 1.4305338E+04 9.5119696E+06 664.92 0.90017 1.4272873E+04 9.4888186E+06 664.81 1.00000 1.4151129E+04 9.4151449E+06 665.33 1.10016 1.3962764E+04 9.3022100E+06 666.22 1.20014 1.3846967E+04 9.2358276E+06 666.99 1.30006 1.3682495E+04 9.1379399E+06 667.86 1.40019 1.3597790E+04 9.0918136E+06 668.62 1.50001 1.3546720E+04 9.0685560E+06 669.43 1.60021 1.3469738E+04 9.0282535E+06 670.26 1.70011 1.3380283E+04 8.9786760E+06 671.04 1.90006 1.3193815E+04 8.8690636E+06 672.21 2.00049 1.3130329E+04 8.8308523E+06 672.55 Note: The tabulated energy releases should be increased 8.35% in order to bound operation at a Tavg of 572°F with a -6.0°F temperature uncertainty. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-18 PRESSURIZER SUBCOMPARTMENT SPRAY LINE DOUBLE ENDED PRESSURIZER BREAK - MASS AND ENERGY RELEASE RATES FOR ORIGINAL DESIGN BASES 3 TIMES(S) MASS FLOW (lb./S) x 10 AVG ENTHALPY (Btu/lb.)

               .00000             0.                      0.00
               .05003          4.864                     611.05
               .10011          4.980                     609.82
               .15001          4.740                     612.01
               .20062          4.726                     612.03
               .25024          4.734                     611.80
               .30016          4.644                     612.62
               .35000          4.667                     612.22
               .40023          4.650                     612.26
               .45080          4.670                     611.88
               .50011          4.689                     611.52
               .55038          4.671                     611.56
               .60015          4.677                     611.34
               .65025          4.671                     611.25
               .70026          4.667                     611.14
               .75041          4.665                     611.02
               .80005          4.650                     611.05
               .85031          4.645                     610.95
               .90033          4.635                     610.93
               .95038          4.622                     610.93 1.00004          4.617                     610.86 1.05002          4.599                     610.93 1.10003          4.589                     610.90 1.15004          4.579                     610.89 1.20016          4.565                     610.95 1.25006          4.556                     610.91 1.30023          4.543                     610.94 1.35011          4.530                     610.97 1.40018          4.522                     610.95 1.45020          4.508                     611.00 1.50020          4.497                     611.02 1.55112          4.487                     611.02 1.60021          4.474                     611.07 1.65033          4.463                     611.09 1.70032          4.453                     611.11 1.76012          4.440                     611.17 1.80021          4.429                     611.20 1.85040          4.418                     611.24 1.90010          4.404                     611.30 1.95032          4.393                     611.34 2.00001          4.382                     611.40 Amendment 61                                                                   Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-18a Pressurizer Subcompartment Spray Double-Ended Pressurizer Break Mass and Energy Release Rates for SGR/PUR Time (sec) Break Flow (lb/sec) Break Energy (Btu/sec) Avg. Enthalpy (Btu/lb) 0.00000 0.0000000E+00 0.0000000E+00 0.00 0.01103 5.5518493E+03 3.2905004E+06 592.69 0.02003 5.7362822E+03 3.3924253E+06 591.40 0.03004 5.7848213E+03 3.4191326E+06 591.05 0.04012 5.8059940E+03 3.4305494E+06 590.86 0.05008 5.6819445E+03 3.3609717E+06 591.52 0.06007 5.6565506E+03 3.3467035E+06 591.65 0.07002 5.7317091E+03 3.3884934E+06 591.18 0.08006 5.7132020E+03 3.3780441E+06 591.27 0.09004 5.6174032E+03 3.3245990E+06 591.84 0.10006 5.5866929E+03 3.3074862E+06 592.03 0.20007 5.5623287E+03 3.2933991E+06 592.09 0.30000 5.4512179E+03 3.2308583E+06 592.69 0.40009 5.3230373E+03 3.1592006E+06 593.50 0.50001 5.2181352E+03 3.1007846E+06 594.23 0.60009 5.1229225E+03 3.0478652E+06 594.95 0.70015 5.0239055E+03 2.9929317E+06 595.74 0.80014 4.9453679E+03 2.9493564E+06 596.39 0.90018 4.8895171E+03 2.9183478E+06 596.86 1.00013 4.9513805E+03 2.9521263E+06 596.22 1.10003 4.9736299E+03 2.9641328E+06 595.97 1.20003 5.0045052E+03 2.9809611E+06 595.66 1.30005 5.0311580E+03 2.9955387E+06 595.40 1.40002 5.0433071E+03 3.0021727E+06 595.28 1.50009 5.0585074E+03 3.0105403E+06 595.14 1.60003 5.0654476E+03 3.0143923E+06 595.09 1.70006 5.0667942E+03 3.0151892E+06 595.09 1.80003 5.0675444E+03 3.0156876E+06 595.10 1.90011 5.0613076E+03 3.0123487E+06 595.17 2.00022 5.0532974E+03 3.0080471E+06 595.26 Note: The tabulated energy releases should be increased 0.4% in order to bound operation at a Tavg of 572°F with a -6.0°F temperature uncertainty. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-19 REACTOR CAVITY SUBCOMPARTMENT PRESSURIZATION MODEL RELAP-4 VOLUME INPUT DATA Analysis(1) GA Drawing Net Free Vol. Height Vol. Floor Volume No. Vol. No. Volume (ft.2) (ft.) Elev. (ft.) 01 2 4250.18 12.00 211.50 02 1 2224.30 12.00 211.50 03 3a 43.55 10.29 223.50 04 3b 74.50 10.29 223.50 05 3c 48.55 10.29 223.50 06 3d 74.50 10.29 223.50 07 3e 48.55 10.29 223.50 08 3f 74.50 10.29 223.50 09 4a 7.73 8.52 223.79 10 4b 5.82 8.52 233.79 11 4c 7.73 8.52 233.79 12 4d 12.82 8.52 233.79 13 4e 7.73 8.52 233.79 14 4f 6.82 8.52 233.79 15 5a 8.56 6.82 242.31 16 5b 11.99 6.82 242.31 17 5c 8.56 6.82 242.31 18 5d 11.99 6.82 242.31 19 5e 8.56 6.82 242.31 20 5f 11.99 6.82 242.31 21 6a 379.59 11.07 249.13 22 6b 559.26 11.07 249.13 23 6c 378.59 11.07 249.13 24 6d 565.11 11.07 249.13 25 6e 378.59 11.07 249.13 26 6f 560.26 11.07 249.13 27 7a 1233.07 25.80 260.20 28 7b 1818.83 25.80 260.20 29 7c 1185.92 25.80 260.20 30 7d 1672.06 25.80 260.20 31 7e 1178.33 25.80 260.20 32 7f 30861.21 25.80 260.20 33 Vol. outside of 2252715.00 220.000 221.000 Reactor Cavity NOTE: 1. See Figure 6.2.1-22 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-20 REACTOR CAVITY SUBCOMPARTMENT PRESSURIZATION MODEL RELAY-4 JUNCTION INPUT DATA Junction Volume Inertia Loss Coefficient No. From To Area Elev. Coefficient Forward Reverse 01 27 33 13.430 286.000 0.2729 1.3647 1.0232 02 28 33 48.980 286.000 0.1885 1.1805 0.6071 03 29 33 41.060 286.000 0.2729 1.0467 0.5219 04 30 33 57.480 286.000 0.1966 1.0442 0.5212 05 31 33 41.060 286.000 0.2729 1.0467 0.5219 06 32 33 820.350 286.000 0.0192 0.9234 0.4829 07 21 27 34.270 260.200 0.4044 0.0220 0.0353 08 22 28 50.530 260.200 0.2807 0.0226 0.0362 09 23 29 34.270 260.200 0.4044 0.0220 0.0353 10 24 30 47.980 260.200 0.2888 0.0220 0.0352 11 25 31 34.270 260.200 0.4044 0.0131 0.0353 12 26 32 47.980 260.200 0.1114 0.9119 0.4840 13 15 21 0.700 249.130 2.4259 0.9991 0.6985 14 16 22 0.980 249.130 1.7296 0.9990 0.6978 15 17 23 0.700 249.130 2.4259 0.9991 0.6985 16 18 24 0.980 249.130 1.7296 0.9990 0.6978 17 19 25 0.700 249.130 2.4259 0.9991 0.6985 18 20 26 0.980 249.130 1.7296 0.9990 0.6978 19 9 15 1.190 242.310 5.1477 0.2175 0.2175 20 10 16 1.670 242.310 3.6699 0.2234 0.2234 21 11 17 1.190 242.310 5.1477 0.2175 0.2175 22 12 18 1.670 242.310 3.6699 0.2234 0.2234 23 13 19 1.190 242.310 5.1477 0.2175 0.2175 24 14 20 1.670 242.310 3.6699 0.2234 0.2234 25 3 9 1.190 233.790 3.0554 0.5981 1.0321 26 4 10 1.670 233.790 2.1785 0.5980 1.0318 27 5 11 1.190 233.790 3.0554 0.5981 1.0321 28 6 12 1.670 233.790 2.1785 0.5981 1.0319 29 7 13 1.190 233.790 3.0554 0.5981 1.0313 30 8 14 1.670 233.790 2.1785 0.5981 1.0319 31 2 3 22.260 223.500 5.6630 0.3485 0.6499 32 2 4 31.160 223.500 4.0460 0.3519 0.6533 33 2 5 22.260 223.500 5.6630 0.3485 0.6499 34 2 6 31.140 223.500 4.0460 0.3514 0.6524 35 2 7 22.260 223.500 5.6630 0.3485 0.6499 36 2 8 31.140 223.500 4.0460 0.3514 0.6524 37 32 27 210.610 273.100 0.0690 0.3444 0.4547 38 26 21 27.060 254.670 0.2149 0.4458 0.4458 39 20 15 1.410 245.720 5.2278 0.3027 0.3027 40 14 9 1.620 238.050 4.1877 0.2569 0.2569 41 8 3 14.750 228.650 0.4589 0.2423 0.2423 42 27 28 122.880 273.100 0.0742 0.0172 0.0172 43 21 22 26.780 254.670 0.2149 0.4537 0.4537 44 15 16 1.410 245.720 5.2278 0.3025 0.3025 45 9 10 1.620 238.050 4.1877 0.2568 0.2568 46 3 4 14.750 228.650 0.4589 0.2423 0.2423 47 28 29 122.880 273.100 0.0742 0.0172 0.0172 48 22 23 27.060 254.670 0.2149 0.4458 0.4458 49 16 17 1.410 245.720 5.2278 0.3027 0.3027 50 10 11 1.620 238.050 4.1877 0.2569 0.2569 51 4 5 14.750 228.650 0.4589 0.2423 0.2423 52 29 30 122.880 273.100 0.0742 0.0172 0.0172 53 23 24 26.780 254.670 0.2149 0.4537 0.4537 Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-20 REACTOR CAVITY SUBCOMPARTMENT PRESSURIZATION MODEL RELAY-4 JUNCTION INPUT DATA Junction Volume Inertia Loss Coefficient No. From To Area Elev. Coefficient Forward Reverse 54 17 18 1.410 245.720 5.2278 0.3027 0.3027 55 11 12 1.620 238.050 4.1877 0.2569 0.2569 56 5 6 14.750 228.650 0.4589 0.2423 0.2423 57 30 31 122.880 273.100 0.0742 0.0172 0.0172 58 24 25 27.060 254.670 0.2149 0.4458 0.4458 59 18 19 1.410 245.720 5.2278 0.3026 0.3026 60 12 13 1.620 238.050 4.1877 0.2568 0.2568 61 6 7 14.750 228.650 0.4589 0.2422 0.2422 62 31 32 210.610 273.100 0.0690 0.4547 0.3443 63 25 26 26.780 254.670 0.2149 0.4537 0.4537 64 19 20 1.410 245.720 5.2278 0.3026 0.3026 65 13 14 1.620 238.050 4.1877 0.2569 0.2569 66 7 8 14.750 228.650 0.4589 0.2422 0.2422 67 2 1 152.873 217.500 0.107 0.0504 0.1097 68 21 33 2.5252 253.750 1.787 1.6899 1.6899 69 22 33 2.2814 253.750 1.965 1.9007 1.9007 70 23 33 2.9708 253.750 1.552 1.6643 1.6643 71 24 33 2.2814 253.700 1.965 1.9007 1.9007 72 25 33 2.5252 253.700 1.552 1.6643 1.6643 73 26 33 2.2814 253.700 3.009 2.0087 2.0086 74 1 33 30.0000 223.500 0.040 1.5000 1.5000 75 0 23 0.5000 253.750 0.000 0.0000 0.0000 76 0 24 0.5000 253.750 0.000 0.0000 0.0000 Fill Junctions for Cold Leg Guill. Break 75 0 22 0.5000 253.750 0.0 0.0 0.0 76 0 23 0.5000 253.750 0.0 0.0 0.0 Fill Junctions for Hot Leg Guill. Break Note: (1) See Figure 6.2.1-22 Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-20A LIST OF PROJECTED AREAS Projected Area for Force Calculation (in2) Volume Number in X-direction in Y-direction in Z-direction 3 6448.26 4515.12 4042.21 4 931.14 10643.03 5659.09 5 7134.34 3326.80 4042.21 6 8751.57 6127.91 5659.09 7 686.08 7841.92 4042.21 8 9682.71 4515.12 5659.09 9 6428.62 4501.37 N/A 10 928.31 10610.62 N/A 11 7112.61 3316.67 N/A 12 8724.91 6109.25 N/A 13 683.99 7818.04 N/A 14 9653.22 4501.37 N/A 15 5140.38 3599.34 N/A 16 742.28 8484.35 N/A 17 5687.31 2652.04 N/A 18 6976.52 4885.01 N/A 19 546.92 6251.37 N/A 20 7718.81 3599.34 N/A 21 8855.03 6200.36 N/A 22 1278.69 14615.47 N/A 23 9797.18 4568.50 N/A 24 12018.03 8415.11 N/A 25 942.15 10768.87 N/A 26 13296.71 6200.36 N/A 27 6364.25 4456.29 4042.21 28 919.01 10504.37 5659.09 29 7041.38 3283.45 4042.21 30 8637.54 6048.07 5659.09 31 677.14 7739.75 4042.21 32 9556.55 4456.29 5659.09 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-20B LIST OF LEVER ARMS Volume Number Lever Arms (ft) relative to nozzle elevation (253.75') 3,4,5,6,7,8 -24.24 9,10,11,12,13,14 -15.70 15,16,17,18,19,20 - 8.03 21,22,23,24,25,26 + 0.92 27,28,29,30,31,32 10.43 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-21 STEAM GENERATOR - LOOP 1 SUBCOMPARTMENT ANALYSIS-VOLUME INPUT DATA Net Free Volume (1) Analysis Volume Vol. Height Floor Elev. (2) 3 Vol. No. GA Drawing Vol. No. (Ft. ) (Ft.) (Ft.) 01 11 8093.87 15.000 221.000 02 13 6034.32 15.000 221.000 03 71 10412.70 23.500 221.000 04 15 4041.25 61.000 221.000 05 21 1288.29 17.750 236.000 06 2 4250.48 12.000 211.500 07 12 18857.39 35.000 221.000 08 10 22158.22 41.500 221.000 09 9 4386.33 15.000 221.000 10 22 499.66 17.750 236.000 11 16 5421.03 61.000 221.000 12 20 4049.88 17.750 236.000 13 70 1694856.7 159.000 282.000 14 23,24,25,26,40,41,42,43, 51912.30 76.750 236.000 57,58,59,60,8(a to d) 15 27,28,29,30,31,32,44,45, 24833.40 46.000 236.000 46,47,48,49,61,62,63,64,65,66 16 1,3 (a to f), 4 (a to f), 2703.75 37.630 211.500 5 (a to f) 17 6 (a to f), 7 (a to f) 40770.82 36.870 249.130 18 14,33,50 365478.26 65.000 221.000 19 67b 8155.33 35.450 298.670 20 67a 2748.20 16.670 282.000 21 51 452.87 8.000 274.000 22 52 503.50 8.000 274.000 23 53 951.99 8.000 274.000 24 54 1343.48 8.000 274.000 25 55 602.85 8.000 274.000 26 56 248.35 8.000 274.000 27 34 1177.05 20.250 253.750 28 35 1059.99 20.250 253.750 29 36 1934.60 20.250 253.750 30 37 3602.54 20.250 253.750 31 38 1135.39 20.250 253.750 32 39 380.78 20.250 253.750 33 17 1652.62 17.750 236.000 34 18 1760.31 17.750 236.000 35 19 2646.17 17.750 236.000 Notes:

1) See Figures 6.2.1-23 and 6.2.1-24

2) See Figures 6.2.1-18 through 6.2.1-20 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-22 STEAM GENERATOR LOOP 1 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA Analysis Volume Loss Coefficient Junction Area Elev. Inertia 2 No. From To (Ft. ) (Ft.) Coef. Forward Reverse Remarks 01 9 18 103.275 227.000 0.0294 1.4567 1.1665 02 8 18 138.516 227.000 0.0417 1.4658 1.2916 03 1 18 255.850 227.000 0.0361 1.5365 1.1124 04 7 18 98.600 227.000 0.0584 1.4485 1.2745 05 2 18 135.006 227.000 0.0652 1.5438 1.1219 06 4 18 58.650 228.950 0.1601 1.3883 1.4121 07 3 18 137.513 231.000 0.0479 1.2868 0.8502 08 1 14 540.838 236.000 0.0563 0.3512 0.2941 09 8 18 17.850 239.500 0.2369 1.5126 1.4881 10 7 18 17.850 227.000 0.2419 1.5089 1.4748 11 24 26 78.200 278.000 0.1453 0.1424 0.1327 12 13 17 1022.35 286.000 0.0204 0.4730 0.9629 13 13 15 153.746 282.000 0.0794 1.1050 1.4097 14 2 4 6.851 230.639 0.6553 1.4997 1.5061 15 11 18 58.225 231.096 0.0753 1.4662 1.3439 16 9 11 14.161 227.833 0.3653 1.4492 1.3758 17 8 9 262.650 228.500 0.0922 0.3741 0.5268 18 3 11 4.301 232.667 1.4107 1.5607 1.4853 19 1 8 161.925 228.500 0.0711 0.9698 0.7475 20 1 7 203.150 228.500 0.0692 0.8348 0.6230 21 6 7 30.000 223.500 0.0562 1.3500 1.2897 22 13 26 12.912 282.000 0.4686 0.7736 1.2568 23 11 13 5.313 282.000 1.1878 1.4999 1.3565 24 11 18 7.650 271.000 0.5294 1.5248 1.4765 25 4 18 63.376 276.743 0.1588 1.4265 1.4247 26 2 7 315.988 229.625 0.0452 0.4695 0.3402 27 11 18 46.750 279.167 0.0921 1.4737 1.3741 28 13 18 2534.30 286.000 0.0140 0.5660 0.8635 29 6 16 152.873 217.500 0.1708 0.0597 0.0998 30 16 17 5.024 249.130 2.2410 2.5071 2.4934 31 8 30 57.639 258.125 0.0437 0.1146 0.1722 32 28 17 2.525 253.750 2.1257 1.6995 1.6762 33 8 32 62.934 258.125 0.0274 0.1142 0.1718 34 11 33 3.400 253.750 0.7843 1.4929 1.5080 35 11 28 13.235 270.750 0.8223 1.2996 1.3976 36 25 13 63.750 282.000 0.1244 1.0006 0.5078 37 4 15 4.726 260.781 0.8930 1.5629 1.5479 38 8 31 70.661 258.125 0.0438 0.1175 0.1751 39 3 15 81.983 236.000 0.1296 1.3465 1.2987 40 7 15 356.762 244.875 0.0239 0.7580 0.6776 41 7 14 347.302 244.875 0.0317 0.7922 0.7011 42 8 14 375.063 244.875 0.0336 0.7612 0.6926 43 13 14 171.340 282.000 0.0501 1.1753 1.4287 44 13 14 201.348 310.875 0.1829 1.2545 1.4727 45 2 15 332.257 236.000 0.1095 0.2551 0.2520 46 8 10 133.238 244.875 0.0241 0.1027 0.1603 47 8 5 142.375 244.875 0.0430 0.1141 0.1717 48 8 12 103.352 244.875 0.0406 0.1139 0.1715 49 34 3 28.900 240.250 0.2765 1.2806 1.1582 50 19 13 1577.12 319.385 0.0088 0.9380 0.5448 51 19 13 246.500 334.120 0.0455 1.4584 0.8599 52 20 19 49.742 298.670 0.2340 0.7127 0.7109 Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Analysis Volume Loss Coefficient Junction Area Elev. Inertia 2 No. From To (Ft. ) (Ft.) Coef. Forward Reverse Remarks 53 9 33 58.327 236.000 0.1237 0.5288 0.8415 54 9 34 28.492 236.000 0.1568 0.9041 1.1647 55 9 35 58.752 236.000 0.0953 0.6865 0.9303 56 9 12 99.450 236.000 0.0710 0.6219 0.7380 57 9 5 18.012 236.000 0.1648 0.9816 1.2477 58 9 10 7.438 236.000 0.6975 0.6538 1.1593 59 33 35 143.225 244.875 0.0817 0.1385 0.1385 60 35 5 63.682 244.875 0.0810 0.7327 0.7327 61 33 34 143.225 244.875 0.0516 0.1368 0.1368 62 35 12 128.622 244.875 0.0348 0.6165 0.6165 63 5 10 62.722 244.875 0.0608 0.3223 0.4100 64 33 27 75.310 253.750 0.0801 0.0102 0.0102 65 34 28 56.610 253.750 0.2853 0.1868 0.1328 66 34 12 148.249 244.875 0.0486 0.3432 0.2835 67 12 30 159.205 253.750 0.1014 0.1258 0.0634 68 12 10 140.820 244.875 0.0337 0.2848 0.2848 69 10 32 10.540 253.750 1.0548 0.3777 0.3152 70 5 31 53.414 253.750 0.3024 0.0260 0.0260 71 35 29 105.213 253.750 0.1535 0.0136 0.0136 72 27 29 81.796 263.875 0.1193 0.1905 0.2231 73 29 31 57.690 263.875 0.0957 0.7414 0.6870 74 31 32 57.690 263.875 0.0924 0.1356 0.1369 75 29 30 131.470 263.875 0.0401 0.4837 0.5564 76 27 28 71.596 263.875 0.1370 0.1551 0.2033 77 30 32 124.848 263.875 0.0401 0.2597 0.2673 78 28 30 105.383 263.875 0.0943 0.3581 0.3048 79 32 26 22.593 274.000 0.5079 0.0173 0.0173 80 31 25 53.414 274.000 0.2127 0.1035 0.1646 81 30 24 114.164 274.000 0.0829 0.1987 0.1560 82 299 23 92.157 274.000 0.1221 0.1361 0.0737 83 28 22 37.400 274.000 0.2429 0.2326 0.2056 84 27 21 53.338 274.000 0.1780 0.2083 0.1677 85 21 23 27.200 278.000 0.3266 0.2718 0.2644 86 23 25 47.600 278.000 0.2421 0.1014 0.1614 87 21 22 27.200 278.000 0.3390 0.1696 0.2114 88 23 24 73.100 278.000 0.1014 0.2517 0.2488 89 25 26 47.600 278.000 0.1161 0.0112 0.0112 90 21 20 30.107 282.000 0.1277 0.3171 0.3966 91 22 20 23.817 282.000 0.1548 0.3179 0.3974 92 22 24 47.600 278.000 0.1577 0.2514 0.2586 93 23 20 26.699 282.000 0.1008 0.3150 0.3945 94 24 20 34.153 282.000 0.0937 0.3158 0.3953 95 23 13 18.785 282.000 0.2615 1.4207 1.1930 96 24 13 11.747 282.000 0.3776 1.4668 1.3446 97 0 34 0.22727 253.750 0.0 0.0 0.0 Fill junctions used for 98 0 12 0.22727 253.750 0.0 0.0 0.0 analyzing the double-ended 99 0 28 0.22727 253.750 0.0 0.0 0.0 hot leg guillotine break 100 0 30 0.22727 253.750 0.0 0.0 0.0 97 0 12 0.50000 244.500 0.0 0.0 0.0 Fill junctions used for 98 0 35 0.50000 244.500 0.0 0.0 0.0 analyzing the double-ended pump guillotine break 97 0 10 0.22727 253.750 0.0 0.0 0.0 Fill junctions used for 98 0 12 0.22727 253.750 0.0 0.0 0.0 analyzing the double-ended 99 0 30 0.22727 253.750 0.0 0.0 0.0 cold leg guillotine break 100 0 32 0.22727 253.750 0.0 0.0 0.0 Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-23 STEAM GENERATOR - LOOP 3 SUBCOMPARTMENT ANALYSIS VOLUME DATA (1) (2) Analysis GA Drawing Net Free Volume Vol. Height Volume Floor Elev. 3 Vol. No. Vol. No. (Ft. ) (Ft.) (Ft.) 01 16 5421.03 61.000 221.000 02 10 22158.22 36.000 221.000 03 12 18857.39 35.000 221.000 04 13 6034.32 15.000 221.000 05 71 10412.70 23.500 221.000 06 15 4041.25 61.000 221.000 07 70 1696555.5 159.000 282.000 08 6 (a to f), 7 (a to f) 40770.75 36.870 249.130 09 1,3 (a to f), 4 (a to f), 5 (a to f) 2703.75 37.630 211.500 10 9,17,18,19,20,21,22,34,35,36, 29712.56 61.000 221.000 37,38,39,51,52,53,54,55,56 11 11,23,24,25,26,40,41,42,43,57, 60006.17 91.750 221.000 58,59,60, 8(a to d) 12 14,33,50 365478.26 65.000 221.000 13 69b 6456.48 35.450 298.670 14 69a 2748.20 16.670 282.000 15 61 585.54 8.000 274.000 16 62 474.95 8.000 274.000 17 63 1005.05 8.000 274.000 18 64 1519.79 8.000 274.000 19 65 706.09 8.000 274.000 20 66 202.38 8.000 274.000 21 44 1802.53 20.250 253.750 22 45 1132.92 20.250 253.750 23 46 2052.32 20.250 253.750 24 47 2240.17 20.250 253.750 25 48 1434.65 20.250 253.750 26 49 180.16 20.250 253.750 27 27 1948.98 17.750 236.000 28 28 1647.16 17.750 236.000 29 29 2471.24 17.750 236.000 30 30 3514.28 17.750 236.000 31 31 1533.70 17.750 236.000 32 32 381.49 17.750 236.000 33 2 4250.48 12.000 211.500 Notes:

1. See Figures 6.2.1-25 and 6.2.1-26

2. See Figures 6.2.1-18 through 6.2.1-20 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-24 STEAM GENERATOR LOOP 3 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA Analysis Volume Inertia Loss Coefficient Junction Coef. 2 -1 No. From To Area (Ft. ) Elev. (Ft.) (Ft. ) Forward Reverse Remarks 01 10 12 103.275 227.000 0.0623 1.3990 1.1088 02 2 12 138.516 227.000 0.0373 1.4423 1.2668 03 11 12 255.850 227.000 0.0337 1.2893 0.8652 04 3 12 98.600 227.000 0.0523 1.4534 1.2794 05 4 12 135.006 227.000 0.0607 1.3190 0.8971 06 6 12 58.650 228.950 0.1575 1.3886 1.3899 07 5 12 137.513 231.000 0.0479 1.2868 0.8501 08 32 5 24.183 240.250 0.4143 1.3698 1.1279 09 2 12 17.850 239.500 0.2033 1.5122 1.4877 10 3 12 17.850 239.500 0.2083 1.5089 1.4748 11 18 20 68.000 278.000 0.1048 0.1080 0.1063 12 19 7 56.687 282.000 0.0619 1.1074 0.5531 13 7 10 107.194 282.000 0.1072 1.0768 1.3715 14 4 6 6.851 230.639 0.5677 1.4990 1.5054 15 1 12 58.225 230.096 0.0650 1.4662 1.3439 16 1 10 14.161 227.833 0.1967 1.4909 1.4738 17 2 10 641.614 237.375 0.0260 0.1757 0.2138 18 1 5 4.301 232.667 1.2014 1.5620 1.5607 19 2 11 606.050 239.000 0.0292 0.3786 0.3112 20 3 11 550.452 237.375 0.0273 0.4224 0.3280 21 31 5 28.900 240.250 0.3568 1.3897 1.2461 22 20 7 8.364 282.000 0.2156 1.2964 0.8319 23 1 7 5.313 282.000 1.0749 1.4999 1.3941 24 1 12 7.650 271.000 0.4510 1.5248 1.5079 25 6 12 46.750 278.667 0.1612 1.4108 1.4119 26 3 4 268.175 228.500 0.0452 0.0054 0.0054 27 1 12 46.750 279.167 0.0793 1.4737 1.3741 28 7 12 2534.30 286.000 0.0140 0.4270 0.6771 29 33 9 152.873 217.500 0.1708 0.0597 0.0998 30 9 8 5.024 249.130 2.2410 2.5071 2.4934 31 18 7 4.675 282.000 0.0284 1.5044 1.4662 32 7 8 1022.353 286.000 0.0204 0.4730 0.9629 33 6 12 16.626 271.333 0.1942 1.4721 1.4725 34 1 10 3.400 254.000 0.7666 1.5278 1.5237 35 1 10 13.235 270.750 0.2093 1.4900 1.4739 36 7 10 58.520 282.000 0.3055 0.9404 1.4344 37 6 19 1.326 278.000 3.1636 1.5212 1.5549 38 6 25 2.125 263.875 1.9624 1.5241 1.5354 39 6 31 1.275 244.875 3.2235 1.5612 1.5701 40 3 27 164.586 244.875 0.0431 0.2205 0.2358 41 16 18 64.600 278.000 0.1432 0.1560 0.1548 42 17 14 30.158 282.000 0.0967 0.3650 0.5293 43 7 11 86.267 282.000 0.0899 1.3566 1.4582 44 7 11 69.850 310.875 0.2369 1.0017 1.4116 45 30 5 28.900 240.250 0.3752 1.4900 1.4238 46 18 14 33.847 282.000 0.0852 0.3652 0.5295 47 17 7 24.650 282.000 0.0417 1.3985 1.1194 48 3 28 192.177 244.875 0.0418 0.2197 0.2350 49 5 10 28.900 240.250 0.1965 1.3691 1.3419 50 13 7 187.000 334.120 0.0324 0.9278 1.6630 51 13 7 1376.72 319.385 0.0061 0.1615 0.1043 Amendment 61 Page 1 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-24 STEAM GENERATOR LOOP 3 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA Analysis Volume Inertia Loss Coefficient Junction Coef. 2 -1 No. From To Area (Ft. ) Elev. (Ft.) (Ft. ) Forward Reverse Remarks 52 14 13 49.742 298.670 0.4455 0.3093 0.3827 53 4 27 66.462 236.000 0.0916 0.2537 0.2537 54 4 28 39.466 236.000 0.1064 0.2516 0.2516 55 4 29 75.013 236.000 0.0794 0.2527 0.2527 56 4 30 107.950 236.000 0.0611 0.2523 0.2523 57 4 31 22.967 236.000 0.1227 0.2511 0.2511 58 4 32 8.500 236.000 0.3897 0.2523 0.2523 59 27 29 131.750 244.875 0.0806 0.2147 0.2194 60 29 31 78.770 244.875 0.0813 0.6248 0.5841 61 27 28 128.138 244.875 0.0568 0.1899 0.1424 62 29 30 128.622 244.875 0.0436 0.5892 0.6526 63 31 32 50.184 244.875 0.0673 0.6293 0.6703 64 27 21 99.382 253.750 0.1583 0.1174 0.0577 65 28 22 82.820 253.750 0.2087 0.1722 0.1138 66 28 30 193.239 244.875 0.0668 0.1641 0.1728 67 29 23 116.858 253.750 0.1332 0.1200 0.0586 68 30 32 114.223 244.875 0.0615 0.2950 0.3446 69 32 26 6.392 253.750 0.9005 0.8140 0.8758 70 31 25 59.313 253.750 0.2442 0.2063 0.2041 71 30 24 147.985 253.750 0.0981 0.1923 0.1897 72 21 23 148.096 263.875 0.0835 0.1752 0.1755 73 23 25 91.477 263.875 0.0937 0.4251 0.4790 74 25 26 39.840 263.875 0.0653 0.8113 0.8768 75 23 24 131.470 263.875 0.0462 0.5342 0.6655 76 21 22 77.495 263.875 0.0613 0.6947 0.6689 77 24 26 100.938 263.875 0.0503 0.2582 0.2617 78 22 24 142.044 263.875 0.0565 0.2595 0.2613 79 26 20 16.218 274.000 0.6973 0.7189 0.7759 80 25 19 60.843 274.000 0.1878 0.1033 0.1638 81 24 18 151.428 274.000 0.0781 0.0457 0.1052 82 23 17 95.226 274.000 0.1065 0.2091 0.2114 83 22 16 64.073 274.000 0.1753 0.1265 0.0645 84 21 15 77.427 274.000 0.1305 0.1823 0.1275 85 15 17 27.200 278.000 0.2113 0.7335 0.7310 86 17 19 47.600 278.000 0.2373 0.2385 0.2518 87 15 16 25.500 278.000 0.1552 0.8092 0.7842 88 17 18 73.100 278.000 0.1170 0.8437 1.1364 89 19 20 40.800 278.000 0.1428 0.4042 0.3513 90 15 14 37.094 282.000 0.1078 0.3670 0.5310 91 16 14 15.164 282.000 0.1170 0.3637 0.5280 92 3 33 30.000 223.500 0.0562 1.2897 1.3500 93 8 22 2.525 253.750 1.7853 1.6371 1.6455 94 8 24 2.281 253.750 2.6804 1.4780 1.7824 95 3 21 22.474 254.875 0.0470 0.9914 0.4979 96 3 22 25.339 254.875 0.2183 0.2183 0.2337 97 0 22 0.22727 253.750 0.0 0.0 0.0 Double-ended Hot 98 0 24 0.22727 253.750 0.0 0.0 0.0 Leg Guillotine 99 0 28 0.22727 253.750 0.0 0.0 0.0 Break Fill 100 0 30 0.22727 253.750 0.0 0.0 0.0 Junctions 97 0 29 0.50000 244.500 0.0 0.0 0.0 Double-Ended 98 0 29 0.50000 244.500 0.0 0.0 0.0 Pump Suction Leg Guillotine Break Amendment 61 Page 2 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-24 STEAM GENERATOR LOOP 3 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA Analysis Volume Inertia Loss Coefficient Junction Coef. 2 -1 No. From To Area (Ft. ) Elev. (Ft.) (Ft. ) Forward Reverse Remarks Fill Junctions 97 0 24 0.22727 253.750 0.0 0.0 0.0 Double-Ended 98 0 26 0.22727 253.750 0.0 0.0 0.0 Cold Leg Guillotine 99 0 30 0.22727 253.750 0.0 0.0 0.0 Break Fill 100 0 32 0.22727 253.750 0.0 0.0 0.0 Junctions Amendment 61 Page 3 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-25 STEAM GENERATOR AND PRESSURIZER - LOOP 2 SUBCOMPARTMENT ANALYSIS VOLUME DATA (1) (2) Analysis GA Drawing Net Free Volume Vol. Height Volume Floor Elev. 3 Vol. No. Vol. No. (Ft. ) (Ft.) (Ft.) 01 70 1674527.50 159.00 282.00 02 9 4386.33 15.00 221.00 03 10 22158.22 36.00 221.00 04 11 8093.87 15.00 221.00 05 12 18857.39 35.00 221.00 06 13 6034.32 15.00 221.00 07 14 81093.87 19.62 216.38 08 15 4041.25 61.00 221.00 09 16 5421.03 61.00 221.00 10 33 146321.08 25.00 236.00 11 50 138063.31 25.00 261.00 12 71 10412.70 23.50 221.00 13 69 9204.68 52.12 282.00 14 67 10903.53 52.12 282.00 15 2 4250.48 12.00 211.50 16 1,3 (a to f), 2703.75 37.63 211.50 4 (a to f), 5 (a to f) 17 6 (a to f), 7 (a to f) 40770.82 36.87 249.13 18 68b 8232.45 35.45 298.67 19 68a 2892.02 16.67 282.00 20 8a 5143.46 25.00 261.00 21 8b 2492.45 10.00 286.00 22 8c 2042.40 8.25 296.00 23 8d 2229.41 8.50 304.25 24 57 2172.09 8.00 274.00 25 58 952.56 8.00 274.00 26 59 1617.60 8.00 274.00 27 60 2543.61 8.00 274.00 28 40 5457.70 20.25 253.75 29 41 1962.58 20.25 253.75 30 42 2647.43 20.25 253.75 31 43 6001.73 20.25 253.75 32 23 5211.83 17.75 236.00 33 24 2478.92 17.75 236.00 34 25 4054.02 17.75 236.00 35 26 4904.51 17.75 236.00 36 17, 18, 19, 20, 21, 25326.23 46.00 236.00 22, 34, 35, 36, 37, 38, 39, 51, 52, 53, 54, 55, 56 37 27, 28, 29, 30, 31, 24833.40 46.00 236.00 32, 44, 45, 46, 47, 48, 49, 61, 62, 63, 64, 65, 66 NOTES:

1. See Figure 6.2.1-27

2. See Figures 6.2.1-18 through 6.2.1-20 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-26 STEAM GENERATOR AND PRESSURIZER LOOP 2 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA Analysis Volume Inertia Loss Coefficient Junction Coef. 2 -1 No. From To Area (Ft. ) Elev. (Ft.) (Ft. ) Forward Reverse Remarks 01 2 7 136.85 227.00 0.035 1.032 0.932 02 3 7 138.55 227.00 0.038 0.944 0.858 03 4 7 255.85 227.00 0.024 0.962 0.797 04 5 7 98.60 227.00 0.027 0.974 0.974 05 6 7 135.01 227.00 0.044 0.875 0.791 06 7 8 46.75 232.50 0.045 0.918 0.962 07 7 12 110.53 228.50 0.174 0.035 0.964 08 7 10 3167.87 236.00 0.004 0.191 0.191 09 3 10 17.85 239.50 0.025 1.461 1.455 10 5 10 17.85 239.50 0.017 1.452 1.443 11 10 12 26.99 238.50 0.230 0.771 1.449 12 1 17 1022.35 286.00 0.0204 0.4730 0.9629 13 1 13 1376.72 319.39 0.192 0.972 1.974 14 6 8 2.55 229.75 0.243 1.482 1.508 15 7 9 4.75 230.83 0.042 0.975 1.064 16 2 9 14.16 227.83 0.091 1.304 1.332 17 2 3 262.65 228.50 0.073 1.068 0.769 18 9 12 4.30 232.67 0.020 1.489 1.477 19 3 4 162.35 228.50 0.097 0.753 0.881 20 4 5 203.15 228.50 0.101 0.863 0.784 21 5 15 51.00 223.50 0.073 1.073 1.175 22 10 11 3701.34 261.00 0.004 0.192 0.192 23 1 9 5.31 282.00 0.066 1.372 1.491 24 1 14 1577.11 319.39 0.141 0.965 1.865 25 8 11 46.75 278.67 0.030 0.751 0.694 26 5 6 268.18 228.50 0.088 0.793 1.123 27 9 11 7.65 271.00 0.051 1.448 1.424 28 1 11 2534.30 286.00 0.003 0.503 1.131 29 15 16 152.87 217.50 0.1708 0.0597 0.0998 30 16 17 5.024 249.13 2.2410 2.5071 2.4934 31 8 37 1.33 278.17 0.456 1.505 1.478 32 1 23 201.35 310.88 0.012 0.552 1.115 33 8 37 3.40 254.00 0.096 1.508 1.493 34 13 37 58.52 282.00 0.286 0.087 1.410 35 6 37 320.36 236.00 0.072 0.151 0.190 36 12 37 81.98 240.25 0.042 1.085 1.112 37 5 37 404.57 246.00 0.046 0.491 0.345 38 5 35 347.30 244.88 0.046 0.103 0.076 39 12 36 28.90 240.25 0.200 0.967 1.108 40 2 36 270.47 236.00 0.071 0.147 0.188 41 3 36 570.20 249.25 0.039 1.494 0.821 42 14 36 58.52 282.00 0.367 0.803 0.873 43 9 36 3.40 254.00 0.076 1.500 1.491 44 1 18 223.98 334.12 0.098 0.686 1.735 45 1 18 1496.96 319.39 0.098 1.337 1.506 Amendment 61 Page 1 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-26 STEAM GENERATOR AND PRESSURIZER LOOP 2 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA Analysis Volume Inertia Loss Coefficient Junction Coef. 2 -1 No. From To Area (Ft. ) Elev. (Ft.) (Ft. ) Forward Reverse Remarks 46 22 23 192.30 304.25 0.031 0.093 0.130 47 21 22 202.02 296.00 0.028 0.010 0.008 48 20 21 119.46 286.00 0.047 0.468 0.565 49 20 24 170.00 278.00 0.060 0.507 0.010 50 20 28 265.86 267.50 0.037 0.485 0.023 51 18 19 59.37 298.67 0.379 0.828 0.982 52 19 24 60.88 282.00 0.073 0.679 0.416 53 19 25 29.33 282.00 0.078 0.675 0.412 54 19 26 25.93 282.00 0.072 0.674 0.411 55 24 25 27.20 278.00 0.264 0.518 0.343 56 25 26 73.10 278.00 0.106 0.233 0.246 57 25 27 47.60 278.00 0.338 0.006 0.007 58 26 27 129.20 278.00 0.092 0.002 0.002 59 27 31 250.60 274.00 0.047 0.034 0.015 60 26 30 182.22 274.00 0.065 0.024 0.007 61 25 29 88.29 274.00 0.132 0.046 0.020 62 24 28 238.54 274.00 0.050 0.004 0.009 63 24 26 64.60 278.00 0.114 0.345 0.346 64 28 29 81.80 263.88 0.088 0.539 0.539 65 29 30 131.47 263.88 0.042 0.479 0.524 66 30 31 241.61 263.88 0.037 0.200 0.473 67 29 31 57.69 263.88 0.133 0.534 0.534 68 28 32 265.91 253.75 0.062 0.088 0.024 69 29 33 106.40 253.75 0.168 0.046 0.037 70 30 34 192.29 253.75 0.086 0.009 0.030 71 31 35 246.71 253.75 0.065 0.018 0.010 72 28 30 140.32 263.88 0.060 0.431 0.431 73 3 32 375.06 244.88 0.048 0.017 0.006 74 32 33 143.23 244.88 0.106 0.036 0.036 75 33 34 128.62 244.88 0.047 0.252 0.284 76 34 35 240.77 244.88 0.043 0.089 0.089 77 33 35 48.59 244.88 0.151 0.562 0.562 78 32 34 194.23 244.88 0.401 0.093 0.093 79 4 32 185.04 236.00 0.062 0.097 0.154 80 4 33 58.75 236.00 0.180 0.211 0.214 81 4 34 118.35 236.00 0.093 0.151 0.194 82 4 35 178.70 236.00 0.067 0.082 0.087 83 3 28 69.06 255.38 0.297 0.027 0.020 84 1 36 107.19 282.00 0.109 0.661 1.211 85 1 25 13.45 282.00 0.078 0.984 1.349 86 1 26 13.69 282.00 0.038 1.222 0.925 87 1 27 74.66 282.00 0.013 1.019 1.360 88 1 37 94.38 282.00 0.098 0.780 1.271 89 1 14 246.50 334.12 0.141 0.774 1.372 90 1 13 187.00 334.12 0.192 0.781 1.405 91 9 36 13.23 270.75 0.048 1.424 1.407 Amendment 61 Page 2 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-26 STEAM GENERATOR AND PRESSURIZER LOOP 2 SUBCOMPARTMENT ANALYSIS JUNCTION INPUT DATA Analysis Volume Inertia Loss Coefficient Junction Coef. 2 -1 No. From To Area (Ft. ) Elev. (Ft.) (Ft. ) Forward Reverse Remarks 92 9 11 46.75 279.17 0.051 0.986 0.887 93 8 11 16.63 271.33 0.025 1.238 1.263 94 28 17 2.525 253.75 1.9250 1.6842 1.6929 95 31 17 2.281 253.75 2.1151 1.6729 1.6808 96 0 28 0.22727 253.75 0.0 0.0 0.0 Double-Ended Hot 97 0 30 0.22727 253.75 0.0 0.0 0.0 Leg Guillotine 98 0 32 0.22727 253.75 0.0 0.0 0.0 Break Fill 99 0 34 0.22727 253.75 0.0 0.0 0.0 Junctions 96 0 33 0.50000 244.50 0.0 0. 0.0 Double-Ended 97 0 34 0.50001 244.50 0.0 0.0 0.0 Pump Suction Guillotine Break Fill Junctions 96 0 30 0.22727 253.75 0.0 0.0 0.0 Double-Ended 97 0 31 0.22727 253.75 0.0 0.0 0.0 Cold Leg Guillotine 98 0 34 0.22727 253.75 0.0 0.0 0.0 Break Fill 99 0 35 0.22727 253.75 0.0 0.0 0.0 Junctions 96 0 20 0.90910 261.50 0.0 0.0 0.0 Double-Ended Surge Line Guillotine Break Fill Junction Spray Line 96 0 23 1.0000 306.13 0.0 0.0 0.0 Double-Ended Break Fill Junction Amendment 61 Page 3 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-27

SUMMARY

OF CALCULATED SUBCOMPARTMENT PEAK PRESSURES FOR ORIGINAL DESIGN BASES Subcompartment Analysis Volume No. Containment Peak Pressure Differ. Time of Compartment Blowdown (psid) P = Occurrence (sec.) Loop 1 DEHLG 22.2 0.010 28 17 Loop 1 DESLG 18.3 0.010 35 18 Loop 1 DECLG 14.3 0.005 32 18 Loop 2 DEHLG 11.3 0.029 33 10 Loop 2 DESLG 22.4 0.014 33 10 Loop 2 DECLG 7.6 0.085 31 17 Loop 3 DEHLG 16.0 0.008 22 08 Loop 3 DESLG 18.9 0.013 29 12 Loop 3 DECLG 29.7 0.005 26 08 Pressurizer P SUR GB 7.0 0.018 20 01 Pressurizer P SPR LB 0.9(See Note 1) 0.053 23 01 Pressurizer DEHLG 8.7 0.038 20 01 2 Reactor Cavity 150 in. CLG 29.8 0.019 23 33 2 Reactor Cavity 250 in. HLG 25.6 0.019 22 33 Note 1: Refer to section 6.2.1-2a for the evaluation and results for SGR/PUR. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-28 CASES ANALYZED AND RESULTS Balances Case Blowdown Reflood Post Reflood Mass Energy (1) A Double Ended Pump Suction Max. S.I. 6.2.1-29a 6.2.1-35 6.2.1-40 6.2.1-43 6.2.1-51 (1) B Double Ended Pump Suction Min. S.I. 6.2.1-29b 6.2.1-36 6.2.1-41 6.2.1-44 6.2.1-52 C Double Ended Hot Leg 6.2.1-33 - - 6.2.1-47 6.2.1-55 NOTES

