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{{#Wiki_filter:CHAPTER 3TABLE OF CONTENTS (Cont'd)SectionTitlePage3.1.24CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL3.1-16 SYSTEMS3.1.25CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS FOR3.1-16 REACTIVITY CONTROL MALFUNCTIONS3.1.26CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY3.1-17 AND CAPABILITY 3.1.27CRITERION 27 - COMBINED REACTIVITY CONTROL SYSTEMS3.1-17 CAPABILITY 3.1.28CRITERION 28 - REACTIVITY LIMITS 3.1-183.1.29CRITERION 29 - PROTECTION AGAINST ANTICIPATED3.1-19 OPERATIONAL OCCURRENCES3.1.30CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE3.1-19 BOUNDARY 3.1.31CRITERION 31 - FRACTURE PREVENTION OF REACTOR COOLANT3.1-20 PRESSURE BOUNDARY3.1.32CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESSURE3.1-21 BOUNDARY 3.1.33CRITERION 33 - REACTOR COOLANT MAKEUP3.1-223.1.34CRITERION 34 - RESIDUAL HEAT REMOVAL3.1-223.1.35CRITERION 35 - EMERGENCY CORE COOLING3.1-233.1.36CRITERION 36 - INSPECTION OF EMERGENCY CORE COOLING3.1-24 SYSTEM 3.1.37CRITERION 37 - TESTING OF EMERGENCY CORE COOLING3.1-24 SYSTEM 3.1.38CRITERION 38 - CONTAINMENT HEAT REMOVAL3.1-253.1.39CRITERION 39 - INSPECTION OF CONTAINMENT HEAT REMOVAL3.1-26 SYSTEM 3.1.40CRITERION 40 - TESTING OF CONTAINMENT HEAT REMOVAL3.1-26 SYSTEM 3.1.41CRITERION 41 - CONTAINMENT ATMOSPHERE CLEANUP3.1-273-iiAm. 3-7/85
CHAPTER 3 TABLE OF CONTENTS (Cont'd) Section Title Page 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.2-1 3.2.1 SEISMIC CLASSIFICATION 3.2-1 3.2.1.1 Seismic and Non-Seismic Interfaces 3.2-2 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATION 3.2-4 3.2.2.1 Quality Group A 3.2-4 3.2.2.2 Quality Group B 3.2-4 3.2.2.3 Quality Group C 3.2-5 3.2.2.4 Quality Group D 3.2-5 3.3 WIND AND TORNADO LOADINGS 3.3-1 3.3.1 HURRICANE WIND CRITERIA 3.3-1 3.3.2 TORNADO CRITERIA 3.3-1 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.4.1 FLOOD ELEVATIONS 3.4-1 3.4.2 PHENOMENA CONSIDERED IN DESIGN LOAD CALCULATIONS 3.4-1 3.4.3 FLOOD FORCE APPLICATION 3.4-1 3.4.4 FLOOD PROTECTION 3.4-1 3.5 MISSILE PROTECTION 3.5-1 3.5.1 MISSILE BARRIERS AND LOADINGS 3.5-1 3.5.2 MISSILE SELECTION 3.5-2 3.5.2.1 Internal Missiles 3.5-2 3.5.2.2 External Missiles 3.5-3 3.5.3 SELECTED MISSILES 3.5-3 3.5.3.1 Internal Missiles 3.5-3
UNIT 1 3-iv Amendment No. 27 (04/15)
CHAPTER 3 TABLE OF CONTENTS (Cont'd) Section Title Page 3.7.1.5 Soil-Structure Interaction 3.7-5 3.7.2 SEISMIC SYSTEM ANALYSIS 3.7-8 3.7.2.1 Method of Analysis 3.7-8 3.7.2.2 Vertical Analysis 3.7-15 3.7.2.3 Torsional Modes of Vibration 3.7-15 3.7.2.4 Comparison of Model Analysis and Time History Methods 3.7-15 3.7.2.5 Overturning Moments 3.7-15 3.7.2.6 Results of Analyses 3.7-16 3.7.2.7 Computer Programs Utilized for Structural and Seismic Analyses 3-7-16 3.7.3 SEISMIC SUBSYSTEM ANALYSIS 3.7-18 3.7.3.1 Seismic Input Data 3.7-18 3.7.3.2 Seismic Analysis - Reactor Coolant System 3.7-18 3.7.3.3 Seismic Analysis - Reactor Internals & Core 3.7-25 3.7.3.4 Method of Analysis - Other-Seismic Class I Systems 3.7-33 3.7.3.5 Torsional Effects of Valves 3.7-40 3.7.3.6 Differential Movement of Piping Supports 3.7-40 3.7.3.7 Interaction With Non-Class I Systems 3.7-41 3.7.3.8 Field Location of Seismic Restraints and Supports 3.7-41 3.7.3.9 Reactor Building Crane Restraints 3.7-41 3.7.3.10 Additional Computer Codes Used in Piping System Stress 3.7-41a Analysis and Support Design 3.7.4 SEISMIC INSTRUMENTATION PROGRAM 3.7-42 3.7.4.1 Compliance with Regulatory Guide 1.12 3.7-42 3.7.4.2 Location and Description of Instrumentation 3.7-43
3-vi Amendment No. 26 (11/13)
DESIGN CRITERIA - STRUCTURES, COMPONENTS, EQUIPMENT AND SYSTEMS CHAPTER 3 LIST OF TABLES TABLE TITLE PAGE 3.2-1 Design Classifications of Structures, Systems & Components 3.2-6 3.2-2 Minimum Code Requirements for Quality Groups 3.2-11 3.3-1 Tornado Wind Speeds and Resulting Static Pressure Loadings 3.3-5 3.5-1 Internal Missile Parameters 3.5-17 3.5-1A Characteristics of Tornado Generated Missiles 3.5-19 3.5-1B Missile Data 3.5-20 3.5-2 Deleted 3.5-3 Deleted 3.6-1 Sensitivity Study for 20" FW Line Outside Containment 3.6-35 3.6-2 Summary of Pipe Whip Analysis 3.6-36 3.6-3 Main Steam Line Inside the Containment Guillotine Break at 3.6-37 Node No. 12 3.6-4 Main Steam Line Inside the Containment Guillotine Break at 3.6-37 Node No. 16 3.6-5 Boiler Feedwater Line Outside the Containment Guillotine Break at 3.6-38 Node No. 7 3.6-6 Boiler Feedwater Line Inside the Containment Guillotine Break at 3.6-38 Node No. 7 3.7.1 Reactor Building Properties Horizontal Model 3.7-48 3.7-lA Soils-Supported Seismic Class I Structures 3.7-49 3.7-lB Side Spring Constants Effect on Soil Structure Interaction 3.7-50 3.7-2 Reactor Building Properties Vertical Model 3.7-51
UNIT 1 3-xii Amendment No. 27 (04/15) CHAPTER 3LIST OF TABLES (Cont'd)TABLETITLEPAGE3.7-3Reactor Auxiliary Building Properties Horizontal Model3.7-523.7-4Reactor Auxiliary Building Properties Vertical Model3.7-533.7-5Fuel Handling Building Properties Horizontal Model3.7-543.7-6Intake Structure Properties Horizontal Model3.7-553.7-7Diesel Generator Building Properties Horizontal Model3.7-563.7-8Comparison of Structural Responses for Response Spectra3.7-57and Time History Seismic Analysis Methods3.7-9Reactor Building Natural Periods of Vibration3.7-583.7-10Reactor Building Horizontal Structural Responses3.7-113.7-11Reactor Building Vertical Structural Responses3.7-603.7-12Reactor Auxiliary Building Natural Periods of Vibration3.7-613.7-13Reactor Auxiliary Building Horizontal Structural Responses 3.7-623.7-14Reactor Auxiliary Building Vertical Structural Responses3.7-633.7-15Fuel Handling Building Natural Periods of Vibration3.7-643.7-16Fuel Handling Building Horizontal Structural Responses3.7-653.7-17Fuel Handling Building Vertical Structural Responses3.7-663.7-18Intake Structure Natural Periods of Vibration3.7-673.7-19Intake Structure Structural Responses3.7-683.7-20Diesel Generator Building Natural Periods of Vibration3.7-693.7-21Diesel Generator Building Structural Responses3.7-703.7-21AComparison Between FRQA and Stardyne3.7-713.7-21BComparison Between RFRM and Other References3.7-723.7-22Natural Frequencies and Dominant Degrees of Freedom3.7-823-xiiiAm. 2-7/84
CHAPTER 3LIST OF TABLES (Cont'd)TABLETITLEPAGE3H-2Allowable Stress and Displacement Criteria -3H-22 Steel Beam Column Assembly3H-3Load Condition 3H-233H-4Comparison of Load Upon Which Radial Stiffnesses Were 3H-24 Based Versus Horizontal Model Forces3-xviiAm. 3-7/85
CHAPTER 3LIST OF FIGURES (Cont'd)FigureTitle3.6-52Main Steam Line Inside Containment Guillotine Break at Node No. 123.6-53Main Steam Line Inside Containment GuillotineBreak at Node No. 163.6-54Boiler Feedwater Line Outside Containment Guillotine Break at Node 73.6-55Boiler Feedwater Line Inside ContainmentGuillotine Break at Node 73.6-56Blowdown Data for Main Steam and Feedwater Lines3.6-57Main Steam Line Inside Containment (Break at Node 16)Reactions at Pipe Whip Restraints (Nodes 13 & 9)Versus Time3.6-5834" Main Steam Line Inside Containment (Break at Node 12)Reaction at Pipe Whip Restraint at Node 9 IN + X & + Z Directions Versus Time3.6-59FW Line Inside Containment Reaction at Pipe WhipRestraint at Node 6 vs. Time3.6-60Main Steam Line Inside Containment Break at Node 16 Displacement of Node 16 Versus Time Max.Plastic Strain = 03.6-6134" Main Stream Line Inside Containment (Break at Node 12) Deflection of Node 12 Versus Time Max. Plastic Strain = .007623.6-6220" F Line Inside Containment Rotation of Node 7 vs. Time3.6-6320" F Line Outside Containment Reactions at Pipe WhipRestraint at Node 5 vs. Time3.6-6420" F Line Outside Containment Displacement at Node 7vs.Time3-xx CHAPTER 3LIST OF FIGURES (Cont'd)FigureTitle3.6-6520" F Line Outside Containment Reactions at Pipe WhipRestraint at Node 5 Versus Time3.6-6620" F Line Outside Containment Displacement of Node 7Verses Time3.6-6720" F Line Outside Containment Reactions at Pipe WhipRestraint at Node 5 Verses Time3.6-6820" F Line Outside Containment Displacement of Node 7Versus Time3.7-1Operating Basis Earthquake Design Response (Housner) Spectrum3.7-2Design Basis Earthquake Design Response (Housner) Spectrum 3.7-3Synthetic Earthquake Intensity Function 3.7-4Comparison of Time History and Housner Response Spectra 3.7-5Reactor Building Horizontal Mathematical Model3.7-6Reactor Building Vertical Mathematical Model3.7-7Reactor Auxiliary Building Horizontal Mathematical Model 3.7-8Reactor Auxiliary Building Vertical Mathematical Model 3.7-9Fuel Handling Building Mathematical Models 3.7-10Intake Structure Mathematical Models3.7-11Diesel Generator Building Mathematical Models3.7-11AComparison Between FRQA and Stardyne 3.7-12Reactor Building EL 18.0 Horizontal Floor Response Spectra 3.7-13Reactor Building EL 24.0 Horizontal Floor Response Spectra3-xxi CHAPTER 3LIST OF FIGURES (Cont'd)FigureTitle3.7-14Reactor Building EL 44.0 Horizontal Floor Response Spectra3.7-15Reactor Building EL 60.0 Horizontal Floor Response Spectra3.7-16Reactor Building EL 68.5 Horizontal Floor Response Spectra3.7-17Reactor Building Vertical Floor Response Spectra 3.7-18Reactor Aux. Bldg. El-0.5 Horizontal Floor Response Spectra 3.7-19Reactor Aux. Bldg. El-19.5 Horizontal Floor Response Spectra 3.7-20Reactor Aux. Bldg. El-43.0 Horizontal Floor Response Spectra 3.7-21Reactor Aux. Bldg. El-62.0 Horizontal Floor Response Spectra3.7-22Reactor Aux. Bldg. El-82.0 Horizontal Floor Response Spectra3.7-23Reactor Aux. Bldg. El 62.0, 43.0, 19.5, -0.5, -10.0 VerticalFloor Response Spectra3.7-24Reactor Aux. Bldg. El 82.0 Vertical Floor Response Spectra 3.7-25Typical Reactor Coolant System Seismic Analysis Model 3.7-26Reactor and Internals Seismic Analysis Model3.7-27Surge Line Seismic Analysis Model3.7-28aOriginal Pressurizer Seismic Analysis Model 3.7-28bReplacement Pressurizer Seismic Analysis Model 3.7-29Representative Node Locations Horizontal Mathematical Model 3.7-30Mathematical Model Horizontal Seismic Analysis3.7-31Mathematical Model Vertical Seismic Analysis3.7-32Core Support Barrel Upper Flange Finite Element Model3-xxiiAmendment No. 21 (12/05)
CHAPTER 3LIST OF FIGURES (Cont'd)FigureTitle3H-7Reactor Building Internal Concrete - Plans and Sections Masonry - Sheet 23H-8Air Flow Diagram for Reactor Cavity, Support Leg and Control Rod Drive Mechanism Cooling Systems3H-9Reactor Cavity Reinforcing - Sheet 13H-10Reactor Cavity Reinforcing - Sheet 23H-11Reactor Cavity Reinforcing - Sheet 3 3H-12MOHR's Failure Envelope Plain Concrete3H-13Reactor Internal Structure Isometric Section3H-14Reactor Support Main Horizontal Model 3H-15Vertical Finite Element Model 3H-16Lower Mass Concrete Boundary Conditions 3H-17Plan-Horizontal Finite Element Model for Lower Ring 3H-18Reactor Cavity Wall Displacements and Forces - Sheet 13H-19Reactor Cavity Wall Displacement and Forces - Sheet 63H-20Reactor Cavity Wall Displacement and Forces - Sheet 5 3H-21Reactor Cavity Wall Displacements and Forces 3H-22Reactor Cavity Wall Displacement and Forces - Sheet 3 3H-23Reactor Cavity Wall Displacement and Forces Sheet 23H-24Reactor Cavity Wall Principal Stress Plot - Sheet 13H-25Reactor Cavity Wall Principal Stress Plot - Sheet 2 3-xxviii Am. 3-7/85
3.1.2CRITERION 2 - DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENAStructures, systems and components important to safety shall be designed to withstand the effects ofnatural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems and components shall reflect: (1) appropriate consideration of the most severe of natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time which the historical data have been accumulated, (2)appropriate combinations of the effects of normal and accident conditions with the effects of thenatural phenomena and (3) the importance of the safety functions to be performed.DISCUSSIONThe structures, systems and components important to safety are designed to withstand the effects ofnatural phenomena without loss of capability to perform their safety functions. Natural phenomenafactored into the design of plant structures, systems and components important to safety aredetermined from recorded data for the site vicinity with appropriate margin to account for uncertainties in historical data.The most severe natural phenomena postulated to occur at the site in terms of induced stresses isthe design basis earthquake (DBE). Those structures, systems, and components vital for themitigation and control of accident conditions are designed to withstand the effects of a loss of coolantaccident (LOCA) coincident with the effects of the DBE. Structures, systems and components vital to the safe shutdown of the plant are designed to withstand the effects of any one of the most severe natural phenomena, including flooding, hurricanes, tornadoes and the DBE.Design criteria for wind and tornado, flood and earthquake are discussed in Section 3.3, 3.4 and 3.7respectively.3.1.3CRITERION 3 - FIRE PROTECTIONStructures, systems and components important to safety shall be designed and located to minimize,consistent with other safety requirements, the probability and effect of fires and explosions.Noncombustible and heat resistant materials shall be used wherever practical throughout the unit,particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems and components important to safety. Fire fighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems and components.3.1-2 Am. 3-7/85