1) Double ended refers to the size and type of break.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-29a DOUBLE ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (Maximum Safeguards) Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 0.00000 0.0 0.0 0.0 0.0 0.00102 92463.8 51485.9 40808.9 22659.2 0.00209 42572.8 23639.1 42166.4 23411.3 0.101 42230.4 23509.3 21593.0 11977.5 0.201 43036.6 24088.9 24339.8 13513.6 0.301 45457.6 25620.3 24771.9 13767.7 0.501 45811.1 26285.6 22806.4 12697.9 0.701 45947.9 26903.5 21102.6 11756.3 0.902 45086.2 26887.8 20243.6 11286.3 1.20 41912.3 25628.0 19759.7 11019.8 2.00 34787.9 22749.2 19342.1 10781.9 2.30 30697.4 20945.5 18738.3 10443.4 2.40 28263.4 19597.2 18335.2 10217.0 2.50 24063.3 16949.4 17767.9 9900.3 2.60 21095.3 15133.5 17472.9 9737.4 2.80 17157.2 12691.7 16980.6 9466.0 3.00 15029.1 11341.2 16485.0 9194.2 3.20 13842.7 10580.6 16106.8 8989.3 3.50 12676.2 9845.0 15550.9 8688.6 4.00 11360.4 9072.2 14676.1 8218.1 4.40 10583.5 8612.8 14578.8 8184.8 4.60 10270.6 8412.5 15940.2 8956.5 5.20 9807.0 8030.8 15032.2 8473.6 5.60 9623.0 7799.4 14607.6 8252.7 6.00 10323.8 8371.1 14478.0 8182.5 6.40 8675.0 7822.4 13902.4 7842.3 6.60 8565.5 7627.3 13748.1 7754.5 7.00 8889.9 7434.6 13352.2 7524.9 7.80 9919.5 7427.4 12618.0 7095.4 8.40 9731.5 7070.2 12176.3 6837.1 9.80 8255.6 6168.3 11248.9 6300.8 11.0 6936.7 5332.5 10207.0 5715.0 13.2 5686.6 4439.0 8872.5 5013.5 14.0 5240.2 4397.4 7716.0 4524.6 14.6 4334.0 4340.9 6611.3 3819.7 15.0 3377.8 3948.0 5996.9 3142.1 15.4 2633.2 3253.1 5328.1 2578.0 15.8 2068.6 2580.7 4590.7 2104.4 16.2 1673.1 2099.8 3986.9 1736.5 16.6 1426.1 1797.1 2580.7 1019.1 16.8 1314.9 1659.6 2272.5 836.3 17.4 967.6 1226.2 2389.1 784.6 17.8 719.0 913.8 3202.4 998.9 18.0 521.7 663.3 3219.9 977.8 18.6 0.0 0.0 1313.0 388.5 19.0 0.0 0.0 234.3 69.8 19.6 0.0 0.0 398.6 121.3 20.0 0.0 0.0 253.5 78.2 21.0 0.0 0.0 237.8 75.3 Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 21.4 0.0 0.0 0.0 0.0

M&E exiting the SG side of the break
- M&E exiting the RV side of the break Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-29b DOUBLE ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (Minimum Safeguards) Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 0.0 0 0 0 0.0000E+00 0.001 92904.62 5.14E+07 40021.14 2.2077E+07 0.002 42111.88 2.32E+07 41727.12 2.3017E+07 0.1 41775.36 2.31E+07 21383.11 1.1785E+07 0.2 42434.96 2.36E+07 24027.03 1.3254E+07 0.3 44296.81 2.48E+07 24482.87 1.3519E+07 0.4 44484.33 2.50E+07 23775.81 1.3142E+07 0.5 44756.41 2.54E+07 22713.19 1.2564E+07 0.6 44857.21 2.57E+07 21879.42 1.2108E+07 0.7 45221.76 2.62E+07 21126.29 1.1694E+07 0.8 45220.21 2.64E+07 20607.74 1.1412E+07 0.9 44768.93 2.64E+07 20257.06 1.1221E+07 1.0 43926.74 2.61E+07 19994.91 1.1077E+07 1.1 42990.71 2.57E+07 19834.07 1.0989E+07 1.2 42033.37 2.54E+07 19761.59 1.0950E+07 1.3 41123.67 2.50E+07 19730.07 1.0933E+07 1.4 40262.59 2.47E+07 19676.48 1.0903E+07 1.5 39432.86 2.43E+07 19598.86 1.0859E+07 1.6 38593.39 2.40E+07 19547.18 1.0829E+07 1.7 37749.23 2.36E+07 19534.42 1.0821E+07 1.8 36929.36 2.33E+07 19503.61 1.0804E+07 1.9 36128.61 2.30E+07 19399.08 1.0744E+07 2.0 35168.7 2.26E+07 19243.53 1.0656E+07 2.1 33968.88 2.21E+07 19084.29 1.0567E+07 2.2 32591.36 2.15E+07 18928.85 1.0481E+07 2.3 31171.78 2.09E+07 18680.8 1.0342E+07 2.4 29035.76 1.98E+07 18293.02 1.0125E+07 2.5 24244.16 1.67E+07 17697.32 9.7938E+06 2.6 21059.5 1.48E+07 17400.18 9.6303E+06 2.7 18726.77 1.34E+07 17137.07 9.4852E+06 2.8 16851.18 1.23E+07 16866.01 9.3358E+06 2.9 15537.18 1.14E+07 16590.26 9.1841E+06 3.0 14575.96 1.08E+07 16349.04 9.0521E+06 3.1 13866.62 1.04E+07 16134.28 8.9348E+06 Amendment 63 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 3.2 13315.52 1.01E+07 15929.25 8.8230E+06 3.3 12854.67 9759914 15718.17 8.7076E+06 3.4 12465.45 9516896 15559.53 8.6221E+06 3.5 12141.96 9320690 15444.52 8.5605E+06 3.6 11845.04 9140668 15241.89 8.4495E+06 3.7 11553.47 8961791 15053.96 8.3470E+06 3.8 11295 8808386 14812.26 8.2150E+06 3.9 11083.43 8689816 14672.57 8.1397E+06 4.0 10882 8573856 14537.04 8.0664E+06 4.2 10514.27 8360908 14268.92 7.9215E+06 4.4 10182.85 8172337 14163.25 7.8668E+06 4.6 9853.062 7971689 15593.25 8.6701E+06 4.8 9552.499 7776162 15544.78 8.6430E+06 5.0 9339.392 7619816 15122.46 8.4122E+06 5.2 9219.103 7511356 14993.72 8.3445E+06 5.4 9132.269 7421199 14747.31 8.2116E+06 5.6 9278.075 7509970 14623.42 8.1486E+06 5.8 9775.336 7935800 14631.98 8.1588E+06 6.0 9138.131 7999520 14372.72 8.0151E+06 6.2 8120.285 7558701 14075.95 7.8488E+06 6.4 7916.252 7306096 13879.19 7.7392E+06 6.6 7997.763 7159023 13713.27 7.6467E+06 6.8 8187.296 7065130 13544.99 7.5512E+06 7.0 8437.297 7016812 13339.36 7.4333E+06 7.2 8709.262 7005002 13142.57 7.3207E+06 7.4 8952.062 7000054 12974.56 7.2248E+06 7.6 9130.552 6980516 12801.97 7.1262E+06 7.8 9221.261 6929837 12626.62 7.0260E+06 8.0 9212.234 6839897 12474.49 6.9392E+06 8.2 9101.602 6709368 12341.66 6.8632E+06 8.4 8950.122 6582546 12221.98 6.7944E+06 8.6 8802.578 6482830 12081.18 6.7135E+06 8.8 8628.309 6378026 11944.32 6.6354E+06 9.0 8449.852 6278278 11817.56 6.5634E+06 9.2 8267.771 6177818 11671.53 6.4808E+06 9.4 8083.11 6075873 11537.09 6.4049E+06 9.6 7900.68 5975670 11402.86 6.3292E+06 9.8 7718.878 5876232 11263.89 6.2508E+06 10.0 7522.295 5766084 11089.76 6.1527E+06 10.2 7316.43 5648842 10917.55 6.0563E+06 10.4 7112.608 5526236 10762.57 5.9699E+06 10.6 6917.401 5396420 10597.41 5.8774E+06 Amendment 63 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 10.8 6752.821 5277266 10431.94 5.7847E+06 11.0 6603.161 5156164 10272.33 5.6953E+06 11.2 6468.664 5037476 10108.02 5.6033E+06 11.4 6348.755 4926110 9963.724 5.5228E+06 11.6 6241.504 4822795 9831.32 5.4490E+06 11.8 6140.548 4724484 9673.41 5.3607E+06 12.0 6062.111 4645379 9616.331 5.3297E+06 12.2 5988.198 4571698 9455.626 5.2405E+06 12.4 5914.275 4504352 9393.91 5.2081E+06 12.6 5838.567 4441787 9259.228 5.1353E+06 12.8 5759.927 4384204 9181.57 5.0961E+06 13.0 5677.502 4331408 9068.448 5.0386E+06 13.2 5590.516 4283017 8976.836 4.9954E+06 13.4 5497.694 4237823 8870.274 4.9468E+06 13.6 5402.477 4197675 8775.678 4.9088E+06 13.8 5310.542 4165651 8445.241 4.7463E+06 14.0 5228.808 4149578 8312.26 4.7245E+06 14.2 5153.193 4164698 7860.201 4.5084E+06 14.4 5024.887 4190134 7700.128 4.4166E+06 14.6 4825.918 4203247 7453.152 4.2167E+06 14.8 4583.533 4206397 7275.305 4.0239E+06 15.0 4315.399 4199916 7125.843 3.8346E+06 15.2 3980.644 4130737 6806.493 3.5546E+06 15.4 3526.648 3951140 6344.497 3.2039E+06 15.6 3068.72 3676747 5919.004 2.9144E+06 15.8 2706.987 3327979 5520.761 2.6624E+06 16.0 2387.094 2956656 5139.264 2.4333E+06 16.2 2133.476 2653902 4778.308 2.2229E+06 16.4 1932.445 2411330 4420.575 2.0199E+06 16.6 1773.314 2217876 3735.696 1.6493E+06 16.8 1631.728 2044964 3095.36 1.2621E+06 17.0 1495.529 1877200 2911.343 1.1071E+06 17.2 1357.029 1706152 3071.554 1.1161E+06 17.4 1225.151 1542811 2876.564 1.0169E+06 17.6 1101.284 1388813 2589.467 9.0040E+05 17.8 985.6415 1244339 2745.666 9.3551E+05 18.0 857.2459 1084240 3159.863 1.0419E+06 18.2 738.6794 935289.6 3470.16 1.1038E+06 18.4 631.8585 800794 3524.288 1.0867E+06 18.6 526.1107 667279.4 3136.854 9.4555E+05 18.8 429.6469 545383.8 2700.743 8.0026E+05 19.0 347.7749 441805.6 2350.773 6.8605E+05 Amendment 63 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 19.2 285.0069 362366.9 1951.271 5.6150E+05 19.4 262.1233 333504.1 1516.271 4.3082E+05 19.6 200.8088 255638.3 1032.456 2.9041E+05 19.8 159.5739 203336.6 492.2412 1.3771E+05 20.0 104.9937 133989.8 0 0.0000E+00 20.2 112.6208 143925.2 0 0.0000E+00 20.4 0 0 0 0.0000E+00 Amendment 63 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-33 DOUBLE-ENDED HOT-LEG BREAK - BLOWDOWN M&E RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 0 0 0 0 0 1.00E-03 43688.44 2.85E+07 43687.19 2.85E+07 2.09E-03 47321.3 3.09E+07 46979.03 3.06E+07 0.1012072 39202.57 2.59E+07 27844.44 1.81E+07 0.2016874 36958.98 2.44E+07 24045.26 1.55E+07 0.30157 35700.72 2.35E+07 21537.16 1.37E+07 0.4016626 34450.62 2.27E+07 20272.73 1.27E+07 0.5011847 33750.91 2.22E+07 19540.41 1.21E+07 0.6013104 33629.93 2.21E+07 19078.68 1.16E+07 0.7010645 33277.37 2.19E+07 18678.92 1.12E+07 0.8021625 32629.22 2.16E+07 18444.73 1.10E+07 0.9015223 32179.52 2.13E+07 18166.1 1.07E+07 1.001001 31781.62 2.12E+07 17975.91 1.05E+07 1.101134 31318.08 2.10E+07 17818.96 1.04E+07 1.202022 30772.77 2.07E+07 17700.96 1.03E+07 1.302057 30212.57 2.05E+07 17617.87 1.02E+07 1.401172 29616.68 2.02E+07 17575 1.01E+07 1.501605 28935.77 1.99E+07 17559.58 1.01E+07 1.602025 28197.88 1.95E+07 17565.08 1.00E+07 1.702078 27446.86 1.91E+07 17585.53 1.00E+07 1.801534 26696.02 1.87E+07 17610.96 9986452 1.901582 25927.11 1.83E+07 17636.62 9975202 2.001132 25153.17 1.79E+07 17659.7 9965092 2.102046 24383.73 1.74E+07 17678.47 9954695 2.202007 23638.41 1.70E+07 17691.96 9943932 2.302126 22927.73 1.66E+07 17698.96 9931795 2.40195 22229.81 1.62E+07 17698.1 9917478 2.501767 21563.74 1.58E+07 17684.54 9898077 2.601148 20974.5 1.54E+07 17658.12 9873457 2.701902 20452.37 1.50E+07 17618.02 9842890 2.801979 20001.97 1.47E+07 17563.4 9806016 2.90169 19629.72 1.44E+07 17493.73 9762371 3.001117 19315.48 1.42E+07 17408.03 9711327 Amendment 63 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 3.10147 19059.91 1.40E+07 17303.72 9651353 3.201951 18855.89 1.38E+07 17180.35 9582147 3.301385 18698.64 1.36E+07 17038.02 9503704 3.401954 18581.02 1.35E+07 16865.84 9409873 3.502226 18499.63 1.33E+07 16667.2 9302571 3.601166 18450.26 1.32E+07 16458.88 9191126 3.701886 18434.34 1.32E+07 16236.93 9073354 3.801023 18461.9 1.31E+07 16013.92 8955895 3.901815 18580.6 1.31E+07 15779.64 8833137 4.002116 18736.04 1.31E+07 15541.34 8708825 4.200067 19056.76 1.31E+07 15097.42 8480102 4.402387 19429.93 1.32E+07 14613.58 8229352 4.60138 19832.56 1.33E+07 14029.98 7921374 4.800666 20260.04 1.34E+07 13470.24 7628088 5.000408 20864.93 1.36E+07 12910.34 7335176 5.200418 15538.79 1.11E+07 12385.5 7061230 5.400113 15824.68 1.12E+07 11893.18 6804416 5.600178 16024.41 1.13E+07 11438.75 6567634 5.800522 16246.63 1.13E+07 11046.44 6364623 6.000829 16450.43 1.13E+07 10679.84 6173444 6.200135 16708.9 1.13E+07 10349.23 6000062 6.400867 16943.47 1.14E+07 10054.79 5845420 6.600744 17144.7 1.14E+07 9788.913 5705173 6.802117 17064.19 1.13E+07 9543.329 5575148 7.001101 17384.61 1.14E+07 9318.732 5455814 7.200984 17658.84 1.15E+07 9104.526 5341619 7.401012 17930.24 1.15E+07 8902.196 5233696 7.600245 18267.85 1.17E+07 8700.18 5125267 7.800224 18570.39 1.18E+07 8491.396 5013054 8.001768 18344.9 1.16E+07 8275.066 4897368 8.20004 17645.11 1.12E+07 8057.417 4781948 8.400576 15782.59 1.01E+07 7832.872 4664023 8.602201 15507.1 9954610 7609.89 4548586 8.801587 15547.18 9942111 7390.665 4436594 9.00193 15572.89 9929198 7182.175 4331928 9.200268 15575.31 9905857 6981.686 4231672 9.401546 15512.82 9847560 6783.328 4132893 9.601906 15227.57 9671024 6588.509 4035974 9.805025 14470.51 9245470 6391.597 3938246 10.00153 13748.38 8842994 6210.81 3849943 Amendment 63 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 10.20095 13396.29 8635079 6027.533 3761202 10.40022 13177.88 8501138 5850.784 3677035 10.60278 12948.9 8366344 5677.693 3595673 10.80236 12660.65 8203820 5512.13 3518203 11.00169 12295.35 8003456 5351.882 3443363 11.20192 11881.82 7780744 5196.395 3371134 11.40024 11476.42 7565134 5047.114 3302028 11.60276 11097.75 7367152 4899.979 3234558 11.80278 10747.16 7188022 4758.789 3170232 12.00241 10373.63 7034326 4623.638 3109343 12.20265 9595.021 6794880 4490.47 3049399 12.40053 9121.757 6638823 4362.469 2991924 12.60178 8694.217 6468444 4235.013 2934589 12.80217 8179.366 6230320 4108.249 2877552 13.00118 7599.651 5956289 3980.227 2820365 13.20063 7069.965 5726657 3841.534 2758715 13.4003 6546.679 5523690 3686.082 2691981 13.60148 5947.967 5271796 3508.801 2621254 13.80067 5381.146 4872683 3305.53 2546066 14.00167 4862.264 4511506 3077.764 2468870 14.20092 4425.477 4236427 2831.621 2389324 14.40087 4069.554 4006095 2573.996 2307772 14.60033 3767.698 3797650 2320.814 2230362 14.80085 3523.746 3604592 2086.706 2152756 15.00015 3280.382 3404238 1885.775 2074668 15.20079 3007.536 3200691 1715.463 1986217 15.40059 2712.642 2982759 1576.453 1888573 15.60072 2396.906 2742627 1452.93 1773989 15.80065 2117.364 2493038 1348.352 1660556 16.0002 1886.449 2257668 1248.008 1544872 16.2005 1711.418 2074086 1145.972 1423289 16.40085 1591.614 1950431 1052.229 1310371 16.60076 1501.408 1852779 975.9382 1217948 16.8005 1391.564 1724180 915.0459 1143892 17.00005 1284.553 1600832 867.8073 1086740 17.20078 1187.933 1486122 835.7102 1048068 17.4005 1101.133 1378778 819.3569 1028686 17.60022 1011 1268316 813.6566 1022182 17.80008 934.9639 1177410 804.3359 1010682 18.00011 875.3536 1103760 763.9021 959784.5 Amendment 63 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 18.20043 802.9268 1015315 678.7601 853080.2 18.40074 710.7604 900126.2 611.5388 770560.6 18.60028 624.6891 791752.1 570.4488 719463.1 18.80079 525.5437 666314.9 514.0261 649313.5 19.00025 480.1958 609251.5 434.6844 548613.8 19.20018 420.3883 534551.8 334.4438 423800 19.4003 362.6664 461871.8 217.2682 275748.7 19.60033 297.6432 380188.3 146.0993 186414.9 19.80039 230.924 295942.8 80.2612 103335.4 20.00016 153.5174 197554.2 0 0 20.20012 53.98095 70014.02 0 0 20.40027 0 0 0 0

M&E exiting from the RV side of the break
- M&E exiting from the SG side of the break Amendment 63 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-35 DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS REFLOOD M&E RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 21.4 .0 .0 .0 .0 21.9 .0 .0 .0 .0 22.0 .0 .0 .0 .0 22.2 .0 .0 .0 .0 22.3 .0 .0 .0 .0 22.4 43.9 51.7 .0 .0 22.5 32.2 37.9 .0 .0 22.6 28.1 33.1 .0 .0 22.7 30.9 36.4 .0 .0 22.8 35.7 42.1 .0 .0 22.9 43.1 50.8 .0 .0 23.0 49.8 58.7 .0 .0 23.1 56.7 66.7 .0 .0 23.2 62.8 74.0 .0 .0 23.3 69.3 81.7 .0 .0 23.4 73.3 86.4 .0 .0 23.5 77.3 91.1 .0 .0 23.6 81.2 95.6 .0 .0 23.7 84.8 99.9 .0 .0 23.8 88.4 104.1 .0 .0 23.9 91.9 108.2 .0 .0 24.0 95.2 112.2 .0 .0 24.1 98.5 116.0 .0 .0 24.2 101.6 119.7 .0 .0 24.3 104.7 123.4 .0 .0 24.4 107.7 126.9 .0 .0 24.5 110.6 130.3 .0 .0 25.5 136.7 161.1 .0 .0 26.0 147.5 173.9 .0 .0 26.5 380.3 450.0 3827.1 514.9 27.6 435.3 515.6 4380.7 609.0 28.6 425.8 504.3 4283.7 599.5 29.6 415.2 491.6 4176.5 588.0 30.0 410.9 486.6 4133.4 583.3 30.6 404.7 479.1 4069.3 576.2 31.6 394.5 467.0 3964.7 564.5 32.7 434.5 514.6 4406.8 604.6 33.7 421.1 498.7 4269.0 591.6 34.7 421.6 488.5 4182.8 581.8 35.6 405.2 479.8 4108.1 573.2 35.7 404.4 478.8 4099.9 572.2 36.7 396.7 469.5 4020.0 563.1 37.7 389.2 460.7 3943.2 554.2 38.7 382.1 452.2 3869.1 545.6 39.7 375.3 444.1 3797.6 537.4 40.7 368.8 436.4 3728.7 529.4 41.7 362.6 428.9 3662.1 521.7 42.2 359.5 425.3 3629.6 517.9 42.7 356.6 421.8 3597.7 514.2 43.7 350.8 414.9 3535.4 507.0 Amendment 61 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-35 DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS REFLOOD M&E RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 44.7 345.2 408.3 3475.0 500.0 45.7 339.9 401.9 3416.5 493.2 46.7 334.7 395.7 3359.7 486.6 47.7 329.7 389.8 3304.5 480.1 48.7 324.9 384.1 3250.9 473.9 49.7 320.2 378.5 3198.7 467.8 50.7 315.7 373.1 3147.9 461.8 51.7 311.3 367.9 3098.4 456.0 52.7 307.1 362.9 3050.2 450.4 53.7 303.0 358.0 3003.1 444.8 54.7 205.5 242.4 500.9 141.9 55.7 208.9 246.5 494.1 141.2 56.7 207.5 244.8 496.7 141.4 57.7 206.0 243.1 499.3 141.7 58.6 204.7 241.5 501.7 141.9 58.7 204.5 241.3 502.0 142.0 59.7 203.0 239.5 504.7 142.3 60.7 201.5 237.7 507.4 142.6 61.7 199.9 235.8 510.2 142.9 62.7 198.3 233.9 513.0 143.2 63.7 196.7 232.0 515.8 143.5 64.7 195.1 230.1 518.6 143.8 65.7 193.4 228.1 521.5 144.2 66.7 191.7 226.1 524.5 144.5 67.7 190.0 224.1 527.4 144.9 68.7 188.3 222.0 530.5 145.2 69.7 186.5 219.9 533.5 145.6 70.7 184.6 217.7 536.6 146.0 71.7 182.8 215.6 539.8 146.4 72.7 180.9 213.3 543.0 146.8 73.7 179.0 211.0 546.3 147.2 74.7 177.0 208.7 549.6 147.7 75.7 175.0 206.3 553.0 148.1 76.7 172.9 203.9 556.4 148.6 77.7 170.8 201.4 559.9 149.1 78.7 168.7 198.9 563.5 149.6 79.7 166.5 196.3 567.1 150.1 80.4 164.9 194.4 569.7 150.5 80.7 164.2 193.6 570.8 150.6 81.7 161.9 190.9 574.7 151.2 82.7 159.5 188.0 578.6 151.8 84.7 157.8 186.0 581.8 151.8 86.7 157.2 185.3 583.2 151.6 88.7 156.6 184.6 584.7 151.3 90.7 156.0 183.9 586.1 151.1 92.7 155.5 183.2 587.5 150.9 94.7 154.9 182.6 588.9 150.6 96.7 154.3 181.9 590.4 150.4 98.7 153.7 181.2 591.8 150.1 100.7 153.1 180.5 593.2 149.9 102.7 152.5 179.8 594.6 149.6 104.7 151.9 179.1 596.0 149.4 105.4 151.7 178.9 596.4 149.3 Amendment 61 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-35 DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS REFLOOD M&E RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 106.7 151.3 178.4 597.3 149.1 108.7 150.8 177.7 598.7 148.9 110.7 150.2 177.0 600.1 148.6 112.7 149.6 176.3 601.5 148.4 114.7 148.9 175.6 602.9 148.1 116.7 148.3 174.8 604.3 147.8 118.7 147.7 174.1 605.6 147.6 120.7 147.1 173.4 607.0 147.3 122.7 146.5 172.7 608.4 147.1 124.7 145.9 172.0 609.7 146.8 126.7 145.3 171.2 611.1 146.5 128.7 144.7 170.5 612.4 146.3 130.7 144.1 169.8 613.8 146.0 132.7 143.4 169.1 615.1 145.8 132.8 143.4 169.0 615.2 145.7 134.7 142.8 168.3 616.5 145.5 136.7 142.2 167.6 617.8 145.2 138.7 141.6 166.9 619.2 145.0 140.7 141.0 166.1 620.5 144.7 142.7 140.3 165.4 621.9 144.4 144.7 139.7 164.7 623.2 144.2 146.7 139.1 163.9 624.5 143.9 148.7 138.5 163.2 625.9 143.6 150.7 137.8 162.4 627.2 143.4 152.7 137.2 161.7 628.5 143.1 154.7 136.5 160.9 629.9 142.8 156.7 135.9 160.2 631.2 142.6 158.7 135.3 159.4 632.5 142.3 160.7 134.6 158.6 633.9 142.0 162.7 134.0 157.9 635.2 141.8 163.0 133.9 157.8 635.4 141.7 164.7 133.3 157.1 636.5 141.5 166.7 132.7 156.4 637.8 141.2 168.7 132.0 155.6 639.2 141.0 170.7 131.4 154.8 640.5 140.7 172.7 130.7 154.1 641.8 140.4 174.7 130.1 153.3 643.1 140.2 176.7 129.4 152.5 644.4 139.9 178.7 128.8 151.8 645.7 139.7 180.7 128.1 151.0 647.1 139.4 182.7 127.5 150.2 648.4 139.1 184.7 126.8 149.4 649.7 138.9 186.7 126.2 148.7 651.0 138.6 188.7 125.5 147.9 652.3 138.3 190.7 124.9 147.1 653.6 138.1 192.7 124.2 146.4 654.9 137.8 194.7 123.6 145.6 656.2 137.6 196.7 122.9 144.8 657.5 137.3 197.0 122.8 144.7 657.7 137.3 198.7 122.2 144.0 658.8 137.1 200.7 121.6 143.3 660.1 136.8 202.7 120.9 142.5 661.4 136.5 204.7 120.3 141.7 662.7 136.3 Amendment 61 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-35 DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS REFLOOD M&E RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 206.7 119.6 140.9 664.0 136.0 208.7 118.9 140.1 665.3 135.8 210.7 118.3 139.3 666.6 135.5 212.7 117.6 138.5 667.9 135.3 214.7 116.9 137.8 669.2 135.0 216.7 116.2 137.0 670.5 134.8 218.7 115.6 136.2 671.7 134.5 220.7 114.9 135.4 673.0 134.3 222.7 114.2 134.6 674.3 134.0 223.8 113.9 134.2 675.0 133.9

M&E exiting from the SG side of the break
- M&E exiting from the RV side of the break Amendment 61 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-36 DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 20.4 0 0 0 0.0000E+00 20.9 4.59E-04 0.5399843 9.18E-05 9.14E-03 21.1 4.59E-04 0.5399857 9.18E-05 9.14E-03 21.2 4.59E-04 0.5399864 9.18E-05 9.14E-03 21.3 4.59E-04 0.5400019 9.18E-05 9.11E-03 21.3 4.59E-04 0.5399752 9.18E-05 9.11E-03 21.4 75.31432 88635.78 9.18E-05 1.01E-02 21.5 27.59599 32466.84 9.18E-05 9.46E-03 21.6 27.16997 31964.96 9.18E-05 9.47E-03 21.8 31.65934 37247.23 9.18E-05 9.53E-03 21.9 38.78347 45629.97 9.18E-05 9.63E-03 22.0 45.33939 53345.19 9.18E-05 9.71E-03 22.1 52.0619 61257.02 9.18E-05 9.80E-03 22.2 58.06995 68329.14 9.18E-05 9.88E-03 22.3 63.33302 74524.86 9.18E-05 9.96E-03 22.4 67.37204 79280.49 9.18E-05 1.00E-02 22.5 71.30774 83914.51 9.18E-05 1.01E-02 22.5 73.21647 86162.05 9.18E-05 1.01E-02 22.6 75.08839 88366.34 9.18E-05 1.01E-02 22.7 78.72913 92653.83 9.18E-05 1.02E-02 22.8 82.24329 96792.59 9.18E-05 1.02E-02 22.9 85.64216 100795.9 9.18E-05 1.03E-02 23.0 88.9355 104675.3 9.18E-05 1.03E-02 23.1 92.13176 108440.7 9.18E-05 1.04E-02 23.2 95.23832 112100.6 9.18E-05 1.04E-02 23.3 98.26167 115662.9 9.18E-05 1.05E-02 23.4 101.2075 119134.2 9.18E-05 1.05E-02 23.5 104.081 122520.5 9.18E-05 1.06E-02 24.5 129.6525 152668.5 9.18E-05 1.10E-02 25.5 262.6908 309967.2 2535.821 320096.4 26.6 418.5477 495464.8 4309.344 578783.2 27.6 411.685 487281.5 4234.953 572976.4 Amendment 63 Page 1 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 28.6 402.6697 476522.1 4140.938 563385.9 29.6 393.5205 465606.6 4045.038 553277.3 30.6 384.5591 454919.1 3950.402 543140.4 31.1 380.19 449710 3903.98 538129.7 31.6 375.9062 444603.5 3858.265 533177.2 32.6 386.9898 457906 4021.158 550007.8 33.6 387.0697 457911.6 3996.533 545528.6 34.6 379.681 449101.8 3918.124 536891.7 35.6 372.6035 440665.8 3842.425 528533 36.6 365.8243 432587.7 3769.359 520448.5 36.8 364.502 431012.3 3755.042 518862.6 37.6 359.3232 424843.2 3698.761 512623.9 38.6 353.0885 417418.2 3630.567 505053.9 39.6 347.0946 410281.9 3564.54 497715.5 40.6 341.3322 403423 3500.654 490603.8 41.6 335.7876 396825.1 3438.805 483708 42.6 330.4485 390473.2 3378.876 477017.3 43.6 325.3019 384351.9 3320.761 470520.2 44.6 320.3359 378446.6 3264.358 464205.9 45.6 315.5394 372744.1 3209.573 458064.2 46.6 310.9022 367232.2 3156.319 452085.6 47.6 306.4147 361899.3 3104.513 446261.2 48.6 298.4865 352442 277.6143 145466.2 49.6 302.0064 356661.7 278.5453 147242.4 50.6 297.5015 351310.5 276.9011 144947 51.1 295.2643 348653.3 276.0872 143811.2 51.6 293.0448 346017.5 275.2814 142687.2 52.6 288.6543 340804.1 273.6916 140472.1 53.6 284.3152 335652.9 272.1263 138294.2 54.6 279.9856 330513.8 270.5705 136133.1 55.6 275.608 325318.7 269.0065 133960.4 56.6 271.6819 320660.3 267.6054 132018 57.6 267.8364 316098.4 266.2379 130124.8 58.6 264.0636 311623.6 264.9013 128276.4 59.6 260.3628 307234.8 263.5952 126472.3 60.6 256.7295 302926.8 262.3181 124710.4 61.6 253.1589 298693.8 261.0683 122988 62.6 249.6579 294544 259.8477 121307.7 63.6 246.2263 290477 258.6561 119669 64.6 242.8639 286492.8 257.4932 118071.4 Amendment 63 Page 2 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 65.6 239.5706 282590.9 256.3587 116514.5 66.6 236.3462 278771.2 255.2524 114997.7 67.6 233.1906 275033.5 254.174 113520.7 68.6 230.1036 271377.6 253.1233 112083 69.6 227.0851 267803.3 252.1001 110684.1 70.6 224.1349 264310.4 251.104 109323.6 71.6 221.2529 260898.6 250.1348 108001.1 72.6 218.439 257567.8 249.1923 106716 73.6 215.6928 254317.6 248.2761 105467.9 74.6 213.0142 251147.7 247.3859 104256.2 75.6 210.4028 248057.7 246.5215 103080.5 76.6 207.8584 245047.4 245.6825 101940.4 77.6 205.3806 242116.1 244.8686 100835.1 78.6 202.9683 239262.6 244.0792 99763.96 79.6 200.6217 236487.1 243.3142 98726.62 80.6 198.3374 233785.6 242.5725 97721.59 81.6 196.1118 231153.7 241.8527 96746.95 83.6 191.8509 226115.6 240.4821 94892.82 85.6 187.8394 221373.4 239.2008 93161.6 87.6 184.0717 216919.9 238.0053 91548.23 89.6 180.5409 212747.2 236.8922 90047.57 90.1 179.6944 211746.9 236.6263 89689.4 91.6 177.2401 208846.7 235.8576 88654.36 93.6 174.1615 205209.4 234.8981 87363.36 95.6 171.2972 201825.5 234.0099 86169.32 97.6 168.6387 198685.2 233.1892 85067.05 99.6 166.1773 195778.1 232.4325 84051.42 101.6 163.8925 193079.7 231.7332 83113.48 103.6 161.7803 190585.3 231.0889 82249.89 105.6 159.8382 188292.1 230.498 81458.34 107.6 158.0575 186189.6 229.9572 80734.25 109.6 156.4297 184267.7 229.4633 80073.37 111.6 154.9461 182516.1 229.0133 79471.47 113.6 153.5982 180925 228.6043 78924.57 115.6 152.378 179484.5 228.2334 78428.88 115.7 152.3202 179416.3 228.2159 78405.38 117.6 151.2775 178185.5 227.8981 77980.81 119.6 150.289 177018.7 227.5957 77576.96 121.6 149.4052 175975.5 227.324 77214.11 123.6 148.6191 175047.7 227.0807 76889.26 Amendment 63 Page 3 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 125.6 147.9239 174227.1 226.8637 76599.56 127.6 147.3132 173506.4 226.671 76342.37 129.6 146.781 172878.3 226.5007 76115.2 131.6 146.3215 172336 226.3512 75915.73 133.6 145.9294 171873.2 226.2208 75741.8 135.6 145.5996 171484 226.1081 75591.4 137.6 145.3272 171162.5 226.0116 75462.66 139.6 145.1078 170903.6 225.93 75353.84 141.6 144.926 170689 225.8595 75259.74 143.6 144.7825 170519.7 225.7999 75180.24 144.2 144.7477 170478.7 225.7843 75159.49 145.6 144.6806 170399.4 225.752 75116.33 147.6 144.6169 170324.2 225.7148 75066.73 149.6 144.5882 170290.4 225.6875 75030.3 151.6 144.5919 170294.7 225.6693 75005.94 153.6 144.625 170333.8 225.6594 74992.66 155.6 144.6853 170404.9 225.6571 74989.52 157.6 144.7703 170505.3 225.6618 74995.73 159.6 144.8781 170632.5 225.6729 75010.49 161.6 145.0066 170784.1 225.6899 75033.08 163.6 145.1539 170958 225.7122 75062.83 165.6 145.3185 171152.3 225.7395 75099.12 167.6 145.4988 171365 225.7713 75141.38 169.6 145.6933 171594.5 225.8071 75189.11 171.6 146.0324 171994.5 225.8853 75291.21 173.6 146.7846 172882.4 226.4977 75626.09 174.5 147.0687 173217.7 226.9471 75791.86 175.6 147.4194 173631.6 227.6468 76023.54 177.6 148.0708 174400.4 229.3034 76520.81 179.6 148.7208 175167.6 231.3793 77095.97 181.6 149.3299 175886.4 233.7845 77720.52 183.6 149.8741 176528.8 236.4426 78373 185.6 150.337 177075.2 239.2915 79037.26 187.6 150.7023 177506.4 242.2841 79700.27 189.6 150.959 177809.4 245.3873 80353.42 191.6 151.1003 177976.2 248.5791 80991.48 193.6 151.1221 178002 251.8451 81611.7 195.6 151.0225 177884.4 255.1766 82212.98 197.6 150.8007 177622.6 258.568 82795.29 199.6 150.4565 177216.3 262.0156 83359.12 Amendment 63 Page 4 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 201.6 149.9901 176665.9 265.5164 83905.3 203.6 149.4023 175972 269.0676 84434.74 205.6 148.6937 175135.7 272.6658 84948.34 206.1 148.4978 174904.5 273.5723 85074.38

M&E exiting the SG side of the break
- M&E exiting the RV side of the break Amendment 63 Page 5 of 5

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-40 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 228.8 125.2 158.8 736.2 142.8 233.8 124.9 158.4 736.6 142.8 238.8 126.2 160.0 735.2 142.3 243.8 125.9 159.6 735.6 142.3 248.8 125.5 159.2 735.9 142.2 253.8 125.2 158.7 736.3 142.2 258.8 124.9 158.3 736.6 142.2 263.8 124.5 157.9 737.0 142.1 268.8 124.2 157.4 737.3 142.1 273.8 125.5 159.1 736.0 141.6 278.8 125.1 158.6 736.3 141.6 283.8 124.8 158.2 736.7 141.6 288.8 124.4 157.8 737.0 141.5 293.8 124.1 157.3 737.4 141.5 298.8 123.7 156.9 737.7 141.5 303.8 125.0 158.5 736.5 141.0 308.8 124.6 158.0 736.8 141.0 313.8 124.3 157.6 737.2 140.9 318.8 123.9 157.1 737.5 140.9 323.8 123.6 156.7 737.9 140.9 328.8 124.8 158.3 736.6 140.4 333.8 124.5 157.8 737.0 140.4 338.8 124.1 157.4 737.4 140.4 343.8 123.7 156.9 737.7 140.3 348.8 123.4 156.4 738.1 140.3 353.8 123.0 156.0 738.4 140.3 358.8 124.2 157.5 737.2 139.8 363.8 123.9 157.1 737.6 139.8 368.8 123.5 156.6 738.0 139.8 373.8 123.1 156.1 738.3 139.7 378.8 122.8 155.7 738.7 139.7 383.8 122.4 155.2 739.1 139.7 388.8 123.6 156.7 737.9 139.2 393.8 123.2 156.2 738.2 139.2 398.8 122.8 155.8 738.6 139.2 403.8 122.6 155.4 738.9 139.1 408.8 122.3 155.1 739.2 139.1 413.8 122.1 154.8 739.4 139.0 418.8 123.4 156.4 738.1 138.5 423.8 123.1 156.1 738.4 138.5 428.8 122.8 155.7 738.6 138.4 433.8 122.6 155.4 738.9 138.4 438.8 122.3 155.1 739.2 138.3 443.8 122.0 154.7 739.4 138.3 448.8 121.8 154.4 739.7 138.2 453.8 123.1 156.0 738.4 137.8 458.8 122.8 155.7 738.7 137.7 463.8 122.5 155.3 738.9 137.7 468.8 122.2 155.0 739.2 137.6 473.8 122.0 154.6 739.5 137.6 Amendment 61 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-40 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 478.8 121.7 154.3 739.8 137.5 483.8 121.4 154.0 740.0 137.4 488.8 122.7 155.5 738.8 137.0 493.8 122.4 155.2 739.1 136.9 498.8 122.1 154.8 739.4 136.9 503.8 121.8 154.4 739.6 136.8 508.8 121.5 154.1 739.9 136.8 513.8 121.2 153.7 740.2 136.7 518.8 122.4 155.3 739.0 136.3 523.8 122.2 154.9 739.3 136.2 528.8 121.9 154.5 739.6 136.2 533.8 121.6 154.1 739.9 136.1 538.8 121.3 153.8 740.2 136.1 543.8 121.0 153.4 740.5 136.0 548.8 122.1 154.9 739.3 135.6 553.8 121.8 154.5 739.6 135.6 558.8 121.5 154.1 739.9 135.5 563.8 121.2 153.7 740.2 135.5 568.8 120.9 153.3 740.5 135.4 573.8 120.6 153.0 740.8 135.4 578.8 121.8 154.4 739.7 134.9 583.8 121.4 154.0 740.0 134.9 588.8 121.1 153.6 740.3 134.8 593.8 120.8 153.2 740.7 134.8 598.8 120.5 152.8 741.0 134.7 603.8 121.6 154.2 739.8 134.3 608.8 121.3 153.8 740.1 134.3 613.8 121.0 153.5 740.4 134.2 618.8 120.8 153.1 740.7 134.2 623.8 120.5 152.8 741.0 134.1 628.8 120.2 152.4 741.3 134.0 633.8 121.3 153.8 740.2 133.6 638.8 121.0 153.4 740.5 133.6 643.8 120.7 153.0 740.8 133.5 648.8 120.4 152.7 741.1 133.5 653.8 120.1 152.3 741.4 133.4 658.8 121.1 153.6 740.3 133.0 663.8 120.8 153.2 740.6 133.0 668.8 120.5 152.8 740.9 132.9 673.8 120.2 152.4 741.2 137.1 678.8 119.9 152.0 741.6 137.0 683.8 120.9 153.3 740.6 136.6 688.8 120.6 152.9 740.9 136.6 693.8 120.3 152.5 741.2 136.5 698.8 119.9 152.1 741.5 136.4 703.8 119.6 151.6 741.9 136.4 708.8 120.6 152.9 740.9 136.0 713.8 120.20 152.4 741.2 135.9 718.8 119.9 152.0 741.6 135.9 723.8 119.5 151.5 741.9 135.8 728.8 120.4 152.7 741.0 135.4 Amendment 61 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-40 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 733.8 120.1 152.3 741.4 135.3 738.8 119.7 151.8 741.7 135.3 743.8 119.4 151.3 742.1 135.2 748.8 120.2 152.5 741.2 134.9 753.8 119.9 152.0 741.6 134.8 758.8 119.5 151.5 742.0 134.8 763.8 120.3 152.6 741.1 134.4 768.8 119.9 152.1 741.5 134.3 773.8 119.5 151.6 741.9 134.3 778.8 119.1 151.1 742.3 134.2 783.8 119.9 152.1 741.5 133.9 788.8 119.5 151.5 741.9 133.8 793.8 119.1 151.0 742.4 133.8 798.8 119.9 152.0 741.6 133.4 803.8 119.4 151.4 742.0 133.4 808.8 119.0 150.9 742.4 133.3 813.8 119.7 151.8 741.7 133.0 818.8 119.3 151.3 742.1 133.0 823.8 118.9 150.7 742.6 132.9 828.8 119.6 151.6 741.9 132.6 833.8 119.1 151.0 742.4 132.6 838.8 118.6 150.4 742.8 132.5 843.8 119.3 151.2 742.2 132.2 848.8 118.8 150.6 742.7 132.2 853.8 119.4 151.4 742.1 131.8 858.8 118.9 150.7 742.6 131.8 863.8 119.4 151.4 742.0 131.5 868.8 118.9 150.8 742.6 131.5 873.8 119.4 151.4 742.0 131.2 878.8 118.9 150.7 742.6 131.2 883.8 119.3 151.3 742.1 130.9 888.8 118.7 150.6 742.7 130.9 893.8 119.2 151.1 742.3 130.6 898.8 118.5 150.3 742.9 130.6 903.8 118.9 150.8 742.6 130.4 908.8 118.3 149.9 743.2 130.4 913.8 118.6 150.3 742.9 130.2 918.8 118.8 150.7 742.6 133.8 923.8 119.1 151.0 742.4 133.6 928.8 118.3 150.0 743.1 133.6 933.8 118.5 150.2 743.0 133.3 938.8 118.6 150.4 742.8 133.1 943.8 118.7 150.5 742.8 132.9 948.8 118.7 150.5 742.8 132.7 953.8 118.7 150.5 742.8 132.6 958.8 118.6 150.4 742.9 132.4 963.8 118.5 150.2 743.0 132.3 968.8 118.3 149.9 743.2 132.1 973.8 118.0 149.6 743.5 132.0 978.8 118.4 150.2 743.0 131.7 983.8 118.0 149.6 743.5 131.7 Amendment 61 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-40 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Thousand Btu/sec) (lbm/sec) (Thousand Btu/sec) 988.8 118.2 149.9 743.2 131.4 993.8 118.3 150.1 743.1 131.2 998.8 118.3 150.0 743.2 131.0 1003.8 118.1 149.7 743.4 130.9 1008.8 117.7 149.3 743.7 130.8 1013.8 117.8 149.4 743.6 130.6 1018.8 64.3 81.5 797.2 144.4 1198.8 61.9 78.5 799.5 144.4 1200.0 61.9 78.5 735.1 183.1 1283.4 61.9 78.5 735.1 183.1 1283.5 71.0 88.8 726.0 180.3 1285.0 70.9 88.8 726.0 180.3 1786.3 70.9 88.8 726.0 180.3 1786.4 65.0 74.8 731.9 121.8 3599.9 54.9 63.1 742.1 123.6 3600.0 54.9 63.1 742.1 123.6 3600.1 43.4 49.9 753.6 112.4 10000.0 31.5 36.3 765.4 114.1 18000.0 27.0 31.1 770.0 114.8 18000.1 26.4 30.4 776.7 97.9 18001.0 26.4 30.4 776.7 97.9 18001.1 26.5 30.5 775.3 101.6 30000.0 23.6 27.1 778.2 102.0 30000.1 23.4 26.9 780.3 96.8 106400.0 16.2 18.6 787.4 97.7 106400.1 16.0 18.4 790.4 89.4 1000000.0 7.0 8.0 799.4 90.4 1000000.1 6.9 7.9 802.8 79.6 2592000.0 4.7 5.4 805.0 79.8 2592000.1 4.7 5.4 805.1 79.5 10000000.0 2.2 2.5 807.6 79.7

M&E exiting from the SG side of the break
- M&E exiting from the RV side of the break Amendment 61 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-41 DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow** Time (sec) (lbm/sec) (Btu/sec) (lbm/sec) (Btu/sec) 206.2 158.8785 199070.5 375.0221 107467.9 211.2 158.2352 198264.4 375.6654 107491.3 216.2 157.5905 197456.7 376.31 107514.8 221.2 158.2797 198320.2 375.6208 107193.3 226.2 157.6228 197497.1 376.2777 107219.5 231.2 156.9643 196672 376.9362 107245.7 236.2 156.3046 195845.3 377.596 107272.1 241.2 155.6433 195016.8 378.2572 107298.6 246.2 154.9806 194186.5 378.92 107325.2 251.2 154.3541 193401.5 379.5464 107344.2 256.2 153.845 192763.6 380.0556 107339.1 261.2 154.6281 193744.8 379.2725 106999.8 266.2 154.1055 193090 379.795 106997.6 271.2 153.5815 192433.4 380.3191 106995.5 276.2 153.0557 191774.6 380.8449 106993.6 281.2 152.5282 191113.7 381.3723 106991.9 286.2 151.999 190450.6 381.9016 106990.4 291.2 151.4681 189785.4 382.4325 106989 296.2 150.9354 189117.9 382.9652 106987.8 301.2 150.4009 188448.2 383.4997 106986.8 306.2 149.8646 187776.2 384.036 106986.1 311.2 149.3265 187102.1 384.574 106985.4 316.2 148.7865 186425.4 385.1141 106985.1 321.2 148.2448 185746.7 385.6558 106984.8 326.2 147.7009 185065.1 386.1997 106984.9 331.2 147.1553 184381.6 386.7452 106985.1 336.2 146.6075 183695.2 387.2931 106985.6 341.2 146.058 183006.6 387.8426 106986.3 346.2 145.5063 182315.4 388.3943 106987.3 351.2 144.9525 181621.6 388.948 106988.4 356.2 144.397 180925.5 389.5036 106989.9 Amendment 63 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 361.2 145.0177 181703.3 388.8828 106686.9 366.2 144.446 180986.9 389.4546 106692 371.2 143.8729 180268.8 390.0277 106697 376.2 143.2984 179549 390.6021 106702.2 381.2 142.722 178826.8 391.1786 106707.5 386.2 142.1409 178098.7 391.7597 106713.8 391.2 141.5571 177367.2 392.3435 106720.4 396.2 140.9712 176633 392.9294 106727.4 401.2 140.3821 175895 393.5185 106734.8 406.2 139.7908 175154.1 394.1098 106742.6 411.2 139.1972 174410.3 394.7034 106750.6 416.2 138.6003 173662.4 395.3003 106759.2 421.2 138.0011 172911.6 395.8995 106768.1 426.2 137.3994 172157.7 396.5012 106777.3 431.2 136.7943 171399.5 397.1063 106787.2 436.2 136.1866 170638.1 397.714 106797.4 441.2 135.5763 169873.4 398.3243 106808 446.2 134.9627 169104.6 398.9379 106819 451.2 134.3643 168354.8 399.5363 106826.9 456.2 133.8212 167674.3 400.0793 106823.4 461.2 133.2753 166990.3 400.6253 106820.2 466.2 132.7252 166301.1 401.1754 106817.9 471.2 132.1722 165608.1 401.7284 106816 476.2 131.6161 164911.3 402.2845 106814.6 481.2 131.0565 164210.3 402.844 106813.7 486.2 130.4928 163503.9 403.4078 106813.6 491.2 129.9257 162793.4 403.9749 106814.1 496.2 129.3553 162078.6 404.5453 106815.1 501.2 128.7813 161359.4 405.1193 106816.8 506.2 128.2024 160634.1 405.6982 106819.4 511.2 127.6198 159904.2 406.2808 106822.6 516.2 127.0335 159169.5 406.8671 106826.5 521.2 126.4432 158429.9 407.4573 106831 526.2 125.8483 157684.5 408.0523 106836.5 531.2 125.2485 156932.9 408.6521 106842.9 536.2 123.7128 155008.8 410.1877 107090.8 541.2 137.8955 172779.3 396.0051 107542.9 546.2 137.9269 172818.7 395.9737 107370.6 551.2 137.0137 171674.5 396.8868 107442.2 556.2 136.0901 170517.2 397.8104 107516 Amendment 63 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 561.2 136.0371 170450.8 397.8635 107364.5 566.2 135.0711 169240.4 398.8295 107448.6 571.2 134.9508 169089.6 398.9498 107313.6 576.2 133.9355 167817.6 399.965 107409.6 581.2 133.7417 167574.7 400.1588 107292.9 586.2 132.676 166239.4 401.2246 107401.1 591.2 132.4051 165900 401.4955 107303.7 596.2 131.2824 164493.3 402.6181 107425.8 601.2 130.9326 164055 402.968 107348.3 606.2 129.7816 162612.9 404.1189 107479 611.2 129.3671 162093.5 404.5334 107418.8 616.2 128.8896 161495.2 405.011 107374.6 621.2 128.3454 160813.3 405.5552 107347.3 626.2 127.7267 160038.1 406.1738 107338.7 631.2 127.0312 159166.7 406.8694 107349.6 636.2 126.2549 158194 407.6457 107381 641.2 125.3919 157112.7 408.5086 107434.3 646.2 124.4357 155914.5 409.4649 107511.4 651.2 124.0055 155375.5 409.8951 107452.1 656.2 123.4061 154624.4 410.4945 107436 661.2 122.6155 153633.8 411.2851 107469.1 666.2 121.6187 152384.9 412.2819 107554.9 671.2 120.9314 151523.8 412.9691 107560.3 676.2 120.4203 150883.3 413.4803 107519.9 681.2 119.4311 149643.9 414.4695 107602.4 1003.18 119.4311 149643.9 414.4695 107602.4 1003.28 70.50493 87516.83 463.3956 116459.9 1006.2 70.46474 87466.48 463.4358 118343.1 1529.426 70.46474 87466.48 463.4358 118343.1 1529.526 63.50435 73069.16 470.3962 51987.08 2210 57.84757 66560.38 476.053 53007.66 2210.1 57.84706 66559.79 435.0529 75371.42 3600 50.75419 58398.61 442.1458 76651.09 3600.1 41.24524 47457.46 451.6548 65986.76 3610 41.19645 47401.32 451.7036 65993.89 3610.1 41.11409 47306.56 452.4859 65203.22 10000 29.94601 34456.38 463.654 66812.54 18000 26.94063 30998.34 466.6594 67245.62 18000.1 26.51904 30513.24 469.481 60140.51 18001 26.51869 30512.85 469.4813 60140.55 Amendment 63 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 18001.1 26.49274 30482.99 469.7072 59699.79 30000 23.73453 27309.34 472.4655 60050.36 30000.1 23.52762 27071.27 474.1724 55999.76 100000 16.90459 19450.71 480.7954 56781.94 106400 16.64603 19153.2 481.054 56812.47 106400.1 16.5006 18985.86 482.7994 52625.14 1000000 7.104993 8175.124 492.195 53649.25 1.00E+07 2.283075 2626.944 497.0169 54174.84 1.00E+07 2.257072 2597.025 498.6429 48368.36

M&E exiting from the SG side of the break
- M&E exiting from the RV side of the break Amendment 63 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-43 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MAXIMUM SAFEGUARDS Time (Sec) 1 2 3 4 5 6

                                                     .00                21.40             21.40                223.8               1283.51             1786.30            3600.00 Mass (Thousand lbm)

Initial In RCS & Accumulator 623.83 623.83 623.83 623.83 623.83 623.83 623.83 Added Mass Pumped Injection .00 .00 .00 162.65 1070.16 1470.87 2916.35 Total Added .00 .00 .00 162.65 1070.16 1470.87 2916.35 Total Available 623.83 623.83 623.83 786.48 1693.99 2094.70 3540.18 Distribution Reactor Coolant 426.13 49.29 56.88 106.91 106.91 106.91 106.91 Accumulator 197.70 146.24 138.65 .00 .00 .00 .00 Total Contents 623.83 195.53 195.53 106.91 106.91 106.91 106.91 Effluent Break Flow .00 428.28 428.28 670.90 1578.41 1979.08 3424.56 ECCS Spill 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total Effluent .00 428.28 428.28 670.90 1578.41 1979.08 3424.56 Total Accountable* 623.83 623.81 623.81 777.80 1685.31 2085.99 3531.47 1-End of Blowdown 2-Bottom of core recovery time, which is identical to the end of blowdown time due to the assumption of instantaneous refill 3-End of Reflood 4-Time at which the Broken Loop SG equilibrates at the first intermediate pressure. 5-Time at which the Intact Loop SG equilibrates at the second intermediate pressure. 6-Time at which both SGs equilibrate to 14.7 psia.