In accordance with NRC staff requirements, a missile protected inter-tie is provided between the Unit 1 auxiliary feedwater pump suction lines and the Unit 2 condensate storage tank (CST) to be used under administrative control. To add to the systems operational flexibility, the provision to supply the Unit 2 auxiliary feedwater pumps from the Unit 1 condensate storage tank is also provided. To prevent inadvertent draining of the Unit 2 CST to the Unit 1 CST, plant procedures for placing the inter-tie in service require that the Unit 1 CST outlet isolation valves be closed prior to placing the inter-tie line in service. This helps to assure that the water level in the Unit 2 CST is maintained at the minimum value required for safe shutdown. In the unlikely event of a loss of offsite power, both St. Lucie Units 1 and 2 have their own 100 percent capacity redundant diesel generator sets which are available for safe shutdown.
In the unlikely event of a station blackout in one unit, i.e., total loss of AC power on-site and off-site, both units can be electrically connected, under administrative control, such that a diesel generator set from the non-blacked out unit is able to provide power to the minimum loads required to maintain both units in a hot standby condition. The ultimate heat sink (a safety related structure) supplies emergency cooling water to both St. Lucie Units 1 and 2. The canal has sufficient cross-sectional water flow area to mitigate the consequences of a LOCA on one unit while safely shutting down the other unit.
UNIT 1 3.1-5 Amendment No. 27 (04/15) THIS PAGE INTENTIONALLY BLANK3.1-6Amendment No. 16, (1/98) 3.1.10CRITERION 10 - REACTOR DESIGNThe reactor core and associated coolant, control and protection systems shall be designed withappropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.DISCUSSIONIn ANSI-N 18.2, plant conditions have been categorized in accordance with their anticipatedfrequency of occurrence and risk to the public, and design requirements are given for each of the fourcategories. These categories covered by this criterion are Condition I - Normal Operation and Condition II - Faults of Moderate Frequency.The design requirement for Condition I is that margin shall be provided between any plant parameterand the value of that parameter which would require either automatic or manual protective action; it ismet by providing an adequate control system (refer to Section 7.7). The design requirement forCondition II is that such faults shall be accommodated with, at most, a shutdown of the reactor, with the plant capable of returning to operation after corrective action; it is met by providing an adequate protective system. (refer to Section 7.2 and Chapter 15)Specified acceptable fuel design limits are stated in Section 4.4. Minimum margins to specifiedacceptable fuel design limits are prescribed in the Technical Specifications (Limiting Conditions forOperations) which support Chapters 4 and 15 of the Safety Analysis Report. The plant is designed such that operation within Limiting Conditions for Operation, with safety system settings not less conservative than Limiting Safety System Settings prescribed in the Technical Specifications, assures that specified acceptable fuel design limits will not be violated as a result of anticipated operationaloccurrences. During non-accident conditions, operation of the plant within Limiting Conditions forOperation ensures that specified acceptable fuel design limits are not approached within the minimum margins. Operator action, aided by the control systems and monitored by plant instrumentation, maintains the plant within Limiting Conditions for Operation during non-accident conditions.3.1.11CRITERION 11 - REACTOR INHERENT PROTECTIONThe reactor core and associated coolant systems shall be designed so that in the power operatingrange the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.DISCUSSIONIn the power operating range, the combined response of the fuel temperature coefficient, themoderator temperature coefficient, the moderator void coefficient, and the moderator pressure coefficient to an increase in reactor power in the power operating range is a decrease in reactivity; i.e., the inherent nuclear feedback characteristics is negative.3.1-7 The reactivity coefficients are listed in Table 4.3-3 and are discussed in detail in Section 4.3.1.3.1.12CRITERION 12 - SUPPRESSION OF REACTOR POWER OSCILLATIONSThe reactor core and associated coolant, control, and protection systems shall be designed to assurethat power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.DISCUSSIONPower level oscillations do not occur. The effect of the negative power coefficient of reactivity (seeCriterion 11), together with the coolant temperature program maintained by control of regulating rods and soluble boron, provides fundamental mode stability. Power level is continuously monitored by neutron flux detectors (Section 7.2.1.1) and by reactor coolant temperature difference measuringdevices.3.1.13CRITERION 13 - INSTRUMENTATION AND CONTROL Instrumentation shall be provided to monitor variables and systems over their anticipated ranges fornormal operation, for anticipated operational occurrences, and for accident conditions as appropriateto assure adequate safety, including those variables and systems that can affect the fission process,the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.DISCUSSIONInstrumentation is provided, as required, to monitor and maintain significant process variables whichcan affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Controls are provided for the purpose of maintaining these variables within the limits prescribed for safe operation.The principal variables and systems monitored include neutron level -(reactor power); reactor coolanttemperature, flow, and pressure; pressurizer liquid level; steam generator level and pressure; and containment pressure and temperature. In addition, instrumentation is provided for continuous automatic monitoring of process radiation level in the reactor coolant system.3.1-8Amendment No. 21 (12/05)
DISCUSSIONThe containment system is designed to protect the public from the consequences of a LOCA, basedon a postulated break of reactor coolant piping up to and including a double ended break of the largest reactor coolant pipe.The containment vessel, shield building, and the associated engineered safety features systems aredesigned to safely sustain all internal and external environmental conditions that may reasonably beexpected to occur during the life of the plant, including both short and long term effects following aLOCA.Leak tightness of the containment system and short and long term performance following a LOCA areanalyzed in Section 6.2.3.1.17CRITERION 17 - ELECTRICAL POWER SYSTEMSAn on-site electrical power system and an off-site electrical power system shall be provided to permitfunctioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of thereactor coolant pressure boundary are not exceeded as a result of anticipated operationaloccurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.The on-site electrical power sources, including the batteries, and the on-site electrical distributionsystem, shall have sufficient independence, redundancy, and testability to perform their safetyfunctions assuming a single failure.Electrical power from the transmission network to the switchyard shall be supplied by two physicallyindependent transmission lines (not necessarily on separate rights-of-way) designed and located so as to suitably minimize the likelihood of their simultaneous failure under operating and postulatedaccident and environmental conditions. Two physically independent circuits from the switchyard tothe on-site electrical distribution system shall be provided. Each of these circuits shall be designed tobe available in sufficient time following a loss of all on-site alternating current power sources and the other off-site electrical power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss of coolant accident to assure that corecoolant, containment integrity, and other vital safety functions are maintained.Provisions shall be included to minimize the probability of losing electrical power from any of theremaining sources as a result of, or coincident with, the loss of power generated by the nuclear power unit,3.1-11Amendment 15, (1/97)
3.1.20CRITERION 20 - PROTECTION SYSTEM FUNCTIONSThe protection system shall be designed (1) to initiate automatically the operation of appropriatesystems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.DISCUSSIONThe reactor protective system monitors reactor operating conditions and automatically initiates areactor trip when the monitored variable or combination of variables exceeds a prescribed operating range. The reactor trip setpoints are selected to ensure that anticipated operational occurrences do not cause acceptable fuel design limits to be violated. Specific reactor trips are described in Section 7.2.Reactor trip is accomplished by deenergizing the control element drive mechanism holding latch coilsthrough the interruption of the CEDM power supply. The CEA's are thus released to drop into the core reducing reactor power.The engineered safety features actuation system monitors potential accident conditions andautomatically initiates engineered safety features and their supporting systems when the monitoredvariables reach prescribed setpoints. The parameters which automatically actuate engineered safety features are described in Section 7.3. Manual actuation is provided to the operator.3.1.21CRITERION 21 - PROTECTION SYSTEM RELIABILITY AND TESTABILITYThe protection system shall be designed for high functional reliability and inservice testabilitycommensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protectionsystem can be otherwise demonstrated. The protection system shall be designed to permit periodictesting of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.DISCUSSIONThe protection systems are designed to provide high functional reliability and in-service testability bydesigning to the requirements of IEEE 279 - 1971 and IEEE 338 - 1971. No single failure will result in the loss of the protection function. The protection channels are independent with respect to sensors and power supplies, piping, wire routing and mounting. This independence permits testing3.1-14 without loss of the protection function.Each channel of the protection system, including the sensors up-to the final actuation device iscapable of being checked during reactor operation. Measurement sensors of each channel used in protection systems are checked by observing outputs of similar channels which are presented on indicators and recorders in the control room. Trip units and logic are tested by inserting a signal into the measurement channel ahead of the readout and, upon application of a trip level input, observing that a signal is passed through the trip units and the logic to the logic output relays. The logic outputrelays are tested individually for initiation of trip action.Protection system reliability and testability are discussed in Sections7.2.2 and 7.3.2.3.1.22CRITERION 22 - PROTECTION SYSTEM INDEPENDENCEThe protection system shall be designed to assure that the effects of natural phenomena, and ofnormal operating, maintenance, testing and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.DISCUSSIONThe protection systems conform to the provisions of the Institute of Electrical and ElectronicEngineers (IEEE) Criteria for Protection Systems for Nuclear Power Generating Stations, IEEE 279 -1971. Four independent measurement channels complete with sensors, sensor power supplies,signal conditioning units and bistable trip units are provided for each protective parameter monitored by the protection systems. The measurement channels are provided with a high degree of independence by separate connections of the channel sensors to the process systems. Power to the channels is provided by independent emergency power supply buses.The protective system is functionally tested to ensure satisfactory operation prior to installation in theplant. Environmental and seismic qualifications are also performed utilizing type tests and specific equipment tests. (Refer to Section 7.1.2)3.1.23CRITERION 23 - PROTECTION SYSTEM FAILURE MODESThe protection system shall be designed to fail into a safe state or into a state demonstrated to beacceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air) or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.3.1-15 DISCUSSIONProtective system trip channels are designed to fail into a safe state or into a state established asacceptable in the event of loss of power supply or disconnection of the system. A loss of power to the CEDM holding coils results in gravity insertion of the full length CEAS into the core. Redundancy, channel independence, and separation incorporated in the protective system design minimize the possibility of the loss of a protection function under adverse environmental conditions. Refer to Sections 7.2 and 7.3.3.1.24CRITERION 24 - SEPARATION OF PROTECTION AND CONTROL SYSTEMSThe protection system shall be separated from control systems to the extent that failure of any singlecontrol system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact asystem satisfying all reliability, redundancy, and independence requirements of the protection system.Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.DISCUSSIONThe protection systems are separated from the control instrumentation systems so that failure orremoval from service of any control instrumentation system component or channel does not inhibit the function of the protection system.3.1.25CRITERION 25 - PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITYCONTROLMALFUNCTIONSThe protection system shall be designed to assure that specified acceptable fuel design limits are notexceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.DISCUSSIONReactor shutdown with CEA's is accomplished completely independent of the control functions sincethe trip breakers interrupt power to the full length CEA drive mechanisms regardless of existing control signals. The design is such that the system can withstand accidental withdrawal of controlling groups without exceeding acceptable fuel design limits. Analysis of possible reactivity controlmalfunctions is given in Sections 15.2.1 and 15.2.2. The reactor protection system will preventspecified acceptable fuel design limits from being exceeded for any anticipated transients.3.1-16 3.1.26CRITERION 26 - REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITYTwo independent reactivity control systems of different design principles shall be provided. One ofthe systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changesresulting from planned, normal power changes (including xenon burnout) to assure acceptable fueldesign limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.DISCUSSIONTwo independent reactivity control systems of different design principles are provided. The firstsystem, using control element assemblies (CEA's) includes a positive means (gravity) for inserting CEA's and is capable of controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences specified acceptable fuel design limits are not exceeded. The CEA's can be mechanically driven into the core. The appropriate margin for a stuck CEA is provided by assuming in the analyses that the highest worth CEA does not fall into thecore.The second system, using neutron absorbing soluble boron, is capable of Compensating for the rateof reactivity changes resulting from planned normal power changes, (including xenon burnout), such that acceptable fuel design limits are not exceeded.Either system is capable of making the core subcritical from a hot operating condition and holding itsubcritical in the hot standby condition. The soluble boron system is capable of holding.the reactor subcritical under cold conditions-Refer to Section 9.3.4 for details.3.1.27CRITERION 27 - COMBINED REACTIVITY CONTROL SYSTEMS CAPABILITYThe reactivity control systems shall be designed to have a combined capability, in conjunction withpoison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.DISCUSSIONThe reactivity control systems provide the means for making and holding the core subcritical underpostulated accident conditions, as discussed in Sections 9.3.4 and 4.3. Combined use of CEA's and soluble boron control by the chemical and volume control system provides the shutdown marginrequired for plant cooldown and long term xenon decay, assuming3.1-17Amendment 15, (1/97)
3.1.29CRITERION 29 - PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCURRENCESThe protection and reactivity control systems shall he designed to assure an extremely highprobability of accomplishing their safety functions in the event of anticipated operational occurrences.DISCUSSION Plant conditions designated as Condition I and Condition II in ANS-N 18.2 have been carefullyconsidered in the design of tile reactor protective system and the reactivity control systems.Consideration of redundancy, independence and testability in the design, coupled with careful component selection, overall system testing, and adherence to detailed quality assurance, assure an extremely high probability that safety functions are accomplished in the event of anticipated operational occurrences.3.1.30CRITERION 30 - QUALITY OF REACTOR COOLANT PRESSURE BOUNDARYComponents which are part of the reactor coolant pressure boundary shall be designed, fabricated,erected and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.DISCUSSIONThe reactor coolant pressure boundary components are designed, fabricated, erected and tested inaccordance with the codes and standards specified in Criterion 14.Containment sump instrumentation is used to detect reactor coolant system leakage (Section 5.2.4)by providing information on sump levels and frequency of sump pump operation. Flow instrumentation indicate and record make-up flow rate and volumes from the primary water system. This instrumentation allows detection of suddenly occurring leaks or those which are gradually increasing. Containment radiation monitors (Section 12.1) provide an additional means of reactor coolant system leakage detection.3.1-19Am. 3-7/85
During normal start-up for power operation, the reactor will not be made critical until the reactorcoolant system temperature is at least 120F greater than the predicted nil ductility transitiontemperature based on plant records of fast neutron dose to the vessel. The stress criteria include themaximum loads associated with the most severe transients during emergency conditions at operatingtemperature. This will assure that a reactivity-induced loading which would contribute to elastic orplastic deformation cannot occur below a reactor operating temperature corresponding to NDTT+120F.The activation of the safety injection systems will introduce highly borated water into the primarysystem at pressures significantly below operating pressures and will not cause adverse pressure or reactivity effects.The thermal stresses induced by the injection of cold water into the vessel have been examined.Analysis shows that there is no gross yielding across the vessel wall using the minimum specified yield strength in the ASME Boiler and Pressure Vessel Code, Section III.Adverse effects that could be caused by exposure of equipment or instrumentation to containmentspray water is avoided by designing the equipment or instrumentation to withstand direct spray or bylocating it or protecting it to avoid direct spray.3.1.32CRITERION 32 - INSPECTION OF REACTOR COOLANT PRESSURE BOUNDARY Components which are part of the reactor coolant pressure boundary shall be designed to permit (1)periodic inspection and testing of important areas and features to assess their structural and leaktightintegrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.DISCUSSIONProvisions are made for inspection, testing, and surveillance of the reactor coolant system boundaryas described in Section 5.2.5.The reactor vessel material surveillance program described in Section 5.4.4 conforms with ASTM-E-185-66. Sample pieces taken from the same shell plate material used in fabrication of the reactor vessel are installed between the core and the vessel inside wall. These samples will be removed andtested at intervals during vessel life to provide an indication of the extent of the neutron embrittlementof the vessel wall. Charpy tests will be performed on the samples to develop a Charpy transition curve. By comparison of this curve with the Charpy curve and drop weight tests on specimens taken at the beginning of the vessel life, the change of NDTT will be determined and operating instructions adjusted as required.3.1-21 Amendment No. 20 (4/04)
3.1.36CRITERION 36 - INSPECTION OF EMERGENCY CORE COOLING SYSTEMThe emergency core cooling system shall be designed to permit periodic inspection of importantcomponents, such as spray rings in the reactor containment vessel, water injection nozzles, and piping to assure the integrity and capability of the system.DISCUSSIONThe capability for periodic inspection of important components of the Emergency Core CoolingSystem (safety injection system) is provided to the extent practicable through the arrangement and location of the components of the system. System components external to the containment structure are accessible for physical inspection at any time. All components (valves and piping) inside the containment and the safety injection tanks can be inspected during refueling. See Section 6.3.3.1.37CRITERION 37 - TESTING OF EMERGENCY CORE COOLING SYSTEMThe emergency core cooling system shall be designed to permit appropriate periodic pressure andfunctional testing to assure: (1) the structural and leaktight integrity of its components, (2) theoperability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the fulloperational sequence that brings the system into operation, including operation of applicable portionsof the protection systems that transfer between normal and emergency power sources, and theoperation of the associated cooling water system.DISCUSSIONThe emergency core cooling system (safety injection system) is provided with testing facilities todemonstrate system component operability. Testing can be conducted during normal plant operation with the test facilities arranged not to interfere with the performance of the systems or with the initiation of control circuits.The safety injection system is designed to permit periodic testing of the delivery capability up to alocation as close to the core as practical. Periodic pressure testing of the safety injection system is possible using the cross connection to the charging pumps in the chemical and volume control system.The low pressure safety injection pumps are used as shutdown cooling pumps during normal plantcooldown. The pumps discharge into the safety injection header via the shutdown heat exchangersand the low pressure injection lines.3.1-24Amendment No. 17 (10/99) With the plant at operating pressure, operation of the safety injection pumps and valves may beverified by recirculation back to the refueling water tank. This will permit verification of flow pathcontinuity in the high pressure injection lines and suction lines from the refueling water tank.Borated water from the safety injection tanks may be bled through the recirculation test line to verifyflow path continuity from each tank to its associated main safety injection header.The operational sequence that brings the safety injection system into action, including transfer toalternate power sources, can be tested in parts as described in Sections 7.3.2.3.1.38CRITERION 38 - CONTAINMENT HEAT REMOVAL A system to remove heat from the reactor containment shall be provided. The system safety functionshall be to reduce rapidly, consistent with the functioning of other associated systems, thecontainment pressure and temperature following any loss of coolant accident and maintain them atacceptably low levels.Suitable redundancy in components and features, and suitable interconnections, leak detection,isolation, and containment capabilities shall be provided to assure that for on-site electric power system operation (assuming off-site power is not available) and for off-site electric power systemoperation (assuming on-site power is not available) the system safety function can be accomplished,assuming a single failure.DISCUSSIONThe containment heat removal system described in Section 6.2.2 consists of the containment spraysystem and the containment cooling system. The containment spray system consists of two redundant subsystems each containing a containment spray pump, shutdown heat exchanger and spray header. The containment cooling system consists of four fan coolers. The containment spraysystem and the containment cooling system are each designed with the capacity to reducecontainment pressure and temperature following a LOCA and maintain them at acceptably low levels.Both the containment spray and the containment cooling systems are provided with emergency on-site power necessary for their operation, assuming a loss of off-site power. The systems together provide a minimum of 100 percent containment cooling capability assuming a single failure in eithersystem or in the emergency on-site power supply.3.1-25Amendment No. 16, (1/98)
DISCUSSIONThe cooling water systems which function to remove the combined-heat load from structures,systems and components important to safety under normal operating and accident conditions, are the component cooling system and the intake cooling water system.The component cooling system is a closed loop system which removes heat from the shutdown heatexchangers, containment cooling system and other essential and nonessential components asdescribed in Section 9.2.2. The system consists of three pumps with two heat exchangers, piping ,valves and instrumentation arranged in two essential headers and one nonessential header. Two essential headers serve redundant safety related components. Only one essential header is needed to remove the heat generated under post-LOCA conditions.The intake cooling water system is an open loop system which removes heat from the componentcooling system and transfers it to the ultimate heat sink as described in Section 9.2.1. The systemconsists of three pumps with piping, valves and instrumentation arranged in two essential headers, one to each component cooling heat exchanger, and branches to two non-essential headers which supply water to the turbine cooling water heat exchangers, which are isolated automatically upon receiving SIAS. Only one essential header is needed to remove the heat generated under post-LOCA conditions.The intake cooling water pumps normally take water from the Atlantic Ocean through the circulatingwater intake conduits and canal. In the event of interruption of water from this source, water is taken through the emergency cooling water canal from Big Mud Creek which serves as the ultimate heat sink. The ultimate heat sink is discussed in Section 9.2.7.The piping, valves, pumps and heat exchangers in each system are designed and arranged so thatthe safety function can be performed assuming a single failure. The essential headers of each system will each be isolated from the nonessential header during the emergency mode of operation.Electrical power for the operation of each system may be supplied from offsite or onsite emergencypower sources, with distribution arranged such that a single failure will not prevent the system fromperforming its safety function.3.1.45CRITERION 45 - INSPECTION OF COOLING WATER SYSTEM The cooling water system shall be designed to permit periodic inspection of important components,such as heat exchangers and piping, to assure the integrity and capability of the system.DISCUSSION The component cooling system and intake cooling water system are designed to permit periodicinspection, to the extent practical of important components, such as heat exchangers, pumps, valvesand accessible piping. Each system is normally pressurized permitting leakage detection by routine surveillance or monitoring instrumentation. Refer to Sections 9.2.1.4 and 9.2.2.4.3.1-29 3.1.46CRITERION 46 - TESTING OF COOLING WATER SYSTEMThe cooling water system shall be designed to permit appropriate periodic pressure and functionaltesting to assure (1) the structural and leaktight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss of coolant accidents, including operation of applicable portions of the protection system and the transfer between normaland emergency power sources.DISCUSSIONBoth the component cooling and intake cooling water systems are in operation during normal plantoperation or shutdown. The structural and leaktight integrity of the component cooling and intake cooling water systems components are demonstrated in this way. Pumps and heat exchangers areoperated as dictated by plant operational modes and tested on a schedule basis to monitoroperational capability of redundant components. Data can be taken periodically during normal plant operation to confirm heat transfer capabilities. Refer to Sections 9.2.1.4 and 9.2.2.4.The systems are designed to permit testing of system operability encompassing simulation ofemergency reactor shutdown or LOCA conditions including the transfer between normal andemergency power sources.3.1-30Amendment No. 17 (10/99)
DISCUSSIONThe material selected for the containment vessel is carbon steel (ASTM-A516 Grade 70) normalizedto refine the grain which results in improved ductility. In addition, the actual mechanical and chemical properties of the material are documented and are within the limits for minimum ductility defined in ASTM-A516.The containment vessel was built to Subsection B of the 1968 edition of Section III of the ASMEBoiler and Pressure Vessel Code, and in accordance with this Code the materials including weldspecimens were impact tested at a temperature at least 30 F below the lowest metal service temperature.The design of the vessel reflects consideration of all ranges of temperature and loading conditionswhich apply to the vessel during operation, maintenance, testing and postulated accident conditions.All seam welds in the vessel have been 100 percent radiographed and the acceptance standards ofthe radiographs ensured that flaws in welds did not exceed the maximum allowed by ASME Code.Since this vessel has been post weld heat treated, residual stresses from welding will be minimal.Steady state and transient stresses have been calculated in accordance with accepted methods.Refer to Section 3.8.2.3.1.52CRITERION 52 - CAPABILITY FOR CONTAINMENT LEAKAGE RATE TESTING The reactor containment and other equipment which may be subjected to containment test conditionsshall be designed so that periodic integrated leakage rate testing can be conducted at containmentdesign pressure.DISCUSSION The containment vessel has been designed so that initial integrated leak rate testing can beperformed at design pressure after completion and installation of penetrations and equipment.Provisions have been made in the containment design to permit periodic leakage rate tests to verifythe continued leak-tight integrity of the containment. Refer to Sections 6.2.1.4 and 16.4.4.3.1-32 3.1.53CRITERION 53 - PROVISIONS FOR CONTAINMENT TESTING AND INSPECTIONThe reactor containment shall be designed to permit (1) appropriate periodic inspection of allimportant areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak tightness of penetrations which have resilient seals and expansion bellows.DISCUSSIONThe absence of insulation on the containment vessel permits periodic inspection of the exposedinterior surfaces of the vessel. The lower portions of the containment vessel are totally encased in concrete and will not be accessible for inspection after the acceptance testing. It is contemplated that there will be no need for any special in-service surveillance program due to the rigorous design, fabrication, inspection and pressure testing the containment vessel receives prior to operation.Provisions have been made to permit periodic testing at containment design pressure of penetrationswhich have resilient seals or expansion bellows to allow leak tightness to be demonstrated. Refer to Section 6.2.1.4.3.1.54CRITERION 54 - PIPING SYSTEMS PENETRATING CONTAINMENTPiping systems penetrating primary reactor containment shall be provided with leak detection,isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associatedapparatus and to determine if valve leakage is within acceptable limits.DISCUSSION Piping penetrating the containment vessel shell is designed to withstand at least a pressure equal tothe containment vessel maximum internal pressure. The isolation system design requires a doublebarrier on all of the above systems not serving accident consequence limiting systems so that nosingle active failure can result in loss of isolation or intolerable leakage. These lines are provided with isolation valves as indicated in Section 6.2.4.2.Valves isolating penetrations serving engineered safety features systems will not automatically closewith a containment isolation signal (CIS), but may be closed by remote manual operation from thecontrol room to isolate any engineered safety feature when required.3.1-33 Proper valve closing time is achieved by appropriate selection of valve, operator type and operatorsize. Refer to Table 6.2-16 for additional isolation valve information.To ensure continued integrity of the containment isolation system, periodic closure and leakage testsshall be performed as stated in Section 6.2.4.4 and Technical Specification 3/4.6.1.3.1.55CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY PENETRATINGCONTAINMENTEach line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:1)one locked closed isolation valve inside and one locked closed isolation valve outsidecontainment, or2)one automatic isolation valve inside and one locked closed isolation valve outside containment,or3)one locked closed isolation valve inside and one automatic isolation valve outside containment.A simple check valve may not be used as the automatic isolation valve outside containment, or4)one automatic isolation valve inside and one automatic isolation valve outside containment. Asimple check valve may not be used as the automatic isolation valve outside containment.Isolation valves outside containment shall be located as close to the containment as practical andupon loss of actuating power, automatic isolation valves shall be designed to take the position thatprovides greater safety.Other appropriate requirements to minimize the probability or consequences of an accidental ruptureof these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for in-service inspection, protection against more severenatural phenomena, and additional isolation valves and containment, shall include consideration ofthe population density, use characteristics, and physical characteristics of the site environs.DISCUSSION Except for the safety injection lines, the reactor coolant pressure boundary as defined in 10 CFR 50 islocated within the containment. The safety injection lines are closed seismic Class I piping systemsoutside containment with isolation valves as indicated in Table 6.2-16. Provisions are made for leaktesting as described in Section 6.2.1. Isolation valves are located as close to the containment as practical.3.1-34Amendment No. 16, (1/98)
to the containment atmosphere shall have at least one containment isolation valve which shall beeither automatic, or locked closed, or capable of remote manual operation. This valve shall beoutside containment and located as close to the containment as practical. A simple check valve maynot be used as the automatic isolation valve.DISCUSSION Except for the shutdown cooling lines each line that penetrates the reactor containment and is neitherpart of the reactor coolant pressure boundary nor connected directly to the containment atmospherehas at least one containment isolation valve located outside the containment as close to the containment as practical. Refer to Table 6.2-16.The shutdown cooling lines arrangement of two locked closed valves, located inside containmentmeets the intent of GDC 57 since no single failure will prevent the recirculation of core cooling wateror will adversely affect the integrity of the containment since the system is designed as seismic ClassI inside containment and as a closed seismic Class I system outside containment.3.1-36
1)remove directly residual heat from the reactor,2)circulate reactor coolant for any safety system* purpose,3)control within the reactor containment radioactivity released or control hydrogen in the reactorcontainment.3.2.2.3Quality Group CQuality Group C applies to those components not in Quality Group A or B whose failure would resultin significant radioactive release to the environment or that are necessary to: a)provide or support any safety system* functionb)control accident airborne radioactivity outside the reactor containment.3.2.2.4Quality Group DQuality Group D applies to those components not related to nuclear safety.*A safety system (in this context) is any system that functions to shut down the reactor, cool thecore or cool another safety system or the containment, and contains, controls, or reducesradioactivity released in an accident. Only those portions of the secondary systems are included (a) that are designed primarily to accomplish one of the above safety functions or (b) whosefailure could prevent accomplishing one of the above functions.3.2-5Amendment No. 17 (10/99) TABLE 3.2-1DESIGN CLASSIFICATIONS OF STRUCTURES, SYSTEMS AND COMPONENTSTORNADOSEISMICWINDFLOODMISSILEQUALITYSTRUCTURECLASSCRITERIONCRITERIONCRITERIONGROUPNOTESShield buildingaaa-Containment vesselbcbBReactor buildinginterior structuresbca,b,c-Reactor auxiliary buildingaaa-Diesel generator buildingaaa (8)-Intake structureaaa-Fuel handling buildingaaa-Cask crane support structure a---(11) Supports for Class I equipmenta,ba,b,ca,b,c-(7)UHS Dam (Barrier Wall)a-a-SYSTEMS AND COMPONENTSA.Reactor Coolant System1.Reactor pressure vessel bc b A2.Reactor pressure vesselinternals bc b-3.Control rod drivemechanismsbcb-4.Control element assemblies bcb-5.PressurizerbcbA6.Steam generatora)Primary sidebc bAbSecondary sidebc bB7.Reactor coolant pumps bc bA8.PipingaPart of RCPB bc bA(1)9.Reactor protectioninstrumentation bc b-(2)B.Safety Injection System1Safety Injection Tanks ---B(9)2Refueling Water Tank B(9)3.Pumps bc bB4.Piping and valves (3)a)Part of RCPB bc bA(1)b)Required only forinitial injection bcbBc)Required for long termpost-accident cooling bcbBd)Normally isolated orautomatically isolatedfrom parts of system covered by (a), (b) or (c) ----D(4)5.Instrumentation bcb-(2)C.Shutdown Cooling System1.Heat exchangersa)Reactor coolant side b c b Bb)Component cooling waterside b c b C3.2-6Amendment No. 20 (4/04) TABLE 3.2-1 (Cont'd)TORNADOSEISMICWINDFLOODMISSILEQUALITYSYSTEMS AND COMPONENTSCLASSCRITERIONCRITERIONCRITERIONGROUPNOTES2.Piping and valves(3)a)Part of RCPBbcbA(1)b)Required for residual heatremovalb cbBc)Normally isolated orautomatically isolatedfrom parts of systemcovered by (a) or (b)----D(4)3.Instrumentationbcb-(2)D.Chemical Volume and Control System1.Charging pumpsbcbB2.Boric acid make-up tanksbcbB3.Boric acid pumpsbcbB4.Letdown heat exchangers----B5.Regenerative heat exchangerbcbB6.Volume control tank----B7.Boric acid batching tank----D8.Ion exchangers----B9.Piping and valves(3)a)Part of RCPBbcbA(1)b)Required for letdown----C(12)c)Required for post-accidentinjection of boric acidb cbBd)Normally or automaticallyisolated from parts ofsystem covered by (a), (b) or (c)----D(4)10.Instrumentationbcb-(2)E.Containment Spray System1.Pumps- - -B2.Nozzles- - - B3.Piping and valves(3)a)Required for spray andrecirculation---Bb)Normally or automaticallyisolated from parts of system covered by (a)----D(4)4.Instrumentation----(2)F.Waste Management System1.Reactor coolant drain tank----D2.Flash tank----D3.Reactor drain pumps----D 4.Holdup tanks----D5.Spent resin tank----D6.Flash tank pumps----D 7.Gas surge tank----D8.Waste gas compressors----D9.Gas decay tanks---D3.2-7Amendment No. 21 (12/05)
TABLE 3.2-2 MINIMUM CODE REQUIREMENTS FOR QUALITY GROUPS Component Quality Group A Quality Group B Quality Group C Quality Group D Pressure Vessels ASME Boiler and Pressure Vessel Code, Section III, Class A ASME Boiler and Pressure Vessel Code, Section III, Class C ASME Boiler and Pressure Vessel Code, Section VIII Division 1 ASME Boiler and Pressure Vessel Code, Section VIII, Division I or Equivalent Containment Vessel - ASME Boiler and Pressure Vessel Code, Section 111, Class B - 15 Psig Storage Tanks - API-620 API-620 API-620 or Equivalent Atmospheric Storage Tanks - Applicable Storage Tank Codes such as API-650, AWWAD100 or ANSI B 96.1 Applicable Storage Tank Codes such as API-650 AWWAD100 or ANSI B 96.1 API-650, AWWAD100 or ANSI B 96.1 or Equivalent Piping1 ANSI B 31.7, Class I (1969 Edition) ANSI B 31.7, Class II (1969 Edition) ANSI B 31.7, Class III (1969 Edition) ANSI B 31.1.0 or Equiva- lent (1967 Edition) Pumps and Valves Draft ASME Code for Pumps and Valves Class I Draft ASME Code for Pumps and Valves Class II Draft ASME Code for Pumps and Valves Class III Valves - ANSI B 31.1.0 or Equivalent Table 3.2-2 reflects minimum code requirements for Quality Groups used in original design. Replacement components may utilize alternate codes and edition/addenda as permitted by the PSL Unit 1 ASME Section XI program.