   *-The difference between total available mass and total accountable mass at later times in the calculation reflect calculation error due to round off, time step changes, ect.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-44 DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MINIMUM SAFEGUARDS Time (Sec) 0.00 20.40 20.40 206.13 1003.28 1529.43 3600.00 Mass (Thousand lbm) Initial In RCS & 624.11 624.11 624.11 624.11 624.11 624.11 624.11 Accumulator Added Mass Pumped Injection 0.00 0.00 0.00 90.33 515.89 796.80 1845.29 Total Added 0.00 0.00 0.00 90.33 515.89 796.80 1845.29 Total Available 624.11 624.11 624.11 714.45 1140.01 1420.92 2469.41 Distribution Reactor Coolant 426.40 37.40 69.04 118.70 118.70 118.70 118.70 Accumulator 197.72 156.73 125.09 0.00 0.00 0.00 0.00 Total Contents 624.11 194.13 194.13 118.70 118.70 118.70 118.70 Effluent Break Flow 0.00 429.97 429.97 586.90 1012.47 1293.37 2341.87 ECCS Spill 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total Effluent 0.00 429.97 429.97 586.90 1012.47 1293.37 2341.87 Total Accountable 624.11 624.10 624.10 705.60 1131.17 1412.08 2460.57 Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-47 DOUBLE-ENDED HOT-LEG BREAK MASS BALANCE Time (Sec) 0.00 20.40 20.40* Mass (Thousand lbm) Initial In RCS and ACC 624.11 624.11 624.11 Pumped Injection 0.00 0.00 0.00 Added Mass Total Added 0.00 0.00 0.00 Total Available 624.11 624.11 624.11 Reactor Coolant 426.40 64.78 96.42 Distribution Accumulator 197.72 145.57 113.93 Total Contents 624.11 210.35 210.35 Break Flow 0.00 413.75 413.75 Effluent ECCS Spill 0.00 0.00 0.00 Total Effluent 0.00 413.75 413.75 Total Accountable** 624.11 624.10 624.10

-This time is the bottom of core recovery time, which is identical to the end of blowdown time due to the assumption of instantaneous refill.
- -The difference between total available energy and total accountable energy at later times in the calculation reflect calculational error due to round off, time step changes, etc.

Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-49 DOUBLE-ENDED PUMP SUCTION BREAK, MAXIMUM SAFEGUARDS PRINCIPAL PARAMETERS DURING REFLOOD Flooding Total Injection Accum Spill Time(sec) Temp(°F) Rate(in/sec) Carrover Core Downcomer FlowFrac Enthalpy (lbm/sec) Fraction Height(ft) Height(ft) (Btu/lbm) 21.4 223.6 .000 .000 .00 .00 .333 .0 .0 .0 .00 22.2 219.0 26.896 .000 .80 1.83 .000 7558.9 7558.9 .0 99.35 22.3 217.9 28.689 .000 1.03 1.82 .000 7521.3 7521.3 .0 99.35 23.5 216.0 2.751 .316 1.50 5.91 .430 7041.5 7041.5 .0 99.35 24.4 215.7 2.651 .436 1.63 9.10 .451 6771.0 6771.0 .0 99.35 27.6 214.4 4.679 .654 2.03 15.59 .684 5293.1 5293.1 .0 99.35 28.6 214.0 4.409 .682 2.15 15.59 .682 5115.4 5115.4 .0 99.35 31.6 213.0 3.928 .722 2.45 15.59 .671 4675.1 4675.1 .0 99.35 32.7 212.8 4.153 .732 2.55 15.59 .692 5163.7 4396.8 .0 98.40 38.7 212.3 3.671 .751 3.05 15.59 .670 4515.8 3732.7 .0 98.25 45.7 212.9 3.323 .758 3.55 15.59 .653 3988.1 3187.7 .0 98.07 53.7 214.6 3.037 .760 4.05 15.59 .635 3514.6 2700.3 .0 97.88 54.7 214.8 2.427 .754 4.11 15.59 .547 848.2 .0 .0 92.99 55.7 215.1 2.439 .754 4.16 15.59 .550 846.4 .0 .0 92.99 62.7 217.8 2.348 .755 4.50 15.59 .541 849.0 .0 .0 92.99 73.7 223.5 2.190 .756 5.01 15.59 .523 853.6 .0 .0 92.99 86.7 231.6 2.014 .756 5.56 15.59 .500 858.3 .0 .0 92.99 98.7 239.2 1.957 .760 6.04 15.59 .501 858.3 .0 .0 92.99 110.7 245.9 1.901 .763 6.50 15.59 .501 858.3 .0 .0 92.99 124.7 252.6 1.836 .766 7.02 15.59 .502 858.3 .0 .0 92.99 138.7 258.2 1.772 .769 7.51 15.59 .503 858.3 .0 .0 92.99 154.7 263.8 1.699 .772 8.04 15.59 .504 858.4 .0 .0 92.99 170.7 268.5 1.626 .776 8.54 15.59 .504 858.5 .0 .0 92.99 186.7 272.4 1.554 .779 9.01 15.59 .505 858.6 .0 .0 92.99 204.7 276.3 1.474 .782 9.51 15.59 .505 858.7 .0 .0 92.99 223.8 279.7 1.389 .786 10.00 15.59 .505 858.9 .0 .0 92.99 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-51 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE - MAXIMUM SAFEGUARDS Time (sec) 1 2 3 4 5 6

                                                                           .00      21.40       21.40      223.8      1283.51      1786.30      3600.00 Energy (Million Btu)

Initial Energy In RCS, Acc, SG 736.00 736.00 736.00 736.00 736.00 736.00 736.00 Added Energy Pumped Injection .00 .00 .00 15.13 103.25 163.01 378.56 Decay Heat .00 5.78 5.78 25.89 98.27 126.30 212.65 Heat From Secondary .00 -.35 -.35 -.35 4.65 6.22 6.22 Total Added .00 5.42 5.42 40.66 206.17 295.54 597.43 Total Available 736.00 741.42 741.42 776.66 942.16 1031.53 1333.43 Distribution Reactor Coolant 254.23 10.77 11.53 28.01 28.01 28.01 28.01 Accumulator 19.64 14.54 13.78 0.01 -0.01 0.01 0.00 Core Stored 21.51 12.21 12.21 3.91 3.74 3.51 2.71 Primary Metal 125.65 118.82 118.82 99.51 62.17 53.11 41.37 Secondary Metal 83.41 83.06 83.06 76.67 49.38 40.48 31.68 Steam Generator 231.56 230.29 230.29 208.93 132.41 110.07 87.03 Total Contents 736.00 469.69 469.69 417.04 275.70 235.19 190.80 Effluent Break Flow .00 271.25 271.25 351.92 658.77 779.73 1127.39 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total Effluent .00 271.25 271.25 351.92 658.77 779.73 1127.39 Total Accountable* 736.00 740.93 740.93 768.96 934.47 1014.91 1318.18 1-End of Blowdown 2-Bottom of core recovery time. This time is identical to the end of blowdown time due to the assumption of instantaneous refill. 3-End of Reflood 4-Time at which the Broken Loop SG equilibrates at the first intermediate pressure. 5-Time at which the Intact Loop SG equilibrates at the second intermediate pressure. 6-Time at which both SGs equilibrate to 14.7 psia.

-The difference between total available energy and total accountable energy at later times in the calculation reflect calculational error due to round off, time step changes, etc.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-52 DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE - MINIMUM SAFEGUARDS Time (sec) 0.00 20.40 20.40 206.13 1003.28 1529.43 3600.00 Energy (Million Btu) Initial Energy In RCS, Acc, SG 703.47 703.47 703.47 703.47 703.47 703.47 703.47 Added Energy Pumped Injection 0.00 0.00 0.00 8.40 47.97 74.10 207.98 Decay Heat 0.00 5.58 5.58 23.69 79.13 109.09 205.22 Heat From Secondary 0.00 3.39 3.39 3.39 7.15 9.23 9.23 Total Added 0.00 8.97 8.97 35.47 134.26 192.42 422.44 Total Available 703.47 712.43 712.43 738.94 837.72 895.89 1125.90 Distribution Reactor Coolant 252.95 8.51 11.51 30.12 30.12 30.12 30.12 Accumulator 19.64 15.57 12.57 0.00 0.00 0.00 0.00 Core Stored 21.43 11.90 11.90 3.86 3.70 3.50 2.71 Primary Metal 135.07 127.76 127.76 100.42 64.91 54.35 41.77 Secondary Metal 42.40 41.47 41.47 38.01 26.30 20.83 15.94 Steam Generator 231.97 237.20 237.20 212.23 143.66 116.31 89.83 Total Contents 703.47 442.41 442.41 384.64 268.69 225.11 180.38 Effluent Break Flow 0.00 269.55 269.55 337.26 552.00 661.81 937.56 ECCS Spill 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Total Effluent 0.00 269.55 269.55 337.26 552.00 661.81 937.56 Total Accountable 703.47 711.95 711.95 721.91 820.69 886.92 1117.94 Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-55 DOUBLE-ENDED HOT LEG BREAK ENERGY BALANCE Time (sec) 0.00 20.40 20.40* Energy (Million Btu) Initial Energy In RCS, Acc, SG 703.47 703.47 703.47 Added Energy Pumped Injection 0 0 0 Decay Heat 0 5.89 5.89 Heat From Secondary 0 2.72 2.72 Total Added 0 8.61 8.61 Total Available 703.47 712.08 712.08 Distribution Reactor Coolant 252.95 14.18 17.18 Accumulator 19.64 14.46 11.46 Core Stored 21.43 8.69 8.69 Primary Metal 135.07 125.76 125.76 Secondary Metal 42.4 40.74 40.74 Steam Generator 231.97 232.1 232.1 Total Contents 703.47 435.93 435.93 Effluent Break Flow 0 275.65 275.65 ECCS Spill 0 0 0 Total Effluent 0 275.65 275.65 Total Accountable 703.47 711.59 711.59

-This time is the bottom of core recovery time, which is identical to the end of blowdown time due to the assumption of instantaneous refill.
- -The difference between total available energy and total accountable* energy at later times in the calculation reflect calculation error due to round off, time step changes, ect.

Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-58A MSLB Full Double-Ended Rupture (1.4 ft2) at 100.34% Power (with MFIV failure) Forward Flow Time (sec) Flow (lbm/sec) Enthalpy (Bthu/lbm) 0.0 0.0 0.0 0.2 9222.0 1189.5 0.4 9060.0 1190.9 0.6 8956.0 1191.4 0.8 8855.0 1191.4 1.0 8759.0 1191.9 1.4 8581.0 1192.2 1.8 8410.0 1192.6 6.2 7163.0 1197.8 6.6 7085.0 1198.0 7.0 7010.0 1198.3 7.4 6938.0 1198.5 7.8 6867.0 1198.8 8.0 6832.0 1198.9 8.2 2203.0 1199.7 8.4 2192.0 1199.8 8.6 2181.0 1199.9 8.8 2170.0 1200.0 9.0 2160.0 1200.0 9.2 2149.0 1200.1 11.0 2050.0 1201.0 11.2 2039.0 1200.6 11.4 2032.0 1200.8 18.6 1651.0 1203.5 18.8 1642.0 1203.4 19.0 1632.0 1204.0 19.2 1623.0 1203.3 19.8 1595.0 1204.4 20.0 1586.0 1204.3 20.2 1578.0 1203.4 20.7 1563.0 1203.5 21.2 1542.0 1204.3 34.7 1169.0 1203.6 35.2 1160.0 1204.3 35.7 1151.0 1204.2 36.2 1143.0 1203.8 36.7 1135.0 1204.4 44.2 1041.0 1204.6 48.2 1008.0 1204.4 48.7 1005.0 1203.0 49.2 1001.0 1203.8 49.7 997.8 1203.6 132.7 963.0 1203.5 132.8 2.75 1150.0 2210.0 2.75 1150.0 The above forward mass & energy releases include the effects of feedwater addition until isolation. They also include mass & energy releases via the main steam header from the intact steam generator until main steam isolation. Reverse flow mass & energy releases from the initial blowdown of the main steam header itself are to be added to 6 the above forward flow M&E releases. These are 10,743.8 lbm/sec and 12.795 x 10 BThU/sec for a duration of 2.084 sec's. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-58B MSLB Full Double-Ended Rupture (1.4 ft2) at 29.4% Power (with MFIV failure) Forward Flow Time Flow Enthalpy (sec) (lbm/sec) (BThuU/lbm) 0.00 0.00 0.00 0.20 9735.00 1187.47 0.40 9482.00 1188.57 0.60 9316.00 1189.35 0.80 9159.00 1190.09 1.00 9007.00 1190.19 2.00 8332.00 1193.59 3.00 7781.00 1195.73 4.00 7316.00 1197.38 5.00 6920.00 1198.55 7.60 6124.00 1201.01 7.80 1918.00 1202.29 8.00 1902.00 1202.42 10.00 1759.00 1202.96 15.00 1581.00 1204.30 20.00 1399.00 1204.43 30.20 1153.00 1204.68 40.20 1017.00 1203.54 50.20 949.20 1203.12 86.20 955.70 1203.31 176.20 954.60 1203.6 176.30 0.0 0.0 2210.00 0.0 0.0 The above forward flow mass & energy releases include the effects of feedwater addition until isolation. They also include mass & energy releases via the main steam header from the intact steam generator until main steam isolation. Reverse flow mass & energy releases from the initial blowdown of the main steam header itself are to be added to 6 the above forward flow M&E releases. These are 11,398.9 lbm/sec and 13.547 x 10 BThU/sec for a duration of 2.094 sec's. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-62 ACTIVE HEAT SINK DATA FOR MINIMUM POST-LOCA CONTAINMENT PRESSURE I Containment Spray System Parameters A. Maximum spray system flow, total 5000 gpm B. Fastest post LOCA initiation of Containment Spray System 0.0 sec. II Fan Coolers A. Maximum number of fan coolers operating 4 B. Fastest post LOCA initiation of fan coolers 0.0 sec. C. Performance data See Figure 6.2.1-303 for fan cooler atmosphere heat removal rate. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-63 PASSIVE HEAT SINK DATA FOR MINIMUM POST-LOCA CONTAINMENT PRESSURE Heat Sink Description Slab 2 Number Description Slab* Material Material Thickness (in.) Surface Area, ft (2)

1. Containment Dome Carbon Steel .50 26546 Concrete 30 (2)

2. External Cylinder Wall Carbon Steel .375 63065 Concrete 54 (2)

3. 1" Steel Liner Carbon Steel 1.0 2280 Concrete Concrete 54

4. Concrete Concrete 45 82525

5. Stainless Steel Liner Stainless Steel .1872 6756 Concrete Concrete .60

6. Sump Concrete 45 29320 (3)

7. Piping Carbon Steel .19656 5703 (3)

8. Piping Carbon Steel .41808 3870 (2)

9. Structural Heat Sink Carbon Steel .312 53810 (5)

10. Electrical Carbon Steel .17448 33066

11. Embedded Stainless Stainless Steel .39024 1030 Concrete 3.2244

12. Not Embedded Stainless Stainless Steel .40068 3242 (2)

13. Structural Heat Sinks Carbon Steel 1.0 30300 (2)

14. Not Embedded Structural Carbon Steel .17375 119467 (2)

15. Structural Heat Sinks Carbon Steel .5004 66753 (2)

16. Embedded Structural Carbon Steel .3405 3472 Concrete 3.2244 (2)

17. Embedded Structural Carbon Steel 1.444 13899 Concrete 3.2244 (4)

18. Ductwork Carbon Steel .1248 5430 (5)

19. Ductwork Galvanized .029028 39672 Carbon Steel (2)

20. Seismic Hangers Carbon Steel .18756 84386

Metal Coatings for individual slabs are defined via superscripts (2), (3), (4), (5). Properties of the coatings are provided in the thermophysical property listing by number.

Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.2.1-63 (Continued) Thermophysical Properties (1) Metal Thermal Conductivity: Carbon Steel: 26 Btu/hr.-ft.-F Stainless Steel: 9.4 Btu/hr.-ft.-F Thermal Capacity: 53.9 BTU/cu.ft. F) (2) Paint is applied to outer and inner surfaces of the bare carbon steel plate casing. Thickness Range: 5 mils Thermal Capacity: 42.6 BTU/cu.ft - F Thermal Conductivity (Paint System): .23 BTU/hr-ft-F (3) Paint is applied only to outer surface of carbon steel, uninsulated pipe Paint Thickness Range: 5 mils Thermal Capacity: 147 BTU/cu.ft-F Thermal Conductivity: .23 BTU/hr-ft-F (4) Paint is applied to outer only of bare carbon steel sheet metal, Thickness Range: 8 mils Thermal Capacity: 42.6 BTU/cu.ft-F Thermal Conductivity (Paint System): .23 BTU/hr-ft-F (5) Galvanizing:Zinc is applied according to ASTM A 525 coating designation 90 (commercial) coating thickness 2 is approximately 0.90 Oz./ft galvanized on both sides. Thickness Range: 1.513 mils Thermal Capacity: 40.6 BTU/cu.ft-F Thermal Conductivity: 64 BTU/hr.-ft-F (6) Concrete Thermal Conductivity: .92 BTU/hr-ft-F Thermal Capacity: 22.62 BTU/cu.ft-F Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-64 SINGLE FAILURE ANALYSIS-CONTAINMENT VACUUM RELIEF SYSTEM Component Identification and Quantity Failure Mode Effect on System Method of Detection Monitor Remarks Compressed air system Fails Loss of normal air supply Low air pressure alarm CRI* Seismic Class I accumulators have to accumulators sufficient stored air to operate their respective vacuum valves Air accumulator Fails Loss of one vacuum relief Periodic testing CRI Redundant vacuum relief subsystem subsystem available Vacuum relief valve or check valve Fails to open Loss of one vacuum relief Valve position indication CRI Redundant vacuum relief subsystem subsystem plus high P alarm available Vacuum relief outside to Fails Loss of one vacuum relief Periodic testing CRI Redundant vacuum relief P switch containment P switch subsystem available Differential pressure sensor Fails Loss of one valve CRI Redundant vacuum relief subsystem actuation signal available Outside air damper Fails to open Partial loss of outside air Damper position indication CRI Redundant vacuum relief subsystem available

CRI = Control Room Indication.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-65 POST-ACCIDENT MONITORING CONTAINMENT ATMOSPHERE TEMPERATURE AND CONTAINMENT SUMP WATER TEMPERATURE Tag. No. Instrumentation Service Time Response Range Accuracy TE-7541A,B,C Thermocouple Cont. Dome Temp. 1 Second 0-400 F +/- 2 1/4 F 32-600 F TE-7542A,B,C +/- 3/8% 600 -1600 F TT-7541A,B,C Thermocouple Amplifier Cont. Dome Temp. 10 millisecond 2-100 mV +/- 0.1% at 5m VDC input span TT-7542A,B,C 0-400 F TY-7541 TY-7542 Isolator Cont. Dome Temp. N/A (-) 10 - +/- 0.1% of signal span 0 - 10 VDC TS-7541 TS-7542 Signal Comparator Cont. Dome Temp. 10 millisecond 0 - 10 VDC +/- 0.35% of input span 0-400 F TR-0005 Temperature Recorder Cont. Dome Temp. 0.5 Second Full Scale 0 - 10 VDC +/- 0.25% of span 0-400 F TI-7541 TI-7542 Temperature Indicator Cont. Dome Temp. N/A 0 - 10 VDC +/-1% of span 0-400 F TE-7133A TE-7133B Thermocouple CS Recirc. Sump Temp. 1 Second 50-250 F +/- 1 1/4 F 32-600 F

                                                                                               +/- 3/8%  600 F - 1600 F Amendment 61                                                                                                                  Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-66 LOCA M&E RELEASE ANALYSIS CORE DECAY HEAT FRACTION Decay Heat Generation Rate Time (sec) (Btu/hr) 1.0000E+00 .65709E-01 1.2000E+00 .64706E-01 1.4000E+00 .63757E-01 1.6000E+00 .62939E-01 1.8000E+00 .62245E-01 2.0000E+00 .61586E-01 2.5000E+00 .60288E-01 3.0000E+00 .59144E-01 3.5000E+00 .58123E-01 4.0000E+00 .57203E-01 4.5000E+00 .56401E-01 5.0000E+00 .55668E-01 6.0000E+00 .54367E-01 7.0000E+00 .53255E-01 8.0000E+00 .52275E-01 9.0000E+00 .51427E-01 1.0000E+01 .50663E-01 1.2000E+01 .49337E-01 1.4000E+01 .48212E-01 1.6000E+01 .47242E-01 1.8000E+01 .46392E-01 2.0000E+01 .45633E-01 2.5000E+01 .44044E-01 3.0000E+01 .42749E-01 3.5000E+01 .41652E-01 4.0000E+01 .40701E-01 4.5000E+01 .39875E-01 5.0000E+01 .39136E-01 6.0000E+01 .37860E-01 7.0000E+01 .36793E-01 8.0000E+01 .35879E-01 9.0000E+01 .35085E-01 1.0000E+02 .34386E-01 1.2000E+02 .33212E-01 1.4000E+02 .32250E-01 1.6000E+02 .31442E-01 1.8000E+02 .30749E-01 2.0000E+02 .30143E-01 2.5000E+02 .28914E-01 3.0000E+02 .27943E-01 3.5000E+02 .27138E-01 4.0000E+02 .26445E-01 4.5000E+02 .25828E-01 5.0000E+02 .25274E-01 6.0000E+02 .24303E-01 7.0000E+02 .23482E-01 8.0000E+02 .22762E-01 9.0000E+02 .22108E-01 1.0000E+03 .21517E-01 Amendment 61 Page 1 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-66 LOCA M&E RELEASE ANALYSIS CORE DECAY HEAT FRACTION Decay Heat Generation Rate Time (sec) (Btu/hr) 1.2000E+03 .20493E-01 1.4000E+03 .19622E-01 1.6000E+03 .18862E-01 1.8000E+03 .18192E-01 2.0000E+03 .17599E-01 2.5000E+03 .16384E-01 3.0000E+03 .15440E-01 3.5000E+03 .14685E-01 4.0000E+03 .14068E-01 4.5000E+03 .13545E-01 5.0000E+03 .13103E-01 6.0000E+03 .12391E-01 7.0000E+03 .11850E-01 8.0000E+03 .11417E-01 9.0000E+03 .11055E-01 1.0000E+04 .10748E-01 1.2000E+04 .10619E-01 1.4000E+04 .10230E-01 1.6000E+04 .99059E-02 1.8000E+04 .96274E-02 2.0000E+04 .93890E-02 2.5000E+04 .88946E-02 3.0000E+04 .85120E-02 3.5000E+04 .81995E-02 4.0000E+04 .79355E-02 4.5000E+04 .77038E-02 5.0000E+04 .75007E-02 6.0000E+04 .71598E-02 7.0000E+04 .68673E-02 8.0000E+04 .66223E-02 9.0000E+04 .64093E-02 1.0000E+05 .62238E-02 1.2000E+05 .58963E-02 1.4000E+05 .56287E-02 1.6000E+05 .53990E-02 1.8000E+05 .51969E-02 2.0000E+05 .50192E-02 2.5000E+05 .46447E-02 3.0000E+05 .43460E-02 3.5000E+05 .40984E-02 4.0000E+05 .38880E-02 4.5000E+05 .37081E-02 5.0000E+05 .35506E-02 6.0000E+05 .32873E-02 7.0000E+05 .30768E-02 8.0000E+05 .29035E-02 9.0000E+05 .27605E-02 1.0000E+06 .26387E-02 1.2000E+06 .24400E-02 1.4000E+06 .22838E-02 Amendment 61 Page 2 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.1-66 LOCA M&E RELEASE ANALYSIS CORE DECAY HEAT FRACTION Decay Heat Generation Rate Time (sec) (Btu/hr) 1.6000E+06 .21563E-02 1.8000E+06 .20489E-02 2.0000E+06 .19560E-02 2.5000E+06 .17679E-02 3.0000E+06 .16229E-02 3.5000E+06 .15069E-02 4.0000E+06 .14117E-02 4.5000E+06 .13335E-02 5.0000E+06 .12668E-02 6.0000E+06 .11579E-02 7.0000E+06 .10731E-02 8.0000E+06 .10026E-02 9.0000E+06 .94319E-03 1.0000E+07 .89113E-03 Amendment 61 Page 3 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-1 CONTAINMENT COOLING SYSTEM COMPONENTS NOTE: All air quantities are actual cfm. CONTAINMENT FAN COOLER SAFETY CLASS 2 UNITS (4) Normal Operating Conditions Design Basis Accident Conditions No. of Units 2 fans per unit and 2 units 1 fan per unit half speed, 4 units operating starting and 2 units operating Fan Cooler Unit Operating Capacity ACFM 125,000 31,250 Actual Air Mixture Flow (ACFM) at Fan Inlet 62,500 31,250 Design Ambient Pressure, psig 0 45.0(1) Ambient Temp, F 120 258 Total Pressure, in. WG 7.9 5.1 Fan RPM 1770 870 Outlet Velocity, FPM 5800 2560 Brake HP 101.2 32.8 Motor HP 125 62.5 Cooling Water Flow - GPM 1360 1360 Entering Water Temp. F 95 95 Leaving Water Temp. F 98 179 Number of Coil Banks 4 4 Number of Rows 6 6 Fins per inch 6 6 Face Area (Sq. Ft.) 160 160 No. of Coils High 4 4 Coil Size (L x W) 60" x 24" 60" x 24" Cooling Coil Entering Air Mixture Temp. 120/98.4 258/258 DB/WB, F Water Pressure Drop Coil and Manifolds 31 31 (Ft H2O) Cooling Coil Leaving Air Mixture Temp. 103.21/97.4 248/248 DB/WB, F Entering Air Mixture Flow lb/hr - 422,400 Steam Condensed lb/hr - 56,746 6(2) 6 (4)(5) Btu/hr at 95F Entering WaterTemp. 2.19 x 10 55.5 x 10 at 45.0 psig . See figures 6.2.1-16, 6.2.2-4 and Table 6.2.2-3 6(2) Btu/hr at 95F Entering Water Temp. and one 1.35 x 10 - fan running full speed Amendment 61 Page 1 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.2.2-1 (Continuted) CONTAINMENT FAN-COIL NNS UNITS (4) Normal Operating Conditions Design Basis Accident Conditions No. of Units 3 units with 2 fans per unit (one - standby and one operating fan) Actual Airflow at Inlet (ACFM) 91,000 - Outlet Velocity, FPM 3218 - Motor Brake HP 81 - Motor HP 100 - Fan RPM 1170 - Cooling Coil Entering Air Temp. 120/98.4 - DB/WB,F Cooling Coil Leaving Air Temp. 99.8/95.0 - DB/WB,F 6 Coil Capacity Btu/hr 1.866 x 10 - Face Area (Sq. Ft.) 110 - Entering Water Temp. F 95 - Leaving Water Temp. F 99.7 - Water Flow GPM 800 - Water Pressure Drop (Ft. H2O) 16.8 - No. of Coils 4 - Size of each Coil (L x W) 94 1/2" x 42 3/4" - No. of Rows 8 - Fins per Inch 10 0 EQUIPMENT TABULATION Safety Related Fan Coolers Each of 4 identical units containing the following:

1. Supply Fan Quantity 2 fans per unit Type Axial Material Carbon Steel Air Flow, Each Fan, cfm 62,500 nominal/31,250 accident Code AMCA -Air Moving and Conditioning Association AFBMA -Anti-Friction Bearing Manufacturers Association

2. Supply Fan Motor Quantity Per Fan 1 Capacity 125/62.5 HP, 460 V, 60 Hz, 3ph., 2 speed Insulation Class RN Enclosure TEAO (Totally enclosed air over)

Code NEMA MG-1 IEEE 334

3. Cooling Coil Type (Service Water) fin tube Material Cu/Ni-90/10 tubes, copper fins Amendment 61 Page 2 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.2.2-1 (Continued) Non Safety Related Fan Coil Units Each Unit containing the following: Quantity 3 Identical Units

1. Supply Fan 2 fans per unit (1 Standby)

Type Axial Material Carbon Steel Air Flow, Each Fan, cfm 91,000 Total Pressure, in. wg 5.0 Code Air Moving and Conditioning Association (AMCA), Anti-Friction Bearing Manufacturers Association (AFBMA)

2. Supply Fan Motors Quantity Per Fan 1 Capacity 100 HP, 460 V, 60 Hz, 3 ph Insulation Type F Enclosure TEAO Code NEMA MG-1 IEEE 334 3.

Type (Service Water) fin tube Material Cu/Ni 90/10 tubes, copper fins Air Flow. acfm per coil bank 22,750

NOTES: (1) 45.0 psig is the design pressure for the containment structure. This pressure is used to establish the design conditions for cooling capacity. (2) For two fans the entering condition for normal operation is 120°F, 265 grains moisture per lbm of dry air. For one fan the entering condition for normal operation is 120°F, 120 grains moisture per lbm of dry air. (3) DELETED BY AMENDMENT No. 51 (4) Fan cooler performance assumed for the peak pressure & temperature containment analyses for MSLB and LOCA DBA's assume a more conservative degraded performance than listed in the above Table 6.2.2-1. Refer to Table 6.2.1-6 for actual performance assumed. (5) The fan cooler performance at 45 psig and 258°F is shown conservatively instead of at higher LOCA/MSLB temperatures resulting from power uprate. Amendment 61 Page 3 of 3

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-3 CONTAINMENT FAN COOLER PERFORMANCE DATA(1)(2)(3) CONTAINMENT GAS CONDITIONS HEAT PRESSURE FLOW X 1000 CFM TEMPERATURE WATER TEMPERATURE LOAD PSIA IN OUT IN °F OUT °F IN °F OUT °F BTU/HX10 59.7 38.46 31.25 258 248 95 179 55.5 41.3 38.15 31.25 220 204 95 153 38.6 30.2 36.55 31.25 180 158 95 128 22.1 25.3 34.74 31.25 150 129 95 112 11.7 16.1 33.24 31.25 120 106 95 101 4.2 59.7 39.06 31.25 258 247 80 171 60.2 41.3 38.87 31.25 220 202 80 144 42.8 30.2 37.38 31.25 180 154 80 118 25.5 25.3 35.57 31.25 150 123 80 101 14.5 16.1 34.33 31.25 120 98 80 90 6.5 NOTE (1) The fan cooler performance data at 59.7 psia and 258°F is shown conservatively instead of at higher LOCA/MSLB temperatures resulting from power uprate. (2) Based on cooling water flow rate of 1360 gpm/unit. (3) Refer to Figure 6.2.2-4 for graphical plots. Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-4 PRIMARY SHIELD COOLING SYSTEM COMPONENTS SAFETY CLASS - 3 UNITS

1. Supply Fans Quantity 2 (one standby)

Type Axial Material Carbon Steel Air flow, each fan, acfm 18,000 Code Air Moving and Conditioning Association (AMCA), Anti-Friction Bearing Manufacturers Association (AFBMA)

2. Supply Fan Motors Quantity per fan 1 Capacity 40 Hp, 460 V, 60 Hz, 3 ph Insulation Class H, Type RH, Class H Enclosure TEAO Code NEMA MG-1 IEEE Std. 334 Note: All air quantities are actual cfm.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-5 REACTOR SUPPORTS COOLING SYSTEM COMPONENTS SAFETY CLASS 3 UNITS

1. Supply Fans Quantity 2 (one standby)

Type Axial Material Carbon Steel Air Flow, each fan, acfm 27,600 Code Air Moving and Conditioning Association (AMCA), Anti-Friction Bearing Manufacturers Association (AFBMA)

2. Supply Fan Motors Quantity per Fan 1 Capacity 50 HP, 460 volt, 60 HZ, 3 ph Insulation Class RH Enclosure TEAO Code NEMA MG-1 IEEE Std. 334 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-6 SINGLE FAILURE ANALYSIS CONTAINMENT COOLING SYSTEM Component Malfunction Comments a) Injection Phase

1) Active failure (a) Containment cooling unit fan Fails to start There are two fans in each cooling unit. Only one fan in each unit is required for accident cooling. The operator can select the alternate fan in the event of fan failure.

(b) One emergency electric train Fails There are redundant power sources. Four equal capacity containment cooling units are provided with two connected to each train. (c) Damper Fails to achieve There is redundancy in the ventilation distribution safe position system. Failure of a single damper will not prevent system from providing adequate air distribution within containment. b) Recirculation

1) Active failure (a) Containment cooling unit fan Fails to operate Same as A.1)a).

(b) One emergency electric train Fails Same as A.1)b) (c) Damper Fails to achieve Same as A.1)c) safe position

2) Passive (a) Fan shaft, blade, etc Fails Same as A.1)a).

(b) Component cooling unit Clogs, rupture or major Same as A.1)b). The operator would be alerted cooling coil leakage and the affected unit could be isolated. (c) Electric cable, (worst case for Fails Same as A.1)b) a complete train) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-7 SINGLE FAILURE ANALYSIS CONTAINMENT SPRAY SYSTEM Component Malfunction Comments A. Injection Phase

1) Active failure a) Containment spray pump Fails to start Two equal capacity pumps are provided. A single pump operating in combination with the Containment Cooling System will provide the required heat and iodine removal capability.

b) CSS header isolation valve Fails to open Two equal capacity headers are provided. A single header operating in combination with the Containment Cooling System will provide the required heat and iodine removal capability. c) Spray additive tank isolation valve Fails to open Two valves are provided in parallel. d) Containment spray pump injection Fails to open Same as A.1)b). supply valve e) One emergency electric power Fails A redundant emergency diesel train generator and associated electric distribution system will supply power for minimum system requirements. B. Recirculation Phase

1) Active failure a) Containment spray pump Fails to operate Same as A.1)a).

b) Containment spray pump Fails to open Same as A.1)b). recirculation supply valve c) One emergency power train Fails to operate Same as A.1)e)

2) Passive Failure a) Spray nozzles Clogged Redundant spray nozzles are provided to fulfill minimum system requirements.

b) Recirculation piping or spray Ruptures or major leakage Same as A.1)b). piping; valve body, or pump casing c) Spray additive educator Fails to operate Two are provided, one for each header. d) CSS pump shaft Fractures Same as A.1)a) e) Electric cable (worst case for a Fails Same as A.1)e) complete train Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-8 CSS PUMP NPSH EVALUATION Flow* Minimum NPSH Required NPSH Available (GPM/Pump) (FT) (Ft) (A) During the Injection Phase of 2464 13.0 92.0 Containment Spray (B) During Recirculation Phase of 2251 12.4 25.5 Containment Spray

The maximum expected flow based on conservative assumptions is used to calculate minimum NPSH required.

Amendment 62 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.2-9 CONTAINMENT SPRAY SYSTEM COMPONENT PARAMETERS A. Containment Spray Pumps Number of Pumps 2 Type of Pump Centrifugal Design Flow, gpm 2275 Design Head, ft 425 Driver Electric Motor Driver horsepower (approximate) 350 Material (casing) SA-182 Type F 304 Code ASME III, Safety Class 2 B. Refueling Water Storage Tank Quantity 1 Material Stainless Steel Type 304 Maximum Volume, gal 469,260 Minimum Volume, (solution) gal. 434,302 Normal Pressure, psig Atmospheric Operating Temperature, F 40-125 Design Pressure, psig Atmospheric Design Temperature, F 200 Boric Acid (as ppm B) 2400-2600 ppm Code ASME III, Code Class 2 C. Cavitating Venturi Quantity (per train) 1 Size, inches 8 Design Flow, gpm 2770 Material 316 SS Code ASME III, Class 2 D. Containment Spray Nozzles Quantity (per train) 106 Nozzle size, inches 3/8 Design Flow, gpm 15.2 Material ASME A-351 GrCF8 Code ASME III, Class 2 E. Motor Operated Isolation Valves Quantity, (per train) 1 Size, inches 8 Type Gate Design, Pressure, psig 300 Design Temperature, F 300 Material 304 SS Code ASME III, Class 2 Operator Motor Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 SIGNAL INDEX FOR TABLE 6.2.4-1

1. Containment isolation Phase A (Signal 'T')

2. Containment isolation Phase 8 (Signal 'P')

3. Safety injection actuation (Signal 's')

4. Two out of four RWST low-low level

5. Main steam line isolation

6. Two out of three low-low level in any steam generator

7. Loss of main feedwater pumps

8. Containment ventilation isolation

9. Containment inside/outside differential pressure< - 2.5 in. wg.

10. Containment inside/outside differential pressure< - 0.25 in. wg.

11. Containment spray actuation

12. (Deleted)

13. Main feedwater line isolation

14. Auxiliary feedwater isolation

15. Loss of offsite power

16. Any steam generator low-low level Amendment 63 Page 1 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 DEFINITIONS FOR TABLE 6.2.4-1

1. ACCIDENT SIGNAL - A signal which automatically initiates positioning of valves other than containment isolation valves to positions required to perform their safety-related function.

2. AUTOMATIC ACTUATION - Initiation of a power-operated valve by automatic means without any action by a plant operator upon receipt of an accident/isolation signal from a protection system.

3. CLOSED SYSTEM - A safety-related piping system which penetrates and is a closed system either inside or outside the Containment. The system is subject to the pump and valve operability test requirements and the inservice inspection requirements of the ASME Code Section XI. Under normal operating or LOCA conditions for closed systems inside Containment, the fluid in the system does not communicate directly with either the Reactor Coolant System or containment atmosphere. Under normal operating or LOCA conditions for closed systems outside Containment. the fluid in the system does not communicate directly with the environs.

4. CONTAINMENT ISOLATION SIGNAL - A signal which automatically initiates positioning of valves other than engineered safety feature valves to positions required to perform their containment isolation function.

5. CONTAINMENT ISOLATION VALVE - A valve which establishes a mechanical barrier in appropriate fluid systems penetrating the Containment. which could otherwise represent open paths to the environment from inside the Containment for fission products.

6. LEAKAGE PATH - A penetration that is not part of a safety-related closed system and that could provide a direct path to the environment.

7. LOCKED CLOSED ISOLATION VALVE - A valve that is in a closed position by administrative controls by one of the following:

a. A mechanical device locking the valve in the closed position.

b. A normally closed valve with a seal or lock on any manual override. if present. and a seal or lock on the power breaker or power source in a manner that prevents power from being supplied to the valve.

8. POWER TRAIN - The source of emergency electrical power from one or both of the redundant A and B emergency buses. See Section 8.3.

9. REMOTE MANUAL ACTUATION - Initiation of a Power-operated valve by a discrete manual action such as operation of a control switch.

10. RESPONSE TIME - Maximum stroke time for a valve to move to its safety- related position. Valves that close on automatic actuation that are not required to close within specific times by assumptions in accident analysis are indicated as closing in 60 seconds or less. Table 16.3-5 "Containment Isolation Valves establishes a typical maximum isolation time (sec) of 60 seconds. Standard Review Plan Section 6.2.4 makes a general recommendation of "less than 1 minute. regardless of valve size."

11. TYPE C TEST - 10CFR50 Appendix J test as described in Section 6.2.6.

Amendment 63 Page 2 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES FOR TABLE 6.4.2-1

1. The inside containment barrier is a closed system inside Containment. The closed system is subject to the inservice inspection requirements of ASME Section XI. The system remains at a pressure greater than 45 psig post-LOCA.

2. The outside containment isolation barrier is a closed ESF system outside the Containment. The system is subject to the pump and valve operability test requirements and the inservice inspection requirements of the ASME Code Section XI.

3. Valves which do not require a Type C test because they do none of the following:

a. Provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal operation.

b. Are required to close automatically upon receipt of a containment isolation signal in response to controls intended to effect containment isolation.

c. Are required to operate intermittently under post accident conditions.

4. Remote manual instrument isolation valve. Valve is maintained open for both normal and post-LOCA operation.

5. Manual or remote-manual which is locked (physically locked or administratively controlled) and remains locked post-LOCA.

6. The valve is used periodically during operation. Any leakage through the valve would be detected during normal operation.

7. Valve is only opened when adding water to the pressurizer relief tank or RCP standpipe. It is interlocked to close on a Phase A containment isolation signal.

8. Valve is only opened when adding water to the accumulators. It is interlocked to close on a Phase A containment isolation signal.

9. Valve is opened only to verify leaktightness of accumulator check valve. It is interlocked to close a Phase A containment-isolation signal.

10. Valve is opened only when charging nitrogen to the accumulators. It is interlocked to close on a Phase A containment isolation signal.

11. Center shaft butterfly valve may be tested in the reverse direction. Leakage is equivalent since the same sealing surfaces are tested when test pressure is applied from either direction.

Amendment 63 Page 3 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES TO TABLE 6.2.4-1 (Continued)

12. Wedge-disk gate valve tested in the reverse direction. Results obtained in this configuration are equivalent to testing in the accident direction. Because of the disk seat design. testing in either direction measures the leakage across both seating surfaces.

13. Globe valve tested in the reverse direction. The results obtained are conservative since test pressure tends to unseat the valve disk.

14. Diaphragm valve tested in the reverse direction. Leakage is equivalent since the same seating surface is tested when test pressure is applied from either direction.

15. These penetrations are connected directly to the sump. During an accident. they will be filled with water. This water seal will exist during the entire post-accident period.

16. Both the pressure sensor and the hydraulic isolators have an internal bellows which serve as isolation barriers for the capillary tubes.

17. The isolation valve is sealed with a seal-fluid with sufficient fluid inventory to assure the sealing function for at least 30 days at 1.10 Pa.

18. Valves which may be opened by operator action for post-accident long-term cooling.

19. Valves are closed on containment Phase A isolation signal. They are then opened remote manually for intermittent sampling.

20. Valves AF-Vl89. 190. and 191 are opened for steam generator Wet-Layup.

21. The system fluid is water. However. the penetration is drained before the isolation valves are closed. (Reference ESR 9600537 and Generic Letter 96-06)

22. The valve stroke time at degraded voltage for valves VB and V9 is 76.9 seconds. At nominal voltage, the stroke time is 60.7 seconds. These stroke times will ensure these valves can perform their safety function in the required response time. For the AFW actuation times. refer to plant operating manual procedure PLP-106.