- Subsequent to the issuance of the ASME Code Section III all materials purchased for this service are qualified to ASME Section III. 1 ANSI B31.7 was the Construction Code, however for piping, ASME BPV Code Section III, 1971 edition through Summer 1973 Addenda is used for Class II and Class III piping. ANSI B31.7 is still used for Class 1 pipe. Reconciliation was performed in accordance with ASME Section XI. UNIT 1 3.2-11 Amendment No. 27 (04/15)
TABLE 3.2-2 MINIMUM CODE REQUIREMENTS FOR QUALITY GROUPS Component Quality Group A Quality Group B Quality Group C Quality Group D Pressure Vessels ASME Boiler and Pressure Vessel Code, Section III, Class A ASME Boiler and Pressure Vessel Code, Section III, Class C ASME Boiler and Pressure Vessel Code, Section VIII Division 1 ASME Boiler and Pressure Vessel Code, Section VIII, Division I or Equivalent Containment Vessel - ASME Boiler and Pressure Vessel Code, Section 111, Class B - 15 Psig Storage Tanks - API-620 API-620 API-620 or Equivalent Atmospheric Storage Tanks - Applicable Storage Tank Codes such as API-650, AWWAD100 or ANSI B 96.1 Applicable Storage Tank Codes such as API-650 AWWAD100 or ANSI B 96.1 API-650, AWWAD100 or ANSI B 96.1 or Equivalent Piping1 ANSI B 31.7, Class I (1969 Edition) ANSI B 31.7, Class II (1969 Edition) ANSI B 31.7, Class III (1969 Edition) ANSI B 31.1.0 or Equiva- lent (1967 Edition) Pumps and Valves Draft ASME Code for Pumps and Valves Class I Draft ASME Code for Pumps and Valves Class II Draft ASME Code for Pumps and Valves Class III Valves - ANSI B 31.1.0 or Equivalent Table 3.2-2 reflects minimum code requirements for Quality Groups used in original design. Replacement components may utilize alternate codes and edition/addenda as permitted by the PSL Unit 1 ASME Section XI program.
- Subsequent to the issuance of the ASME Code Section III all materials purchased for this service are qualified to ASME Section III. 1 ANSI B31.7 was the Construction Code, however for piping, ASME BPV Code Section III, 1971 edition through Summer 1973 Addenda is used for Class II and Class III piping. ANSI B31.7 is still used for Class 1 pipe. Reconciliation was performed in accordance with ASME Section XI. UNIT 1 3.2-11 Amendment No. 27 (04/15)
The diesel generator building is designed for 300 mph wind speed on the basis of its low height. Thefuel handling building is designed for the full 360 mph wind speed because this structure is too narrowand too low to meet the 300 mph wind speed criteria established for other Class I structures.The shield building has a radius of 154 feet while the overall plan dimensions of the reactor auxiliarybuilding are approximately 115 by 240 feet. Data extrapolated from Reference 2 indicate an average band width of approximately 50 to 80 feet over which the combined velocity distribution of 360 mph is postulated to act. On this basis, a uniform wind speed of 300 mph for large Class I structures wasadopted for design of the shield building and reactor auxiliary building. In addition to the effect of thedesign wind speed associated with this tornado, a 3 psi pressure differential in 3 seconds is applied simultaneously with the wind loading to the seismic Class I structures except for diesel generator building. Because of the ventilation openings in the diesel generator building, the structure is designed for a 2.25 psi pressure differential.The tornado wind speed is converted into equivalent static pressure loading and the computations forwind pressure, their distribution on surface area of buildings, shape factors and drag coefficient are based on the procedures outlined in Reference 1. Because of the unique characteristics of tornadoes, gust factor and velocity variation with height are not considered. With respect to the pressure distribution around the dome-cylinder shield structure wind force data reported in Reference 3 was used in the design. Equivalent static pressure loading for the various structures are given on Table3.3-1.The turbine building is the only structure not designed for tornado wind which can be considered inproximity to safety related equipment and structures. Under tornado loading, the first failures anticipated in the turbine building would occur in the vertical bracing system. The buckling of some ofthese members would force some beam-to-column connections to utilize their inherent moment-resisting capability and behave as moment connections. This in turn would result in the local overstress of some connections. As they begin to yield, the load would redistribute itself among other parts of the frame and the structure would behave in a plastic manner.The failures that would occur under these conditions are anticipated as being of a local nature. It isnot credible that the building will collapse because the turbine pedestal will also act to restrain thestructure. In summary, it is not anticipated that any structure or equipment necessary for safe plant shutdown would be affected by local failures in the turbine building due to tornado loadings.3.3-2 Tornado generated missiles considered in the plant design include a 10ft long 2 inch x 4 inch timbertraveling at 360 mph or a 4000 lb automobile traveling at 50 mph. The analysis to determine the effectof missiles is described in Section 3.5.3. Missile loadings are not applied simultaneously with wind loadings. The purpose of the tornado missile analysis is to determine that the structure can absorb sufficient energy to completely stop the missile without penetration. The capability of the facility to accommodate tornado generated missiles is discussed in Appendix 3F.3.3-3Amendment 18, (04/01)
TABLE 3.3-1TORNADO WIND SPEEDS ANDRESULTING STATIC PRESSURE LOADINGS ExternalLoadingTornadoExternalExternalwithWind SpeedGustPressureLoadingPressureStructure(mph)FactorCoefficient(Psf)Diff.Reactor Building 300 1 See Figures 3.8-35 to 3.8-38Reactor Aux. Building3001.9217 (1)220 (1).5115 (2)550 (2).5115 (3)270 (3)Fuel Handling Building3601.9300 (1)240 (1).5166 (2)706 (2).8266 (3)525 (3)Diesel Generator Building3001.9217 (1)110 (1).5115 (2)440 (2).5115 (3)230 (3)(1) Windward(2) Leeward (3) Roof (includes external wind pressure, internal pressure differential and slab dead weight)3.3-5 3.4 WATER LEVEL (FLOOD) DESIGN 3.4.1 FLOOD ELEVATIONS The plant area is situated above the highest possible water levels attainable except for wave runup resulting from probable maximum hurricane (PMH) considerations. During this condition, wave runups to 17.2 ft* mean low water (MLW) are possible. Plant grade is at elevation 18 ft and minimum entrance elevation to all safety related buildings is +19.5 ft. The maximum elevation of roadways on the plant site is +19.0 ft., thus any ponding of water that might result will be below the building entrances. 3.4.2 PHENOMENA CONSIDERED IN DESIGN LOAD CALCULATIONS All seismic class I structures are designed to withstand buoyant and static forces associated with high water levels. Only minimum structure and equipment deadweight are used in these calculations. 3.4.3 FLOOD FORCE APPLICATION In essence, structural components of all seismic Class I structures are subjected to a buoyant soil loading condition up to elevation +16.2 ft and a saturated soil loading condition from elevation + 16.2 ft to grade. This condition is the buoyant loading condition and accounts for conditions of maximum buoyance and flooding. 3.4.4 FLOOD PROTECTION Structures and components whose failure could prevent safe shutdown of the plant or result in significant uncontrolled release of radioactivity are protected from the effects of high water levels and wave runup associated with PMH conditions by one or more of the following: a) Design of structures and components to withstand such effects where functionally required. b) Positioning of the structures and components such that they are located at sufficient grade to preclude inoperability due to external flooding. c) Housing within waterproof structures: The shield building and reactor auxiliary building are the only seismic Class I structures with basements. These structures are completely waterproofed to finish grade with Nob-Lok waterproofing. All construction joints are waterstooped with 6 in. polyvinyl chloride. Table 3.2-1 lists the flood protection criteria applied to plant structures, systems and components. The a, b, or c designation in the table refers to items a, b, or c above.
The list below designates each seismic Class I structure and the associated means of protection from flooding.
- Reference Section 2.4.5.9 for updated surge levels and wave runup analysis.
UNIT 1 3.4-1 Amendment No. 27 (04/15) Structure Flood Protection Shield Building No openings below elevation +22 ft Reactor Auxiliary Building Ground level openings at 19.5 ft Fuel Handling Building Ground level openings at 19.5 ft Diesel Generator Building Floor and equipment above elevation +22 ft Intake Structure Motors located above elevation +22 ft
UNIT 1 3.4-2 Amendment No. 27 (04/15) All buildings with the exception of the turbine building are of the enclosed building type. The turbinebuilding will be subjected to wind driven water spray, consequently, all equipment inside this buildingis designed for outdoor service.The flood protection for the emergency diesel generator system includes protection for the oil storagetanks which rest at elevation 22 ft. The outlet nozzle for pump suction is located at elevation 22 ft 6 in., with filling connections at approximately 37 ft and vent connections at approximately 38 ft.All permanent door openings in the exterior walls of the reactor auxiliary, fuel handling and dieselgenerator buildings are provided with either roll-up or swing type doors for protection from rain, wind and other atmospheric effects. The rolling shutter doors are fabricated of interlocking slats, curved and jointed to shed water. Large doors are furnished with a continuous, adjustable rubber stripping at jambs, head and floor to provide a positive weather-tight closure. Access doors may not be provided with weather-stripping in all cases, however, the amount of leakage-induced flooding through thesedoors is not more adverse than that considered in the analysis presented in Section 3.1.3 of Chapter9.5A on the rupture of non-seismic Class I equipment (fire system piping).3.4-3Am. 9-7/90
The site drainage system is designed to preclude flooding of safety related structures and components under PMH conditions however total flooding of the drain lines will not cause water to backup into areas which would jeopardize the required function of a safety related system. PMH conditions would produce an overland flow which would exceed the capacity of the drain lines, however, excess waters would run off the plant island. Section 2.4.2.3 addresses drainage of water from the southern site property and the effect of water pooling caused by the intake canal berm. The Unit 1 and Unit 2 site drainage plans are shown in Figures 3.4-3 and 3.4-4, respectively. In areas where Unit 1 drain lines are to carry storm water from both units, the lines are oversized to accommodate the additional flow. The interfaces between Unit 1 and Unit 2 drainage lines are shown in Figure 3.4-3. Drain lines are sized to accommodate runoff in the plant area. Runoff in the plant area is estimated by relating the tributary area and the rainfall intensity to an estimated proportion of the rainfall reaching the catch basin as direct runoff. This procedure is represented by the following formula: Q = ACIP where Q = design discharge, cfs A = tributary drainage area, sq ft C = runoff coefficient based on surface conditions I = intensity of rainfall, in/hr P = coefficient based on percent of full pipe flow The design considered values of C consistent with use in the Rational Method for various ground surface types. The intensity of rainfall, I, used in the calculations was 6 inches per hour. The tributary drainage area was determined by the location of surrounding catch basins and storm drain lines. The ISFSI drainage plan is shown in Figure 3.4-5. Catch basins are constructed to provide ready access to storm drains for inspection and maintenance as well as to serve as points of concentration for runoff. Runoff computations for catch basins include roof, floor and equipment drains having no potential for contamination.
The analyses of a postulated failure of a pressurized fire main within the reactor auxiliary building is provided in Subsection 3.1.3 of Chapter 9.5A. This is the only plant structure that houses both pressurized fire system piping and safety related equipment. The consequences of this postulated failure are acceptable.