Amendment 63 Page 4 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 ABBREVIATIONS FOR TABLE 6.2.4-1 Fluid Valve Type A - Air BF - Butterfly G - Gas CK - Check S - Steam DA - Diaphragm W - Water GA - Gate CF - Capillary Filling GL - Globe RG - Regulating Actuator RL - Relief AO - Air Operator XC - Excess Flow Check Valve EH - Electro-hydraulic M - Manual Primary/Secondary MO - Motor Operator Actuation Modes SA - Self-Actuating A - Automatic SO - Solenoid Operator M - Manual RF - Reverse Flow Valve Position RM - Remote Manual AI - As Is C - Closed Cy - Cycle LC - Locked Closed LO - Locked Open O - Open TH - Throttled TL - Locked Throttled Amendment 63 Page 5 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT R1 - 4 RL SA - - - - - C C C C C YES YES NO R4 - 6 RL SA - - - - - C C C C C YES YES NO R7 - 8 RL SA - - - - - C C C C C YES YES NO R10 - 10 RL SA - - - - - C C C C C YES YES NO R13 - 13 RL SA - - - - - C C C C C YES YES NO MAIN STEAM 042 57 S YES NO P18 A GL EH A RM - - - C C CY C - NO YES NO 1 LOOP A V1 A&B 27 GL AO A RM 5 - 5 O C C C C NO YES NO F1 A&B 37 GL AO A RM 5 - 10 C C C C C NO YES NO V59 A&B 25 GL AO A RM 5 - 60 O C C C C NO YES NO V122 A&B 4 GL AO A RM 1 - 60 O C C C C NO YES NO R2 - 4 RL SA - - - - - C C C C C YES YES NO R5 - 6 RL SA - - - - - C C C C C YES YES NO R8 - 8 RL SA - - - - - C C C C C YES YES NO R11 - 10 RL SA - - - - - C C C C C YES YES NO R14 - 13 RL SA - - - - - C C C C C YES YES NO P19 B 30 GL EH A RM - - - C C CY C - NO YES NO V2 A&B 27 GL AO A RM 5 - 5 O C C C C NO YES NO 1 F2 A&B 37 GL AO A RM 5 - 10 C C C C C NO YES NO MAIN STEAM 042 57 S YES NO V60 A&B 28 GL AO A RM 5 - 60 O C C C C NO YES NO LOOP B V8 A 26 GA MO A RM - 15,6 60.7 C C O AI C YES YES NO 22 V124 A&B 4 GL AO A RM 1 - 60 O C C C C NO YES NO Amendment 63 Page 6 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT R3 - 4 RL SA - - - - - C C C C C YES YES NO R6 - 6 RL SA - - - - - C C C C C YES YES NO R9 - 8 RL SA - - - - - C C C C C YES YES NO R12 - 10 RL SA - - - - - C C C C C YES YES NO R15 - 13 RL SA - - - - - C C C C C YES YES NO 1 MAIN STEAM P20 A 30 GL EH A RM - - - C C CY C - NO YES NO 042 57 S YES NO LOOP C V3 A&B 27 GL AO A RM 5 - 5 O C C C C NO YES NO F3 A&B 37 GL AO A RM 5 - 10 C C C C C NO YES NO 22 V61 A&B 28 GL AO A RM 5 - 60 O C C C C NO YES NO V9 B 26 GA MO A RM - 15,6 60.7 C C O AI C YES YES NO V126 A&B 4 GL AO A RM 1 - 60 O C C C C NO YES NO V26 A&B 4 GA EH A RM 13 - 8 O C C C C NO YES NO V89 - 9 GL M M - - - - LC LC LC LC LC NO YES NO FEEDWATER 044 57 W YES NO 1 LOOP A V90 - 4 GL M M - - - - LC LC LC LC LC NO YES NO V27 A&B 4 GA EH A RM 13 - 8 O C C C C NO YES NO V91 - 9 GL M M - - - - LC LC LC LC LC NO YES NO FEEDWATER 044 57 W YES NO 1 LOOP B V92 - 4 GL M M - - - - LC LC LC LC LC NO YES NO Amendment 63 Page 7 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V28 A&B 4 GA EH A RM 13 - 8 O C C C C NO YES NO V93 - 9 GL M M - - - - LC LC LC LC LC NO YES NO FEEDWATER 044 57 W YES NO 1 LOOP C V94 - 4 GL M M - - - - LC LC LC LC LC NO YES NO R500 - 17 RL SA - - - - - C C C C C NO YES YES V511 A 17 GL AO A RM 1 - 10 CY C C C C NO YES YES CVCS - NORMAL V512 A 16 GL AO A RM 1 - 10 O C C C C NO YES YES 803 55 W YES NO LETDOWN V513 A 16 GL AO A RM 1 - 10 CY C C C C NO YES YES V518 B 1 GL AO A RM 1 - 10 O C C C C NO YES YES V515 - 2 CK SA - - - - - O C C - C NO YES YES CVCS - NORMAL 803 55 W YES NO CHARGING V610 A 1 GA MO A RM 3 - 10 O C C AI C NO YES YES V25 - 2 CK SA - - - - - O O O - C YES YES NO CVCS - SEAL 803 55 W YES NO WATER TO RCP 2, 17 A V522 B 1 GL MO RM M - - - O O O AI C YES YES NO Amendment 63 Page 8 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V26 - 2 CK SA - - - - - O O O - C YES YES NO CVCS - SEAL 803 55 W YES NO WATER TO RCP 2, 17 B V523 B 1 GL MO RM M - - - O O O AI C YES YES NO V27 - 2 CK SA - - - - - O O O - C YES YES NO CVCS - SEAL 803 55 W YES NO WATER TO RCP 2, 17 C V524 B 1 GL MO RM M - - - O O O AI C YES YES NO V67 - 4 CK SA - - - - - C C C - C NO YES YES CVCS - SEAL WATER RETURN V516 A 2 GL MO A RM 1 - 10 O O C AI C NO YES YES 803 55 W NO NO

                                                                                                                                     & EXCESS LETDOWN V517            B             2                      GL           MO         A                        RM                      1                  -                 10              O        O          C               AI              C      NO                  YES                     YES SAFETY            V581            -             1                      CK           SA         -                            -                   -                  -                 -               C        O          O               -               O      YES                 YES                     NO INJECTION - LOW 810          55               W       YES                 NO                                                                                                                                                                                                                                                                                                                                                   2, 17 HEAD TO COLD LEGS             V579            A             4                      GA           MO         RM                         M                     -                  -                 -               O        O          O               AI              O      YES                 YES                     NO SAFETY            V580            -             1                      CK           SA         -                            -                   -                  -                 -               C        O          O               -               O      YES                 YES                     NO INJECTION - LOW 810          55               W       YES                 NO                                                                                                                                                                                                                                                                                                                                                   2, 17 HEAD TO COLD LEGS             V578            B             4                      GA           MO         RM                         M                     -                  -                 -               O        O          O               AI              O      YES                 YES                     NO R501            -             4                      RL           SA         -                            -                   -                  -                 -               C        C          C               -               C      NO                  YES                     NO RHR SUCTION          V502            B             100                    GA           MO         RM                         M                     -                  -                 -               C        O          C               AI              O      YES                 NO                      NO 824          55               W       YES                 NO                                                                                                                                                                                                                                                                                                                                                   2, 18 FROM HOT LEG V503            A             12                     GA           MO         RM                         M                     -                  -                 -               C        O          C               AI              O      YES                 YES                     NO Amendment 63                                                                                                                                                                                                                                                                                                                                                                                            Page 9 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT R500 - 4 RF SA - - - - - C C C - C NO YES NO RHR SUCTION V500 B 100 GA MO RM M - - - C O C AI O YES NO NO 824 55 w YES NO 2, 18 FROM HOT LEG V501 A 12 GA MO RM M - - - C O C AI O YES YES NO V17 - 126 CK SA - - - - - C C O - C YES YES NO V23 - 26 CK SA - - - - - C C O - C YES YES NO V29 - 145 CK SA - - - - - C C O - C YES YES NO SAFETY V440 - 124 GL M M - - - - TL TL TL - TL YES YES NO INJECTION - 808 55 W YES NO 2, 17 HIGH HEAD TO V439 - 25 GL M M - - - - TL TL TL - TL YES YES NO COLD LEG V438 - 144 GL M M - - - - TL TL TL - TL YES YES NO V505 B 2 GA MO A RM - 3 10 C C O AI C YES YES NO V506 A 3 GA MO A RM - 3 10 C C O AI C YES YES NO V510 - 53 CK SA - - - - - C C C - C YES YES NO SAFETY INJECTION - LOW V511 - 55 CK SA - - - - - C C C - C YES YES NO 810 55 W YES NO 2, 17, 18 HEAD TO HOT LEGS V587 A 1 GA MO RM M - - - C C C AI C YES YES NO 19 SPARE Amendment 63 Page 10 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V84 - 82 CK SA - - - - - C C C - C YES YES NO V90 - 36 CK SA - - - - - C C C - C YES YES NO V96 - 140 CK SA - - - - - C C C - C YES YES NO SAFETY INJECTION - V431 - 81 GL M M - - - - TL TL TL - TL YES YES NO 808 55 W YES NO 2, 17, 18 HIGH HEAD TO HOT LEGS V430 - 35 GL M M - - - - TL TL TL - TL YES YES NO V429 - 139 GL M M - - - - TL TL TL - TL YES YES NO V500 A 1 GA MO RM M - - - C C C AI C YES YES NO V39 - 122 CK SA - - - - - C C C - C YES YES NO V45 - 34 CK SA - - - - - C C C - C YES YES NO SAFETY V51 - 136 CK SA - - - - - C C C - C YES YES NO INJECTION - V434 - 120 GL M M - - - - TL TL TL - TL YES NO NO 808 55 W YES NO 2, 17, 18 HIGH HEAD TO HOT LEGS V433 - 31 GL M M - - - - TL TL TL - TL YES NO NO V432 - 138 GL M M - - - - TL TL TL - TL YES NO NO V501 B 1 GA MO RM M - - - C C C AI C YES YES NO V63 - 155 CK SA - - - - - C C C - C YES YES NO V69 - 24 CK SA - - - - - C C C - C YES YES NO V75 - 153 CK SA - - - - - C C C - C YES YES NO SAFETY INJECTION - V437 - 152 GL M M - - - - TL TL TL - TL YES NO NO 808 W YES NO 2, 17, 18 HIGH HEAD TO V436 - 23 GL M M - - - - TL TL TL - TL YES NO NO COLD LEGS V435 - 80 GL M M - - - - TL TL TL - TL YES NO NO V502 A 1 GA MO RM M - - - C C C AI C YES YES NO V27 - 3 CK SA - - - - - C C O - C YES YES YES CONTAINMENT 050 56 W YES NO SPRAY V21 A 2 GA MO A RM - 11 10 C C O AI C YES YES YES Amendment 63 Page 11 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V51 - 3 CK SA - - - - - C C O - C YES YES YES CONTAINMENT 050 56 W YES NO SPRAY V43 B 2 GA MO A RM - 11 10 C C O AI C YES YES YES SERVICE WATER 047 57 W NO NO TO FAN COOLER B46 A 1 BF MO RM M - - - O O O AI O YES YES YES 1, 3 AH-3 SERVICE WATER 047 57 W NO NO TO FAN COOLER B45 A 1 BF MO RM M - - - O O O AI O YES YES NO 1, 3 AH-2 SERVICE WATER 047 57 W NO NO TO FAN COOLER B52 B 1 BF MO RM M - - - O O O AI O YES YES NO 1, 3 AH-1 SERVICE WATER 047 57 W NO NO TO FAN COOLER B51 B 1 BF MO RM M - - - O O O AI O YES YES NO 1, 3 AH-4 Amendment 63 Page 12 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT B47 A 1 BF MO RM M - - - O O O AI O YES YES NO SERVICE WATER 047 57 W NO NO FROM FAN 1, 3 COOLER AH-3 R1 - 1 RF SA - - - - - C C C - C YES YES NO B49 A 1 BF MO RM M - - - O O O AI O YES YES NO SERVICE WATER 047 57 W NO NO FROM FAN 1, 3 COOLER AH-2 R3 - 1 RF SA - - - - - C C C - C YES YES NO B48 B 1 BF MO RM M - - - O O O AI O YES YES NO SERVICE WATER 047 57 W NO NO FROM FAN 1, 3 COOLER AH-1 R2 - 1 RF SA - - - - - C C C - C YES YES NO B50 B 1 BF MO RM M - - - O O O AI O YES YES NO SERVICE WATER 047 57 W NO NO FROM FAN 1, 3 COOLER AH-4 R4 - 1 RF SA - - - - - C C C - C YES YES NO V408 B 2 GL SO A RM 1 - 5 C C CY C C NO YES YES GAS SAMPLE RETURN FROM 052 56 A NO YES 19 POST ACCIDENT SKID #2 V409 A 2 GL SO A RM 1 - 5 C C CY C C NO YES YES Amendment 63 Page 13 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT ILRT 416 56 A NO YES V2 - 3 G M M - - - - LC LC LC - O NO YES YES 5 ROTOMETER V171 - 1 CK SA - - - - - O O C - C NO YES YES COMPONENT 821 56 W NO NO COOLING WATER

                                                                                                                                      - TO RCP V170            B             2                     GA           MO         A                        RM                      2                  -                 10              O        O          C               AI              C       NO                  YES                     YES V51             -             3                     CK           SA         -                            -                   -                  -                 -               C        C          C               -               C       NO                  YES                     YES COMPONENT 821          56               W       NO                 NO                     COOLING WATER        V184            A             2                     GA           MO         A                        RM                      2                  -                 10              O        O          C               AI              C       NO                  YES                     YES FROM RCP V183            B             2                     GA           MO         A                        RM                      2                  -                 10              O        O          C               AI              C       NO                  YES                     YES V173            -             2                     CK           SA         -                            -                   -                  -                 -               O        O          C               -               C       NO                  NO                      NO COMPONENT COOLING WATER        V172            B             2                     GA           MO         A                        RM                      1                  -                 10              O        O          C               AI              C       NO                  YES                     NO TO REACTOR COOLANT DRAIN 821          57               W       NO                 NO                                          V182            B             2                     GA           MO         A                        RM                      1                  -                 10              O        C          C               AI              C       NO                  YES                     NO              1 TANK AND EXCESS LETDOWN HEAT         R5              -             42                    RL           SA         -                            -                   -                  -                 -               C        C          C               -               C       NO                  YES                     NO EXCHANGERS R6              -             60                    RL           SA         -                            -                   -                  -                 -               C        C          C               -               C       NO                  YES                     NO V50             -             3                     CK           SA         -                            -                   -                  -                 -               C        C          C               -               C       NO                  YES                     YES COMPONENT COOLING WATER        V191                          2                     GA           MO         A                        RM                      2                  -                 10              O        O          C               AI              C       NO                  YES                     YES 821          56               W       NO                 NO                       FROM RCP THERMAL BARRIERS V190                          1                     GA           MO         A                        RM                      2                  -                 10              O        O          C               AI              C       NO                  YES                     YES Amendment 63                                                                                                                                                                                                                                                                                                                                                                                          Page 14 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V525 8 CK SA - - - - - O C C - C NO YES YES DEMIN WATER TO 801 56 W NO YES 7 PRT D525 4 DA AO A RM 1 - 60 O C C C C NO YES YES V15 - 2 CK SA - - - - - C C C - C NO YES YES 300 56 A NO YES SERVICE AIR 5 V14 - 8 GL M M - - - - LC LC LC - LC NO YES YES D653 - 6 DA M M - - - - O O O - O NO NO NO D654 - 4 DA M M - - - - O O O - O NO NO NO RCDT PUMP 813 56 W NO YES D651 - 7 DA M M - - - - LC LC LC - LC NO YES YES DISCHARGE L600 A 6 GL AO A RM 1 - 10 O O O C C NO YES YES D650 B 4 DA AO A RM 1 - 10 O O O C C NO YES YES 43 SPARE D164 - 1 DA M M - - - - LC LC LC - LC NO YES YES REFUELING 061 56 A NO YES CAVITY CLEAN- 5, 14, 21 UP D165 - 2 DA M M - - - - LC LC LC - LC NO YES YES D25 - 2 DA M M - - - - LC LC LC - LC NO YES YES REFUELING 061 56 A NO YES CAVITY CLEAN- 5, 14, 21 UP D26 - 2 DA M M - - - - LC LC LC - LC NO YES YES 46 SPARE Amendment 63 Page 15 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT CONTAINMENT 810 56 W YES NO SUMP TO RHR V571 A 35 GA MO A RM - 3, 4 20 C C O AI C YES YES NO 15 PUMP CONTAINMENT 810 56 W YES NO SUMP TO RHR V570 B 35 GA MO A RM - 3, 4 20 C C O AI C YES YES NO 15 PUMP CONTAINMENT 050 56 W YES NO SUMP TO CT V6 A 32 GA MO A RM - 4 102 C C O AI C YES YES NO 15 PUMP CONTAINMENT 050 56 W YES NO SUMP TO CT V7 B 32 GA MO A RM - 4 102 C C O AI C YES YES NO 15 PUMP Amendment 63 Page 16 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT P6 B 55 GL AO A RM 3 - 60 0 0 C C C NO NO NO V3 - 55 GL M M - - - - C C C - C NO NO NO STEAM 051 57 W YES NO GENERATOR A 1 BLOWDOWN V11 A 10 GL AO A RM 3 - 60 0 0 C C C NO YES NO V183 - 10 P7 B 26 GL AO A RM 3 - 60 O O C C C NO NO NO STEAM V15 A 9 GL AO A RM 3 - 60 O O C C C NO YES NO 051 57 W YES NO GENERATOR B 1 BLOWDOWN V184 - 9 GL M M - - - - C C C - C NO YES NO P8 B 51 GL AO A RM 3 - 60 O O C C C NO NO NO V9 - 52 GL M M - - - - C C C - C NO NO NO V19 A 10 GL AO A RM 3 - 60 O O C C C NO YES NO STEAM 051 57 W YES NO GENERATOR C 1 BLOWDOWN V185 - 10 GL M M - - - - C C C - C NO YES NO Amendment 63 Page 17 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V89 - 56 GL M M - - - - O O O - O NO NO NO V117 - 54 GL M M - - - - O O O - O NO NO NO STEAM V90 B 54 GA AO A RM 3, 6, 7* - 60 C C C C C NO NO NO 051 57 W YES NO GENERATOR A 1, 19 SAMPLE V91 B 53 GA AO A RM 3, 6, 7* - 60 O O C C C NO NO NO 3, 6, 7* V120 A 1 GL SO A RM - 60 O O C C C NO YES NO

                                                                                                                                                                                                                                                                                   *15,16 V84             -             34                    GL           M          M                            -                    -                  -                 -               O        O          O               -               O       NO                  NO                      NO V118            -             34                    GL           M          M                            -                    -                  -                 -               O        O          O               -               O       NO                  NO                      NO STEAM             V85             B             33                    GA           AO         A                        RM                    3, 6, 7*              -                 60              C        C          C               C               C       NO                  NO                      NO 051          57               W       YES                 NO                    GENERATOR B                                                                                                                                                                                                                                                                                                                 1, 19 SAMPLE            V86             B             33                    GA           AO         A                        RM                    3, 6, 7*              -                 60              O        O          C               C               C       NO                  NO                      NO 3, 6, 7*

V121 A 1 GL SO A RM - 60 O O C C C NO YES NO

                                                                                                                                                                                                                                                                                   *15,16 V79             -             64                    GL           M          M                            -                    -                  -                 -               O        O          O               -               O       NO                  NO                      NO V119            -             64                    GL           M          M                            -                    -                  -                 -               O        O          O               -               O       NO                  NO                      NO V80             B             62                    GA           AO         A                        RM                    3, 6, 7*              -                 60              C        C          C               C               C       NO                  NO                      NO V81             B             62                    GA           AO         A                        RM                    3, 6, 7*              -                 60              O        O          C               C               C       NO                  NO                      NO STEAM 051          57               W       YES                 NO                    GENERATOR C                                                                                                                                                                                                                                                                                                                 1, 19 SAMPLE 3, 6, 7*

V122 A 1 GL SO A RM - 60 O O C C C NO YES NO

                                                                                                                                                                                                                                                                                   *15,16 Amendment 63                                                                                                                                                                                                                                                                                                                                                                                           Page 18 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT B1 A 2 BF AO A RM 8, 10 - 3.5 CY C C C C NO YES YES B3 A 3 BF AO A RM 8 - 15 LC O LC LC LC NO YES YES CONTAINMENT 517 56 A NO YES ATMOSPHERE B4 B 3 BF AO A RM 8 - 15 LC O LC LC LC NO YES YES 11 PURGE MAKE-UP B2 B 3 BF AO A RM 8, 10 - 3.5 CY C C C C NO YES YES B7 A 3 BF AO A RM 8 - 15 LC O LC LC LC NO YES YES B5 A 3 BF AO A RM 8 - 3.5 O C C C C NO YES YES CONTAINMENT 517 56 A NO YES ATMOSPHERE B8 B 3 BF AO A RM 8 - 15 LC O LC LC LC NO YES YES 11 PURGE EXHAUST B6 B 3 BF AO A RM 8 - 3.5 O C C C C NO YES YES V1 - 3 CK SA - - - - - C C C - C NO YES YES CONTAINMENT 517 56 A NO YES VACUUM RELIEF B2 B 3 BF AO A RM 8 9 5 C C C C C NO YES YES 60 SPARE V1 - 1 CK SA - - - - - C C C - C NO YES YES H2 PURGE MAKE-517 56 A NO YES UP B6 - 4 BF M M - - - - LC LC LC - LC NO YES YES Amendment 63 Page 19 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT 416 56 A NO YES ILRT V4 - 3 G M M - - - - LC LC LC - O NO YES YES 5 B5 A 2 BF AO RM - - - - LC LC LC LC LC NO YES YES H2 PURGE 517 56 A NO YES 5 EXHAUST B4 - 5 BF M M - - - - LC LC LC - LC NO YES YES 64 SPARE TYPE FUEL TRANSFER 65 - - - - - - - - - - - - - - - - - B TUBE TEST OPENABLE DURING TYPE OUTAGES REFUELING

                                                                                                                                                        -              -             -                     -            -          -                            -                   -                  -                 -               -        -          -               -               -          -                     -                    B               FOR ACCESS SLEEVE TEST            ACCESS REF. DWG:

2165-G-065 67 SPARE 68 SPARE Amendment 63 Page 20 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT CONTAINMENT PRESSURE TYPE

                                                   -           57               CF      NO                 NO                      SENSING           -              -             -                     -            -          -                            -                   -                  -                 -               -        -          -               -               -          -                     -                    A TEST RPS-IV CONTAINMENT PRESSURE                                                                                                                                                                                                                                                                                                    TYPE
                                                   -           57               CF      NO                 NO                      SENSING           -              -             -                     -            -          -                            -                   -                  -                 -               -        -          -               -               -          -                     -                    A TEST RPS-II CONTAINMENT PRESSURE                                                                                                                                                                                                                                                                                                    TYPE
                                                   -           57               CF      NO                 NO                      SENSING           -              -             -                     -            -          -                            -                   -                  -                 -               -        -          -               -               -          -                     -                    A TEST RPS-I Amendment 63                                                                                                                                                                                                                                                                                                                                                                                      Page 21 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT CONTAINMENT PRESSURE TYPE

                                                       -           57               CF      NO                  NO                       SENSING              -              -             -                     -            -          -                            -                   -                  -                 -               -        -          -               -               -          -                     -                    A TEST RPS-III 73 (SEE PAGE NUMBER 19 OF THIS TABLE)

V36 A 1 GA MO A RM 1 - 60 O O C AI C NO YES YES CONTAINMENT 185 56 W NO YES SUMP PUMP DISCHARGE V77 B 5 GA MO A RM 1 - 60 O O C AI C NO YES YES 75 SPARE V150 - 3 CK SA - - - - - C C C - C NO YES YES ACCUMULATOR 809 56 W YES YES 8 FILL FROM RWST V554 B 2 GL AO A RM 1 - 10 C C C C C NO YES YES V555 A 2 GL AO A RM 1 - 10 C C C C C NO YES YES ACCUMULATOR 809 56 W YES YES 9 TO RWST V550 B 1 GL AO A RM 1 - 10 C C C C C NO YES YES Amendment 63 Page 22 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V188 - 3 CK SA - - - - - CY C C - C NO YES YES NITROGEN TO 809 56 G YES YES 10 ACCUMULATORS V530 B 1 GL AO A RM 1 - 10 CY C C C C NO YES YES D528 A 4 DA AO A RM 1 - 10 C C C C C NO YES YES PRESSURIZER 801 56 G NO YES RELIEF TANK CONNECTION D529 B 3 DA AO A RM 1 - 10 C C C C C NO YES YES D590 A 3 DA AO A RM 1 - 10 O C C C C NO YES YES 813 56 G NO YES RCDT H2 SUPPLY D291 B 3 DA AO A RM 1 - 10 O C C C C NO YES YES V111 B 19 GL SO A RM 1 - 60 O C C C C NO YES YES REACTOR 052 55 W YES YES COOLANT 19 SAMPLE V23 A 3 GL SO A RM 1 - 60 O C C C C NO YES YES V11 B 4 GL SO A RM 1 - 60 C C C C C NO YES YES PRESSURIZER 052 55 W YES YES LIQUID SAMPLE V12 A 3 GL SO A RM 1 - 60 C C C C C NO YES YES Amendment 63 Page 23 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V1 B 2 GL SO A RM 1 - 60 C C C C C NO YES YES PRESSURIZER 052 55 S YES YES STEAM SAMPLE V2 A 3 GL SO A RM 1 - 60 C C C C C NO YES YES V113 B 9 GL SO A RM 1 - 60 O C C C C NO YES YES V114 B 3 GL SO A RM 1 - 60 O C C C C NO YES YES ACCUMULATOR 052 55 W YES YES SAMPLE V115 B 3 GL SO A RM 1 - 60 O C C C C NO YES YES V116 A 3 GL SO A RM 1 - 60 O C C C C NO YES YES V48 - 3 CK SA - - - - - C O C - C NO YES YES FIRE WATER 388 56 W NO YES STANDPIPE SUPPLY V44 - 1 GA M M - - - - LC O LC C LC NO YES YES V33 - 3 CK SA - - - - - O O C C C NO YES YES INSTRUMENT AIR 301 56 A NO YES SUPPLY V192 A 2 GL AO A RM 1 - 60 O O C C C NO YES YES 81 SPARE 82 SPARE 83 (SEE PAGE 6.2.4-31C) Amendment 63 Page 24 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT 84 SPARE 85 SPARE 86 (SEE PAGE 6.2.4-31B) 87 SPARE V406 B 3 GL SO A RM 1 - 5 C C CY C C NO YES YES LIQUID SAMPLE RETURN FROM 052 56 W NO YES 19 POST ACCIDENT SKID #1 V407 A 2 GL SO A RM 1 - 5 C C CY C C NO YES YES 89 SPARE V121 - 1 CK SA - - - - - C C C - C NO YES YES DEMIN WATER 299 56 W NO YES 5 SUPPLY V120 - 5 GA M M - - - - LC LC LC - LC NO YES YES B89 A 2 BF AO A RM 1 - 60 O C C C C NO YES YES SERVICE WATER 047 56 W NO YES FROM NNS FAN B90 B 1 BF AO A RM 1 - 60 O C C C C NO YES YES 11 COILS R18 - - RL SA - - - - - C C C C C NO YES YES Amendment 63 Page 25 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V142 - 2 CK SA - - - - - O C C C C NO YES YES SERVICE WATER 047 56 W NO YES TO NNS FAN COILS B88 AB 1 BF AO A RM 1 - 60 O C C C C NO YES YES 93 SPARE 94A AND 94B (SEE PAGE 18 OF THIS TABLE) CONTAINMENT A - - GL M M - - - - LO LO LO LO LO YES NO B430 OUTSIDE TYPE 56 A NO NO DIFFERENTIAL A 31.190 PRESSURE TEST SENSING B - - XC SA SA - - - - SA SA SA SA SA YES YES 95A AND 95B (SEE PAGE 18 OF THIS TABLE) 95C SPARE 416 56 A NO YES ILRT V1 - 8 G M M - - - - LC LC LC - O NO YES YES 5 97 SPARE V2 - 3 CK SA - - - - - C C C - C NO YES YES CONTAINMENT 517 56 A NO YES VACUUM RELIEF B2 B 3 BF AO A RM 8 9 5 C C C C C NO YES YES 99 SPARE Amendment 63 Page 26 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT 100 SPARE 101 SPARE OPENABLE TYPE DURING REFUELING B OUTAGE ACCESS SLEEVE TEST FOR ACCESS 103 (SEE PAGE 20 OF THIS TABLE) 104 SPARE V46 - 2 CK SA - - - - - O O C - C NO YES YES FIRE WATER 388 56 W NO YES SPRINKLER SUPPLY B1 A 2 BF AO A RM 1 - 60 O O C C C NO YES YES 106 107 (SEE PAGE 21 OF THIS TABLE) V162 - 5 GL M M - - - - LC LC LC LC LC NO YES NO V163 - 5 GL M M - - - - LC LC LC LC LC NO YES NO V153 - 155 CK SA - - - - - O O O - C YES NO NO AUXILIARY 044 57 W YES NO 1, 20 FEEDWATER V10 B 21 GA MO A RM 14 16 24 O O O AI C YES YES NO V116 A 22 GA MO A RM 14 16 24 O O O AI C YES YES NO V189 - 6 GL M M - - - - LC O LC LC LC NO YES NO Amendment 63 Page 27 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V164 - 5 GL M M - - - - LC LC LC LC LC NO YES NO V165 - 5 GL M M - - - - LC LC LC LC LC NO YES NO V154 - 70 CK SA - - - - - O O O - C YES NO NO AUXILIARY 044 57 W YES NO 1, 20 FEEDWATER V19 B 21 GA MO A RM 14 16 24 O O O AI C YES YES NO V117 A 22 GA MO A RM 14 16 24 0 0 0 AI C YES YES NO V190 - 6 GL M M - - - - LC O LC LC LC NO YES NO V166 - 5 GL M M - - - - LC LC LC LC LC NO YES NO V167 - 5 GL M M - - - - LC LC LC LC LC NO YES NO V155 - 155 CK SA - - - - - O O O - C YES NO NO V23 B 21 GA MO A RM 14 16 24 O O O AI C YES YES NO AUXILIARY 044 57 W YES NO 1, 20 FEEDWATER V118 A 22 GA MO A RM 14 16 24 O O O AI C YES YES NO V191 - 6 GL M M - - - - LC O LC LC LC NO YES NO Amendment 63 Page 28 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT A - 5 GL M M - - - - LO LO LO LO LO YES NO B430 CONTAINMENT TYPE 56 A NO NO VACUUM RELIEF A 31.18 - SENSING B TEST B - - XC SA SA - - - - SA SA SA SA SA YES YES A - 5 GL M M - - - - LO LO LO LO LO YES NO B430 CONTAINMENT TYPE 56 A NO NO VACUUM RELIEF A 31.18 - SENSING B TEST B - - XC SA SA - - - - SA SA SA SA SA YES YES A - 5 GL M M - - - - LO LO LO LO LO YES NO B430 CONTAINMENT TYPE 56 A NO NO VACUUM RELIEF A 31.17 - SENSING A TEST B - - XC SA SA - - - - SA SA SA SA SA YES YES Amendment 63 Page 29 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT A - 5 GL M M - - - - LO LO LO LO LO YES NO B430 CONTAINMENT TYPE 56 A NO NO VACUUM RELIEF A 31.17 - SENSING B TEST B - - XC SA SA - - - - SA SA SA SA SA YES YES V300 A 2 GL SO A RM 1 - 60 O C C C C YES YES YES HYDROGEN 105 56 A NO 19 ANALYZER V348 A 1 GL SO A RM 1 - 60 O C C C C YES YES YES V301 A 2 GL SO A RM 1 - 60 O C C C C YES YES YES HYDROGEN 105 56 A NO 19 ANALYZER V349 A 1 GL SO A RM 1 - 60 O C C C C YES YES YES Amendment 63 Page 30 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT V308 B 1 GL SO A RM 1 - 60 C C C C C YES YES YES HYDROGEN 105 56 A NO 19 ANALYZER V314 B 1 GL SO A RM 1 - 60 C C C C C YES YES YES V309 B 2 GL SO A RM 1 - 60 C C C C C YES YES YES HYDROGEN 105 56 A NO 19 ANALYZER V315 B 1 GL SO A RM 1 - 60 C C C C C YES YES YES FILL VALVE IS SEALED TYPE 54, AFTER 844 W YES NO RVLIS - B - - - - - - - - - - - - - NO NO A 55 USE TEST 16 Amendment 63 Page 31 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT FILL VALVE IS SEALED TYPE 54, AFTER 844 W YES NO RVLIS - B - - - - - - - - - - - - - NO NO A 55 USE TEST 16 FILL VALVE IS SEALED TYPE 54, AFTER 844 W YES NO RVLIS - B - - - - - - - - - - - - - NO NO A 55 USE TEST 16 V-448 A GL SO A RM 1 - 60 O O C C C NO YES YES V-449 B GL SO A RM 1 - 60 O O C C C NO YES YES RCPB LEAK DETECTION 105 56 A NO RADIATION MONITOR V-550 A GL SO A RM 1 - 60 O O C C C NO YES YES V-451 B GL SO A RM 1 - 60 O O C C C NO YES YES Amendment 63 Page 32 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-1 CONTAINMENT ISOLATION SYSTEM DATA PENETRATION DATA VALVE DATA Valve Position PRIMARY ACTUATION MODE SECONDARY ACTUATION CONTAINMENT ISOLATION BYPASS LEAKAGE PATH ENGINEERED SAFETY HIGH ENERGY LINE LENGTH OF PIPE, FT. ISOLATION SIGNAL ACCIDENT SIGNAL GENERAL DESIGN VALVE NUMBER RESPONSE TIME POST-ACCIDENT POWER FAILURE PENETRATION DETAIL NOTE FD NUMBER SYSTEM TITLE POWER TRAIN VALVE TYPE ACTUATOR MODE SHUTDOWN FEATURE VALVE TYPE C TEST CRITERION FLUID NORMAL ILRT FILL VALVE IS SEALED TYPE 54, AFTER 844 W YES NO RVLIS - A - - - - - - - - - - - - - NO NO A 55 USE TEST 16 FILL VALVE IS SEALED TYPE 54, AFTER 844 W YES NO RVLIS - A - - - - - - - - - - - - - NO NO A 55 USE TEST 16 FILL VALVE IS SEALED TYPE 54, AFTER 844 W YES NO RVLIS - A - - - - - - - - - - - - - NO NO A 55 USE TEST 16 Amendment 63 Page 33 of 33

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-2 CONTAINMENT ISOLATION VALVE POSITION FOLLOWING AN ACCIDENT Penetration Essential or No. Penetration Name Non Essential Valve Position M1 MS-SG A E CLOSED M2 MS-SG B E CLOSED M3 MS-SG C E CLOSED M4 FEEDWATER SG A NE CLOSED M5 FEEDWATER SG B NE CLOSED M6 FEEDWATER SG C NE CLOSED M7 NORMAL LETDOWN NE CLOSED M8 CVCS NORMAL CHARGING NE CLOSED M9 SEAL INJECTION RC PUMP A E OPENED M10 SEAL INJECTION RC PUMP B E OPENED M11 SEAL INJECTION RC PUMP C E OPENED M12 RC PUMP SEAL INJECTION AND EXCESS LETDOWN EXCH NE CLOSED OUTLET M13 LOW HEAD SI TO COLD LEG E OPENED M14 LOW HEAD SI TO COLD LEG E OPENED M15 RHR LOOP 1 (NORMAL OPERATION MODE) NE CLOSED M16 RHR LOOP 2 (NORMAL OPERATION MODE) E CLOSED M17 HIGH HEAD SI TO COLD LEG E OPENED M18 LOW HEAD SI TO HOT LEG E CLOSED(5) M19 SPARE E N/A M20 HIGH HEAD SI TO HOT LEG E CLOSED(5) M21 HIGH HEAD SI TO HOT LEG E CLOSED(5) M22 HIGH HEAD SI TO COLD LEG E CLOSED(5) M23 CONTAINMENT SPRAY E OPENED** M24 CONTAINMENT SPRAY E OPENED** M25 CONTAINMENT FAN COOLER AH3-SW IN E OPENED M26 CONTAINMENT FAN COOLER AH2-SW IN E OPENED M27 CONTAINMENT FAN COOLER AH1-SW IN E OPENED M28 CONTAINMENT FAN COOLER AH4-SW IN E OPENED M29 CONTAINMENT FAN COOLER AH3-SW OUT E OPENED M30 CONTAINMENT FAN COOLER AH2-SW OUT E OPENED M31 CONTAINMENT FAN COOLER AH1-SW OUT E OPENED M32 CONTAINMENT FAN COOLER AH4-SW OUT E OPENED (3) M33 POST ACCIDENT GAS SAMPLE RETURN NE CLOSED M34 ILRT ROTOMETER NE CLOSED M35 COMPONENT COOLING WATER - RC PUMP NE CLOSED M36 COMPONENT COOLING WATER - RC PUMP NE CLOSED Amendment 61 Page 1 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-2 CONTAINMENT ISOLATION VALVE POSITION FOLLOWING AN ACCIDENT Penetration Essential or No. Penetration Name Non Essential Valve Position M37 COMP COOLING WATER EXC LETDN & RCDT NE CLOSED M38 COMP COOLING WATER - EXC LETDN & RCDT NE CLOSED M39 COMP COOLING WATER - RC PUMP THERM BARR NE CLOSED M40 MAKEUP WATER TO PRESSURIZER NE CLOSED M41 SERVICE AIR SUPPLY NE CLOSED M42 RCDT PUMP DISCHARGE NE CLOSED M43 SPARE N/A M44 S F PURIFICATION PUMP TO REFUELING CAVITY NE CLOSED M45 REFUELING CAVITY WATER CLEANUP- OUT NE CLOSED M46 SPARE N/A (4) M47 SUMP RECIRC (RHR A) E OPENED (4) M48 SUMP RECIRC (RHR B) E OPENED (4) M49 SUMP RECIRC (CONT SPRAY A) E OPENED (4) M50 SUMP RECIRC (CONT SPRAY B) E OPENED M51 SG A BLOWDOWN NE CLOSED M52 SG B BLOWDOWN NE CLOSED M53 SG C BLOWDOWN NE CLOSED M54 SG A BLOWDOWN SAMPLE NE CLOSED M55 SG B BLOWDOWN SAMPLE NE CLOSED M56 SG C BLOWDOWN SAMPLE NE CLOSED M57 CONTAINMENT PURGE MAKEUP NE CLOSED M58 CONTAINMENT PURGE EXHAUST NE CLOSED M59 VACUUM RELIEF A NE CLOSED M60 SPARE N/A M61 H2 PURGE MAKE-UP NE CLOSED M62 CONTMT LEAK RATE TEST PRESS INDIC. NE CLOSED M63 H2 PURGE EXHAUST NE CLOSED M64 SPARE N/A M65 FUEL TRANSFER TUBE NE CLOSED M66 REFUELING ACCESS SLEEVE N/A M67-M68 SPARE N/A M69 CONTAINMENT PRESSURE SENSING A E N/A M70 CONTAINMENT PRESSURE SENSING B E N/A M71 CONTAINMENT PRESSURE SENSING C E N/A M72 CONTAINMENT PRESSURE SENSING D E N/A M73A CONTAINMENT HYDROGEN ANALYZER NE CLOSED* M73B CONTAINMENT HYDROGEN ANALYZER NE CLOSED* M74 CONTAINMENT SUMP PUMP DISCHARGE NE CLOSED M75 SPARE N/A M76A ACCUMULATOR FILL NE CLOSED M76B ACCUMULATOR TO RWST NE CLOSED M77A N2 TO ACCUMULATOR NE CLOSED M77B PRT N2 & CDT CONNECTION NE CLOSED M77C RCDT H2 SUPPLY & GAS SAMPLE NE CLOSED Amendment 61 Page 2 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-2 CONTAINMENT ISOLATION VALVE POSITION FOLLOWING AN ACCIDENT Penetration Essential or No. Penetration Name Non Essential Valve Position (3) M78A RC LOOP 2 & 3 SAMPLE NE CLOSED M78B PRESS. LIQUID SAMPLE NE CLOSED M78C PRESS. STEAM SAMPLE NE CLOSED M78D ACCUMULATOR SAMPLE NE CLOSED M79 FIRE PROTECTION-STANDPIPE SUPPLY NE CLOSED M80 INSTR AIR SUPPLY NE CLOSED M81-M85 SPARES NE N/A M86A CONTAINMENT HYDROGEN ANALYZER NE CLOSED* M86B CONTAINMENT HYDROGEN ANALYZER NE CLOSED* M87 SPARE N/A (3) M88 POST ACCIDENT LIQUID SAMPLE RETURN NE CLOSED M89 SPARE N/A M90 DEMIN. WATER TO FUEL TRANSFER SYSTEM CONTR NE CLOSED PANEL & REFUELING CAVITY DECON M91 CONTAINMENT FAN COIL UNITS SW - OUT NE CLOSED M92 CONTAINMENT FAN COIL UNITS SW - IN NE CLOSED M93 SPARE N/A M94A,B CONTAINMENT VACUUM RELIEF SENSING LINES E OPEN M95A,B CONTAINMENT VACUUM RELIEF SENSING LINES E OPEN M94C CONTAINMENT OUTSIDE DIFFERENTIAL PRESSURE E OPEN SENSING M95C SPARE N/A M96 CONTAINMENT LEAK RATE TEST SUPPLY & EXHAUST NE CLOSED M97 SPARE N/A M98 VACUUM RELIEF B NE CLOSED M99-M101 SPARES N/A M103A RVLIS E N/A M103B RVLIS E N/A M103C RVLIS E N/A M104 SPARE N/A M105 FIRE PROTECTION SPRINKLER SYS HDR NE CLOSED M106 SPARE N/A M107A RVLIS E N/A M107B RVLIS E N/A M107C RVLIS E N/A M108 AUX FEEDWATER TO SG A E OPENED+ M109 AUX FEEDWATER TO SG B E OPENED+ M110 AUX FEEDWATER TO SG C E OPENED+ M102 REFUELING ACCESS SLEEVE N/A

ISOLATION VALVE CLOSED ON PHASE A CONTAINMENT ISOLATION SIGNAL. REOPEN MANUALLY FOR POST ACCIDENT H2 SAMPLING.
- NORMALLY CLOSED. OPEN ON CONTAINMENT SPRAY ACTUATION SIGNAL.
  - A "P" SIGNAL IS DEFINED AS A CONTAINMENT PHASE B SIGNAL.

Amendment 61 Page 3 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.4-2 (Continued)

+ WILL BE CLOSED TO ISOLATE FAULTED STEAM GENERATOR (ie., LOSS OF SG PRESSURE BOUNDARY)

1) ESSENTIAL: LINES REQUIRED TO MITIGATE AN ACCIDENT, OR WHICH, IF UNAVAILABLE COULD INCREASE THE MAGNITUDE OF THE EVENT.

2 NON-ESSENTIAL: LINES WHICH ARE NOT REQUIRED TO MITIGATE AN ACCIDENT, AND WHICH IF REQUIRED AT ALL WOULD BE REQUIRED FOR LONG TERM RECOVERY ONLY; i.e., DAYS OR WEEKS FOLLOWING AN ACCIDENT.

3) VALVES ARE OPENED INTERMITTENTLY FOR POST-ACCIDENT SAMPLING.

4) INITIALLY CLOSED, OPEN ON LOW WATER LEVEL IN RWST.

5) OPENED BY OPERATOR ACTION FOR LONG-TERM COOLING.

Amendment 61 Page 4 of 4

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5 Deleted by Amendment 62 Amendment 62 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5-2 CONTAINMENT HYDROGEN PURGE SYSTEM COMPONENTS NON NUCLEAR SAFETY UNITS

1. Exhaust Fans Quantity 1 (one)

Type Centrifugal type, direct-driven Material ASTM-A36, carbon steel Actual air flow inlet, per fan, cfm 500 nominal Static pressure, in wg 16.64 Code Air Moving and Conditioning Association Inc. (AMCA). Anti-Friction Bearing Manufacturer's Association (AFBMA)

2. Motors Quantity One Type 5 hp, 460 volt, 60 Hz 3 phase horizontal induction type Insulation Class H, Type RH Enclosure & Ventilation Dripproof/Guarded Code NEMA

3. Medium Efficiency Filter Quantity One bank per filter train Air Flow, cfm 500 Face Velocity, fpm 125 Material Glass Fiber

4. HEPA Filters Quantity (2) Two banks per train Air Flow, cfm (total) 500 Cell (Unit) Size 24 in. high, 24 in. wide, 11 1/2 in. deep Cell Arrangement (Units) 1 Unit Max. Resistance Clean, in wg 1.0 Max. Resistance Loaded, in wg 2.0 Efficiency 99.97 percent when tested with 0.3 micron DOP Material Glass or glass asbestos paper separated by aluminum inserts supported on cadmium plated steel frame

5. Charcoal Adsorbers Quantity One bank per filter train Air Flow, cfm (total) 500 nominal Bed depth, inches 2 inch total Max. Air Resistance, in wg 1.1 Efficiency New activated carbon 99.5 percent of elemental iodine when tested at 25°C and 95 percent relative humidity. 95.0 percent of methyl iodide when tested at 25°C and 95 percent relative humidity Lab test for representative samples of used carbon (18 month test requirement) 90.0 percent of methyl iodide when tested at 25°C and 70 percent relative humidity Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Loading Capacity 2.5 mg of iodine per gram of charcoal elemental and organic

5. Charcoal Adsorbers (Contd)

Material Adsorber, activated coconut shell charcoal enclosure, stainless steel Type 316 ASTM gaskets, Neoprene ASTM D1056, ASTM D1056, Grade SCE-43 frame, Steel ASTM-A36 Type Deep bed

6. Demisters Quantity, per fan 1 bank Air Flow, cfm per bank 500 Max. air resistance, clean, in. wg 1.0 Max. air resistance, loaded in. wg 2.0 Efficiency 99 percent when exposed to entrained water particles of 1 to 5 micron size

7. Electric Heating Coil Quantity per fan 1 bank Capacity 5 kW Code Underwriter Laboratory (UL) National Electrical Manufacturing Association (NEMA)

Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5-3 PARAMETERS FOR ANALYSIS OF HYDROGEN GENERATION AND CONTROL Hydrogen Dissolved in Reactor Coolant 934 scf Release Rate for Dissolved Hydrogen Instantaneous Amount of Zircaloy in core 37,483 lbs Fraction of Zirconium Assumed to Oxidize for Purposes of Hydrogen Generation Analysis 5% Release Rate from Zirconium-Water Reaction Instantaneous Fission Product Distribution Model 50% of the halogens and 1% of the solids present in the core are mixed with the coolant water All noble gases are released to the Containment 99% of other fission products remain in fuel rods Fraction Fission Product Radiation Energy Absorbed by the Coolant (a) Beta Percent of beta energy absorbed by coolant: 0% (b) Gamma Percent of gamma energy absorbed by coolant: 10% Hydrogen Yield Rate G (H2) 0.5 molecule per 100 ev Oxygen Yield Rate G(O2) 0.25 molecule per 100 ev Reactor Thermal Power, mwt 2958 Inventory of Corrodible Metal Table 6.2.5-4 Assumed Hydrogen Generation Rate Due to Aluminum Corrosion Figure 6.2.5-3 Assumed Hydrogen Generation Rate Due To Zinc Corrosion Figure 6.2.5-4 Containment Net Free Volume, ft.3 2.266 x 106 Initial Bulk Average Containment Temperature, F 135° Initial Containment Pressure, psia 16.3 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5-3a POST-LOCA CONTAINMENT TEMPERATURES Time Interval, sec Temperature (°F) 0-5 232 5 - 10 254 10 - 175 265 175 - 3,600 260 3,600 - 6,000 247 6,000 - 10,000 231 10,000 - 18,000 215 18,000 - 50,000 194 50,000 - 100,000 177 100,000 - 500,000 165 500,000 - 1,000,000 152 1,000,000 - 10,000,000 136 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5-4 ALUMINUM INVENTORY IN CONTAINMENT Surface Area* Mass Item (ft.2) (lbm)

1. Flux Mapping Drive System 82.5 171

2. Source, Intermediate and Power Range Detectors 91.3 244

3. Control Rod Drive Mechanism Connection 71.5 191

4. Rod Position Indicators 86.9 139

5. Miscellaneous Valves 94.6 230

6. Contingency 82.5** 200**

7. Containment building circular bridge crane 41.0 71.5

8. Jib Crane - Removed 0 0

9. Hoist 28.6 50

10. Elevator 28.6 10

11. Manual Pull Stations 0.5 8.1

12. Fire Detectors 4.8 1.7

13. Additional Inventory 29.8 63.5

10% of uncertainty is included.
- Original design value. Available contingency is tracked administratively.

Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5-5 GALVANIZED ZINC INVENTORY IN CONTAINMENT** Group* Surface Area (ft.2) Thickness (Mils) A. Ductwork Conduits, Cable Trays, Pull Boxes and Junction Boxes 1(1) 41188.4 1.5 2(2) 23610.7 1.5 3(2) 10092.4 4 4(2) 4457.4 2 5(2) 118.5 5 B. Grating and Stair Treads(3) 56668.5 1.7 C. Inorganic Zinc on the surface of neutron streaming shield(1) 127.6 5.0 D. Zinc on the surface of damper actuators(1)

55. 5.0 E. Tube track in RCB(1) 11458.4 5.0 F. Additional Inventory 1,754.5 Notes:

(*) Groups were determined by thickness and uncertainty (1) Includes 10% uncertainty (2) Includes 15% uncertainty (3) Includes 5% uncertainty (**) Additional zinc inventory may be evaluated and tracked as an equivalent amount of aluminum. (See Table 6.2.5-4) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5-6 ZINC-BASE PAINT INVENTORY IN CONTAINMENT Surface Area Item (ft.2)

1. Reactor Coolant Drain Tank Pumps (2) 10.46

2. Reactor Coolant Drain Tank Heat Exchanger 56.79

3. Integrated Head 1083.85

4. Regenerate Heat Exchanger 127.16

6. Fuel Transfer System Control Panel 57.58

7. Steam Generator Upper Section Note 1

8. Steam Generator Lower Section Note 1

9. Pressurizer 1138.04

10. Reactor Vessel 2217.73

11. Other NSSS Equipment 7739.0 TOTAL 12727.49 TOTAL with 20% uncertainty: 15273.0 Note 1 - No coating applied to Delta-75 Steam Generators. Containment Hydrogen analysis based on previous coating area of 7988.67ft2 for Steam Generators and a TOTAL (with 20%

uncertainty) of 24859.39 Amendment 62 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.2.5-7 FAILURE MODE AND EFFECTS ANALYSIS HYDROGEN MONITORING SYSTEM Method of Component Failure Mode Effects on System Detection Monitor Comments Sample line Break or Plug Loss of sample flow Low flow alarm MCRI* Redundant hydrogen analyzer available Sample pump Fails Loss of sample Low flow alarm MCRI Redundant hydrogen analyzer available Vacuum pump sample dilution Fails No backup grab sample available Operator can -- Use redundant panel distinguish from analyzer for a sample pump backup sample Recorder Hydrogen concentration Anomalous compared to the MCRI Redundant indicated higher or lower than result of the grab hydrogen analyzer actual sample available Power supply Failure or loss of power No power to analyzer; Sample and No sample flow isolation valves fail close Low flow alarm MCRI Redundant analyzer powered from redundant power bus

Main Control Room Indication Amendment 62 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks*

1. Motor Fails to Provides isolation of Failure reduces redundancy of Valve open/close Valve is electrically operated close on fluid discharge from the providing tank discharge isolation. position indication on interlocked with isolation gate valve demand. VCT to the suction of Negligible effect on system Valve close position valve 1-LCV-115B (l-LCV-1-LCV- HIISI/CHG pumps. operation. Alternate isolation valve l- monitor light and 115D) and the 115C (1- LCV-l15E (1-LCV-l15C) provides alarm for group instrumentation that monitors LCV-115E backup tank discharge isolation. monitoring of fluid level of the VCT. Valve analgous) components at MCB. closes upon receipt of an SIAS or upon receipt of a VCT "low" water level signal providing that isolation valve 1-LCV-115B (l-LCV-115D) is at full open position.

2. Motor Fails to Provides isolation of Failure reduces redundancy of Same methods of Valve is electrically operated Open on fluid discharge from the providing fluid flow from RWST to detection as that interlocked with the gate valve demand. RWST to the suction of suction of HHSI/CHG pumps. stated for item #1 instrumentation that monitors 1-LCV-115B HHSI/CHG pumps and Negligible effect on system except open position fluid level of the VCT. Valve (1-LCV- an electrical interlock to operation. Alternate isolation valve monitoring of opens upon receipt of an 115D the closing of isolation l/LCV-115D (1-LCV-ll5B) opens to components at MCB. SIAS or upon receipt of a analogous) valve 1-LCV-115C (l- provide backup flow path to suction VCT "low" water level signal LCV-115E). of HHSI/CHG pumps. . (Except when the control switches are in the pull-to-During the recirculation phase, lock position) activation of pull-to-lock switches will maintain RWST isolation valves in the shut position.

Amendment 61 Page 1 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks*

3. Centrifugal Fails to Provides fluid flow of Failure reduces redundancy of HHSI/CHG pump One HHSI/CHG pump is charging deliver emergency coolant providing emergency coolant to the discharge header used for normal charging of pump 1 working through the BIT to the RCS at high RCS pressures. Fluid pressure and flow RCS during plant operation.

(pump 2 fluid RCS at the prevailing flow from HHSI/CHG pump 1 (pump indication at MCB. analogous) incident RCS pressure. 2) will be lost. Minimum flow Open/close pump Charging pump 3 is lined up requirements for HHSI will be met by switch-gear circuit to SSPS train "A" when HHSI/CHG pump #2 (pump 1). breaker indication on replacing pump 1 or on MCB. Circuit breaker SSPS train "B" when close position replacing pump 2. monitor light for Replacement requires group monitoring of operator action for the line component at MCB. up of pump and line up of Common breaker trip isolation valves. alarm at MCB. Technical specifications limiting conditions of operation requires inoperable ECCS subsystem to be restored to an OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours. Analysis of HHSI/CHG pump 3 being on line is analogus to that presented for HHSI/CHG pumps 1 and 2.

4. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #1 operated close on fluid flow from the providing isolation of HHSI/CHG gate valve demand. HHSI/CHG pump pump miniflow line. Negligible effect 1-8106 discharge header to the on system operation. Alternate seal water heat isolation valves l- 8109A and 1-exchanger via minimum 8109B in HHSI/CHG pump flow bypass line. discharge lines provides backup miniflow line isolation.

Amendment 61 Page 2 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks*

5. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #1. Valve 1-8109C provides operated close on fluid flow from providing isolation of HHSI/CHG isolation to miniflow line if globe valve demand. HHSI/CHG pump 1 pump miniflow line. Negligible effect HHSI/CHG pump 3 is on 1-8109A (1- (pump 2) to the seal on system operation. Alternate line. Analysis for this valve 8109B water heat exchanger isolation valve 1-8106 provides being in service is analogous analogous) via minimum flow backup miniflow line isolation. to that presented for valves bypass line. 1-8109A and 1-8109B.

6. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #1 operated close on fluid flow from the providing isolation of HHSI/CHG except no valve close gate valve demand. HHSI/CHG pump pump discharge to normal charging monitor alarm for 1-8107 (1- discharge header to the line of CVCS. Negligible effect on group monitoring.

8108 CVCS normal charging system operation. Alternate isolation analgous) line to the RCS. valve 1-8108 (1-8107) provides backup normal CVCS charging line isolation.

8. Motor Fails to Provides isolation of Failure reduces redundancy of Same as in item #2.

operated open on fluid discharge from the providing fluid flow from BIT to high gate valve demand. BIT to high head head injection header feeding the 1-8801A (1- injection header cold legs of RCS loops. Negligible 8801B connected to the cold effect on system operation. analogous) legs of RCS coolant Alternate isolation valve 1-8801B (1-loops. 8801A) opens to provide backup flow path to header.

11. Motor Fails Provides regulation of Failure reduces working fluid Same as item #1. In Valves are regulated by operated open. fluid flow through delivered to RCS from LHSI/RHR addition, pump signals from flow transmitter globe valve miniflow bypass line to pump 1 (pump 2). Minimum flow discharge header located in each pump 1-FCV- suction of LHSI/RHR requirements will be met by pressure and flow discharge header. The 602A (1- pump 1 (pump 2) to LHSI/RHR pump 2 (pump 1) indication at MCB. control valves open when a FCV-602B protect against delivering working fluid to RCS LHSI/RHR pump discharge analogous) overheating of the pump flow is less than 746 gpm and loss of suction flow and close when the flow to the pump. exceeds 1402 gpm.

Amendment 61 Page 3 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks* Fails Failure results in an insufficient fluid closed. flow through LHSI/RHR pump 1 (pump 2) for a small LOCA or steam line break resulting in possible pump damage. Minimum flow requirements will be met by LHSI/RHR pump 2 (pump 1) and HHSI/CHG pump 2 (pump 1) delivering coolant fluid to RCS.

12. Residual Fails to Provides fluid flow of Failure reduces redundancy of Same as that stated The LHSI/RHR pumps are heat deliver emergency coolant to providing emergency coolant to the for item #3 except sized to deliver reactor removal working the RCS when the RCS at low RCS pressure. Fluid LHSI/RHR pump coolant through the Residual pump 1 fluid. incident RCS loop flow from LHSI/RHR pump 1 (pump discharge pressure Heat Exchanger to meet the (pump 2) pressure drops below 2) will be lost. Minimum flow and flow indication at plant cooldown requirements shutoff head of pump requirement for LHSI will be met by MCB. and are used during plant (160 psig) and provides LHSI/RHR pump 2 (pump 1). cooldown and startup long term recirculation operation.

capability for core cooling following the injection phase of LOCA.

13. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #2. Valves open automatically operated open on fluid discharge from providing fluid flow from the on receipt of a 2/4 RWST gate valve demand. containment sump to Containment Sump to the RCS. Lo-Lo) level signal in 1-8811A (1- suction line of LHSI/RHR pump 1 not available for coincidence with SI S 8812A LHSI/RHR pump 1. recirculation. Minimum flow signal being present.

analogous) requirements will be met by LHSI/RHR pump 2 through opening Administrative procedures of isolation valves 1-8811B and 1- require reactor operator to 8812B. Negligible effect on system verify opening of sump operation. isolation valves.

14. Motor Fails to Same function as stated Same as item #13 except isolation Same as item #2.

operated open on for item #13 except valves 1-8811A and 1-8812A gate valve demand. applies to LHSI/RHR automatically open with flow 1-8811B (1- pump 2. provided by LHSI/RHR pump 1. 8812 B analogous) Amendment 61 Page 4 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks*

15. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #1.

operated close on fluid discharge from the providing RWST isolation from gate valve demand. RWST to suction line of suction line of LHSI/RHR pump 1 1-8809A (1- LHSI/RHR pump 1 (pump 2). Negligible effect on 8809B (pump 2). system operation. A series check analogous) valve 1-8958A (1-8958B) provides backup isolation against fluid flow from the suction of LHSI/RHR pump 1 (pump 2) to the RWST.

16. Deleted by Amendment No. 39

17. Motor Fails to Provides isolation of Failure reduces flow of recirculation Same as item #1. In Hot legs RCS coolant loop operated close. fluid flow from coolant to hot legs of RCS coolant addition LHSI/RHR recirculation required to gate valve LHSI/RHR pump 1 loops from LHSI/RHR pump 1 (pump pump discharge prevent boron precipitation 1-8888A (1- (pump 2) to cold leg 2). Minimum flow requirements to header pressure and problem for long-term core 888B is injection header of RCS hot leg of RCS coolant loops will be flow indication and cooling.

analogous) coolant loops. met by delivery of coolant from miniflow valve LHSI/RHR pump 2 (pump 1) and two monitoring at MCB. HHSI/CHG pumps.

18. Motor Fails to Provides isolation of Failure prevents fluid flow from Same as item #2. In LHSI will be realigned to the operated open on fluid flow from LHSI/RHR pumps to hot leg injection addition, LHSI/RHR cold legs and HHSI will be gate valve demand. LHSI/RHR pumps to hot header fo RCS coolant loops. pump discharge aligned to the hot legs. This 1-8889 leg injection header of header pressure and action will provide sufficient RCS coolant loops. flow indication and cooling to the core and miniflow valve prohibit boron precipitation monitoring at MCB. (Reference 6.3.1-1).

Amendment 61 Page 5 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks* Fails to Failure reduces redundancy of close on providing isolation of recirculation of demand. fluid into hot legs of RCS coolant loops by LHSI/RHR pumps. Negligible effect on recirculation into cold legs of RCS coolant loops. Two HHSI/CHG and two LHSI/RHR pumps can meet minimum flow requirements for RCS cold leg recirculation even with simultaneous flow provided to LHSI hot leg recirculation penetration.

19. Motor Fails to Provides isolation of No effect on system operation. Same as that stated operated open on fluid flow from HHSI/CHG pumps 1 and 2 will be for item #2. In gate valve demand. LHSI/RHR pump 1 provided suction head by LHSI/RHR addition, HHSI/CHG 1-8706A. (pump 2) via RHR Heat pump 2 (pump 1) via the common pump 1 (pump 2)

(1-8706B Exchanger, (exchanger charging pump suction header. flow indication at analogous) 2) to suction line of MCB. HHSI/CHG pump 1 (HHSI/CHG 2).

20. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #2. Valve is activated to open by operated close. fluid discharge from the providing isolation of fluid discharged a VCT low water level or by gate valve RWST to the suction of from residual Heat Exchanger 1 an SIAS. Prior to the closing 1-LCV- HHSI/CHG pump 1 (Exchanger 2) to RWST. No of the valve following an 115B/((1- (pump 2) and an immediate effect on system SIAS, reactor operator LCV-115D electrical interlock to the operation during recirculation. resets SIAS.

analogous) closing of isolation valve Alternate isolation check valve 1-1-LCV-1154C )1-LCV- 8926 in common line from RWST 115E). provides backup tank isolation.

21. Motor Fails to Provides isolation No effect on system operation. Same as item #1. The normal operating operated close. barrier to form two Backup isolation is provided by position of the valve during gate valve independent flow paths closing alternate isolation valve 1- recirculation changes if 1-8132A (1- in the vent of a single 8132B (1-8132A) HHSI/CHG pump #3 is on 8132B passive failure. line and is in operation and analogous) HHSI/CHG pump #1 is out-of-service.

Amendment 61 Page 6 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks*

22. Motor Fails to Provides an isolation No effect on system operation. Same as item #1. The normal operating operated close. barrier to form two Backup isolation is provided by position of the valve during gate valve independent flow paths closing alternate isolation valve 1- recirculation if HHSI/CHG 1-8133A (1- in the event of the 8133B (1-8133A) pump #3 is on line and is in 8133B single passive failure. operation and HHSI/CHG analogous). pump 2 is out-of-service.

23. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #2. In Valve is positioned open by operated open on fluid flow from providing fluid flow from HHSI/CHG addition HHSI/CHG reactor operator for gate valve demand. HHSI/CHG pump 1 pumps to cold legs of RCS coolant pump 1 flow recirculation into cold legs of 1-8885. discharge, line to cold loops. Minimum flow requirements indication at MCB. RCS coolant loops and legs of RCS coolant will be met by HHSI/CH pump #2 closed by the operator when loops. providing flow of coolant to cold legs recirculation into hot legs of via BIT cold leg injection line. RCS coolant loops is desired during long term incident recovery periods.

Fails to Failure reduces flow delivery of close on HHSI/CHG pumps to RCS hot legs. demand. Minimum flow will be met by HHSI/CHG pump #2 providing flow to its hot leg recirculation flow path.

24. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #2. Valves are activated to open operated close on fluid flow from providing isolation of fluid flow from by an SIAS. Prior to the gate valve demand. HHSI/CHG pump #2 HHSI/CHG pump 2 to cold legs of closing of the valves, reactor 1-8801A (1- discharge line via BIT to RCS coolant loops. Failure reduces operator resets the SIAS.

8801B cold legs of RCS flow delivery of HHSI/CHG pumps to Valves are closed by the analogous) coolant loops. RCS hot legs. Minimum flow will be reactor operator for met by HHSI/CHG pump No. 1 recirculation into hot legs of providing flow to its hot leg RCS coolant loops and open recirculation flow path. by the operator when recirculation into cold legs of RCS coolant is desired during long term incident recovery period.

25. Deleted by Amendment No. 39.

Amendment 61 Page 7 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks*

26. Motor Fails to Provides isolation of Failure reduces redundancy of Same as item #19. Valve is positioned open by operated open on fluid flow from providing fluid flow from HHSI/CHG reactor operator for gate valve demand. HHSI/CHG pump 1 pumps to hot legs of RCS coolant recirculation into hot legs of 1-8884 (1- (pump 2) discharge line loops. Minimum flow requirements RCS coolant loops and 8886 to hot legs of RCS will be met by HHSI/CHG pump 2 closed by the operator when analogous) coolant loops. (pump 1). recirculation into cold legs of RCS coolant loops is desired during long term incident recovery period.

Fails to Failure allows for the simultaneous Same as item #19. close on recirculation of coolant into hot and demand. cold legs at RCS coolant loops during cold leg recirculation operation. Minimum flow requirements will be met by HHSI/CHG pump 2 (pump 1) and LHSI/RHR pump flow to cold legs of RCS coolant loops.

27. Motor Fails to Protects HHSI pump Failure could result in failure of the Valve open operated open on from dead heading weak HHSI pump. However, pump indication. Monitor globe valve demand. subsequent on SI from other train is still available and 1-8489A (1- coincident with high sufficient. Light Box 3A on the 8489B RCS pressure. MCB.

analogous) Fails to Isolates to maximize Failure would result in reduction of Valve close The valve may open close on HHSI flow during a SI flow by as much as 65 gpm on indication. subsequent to an SI since SI demand. LOCA, MSLB etc. On one train. However, flow from other may actuate before the high SI coincident with low train is still available and sufficient. Monitor Light Box 3A RCS pressure signal clears. RCS pressure. on the MCB. As the associated event progresses, the valve will reclose.

28. Manual Fails to Provides fluid from the Failure prevents use of the Operator unable to operated open on RWST to the suction of hydrostatic test pump. However, the open the valve.

globe valve demand. the hydrostatic test pump does not perform a safety 2CT- pump. function. Hence no safety V144SAB-1 significance. Amendment 61 Page 8 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.1-1 EMERGENCY CORE COOLING SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Failure Detection Component Mode Function* Effect on System* Method** Remarks* Fails to Provides boundary Failure could allow loss of fluid from Operator unable to This is a 2 1500 lb ss globe close on isolation between the the RWST through a line rupture in close valve. valve in a very low demand. safety RWST and the the non-safety hydrostatic test pump temperature and pressure non-safety hydrostatic piping. service. Failure of this valve test pump. to close when required is not a credible event. List of acronyms and abbreviations BIT - Boron injection tank CHG - Charging HHSI - High head safety injection LHSI - Low head safety injection LOCA - Loss-of-coolant accident MCB - Main control board RCS - Reactor Coolant System RHR - Residual heat removal RWST - Refueling water storage tank SIAS - safety injection actuation signal SIS - Safety Injection System SSPS - Solid state protection system VCT - Volume control tank Amendment 61 Page 9 of 9

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-1 EMERGENCY CORE COOLING SYSTEM COMPONENT PARAMETERS Accumulators Number 3 Design pressure (psig) internal 700 external 60 Design temperature (F) 300 Operating temperature (F) 120 Normal pressure (psig) 665 Minimum operating pressure (psig) 585 3 Total volume (ft. ) 1450 each 3 Normal operating water volume (ft. ) 1012 each 3 Volume N2 gas (ft. ) 438 Boron concentration, (ppm) 2400 - 2600 Centrifugal Charging Pumps Number 3 Design pressure (psig) 2800 (Note 1) Design temperature (F) 300 (a) Design flow (gpm) 150 Design head (ft.)* 6300 Maximum flow (gpm)** 685 Head at maximum flow (ft.) 3100 Motor rating (hp) 900 Residual Heat Removal Pumps Number 2 Design pressure (psig) 600 Design temperature (F) 400 Design flow (gpm) 3750 Design head (ft.) 240 NPSH required @ 4500 gpm (ft.)* 19 Available NPSH (ft.) 20.85 Motor Rating (HP) 300

Orifices are installed in the safety injection headers, to limit runout flow to a maximum of approximately 4500 gpm.

Residual Heat Exchangers (See Section 5.4.7 for Design Parameters) Hydrostatic Test Pump Number 1 Design pressure (psig) 3300 Design temperature (F) 300 Normal operating temperature ambient Design flow rate (gpm) 24.5 Develop head (ft) at design flow 7000 Boron Injection Tank Number 1 Total volume (gal.) 900 Boron concentration (ppm) 0 - 2,600 Design pressure (psig) 2735 (Note 2) Operating pressure 2712 Design temperature (F) 300 Operating temperature (F) 120 Amendment 62 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-1 EMERGENCY CORE COOLING SYSTEM COMPONENT PARAMETERS Motor Operated Valves Maximum Opening or Closing Time Fast (b) 3" and 4" 1500# valves 10 sec. (b) 6" - 12" valves 15 sec. 14" 20 sec. Slow Up to and including 8 in. 12 in./min./in. of nominal valve size Over 8 in. 2 min. NOTES: (a) Includes 60 gpm allowance for miniflow. (b) Stroke times of the following valves are 15 seconds: LC V115 B/D and 8888 A/B. Stroke times of the following valves are 20 seconds: 8808 A/B/C, 8811 A/B, and 8812 A/B. Stroke time for 8889 is < 30 seconds. Stroke time for 8706 A/B 30 seconds. Note 1: With CVCS alignment in normal or alternate miniflow, a limited portion of the system piping and components may experience a momentary increased pressurization, above the system design pressure, due to the reduction in the flow paths. The piping and components within these flow paths have been qualified to a pressure equal to or greater than this pressure anomaly (up to 3100psig). Note 2: The Boron Injection Tank has been evaluated for 2800 psig. Amendment 62 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-2 EMERGENCY CORE COOLING SYSTEM RELIEF VALVE DATA Maximum Fluid Inlet Backpressure Total Fluid Temperature Set Pressure Constant Backpressure Description Discharged Normal (psig) (psig) (psig) Capacity N2 supply to N2 120 700 0 0 1500 scfm accumulators Residual heat Water 120 600 3 50 20 gpm removal pump safety injection line Accumulator to N2 gas 120 700 0 0 1500 scfm Containment Hydrostatic Test Water 120 700 0 0 30 gpm Pump Discharge Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-3 MOTOR OPERATED ISOLATION VALVES IN THE EMERGENCY CORE COOLING SYSTEM Valve Position Location Identification Interlocks Automatic Features Indication Alarms Accumulator Isolation Valves 8808 A,B,C "S" signal, RCS pressure > unblock Opens on "S" signal if closed and RCS MCB Yes-out of position pressure > unblock Recirculation Containment 8811 A,B "S" signal, RWST "Lo-Lo" signal Opens on coincident "S" and RWST "Lo- MCB Yes-out of position Sump Isolation Valves 8812 A,B Lo" signals CVCS Suction from RWST LCV-115 B,D "S" signal Opens on "S" signal MCB None CVCS Normal Suction LCV-115 C,E "S" signal Closes on "S" signal if CVCS suction MCB Yes-out of position valves from RWST open CVCS Normal Discharge 8107, 8108 "S" signal Closes on "S" signal MCB None Boron Injection Tank 8801, A,B "S" signal Opens on "S" signal MCB Yes-out of position Discharge RWST to RHR Pump 8809, A,B None None MCB Yes-out of position Suction Charging Pump Miniflow 8109, A,B "S" signal Closes on "S" signal MCB Yes-out of position 8106 Charging Pump Miniflow 8109C "S" signal Closes on "S" signal MCB None HHSI-HL Recirculation Gate 8884, 8886 None None MCB Yes-out of position Valves HHSI-CL Recirculation Gate 8885 None None MCB Yes-out of position Valve LHSI Crossover 8887A,B None None MCB Yes-out of position LHSI-Recirculation Gate 8888A,B None None MCB Yes-out of position Valves LHSI to RCS Hot Legs 8889 None None MCB Yes-out of position RHR Discharge to Charging 8706A,B Cannot be opened unless at least None MCB Yes-out of position Pump Suction one RHR suction isolation valve in corresponding subsystem closed Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-3 MOTOR OPERATED ISOLATION VALVES IN THE EMERGENCY CORE COOLING SYSTEM Valve Position Location Identification Interlocks Automatic Features Indication Alarms CHG Pump Suction 8130A,B None None MCB Yes-out of position Crossover 8131A,B Charging Pump Discharge 8132A,B None None MCB Yes-out of position Crossover 8133A,B Charging Pump Alternate 8489A,B "S" signal with LHSI crossover Open/close on "S" signal and RCS Press MCB No Miniflow isolates 8706A,B to prevent recirc. 8490A,B water from entering RWST None MCB No Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-4 MATERIALS EMPLOYED FOR EMERGENCY CORE COOLING SYSTEM COMPONENTS Component Material Accumulators Carbon steel clad with austenitic stainless steel Boron Injection Tank Austenitic stainless steel Pumps Centrifugal charging Austenitic stainless steel Residual heat removal Austenitic stainless steel Hydrotest Austenitic stainless steel Residual heat exchangers Shell Carbon steel Shell end cap Carbon steel Tubes Austenitic stainless steel Channel Austenitic stainless steel Channel cover Austenitic stainless steel Tube sheet Austenitic stainless steel Valves Motor operated valves containing radioactive fluids Pressure containing parts Austenitic stainless steel or equivalent (Refer to Table 6.1.1-1). Body-to-bonnet Bolting and nuts Low alloy steel Seating surfaces Hard faced Stems Austenitic stainless steel or 17-4 pH stainless Motor operated valves containing nonradioactive, boron-free fluids Body, bonnet and flange Carbon steel Stems Corrosion resistance steel Diaphragm valves Austenitic stainless steel Accumulator check valves Parts contacting borated water Austenitic stainless steel Clapper arm shaft 17-4 pH stainless Relief valves Stainless steel bodies Stainless steel Carbon steel bodies Carbon steel All nozzles, discs, spindles and guides Austenitic stainless steel Bonnets for stainless steel valves without a balancing Stainless steel or plated carbon steel bellows All other bonnets Carbon steel Piping All piping in contact with borated water Austenitic stainless steel Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-5 EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS LONG TERM PHASE Indication of Loss of Flow Flow Path Path Alternate Flow Path Low Head Recirculation From containment sump to Accumulation of water in a Via the independent, identical low head injection header via residual heat removal pump low head flow path utilizing the residual heat removal compartment or reactor the second residual heat pumps and the residual heat auxiliary building sump exchanger and residual heat exchangers removal pump High Head Recirculation From containment sump to Accumulation of water in a From containment sump to the high head injection header residual heat removal pump the charging headers via via residual heat removal compartment or the reactor alternate residual heat pump, residual heat auxiliary building sump or removal pump, residual heat exchanger and the charging charging pump compartments exchanger pumps Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-6 SEQUENCE OF SWITCHOVER OPERATION FROM INJECTION TO RECIRCULATION Manual operator actions are required to complete the switchover from the injection mode to the recirculation mode. During the injection mode, the operator verifies that all ECCS pumps are operating and monitors the RWST and reactor building recirculation sump levels in anticipation of switchover. Also during the injection mode, operator action is required to close the power supply breakers for Charging Pump Discharge Header Crossover valves 8132 A/B and 8133 A/B in preparation for their operation per step six below. By closing the Charging Pump Discharge Header Crossover valve breakers during the injection mode, the time required to perform the actions of Table 6.3.2-6 is unaffected. Charging Pump Suction Header Crossover valves 8130 A/B and 8131 A/B are not required for train separation but their MOV supply breakers are also closed at this time to provide for passive failures during the recirculation mode as required. Upon receipt of the RWST low-low level signal in conjunction with the safety injection signal, the containment sump isolation valves automatically open. Following this automatic action, the operator is required to complete the switchover. The operator normally opens the component cooling water inlet isolation valves to the residual heat removal heat exchanger prior to switchover. The following manual actions must be performed to align the charging pump suction to the residual heat removal pumps discharge.

1. Verify that the containment sump isolation valves are open and close the residual heat removal pump suction valves from the refueling water storage tank.

2. Close one (not both) of the cold leg header isolation valves associated with the RHR pumps. (This action prevents RHR pump runout in the recirculation condition.) Close the charging pump alternate miniflow isolation valves.

3. Open residual heat removal pump discharge valves to the charging pump suction.

All ECCS pumps are now aligned with suction flow from the containment sump. The operator verifies proper operation and alignment of all ECCS components and proceeds to complete the following manual actions to align the ECCS in redundant flow path for long term recirculation operation.

4. Close refueling water storage tank valves to charging pump suction and place associated control switches into pull-to lock.

5. Open valve in the alternate high head cold leg recirculation line.

6. Close valves (depending on operating charging pumps) in the discharge header to establish two separate high head recirculation systems.

Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.3.2-6 (Continued) The following manual operator actions are required to perform the change-over operation from the cold leg recirculation mode to the hot leg recirculation mode.

1. Close the cold leg header isolation valves associated with the RHR pumps.

2. Deleted.

3. Open the hot leg header isolation valve from the RHR pumps. If the isolation valve does not open, re-align the RHR pumps to the cold leg header.

4. Stop charging pump No. 1. If pump No. 1 was out of service prior to the accident, stop the swing pump (charging pump No. 3).

5. Close the alternate high head cold leg header isolation valve and open the corresponding high head hot leg header isolation valve.

6. Restart the charging pump stopped in Step 4.

7. Stop charging pump No. 2. If pump No. 2 was out of service prior to the accident, stop the swing pump (charging pump No. 3).

8. Close the boron injection tank discharge isolation valves and open the corresponding high head hot leg header isolation valve.

9. Restart the charging pump stopped in step 7.

Contingency actions are required in the event either high head hot leg isolation valves is pressure locked. The following sequence is used to open the pressure locked valve.

1. Open the normal miniflow valves of the associated charging pump for the affected high head hot leg isolation valve.

2. Start the charging pump.

3. Open the affected high head hot leg isolation valve.

4. Shut the normal miniflow valves for the charging pump.

Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-7 EMERGENCY CORE COOLING SYSTEM SHARED FUNCTIONS EVALUATION Normal Operating Component Arrangement Accident Arrangement Refueling water storage tank Lined up to suction of residual Lined up to suction of heat removal pumps centrifugal charging and residual heat removal pumps Centrifugal charging pumps Lined up for charging service Suction from refueling water suction from volume control storage tank, discharge lined tank, discharge via normal up to boron injection tank. charging line Valves for realignment meet single failure criteria Residual heat removal pumps Lined up to cold legs of Lined up to cold legs of reactor coolant piping reactor coolant piping Residual heat exchangers Lined up to cold legs of Lined up to cold legs of reactor coolant piping reactor coolant piping Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-8 NORMAL OPERATING STATUS OF EMERGENCY CORE COOLING SYSTEM COMPONENTS FOR CORE COOLING Number of charging pumps operable 2* Number of residual heat removal pumps operable 2 Number of residual heat exchangers operable 2 Refueling water storage tank minimum contained volume (gal.) 434,302** Boron concentration in refueling water storage tanks, minimum (ppm) 2,400 Boron concentration in accumulator, minimum (ppm) 2,400 Number of accumulators 3 Minimum accumulator pressure (psig) 585 Nominal accumulator water volume (ft.3) 1,012 System valves, interlocks, and piping required for the above components which All are operable

Three charging pumps are installed. A maximum of two may be operable at one time.
- Lower limit of "low" alarm. Note: Technical Specification conservatively adjusts this value per the Technical Specification Bases.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-9 RWST OUTFLOW LARGE BREAK - NO FAILURES TIME TOTAL CHANGE IN TOTAL REQUIRED ELAPSED RWST FLOW RWST VOL. RWST VOL. PER STEP TIME RATE PER STEP PER STEP CHANGE (1) (3)(5) (2)(6)(7) STEP (SEC) (SEC) (GPM) (GAL) (GAL) (4) 0 20 20 17,820 5,940 5,940 1 62 82 25,055 25,890 31,830 (8) 2&3 90 172 10,523 15,785 47,615 Completed RWST 39 211 2,826 1,837 49,452 (9) Isolation NOTES: (1) See Table 6.3.2-6 for a description of the steps. (2) Flow rates are based on pump flows as follows: RHR pump = 3096.5 gpm per pump Charging pump = 426 gpm per pump CS pump = 2137 gpm per pump (3) Valve operating times are maximum operating times. (4) Time for valves 8811A/B and 8812A/B to automatically open. (5) Time required to complete the required action includes a conservative 30 seconds for operator response time for each manual step. (6) The flow rate in this column represents an average value during the entire time interval for its respective step. (7) Flow out of the RWST during switchover includes allowances for both pumped flow to the RCS and containment and backflow to the containment sump. (8) Following the completion of this step, RHR and Charging pumps are aligned with suction flow from the containment sump. (9) Due to the long stroke times of the containment spray valves, the containment spray pump suction is not isolated from the RWST until after the ECCS pumps have been isolated. Amendment 62 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 Table 6.3.2-9 (Continued) TIME TOTAL CHANGE IN TOTAL REQUIRED ELAPSED RWST FLOW RWST VOL. RWST VOL. PER STEP TIME RATE PER STEP PER STEP CHANGE (1) (3)(5) (2)(6)(7) STEP (SEC) (SEC) (GPM) (GAL) (GAL) (4) 0 20 20 17,586 5,862 5,862 1 62 82 25,444 26,292 32,154 (8) 2&3 90 172 17,314 25,970 58,125 RWST (11) 39 211 10,098 6,564 64,688 Isolation NOTES: (1) See Table 6.3.2-6 for a description of the steps. (2) Flow rates are based on pump flows as follows: RHR pump = 3096.5 gpm per pump Charging pump = 426 gpm per pump CS pump = 2137 gpm per pump (3) Valve operating times are maximum operating times. (4) Time for valves 8811A/B and 8812A/B to automatically open. (5) Time required to complete the required action includes a conservative 30 seconds for operator response time for each manual step. (6) The flow rate in this column represents an average value during the entire time interval for its respective step. This is conservative since valve repositioning may reduce the flow rate during the time interval. (7) Flow out of the RWST during switchover includes allowances for both pumped flow to the RCS and containment and backflow to the containment sump. (8) Following the completion of this step all ECCS pumps are aligned with suction flow from the containment sump with the exception of one residual heat removal pump due to the single failure. (9) Based on Large Break LOCA in conjunction with a single failure of one of the RWST to residual heat removal pump isolation valves (8809A or 8809B) to close on demand. (10) Deleted by Amendment No. 49. (11) Due to the long stroke times of the containment spray valves, the containment spray pump suction is not isolated from the RWST until after the ECCS pumps have been isolated. Amendment 62 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.3.2-10 PUMPS AND VALVES REQUIRED FOR ECCS OPERATION Tag # System Train Safety Class Operator* Valves: 9431 A/B CC SA/SB 3 M 9370 CC SA 3 M 9371 CC SB 3 M 9384 CC SA 3 M 9385 CC SB 3 M 8888 A/B SI SA/SB 2 M 8887 A/B SI SA/SB 2 M 8889 SI SA 2 M 8811 A/B SI SA/SB 2 M 8812 A/B SI SA/SB 2 M 8809 A/B SI SA/SB 2 M 8808 A/B/C SI SA/SB/SA 2 M 8706 A/B RH SA/SB 2 M 8801 A/B SI SA/SB 2 M 8803 A/B SI SA/SB 2 M** 8885 SI SA 2 M 8886 SI SB 2 M 8884 SI SA 2 M FCV 113 A CS SN 3 A 8105 CS SB 3 M 8106 CS SA 2 M 8108 CS SB 2 M 8133 A/B CS SA/SB 2 M 8132 A/B CS SA/SB 2 M 8109 A/B/C CS SB/SB/SB 2 M 8131 A/B CS SA/SB 2 M 8130 A/B CS SA/SB 2 M LCV 115 C/E CS SA/SB 2 M LCV 115 B/D CS SA/SB 2 M 8104 CS SB 2 M 3SW - B1SA-1 SW SA 3 M 3SW - B2SB-1 SW SB 3 M 3SW - B3SA-1 SW SA 3 M 3SW - B4SB-1 SW SB 3 M 3SW - B5SA-1 SW SA 3 M 3SW - B6SA-1 SW SA 3 M Pumps: APCH 1/2/3 APHR 1/2 APCC 1/2/3 APSN - 1A-SA APSN - 1B-SB

NOTE: M = MOTOR, A = Air
- Locked Open Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.4.2-1 CONTROL ROOM BUTTERFLY VALVES LEAKAGE RATE ESTIMATE

1. COMPONENTS: Butterfly valves in:

a) Exhausts b) Normal Outside Air Intake SIZE: 12 inch diameter (exhaust) 16 inch diameter (intake) (1) QUANTITY: Four (2 valves arranged in series in each of two paths) LEAK RATE AT 13.8 PSIG: 0.018 (0.024) cubic feet per day per exhaust (2) (intake) valve (3) -6 LEAK RATE AT + 1/8 INCH W.G. : 0.53 X 10 cfm per two valves

2. COMPONENTS: Butterfly valves in:

a) Purge Exhausts b) Purge Make-Up SIZE: 30 inch diameter (exhaust) 36 inch diameter (make-up) (1) QUANTITY: Four (2 valves arranged in series in each of two paths) LEAK RATE AT 13.8 PSIG: 0.045 (0.054) cubic feet per day per exhaust (2) (make-up) valve (3) -6 LEAK RATE AT + 1/8 INCH W.G. : 1.24 X 10 cfm per two valves

3. COMPONENTS: Butterfly valves in:

Post-Accident Air Intakes (two) SIZE: 12 inch diameter (1) QUANTITY: Four (2 valves arranged in series in each of two paths) (2) LEAK RATE AT 13.8 PSIG: 0.018 cubic feet per day per valve (3) -6 LEAK RATE AT + 1/8 INCH W.G. : 0.45 X 10 cfm per two valves

                                                                               -6 TOTAL LEAKAGE TO THE OUTSIDE FROM VALVES:                        1) 0.53 X 10 cfm              (For conservatism,
                                                                              -6                       -6

2) 1.24 x 10 cfm 3.0 x 10 cfm is

-6

3) 0.45 x 10 cfm used.)

                                                                                -6 TOTAL =          2.22 X 10 cfm NOTES:

1. There are a total of 12 isolation valves, two in series in each air path. However, it has been assumed that only one valve closes in each path following control room isolation.

2. Based on AEC R&D Report NAA-SR-101000, Reference 2, Section A-2, p III 105.

3. For control room positive pressure +1/8 inch w.g.

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.4.2-2

SUMMARY

OF MAIN CONTROL ROOM LEAK RATE CALCULATION(1) NUMBER NUMBER OF LEAKAGE PER TOTAL CFM PATH OF REFERENCE LEAKAGE COEFFICIENT UNIT COMPONENT (1) (2) NO. COMPONENT UNIT UNITS DETAIL A B AP+BP1/2 LEAKAGE

1. Hollow metal door, metal interlocking 3' 0x7' 0 6 ADS III-A-2 4.0 22.0 8.28 49.68 gasketed weatherstripping, door opening in (4 single and 1 double)

                                                                                                 -6                           -7

2. Door Frames Ft. 106 ADS I-A-7 4x10 0 5x10 .00006

                                                                                                 -6                              -7

3. Walls Ft.2 6,000 ADS I-A-2(1) 1x10 0 1.25x10 .00075

                                                                                                 -6                              -7

4. Slab Ft.2 10,800 ADS I-A-2(1) 1x10 0 1.25x10 .00135

                                                                                                    -3                         -4

5. Juncture of floor slab and wall Ft. 450 ADS I-A-3(1) 1.6x10 0 .2x10 .09

                                                                                                 -5                             -5

6. Eave Ft. 450 ADS I-A-5 6x10 .75x10 .0034

                                                                                                    -5                         -5

7. Corners, columns and wall joints with Ft. 340 ADS I-A-6 Case 1 1.6x10 0 .2x10 .0007 caulking

                                                                                                    -4                            -4

8. Penetrations for electrical cables Ft. 730 ADS III-D-1 1.3x10 0 .1625x10 .0118

                                                                                                    -5                            -5

9. Penetrations for HVAC ducts In. of 1,040 ADS III-D-1 Case 2 1.3x10 0 .1625x10 .00169 Seal

                                                                                                                                                -6(4)

10. Isolation Butterfly Valves ADS A-2 Case 2 3x10

                                                                                                    -5                            -5

11. Pipe Penetrations In. of 116 ADS III-D-1 Case 2 1.3x10 0 .1625x10 .0002 Seal

12. HVAC Equipment and Ductwork 15.8 (Outside of Envelope)

(1) Subtotal (1-12) 66 x 2

13. Opening and closing of doors Note (3) 10.00 Total 142 (1) Based on AEC R+D Report NAA-SR-10100 (2) Leakage estimate based on AP=0.125 in w.g.

(3) See standard review plan Section 6.4 III3d2ii (4) See Table 6.4.2-1 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.4.4-1 TOXIC CHEMICALS STORED ONSITE HORIZONTAL DISTANCE FROM THE CONTROL NO. OF ROOM NORMAL TOXIC CHEMICAL LOCATION TANKS/CAPACITY, EACH VENTILATION INTAKE, FT. Sulfuric Acid (H2SO4) (100%) At Cooling Tower 1/7800 gal. 950 At Turbine Bldg 1/5473 gal. 400 At Water Treat. Bldg. 1/7820 gal. 530 Sodium Hydroxide At Cooling Tower 1/1700 gal. 1000 At Turbine Bldg 1/8883 gal. 380 At Water Treat. Bldg. 1/10,500 gal. 750 Nitrogen (N2) (Liquid) Gas Storage Area 1 system/10,584 lbs. 700 Carbon Dioxide (CO2) (Liquid) Gas Storage Area 1 system/4,000 lbs. liquid 700 1,290 lbs. vapor Oxygen (O2) Gas Storage Area 1 System/60,400 scf 700 Hydrogen (H2) (Liquid) Gas Storage Area 1 System/1,500 gal. 700 Nitrogen (N2) (Liquid) Turbine Bldg. 1 System/6,020 gal. 375 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-1 DESIGN DATA FOR FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM Quantity Two (2) Identical Units, One Standby Each Unit contains the following:

1. Exhaust Fan Quantity 1 Type Centrifugal Direct Drive Air Flow, Per Fan, acfm 6600 Static Pressure, in. wg. 16.1 Code Air Moving and Conditioning Association (AMCA),

Anti-Friction Bearing Manufacturers Association (AFBMA)

2. Exhaust Fan Motors Quantity, Per Fan 1 Type 30 HP, 460 V, 60 Hz 3 phase Induction Type Insulation Class B, Powerhouse Enclosure Drip-proof Code NEMA Class B, IEEE Class 1E

3. HEPA Filters Quantity, Per System Two banks Air Flow, acfm 6600 Cell (Unit) Size 24" H x 24" W x 11 1/2" deep Max. Resistance Clean, in. wg. 1.0 Max. Resistance Loaded, in. wg. 2.0 Efficiency 99.97 percent when tested with 0.3 micron Dioctylphtalate smoke Material Meets the requirements of ANSI/ASME N509-1980

4. Charcoal Adsorbers Type Multiple gasketless bed cells in air-tight housing Quantity, Per System 1 New media Impregnated coconut shell (Meeting the requirement of ANSI/ASME N509-1980 Table 5.1, with the exception that the 30°C/95% relative humidity methyl iodide test is done per ASTM D3803-1989 Depth of Bed (in.) 2" Face Velocity 40 Average Atmosphere Residence Time 0.25 seconds per two in. of adsorber bed Adsorber Capacity of Iodine Loading 2.5 mg of total iodine (radioactive plus stable) per gram of activated carbon Efficiency:

Elemental iodine 95% at 70% RH Organic iodine 95% at 70% RH Adsorbent Acceptance and Inplace Leak Carbon Laboratory Acceptance Testing will be performed in accordance Test Criteria with, and will meet the requirements of, position C.6 of R.G. 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2. Adsorber In place Leak Testing will be performed in accordance with, and will meet the requirements of, position C.5.d of R.G. 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2. Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-1 DESIGN DATA FOR FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM

5. Prefilters Quantity, Per System One bank Type Medium efficiency, dry and replaceable Material Ultra-fine glass fiber

6. Heating Coil 1 per filter train Quantity Electric Capacity (kw) 40 (Sufficiently sized to reduce the relative humidity of the inlet air from 100% to 70%)

Code Underwriter Laboratories (UL), National Electrical Manufacturers Association (NEMA), National Electric Code (NEC), IEEE Class 1E Material Galvanized Steel

7. Demister Quantity Per System 1 bank Air flow acfm 6600 Max. Resistance Clean in. wg 1.28 Max. Resistance Loaded in. wt 2.0 Material Woven stainless steel and glass fiber mesh

8. Valves Quantity Per System Two per system Type Manual and motorized Air Butterfly valve Flow Per Fan, acfm 6600 Material Stainless steel Code ASME III, Class 3 IEEE Class 1E (Motor Operated Valves)

Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-2 COMPARISON OF FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM, REACTOR AUXILIARY EMERGENCY EXHAUST SYSTEM AND CONTROL ROOM EMERGENCY FILTRATION SYSTEM WITH REGULATORY POSITIONS OF R.G 1.52, REVISION 2 Regulatory Position System Design Features 1a,b,c,d,& e Comply. 2a Comply. 2b Comply. Each air cleaning unit and corresponding channel motorized valves are physically separate from each other. 2c Comply. 2d Not applicable. The systems are located outside the Containment and therefore not subject to pressure surges. 2e Comply. 2f Comply. 2g The system is instrumented to signal, alarm and record pertinent pressure drops, temperatures and flow rates at the main Control Room as described in Chapter 7. 2h Components comply with IEEE Standards. Refer to Chapter 7 for detailed information. 2i Comply. FHB Emergency Exhaust System is automatically actuated by redundant seismic Category I radiation monitors. RAB Emergency Exhaust System is automatically actuated by redundant SIS. Control Room Emergency Filtration System is automatically actuated by redundant SIS or seismic Category I radiation monitors. 2j The system is designed to facilitate maintenance in accordance with R.G. 8.8. Isolation valves are provided at the inlet and outlet of each filter train. The plant layout and the filter train design permit replacement of each air cleaning unit as two segmented sections without removal of individual components. 2k The FHB and RAB air cleaning units are exhaust systems and have no outside air intake openings. The outside air intake openings on the control room air cleaning unit are adequately protected and have radiation detectors. 2l Comply. 3a,b,c,d,e,f,g,h,i,j, Comply, with the exception that the activated charcoal is manufactured and tested per ANSI/ASME N509-1980 with the exception that the 30°C/95% relative humidity methyl iodide test is done per ASTM D3803-1989. 3d Reg. Guide 1.5.2 and ANSI-N509-1980 require HEPA filters to be in accordance with MIL-F-51068. MIL-F-50168 has been canceled and replaced by ASME AG-1; therefore, HEPA filter requirements will be allowed to either specification. 3k The RAB Emergency Exhaust units do not require a low-flow bleed air system; however, the interconnecting duct originally installed for this purpose has been left in place. The FHB emergency exhaust units do not require a low-flow bleed air system and the interconnecting duct originally installed for bleed air purposes was blanked off. The Control Room Emergency Filtration System does not require a low-flow bleed air system. 3l,m,n,o,p Comply. 4a,b,c,e, Comply. Amendment 63 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-2 COMPARISON OF FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM, REACTOR AUXILIARY EMERGENCY EXHAUST SYSTEM AND CONTROL ROOM EMERGENCY FILTRATION SYSTEM WITH REGULATORY POSITIONS OF R.G 1.52, REVISION 2 Regulatory Position System Design Features 4d This section required that each ESF atmosphere cleanup train should be operated at least 10 hours per month, with the heaters on (if so equipped), in order to reduce the buildup of moisture on the absorbers and HEPA filters. The duration of the monthly operation of each ESF atmosphere cleanup train was changed from requiring 10 continuous hours to 15 continuous minutes with implementation of License Amendment 156, which was based on NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 0, Revise Ventilation System Surveillance Requirements to Operate for 10 hours per Month. Heaters are not required to be on during the monthly operation of each ESF atmosphere cleanup train as a result of License Amendment 170, which is also based upon TSTF-522. 5a,b,c,d, Comply with the exception that the In-Place Testing be performed in accordance with ANSI/ASME N510-1980. Test agent injection and sampling points are provided as indicated in the Ebasco Equipment Specification CAR-SH-BE-31. 6a, b Comply with exceptions: The new activated carbon is manufactured and tested per ANSI/ASME N509-1980 with the additional exception that the 30°C/95% relative humidity methyl iodide test is performed per ASTM D3803-1989. Laboratory tests of representative samples of used activated carbon are to be performed per ASTM D3803-1989 at 30°C and 95% relative humidity with a methyl iodide penetration of 2.5% for 2 inch beds outside of primary containment and a methyl iodide penetration of 0.5% for 4 inch beds outside of primary containment. Amendment 63 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-3 DESIGN DATA FOR REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM Quantity Two (2) identical units one standby Each unit contains the following:

1. Exhaust Fans Quantity, Per System 1, 100% each, centrifugal with variable inlet vanes, single width, single inlet, belt driven Capacity, Per Fan acfm 6800 Code Air Moving and Conditioning Association (AMCA), Anti-Friction Bearing Manufacturer Association (AFBMA)

2. Motors Quantity, Per Fan 1 Type 30 HP, 460 V, 60 Hz 3 phase, horizontal induction type Insulation Class H Enclosure and Ventilation TEFC-XT Code NEMA IEEE Class 1E

3. Electric Heating Coils Quantity, Per System 1 Type Electric Capacity (kW) Per Coil 40 (Sufficiently sized to reduce the relative humidity of the inlet air from 100%

to 70%) Code Underwriter Laboratories (UL), National Electrical Manufacturers Association (NEMA), National Electric Code (NEC) IEEE Class 1E Material Galvanized steel

4. Medium Efficiency Filters Quantity, Per System 1 Bank Type Extended media Material Glass fiber

5. HEPA Filters Quantity, Per System 2 banks Cell Size 24 in. high, 24 in. wide, 11 1/2 in. deep Max. Resistance Clean, in. wg. 1.0 Max. Resistance Loaded, in. wg. 2.0 Efficiency 99.97 percent when tested with 0.3 micron DOP Material Meets the requirements of ANSI/ASME N509-1980

6. Charcoal Adsorbers Type Multiple gasketless bed cells in air-tight housing Quantity, Per System 1 New Media Impregnated coconut shell (Meeting the requirements of ANSI/ASME N509 1980, Table 5.1 with the exception that the 30°C/95% relative humidity methyl iodide test is done per ASTM D3803-1989.