3.4-6 Amendment No. 26A (05/15) FIGURE 3.4-1HAS BEEN INTENTIONALLY DELETEDAm. 7-7/88 FLORIDA POWER & LIGHT COMPANYST. LUCIE PLANT UNIT 1REACTOR AUXILIARY BUILDING EXT. WALLS-MISC DETAILS-M&RFIGURE 3.4-2Amendment No. 15 (1/97) FLORIDA POWER & LIGHT COMPANYST. LUCIE PLANT UNIT 1GRADING & DRAINAGE - SH.1FIGURE 3.4-3Amendment No. 15 (1/97) FLORIDA POWER & LIGHT COMPANYST. LUCIE PLANT UNIT 1SITE GRADING & DRAINAGE-SH.1FIGURE 3.4-4Amendment No. 15 (1/97)
FLORIDA POWER & LIGHT COMPANYST. LUCIE PLANT UNIT 1REACTOR BUILDING HATCH COVERS & MISCDET'S - M&R, SH1Amendment No. 15 (1/97)
b)high pressure safety injection system (piping which is part of reactor coolant pressure boundaryonly)c)all lines used for shutdown cooling which includes portions of the low pressure safety injectionsystemd)chemical and volume control system (letdown and charging lines)e)main steam system f)main feedwater system g)steam generator blowdown system h)auxiliary steam system The analysis of high energy line breaks outside the containment is presented in Appendix 3C for items e)and f) above. Appendix 3D includes analyses for items c), d), g) and h).3.6.2DESIGN BASIS PIPING BREAK CRITERIA In analyzing the effects of LOCA pipe rupture, both circumferential (guillotine) and longitudinal (slot)breaks are considered capable of occurring at any location along the piping.(1) Guillotine breaks and slotbreaks with an area up to the cross-sectional pipe flow area are assumed. The effects of resulting pipe whip are considered in the design. The effects for jet impingement resulting from slot breaks are also considered. Piping 1 inch and under is not considered to rupture.AEC Regulatory Guide 1.46 required that rupture locations for piping systems inside the containment bechosen based on stress limit and usage factor criteria, with a minimum of four such locations analyzedinclusive of terminal ends. The guide further required that protection for pipe rupture (pipe restraints,separation) be provided for piping systems in which operating temperatures exceed 200 F or operatingpressures exceed 275 psig. (Note that for any future changes to the plant design, the Standard ReviewPlan section 3.6 has superseded Reg. Guide 1.46, which has been withdrawn.) (1) See Section 6.2.1.3.3.a.3.6-1aAmendment No. 18, (04/01)
Generic Letter 87-11 revised Branch Technical Position MEB 3-1, "Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment", as contained in the Standard Review Plan (SRP), Section 3.6.2, "Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping," Reference 28. All modifications to Class 2 and 3 piping systems may invoke this new criteria (see Appendix 3.J) in lieu of the original criteria provided that the requirements stipulated in Appendix 3J are fully complied with. The new criteria is based on portions of the revised Branch Technical Position MEB 3-1 applicable to St. Lucie Unit 1. As stated in NRC Generic Letter 87-011, Reference 27, "Licensees of operating plants desiring to eliminate previously required effects from arbitrary intermediate pipe ruptures may do so without prior NRC approval unless such changes conflict with the license or technical specifications. ...the licensees' updated FSARs should reflect eliminated hardware associated with arbitrary intermediate pipe ruptures." Arbitrary intermediate pipe ruptures, which previously were specified are no longer mentioned or defined. However, requirements for postulated terminal end pipe ruptures, postulated intermediate pipe ruptures at locations of high stress and high usage factor and for leakage cracks are retained. 3.6.3 DESIGN LOADING CONDITIONS The design loading combinations, design condition categories and design stress limits for safety related piping systems are given in Table 3.9.3. All safety related equipment is designated in accordance with the rules stipulated in Sections 5.2.1.4 and 3.9.2.
The following loads are considered in the design of the pipe whip restraints inside the reactor containment and reactor auxiliary buildings: D = Dead load of the pipe whip restraint
PR = Steam/water jet forces and/or pipe whip reactions resulting from a ruptured pipe. A dynamic load factor of 2.0 is used in the design of all pipe whip restraints to account for the dynamic nature of the load.
3.6-2a Amendment 15, (1/97)
the internals are capable of accommodating loads resulting from the largest cold and hot leg breaks. Theanalyses of the fuel indicates that for an unrestrained inlet break localized crushing of the grid spacers occurs near the top of some of the peripheral assemblies. However, this limited crushing does not prevent maintaining the coolable geometry of the assembly and the core overall. See Reference 22.In Reference 23, the NRC staff concluded that there is reasonable evidence that the Unit 1 reactor wouldwithstand the effects of asymmetric LOCA loads.3.6-4 Amendment No. 13, (7/94) 3.6.4 DYNAMIC ANALYSIS3.6.4.1 Pipe WhipThe reaction force on the ruptured pipe and the force resulting from jet impingement are calculated asfollows: Reaction Force = kPo A (lb) Jet Impingement Force = kU m Po A(1)where k is the thrust coefficient which is used to compute the impingement force, and U m is the ratio ofthe peak axial velocity at some distance "x" from the origin of the jet to the initial jet velocity. The peak velocity at any distance from the origin of the jet is obtained from equations and charts of Reference 1.PoA is the product of the initial pressure inside the broken piping and the break flow area.The thrust coefficients used for guillotine and slot breaks are as follows:THRUST COEFFICIENT, kTwo Phase, Flashing Sub-CooledType of BreakDry Saturated Steam Water or Wet Steam WaterGuillotine1.011.121.6Slot0.630.680.97The design for the protection against dynamic effects of pipe rupture is based on the premise that breakscould occur anywhere in the pipe run, and not on the criteria formulated on the hypothesis that satisfactory protection from whipping pipes can be achieved by restraining against breaks postulated at stress related locations. Consequently many more restraints are utilized than would have been provided had the stress related criteria been used. The spacing of such restraints is based on the ultimate moment carrying capability of the restrained pipe such that for a break between consecutive restraints additional breaks would not result at either restraint or any other one. As a result, the protection provided against pipe rupture is at least the equivalent of that which would be achieved according to the stress related pipe break criteria and the pipe whip analysis methodology of Reference 1. Even assuming that an individual restraint could not bear the entire pipe rupture load, the adjacent restraints would share the load and still result in an adequately restrained system.To verify whether the restraints are adequately designed, since the design is based on a static forceapplied with a dynamic load factor of 2.0, it is necessary to examine the dynamic characteristics of the force as well as the structural response of the restraint.The dynamic behavior of the reaction force on any given pipe and any given break depends on severalfactors among which the more salient are:3.6-5Amendment No. 18, (04/01) a)characteristics of contained fluid (steam, compressed water, pressure, temperature)b)capacity of reservoir behind break c)configuration of piping system d)location of the break with respect to the reservoir (friction effects) e)presence of flow restrictors in the line. The initial peak value of the reaction force is only slightly affected by friction effects, phase change effectsand flow restrictors. However, the subsequent transient phase is very much affected by the above factors and also the actual configuration of the system. Thus, for each given break, it is far easier to predict fairly accurately the first peak value of the reaction force than it is to predict the subsequent transient phase. From experiments (Reference 2) and also published literature (Reference 3), the following can be stated with regard to this transient regime:a)For subcooled decompression (such as would occur in feedwater line, shutdown lines, raincoolant lines), the initial force peak decays rapidly as the pressure drops to the saturationpressure for the given temperature, then stabilizes at a slowly decaying value, as shown in Figures 3.6-5 and 3.6-6. The duration of this sharp transient is dependent on the location of the break with respect to the reservoir, the presence of any flow obstacles between the reservoir and the break, and the friction in the line. In general, the duration will be less than 1-2 milliseconds.b)For steam breaks, again the peak force value lasts only a few milliseconds. Prior or subsequentvalues of the force are lower, although the effect is not as pronounced as in subcooleddecompression.Since, due to the gap between piping and restraint, the broken pipes also require times of tho order of milliseconds to strike the restraints, the peak force will not be acting on the break constantly, but rather the total impulse received will be less than that which would he calculated by using the peak force as constant. Furthermore, the energy imparted will be even less, e.g., half the impulse corresponds to one quarter of the energy. Thus a constant force of lower magnitude could be justified in a static analysis.The dynamic load factor (DLF) is the product of the force actually acting on the restraint times a dynamicamplification factor accounting for the response of the restraint, the maximum value of which would be 2.0 if the restraint behaved entirely elastically. In reality the restraint will respond elasto-plastically and the magnitude of the dynamic amplification factor (DAF) will be less than 2.0. The force3.6-6Amendment No. 18, (04/01)
Abramovitch has shown that the velocity profile at any point "x" along the jet axis is given by:(2)where d(x) is the jet diameter at point "x".and y(k) is the distance from the axis to the jet at point "x", atwhich U(x) is measured.Equation (1) calculates the jet force on a structure as if the average velocity at "x" corresponded to thepeak velocity at "x", whereas, in reality the average velocity is only 0.26 Um(x), and the correspondingaverage force is 0.52. The average force on a wall intercepting the entire jet is thus overestimated by using equation (1).If the struck surface does not intercept the entire jet, the impingement force is further reduced to accountfor the projected area of the target. If the struck surface is very small with respect to the jet and such that the jet completely surrounds it, the force is equal to the drag force on the target and is calculated by multiplying equation (1) by the proper drag coefficient.For pipes, the jet impingement force on the analyzed components are calculated by multiplying the jetimpingement pressure by the projected area of that portion of pipe which is contained in the area of the jet and multiplying the result by a factor of 0.6 to account for the curvature of the component.In evaluating the load carrying capability of pipe the internal pressure and the effects of strain hardeningare considered. In the case of a simply supported pipe experiencing a slot failure, the cross-sectional area of the pipe capable of sustaining a moment is considered to be the defect area which accounts for the presence of the slot break instead of the original pipe area.3.6-8 3.6.4.3 Pipe Whip Analysis - Main Steam and FeedwaterAs shown on Figure 3.6-52, a break location was established at node 12 for the main steam line. A breakat node 12 results in the maximum impact at the restraint located at node 9 and the maximum total strain in the pipe.Span lengths between pipe whip restraints are as shown on Figures 3.6-52 through 3.6-55,for main steamand feedwater piping. The maximum span lengths depicted were established using the design criteria presented in Section 3.6.5-1. As stated in 3.6.5.1, failure stress is limited to that value which corresponds to 50 percent of the true ultimate strain when related to a simplified stress-strain curve (Figure 3.6-9A).For the steam line break selected (node 12 on Figure 3.6-52 ), the moment required for full plasticity, foryielding and the actual moment computed for that limiting case are 35.2 x 103 in-kip, 27.8 x 103 in-kip and35.2 x 103 in-kip, respectively. The actual computed moment and the moment to full plasticity are equalsince for this limiting span length the pipe does plastically deform, but does not whip.Sensitivity studies for the main feedwater line outside containment are summarized for variations in gaplength and pipe wall thickness in Table 3.6-1 and Figures 3.6-63 through 3.6-68. A reduction in gapreduces peak restraint reactions while decreasing wall thickness seems to increase reactions.3.6-9Amendment 15, (1/97) The span method of restraint placement (See Section 3.6.5.1) does not provide for a margin to fullplasticity since the method itself assumes the pipe to go plastic. The span method does, however, prevent maximum calculated strain from exceeding one-half of the ultimate strain. For example, at the instance when a pipe is fully plastic, the pipe can still carry moments - only additional strain energy into the pipe will cause further straining up to the ultimate.If the strain hardening is ignored, as was done in this analysis, then the span method does not predict thestrain corresponding to imposed moment since this strain is not unique. However, Figure 3.6-61 shows that the maximum strain does not even approach half ultimate strain values. Since zero strain hardening was employed, it follows that the calculated moment at maximum strain and the moment required for full plasticity are identical, 35.2 in-kip.At this moment value, the pipe has not collapsed and can continually carry equal or diminished loads. Themoment necessary to yield the outer fibers of the pipe is 27.8 in-kip.Figures 3.6-52 through 55 provided the span lengths, restraint locations and node locations using nodalbreakdown requirements for pipe whip analysis. Nodal breakdown requirements for pipewhip analysis are different than those required for stress analysis as reflected in the piping isometrics of Section 3.6.Because node points were included in the numbering scheme in Figures 3.6-52 through 55 and sincenodes were not shown in other Section 3.6 figures, no correspondence should be expected between the two except that piping dimensions and location of pipe whip restraints are identical.To illustrate this point refer to Figures 3.6-36 and 3.6-52. Figure 3.6-36 restraint locations MS-2, MS-3and MS-5 correspond to restraint locations 2, 4 and 8 in Figure 3.6-52, respectively. All the relatedfigures have been reviewed for accuracy. a) ConclusionsFour cases (2 feedwater line breaks and 2 main steam line breaks) of circumferential pipe rupture wereanalyzed for maximum restraint reactions and maximum pipe strain. All analyses were extended for a period of 0.2 seconds past initiation of pipe rupture when steady state oscillations of the deflection of the rupture point occurred with decreasing amplitude. Strain hardening in the pipe was assumed to be zero for conservatism.Figure 3.6-56 indicates that blowdown forces for the feedwater line rupture reach a steady state value of110,000 lbs at 0.034 seconds; for the main steam line, the blowdown forces reach a value of 135,000 lbs at 0.1 seconds, decrease exponentially to 100,000 lbs at 0.15 seconds, and continue to decrease at the same exponential rate thereafter.3.6-10Amendment No. 16, (1/98) Since the steam line is multi-planar, restraint reactions can occur in more than one direction and in morethan one restraint. This is evident from the reaction force results plotted on Figures 3.6-57 and 3.6-58. Peak reactions in all cases except two were below the 2 KPA factors applicable to the line under analysis. In the two exceptions (feedwater line break at node 6, Figure 3.6-59 and main steam line break at node 12, Figure 3.6-58), the peak duration is approximately 0.002 seconds or less.Since the natural periods of the restraint system, consisting of the steel frame restraints, the embedmentsand the concrete wall are of 0.002 seconds or less, this system was reviewed to determine to what extent, if any, its Primary function of pipe restraint during blowdown may be impaired.A very conservative analysis of the steel frames, in which the stiffening effects of collar plates and webswere ignores and the pipe whip impulse loads were treated as step functions constant in time, revealed that yield would occur in the structure. Since the pipe whip dynamic analysis was based on the assumption of elastic, non yielding restraints, the effect of yielding would be to reduce the loading peaks shown in Figures 3.6-58 and 3.6-59. However, the nonyielding assumption used in the pipe whip analysis is taken, as the more conservative approach. In either case, yielding or nonyielding, the steel frames will perform their function of adequately restraining the pipes against excessive movement.Assuming complete rigidity of the steel frame restraints and concrete, the bolts, subjected to a pulse (hatfunction) loading of 0.002 seconds duration were shown to reach a peak strain of 0.0155 in/in (based on a bilinear stress-strain curve in which Young's modulus E=30x106 psi and the strain hardening modulus S-0.05E) for carbon steel. This strain is well below 1/2 for carbon steel (taken as 0.1) and is confined tothe threaded portion of the bolt. Once more it is seen that under very conservative assumptions the bolts do not rupture, and that yielding results in lowering the applied pulse peaks due to pipe whip.Assuming, once more, that the restraint frames remain rigid, and that a step function load is appliedthrough the bolt anchor plates to the concrete with a peak equal to the applied pipe whip impact-pulse distributed to the embedded bolts, it was shown that the concrete would not fail in shear (i.e., pullout) and that the concrete wall was adequate to resist these loads.In summary the design of the pipe whip restraints and embedments is considered adequate to performtheir primary function of limiting pipe motion and secondary damage following a pipe break because there will be no intolerable loadings as a result of exceeding the factor 2.0 for the k load factor.Maximum strains in the feedwater and main steam lines are found by adding the yield strain to themaximum plastic strain. These peak strains, in all cases, are considerably less than half the ultimate strain of the materials (main steam - steel, A 155 GR-KC 65; feedwater - steel, A 106 GRB).3.6-11
where:f= Fanning friction factorAw= Wetted wall areaA= Cross sectional area of flowb)For junctions having flow, the code calculates the friction factor internally from where:Wj and Pj are the mass flow rate and density of the fluid at junction j, and !Ppump,j is the pressure head due to a pump located at junction j; Pi and Pi+" are the total pressures involumes i and i+" across junction j, including gravity head.3.6.5 PROTECTIVE MEASURESPiping within the plant is arranged or restrained such that in the event of a LOCA, the dynamic effectsassociated with the pipe rupture will not result in loss of containment integrity or prevent engineered safety features from mitigating the effects of the LOCA.The containment vessel is protected from the effects of LOCA pipe rupture by the secondary shieldwall. The secondary shield wall encloses all piping whose failure could cause a LOCA. Pipe whip restraints are provided on all such lines which are connected to containment penetrations to limit the pipe rupture loads on the penetrations.3.6-23Amendment No. "9 ("0/02) Protection of engineered safety features against the effects of LOCA pipe ruptures is provided by anyor a combination of the following:a)Spatial separation of redundant pipings and components such that pipe whip or jetimpingement resulting from the LOCA cannot damage both redundant components.b)Placing of pipe whip restraints such that resulting pipe movement cannot damage adjacentpiping or components.c)Placing of barriers between redundant components. Where required, protection against damage to adjacent piping or equipment from pipe whip isprovided by limiting the amount of movement of the ruptured pipe. This is achieved by placing pipe whip restraints such that the reaction forces on the broken pipe do not cause formation of a plastic hinge.Restraint spacings have been developed for rupture in all typical piping configurations as shown onFigure 3.6-". Reaction forces ire calculated using the relationships in Section 3.6.4.The design loads for pipe whip restraints are determined by the location of the pipe rupture in relation to the physical location of the pipe restraint. To assure conservatism, pipe ruptures are postulated in a location on the system which would create the largest forces and moments on the restraints. The magnitude of the load (static) corresponds to the pipe reaction or jet impingement forces.Since this load is applied suddenly, the restraints are designed for impact loading. This isaccomplished by increasing the static nature of the load by a dynamic load factor and the restraints and restraint structures are analyzed in terns of an equivalent static load.The magnitude of the dynamic load factor varies depending, on the rapidity and inelasticity of the loadapplication and the dampening characteristics of the system. It is known, however, that the maximum value, based on a single degree of freedom system is 2.0. To assure conservatism, a design load factor of 2.0 has been used for design of pipe whip restraints.Where required, protection against damage to piping or equipment from jet impingement from anadjacent ruptured pipe is provided by ensuring that the spacing between piping or equipment is such that the jet impingement loads do not result in failure of the piping or equipment being impinged upon. the jet impingement forces are calculated using the relationships given in Section 3.6.4. The impingement force is attenuated with distance from the break according to the variation of jet velocity and divergence with distance.There is no critical equipment exposed to a reactor coolant pipe break blowdown. Critical pipingunderwent analysis to assure that there would be no failures as a result of the LOCA jet impingement. The critical piping system pipe break forces were compared to those that could re-3.6-24 sult from the jet impingement (this includes the cantilever effect on the nozzles where the criticalpiping penetrates the vessel) to determine which was larger. Additional restraints were added where the jet impingement forces were larger.3.6.5."Spacing of Pipe Restraintsa)Design approachFor the pipe rupture loading condition, the design philosophy is that the ruptured pipe itself must berestrained in such a way that it does not develop pipe whipping and consequently impair a nearby pipe, critical structure or pieces of equipment.To implement the above, the load carrying capacities of the pipe under the postulated "maximumcredible load evaluated in terms of its ultimate capabilities to resist torsion and/or bending;i.e. the pipe is allowed to experience permanent deformation without loss of function of the system. once the ultimate load carrying capabilities of the pipe are established, the spacing of the pipe restraints can then be designed to prevent pipe whipping.b) Ultimate Load Capability In the evaluation of the ultimate load capability, it is assumed that a structure fails when the appliedloads produce maximum primary stresses equal to those corresponding to 50 percent of the total strain on the trapezoidal stress-strain curve as shown in Figure 3.6-9A. This consideration accounts for the difference between actual and theoretical structures due to the presence of welding or connections. The basis for the above assumption is the work performed at the United States Naval Ordnance Laboratory. The Naval Lab work shows that models of vessels with welding and connections burst when the maximum strain was of the order of 50 percent