Depth of Bed (in.) 2 in. Face Velocity (fpm) 40 Average Atmosphere Residence 0.25 seconds per 2 in. of adsorber bed Time Adsorber Capacity of Iodine 2.5 mg of total iodine (radioactive plus stable) per gram of activated carbon Loading Amendment 61 Page 1 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-3 DESIGN DATA FOR REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM Efficiency: Elemental iodine 95% at 70% RH Organic iodine 95% at 70% RH

6. Charcoal Adsorbers (Contd)

Adsorbent Acceptance and Carbon Laboratory Acceptance Testing will be performed in accordance with, Inplace Leak Test Criteria and will meet the requirements of, position C.6 of R.G 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2. Adsorber Inplace Leak Testing will be performed in accordance with, and will meet the requirements of, position C.5.d of R.G. 1.52, Revision 2, with the exceptions listed in Table 6.5.1-2.

7. Demister Quantity, Per System 1 bank Air Flow acfm 6800 Max. Resistance Clean, in. wg 1.0 Max. Resistance Loaded, in. wg 2.0 Material Woven stainless steel and glass fiber mesh Amendment 61 Page 2 of 2

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-4 FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM SINGLE FAILURE ANALYSIS COMPONENT IDENTIFICATION AND METHOD OF QUANTITY FAILURE MODE EFFECT ON SYSTEM DETECTION MONITOR REMARKS Exhaust Fans (2)Fails to operate Loss of Suction Low flow alarm C.R.I** 100% capacity stand-by unit provided Exhaust Fan Inlet Valve (2) Fails to open Loss of Suction Low flow alarm C.R.I. 100% capacity stand-by unit provided Exhaust Fan Discharge Damper (2) Fails to open during Loss of Suction Low flow alarm C.R.I. 100% capacity stand-by exhaust phase unit provided Exhaust Fan Inlet Valve (2)Fails to close Reverse air flow Indicating light C.R.I. Outlet gravity damper (in same train) will close and prevent reverse flow HEPA Filter or Demister (4)Clogs Air Flow reduction Low flow alarm C.R.I. 100% capacity stand-by (2) unit provided Electrical Heating Coil (2)Fails to function Methyl iodide trapping High relative humidity C.R.I. 100% capacity stand-by efficiency may reduce alarm unit provided Diesel Generator (2)Fails to function Loss of one fan and filter Diesel generator C.R.I. 100% capacity stand-by train malfunction alarm unit provided Isolation Damper (12)Fails to close None Redundant isolation damper provided in series

- CONTROL ROOM INDICATION Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.1-5 REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM SINGLE FAILURE ANALYSIS COMPONENT IDENTIFICATION AND METHOD OF QUANTITY FAILURE MODE EFFECT ON SYSTEM DETECTION MONITOR REMARKS Exhaust Fans (2)Fails to operate Loss of Suction Low Flow Alarm C.R.I* Redundant capacity standby unit provided. Filtration train inlet valve (2)Fails to open Loss of Suction Low Flow Alarm C.R.I. Redundant capacity standby unit provided. Exhaust Fan inlet valve (2)Fails to open Loss of Suction Low Flow Alarm C.R.I. Redundant capacity standby unit provided. Exhaust Fan Discharge Damper (2)Fails to open during Loss of Suction Low Flow Alarm C.R.I. Redundant capacity exhaust phase standby unit provided. Exhaust Fan Inlet Valve (for Decay (2)Fails to close Reverse Air Flow through Low Flow Alarm for C.R.I. Gravity damper will close Cooling mode) idle fan operating systems and prevent reverse flow. HEPA filter or Demister (4)Clogs Air Flow Reduction Low Flow Alarm C.R.I. Redundant capacity (2) standby unit provided. Decay Cooling Valve in Filter Train (1)Closed and cannot be No effect. Valve position C.R.I. Forced air cooling not discharge interconnecting pipe reopened required for decay heat removal. Electric Heating Coil (2)Fails to function Methyl Iodide Trapping High relative humidity C.R.I. Redundant capacity efficiency may reduce standby unit provided. Diesel Generator(2) Fails to function Loss of one fan and filter D.G. malfunction alarm, C.R.I.* Redundant capacity train flow switch at fan standby unit provided. discharge. Isolation Damper On inlet for each Fails to close None Alarm on main Control C.R.I. Redundant isolation cubicle (34) Board. damper provided in series. On outlet for each cubicle (20)

Control Room Indication Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.2-1 IODINE REMOVAL SYSTEM COMPONENTS A - Containment Spray Additive Tank Volume, gallons 7098 Minimum Liquid Volume in tank, gallons 3268 Design Temperature, °F 200 Design Pressure, psig 15 Operating Temperature, °F 100 Operating Pressure, psig 2 Fluid 27-29% by weight sodium hydroxide Solution with nitrogen (N2) cover gas Material 304 SS Code ASME III, Code Class 3 B - Motor Operated Valves Quantity 2 Size, Inches 2 Type Globe Design Pressure, psig 50 Design Temperature, °F 200 End Connection SW Pipe Schedule 40S Material 304 SS Fluid 27-29% by weight sodium hydroxide Solution Operator Motor Code ASME III, Code Class 3 C - Eductor Quantity 2 Design Pressure, psig 300 Design Temperature, °F 300 Material 304 SS Code ASME III, Code Class 2 D - All other Valves Material 304 SS Code ASME III Code Class 2 and 3 E - Pipings and fittings are of ASME III, Code Class 2 or Class 3 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 TABLE 6.5.3-1 PRIMARY CONTAINMENT OPERATION FOLLOWING A DESIGN BASIS ACCIDENT General Type of Structure Steel-lined, reinforced concrete structure Appropriate Internal Fission Product Removal System Containment Spray System 6 3 Total Free Volume 2.344 x 10 ft 6 3 Sprayed Volume of Primary Containment 2.014 x 10 ft Methods of Hydrogen Removal Purging by the Hydrogen Purge System Time-Dependent Parameters Anticipated Conservative Leak Rate of Primary Containment Less than 0.1% of the Containment 0.1% of the Containment free volume free volume per day following a per day following a LOCA LOCA Leakage Fractions to Volumes 60% to the building 40% to the 60% to the building 40% to the Outside the Primary Containment environment environment

                                             -1                                      -1 Effectiveness of Fission Product       37.1 hr                                  20 hr Removal System (elemental iodine removal constant)

Initiation of Hydrogen Purge Not required 9 days after a LOCA (if containment pressure is reduced to atmospheric.) Hydrogen Purge Rate Not required 125 acfm Amendment 62 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-1 CONTAINMENT TEMPERATURE FOR MOST SEVERE HOT LEG BREAK (DEHLG) 6.2.1-1a SUMP TEMPERATURE FOR MOST SEVERE HOT LEG BREAK (DEHLG) 6.2.1-2 CONTAINMENT PRESSURE FOR MOST SEVERE HOT LEG BREAK (DEHLG) 6.2.1-3 CONTAINMENT PRESSURE FOR DBA (MOST SEVERE PUMP SUCTION LEG BREAK DEPSLG) MINIMUM SAFETY INJECTION 6.2.1-4 CONTAINMENT PRESSURE FOR DEPSG MAXIMUM SAFETY INJECTION 6.2.1-5a CONTAINMENT TEMPERATURE FOR DBA (MOST SEVERE PUMP SUCTION LEG BREAK - DEPSLG) MINIMUM SAFETY INJECTION 6.2.1-5b CONTAINMENT SUMP TEMPERATURE FOR DBA (MOST SEVERE PUMP SUCTION LEG BREAK - DEPSLG) MINIMUM SAFETY INJECTION 6.2.1-6a CONTAINMENT TEMPERATURE FOR DEPSG MAXIMUM SAFETY INJECTION 6.2.1-6b CONTAINMENT SUMP TEMPERATURE MOST SEVERE PUMP SUCTION BREAK-DEPSLG MAXIMUM SAFETY INJECTION 6.2.1-7 DELETED BY AMENDMENT NO. 51 6.2.1-8 DELETED BY AMENDMENT NO. 51 6.2.1-9 CONTAINMENT PRESSURE - WORST MSLB (MFIV FAILURE, 30% POWER FULL DEB) 6.2.1-10a CONTAINMENT TEMPERATURE - WORST MSLB (MFIV FAILURE, 102% POWER FULL DEB) 6.2.1-10b CONTAINMENT SUMP TEMPERATURE WORST MSLB (MFIV FAILURE, 102% POWER FULL DEB) 6.2.1-11 TAGAMI CONDENSING HEAT TRANSFER COEFFICIENT FOR DBA 6.2.1-12 UCHIDA HEAT TRANSFER COEFFICIENT-WORST MSLB 6.2.1-13 TYPICAL ENERGY DISTRIBUTION IN CONTAINMENT FOR DBA 6.2.1-14 TYPICAL TRANSIENT CONTAINMENT LINER SURFACE TEMPERATURE FOR THE CONTAINMENT TEMPERATURE DBA 6.2.1-15 PRESSURE FOLLOWING INADVERTENT SPRAY ACTUATION 6.2.1-16 CONTAINMENT FAN COOLER PERFORMANCE CURVE FOLLOWING A DBA 6.2.1-17 CONTAINMENT FAN COOLER NORMAL MODE PULLDOWN DATA FOR MINIMUM CONTAINMENT PRESSURE (VACUUM ANALYSIS) 6.2.1-18 SUBCOMPARTMENTS CONTAINMENT BUILDING - PLAN EL 221.00' & 236.00' 6.2.1-19 SUBCOMPARTMENTS CONTAINMENT BUILDING - PLAN EL 261.00' & 286.00' 6.2.1-20 SUBCOMPARTMENTS CONTAINMENT BUILDING SECTIONS A-A & B-B Amendment 63 Page 1 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-21 REACTOR CAVITY MODEL AND COORDINATE SYSTEM 6.2.1-22 REACTOR CAVITY SUBCOMPARTMENT PRESSURIZATION MODEL (FILL JUNCTIONS NOT SHOWN) 6.2.1-23 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 1 6.2.1-24 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 1 6.2.1-25 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 3 6.2.1-26 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 3 6.2.1-27 SUBCOMPARTMENT PRESSURIZATION MODEL OF PRESSURIZER COMPARTMENT AND STEAM GENERATOR/LOOP 2 6.2.1-28a PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 1 - VOL. 33 (PSID) 6.2.1-28b PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 2 - VOL. 33 (PSID) 6.2.1-28c PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 3 - VOL. 33 (PSID) 6.2.1-28d PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 4 - VOL. 33 (PSID) 6.2.1-28e PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 5 - VOL. 33 (PSID) 6.2.1-28f PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 6 - VOL. 33 (PSID) 6.2.1-28g PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 7 - VOL. 33 (PSID) 6.2.1-28h PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 8 - VOL. 33 (PSID) 6.2.1-28i PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB VOL. 9 - VOL. 33 (PSID) 6.2.1-29 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 10 - VOL. 33 (PSID) 6.2.1-30 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 11 - VOL. 33 (PSID) 6.2.1-31 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 12 - VOL. 33 (PSID) 6.2.1-32 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 13 - VOL. 33 (PSID) 6.2.1-33 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 14 - VOL. 33 (PSID) 6.2.1-34 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 15 - VOL. 33 (PSID) 6.2.1-35 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 16 - VOL. 33 (PSID) 6.2.1-36 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 17 - VOL. 33 (PSID) 6.2.1-37 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 18 - VOL. 33 (PSID) 6.2.1-38 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 19 - VOL. 33 (PSID) Amendment 63 Page 2 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-39 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 20 - VOL. 33 (PSID) 6.2.1-40 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 21 - VOL. 33 (PSID) 6.2.1-41 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 22 - VOL. 33 (PSID) 6.2.1-42 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 23 - VOL. 33 (PSID) 6.2.1-43 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 24 - VOL. 33 (PSID) 6.2.1-44 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 25 - VOL. 33 (PSID) 6.2.1-45 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 26 - VOL. 33 (PSID) 6.2.1-46 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 27 - VOL. 33 (PSID) 6.2.1-47 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 28 - VOL. 33 (PSID) 6.2.1-48 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 29 - VOL. 33 (PSID) 6.2.1-49 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 30 - VOL. 33 (PSID) 6.2.1-50 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 31 - VOL. 33 (PSID) 6.2.1-51 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG VOL. 32 - VOL. 33 (PSID) 6.2.1-52a PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 1 - VOL. 33 (PSID) 6.2.1-52b PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 2 - VOL. 33 (PSID) 6.2.1-52c PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 3 - VOL. 33 (PSID) 6.2.1-52d PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 4 - VOL. 33 (PSID) 6.2.1-52e PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 5 - VOL. 33 (PSID) 6.2.1-52f PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 6 - VOL. 33 (PSID) 6.2.1-52g PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 7 - VOL. 33 (PSID) 6.2.1-52h PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 8 - VOL. 33 (PSID) 6.2.1-52i PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB VOL. 9 - VOL. 33 (PSID) 6.2.1-53 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 10 - VOL. 33 (PSID) 6.2.1-54 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 11 - VOL. 33 (PSID) 6.2.1-55 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 12 - VOL. 33 (PSID) 6.2.1-56 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 13 - VOL. 33 (PSID) 6.2.1-57 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 14 - VOL. 33 (PSID) Amendment 63 Page 3 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-58 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 15 - VOL. 33 (PSID) 6.2.1-59 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 16 - VOL. 33 (PSID) 6.2.1-60 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 17 - VOL. 33 (PSID) 6.2.1-61 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 18 - VOL. 33 (PSID) 6.2.1-62 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 19 - VOL. 33 (PSID) 6.2.1-63 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 20 - VOL. 33 (PSID) 6.2.1-64 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 21 - VOL. 33 (PSID) 6.2.1-65 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 22 - VOL. 33 (PSID) 6.2.1-66 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 23 - VOL. 33 (PSID) 6.2.1-67 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 24 - VOL. 33 (PSID) 6.2.1-68 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 25 - VOL. 33 (PSID) 6.2.1-69 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 26 - VOL. 33 (PSID) 6.2.1-70 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 27 - VOL. 33 (PSID) 6.2.1-71 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 28 - VOL. 33 (PSID) 6.2.1-72 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 29 - VOL. 33 (PSID) 6.2.1-73 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 30 - VOL. 33 (PSID) 6.2.1-74 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 31 - VOL. 33 (PSID) 6.2.1-75 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG VOL. 32 - VOL. 33 (PSID) 6.2.1-76 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 5 - VOL. 18 (PSID) 6.2.1-77 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 10 - VOL. 18 (PSID) 6.2.1-78 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 12 - VOL. 10 (PSID) 6.2.1-79 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 12 - VOL. 16 (PSID) 6.2.1-80 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 12 - VOL. 17 (PSID) 6.2.1-81 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 12 - VOL. 35 (PSID) 6.2.1-82 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 21 - VOL. 11 (PSID) 6.2.1-83 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 21 - VOL. 13 (PSID) 6.2.1-84 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 21 - VOL. 18 (PSID) Amendment 63 Page 4 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-85 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 22 - VOL. 11 (PSID) 6.2.1-86 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 22 - VOL. 13 (PSID) 6.2.1-87 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 22 - VOL. 17 (PSID) 6.2.1-88 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 22 - VOL. 21 (PSID) 6.2.1-89 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 22 - VOL. 24 (PSID) 6.2.1-90 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 23 - VOL. 13 (PSID) 6.2.1-91 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 23 - VOL. 18 (PSID) 6.2.1-92 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 23 - VOL. 21 (PSID) 6.2.1-93 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 23 - VOL. 25 (PSID) 6.2.1-94 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 24 - VOL. 13 (PSID) 6.2.1-95 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 24 - VOL. 16 (PSID) 6.2.1-96 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 24 - VOL. 17 (PSID) 6.2.1-97 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 24 - VOL. 23 (PSID) 6.2.1-98 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 24 - VOL. 32 (PSID) 6.2.1-99 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 25 - VOL. 13 (PSID) 6.2.1-100 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 25 - VOL. 18 (PSID) 6.2.1-101 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 26 - VOL. 13 (PSID) 6.2.1-102 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 26 - VOL. 18 (PSID) 6.2.1-103 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 27 - VOL. 11 (PSID) 6.2.1-104 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 27 - VOL. 18 (PSID) 6.2.1-105 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 27 - VOL. 33 (PSID) 6.2.1-106 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 28 - VOL. 11 (PSID) 6.2.1-107 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 28 - VOL. 17 (PSID) 6.2.1-108 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 28 - VOL. 27 (PSID) 6.2.1-109 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 28 - VOL. 30 (PSID) 6.2.1-110 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 28 - VOL. 34 (PSID) 6.2.1-111 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 29 - VOL. 18 (PSID) Amendment 63 Page 5 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-112 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 29 - VOL. 27 (PSID) 6.2.1-113 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 29 - VOL. 31 (PSID) 6.2.1-114 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 29 - VOL. 35 (PSID) 6.2.1-115 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 30 - VOL. 12 (PSID) 6.2.1-116 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 30 - VOL. 16 (PSID) 6.2.1-117 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 30 - VOL. 17 (PSID) 6.2.1-118 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 30 - VOL. 29 (PSID) 6.2.1-119 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 30 - VOL. 32 (PSID) 6.2.1-120 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 31 - VOL. 5 (PSID) 6.2.1-121 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 31 - VOL. 18 (PSID) 6.2.1-122 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 32 - VOL. 10 (PSID) 6.2.1-123 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 32 - VOL. 18 (PSID) 6.2.1-124 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 33 - VOL. 11 (PSID) 6.2.1-125 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 33 - VOL. 18 (PSID) 6.2.1-126 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 34 - VOL. 3 (PSID) 6.2.1-127 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 34 - VOL. 11 (PSID) 6.2.1-128 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 34 - VOL. 12 (PSID) 6.2.1-129 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 34 - VOL. 33 (PSID) 6.2.1-130 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 35 - VOL. 5 (PSID) 6.2.1-131 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 35 - VOL. 18 (PSID) 6.2.1-132 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-1 VOL. 35 - VOL. 33 (PSID) 6.2.1-133 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 13 - VOL. 7 (PSID) 6.2.1-134 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 14 - VOL. 13 (PSID) 6.2.1-135 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 15 - VOL. 7 (PSID) 6.2.1-136 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 15 - VOL. 12 (PSID) 6.2.1-137 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 15 - VOL. 14 (PSID) 6.2.1-138 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 16 - VOL. 7 (PSID) Amendment 63 Page 6 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-139 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 16 - VOL. 14 (PSID) 6.2.1-140 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 16 - VOL. 15 (PSID) 6.2.1-141 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 16 - VOL. 18 (PSID) 6.2.1-142 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 17 - VOL. 7 (PSID) 6.2.1-143 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 17 - VOL. 12 (PSID) 6.2.1-144 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 17 - VOL. 14 (PSID) 6.2.1-145 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 17 - VOL. 15 (PSID) 6.2.1-146 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 17 - VOL. 18 (PSID) 6.2.1-147 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 17 - VOL. 19 (PSID) 6.2.1-148 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 18 - VOL. 7 (PSID) 6.2.1-149 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 18 - VOL. 8 (PSID) 6.2.1-150 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 19 - VOL. 6 (PSID) 6.2.1-151 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 19 - VOL. 7 (PSID) 6.2.1-152 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 19 - VOL. 8 (PSID) 6.2.1-153 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 19 - VOL. 12 (PSID) 6.2.1-155 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 20 - VOL. 8 (PSID) 6.2.1-156 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 20 - VOL. 14 (PSID) 6.2.1-157 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 20 - VOL. 18 (PSID) 6.2.1-158 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 20 - VOL. 19 (PSID) 6.2.1-159 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 22 - VOL. 8 (PSID) 6.2.1-160 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 21 - VOL. 12 (PSID) 6.2.1-161 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 21 - VOL. 27 (PSID) 6.2.1-162 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 22 - VOL. 21 (PSID) 6.2.1-163 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 22 - VOL. 24 (PSID) 6.2.1-164 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 22 - VOL. 28 (PSID) 6.2.1-165 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 23 - VOL. 12 (PSID) 6.2.1-166 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 23 - VOL. 21 (PSID) Amendment 63 Page 7 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-167 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 23 - VOL. 24 (PSID) 6.2.1-168 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 23 - VOL. 25 (PSID) 6.2.1-169 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 23 - VOL. 29 (PSID) 6.2.1-170 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 24 - VOL. 8 (PSID) 6.2.1-171 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 24 - VOL. 30 (PSID) 6.2.1-172 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 25 - VOL. 6 (PSID) 6.2.1-173 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 25 - VOL. 8 (PSID) 6.2.1-174 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 25 - VOL. 12 (PSID) 6.2.1-175 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 25 - VOL. 31 (PSID) 6.2.1-176 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 26 - VOL. 8 (PSID) 6.2.1-177 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 26 - VOL. 24 (PSID) 6.2.1-178 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 26 - VOL. 25 (PSID) 6.2.1-179 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 26 - VOL. 32 (PSID) 6.2.1-180 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 27 - VOL. 12 (PSID) 6.2.1-181 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 27 - VOL. 28 (PSID) 6.2.1-182 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 27 - VOL. 29 (PSID) 6.2.1-183 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 28 - VOL. 9 (PSID) 6.2.1-184 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 28 - VOL. 30 (PSID) 6.2.1-185 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 29 - VOL. 12 (PSID) 6.2.1-186 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 29 - VOL. 31 (PSID) 6.2.1-187 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 30 - VOL. 7 (PSID) 6.2.1-188 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 30 - VOL. 9 (PSID) 6.2.1-189 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 30 - VOL. 29 (PSID) 6.2.1-190 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 30 - VOL. 32 (PSID) 6.2.1-191 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 30 - VOL. 5 (PSID) 6.2.1-192 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 31 - VOL. 6 (PSID) 6.2.1-193 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 31 - VOL. 12 (PSID) Amendment 63 Page 8 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-194 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 32 - VOL. 5 (PSID) 6.2.1-195 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-3 VOL. 32 - VOL. 31 (PSID) 6.2.1-196 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 3 - VOL. 16 (PSID) 6.2.1-197 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 3 - VOL. 10 (PSID) 6.2.1-198 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 20 - VOL. 1 (PSID) 6.2.1-199 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 20 - VOL. 3 (PSID) 6.2.1-200 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 21 - VOL. 1 (PSID) 6.2.1-201 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 22 - VOL. 1 (PSID) 6.2.1-202 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 23 - VOL. 1 (PSID) 6.2.1-203 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 24 - VOL. 1 (PSID) 6.2.1-204 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 24 - VOL. 11 (PSID) 6.2.1-205 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 24 - VOL. 17 (PSID) 6.2.1-206 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 24 - VOL. 19 (PSID) 6.2.1-207 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 25 - VOL. 1 (PSID) 6.2.1-208 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 25 - VOL. 11 (PSID) 6.2.1-209 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 25 - VOL. 19 (PSID) 6.2.1-210 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 25 - VOL. 24 (PSID) 6.2.1-211 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 25 - VOL. 26 (PSID) 6.2.1-212 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 25 - VOL. 27 (PSID) 6.2.1-213 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 26 - VOL. 1 (PSID) 6.2.1-214 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 26 - VOL. 17 (PSID) 6.2.1-215 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 26 - VOL. 19 (PSID) 6.2.1-216 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 26 - VOL. 24 (PSID) 6.2.1-217 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 26 - VOL. 27 (PSID) 6.2.1-218 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 27 - VOL. 1 (PSID) 6.2.1-219 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 27 - VOL. 11 (PSID) 6.2.1-220 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 27 - VOL. 17 (PSID) Amendment 63 Page 9 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-221 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 28 - VOL. 11 (PSID) 6.2.1-222 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 28 - VOL. 17 (PSID) 6.2.1-223 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 28 - VOL. 24 (PSID) 6.2.1-224 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 28 - VOL. 32 (PSID) 6.2.1-225 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 29 - VOL. 11 (PSID) 6.2.1-226 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 29 - VOL. 25 (PSID) 6.2.1-227 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 29 - VOL. 28 (PSID) 6.2.1-228 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 29 - VOL. 30 (PSID) 6.2.1-229 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 29 - VOL. 31 (PSID) 6.2.1-230 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 29 - VOL. 33 (PSID) 6.2.1-231 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 30 - VOL. 17 (PSID) 6.2.1-232 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 30 - VOL. 26 (PSID) 6.2.1-233 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 30 - VOL. 28 (PSID) 6.2.1-234 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 30 - VOL. 31 (PSID) 6.2.1-235 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 30 - VOL. 34 (PSID) 6.2.1-236 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 31 - VOL. 11 (PSID) 6.2.1-237 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 31 - VOL. 17 (PSID) 6.2.1-238 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 31 - VOL. 27 (PSID) 6.2.1-239 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 31 - VOL. 35 (PSID) 6.2.1-240 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 32 - VOL. 10 (PSID) 6.2.1-241 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 32 - VOL. 16 (PSID) 6.2.1-242 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 33 - VOL. 10 (PSID) 6.2.1-243 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 33 - VOL. 34 (PSID) 6.2.1-244 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 33 - VOL. 35 (PSID) 6.2.1-245 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 34 - VOL. 16 (PSID) 6.2.1-246 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 34 - VOL. 35 (PSID) 6.2.1-247 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 35 - VOL. 10 (PSID) Amendment 63 Page 10 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-248 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP-2 VOL. 35 - VOL. 16 (PSID) 6.2.1-249 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 3 - VOL. 10 (PSID) 6.2.1-250 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 3 - VOL. 16 (PSID) 6.2.1-251 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 20 - VOL. 1 (PSID) 6.2.1-252 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 20 - VOL. 3 (PSID) 6.2.1-253 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 21 - VOL. 1 (PSID) 6.2.1-254 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 22 - VOL. 1 (PSID) 6.2.1-255 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 23 - VOL. 1 (PSID) 6.2.1-256 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 24 - VOL. 1 (PSID) 6.2.1-257 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 24 - VOL. 11 (PSID) 6.2.1-258 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 24 - VOL. 17 (PSID) 6.2.1-259 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 24 - VOL. 19 (PSID) 6.2.1-260 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 25 - VOL. 1 (PSID) 6.2.1-261 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 25 - VOL. 11 (PSID) 6.2.1-262 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 25 - VOL. 19 (PSID) 6.2.1-263 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 25 - VOL. 24 (PSID) 6.2.1-264 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 25 - VOL. 26 (PSID) 6.2.1-265 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 25 - VOL. 27 (PSID) 6.2.1-266 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 26 - VOL. 1 (PSID) 6.2.1-267 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 26 - VOL. 17 (PSID) 6.2.1-268 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 26 - VOL. 19 (PSID) 6.2.1-269 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 26 - VOL. 24 (PSID) 6.2.1-270 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 26 - VOL. 27 (PSID) 6.2.1-271 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 27 - VOL. 1 (PSID) 6.2.1-272 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 27 - VOL. 11 (PSID) 6.2.1-273 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 27 - VOL. 17 (PSID) 6.2.1-274 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 28 - VOL. 11 (PSID) Amendment 63 Page 11 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-275 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 28 - VOL. 17 (PSID) 6.2.1-276 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 28 - VOL. 24 (PSID) 6.2.1-277 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 28 - VOL. 32 (PSID) 6.2.1-278 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 28 - VOL. 24 (PSID) 6.2.1-279 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 29 - VOL. 25 (PSID) 6.2.1-280 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 29 - VOL. 28 (PSID) 6.2.1-281 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 29 - VOL. 30 (PSID) 6.2.1-282 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 29 - VOL. 31 (PSID) 6.2.1-283 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 29 - VOL. 33 (PSID) 6.2.1-284 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 30 - VOL. 17 (PSID) 6.2.1-285 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 30 - VOL. 26 (PSID) 6.2.1-286 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 30 - VOL. 28 (PSID) 6.2.1-287 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 30 - VOL. 31 (PSID) 6.2.1-288 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 30 - VOL. 34 (PSID) 6.2.1-289 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 31 - VOL. 11 (PSID) 6.2.1-290 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 31 - VOL. 17 (PSID) 6.2.1-291 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 31 - VOL. 27 (PSID) 6.2.1-292 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 31 - VOL. 35 (PSID) 6.2.1-293 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 32 - VOL. 10 (PSID) 6.2.1-294 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 32 - VOL. 16 (PSID) 6.2.1-295 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 33 - VOL. 10 (PSID) 6.2.1-296 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 33 - VOL. 34 (PSID) 6.2.1-297 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 33 - VOL. 35 (PSID) 6.2.1-298 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 34 - VOL. 16 (PSID) 6.2.1-299 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 34 - VOL. 35 (PSID) 6.2.1-300 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 35 - VOL. 10 (PSID) 6.2.1-301 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT VOL. 35 - VOL. 16 (PSID) Amendment 63 Page 12 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.1-302 DELETED BY AMENDMENT NO. 46 6.2.1-303 HEAT REMOVAL RATE OF EMERGENCY COOLER UNIT 6.2.1-304 DELETED BY AMENDMENT NO. 46 6.2.1-305 DELETED BY AMENDMENT NO. 46 6.2.1-306 CONTAINMENT VACUUM RELIEF SYSTEM 6.2.1-307 FORCES ON THE REACTOR VESSEL COLD LEG NOZZLE 150 IN2 BREAK 6.2.1-308 MOMENTS ON THE REACTOR VESSEL COLD LEG NOZZLE 150 IN2 BREAK 6.2.2-1 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-2 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-3 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-4 CONTAINMENT FAN COOLER PERFORMANCE CURVE 6.2.2-5 DELETED BY AMENDMENT NO. 48 6.2.2-6 CONTAINMENT SPRAY NOZZLE DROP SIZE HISTOGRAM 6.2.2-7 CONTAINMENT SUMP PLAN 6.2.2-8 CONTAINMENT SUMP SECTION "A-A" 6.2.2-9 CONTAINMENT SUMP SECTION "B-B" 6.2.2-10 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-11 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-12 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-13 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-14 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-15 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-16 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-17 CONTAINMENT SPRAY PUMP PERFORMANCE CURVE 6.2.2-18 RESIDUAL HEAT REMOVAL PUMP PERFORMANCE CURVE 6.2.2-19 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.2.2-20 CONTAINMENT BUILDING RECIRCULATION SUMP STRAINER ISOMETRIC Amendment 63 Page 13 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE TITLE 6.2.5-1 DELETED BY AMENDMENT NO. 62 6.2.5-2 DELETED BY AMENDMENT NO. 62 6.2.5-3 DELETED BY AMENDMENT NO. 62 6.2.5-4 DELETED BY AMENDMENT NO. 62 6.2.5-5 DELETED BY AMENDMENT NO. 58 6.2.5-6 DELETED BY AMENDMENT NO. 62 6.2.5-7 POST ACCIDENT HYDROGEN MONITORING SYSTEM 6.2A-1 TEMPERATURE GRADIENT IN GASEOUS & LIQUID BOUNDARY LAYERS DURING HEAT SINK SURFACE CONDENSATION 6.2A-2 SPRAY EFFICIENCY VS STEAM/AIR RATIO 6.3.2-1 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.3.2-2 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.3.2-3 REFER TO FSAR TABLE 1.6-3 FOR DESIGN DOCUMENT INCORPORATED BY REFERENCE 6.3.2-4 EMERGENCY CORE COOLING SYSTEM PROCESS FLOW DIAGRAM SHEET 1 6.3.2-5 EMERGENCY CORE COOLING SYSTEM PROCESS FLOW DIAGRAM SHEET 2 6.3.2-6 EMERGENCY CORE COOLING SYSTEM PROCESS FLOW DIAGRAM SHEET 3 6.3.2-7 DELETED BY AMENDMENT NO. 27 6.3.2-8 RHR PUMP PERFORMANCE CURVE 6.3.2-9 CHG PUMP PERFORMANCE CURVE 6.4.2-1 DELETED BY AMENDMENT NO. 15 6.5.2-1 DELETED BY AMENDMENT NO. 51 6.5.2-2 CONTAINMENT SPRAY PH TIME HISTORY OF CONTAINMENT SUMP & SPRAY CASE 1 6.5.2-3 CONTAINMENT SPRAY PH TIME HISTORY OF CONTAINMENT SUMP & SPRAY CASE 2 6.5.2-4 DELETED BY AMENDMENT NO. 27 6.5.2-5 DELETED BY AMENDMENT NO. 27 6.5.2-6 DELETED BY AMENDMENT NO. 27 Amendment 63 Page 14 of 14

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-1 CONTAINMENT TEMPERATURE FOR MOST SEVERE HOT LEG BREAK (DEHLG) Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-1A SUMP TEMPERATURE FOR MOST SEVERE HOT LEG BREAK (DEHLG) Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-2 CONTAINMENT PRESSURE FOR MOST SEVERE HOT LEG BREAK (DEHLG) Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-3 CONTAINMENT PRESSURE FOR DBA (MOST SEVERE PUMP SUCTION LEG BREAK - DEPSLG) MINIMUM SAFETY INJECTION Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-4 CONTAINMENT PRESSURE FOR DEPSG MAXIMUM SAFETY INJECTION Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-5A CONTAINMENT TEMPERATURE FOR DBA (MOST SEVERE PUMP SUCTION LEG BREAK - DEPSLG) MINIMUM SAFETY INJECTION Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-5B CONTAINMENT SUMP TEMPERATURE FOR DBA (MOST SEVERE PUMP SUCTION LEG BREAK - DEPSLG) MINIMUM SAFETY INJECTION Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-6A CONTAINMENT TEMPERATURE FOR DEPSG MAXIMUM SAFETY INJECTION Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-6B CONTAINMENT SUMP TEMPERATURE MOST SEVERE PUMP SUCTION BREAK - DEPSLG MAXIMUM SAFETY INJECTION Amendment 63 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-9 CONTAINMENT PRESSURE - WORST MSLB (MFIV FAILURE, 30% POWER FULL DEB) CP&L HARR]S NP MAXIMUM PRESSURE: 0£ MSLB 307. POHER. COOLINQ

                     ~...-- --.---~-----..-- ---;-----=------:- ---:-------..:---~

0 I i * . * , S* .................. ,..................r***************:***.***: ********Y ...**********i--*****... ******,**************.. ***:***********....**:--....... .* **** 0 . I i ! i j' ; . f : 5i ..................r***************r**************::****..******..****:****************:******** ********** O a: l*****. . . . .....\. . . . . . . . . l.. . . ... . . ..l............... ... l., ... . .... .... t ..........******-;-***........ ... . i

                     $i- ................. r-*****...........

Vlo 1 . * *

                 ~
                 ~                                                   :                  :                  :                                       :
                 ~                                                                      ;                  :                                       ~
                             . ************r****. **T*** . *****-r*******..... :                               *********.. r**** ***+************+***** . ... .............

U') C

~~

9

                     $i:;I * .... .._. .. .... . "'T****......... ""T .................T"" .............T... ............*i....... .. ............!. ..... . ... . . 1, *****" ..'"..... :**~----.... ' ....
                                                    =                =                  *

1 I t

                                                   .                                                       j
                     ~

0 . 1-----ir--- 20.a - r- - ---r- -- -,------,-----;----j----...;.,---1 0.0 iO.a GO.O 00.0 100.0 120.n 110 . 0 160.0 100 , 0 Amendment 61 TIME ISE:CI Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-10A CONTAINMENT TEMPERATURE - WORST MSLB (MFIV FAILURE, 102% POWER FULL DEB) Cf->IL HARR I S NP MAXI NUM TEMPE~flil!RE tlSL8 102Z fl V fA I LURf,_PJ!

                        §---,-----------:-- ----:------:--- - ,--- ____,. _ _ ______

0 i 0 Q

                           .                                    T* .................. .*               !. ........,. . . . ,*******.. . . .. :*..... . .. .. . .*T***
                                                                                                                                                                                                                   *           ... ......****r**. ...... ...... .
                        ;,i; ... . . . ...... . . .. ..                                                                                                                                                                                                                                              ....... , .......

I Cl i:. § * **:**1*****... .....

                                                                 !. .................\' .......... . . .'j".................!................ **(***. . ............ ]. ,........ . . :
                                                                                                       *                                                                        *                                  *                                      ~

1

-~-

(.!) = i

                 ~                                                                                                                               ..
                                                             *i*;..***                                  .

0

                 ~           ....                                                                                                                 .                             :                                                                          ;                                :~
                                                     * **                                         * * *~ * -I ii, * ~+ , ,* *

r* ~.* * * *! *-! -I -t, * ~

* *

1 * ** * * **** * * ** * * * *

* *

I** L L * *: *
I I I* ****
I 1 ,
p * * : *** I . * * * * *
r I , , *
I L *: I * ,i *
i,,,
L * , , I*
I

)

I * *

                                                                                                                                                                                                                   ~                                                                          '

1-c::: a:: w .. \.i_-----+--- ~----------+----- llR

~C!

w 8-

                 ,..;.. N
                                ** j

L * * * ,-1 *-I **

                                                              . *:

0 0 * ** * ** L L L,j. * * * .

' * * *

r
I ~ .. .. ..

                                                                                                                                         *- -t* *-********-                    ..
                                                                                                                                                                  . . ... ... 1 . . .. .. . . .. .... ,            r*. ... . . . .... T. .......           .
                        ~- .. ................( ..................\................. .1........ ....... *-l ***..... ** *** .,; .. ****** *..... ) .......... .. .

l , I r : i Cl 0-1-----,-----,-----r-a.a - -- ;--- --j- - -- +-- - -;r--- - -i-- - - -l 20.0 iO. O 60, 0 eo.o 100 ,0 120 .0 110. 0 160 .0 IM , O Amendment 61 TIME: (S[CI Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-10B CONTAINMENT SUMP TEMPERATURE WORST MSLB (MFIV FAILURE, 102% POWER FULL DEB) CP&L HARRIS NP MAXIMUM TENPERATURE MSLB 102% flV fAILURE PU a

                       § --r----~-----:------,----~--:------,-------:-----,------.

0

                       §-******************r-*************r-************** r*************:** **************:****************
                   .: ~- .... :.............. l**************.. **t*****.............1...................!............. ....!.................

a : : : :. . (!) ! : : : : w ; 0~ . !

                   ~8 ................... :.................. -r .. ******"........ ,..................:.............. .

i=! M a: l l : i 0:: w a..o

c . . 'J.*** * . -:---..;._---.l_-~--iTB

                   ~~
                                                          !                                  I                                                                                                            .
                              * ** * * *' * +* O* o

I * *; *
0
oo 11'1 * '"* I** ol 0 *+I
I * * , . o+ * * * ~ * * . ** *+* ** * * * * '* * *: * * * * * " * * *O
o.,: '* o o * * " * ' " ' . * *
t * ' * * * * ** * * * * *. *. * * * * * * ' **+ *
I " * * ' "' * ' . ** . + *

                       ~~*  ...................... .       ~: . . . . . . . .............. . :.i: . .... . . . . . . . . ... . . . .. . .
                                                                                                                                                                             * ** ** ** ** +* * *
                                                                                                                                                                          ~

l 1

                       ~                                  !

Q+-----;-----;.----;-----;-------;-----;.----;----..;...----1 [ 0.0 20.0 10.0 60.0 80.0 100.0 120.0 UO.D 160,0 180 . 0 TIME ISECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-11 TAGAMI CONDENSING HEAT TRANSFER COEFFICIENT FOR DBA CP&L HARRIS NP MAXIMUM TENPERATURE NSLB 102% flV fAILURC PU 0

                        §--r ----- ~--- ---= ---~ --~- ---,. ----, ----- ,---- ---.

Cl

                        ~ .............. ......;................... \............... ... +******************l**.... ********.. *; ................
                                                                          ~
                                                  ~                                              ~

l i

                                                                                                                ***t,.. .... *........ **t...

o i

                   <<::: ~ * ....:... *.. ........ \................ ***(****.... *********1".............

(!) : i : : w

l. ..... .. . . . T:.............. .. T............, .

00

                   ~~        ... . . . .. . ...l. . . . . . ... .T
                              ** ** ***r. *****... T-*                t.----r:-- --;- --+-,-~_ _j__ __;_ ~TB 0::

w

                                                                                 * * *, ** * *****- T * * * *** ****** *** **-*-**** **** * * **

Q. 0 i::

w8 J-N g ,.,... . . .... .*****:*f ********. . ........~j.............. . .....: ...,***********.-~.-,.......*****---~.-.-******. . . ...

0 l 1 d+----...;-l___--il-- --;-- ---;- ----...;---- -i--- -.--- --i-- --~ o.o 20.0 10.0 60.0 80 . D 100.0 120.0 110.0 160 ,0 180.0 TIME ISECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-12 UCHIDA HEAT TRANSFER COEFFICIENT - WORST MSLB 100 80 U: 0 I a:

c I

N 60

                          --In I-u.
                          -f:
                           ~

c 40 20 20 60 100 140 180 220 260 300 TIME (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-13 TYPICAL ENERGY DISTRIBUTION IN CONTAINMENT FOR DBA 900 800 LEOl!.ND: 700 - ENlROY REMOVED 8Y PAN COOLEIII

                                            - - *-* lHEROY ITEAM-AIII MIXTUIII
                                            - - l!NEIIGY IN SUMl'WATEII IINCLUDES THE EFFECT OF SPRAY) 600           - * - ENl!.IIGV IN HEAT SINKS
                                            -         INEROY REMOVID l'I' RHII lflEAT EXCHANGERS 600
                         ~
                        ..,a:

Cl 400 zIll 300 200

                                                            ,-- -----...___           -.. ... ____ ---*------.l     / .....      -

__,* /*,

                                                          ~
                                                                                                           / ,.,,.*
                                                                                                                        ~
                                              ., ~'                                                      ,_,,,,,.
                                                                                        ------~;:>

100 ~ 0

                                                --------:;-*-- .........__.._,,,,,,~ -----*

0 10 1 1ol 1a' TIME ISECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-14 TYPICAL TRANSIENT CONTAINMENT LINER SURFACE TEMPERATURE FOR THE CONTAINMENT TEMPERATURE DBA 250 240 220 L,_

                          ~200 w

0:::

)

f- 180

<(

0::: w 0.... 160 2 w f-140 120 100 1 10 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-15 PRESSURE FOLLOWING INADVERTENT SPRAY ACTUATION 16.00, 15.20 lr-------------------------

                                  \
                                                                                                       ----AA      B 14.40 13.60
                    ";i'                                 -.                                                   - CONTAINMENT
                    ~
                    .E:.                                        ~~
                     "'cc  12.80
                     =>

t:i w cc 12.00 11.20 10.40 9.60 ,___ _ _...___ __ .....,____ _ _ _ _ _ _......__ __ _ _ _ _.......__ _ __,__ _ _........_ _ __,___ _ _...... 0.00 80.00 100,QQ :320 ,00 400.00 480.00 560 ,00 640.00 720 .00 800.00 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-16 CONTAINMENT FAN COOLER PERFORMANCE CURVE FOLLOWING A DBA DESIGN CONDITIONS: Al FOULING FACTOR .001 Bl 31,250 CFM LEAVING AIR FLOW Cl 1 360 GPM WATER FLOW RAT£ LL. DI WATER TEMPERATURE 950f D El OBA MODE UJ a:

, CONDENSATE FLOW RATE/UNIT lLBM/HR X 10 3 >

l-et a: 0 10 20 30 40 50 60 10 UJ

                              ~      280                                                                  280 I.LI 1--               CONDENSATE FLOW RATE z            -    HEAT REMOVAL RATE Cl
                              ~      240 t-----lf-----4------+-----+-----::.,,-,=.+-----t------l          240 a:

J 1-

                               <C V)
                               ~     200 r---------1-----t-----,-.~"'-------t------t------t-----~ 200 UI

c Ill..

                              'l/)

C>

                              ..:::E
                              ~

t-160 t--'------l'--r~---t------+-----+----+----+-----1 160

                              .z
                              ~
                              .z ct
                              ~      120 O                                                                120 10         20        30         <10       50       60 70
                              *c..>

HEAT REMOVAL RATE/UNlT ~BTU/HR X 10 6 ) HEAT REMOVAL RATE USED IN THE ANALYSIS WAS REDUCED BASED ON 1300 CPM FLOW RATE Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-17 CONTAINMENT FAN COOLER NORMAL MODE PULLDOWN DATA FOR MINIMUM CONTAINMENT PRESSURE (VACUUM ANALYSIS) 160 140 16.6@ 60% R.H. 120 u. 0 I w a::

, 100

                 !ia::

w Q..

t w

I- 80 z w

z

<(

I- 60 z 1.197 @ 0 (.) 100% R.H. Al 0 FOULING 40 B} 33°F WATER T EMPERATUR E C) 125,000 CFM AIR FLOW D) 1,500 GPM WATER FLOW 20 Amendment 61 0"---------....1...----------'-------------' 5 ' 10

                                                                                 -- ---------'20 15                              Page 1 of 1 HEAT TRANSFER - BTUH x 106

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-18 SUBCOMPARTMENTS CONTAINMENT BUILDING - PLAN EL 221.00 & 236.00 Security-Related Information - Figure Withheld Under 10 CFR 2.390 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-19 SUBCOMPARTMENTS CONTAINMENT BUILDING - PLAN EL 261.00 & 286.00 Security-Related Information - Figure Withheld Under 10 CFR 2.390 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-20 SUBCOMPARTMENTS CONTAINMENT BUILDING SECTIONS A-A & B-B Security-Related Information - Figure Withheld Under 10 CFR 2.390 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-21 REACTOR CAVITY MODEL AND COORDINATE SYSTEM z t r V0L33 EL 286'0"' SECTSON A*A f .- -.- -.- -- .- V0L32 VOL 27 TO VOL 31 IA A El268' 116/16::- .. . . EL 260' 2,7/18" f

                                                                            -i---*~ *x-v       PLANE (EL. 253. 75')

f EL 249'13' EL 242' 3-3i4 VOL 15 TO VOL 20 .y X . SECTION EHl SECTIONC-C VOL .9 TO VOL 14 (EL, 253. 75' ) VOL3TOV0L8 ______ _____ ___~------ VOL2 VOL 1

                           ._                   __.               ....._           EL :211' 6" PRO Fl LE v :I Ew

{NOT TO SCALE)

SE.CT ION 0 -D SECTLON E*E r

{SE.E FIGURE 62.1-20 FOR ACTUAL PLAN VIEWS Amendment 61 COMPARTMENT DIMENSIONS) (NOT TO SCALE) Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-22 REACTOR CAVITY SUBCOMPARTMENT PRESSURIZATION MODEL (FILL JUNCTIONS NOT SHOWN) VOL33 J.1 J.2 J-3 J-4 J-5 J-6 CONNECTS TO VOL 32 ... J-37 VOL27 , J-42 VOL 28 ,~7 VOL 29 ,*J-52 VOL 30 t J-57 VO L31 ,tJ-62 VOL 32 J-37

                                                                                                                           ** CONNECTS TO VOL 27 J.J           J:.8          J:..9         J.jO           J.: 11         J.:12 CONNECTS TO VOL 26
                                 -    tJ.38 VOL 21 , J-43 VOL 22 ,J-48 VOL 23 i 68  J:._13  !69 J:14     ~70 J.15 J-53 VOL24 ,J-58 VOL 25, J-63 VOL 26 t11   J-16 J-72   J-17 J-73   J-18 J.38
                                                                                                                           ** CONNECTS TO VOL 21 CONNECTSTO VOL 20  _, *J.39 VOL 15    *J-44 VOL 16 tJ-49 VOL 17     J-54 VOL 18 J.39 J-59 VOL 19 ' J-64 VOL 20 ** CONNECTS TO VO L 15 I

J-19 J:20 J:21 J.;.22 J:..23 J;.24 J-40 CONNECTS TO VOL 14 ,*J-40 VOL 9 *J-45 VOL 10 J-50VOL11 J-55 VOL 12 *J-60 VOL 13 1 J-65 VOL 14 - CONNECTS TO VOL 9 J-_25 J-26 J-27 J-28 J-29 J-3D CONNECTS TO VOL 8 ..., J-41 VOL 3 1 J-46 VOL 4 J-5 1 VOL 5

J-56 VOL 6 *J-61 VOL 7 lJ-66 VOL8 J-41

                                                                                                                            - CONNECTS TO VOL 3 J.31           J-32          J:33          JJ4          J-~5            J)6                           JUN-74
                                              .             -                                                        -                           ~

VOL2 J-67 VOL1 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-23 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 1

                                                                                        -----x
                                                                                     +_J
                                                                        -----~-1---=-

LEGEND: 6 JUNCTION NUMBER QvoLUME NUMBER Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-24 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 1 LEGEND, 6 JUNCTION NUMBER QvoLUME NUM!ER Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-25 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 3 LEGE.,O: 6 JUNCTION NUMIIER Qvou.JME lolUMBER Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-26 SUBCOMPARTMENT PRESSURIZATION MODEL OF STEAM GENERATOR/LOOP 3 LEGEND, 6JUNCTION NUMBER QvoLUME NUMBER Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-27 SUBCOMPARTMENT PRESSURIZATION MODEL OF PRESSURIZER COMPARTMENT AND STEAM GENERATOR/LOOP 2 Lt GEND: 6 JUNCTION NUMBE R QvoLUME NUMBER Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28A PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB 2.4-t6

    ,0
    ~

1;2

    ~
     ..J
    -§?

Q l o.a:

I

   - ..i 0

0.4 0.0

         -0.4L--    - -. . J - - - ---L.---......l'---- --.L....--- ~ _;i_- --_,.I 0.,0,    0;2:          0.4        0.6-        0.8       1.0        1.2:

TIME (SEC} Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28B PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB 2,4.,

2.0 1.6 C l,

     '1J)
     ~     1:.2 *

(If)

     ,('?

_j 0 j

     -N
     ,o,   0.8

_j

(1.4 o.O

          -0.4 , ___ _ __,___ _ __.__ _ _ _.,___ _ _.....___ _ ____.__ _ _- J 0.0     0.2       0.4       O.ti    0.8           1.0       1.2 TIME {SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28C PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB 5 4 3 0 (I)

     ~     2:

(") (')

     .J 0

M

     ...i 0

0

          -1
          ~2 1..-____..:~--.L~~--.J..._--- - - 1 . - - --  ..J...._~ - ---"- -~ ----1 oJ>          0.2     o.4       o.s,         -o.a            1.o       1.2 TIME (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28D PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB 4 3 21

0 1

_3.___ _ __.__ _ ____.,_ _ __ ...__ _ __,_,_ __ _ L...-_ __ _ _..., CUJi 0.2 0 .4 0.6 0.8 1.0 , .2 TIME (S.ECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28E PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB

           !5 .

4 3

    ....0
    ~      2 fl')

M _j 0 in 1 J 0 0

          -1
          -2"-----------1------------------------i OJ)     0,i        0.4        0,6       0*.8  1.0       1. 2 l ME tSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28F PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB 4 c,i:n

     ~      2
     *M I)')

_j 0 I

           -2  L----.L.---- ..L...----'----- --....1..------'-- - - - - '

0.0 0.2 0.4 0.6 0.8: 1.0 1.Z TIME (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28G PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB 5 4 _j 0 I

      ~   1

_j

*~

0

         -1
         -2 .___ _ _Jb-_ _ _....____ _ __,___ _ __,___   _ _....,__  _ __,

0.0 I0.2 0,4 0..6 0.8 1.Cl 1*.2 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28H PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB I 4 3 C

     !e     2
     ~
     ..J 0

I o:i 1.

     ...ii 0
           -1
           -2 ...__ _ __,___ _ _---'-_  _ ____.,_ __ __.____ _ _-1...._  _ ____,

0.0 0,2: 0.4 0.6 0.8 1..0 1,2: TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-28I PRESSURE DIFFERENTIAL IN REACTOR CAVITY, COLD LEG DEB 7 51

       ;Q
       ,Cl.

4

       *M
       *C')
        ..i 0

I al 3

        ..i 0

2 1 0 '-'---_ _...,__ _ _____._ _ _ _. J . -_ _ __..__ _ _ _ ,....__ _ __. o.o 0.2 0.4 0.6 0.8: 1.0 1.2 TIME ($EC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-29 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG

            -I 0

c{l"I

            ...J
            ~

N

0. 6 0. 8 La I.* 2 1.4 1. 6 1. 6 dl. 0 0.2 0.4 T!H E {SEC )

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-30 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG c:J, .... ( J) a..

o. e; o. e 1. o 1..2: 1.' 1.8 1. e TIME lSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-31 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG 0

              -1 0

N "" N 0 c;b. 0 0.2 o. 4 0.6 o. 6 1. 0 l. 2 l.4 l. 6 l.B TIHE (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-32 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG w 01/') (f) CL Cl1 en _.J 0 N d). 0 0.2 o. 4- Q.6 o. 8 1. 0 1. 2 1. 4 1. 6 1. 8 TIME (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-33 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG c.n ll.. M (Y'l ....

             ...J 0

_J 0 N dl. C o. 2 0.4 0.6 O. 8 I. 0 1.2 l. 4 l. 6 I. 8 TIME (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-34 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG N

              ~c o-en 0..
              ...J EJ IJ'I   (D
               ...J 0

N d). 0 0.2 0. 4 0.6 0. 8 l, 0 1. 2 1.4 1.6 1. 6 T !HE !SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-35 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG C'II

              ...... o
              ....U'l Cl-'

0..

               ...J 0

WCO

                ...J 0

N l.B c0.0 c. 2 0. 4 0.6 0. 8 L 0 1.2 1.4 1.6 TIME ISEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-36 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG 0 t-W

             ....J 0

N , cO.O 0.2 o. 4 0.6 0.8 1.0 1.2 1. 6 1.8 TIME CSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-37 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG N

                ~o c-en Q..

er, cr, (I)

                 ...J 0

0) w

                 ....J 0
                         ~
                                \

I I"! I

                                  \r
                                   /

c!J.C 0.2: o. 4 0.6 o. 8 i. 0 i. ~ 1. 4 LS l. 8 TIHE (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-38 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG N

              ..... o o-C/1 a...
              .,.J E:)
              -O') " '
              ..J Cl N

d).0 o. 2 0.4 0.6 0. 8 l. C 1. 2 I. 4 l.6 1.8 TIME (SECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-39 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG 0 oco U1 a... _J El N _J 0 0

0. 4 0. 8 LC 1.2 1.4 1.6 1.8

                  <'fu. 0 0.2              0. 6 TIHE (SECl Amendment 61                                                                                      Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-40 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG ('\I,

             ---o
             ,en a..

m en IQ

              .,.J IC
              ....J 0
                       <;;{J, 0 o. 2     0. 4             0. B;      1. 0      l, 2 1.4    1.8 TIHE    CSEC l Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-41 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG N

               ~a o-Ul c..

_j C NW N _j 0 d). 0 0.2 0.4 0.6 o. 8 1. 0 1. 2 1. 4 1.6 I. 8 TIME ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-42 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG 0 (")

               ~     Ill ON Ul a..

('l")o (YlN

               ....I 0

Ill ('l")- ('\j

               ....I 0

0 L/1 d). 0 o. 2 0. 4 0.6 o. 8 1. 0 l.2 1. 4 l.6 J.B TIHE !SECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-43 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG u, en 0 (")

               ~   \l}

QN en a.. cr1o en N _J 0 lf}

N

               ..J a

D cO. 0 0.2 0.4 0. 6 0. 8 1. 0 I. 2 T!HE CSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-44 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG

               .---o Cl""'

C f) Q_

               ..J 0

0 cD.C 0.2 o. 4 o. 6 0. 8 l. 0 1. 2 1. 4 T!ME (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-45 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG 0 0

            ~n a....

co, I

            ...J 0

__J a N 0 1J. 0 C.2 0."' 0.6 0. 8 l. Q l, 2 l. 6 1. 8 TIME (SEC, Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-46 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG Ill a..,. if) a... _J 0 _J 0 C

10. o 0.2 o. 4 0.6 o. 8 1. 0 1.2 1.4 L.6 1.8 TIME ISECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-47 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG

                    =[
              §~l

.:i c..

              <"rlco M
              ...J 0

l co ~ N

              ....I 0

A/ C\I 0 J

                    %.o     0.2     0. 4   0. 6     O. 8       1. C     1. 2  1. 4    l. 6 l. 8 T !ME    (SECl Amendment 61                                                                                          Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-48 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG 0

                ....en Q      ID 0..

en.,,. N N 0

                       %.o   0.2      0.'    0. 6    O. 8      1. C    1. 2   1. 4    l. 6 1. 8 TIME   (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-49 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG

              .... o c-en Q...
              ...J 0

I otO ('() _J 0 d}.0 0.2 o. 4 0.6 a. s TIME

1. o (SEC)

1. 2 l.4 ..

                                                                                         ~ 1.8 Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-50 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG _J 0

             .J 0

N dl, C 0.4 0. 6 0. 8 LC 1.2 l. 4 1.6 l.8 TIME l"SEC:l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-51 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY COLD LEG 0 c-,i lt1

                ~       '

en a... CrJ 0 ('fl..: _.l 0 I .tl

\\

1 I

                               \!
                               *a o   ~I_ _ __.___ _ _.,L__ _ _. _ __ __ . ._ _ _--L_ _ ___.__ _ ___.___ _ _..,___ ___..
                      "7'c.;;     0.2      0. 4    o. 6      0. 8      1, 0     l. 2  1- 4      l. 6      1.8 TIME    (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52a PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 1,2 1.0 0.8 0 i:;

          ~   0.6
           ~

_j 0

          .- 0.4
          .J 0

0.2 0.0 0,2 0.4 0.6 0.8 1.0 1.2 TI ME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52b PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 1.2 1.0 0.8

          .Q en
          ~       0.6
          ~
           ...i 0
          ""..J   0.4
         *O 0.2 0.0
                - 0.2 '--- - - -----'-- -- - -'-- --   - -'--- - - --'- - - - ........- - - ---'

0.0 0..2 0.4 0.6 0.8 1.0 1.2 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52c PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 4 3 2 0 iii

           ~     1
           ~
           ...I 0

I

           ..,   0

_j

          *O
                -2
                -3L-_ _ _....1._ _ _ __.__ _ _ __.__ __ ___........_ _ _......._ __ __.

0.0 0.2 0.4 0.6 0.8 1.0 1.2 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52d PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 4. I 3 2 0

          ~    1
          ~
          .J g>

I

          .,. 0
          ..i 0
              -1
              -2
              -3L-- _ __J,__ __ __,___ _ _--1,_ __  _ ...L._ _ _ _.___ _ __,J 010     0.2       0.4       0.6       0.8         1.0      1.2 TIMf (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52e PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 4 3 2 Cl f 1

          ~
          ..i a

U) 0

          ..i g
              -2
              -3 ~ - - - ~ - - -- ..___ _ _....__ _ ____,_ _ _ __.___ _ ____.

0.0 0.2 0.4 0.6 0.8 1.0 1.2 TIME !SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52f PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 4 3 2 Q Cl.I 2:. 1 M M

          .J 0

I IP 0

          .J 0
               -1
               -2
               - 3 .___ _ ___.__ _ _ _..,__ _ _ _.__ __ .....__ _ __ .......__ __ __,

0.0 0.2 0.4 0.6 0.8 1.0 1.2 T IME {S EC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52g PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 4 3 I 2 0 g 1 I") I")

           ...l g
           ,...t  0
           ..i
          'O
                 -T
                 -2
                 -3~- - - ~ - - - - ~ - ---...__ ____._ ___ _______.
                                                           ~

0.0 0.2 0.4 0.6 0.8 1.0 1.2 TIME (SECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52h PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 4 3 2 c-1ii M (') J 0 I 00

           ..i, 0
         *O
               -1
               -2
               -3 0.0
                  ~ - - -0.2
                         - - -- - - - - - - - - - ~ - - - - ~ - - - ~

0.4 0.6 o.s 1.0 1.2 TrME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-52i PRESSURE DIFFERENTIAL IN REACTOR CAVITY, HOT LEG DEB 10 8 6 0

         ~
         ~

_j 0 4 I (b _j 0 2 0 0.6 TIME !SECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-53 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

               ~ . Q                0 .. 4  0.6      10 *. 8     I. 0

[ -

l. 2 '

l. -4 i. e TINE rSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-54 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG II I

                   ~ .o       0.2      0. 4  *O. 6    o. a     1.0      1.2           Le 1. 6 THt£ rsE.: ;

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-55 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG 0 Qin i. 0.4 fl. 6 n. e j. 0 T !H E CS EC i

1. 2 .! * .

                                                                                   . l. 6 Amendment 61                                                                                     Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-56 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

                ~l I\

I' I f I I I I

                                 'J 0.6      o. e      Lo     ... 2           I. 6 TIME !SEC l Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-57 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

~,

r[

              ;w i~
                * *11
              ~* I!

I'\

              -.J,      I
              ~~   N           I\  \;
                             '                 I

I

                   %. o     0.2      o.4     o. ,s    o. e     1.0    ~.z      i . ~.  ~ -~ ;.a TIME lSECl Amendment 61                                                                                          Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-58 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG N 0 0 (P (I") 0...

                ...J Q

LJ1 ....

                .J 0

N

                       ~.o  0.2       0. 4   0.6     0. 8     1. 0   1. 2     1.4     l.6 L8 TIME !SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-59 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

              -w Ul 0..
              ...J 0

lD (I) _J 0 0

                   "'fo. o 0.2      o. 4    0. 6     0. 8      l. 0     1.2    1.4    l. 6 l. a TIME   {SEC I Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-60 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG C

CX)

Cf) a..

              ...J 0
              ...J 0

0

                   %.o     o. 2      0. 4   0. 6      o. 8      l. 0     1. 2  l. 4   1.6 1. 8 TIME   (SEC I Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-61 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG 0 C)<D I.fl Q.. _J _J 0

                  ';ti. 6 0. 2      0. 4    0. 6     O. 8      I. 0    1. 2    t. 4   1.6 1.B TI ME {SEC l Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-62 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

                    ~l (D

(fl 0.. __J (Y") n) U'.l _J Q cr, .... _J 0 N I

01 I. 6 1.8

                    ~-*      0.2    0. 4    0.6      0. 8      1. 0    1. 2 TIHE   (SECl Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-63 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

             ~~ I I~
                    - 11
             ~          I

_j I

             ~  01
                    ~II~

11 I A rJ\ f\ I !I' W \ 1 l _.l_

                 *u. 0      0.2      0  ~~~---                1.0 - -I.;2 :-   1.4
                                                                                 .1,  6 I. 8 o.s      o.s TIME   (SECl Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-64 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

               - r.o o --

(n a. m"' l _(V") _ II __j

              ~
              ~

a:, I ('\I a

> I "To. o 0.2 o. 4 0. 6 0. 8 1. 0 1- 2 1.4 1. 6 1. 8 TI ME (SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-65 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

1 rr) (O 11 r"l") - 11

'\

__, I\ u_

I I

I I \

                ~           I       \
                >                    \

OJ c:.C. 0 0.2 o. 4 o. 6 0. 8 l. 0 l. 2 1. 4 l, 6 i. e TIME ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-66 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

              -o or-.

(fl a... 1

(l")l!) i

              ..J El I
                               \

N (>"]- ("\I I _j i 0 \\

(0

d). 0 0. 2 0. 4 0. 6 0. 8 l. 0 1. 2 l. 4 I. 6 l.8 TIME (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-67 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG 0 u'I a.. O'] .-, m ~- .

Ii

/!

r _j 0

             ~
                       -- 1
             ~Q'.l     '. I**

I *

                       ,. I
                       *I I

(\I _,J 0 ii II I i

                 ..... if ';
                                ~

fll ,

                                ':     (1     ,*'    l     1\   '""'

1 ./v 1l /

                                                      \\ (I \ /    \ ~ ~ ~ - - - - - - - - - - - --

I I 0 d \1 V

                       ----         -  - - ~ -- --                    _ __ _ _  ,1

__- - - ~- - ~- - - - ~ -- -- - - - - - - - - - - - - - -- - - - -

0. 2 o. 4 0. 6 0. 6 .I . 0 l. 2 ,. 4 .a 1 ! ME. fS Cl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-68 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

             ; ; CD 0
             -      11
             ~       !
                          *- 2      o. 4    *. 6     0. 8       LO      L 2    l. 4   I. 6 I, 8 TI HE  [SEC l Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-69 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG c:b.o 0. 2 a. 4 0. G 0. 8 I. 0 i.2 l. 4

                                                                                      . ' .;, i. B TIME    ISECl Amendment 61                                                                                           Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-70 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG 0 om ( I) 0-

            ...J 0
            ...J 0

N 0

                 <;'b. 0 0.2        0.4     o. 6     o. 8     1. 0    l. 2     l. 4   1.5 1.8 TIME (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-71 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG 1'o. o 0.2 0.1 0.6 0. 8 1. 0 1.2 1. 4 ,. 6 1.6 TIHE CSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-72 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG ow r.n a... ('(") rn

01 ('I N

_.J D 0 ('\I 0.2 o. 4 0.6 0. 6 1. 0 1. 2 1.. 1. 6 l.6 T IHE ISEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-73 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG (D ow (fl a..

            ...J 0

ON (T')

            ...J 0

0 N I To. o o. 2 o. 4 0.6 o. 8 1. 0 1.2 l, 4 l. 6 1. 8 TI HE {SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-74 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG 0 0 CJ') a.. a) _I C

            ...J 0

N 0 c;'b. 0 0. 2 0. 4 0. 6 0. B 1. 0 1.2 l. 4 1.6 1.8 T1 HE (SECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-75 PEAK PRESSURE DIFFERENTIAL IN REACTOR CAVITY HOT LEG

           ...J C

IV ("\1- \ (Yl

0 C

                -I "ab. 0   0.2       0. 4     0. 6    o. a       1. o    L. 2    1.4    l. 6 l. 8 T I HE CSEC l Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-76 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                 -N
            --o a-en 0..

CX)

            .... (ii)
            ...J a

LJ') (D 0 0 c:0,0 0.2 o** o. 6 0. 8 1. 0 1. 2 l.

1. 6 1. a TIHE (SECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-77 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

             -o o-cn c..

tC

-(D

_J 0 o"' 0 cO. 0 o. 2 0.4 0.6 o. 8 l. 0 I. 2 1.4 1.6 l.8 TIHE £SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-78 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 D Cl) Ill CL

-~

CJ

            ...J D
            ...J 0

dl. D 0. 2 0.4 0.6 0.8 1.0 l. 2 1. 6 l. 8 TIME £SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-79 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N

               -o o-( /)

a.. tD

                -co
                ...I 0

N ID

               ...I C

N d). 0 0. 2 0. 4 0. 6 o. 8 LO . 1. 2 l. 4 1. 6 1. 8 TIME <SECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-80 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 dl. 0 0.2 0.4 0.6 o. e 1. o l.2 l. 4 1. 6 1. e TJ11E ISECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-81 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 o* (('J CL

           ....I 0
                        ,vv--~---------'"-"------

No

           ..J Cl N

I I

                 'fu. 0  0.2      0.4     o. 6     o. 8      1. 0   . 1. 2   1. 4     I. 6 1.8 TIME (SEC l Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-82 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 0 cca (fl a...

              - co
              ..J 0

(\/

              ..J C

0 r:b.o 0.2 o. 4 0. 6 o. e 1. a 1. 2 1. 4 1.6 1. 8 TIME CSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-83 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                -c V)

CL

                ...I ID
                .... (.0 N
                ...I C

N dl. 0 o. 2 0. 4, 0.6 o. 8 1. 0 l. 2 1. 4 1.6 1.8 TIME CSEC J Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-84 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N

             ~o o-(.f) 0...

CX)

-w

_J 0 to (V 0 d), 0 o. 2 0. 4 o. 6 0.6 1.0 1.2 ' 1. 4 l. 6 l. 8 TIHE (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-85 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                -0 c-en a..
                -GO
                ...J 0
                ...J C

dJ, 0 o. 2 0.4 0.6 0. B 1. 0 1.2 1. 4 1.6 1.6 TINE CSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-86 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

            -o
            ....en ON 0..
             ~

en _J ID 0 N N-N _J 0 di . 0 0.2 0. 4 o. 6 o. 8 1. 0 l. 2 I. 6 l. 8 TIME ISECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-87 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N

             ,-.o ON VJ IL

['-U) _J 0 N N .... N

             ..J 0

c:0. 0 o~ 2 o. 4 o. 6 0.8 1.0 1. 2 1.4 1.6 1.8 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-88 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

            ...o-
            -r,i en a...

_J 0

            -I
                        ~---

0 0

                 ....I
                 'lb. 0  0.2       o. 4    o. 6     o. 8     1. 0    1.2     l. 4     L. 6 1.8 TIME £SEC l Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-89 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 J 0 C\IN N

             ...J 0

0 C\I I 1b. o 0.2 o. 4 o. 6 0. 8 I. 0 1.2 I. 4 1.6 LB TrHE (SEC 3 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-90 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N

               -o c-U')

(L ('l'l

               -     Cl)
               .J 0
               .J 0

dl. 0 o. 2 o. 4 0. 6 0.8 l.O 1.2 l. 4 l. 6 l. 8 TIHE (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-91 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

              -o c-en n...*
              ,.J 0

N d). 0 0. 2 0.4 0. :6 O. B T!NE

1. 0 (SEC )

l. 2 1..' 1. 6 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-92 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 ('I'

             ,,o Cl)

CL

              ...J
              'D
                           !rv,/,,-*,..--     --------..~*- ..,,.-----

C'T') I N

              -..J 0
             >      r-,

i

                    *ro. o   0. 2      0 . .f    0.6            0. B     1. 0   t. 2     1.6 l. B TIME (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-93 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

             ,c ro J

0 _l' 0 0

                   - I
                   ~ .o   ,a. 2     o. 4    0. 6    0. 8      1. 0    L2     1.4      1. 6 l. 8 T IHE CSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-94 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N 0... a:

                -o o-t i}

Q... (t")

                -io
                ...J 0

I I

                ..,. a)

(",)

                .J 0

N d). 0 0. 2 0 ... o. 5 O. 8 l, 0 L, 2 L4 1.6 l. 8 TIME <SEC? Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-95 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N

                -o c-q) a..

w

                .,..... *CC .
                ..J l'

C) I

                ..J 0

dl.O 0.2 0.4 0. 6 O. 8 L 0.. - t. 2 1.* I. 6 1. 8 TIH.E lS~L Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-96 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 C'II

                -c 1/V I ,

c -, j I (.I') a..

I j r--

                   .          I j
                ...:I C)
                >            I             \

N C)

                       ,;:(J.0       0,2      0. 4  0.6     D.8      l.O     l.2   1..'     l. 6 1. e TIHE CSEC1 Amendment 61                                                                                                Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-97 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 ('I') ' C C'II

                ...J C

I

                     ~-0     0. 2     0.(    0.6     O. 8     1. 0    l. 2  l. 4     l- 6 TIHE (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-98 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

0. 2 0.-t 0.6 0+ 8 l. 0 t .. 2 I, 4 1. 6 1. e TJNE lSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-99 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                --o o-(/)

Q..

                -~. fi lN er)
                .J 0

ij lJl (0 N

                -l
                ,c N

dl . 0 0.2 O.

0,6 0.8 1.Q l. 2 1** ** 6 1. a ilNE !SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-100 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 (I) ll..

                ..j Cl dl. D   I!. 2     0. 4 , 0. 6    o. 6         1. Q l.2  1. 4    l.6 LB T f ME (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-101 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 f

                 -*         I I\I

( I} Q._

                ~     co      I I

I'

                ..J
                                              '\

CJ U) ID j, I \ r,i I

                                                   ~
                ,...J 0

c:O. 0 0.2 O,t 0. 6 0. 8 1. 0 t. 2 J.4 l. ti LB TIME CSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-102 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

              -c o-( f)

Q., I I t

\\

t

                 . iI W ID
                                      ~

N

              ...J C

cO. 0 0. 2 0.( o. 6 D.8 LO l.2 1. 6 t. a T FHE tSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-103 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

              -a, cO. 0      o. 2      0.4             0,8      1.0     1,2    1. 4      l. 6 1.8 T.JHE (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-104 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

              -o ON vi c...

(D Ul

              ..J 0

N r-- (:',I I

                                 \_-.....,

_I 0 '-* a:, l: J l i I d), 0 0,. 2 o. 4 0. 6 0. 8 J. 0 l. 2 1.4 t. s 1. a T IHE CSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-105 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 fl) 0..

               ...I 0

To. 0 0 .. 2 o. ,t 0.6 O. B l. 0 L 2 1.4 '. 6 l. 8 TIHE ISE.C) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-106 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

              -(()

cll. 0 0.2 a.--t o. 6 - o..8 1. 0 l. 2 1. 4 l. :6 l. 8 TINE CSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-107 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

              -o ON en a...    'il I

L r-- w ' dl. 0 0. 2 D. 4 o.s o. 8 1. 0 l.. 2 I. 4 1. 5 1. 8 TrHE CSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-108 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 o-

                     .1,n

( I)

              ,0....

r'-:c N_ 0 ,..__ __..,_.__ _ _ ..___~-..i......---""----...........--_.,,____ _..-L--_ __.............._ ___,

                      "';b.o     0.2         0.(    0.6        0. B       l. 0           l~ 2  t. 4       l. 6            1.* a T IHE     !SEC.)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-109 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 0 N o-

              -  (,0 :

ON (Tl.,. 0

                 "'ro.o   0. 2     0 *. 4  o. 6    0. 8      l. 0    l. 2   l. t      l. 5 1. 8 Tl ME lSECl Amendment 61                                                                                          Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-110 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                     -Q M,..

0

                     *o. o   0.2      0. t   0. 6    0.6      1.0    l. 2   1. 4     I. 6 I. 8 TIME (SECj Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-111 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

a. 2 0. 4 0.6 0. 8 l. 0 I. 2 1. 4 l. 6 1.8 TI ME tSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-112 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                  ~l
              ~~l CL          l'l I

t-- ' (\J f,.J

                                                    ~-- ,... . _.. -.-~----~-----
                   ~

I C\b. 0 0. 2 0. 4 0. 6 0.8 l.D l.2 l. ' LG 1. 8 TIME fSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-113 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                  ~l
              ~~l CL          l'l I

t-- ' (\J f,.J

                                                    ~-- ,... . _.. -.-~----~-----
                   ~

I C\b. 0 0. 2 0. 4 0. 6 0.8 l.D l.2 l. ' LG 1. 8 TIME fSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-114 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 l

                       ,.I' C?o. 0      0.2      0.4    0. 6    0 .8      l.Q     1.2    Lt       l, 6 i. 8 TIME   (SEO Amendment 61                                                                                          Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-115 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N

              ..J 0

0 ... i:r;i,

              ....I 0

I

                     %.o   o. 2      0. 4   0.6      0. 8      i.O    L2      1, ~     I. 6 I. 8 TlME lSEC ,

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-116 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N

              ..J 0

0 ... i:r;i,

              ....I 0

I

                     %.o   o. 2      0. 4   0.6      0. 8      i.O    L2      1, ~     I. 6 I. 8 TlME lSEC ,

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-117 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                               \ ....
                                     "\
                                        \
                 -   m                    \
                                            \
                                              \

_.I 0 \ N dl . 0 0.2 o.* o. 6 0.8 \. 0 1.2 I.~ 1. 6 1. a T !ME. rSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-118 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                   ~r   I!

if) {L

                    ,o. 0 a. 2       0. 4   0, 5    0. 8       l.. 0    L 2  I. 4     1. 6 1. 8 TI ME    (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-119 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 dJ. 0 0.2 0. 4 0.6 Q.8 1.0 1.2 I.* 6 l-6 TIME (SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-120 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                  --;-- . 0  O* .::?   0. 4   o.s     0 .8     l.O     1.2   1.4 TIME rSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-121 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

              -o o-lf) a.
               ~             .Ii tV'\~

a> 0:, i J IV\}ll

I

;-L

_w I

              ; l;
                  ~ I I

I I q. N~ .__- ---'----,-..--------'-_ ____.____ _........___ _ _ . __ __ . . . . . _ ~ . . . . , _ _ _ _..............._ _ __ ,

o. 2 0, 4. 0.6 0. 8 I. 0 L2 l.4 l. 5 1.8 THt E CSE CJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-122 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 I - _J dl. o 0.2 0.4 0.6 o. 8 LO 1. 2 1.4 1.5 l, 8 TINE (S[Cl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-123 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

                     ~[
              .,.J D

_j j

dJ.O'- l \ _ ~~ ~ I. 0 ~ l.2 1L 4 . l.G 6

0. 4 0. (l 0. e T! ,E rs EC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-124 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 c'.). 0 O.l o. 4 0. 6 o. 8 1..0

                                                                    - l._;___7~-~-

2 l.4 1. 6

1. 8 TIME [SEC!

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-125 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 d) , 0 o. ~ o. 6 0, 8 1.0 l.2 1. 4 1.6 1. 8 TI ME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-126 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

0. 2 o. 4 0.6 o. 6 l. 2 1. 4 t.6 1. 6 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-127 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 d), 0 0.2 0. 4 o. 6 o, e 1. o J. 2 J. f 1. 6 1.6 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-128 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 c:t. 0 o. 2 0. 4

0. 6 o.a 1.0 1.2 l. 4 1. 6 LS TIME (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-129 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 w

              .,..._('I
              ,C) *...

ITT l:l...

                             !i
                  .               ,. l al
                             *1 i! '*tI
              .....J
                            .t !Ir: lIJ,~"'\J-.....r.....

0 *- *~ - --*-- -~---- - -- - *- - . - - - - -- --

1J 0 1: ~

l

                          .r I
                          ¥
                     °rn. 0                  0 .2                   o. 4              a. a        Q.8      L 0    l. 4 1.6 LB T!ME {SEO Amendment 61                                                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-131 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 (IJ '

              -o o *-

(I) CL. co,

              -ID
               ..Ji I!:)
               ...I 0

N dl . a 0.2 0.4 0.6 o. 6 \. 0 l. 2 1.4 1. 6 l. 8 TIHE (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-131 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1 N '

              -o O *-

V) CL.

              ..Ji C
              ..J c,

I ' N d) . 0 o. 2 0 . -4 0.6 o. 6 \.0 1. 2. LS t. a T!HE <SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-132 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 1

               ..J 0
               ..J 0

('1 1, u;\).o 0.2 o. 4 C.6 O. 8 l. 0 1.2 ' 1." l. 6 1. 8

                                                         !IME lSECl Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-133 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 0

d oo en a..

             ~-......ci 0          '
             ...I 0
                   ,:0 10 cnO o *~
             >o 0

D 0 ci "I' O L ...1..._ _ _ _..t,__ _ _ _.l._._ _ _ _i.._._ _ _......1_ _ _ __:L----...J...----..:....---___..1 9o. 0 1 _ _ __

0. 2 0. 4 . 0.6 O. 8 l. 0 1. 2 t. 4 j, s J. 8 T1 ME fSEC )

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-134 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                     ~r   I
              -o q .....

{ l'1 l,. 11...

              -/:f'j a:,

_J 0 I CJ.

                     <I;'
                          -1 L

dJ. 0 o. 2 0,6 0. 8 l. 0 1. 2 l. 4 i. G 1 . fl TI ME (SEC ) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-135 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

              .,...J D

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-136 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

             !'--<0 o -.

( f}

                       -IL\ /\
             ~             ~
             ~w
             ~
             ~Ji I      -

lfl (0 I l

                       *f
                   '" Ii.:....__

d]. 0

0. 2 o. 4 il - - - i . . . - - L . - - - L - . . - - - - - - - - - 1 ~

0. G 01. a 1. D l. 2 1. 4 l- 6 L6 TIME {SE C)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-137 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 en (I._

                  -I r:n.o            0. 4   0.6      0.6      l.O     1.2    1.4      ;. 6 .

TIME rSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-138 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 _J Cl ,

o. 6 0 ,8 I.* C '_l , 2 1.4 : ::;

1. 8 Tl ME (_SEC )

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-139 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                     . if)
              ....-c :~

VI (l_

                                  ~c'.'     -' s ~

o: o o. e ~ TI NE : L 0 - - : 1*. (SECJ. J ,2 -t 1. G l. 8 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-140 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 i

                                      '..              I   .,
                                                                    .I          L 0.2      0. 4   o.s     ,0 .' 8         1. 0   .... 1. 2 l.4 1. G LS TI ME

CS EC-J Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-141 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 (f")

              *    (\I (fl 0...

(X) _J . El

                         -I tD 0 l     ~~J I~

_J 0 i

/~

I

                    ~ - 0..

0. 2 o.

4

0. 13. 0. B I. 0 I. 2. 1. 4 l. G t. 8 TI ME (S EC )

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-142 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 I 1 c.D. e 0. 2 0. 4 O.G 0.8 i . C! l. 2 1.4 I. G i.8 TIME [SEC J Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-143 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                  ~r I
              - c ;
              ,:::i-c.D. 0  0. 2      0. 4   O. G    0 .8       1.0      1.2  I. 4     :*6 1.8 T IME   IS ECl Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-144 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

              ...J 0
                   -   1 I
                       ' - -- - - ' - ~- - L -- - - ' - -- - L - - - - - ' - - -- - - L - -- -- L - - - . _L   __J
                   ~ .O          0.2        0. 4   0. 6       0. 8        1.-0        1.2      1.4         I.G   1. 8 TIM E tS ECl Amendment 61                                                                                                                Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-145 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

               ~-0                 0.-!   0.6      0. 8     1.0     1. 2. l, -4      L8 rME (sECl Amendment 61                                                                                   Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-146 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 (") I

                -ro, o     0.2.             0.6      0,8       LO     l. 2   l. 4      l. 6 l.8 TIM E fS ECl Amendment 61                                                                                          Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-147 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

              -Q(!')

en 11.. _J 0

              ...J IQ 0

I

                   %.o    0.2      0. 4            O. 8       L D      J. 2          l.6 LB T JHE   ( SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-148 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

              ~c o-VI (L

r-- ' Occ*

                 +
              ...J 0

miD J 1, 0 d).0 0.,2 o,. " o. 6 O.B 1.0 L2 1. ,6 1. e TI H£ ,{SEO Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-149 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

              ~o o-( fl Q..
                 +
              ..J Q

J 0 d).0 0.2 o. 6 O,. 8 l. l'I L2 l. 4 l. B l.B TI HE lSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-150 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

               -o en a...

d).0 0 * .i:! o.-t o.s 0. 8 l. 0 L2 l.

J.6 l. 6 TIHE (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-151 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 d}. 0 0.2 0. f 0.6 0. 8 I. 0 I. 2 ** 4 1. 6 1. a TINE ISECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-152 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                .- o o-CJ'l CL 0

Ol <D

                ...J C

di. 0 0.2 o.* 0.6 0. H 1. Q J.2 l. 4 i. e 1. e TINE tSEC~ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-153 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 N a (/'l c.. CJ 0.4 o. 6 o. e LO L2 L t3 T lHE (SE Cl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-154 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 dl. Q o. 2 0.6 o. e 1.0 1.2 1. 4 I, S T!HE (SEC ) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-155 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

               ...J D
               ...J 0

Q. 2 o.. 4 0. ,6 0. 8 LO 1. .2 1.6 l. s THIE (SEO Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-156 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 (,')

             ,a..
                 +
              ..J 0
                   "l'a. o L
                             ------:-*~----;,LG--(~~~~~~lLO 0.4

0. 6 0.13 TIME iSECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-157 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 ON (f) 0...

                   -I
                   ~ IL........_.....t...._________,.______........_______L.------------~'----_______..______
                   ~ -o         n. .2       o. 4    0. 6     0. s        l. (! -  1 . ,2                    1.. 6 T!ME    CSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-158 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

i. e Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-159 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                     ~[
               --o l

( l) Q. m Oin

               ...J 0

I 0.4 - 0.' 6- - - - - - - : - ' ; - ~-~ ~ 1.~ 1.2 ~ L 6 I. 0 l. G 0'. a

T IHE (SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-160 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 I - 0.6 0.8 1 * .Q 1.2 TI HE !SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-161 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 0 en QN (f) a... f"\I

                -i 0
                ;,.-a
                                               -.. . . ,_~/-

O "-- - - -"------L---==-L----'.-....::.=:::::=L-- --L- ---L----L----1 d!. 0 Q. 2 0. 4 Cl. G o. a 1.0 !. 4 l lHE lSECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-162 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 a ( fl

               .J 0
                                  - --~;;------;t;~~'

D. ~ ,:c) * () -~~~~- I..? J. * *. 6 - 11. a o.s

0 .6 TI ME (SEC I Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-163 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                '?    ' L
                               -'-----   o. ,

0. 2 0 -

                ':"Q .C                         0. 6 Amendment 61                                                                                     Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-164 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 I

                                                                                      - - 1.8 e     LO      L2    l. 4

o. TIME £SEC J Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-165 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 C) _ , tr) Q., _J D cl!.()' 0. 2 0. ~ 0.6 C, 8 l.O 1.2 1. 4 1 - (i 1.13 TI ME (SEC} Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-166 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

               ..J 0
                   ":'ti. 0                C. G    10 . 8       LO      L 2 TI ME  ISECl, Amendment 61                                                                                Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-167 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                  .~ ,o
                  ,(r,
                  *!l..                                          -

1rv

('~ L

("!""!

                 "-I 0
                -~

l. 2 1. 8 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-168 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 c:i

                ~ ,C L
                    ,, - ,_ - - ~ ~ : - : : 7a. ~-~ ~-~

0.,2 (L 4 1. } * .2 l. 4

                                                                                      -1.e

o. s 0.8 TIME {SE:CJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-170 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 0.-!

                                            - -~':----;-';,---;-_',---;-_7--;-L O. G              L ll'   l. .2  1. 4      _'";;"6 - ii. 13

o. e TIHE ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-170 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                  ...J 0

N I cl!. 0 () . 2 0. 4 I?. (l 0. d l . t! E. .2 1.4 LG J.8 i IKE !SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-171 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

I I l D"' (f) o.._ 0

                 ,.,., C\I
                 .....J 0
                                     ~ ~ ~-----
                 .....J 0

a

                        <:'b. 0   0. 2     0. 4  O.G     o. t1      l. 0   1.2  I. 4   l. G 1.tl TI ME ISECJ Amendment 61                                                                                           Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-172 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

-1\

l.,;"'; [O / ("~ 0

                       ., .f
               .>           I I

I c:O . 0 0 . .,2 0.

c. f.l a. a 1.D 1* .2 1. 4 I
U TJME !SEC ;

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-173 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                    ~l          '\
                               /\I -(\
                -c o-lf) a...           f    \I \ ,~ ~--
                                         \I I
               <:C O i:,;:
                         -    I I
               -!             I 0
                         -l{\

LJ") (!) f"\J _J D

                    ... r
                        .___ _ __..__ _ __ _ _ _ _ ......__ _ __.____ _ _ _ ,...__ __   .__ _ _ . __ _ _ _ _ _ _ _ i dL Cl             Q. 2      o. 4   c. G     o. a       l. a      1. 2     l. 4      t. G     1. S TIME    iSEC 1 Amendment 61                                                                                                                 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-174 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                                                 ~-
                 ~-C        0. 2             0.6      0.8        l.O      1* .2         l. 6 "I !ME   (SEC l Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-175 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                                              ---~~~-        II. 0 0 .4    (L 6

o. a TINE ISEC i Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-176 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 0 m c:;) C o,.. _J 0 0

                          .___ _ __  ..__ _ _ _,___ _ _ _.,___ _ _ _ ,.__ _ _ _ _ c__ _ _ _..,__ _ _ _ _ , __ _ _ _...,__ _ __   1
                        ~ -0        o. .?     0.4       O.G        0. 8         J . C!     L2                      i. G       I. 8 TIME fSE CI Amendment 61                                                                                                                              Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-177 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                  ~Vl N-

_I l!:I 0 w f\l _,j 0 0

                       ";b. 0        o. 4           o. a      1. o      l. 2 1. 4 l. 6 l. a TIHE    ISEC:l Amendment 61                                                                                      Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-178 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 0 ffl I/lo NN 0

               ,...J 0

0 d) . 0 0.2 0. 4 0. 6 0. 8 1, 0 l. 2 L. 4 I. S l, 8 TIME CSECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-179 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 0 N'

                ...)

Cl

                     ~o. o  D. 2      0. 4  O.G      o. 8      1. 0     1. 2 t. "      1. a TIHE    !SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-180 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 N

C ._, o l l) c.. 0 dl. 0 0. 4 0.6 0. a l. 0 1. 4 1.6 I, 8 TIME {SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-181 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 ON

                <.n Q_

(',I I To. 0 0. 2. 0. 4 a. fi o. a Lo 1.2 1. 6 l. 8 THIE (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-182 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 oo Vl a_ 0)- NI

(',I

C I

                       ~.o     o. 2    o.

o.s o. 8 l. 0 1.2 l. 6 l.8 THIE fSEC J Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-183 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                      ~[
                -ol pJ.

8--... . . t,"l 0.... r I 01

               ,Q a;)
                ....J 0
                >            r I

row

               ,l;',j

_J i::l

u. 2 0." 0.6 o. e Lo l.2 l. 4 1.8 T1ME fSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-184 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                   ,,, en' CJ CD I

('J _.J 0 l.t'l f (0

                       '7o. o  0.2     0.4   O. G    0.8       L 0     t. 2 I, 4       I. 6 T IKE   !SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-185 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                        '                          - ~-
                                            ~ *, ...    --------
                   ,,,Oen' CD I

('J _.J cp Ill f

                        '7o. o 0.2     0.4   0. 6          o. e     i. o  1.2 I, 4 l. 6 I. 6 TIME (SH)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-186 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 _J D 0 V

                  "":'o. o   0. 2      0. 4   o. 6    0.8      1.0      1.2  1-4     J. 6 l. 8 TINE   ISEC l Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-187 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

                           ~r
                           "'            i\
                                        /(r i\
                   - o
                   ....oen ~          J I

l I a... I r-I Oc:, _j 0 I

> I I I o "' ) ',

I ('r"1 _j 0

                  >            I
                           ~1 ;

d} . (] o. 2 Q, 4 0. 6 0. 8 1. D J.2 I. 4 l.6 l.8 TIM E !SEC ) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-188 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 N

                  -c en il..

Cl} Oo:, _J 0

                  ..J 0

d). 0 0.2 0.4 0.6 o. a 1. a L 2 1. 4 l. 6 I. 8 TINE (SHJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-189 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 Cl c:, ( )1 a... m 0J <<)

               ...J' CJ o""

(f)

               -l 0

N .~ 0 1~-,-~

                     ":'b. 0    0.2
                                           --------.l.--'--------1....--L--1-----------1--'

0. 4 0.6 0. 8 T !ME

l. 0 (SEC J

l. 2 1. 4 l. 6 Le Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-190 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 r co Ul c.. N...,. ' (Tl '

                   ....J 0

C'.) o ' fn

                   .J 0

>('I t

t

                          ,:j, ~ - - -.........- - - - ~ - -- ~~ -- - - , . __ _ ___._ _ _ _....__ _ _ _...__ _ _ __.___ _ ___,
                         'i\). D        0.2            0. 4    0. 6      0. 6       I. 0    1. 2                   l. 6      LB THI E (SEC )

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-191 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3

             ,..... 0 o-
              ~<),

c.... 1/'l 0 a,,

             ..J 0

l l,Q , (T'l

             ..J C) c;O.O    0. 2     0. 4   0.6      0. 8      l. 0     l.2 TIME    (SEC}

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-192 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 N ' 0.6 o. 6

                                                                 - _-.1._--;-7-~-

l,".2 1.4 Le ll.8 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-193 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 c:i. tn Ill..

-N

(:Q l

             ...J 0
             ..... to  I, tr')
             .J C

N 0.,. 0,. 6 0, S l, 0 l. 2* I. 4 L S 1. 6 TINE fSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-194 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 N

             ~o QN u;

0.. C N N-m

             ...I C

ID di. 0 0.2 0.4 0.6 ,o. 8 l. 0 L2 1. 4 l, 6 l. 8 TIME fSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-195 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 3 N

                .J, 0
                ..J 0
                    .c:(J. 0 o. 2      o. 4  o. 6    0. 8     L 0     1. 2   l. 4       1.8 TIHE ISECl Amendment 61                                                                                      Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-196 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

~ '

1,

                    .         I.

I

               ..J C
                      -1~!_*_ ~-~L--_.-.-,..,l....-- . . . . . . . . . .----~----'---------'-----.______1 dl.0       0. 2       ~4                 Q.6           0. 6       1.0     1.2 1.4 t .5 t.~

T lME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-197 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 fl- l* o~ ......

                  ..J Cl

("J-I t;;J

                       -r:l.

re;.

                 >o 0

cb.O 0.2 _o. 4 0. 6 o. a 1. o 1.6 i. e TI ME CSE.Ci Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-198 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 C\I

               -a o *-

(/) a.

               ._.l 0

di- (I 0. 2 o. 6 o. 8 1. 0 1. 2 l.4 l. 6 1.6 TIHE ESECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-199 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

             .-i 10 oj',I N

J 0 Q ('J I L l-2 L4 I..8

                 ""fo. o 0.2       Q.* 4  0. 6     0, 8      l. 0 TIME    (SEC 1 Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-200 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                ....in c ...

0.. 0~ _J 0 0

                       ~-0   0.2      o. (  0. 6    o. a      l. o   1.2   1. 4    LS  L8 Tlt1E fSECJ Amendment 61                                                                                    Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-201 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 o..,

                 ...J C

N ' N N j 0

                      "ib. 0   0.2     a.~    o. 6    O, 8      l. 0     l. 2 1. ~  1.6 t. 6 TIME    [SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-202 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                      *Q r

c~ en 0... _.o I 0~ I

               ..J 0
                      ,n.

('!') '- IN L

               ...J C
               >o, 0

dl. o

                           &.....--- 0....L...----------~-

2 0. 4 0. 6

                                                               ----i--~-.

10, e

                                                                            . . . . . . . . . . . . . . . .1.2 L 0
                                                                                                           . .~--- .....----1.-

l.. 4 LG

                                                                                                                                - - - l... 8 TIME CSEC l Amendment 61                                                                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-203 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

              ..... c,1 CJ-*

(n

             ,ct.,
              ..J C

N d). 0 o. 2 ,o.

0 .. 6 o. a 1. o t. 2 l.4 LS l. 6 T !ME rsECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-204 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

             .-t CD
             .J 0
             ..J IO d). 0 o. 2     0 ... 0. 6    0. 6      l. 0    1. 2   l ,. -4  1. 6 l. 8 Tlt1E tSECl Amendment 61                                                                                            Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-205 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 N

               ~c o-en c..
               ...J 0
               ~

0 d). 0 0. 2 o. 4 0.6 o. a l. o l. 2 l.4 1. 6 l.8 TINE (SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-206 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

          =..,.
          -ci IQ
          ~]
          ~;:     l-c,..

I'----'----------"-----'-_____.___._______.___________._________

ih. 0 0.2 o.* o. 6 O. 6 LU L 2 l. 4 i. 6 1.6 TI ME rnEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-207 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                     ~r l
                §~l tr\

Q. a ce

                ...J 0

I_ dl. o 0 .. 2 0. 4 0.6 Oytl 1. 0 1. 2 l. 4 LS l. 6 TIME (SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-208 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

~[

U") ll..

                -    <<)
                ..J C)

Ul Ul N _j C Nl...___, _ _ _ , . _ _ _i , . _ _ i~ , _ _ _ , . _ _ _ ~ dJ. D 0. 2 o. ( 0.6 0. a l, 0 1.2 l. 6 1. 8 TIM E C:5 ECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-209 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

[

                           '° .  ,

I 1 o,- {/) tl. m~ {\

                    - ci
                    ....I 0

1 ... r.n ci ("\J _l

0

,,. Q .

0 IX) C;b. 0 Q.2 0.4 o. 6 o. 8 J. 0 1.2 1. 4 I. 6 l. 8 TrME. tSEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-210 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                   -. on c:::,- '

f,/) c...

                   "SI" C,    4, 01...:
                   ...,       !l C              1
                                  \I I '1'l ind
                   <"\I I~~
                   ..J IO
                   >c a

con ci Cl

                        -:ti. 0     0,2    0.4    0. 6  - 0. 8      1.0     1.2 Le     LS  l.8 TIME   !SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-211 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 o-I ll Q_

                ...J 0
                ,I) 0

('-J 0

                >o a

ID .. _ _ __ _. . __ __.__,___ _. . , _ __ _ _ . _ __ __,____ _ __,__ _ _. . L _ __ _ __.___ ___, 913. 0 o. 2 0. ~ 0.6 o. s 1.*o i. 2 l. 6 1. a TI ME ISECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-212 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

o. 6 1. 0 1. 2 1.e 0.8 TIME [SEC) -

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-213 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                  ~o o -

II (fl CL. O a;, _J 0 (O ta r-.

                 ..,..I ID r

II I

                        ~ Ii ----~-------<.- - - -_. .1. _ _ _ ___.1_ __ _ _ .__ _ __  _.___ _ ~ -L- -- - ~ - -- - l c:n. a     0. 2    o. 4   O. 6          0. 8        I. 0      1. 2       l. 4     I. 6    I. 8 TINE   iSECJ Amendment 61                                                                                                                 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-214 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 dl. O 0. 2 0.4 0. 6 O. 8 L 0 l. 2 l. 4 l. 6 i.e TI ME ISEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-215 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 m

                 -N
                ..I 0

(O-N

                ...J 0

0

I._________.___............... ~~'------'!

                     %.o      0,2      0. 4   0 .6    0. 8     L. 0    1. 2           ,. 6 I. 8 TIME ISEC 1 Amendment 61                                                                                          Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-216 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 I i1 I 1 \ 1 r\A

                               \    J\
                               . I j ~-

I 1

                     'm. 0       0.2     0.4   0. 6    o. e       1.0     I. 2 I. 4 I. 6 Le T fHE   CSECI Amendment 61                                                                                      Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-217 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                 ~-0        0.2       o. 4. O.G      0.8       1.0      l.2 I. (     I. 6 i. a TI ME   !SEC 1 Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-218 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                 -o o-Om
                 ..J D
                                            --~---

cil.O 0.2 o. 4 0.6 0. 8 l. 0 1. 2 1. 4 L6 1.8 TI HE CSE.Cl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-219 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                    ~r
               -o o -

i f) l CL d}. l 0 o. 2 0. 4 0, 6 0. B 1. 0 1. 2 L4 I. 6 TI ME lSEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-220 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                      '.!    r
                 ~o o-(fl I~

o_ f l r- I

                 -     Q;)

l _j

                 .EJ I
                              !j
                ,;-.0 I

N __J 0 d}. 0 o. ,.! 0.4 o. 6 o. 8 I. 0 l. 2 1.4 1. s 1. 1:l TINE ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-221 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 I Oil'

(D,.,.,

                  .J C

d). 0 o. 2 0. 4 o.s o. a 1. a l. 2 1. ,4 I. 6 L8 TH~E (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-222 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 c:, .,, I f) Cl...

                    .J IC N

dl. o 0. i 0. i 0. 6 o. 8 1.0 l. 2 I. 4 1. 6 l. B TIME !SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-223 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 r~1 ~

                  ~-      f     r

_J ) [ I I J

                     " J
                     %. c         0. 2    0. 4   0. 6   0 .8      l. O    1. 2 1. 4     I. 6 1.8 TI ME lSEC l Amendment 61                                                                                           Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-224 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

w tn ll..

                      ~o. o  0. 2     0. 4   !l. 6   0, 8      1. 0      1. 2 l. (   1. 6 1.8 TIHE    (S EC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-225 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                -o
                - L lfl 0..
                ;m         1) l1,.
                             *I m '°        I \f

('.J* I

                 ...)

0 V' N di. (I 0. 2 0.4 0 ,6, O. a l. 0 1.2 1. 4 !. 6 I. 8 r ME CSE C I Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-226 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 "lh. 0 0, 2 o. 4 0. 6 0, 8 I. D I. 2 I, 6 l. 6 TIME [S ECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-227 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                      .... I .

Q"l _J 0 _j 0 D 71 L

                      %.c      0. 2    0. 4  0.6     0. 8      !. 0     1. 2 1. 4  ,. 6 i. a TIME   !5EC l Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-228 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

p p t

1l

                   - Ij od (f}

j~

[ I 0~ mo _J 0

                        .... I N
                    *ol I m'

_J C,:c

                   > .           I 9
                       ~

I

                       ~
                       -;)J, 0

0. 2 0 * .f, 0. 6 0.8 1.0 J.,?

1. 4 l. 6 TIME ISEC J Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-229 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 "l.

                   ,(lj '

f "1b. o o. 2 o. 4 0.6 0. 8 l. 0 l. 2 1. 4 1. 8 TINE (SEC:} Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-230 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                        .. r I ['

D I\ . O _.,. *I V (/") Cl.. _J Cl I "' ml N

                  .-I 0
                  > '° "i

0 (", I

                       ~ ' - - - - - - L - - - - ' - - - - - - ' - - - - . . _ - - - - - ' - - -->-- - - - ' - - - - - - - ' - - - - -l
                       % .O         0.2       0.4          0,6          0.8         LO           1.2         1.4        1.6          1, 6 TIHE.   (SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-231 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                   ...I 0

_ __ i _ _ ~ _ 'l.B d! . 0 0.2 G. 4 0.6 o. 6 i. C 1. 2 l. 4 1. G TIME ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-232 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 Q(\/ ( I]

w N-

                   ...J 0

_J 0 I

                        ~.o    o. 2    0. 4  0. 6    <l. 8       l.0 Til1. (SECl l, 2 l.. 1.6  1. fl Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-233 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 CO l">IN ji

                 -'         /\

0 I o-m I I

0

                          .I j

cl I J

                     *1                   t
                     ~~ 0       0. 2    0. 4  o. 6    Cl. 8     I. 0    I. 2 1. -4   1. 6 l. 8 THIE (SEC )

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-234 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

~~

QN

               ,(I')

0... rry _ ( 8

                     °l
               <<)

0 C

                     ~I "l

J

                     '-0.o  0. 2      o. 4  C,6      o. a      1.0     1.2          l. 6 1. 6 i HIE lSfC l Amendment 61                                                                                        Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-235 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 0"" m CL.

             *"'1 me, Q

m

             ...J C

03 I

,_ _ _ _ _. . . _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ , _ _ ~ 1

                    ';b.o      0.2     o.-t    o.e     o.a       t.o     1.2    1. .c    l.6 1.e TI ME (SECJ Amendment 61                                                                                           Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-236 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                 - o U'l CL
               -_ . (0 0
                       ~....., ' I                  -...___----~----

d). 0 ~---..L,------;;"-;-----;'.,;----;o~ --~sl--}l~-~o~-::-~l~.2;----1.1-~"~--~1~-.~s~--- It .. a

o. 2 o. 4 o. 6 TI ME fSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-237 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

               -o o-N  /, , . . _ _} . . . . _ _ _ _ . _...________.___~.____________._J_____________..J dJ. O         0.2                 0.~      0.6    0.8       I.Q      1.2 !.~ 1.6   l.B i !ME   tSEC J Amendment 61                                                                                                 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-238 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 oui ( /1 (L_

-N

('") 0 N, 1'o. o [}. 2 0. 4 0.6 0. 8 l. 0 l. 2 l. 4 LB 1.8 T !ME tSEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-239 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 ITT _,J C Q I

                    ~-0     0. 2       o. 4  0.6     o. e     1. o    1. 2   1. 4    l. 6 l. 8 TIME ISECJ Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-240 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 r. II

                       "' I I
                                       ,j r1 I

I i I I j I I ~1

                                   .        n     i\

[ I1 Jl

                   ..J C              it 'I '\1/ I\1 \

I \~~- I I I I f I

                                '         J dl . 0                 o. 2     0. 4 0, 6    o.e      t.o      1.2 1. 4  l. 6 1. 8 TIME   !SEC J Amendment 61                                                                                                    Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-241 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                         ~r  l
                            !I
                                  ~
                         -l otn      i
                    ~        r l

c.o _j 0 Nl"l' ('1

                    ...J
                    ~

Ni I I I

                         -1
                            ~

d). (! 0.2 0.4 0,6 0. 6 I. 0 1. 2 I. 4 1. c 1.8 TIME !SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-242 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 _J 0 _J D di.(; 0.2 o. 4 0.6 o. e 1.lJ l 4 I 5 l , 8 TIHE rsEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-243 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                               - I            _I d]. 0    0. Z   0. 4   a. s    0. 8       I. Cl    I. l I. 4 1. 6 1. 8 TI ME    ISECl Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-244 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                  .=, ;::;,
                         -\
                 \fJ'

~

G.. j j

                  ,L-'l
                   . ,,  ~
                 ~
                        ,r,.J
                 '"')
                 "1
                .e;
                        ';t:}
                              '------~--------~c.._..-___,_,_____...___________. . .,_____   ~----1

0. 2 o. 4 0,6 0.8 l,Q l.6 TIME iSE(l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-245 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 [

""N

_ ,o ON ti') Cl-lD rD

0

I i I d) . 0 0. 2 0.4 a. G O. 'B 1,0 l. 2 1. 4 l. 6 I. 6 T iM E !S ECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-246 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

                      ~,i (f) 0...

0

                        . ______ ___,_____. . .,_____1
                                                     ,~----l.-.----.......-----'--------'---- . . .,_____ _J ro.n         0.2     0, 4     Q.6     o. e       1. e            l 4     L. G   : ... 9 TIM!:    tSF.C I Amendment 61                                                                                                           Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-247 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2

o. '! 0.6 :~';;---;-.',-1 1.0 I +2 l. 4 ~---i-1.

1. 6 B 0.13 TIME. ISE Cl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-248 PEAK PRESSURE DIFFERENTIAL IN STEAM GENERATOR LOOP - 2 1 IJ . 4 C,6 l.O _l. _2_ _ . . . LG l.l -!I. . . . _ _ _ ~ ~ ~l. 8 Cl . 8 TIN'E (SECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-249 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0

               ...J C

dl. O o. 2 0. -4 0. 6 0, B ~- 0 1. 2 l. 6 1. 6 TIM[ JSECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-250 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

               ..J 0

_j D

                   -1  . " - - - -~ . . _ _ _ _ . _ _ _ ~ ~ (

dl . O o., o. 6 0. 6 l. 0 L2 I. 4 1.S l. 8 TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-251 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT N

              ~o o-0   ID   1 0

_I 0 d}. O 0,2 o. 4, o. 6 o. 8 I. 0 I:. i 1.6 L6 TIHE ISEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-252 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0 (T"1 O.o o"" N

                ..J 0

0

                     'lti. 0 0.2     () , 4 0.6     0. 8         1. 0    1. 2 l. 4 I. S L6 TI f'tE   lSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-253 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT Qi/) en a..

                 ..J 0

dl . O 0.2 0.4 0. 6 0. 8 1. 0 1.2 l. 4 1. 5 l,a TIHE (SEC J Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-254 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT (D

                -D In

( f) CL _J C

                ..J 0

c:O. 0 0. 2 0.4 0.6 0. B l. 0 l. 2 1. 4 1.B i. a Tl KE fSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-255 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

               -out

{ fl ll. a .., _J 0 en ("I N

                -I 0
                ;=--

dJ,0 0.2 o. 4 0.6 0. B l. 0 J. 2 l.4 1. 6 1. a TINE ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-256 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                  -o o-Ul a.

Oo:t

                 ..J
                -o
                 ..J 0

d'.J. 0 0.2 0.4 o. 6 o. a i. o J.2 1.4 l.6 I. 8 T Tl"IE tSEC, Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-257 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

-o

( 11 (L 0 c;O. a 0.2 0.4 0 .. 6 o. e 1. o L. 2 1. 4 1, 6 1.8 Til1E CSECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-258 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

-o

( fl a..

                 -l 0

N d). 0 o. 2 o. t 0. 6 o. 8 t.. 0 1.2 1** l.6 l.8 T IHE !SECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-259 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT Tu.o o. .z 0.4 0.6 0. 8 I. 0 I. 2 I. 4 I. 6 Lfl TIHE (SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-260 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT (/'I 0.. C

                ...J 0

dl. 0 0.2 0.4 ,O. 6 o. 6 J. 0 1. 2 l. 4 t .* 6 1. 6 T Jr1E fSEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-261 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                 - n~
                 -o o-(./"J

{l..

                    .       Ji'I
                ...... <D

_.l 0 I ! j I l/J 0:0 {"\J

                ...J
                ~

N d). 0 0.2 0 *. 4 0.6 0. 6 l. 0 I. 2 I, -4 J.6 l. 8 T JNE lSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-262 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0)

                  -N
                 ...J 0
                 ....J' Cl 0
                        %. 0  0.2     0. 4   0. 6     a. a      1. o         l.

l. 6 l. 8 TIHE CSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-263 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT a.. a: II) oo V) a...

..)

0 Ill I

0 U? ,I

                   <">I Ill f
                              '-------'-----.L...----.1~----'-------11.....---...L----.._______,____
                        ':tl. 0 Q                                                                            ____,1 0.2     0.4      o. 6    o.a      1.a     1.2     t. 4  1.1.i     1.6 THIE   {SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-264 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT CN Q) a... (0 C\1-

                .-1 0

I ('1 I

                      ~.o   0-2      o. 4   0.6      o. !      1, 0-  l. 2  1. 4   1.6 l.8 TI NE (SEC>

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-265 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                ...J Cl lfl -

N

                ...J Cl 0
                     <;ti. 0 0.2     0. -f, 0. 6     0. 1:S     L. O TJHE ISEC )

t.' t. 6 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-266 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

-o

( I} Q,_

                -I ID
                .J C

dl , 0 0.2 o. 4 0. S o.e- 1.0 1. 2 l. 5 I. 8 iIHE rSEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-267 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                 -o o-( /)
                ~

_j a

                .J C

d)..ll o.i 0.4 0.6 0. 8 l. 0 l. 2 l. S I. S T IKE lSECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-268 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0)

                 -(",j

_,J C

                ..J 0

0 1

                    %.o     0.2      o. 4   D, 6     0. 8     l. 0    1. 2         1. 6 TIHE CSE.CJ Amendment 61                                                                                  Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-269 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT "It' N-J C I t.00 (',,I,

                ...J' C

I

                       <>,I I
                       %.o   0.2     0. 4   o. 6     Cl.ft     I. 0    1.2  1. 4   1. 6 J. 8 THIE ,(SEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-270 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                      ... I
                *OO"l U) a..

0

                     -1
                     %.o    0.2      o. 4   0.6      o. e      1. o         1.4    1. 6 l. 8 TIHE    fSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-271 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT N

                  -o o-
                  <n 0....

Om 0

                  ...J IO d). 0 o. 2    0.4    0. 6     O. B     1. 0   1.2   I. 4   l. 6 l. 'i.!

TIME rSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-272 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                 ~o Q ,-

r,h CL

                - cc
                ...J 0
                ...J Cl dl. 0   o. 2    0.4    c. 6     o. a      1. a     l. 2 t. 4  t. 6 l.6 TINE    ISECJ Amendment 61                                                                                      Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-273 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT N

                -o c-(/:)

a.. C _j 0 dl. 0 0. 2 0. 4 o.s o. 6 1, 0 1.2 l. 4 l. 6 1.B TIME CSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-274 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

              ~o c

l.n 0..

              ...J C

mW N __J 0 d) , 0 0 .. 2 o. ~ 0.6 0.8 1-0 l. 2 l. El l. 8 TINE tSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-275 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                  -o a-
                  ~

IL c--

                 -     ID a

d). 0 o. 2 o. 4 0.6 o. e 1. o t. 2 l.. 1.6 l.8 TI ME ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-276 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT C)IO ( f) IL _j 0 0 N I 1b. 0 0.2 o. ,c 0.6 0,8 1.. 0 l. 2 1.-1 1- 6 l. 8 TIME. (SECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-277 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT Ii\ aon ( /) 0...

                  ...J a

IO 0

                       - I
                       'itl.o         0.4    a. s    0. 8     I. 0   1. 2  1.4   !. 6 1. 8 T!HE rSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-278 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT co DID U1 Q_ _J 0

               \'.O "'

N C ' N 1

                     'To. o o. 2      o. 4  0.6      0. 6       1. 0     l. 2 l. 4 l.6 1.6 T IHE    CSECJ Amendment 61                                                                                     Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-279 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0 m"' N 0 Q

o. 2. 0. 4 0. 6 ll, 8 1. 0 J. 2 1. 4 1.S 1. 8 TI ME (SEC}

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-280 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT o-( /) ll..

                  ..I D

Ol I N (I') To, 0 o. 2 o. 4 0.6 a. a 1. o 1.2 l. 4 I. 6 I. 8 T rHE CSE Cl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-281 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT ON f J) 0 mo

                  ......J
                  ,o N

0) '

(,J I U) I

o. 2 0. 4 0.6 o. 6 1.0 1. 2 1, 4 1.6 1. a TIME ISECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-282 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0, 2 0.4 0.6 o. a 1. o t. 2 l. 4 I. 6 !. B TI ME ISE Cl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-283 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT o*

                  <fl 0..

('I] (T}"'

                  ...J 0

Cl) 0 N I 0.2 0. ,t 0.6 O. 8 l, 0 I. 2 J. 4 1. 6 1, 8 TINE CSECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-284 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                ...... o o-en
               .a_

_I C c;{l. 0 o. 2 o. 4 0.6 0, 8 1. 0 1.4 l.6 1.8 TIME IS EC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-285 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

              .J 0

0 (',I 1

                 "!'o, o   0.2      0.4    o. s     o.a       Lo     1.2     l. 4   l. 6 1.6 Tl ME  lSEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-286 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT oco ( f') 0...

                -I C
                -l C

0 N To. o o. 2 0.. 4 0. 6 o. e 1. o 1. 2 l, 4 l, 6 l.8 TIME CSECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-287 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0 0 ID en 0..

               ..J C
               ...J C

N 0

                      ~-0   a. 2       0.4   0.6      O. tJ      l, 0     1. ,2 l. 4 l. 6 1. a TrHE    CSECJ Amendment 61                                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-288 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT Oto {I} CL

               ....I D

ON (n _I 0 C

                     <'I I
                     "to. o o. 2      0.4    0. 6     0,8      1.0     l.. 2  1."    LS 1. 6 TIME   ISEC)

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-289 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

_J 0 d).0 0.2 o. 4 0,6 0. 8 l. 0 I. 2 1. ( l. 6 t. 6 TIME CSECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-290 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT N N , en. o 0, 2 o., o. 6 D. 8 l. 0 t.2 l. 4 l. 6 ?* B TIME (SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-291 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0"' ( /) 0..

                  ...J 0
                       -I N
                       ";b_ 0 0.2    0.4    0.6     o. 8     1. 0   I. 2   t.

1. 6 I. a TIKE CSECI Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-292 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT C"' en

                  .J Q

0 0.2 0. 4 o. s O. il I. 0 T !NE (SEC> Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-293 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT

                ~o c-( /"j a...

0 _.. ID 0 _l C N dl. Cl 0.2 o... 0.6 o. a 1. o i.2 1. 4 t. 6 I. a Tr ME lSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-294 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT di. D o. 2 0.4 ci. 6 o. 8 t. 0 l. 2 l. 4 l,6 l. 6 TIHE ISECJ Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-295 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT dl.O 0.2 o. .. 0 .. 6 0. B t. 0 l. 2 1.4 1. 6 1.a TINE CSEC1 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-296 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT QN U) 0..

                  ...J 0

1 N (r} , (0 I

                        ~- 0  0, 2    o.'   o.s     o. 8     1. 0   L2     l. 4      1. 8 rtME [SECl Amendment 61                                                                                    Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-297 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT our Ul 0..

                .J 0
                      "'fo. o 0.2     0 .. 4 0.6     0. a     l. 0   t. 2   1. 4  l.5 l. 8 TIME {SECl Amendment 61                                                                                     Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-298 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT t:O . C 0.2 0.4 0.5 O, 8 l. 0 1, 2 l. 4 1. 6 1. 8 TIME (SEC l Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-299 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT 0 Octl en a... _J D

                ..J Cl N

0

                     ~-0    0,2      0, 4   o. 6     0.8       1.0     1.2  I. -4  1. 6 l. 8 T lt1E !SEC 1 Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-300 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT d). IJ 0.2 0.4 0.6 o. 13 l. 0 1. 2 t. 4 l.6 1.8 TJME ISECl Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-301 PEAK PRESSURE DIFFERENTIAL IN PRESSURIZER COMPARTMENT dl. 0 o. .2 0.4 0.6 O. 8 l. 0 1- 2 !. 6 r. a THIE [SEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-303 HEAT REMOVAL RATE OF EMERGENCY COOLER UNIT 300 ZERO FOU L ING 33°F WATER TEMPERATURE

u. 31.250 C FMI AIR FLOW 0

1500 GPM WATER FlOW 100 l?ERCENT SATURATION 200 I. 100 - - - 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80- 85 9U 95 100, 105 110 H EAT REMOVAL.* RATE PER FAN {BTUi ttR X 1,0 6) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-306 CONTAINMENT VACUUM RELIEF SYSTEM ELEME!iI HQ, DESCRIPTION t ENTFIY WITH SCREEN 2 LOUVER

                                    ~       EXIT TO PLENUM 4       ENTRY WITH SCREEN 5       TORNADO DAl,IPl:II 6       PAMPER 7       EXIT Wl1H SCREEN 8       BEU MOUTH ENTRY WITH SCREEN
                                    ~       BIITTI;RFLY VALVE to       TRANSITION (CONTRACTION}

11 PIPE: RUN t2 TflAN$JTION (EXPANSION) t3 VACUUM REUliF VA!.VE 14 SPOOL PIECE 15 l:](IT 16 DIVERGING TEE Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-307 FORCES ON THE REACTOR VESSEL COLD LEG NOZZLE 150 IN2 BREAK 140*.00* NOTE: *THE P -EAK FORCE IN X-DIRECTION OCCURRED AT 0.01625SEC. THE CORRESPONO, NG FORCE. lN. Y -DIRECnON lS 9.41 )(. 1o4 LBF. 120.00 100.00 80.00 "81.05 x 104 LBF ii;'

  ~
  ~        60.00
    )C

ii!;

v.l LL X-Direction 40.00 20.00 Y-D l rec tlon o.oo ' z-otrect1on

          -20.00 - - - - - - - - - - - - - - - - ~ - - - - ~ - - - - - ~ - - - - -

0.00 0 . 20 *0.40 0.60 0.80 1.00 1.20 TIME CSEC) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.1-308 MOMENTS ON THE REACTOR VESSEL COLD LEG NOZZLE 150 IN2 BREAK 480.00 400.00 320.00 240.0(J u:-

       ....i,,:.

co

      'LL.

0

E 160.00

l

80.00 ABOUTY AXIS ABOlITXAXIS 0 . 00

                  *80,00
                 *160.00 ...__ _ _ ___.__ _ _ _ _..,___ _ _ ___._ _ _ _ _......__ _ _~__,...__ ~ - - - '

0 . 00 0 .20 0.40 0.60 0 . 80 1.00 1.20 T IME (SEC.) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-4 CONTAINMENT FAN COOLER PERFORMANCE CURVE 70 AAF COIL 16-60-6W5-6C 60 TUBES 90-10 COPPER

NICKEL GAS FLOW 31,250 CFM DELIVERED WATER FLOW 1360 GPM FOULING FACTOR 0.001 so

             .c
             ~ 40
             '5.

C

             ~
             ..I
             ~ 30 w

c Inlet water Temperature 20

                                                                 -+-80F
                                                                 ----83F
                                                                 --.!r-86F
                                                                 -w-89F 10                                             --92F
                                                                 ...,_9SF 0+---+---+---+---+---+---+---+---~

100 120 140 160 180 200 220 240 260 GASTEMPcRATURE(F) Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-6 CONTAINMENT SPRAY NOZZLE DROP SIZE HISTOGRAM

       .10 -
       .09 ....
       .08 z

w

   ~ .06 0

w fl: c-

      .05 w

i== j

  &   .04
                                -       c-
      .03  -
      .02
      *01 ,-
      .OD       I                                                                     - -                ~mn rh I  I  !  I      I       I 0        10                20         30        40       50              60      70        80   90 100 I 110    120     130 MICRONS (X10 1J Amendment 61                                                                                                           Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-7 CONTAINMENT SUMP PLAN

     -  ----:_..- N I

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-8 CONTAINMENT SUMP SECTION A - A Security-Related Information - Figure Withheld Under 10 CFR 2.390 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-9 CONTAINMENT SUMP SECTION B - B Security-Related Information - Figure Withheld Under 10 CFR 2.390 Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-17 CONTAINMENT SPRAY PUMP PERFORMANCE CURVE 1 6 0 0

,c:

5 TOTAL H EAD I-w w u.. 4 z C

               <(      3 w

i:::

               ..JI
               ..... 2
               ~

O* I-1 - 0 0 500 1000 1500 2000 2500* 3000 3500

                                                   *GAL:LONS PER MI NUTE Amendment 61                                                                                       Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-18 RESIDUAL HEAT REMOVAL PUMP PERFORMANCE CURVE

                        .o. *-------------"!'-__________,.,___

0 2000 *:uioo

                                                  *aAi..LONS Psi~ ft!l~UT6° Amendment 61                                                                         Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.2-20 CONTAINMENT BUILDING RECIRCULATION SUMP STRAINER ISOMETRIC REORCULATION SUMP TOP HAT STRAINER ARRANGEMENT TYPICAL OF SUMPS lA AND 1B Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2.5-7 POST ACCIDENT HYDROGEN MONITORING SYSTEM 1'r--- 1====-=- -t-Ni PUA.ct--- -- ~ REMOTE CONTROL LO SPMil CAUillllA,nOH - - - - ~ PMIEL A I tll'Sl'AN"CAs~---- Cl.ASS IE. II I

                                                                                      ~~~-'~:~~'                     ~------------7 f-<,..,_..,__..,,.~---~----~-.i HY'OROGEN AKA.1...Y2ER REUOTE SAAOP1.E OJLUTIOH PAN.EL
                                                                                                                                                    *NNS
                                                                                                                                                               - l'll7rN;a (PURct) I CABINET A L---,----,---r-1--~--*MY ... """*       I Cl.ASS 1E                                                                              I l

I

                                                                 .,.....,. ______                                 ,------------_J r-----1:-

1====r=

RE ... 01E CONTROL t o ~ CAUMAtIQN - - - - - l)AIIICL B Ml. 5PM OOts - - - - Cl.ASS u: I I

                                                                                     ,--'----.L---'-1-'-I.....,

I c;.eJNET B I I I MAIN CONTROL ROOM (RAB) I I I I I CONTAlNMENT REACTOR AUXILIARY BUILDING I Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2A-1 TEMPERATURE GRADIENT IN GASEOUS & LIQUID BOUNDARY LAYERS DURING HEAT SINK SURFACE CONDENSATION HEA! SlNK INU:AlOi:t Sa.II( SUAFACS CONDENS,AA~Tl;;--f!---- -- - -- .f:....-------

                &OUNOARV
                   ,LAYeR_ _
                               - T- - *:.....--~ L _ - - - - - U0Ulo.<3AS 1N1'1!RFACE I

I l

1 I

GASEOUS I BOUNDARY I LAYER I I auu:; CONTAlNME.N1 t:i1HHON Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.2A-2 SPRAY EFFICIENCY VS STEAM/AIR RATIO 1 .0

                  .9  -
                  .8
                  .7
           >- .6 u          -

zw u LI.. .5 LI.. w

a: .4 a.. (/)

                  .3
                  .2
                  .1 0
                        .1      .2 .3    .4        .5     .6     .7       .8     .9    1.0 1. 1 1 .2    1.3 AA T IO OF MASS STEAM TO H t f MASS OF AIR Amendment 61                                                                                                 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES TO FIGURES 6.3.2-4 THROUGH 6.3.2-6 The process flow diagrams are provided for illustrative purposes only and are not intended to represent the flow rates or temperatures used in various accident analyses. The process flow diagrams are developed to provide representative system performance data based on minimum safeguards systems alignment.** The flow rates in the FSAR accident analyses are conservatively applied. Valve alignments are provided for all principle modes of ECCS operation. Amendment 61 Page 1 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES TO FIGURES 6.3.2-4 THROUGH 6.3.2-6 (Continued) VALVE ALIGNMENT TABLE PRINCIPLE MODES OF ECCS OPERATION C E G Injection D Cold Leg F Hot Leg B Minimum Cold Leg Recirculation Hot Leg Recirculation A Injection Safeguards Recirculation Minimum Recirculation Minimum Valve Normal Maximum (Train A Maximum Safeguards Maximum Safeguards No. Standby Safeguards Only) Safeguards (Train A Only) Safeguards (Train A Only) 1A O O O C C C C 1B O O O C O C O 2A O O O O O O O 2B O O O O O O O 3A C C C C C C C 3B C C C C C C C 4A O C C C C C C 4B O C O C O C O 5A C C C O O O O 5B C C C O C O C 6A O O O O O O O 6B O O O O O O O 7A O O O O O C C 7B O O O C* C* C O 8 C C C C C O O 9A C C C O O O O 9B C C C O C O C 10A C C C O O O O 10B C C C O C O C 11A C C C C C C C 11B C C C C C C C 12A C C C C C C C 12B C C C C C C C 13A C O O C C C C 13B C O C C C C C 14A O C C C C C C 14B O C O C O C O 15A O O O O O O O 15B O O O O O O O 16A O O O O O O O 16B O O O O O O O 17A O O O C O C O 17B O O O C O C O 18A O O O C O C O 18B O O O C O C O 19A O C O C O C O 19B O C O C O C O 19C O C O C O C O 20 O C C C C C C 21A O C C C C C C 21B O C O C O C O 22A O O O O O O O 22B O O O O O O O 23A C O O O O C C 23B C O C O C C C 24 C C C C C O C Amendment 61 Page 2 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 C E G Injection D Cold Leg F Hot Leg B Minimum Cold Leg Recirculation Hot Leg Recirculation A Injection Safeguards Recirculation Minimum Recirculation Minimum Valve Normal Maximum (Train A Maximum Safeguards Maximum Safeguards No. Standby Safeguards Only) Safeguards (Train A Only) Safeguards (Train A Only) 25 C C C C C O O 26 C C C O C C C 29A O O O O O O O 29B O O O O O O O 29C O O O O O O O Amendment 61 Page 3 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES TO FIGURES 6.3.2-4 THROUGH 6.3.2-6 (Continued) PROCESS TABLES MODES OF OPERATION MODE C - INJECTION/MINIMUM SAFEGUARDS - TRAIN A OPERATING This mode represents the process conditions for the case of minimum safeguards with RHR pump 1 and CC pump 1 taking suction from the RWST and delivering to the reactor through three cold leg connections. MODE E - COLD LEG RECIRCULATION/MINIMUM SAFEGUARDS - TRAIN A OPERATING This mode represents the case of cold leg recirculation with RHR pump 1 on and CC pump 1 operating. In this mode the safeguards pumps operate in series, with only RHR pump 1 capable of taking suction from the containment sump. The recirculated coolant is then delivered by RHR pump 1 to CC pump 1, which delivers to the reactor through three cold leg connections. The RHR pump also delivers flow directly to the reactor through the same three cold leg connections. MODE G - HOT LEG RECIRCULATION/MINIMUM SAFEGUARDS - TRAIN A OPERATING This mode represents the case of hot leg recirculation with RHR pump 1 and CC pump 1 operating. In this mode, the safeguards pump again operate in series with only RHR pump 1 taking suction from the containment sump. The recirculated coolant is then delivered by RHR pump 1 to CC pump 1, which delivers to the reactor through three hot leg connections. The RHR pump also delivers directly to the reactor through two hot leg connections. Amendment Amendment 61 Page 4 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES TO FIGURES 6.3.2-4 THROUGH 6.3.2-6 (Continued) PROCESS TABLE MODE C Location Fluid Pressure (psig) Temperature (F) Flow (gpm) 1 Refueling Water 0 70 - 2 Refueling Water 0 70 6160 3 Refueling Water 0 70 2400 4 Refueling Water 0 70 650 5A Refueling Water 0 70 3760 5B Refueling Water - - 0 6A Refueling Water 110 70 3760 6B Refueling Water - - 0 7A Refueling Water - 70 3760 7B Refueling Water - - 0 8A Refueling Water 80 70 3760 8B Refueling Water - - 0 9A Refueling Water - - 0 9B Refueling Water - - 0 10A Refueling Water - - 0 10B Refueling Water - - 0 11A Refueling Water ~50 70 3760 11B Refueling Water - - 0 12 Refueling Water 0 70 1490 13 Refueling Water 0 70 1470 14 Refueling Water 0 70 1400 15 Refueling Water - - 0 16 Refueling Water - - 0 17 Refueling Water - - 0 20A Reactor Coolant - - 0 20B Reactor Coolant - - 0 21 Recirculating Coolant - - 0 22A Recirculating Coolant - - 0 22B Recirculating Coolant - - 0 40 Refueling Water 0 70 650 41A Refueling Water 0 70 650 41B Refueling Water - - 0 42 Refueling Water - - 0 43A Refueling Water 0 70 650 43B Refueling Water - - 0 43S Refueling Water - - 0 44 Refueling Water - - 45A Refueling Water 1430 70 650 45B Refueling Water - - 0 45S Refueling Water - - 0 46 Refueling Water - - 0 47 Refueling Water 1300 70 50 48 Refueling Water 1300 70 600 49 Refueling Water - 70 600 50A Refueling Water - - 0 50B Refueling Water 1000 70 600 51A Refueling Water - - 0 52A Refueling Water - - 0 53A Refueling Water - - 0 51B Refueling Water 100 70 200 52B Refueling Water 100 70 200 53B Refueling Water 100 70 200 54A Refueling Water - - 0 55A Refueling Water - - 0 54B Refueling Water - - 0 55B Refueling Water - - 0 56 Refueling Water - - 0 57A Refueling Water - - 0 Amendment 61 Page 5 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 57B Refueling Water - - 0 60 Nitrogen 0 120 0 61 Nitrogen 0 120 0 62 Nitrogen 0 120 0 Amendment 61 Page 6 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES TO FIGURES 6.3.2-4 THROUGH 6.3.2-6 (Continued) PROCESS TABLE MODE E Location Fluid Pressure (psig) Temperature (F) Flow (gpm) 1 Refueling Water - - - 2 Refueling Water - - 0 3 Refueling Water - - 0 4 Refueling Water - - 0 5A Recirculating Water 12 244 3820 5B Recirculating Water - - 0 6A Recirculating Water 115 244 3820 6B Recirculating Water - - 0 7A Recirculating Water - 244 3820 7B Recirculating Water - - 0 8A Recirculating Water 85 180 3820 8B Recirculating Water - - 0 9A Recirculating Water 85 180 650 9B Recirculating Water - - 0 10A Recirculating Water - - 0 10B Recirculating Water - - 0 11A Recirculating Water ~60 180 3160 11B Recirculating Water - - 0 12 Recirculating Water 0 180 1620 13 Recirculating Water 0 180 1170 14 Recirculating Water 0 180 970 15 Refueling Water - - 0 16 Refueling Water - - 0 17 Refueling Water - - 0 20A Recirculating Water - - 0 20B Recirculating Water - - 0 21 Recirculating Water 0 244 3820 22A Recirculating Water 0 244 3820 22B Recirculating Water - - 0 40 Recirculating Water - - 0 41A Recirculating Water - - 0 41B Recirculating Water - - 0 42 Recirculating Water - - 0 43A Recirculating Water 65 180 650 43B Recirculating Water - - 0 43S Recirculating Water - - 0 44 Refueling Water - - 45A Recirculating Water 1475 180 650 45B Recirculating Water - - 0 45S Recirculating Water - - 0 46 Recirculating Water - - 0 47 Recirculating Water 1345 180 50 48 Recirculating Water 1345 180 600 49 Recirculating Water - 180 600 50A Recirculating Water - - 0 50B Recirculating Water 1045 180 600 51A Recirculating Water - - 0 52A Recirculating Water - - 0 53A Recirculating Water - - 0 51B Recirculating Water 100 180 200 52B Recirculating Water 100 180 200 53B Recirculating Water 100 180 200 54A Refueling Water - - 0 55A Refueling Water - - 0 54B Refueling Water - - 0 55B Refueling Water - - 0 56 Refueling Water - - 0 57A Refueling Water - - 0 Amendment 61 Page 7 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 57B Refueling Water - - 0 60 Nitrogen 0 120 0 61 Nitrogen 0 120 0 62 Nitrogen 0 120 0 Amendment 61 Page 8 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 NOTES TO FIGURES 6.3.2-4 THROUGH 6.3.2-6 (Continued) PROCESS TABLE MODE G Location Fluid Pressure (psig) Temperature (F) Flow (gpm) 1 Refueling Water - - - 2 Refueling Water - - 0 3 Refueling Water - - 0 4 Refueling Water - - 0 5A Recirculating Water 12 180 3710 5B Recirculating Water - - 0 6A Recirculating Water 115 180 3710 6B Recirculating Water - - 0 7A Recirculating Water - 180 3710 7B Recirculating Water - - 0 8A Recirculating Water 85 125 3710 8B Recirculating Water - - 0 9A Recirculating Water 85 125 650 9B Recirculating Water - - 0 10A Recirculating Water - - 0 10B Recirculating Water - - 0 11A Recirculating Water ~60 125 3060 11B Recirculating Water - - 0 12 Recirculating Water - - 0 13 Recirculating Water - - 0 14 Recirculating Water - - 0 15 Recirculating Water 55 125 3060 16 Recirculating Water 0 125 1730 17 Recirculating Water 0 125 1730 20A Recirculating Water - - 0 20B Recirculating Water - - 0 21 Recirculating Water - - 0 22A Recirculating Water 0 180 3710 22B Recirculating Water - - 0 40 Refueling Water - - 0 41A Recirculating Water - - 0 41B Recirculating Water - - 0 42 Recirculating Water - - 0 43A Recirculating Water 65 125 650 43B Recirculating Water - - 0 43S Recirculating Water - - 0 44 Refueling Water - - 0 45A Recirculating Water 1480 125 650 45B Recirculating Water - - 0 45S Recirculating Water - - 0 46 Recirculating Water - - 0 47 Recirculating Water 1350 125 50 48 Recirculating Water - - 0 49 Recirculating Water - - 0 50A Recirculating Water - - 0 50B Recirculating Water - - 0 51A Recirculating Water - - 0 52A Recirculating Water - - 0 53A Recirculating Water - - 0 51B Recirculating Water - - 0 52B Recirculating Water - - 0 53B Recirculating Water - - 0 54A Recirculating Water 100 125 200 55A Recirculating Water 100 125 200 54B Recirculating Water - - 0 55B Recirculating Water - - 0 56 Recirculating Water 0 125 200 57A Recirculating Water 1125 125 600 Amendment 61 Page 9 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 57B Recirculating Water - - 0 60 Nitrogen 0 120 0 61 Nitrogen 0 120 0 62 Nitrogen 0 120 0 Amendment 61 Page 10 of 10

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.3.2-4 EMERGENCY CORE COOLING SYSTEM PROCESS FLOW DIAGRAM SHEET 1 I IIINSl/tH I IC$ 1111, ....1.4.,-1--....... -v l t .C

                 . *IIU H0l _ ......._ _......_...,..,_. .,- ..

ltG i

                 . *ts          _..,.._-'-.--r~.--                                                           I I.

Clll.11 _ UG u:~ (ot.~1 .....-L,,,,:..:.......-.i..,.,.... tl~

                *c:s:
             - t~\Jf*..L..li/1..1-....L-L,,-il-,I
              ' tlC fl 101 U5                                                           I
                                                      ,u.                        l:

a, lllll 1U. INII 0111 l~C elllC Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.3.2-5 EMERGENCY CORE COOLING SYSTEM PROCESS FLOW DIAGRAM SHEET 2

LHSI/RHR I

                                                                                                      ~i20 PUM:P 1 (SHEIT 1) l I       s u-rst/COLD l[G                                                        ~        ;-{

IMJECrJON __ _ _:; __ ,_ UNES_.:_

                                                                                                           ~I -t~j (SHE,;J 1)

,.,s s T c;; (

eves A NORMAL.- -,r.,- - -;_-,.:;- I I 1 I .,,, CHG.. 21B 21A ~ f---'--+-I-v1---'--,~_,__,__~.__, I I HHS.I/CHG LHSI/HOT LEG I PUMP 1 R[CIRCULATJON

UNES-*I _.,,., .

I ,s I Q

                                                                                                                                              ,. ~

(SHE(T 1) r-~Ei7 I I J., 1-

'~

I RCS HOT LEG _ ___._,_,.,,...,.... I (S><m1J

                                                                                                                            -+-'-+---          -1*2 i-ci,..VCT_

14A 148 1---v1--- RWST

                                                                                                                                                                   -   (SHE£T 1) 16" LHSI /HOT LEG RECIRCULATION LINES       .

(SH<IT 1) 168 LHSI/COLD LEG

                      ..~-~-[~~;TJON ___.

(SH<IT 1) eves Ll<SI/RHR RCP S£AL r:iuw> 2 COOi.iNG (SHEIT 1) NOTE 1: POWER REMOVED Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.3.2-6 EMERGENCY CORE COOLING SYSTEM PROCESS FLOW DIAGRAM SHEET 3 r---------,---...,.1---~ 'IEMf IMII ~II I CS COLD LEu 211!

                                                                                                   ~N 2 SU~PtY I

I I I

                                                                                        ., T I

RCS C0'.0 I I IITOROTEST l[G 29C I l'IM' IRtl ORC I Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.3.2-8 RHR PUMP PERFORMANCE CURVE

                                                   ~ - - - - ~ TOTAL. HEAD 300
                        ....w w

IJ,. 2 0 ct

                        ~
                        %     200 i
                        -I                                                                I  ,
                                                                                               -1 0  L.---____.1----_.....,_____'--________________

O tOOO 2000 3000 4000 5000 GALLONS PE:R MlNlJTE Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.3.2-9 CHG PUMP PERFORMANCE CURVE 7000 6000 5000 I-w

w u.

z 4000 C w 3000 J::

                             ..I g;

I- 2000 zw a: w LL LL. 1000 0 0

                                 -1000
                                 .2000 a 100      200        300          -400 500 ,600 700 U.S. GALLONS PER M INUTE Amendment 61                                                                                      Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.5.2-2 CONTAINMENT SPRAY PH TIME HISTORY OF CONTAINMENT SUMP & SPRAY CASE 1 FSAR Case 1- Mfnimum pH Z7" Nil-OH, Minimum Spray Flow (One CTTral n RUnntng. 5p'11't' 'A.' In S!r,tlce) 11 I I tnJecUon I Reciro.Jl~tron I t 1 I I 10

1 I 1 I 1 g,

I I Sptav'A'pH ' 19.107 I I I

                                                     ,r                                                          Sump pH                 _J   a.sos...

I 8.276

i. B

                               .,      f
                                            ~

7.146 I I I I I

                                                   'I                                                                                      I I

I t I l 5 I 1Switchover End o.l NaOH lnjedlan, 1 4 I {t,.1,973 s.e C) I I (t: 17,538 $,:c) l l (I 100 150 41]0 UD ~00 350

SumppH
1.479 *I 30 da'I' (not plott11d) d'- 10 I.... lime (mlr:Mft1) 111111 pro6.1'11on of ~*11!1 from ndlcilyll~

Amendment 61 Page 1 of 1

Shearon Harris Nuclear Power Plant UFSAR Chapter: 6 FIGURE 6.5.2-3 CONTAINMENT SPRAY PH TIME HISTORY OF CONTAINMENT SUMP & SPRAY CASE 2 Amendment 63 Page 1 of 1}}

ML20147A023

Text

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

REFERENCES:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Navigation menu

Search