of the total strain.To simplify the analysis, the stress-strain diagram is approximated by a trapezoidal curve, as shown inFigure 3.6-9A. The design failure stress, Su*. material is obtained from the resulting diagrams at thestrain #u*. = 0.50 #u.This linearized approach introduces further conservatism because the failure stress thus selected corresponds to a strain less than 50 percent of the total strain on the actual stress-strain diagrams. For ferritic material with a pronounced yield point this difference is small, but significant. For stainless steel, the difference is greater. The slope of the actual stress-strain diagram is much steeper than the linear model used in the calculations of the failure stresses in the region of the yield point indicating that most of the strain hardening resulting in increased strength occurs as soon as the structure starts to yield.Structures having hollow-circuit cross sections are considered approximating nozzles or pipes carryingsteam or water under pressure, when subjected to mechanical loads across their total cross section. The method used by Stokey, Peterson and Wunder (7) to evaluate limit3.6-25 loads for tubes under internal pressure, bending moment, axial .force and torsion rigid-plastic materialwithout strain-hardening has been adopted and amplified to account for the effect of strain hardening. This effect has been evaluated following the method outlined in References (8), (9), and ("0). TheTresca, or maximum shear stress, theory has been applied in this development.c)Analytical Methods The principal stresses in a Mohr's circle are = + (") = circumferential stress = - (2) = axial tensile stress = shear stress(3) = Internal pressureIn the case of the stress intensity, i.e., is given by: - S = + + (4)In the case of - S(5)Using the maximum shear stress criterion, the limit condition is:(6)that is, when + (7)3.6-26 and, when 2= (8)Equations (7) and (8) yield the lower limits on the allowable axial tensile stress. Rewriting the aboveequations: (9) ("0)where: ("")!!("2)For the case of no shear stress, Equations (9) and ("0) give:("3)("4)For the combinations of stresses present in the components of piping systems, is usually greaterthan therefore the equations that normally apply are (4), (7), (9) and ("3).For the compressive stress region, the previously mentioned cases, i.e., and alsooccur. In the components of piping systems is algebraically smaller than .3.6-27
It has been shown that ultimate load capabilities for pipe can very well be based upon assuming thinwalled tubing. Introducing cylindrical coordinates and assuming a thin-walled tube, the following expressions can be obtained:(20)(21)For the example illustrated in Figure 3.6-9B, the axial force, , and the bending moment, , are thengiven by:
(22)
(23)By proportioning of the stress distribution diagram in Figure 3.6-9B and can be expressed as(24)(25)The bending moment increases as the angle, , approaches zero.By inspection of equations (9) and (16), it is found that all.is always greater than Since is less than conceivablycan equal The axial force, N, is calculated from equations (22), (24) and (25) and it is consequently assumed tobe zero. The effect of axial force on the maximum moment capability is negligible for reasonable pressures and standard pipe sizes (7). Also a normally restrained pipe run does not exhibit any axial loads under the considered loading conditions.3.6-29Amendment 19 (10/02)
generator is the only leg expected to form a plastic hinge. This is confirmed by the analysis describedabove. The force will cause the pipe to be driven to the floor.There is no critical equipment exposed to a LOCA blowdown. critical piping underwent analysis toassure that there would be no failures as a result of the LOCA jet impingement. The critical piping system pipe break forces were compared to those that could result from the jet impingement (this includes the cantilever effect on the nozzles where the critical piping penetrates the vessel) to determine which was larger. Additional restraints were added where the jet impingement forces were larger. 3.6-32a Amendment No. "2, ("2/93)
TABLE 3.6-1SENSITIVITY STUDY FOR 20" FW LINE OUTSIDE CONTAINMENTMax. PlasticMax. Defl.Node No. GapWall Thick.Max. Reaction Strain in At LoadedAt Restraint(in) (in) (lbs)PipeNode521.8721 x 106.00656.381"521.5.68 x 106.00725.446"541.5.7806 x 106.001489.255"3.6-35Amendment 15, (1/97) TABLE 3.6-2SUMMARY OF PIPE WHIP ANALYSISLine Description Guillotine Break Active Restraint Maximum Deflection Maximum Reaction Maximum Plasticat Node NumberNode Numberat breakat RestraintStrain in Pipe(pounds x 106)Feedwater Outsidecontainment 1.5" 7 5 9.255" 0.7806 0.00148wall and 4.00" gapFeedwater InsideContainment 7 6 3.502" 1.34 0.00259Mainsteam Inside 2.72* (+X)Containment 12 9 11.46" 2.34 (+Z) 0.00762Mainsteam Inside 9 1.0 (+Z)Containment 16 13 13.36" 1.78 (+X) 0.0000 1.52 (+X,-Z)*Values exceeding 2KPA have a duration of <2 milliseconds 3.6-36 TABLE 3.6-3MAIN STEAM LINE INSIDE THE CONTAINMENTGUILLOTINE BREAK AT NODE #12RestraintNodeI.D. No,El. Spring Const.Pl. Spring Const.Gap (in.)2 MS-12 8.536 x 106#/in 3.570 x 105 #/in 6.004 MS-13 8.786 x 106#/in 3.650 x 105#/in 6.00 8 MS-15 8.786 x 106#/in 3.650 x 105#/in 5.509 MS-16 8.786 x 106#/in 3.650 x 105#/in 5.5011MS-17 8.786 x 106#/in 3.650 x 105#/in 4.00TABLE 3.6-4MAIN STEAM LINE INSIDE THE CONTAINMENTGUILLOTINE BREAK AT NODE #16Restraint InformationRestraintNodeI.D. No.El. Spring Const.Pl. Spring Const.Gap (in.)2MS-128.536 x 106 #/in3.570 x 106 #/in6.004MS-138.786 x 106 #/in3.650 x 106 #/in6.008MS-158.786 x 106 #/in3.650 x 106 #/in5.509MS-168.786 x 106 #/in3.650 x 106 #/in5.5011MS-178.786 x 106 #/in3.650 x 106 #/in4.0013MS-188.786 x 106 #/in3.650 x 106 #/in4.003.6-37Amendment No. 16, (1/98) TABLE 3.6-5BOILER FEEDWATER LINE OUTSIDE THE CONTAINMENTGUILLOTINE BREAK AT NODE #7Restraint InformationNodeEl. Spring Const.Pl. Spring Const.Gap (in.)28.536 x 106 #/in3.570 x 106 #/in4.0048.536 x 106 #/in3.570 x 106 #/in4.0058.536 x 106 #/in3.570 x 106 #/in4.00TABLE 3.6-6BOILER FEEDWATER LINE INSIDE THE CONTAINMENTGUILLOTINE BREAK AT NODE #7Restraint InformationNodeEl. Spring Const.Pl. Spring Const.Gap (in.)48.786 x 106 #/in3.650 x 106 #/in2.5068.786 x 106 #/in3.650 x 106 #/in4.003.6-38
- CASE I CMTILEYER llEMDl<G GIJILLOTINE llREl.k CASE Z tF tlAX. ALLO.A8l.E SPAN I IS LESS THAN L PLACE T0RS&0M RESTRAINTS HE!tE AS AEO"O. liUIU.OTllE BRE;.IC " r ?_J_)* CASE l llEAll llEJl)IHG SLOT BRHK CASE 4 CANTiLfVER SLOT BREU F CASE S TORSION SLOT BREAk
- CASE 6 CURVFD BEAM SLOT BREAK For !-8 L -;_90'tof MAX SPCCIFIED BEAM CHECK FOR (2) CONDITIONS: 1. l ! .90 x SPECIFIED BEAU 2. L1 $ SPECIFIED CANTILEVER CASE 7 BEAM BENDlllG SLOT BP.EAK PREFERRED ARRANGEMENT CASE8 _ 815 AM BENDING Ef==-_ _ __ -:--:::1 SLOT BREAK * ** j) USE CASE 8 OHL Y IF CASE 7 CAN.iOT BE ACCOMMODATED --J. BE THE SHOfHEA OF THE 1WO SPf< l"EQ CANT llfVER ir CANTILEVER SPA.NS SPEC!flEO ARE fl.tr.SE:DDN GllllLOl!Nl rHfN l SLOl t (01!ll l Ill 1N£ II FORCE f ORCF
- F aEA:;::0 .. c DL --I SLOT AND GUil BREAK PREFERRED ARRANGEMENT , LJ L,,. LEFT HAND SIDE lOENTICAL TO , CAS:: 1 R H StOE BECOMES A -BEAM WITH GUILLOTINE FORCE OF BRANCH 11\._ THEN F l BEAM GU1L. :: l BEAM SLOT 1. FORCE SLOT 6EAM Fa r-U111,,,t: uUIL. BAA.NCH BUT NOT Ge\EATER THAN L CASE 10 BEAM 8ENOING SLOT A.Ht; CUil Bf!FAKS USJ:: CASE 10 ONLY IF CASE 9 CANNOT BE ACCOMOD.t Tf) FOR SLOT BREA.If APPLY PROCEDUAE OF CASE I* FOR GUILLOTINE BREAK IN BRANCH LL AN:' Ls TO MEET THE RESPECTWE LEVEf REQUIREMENTS tllOOIFIEO FOR THE BRANCH GUILLOTINE FORCE. THEN L = l CANTILEVFA 1. FORCE GUil. L '-L FORCE GUil., BRANCH Ls Li CANTILEVER 1t CASE II Y-LOOP CANTtLEYEF! BENDING GUILLOTINE llllEAk *1 .----1 I IF MAX. ALU)¥ABLE SPAN_ 15 LESS THAN l PLACE RESTF:AINT HERE /'-CASE 11 CANTILEVER BENDING BRE4K \ \t-:+---* !:.. ___________________ ,. FLORIDA POWER & LIGHT COMPANY St. Lucie Plant
- RESTRAINT SPACING FOR PIPE RUPlURE -SHEET l FIGURE 3.6-1
- Refer to drawing 8770-G-799 Sheet 13 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING PIPE RESTRAINTS SH.13 FIGURE 3.6-2 Amendment No. 15 (1/97) m I I 11. i __ -' !"-!"-111 {' -u : ---.. ifll,U I "" -I l'!i r,.;*E 10*1 r* P; -;:_E\A.1,)1...... A.T*A.T W.t..U .. (>r><,,,..,ee;...,* r., o W ... LL SE.Ci *,*.**o -?E CT SE Dtt><o>) ""' ... H r* r.f ---1 ., ... l,,.. . .!:i,..::::J
- I *I ti. v.o-, i C11h) RA8 R'1 : . <:Oo.JC..-.'T<; a.; Pl..t.tt RAD I l'>**t .... ,,. \ '-,{' ;,; ,; : p .,. ... '! "') QA.I*', I I WES\ WA.ll cco1t*1<<t. NOP1H WALL *ht*,.,.., ... ,,. 'loti.,,1 .. 1 "" 1YP OE:-1 .._ 1 C;t011t* .... .. "'1 1 * ,,.,,., ..... t* <.l..'11<* <#ti.la.I .. \
- 04\11..it.1 Cl.I\ t M* *** -.,.., ....... -io euMH<\1 ' t.LI.. ..... , A 1*"lll' ......... 111;1 ""' ""' I W)..ll Htv WH1) *"=' l Wt* . oe:-1,A.,1l'S ' * ..,. .,,O*l\i '!><..*) ol /' '""'\ " -_-< :: ! ;:y i "' 1 _._ 'YP OE.l 'Y' ltl*v,.,,01.:11 Stc.1 A...,.,,.., i'l'P DE.r 'C I 12. WA.LL ii! ::t L * ** ** 1L 1'h<*N,.,O ... \-*lft** >'OO 'I: ... .....,.. .. "*>wt _,Ul\10Jo 1**t ""'""* HO I°""'°' -'"'"'T PLA..t-..l i'<P DEl O<Jl'I.*!)* '. ""°" * ' t-... t.t __, "'---o*t.<ll 3-' t.\9 1 i Pl,6.,,kJ ( o .... DE:"'ti.1..,. -::;to 01PE llG t l j lJ"-Nl !11(:;'$ 1llH 'l1 J*hl'.I *#MH _,IOI \lHl \H -* &hJCMI .. 'loe ... 'NU ::,',": il;jl1J1*11'!'1. :'Jl:l t110 .... _"_"_' __ *ov1*llfi , .... u "' o*n*11111l!U* ., tih* ...... -.. ," .. ""'!) 'I* '1<>.,.-"H"' r!:\0Jf.f11 .._)A'!.,.. .... l:t,,.,;CAlft, .,,_,._ *' 1 ........ , .......... ,. ' -... _ ..... .... . ::__ i_ I , ..... .......... . --* --r -----** I : / -:t/ ____ I ' . --,1 ' I f " _._, ,...,,,._..,,..., .. t \<At ,,.., ... , 'Y<Ov*O lftl < .. .,.,.,.,, >J(<>*-'" <.>'l\l * .,._ ,.,. ..... _., ( ... --,... .. ... ., ......,. ..... """'° """'"' .,,., --* ....... ............. ... fl .. *. DElt.tlS Al QE:Jl':.E-0 1 ... ***t\ , .... H *1 **t.ltl'.l **l .. t.nrr,it. o;':" .. "'t:!:. t. .. ..... , ...... ... , .. .. , "" .... """'" *t.,,V'Osff 'SlUt. .. -...,,,,.. ....,..,, __ ,,,, .. ,,,... .. (....,, """ ** .... l* f', .. N*aM ..... .. ........... t,..,., .. t llChCAU* 1)1tc.1P1.1ft ..<t; '"' **"-.. * ,.,&u:4'.1'( .. * . ._ ... ", ... -VffEUhJCE CR'"-V../lJC1"'1 \T Cl'.._,..,..., f'tttl. ufli u-... 'k"4**uu
- 4..:> o I\ fl l*,M MJI* el110*C,. ...,,, ... , .. , .......... ,..., " , ..... 4**1 FLORIDA POWER & LICHT COMP Ml'( ST. LUCIE PU.HT UNIT COtH !(;IJJ</iTION AND DE1 Al LS ro:1 f'lf'E WHIP RESTRAIMTS TYPICAL FIGURE 3 .6-3
2400
- 2200 2000 1800 ]. w er. 1600 :::> Cl) (I) w ex: c..
- 1400 1200 1000 0
- 2 4 ---Experimental c 6 TIME (msec) AOR-216 8 10 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 PIPE RUPTURE PRESSURE HISTORY FIGURE 3.6-5
- Qi "' "' "' QI QI > <c QI :; QI 'N "' N "' QI 0 0:: 2 I I > z c.. <t:
- 0 0 0 0 0 0 0 & 0 0 0 o* CD* N* co N ... ... (qL) .lSm:IH.l
- I I I I I (! -z I <t: x I > I I I I I I I I 0 N C!J "';' ex: 0 <t: co C!J u w i== '<t N c 0 0 0 0 f o. °i' FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 COMPARISON OF MEASURED AND CALCULATED THRUSTS FIGURE 3. 6-6
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 GROWTH ALONG THE AXIS OF THE NESS OF A NONISOTHERMAL DIMENSIONAL GAS JET IN A STATIONARY MEDIUM, (INFLUENCE OF PARAMETER 0). FIGURE 3.6-7
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT l INCREASE OF THE THICKNESS OF A ISOTHERMAL AXIALLY SYMMETRIC JET OF GAS IN A STATIONARY ENVIRONMENT (INFLUENCE OF PARAMETER 0) FIGURE 3.6-8 I *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PL.A.MT UHIT 1 THE LATERAL BOUNDARY OF A TWO-PHASE JET FIGURE 3 .. 6-9
- *
- S* u E* u ----Typical Curve -* -Simplified Stress-Strain Curve I FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 TYPICAL AND SIMPLIFIED STRESS STRAIN CURVES FIGURE 3.6-9A
- * (A)
- y COMPRESSIVE REGION TENSILE REG I OH (B} St (Sc) al 1 Sy < s't < It Sc) a 111 (C) S' t FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UHIT 1 TYPICAL STRESS DISTRIBUTION FOR PIPE COLLAPSE FIGURE 3.6-9H
Amendment No. 18, (04/01)
- +Y +Z +X 12"*$1-814-t>f
- fc0 "i.i, '* *'19"lt*1 112" cs.e P.W.RESTR. PIPE VIHIP RESTRAINT
- FLORIDA POWER &. LIGHT COMPANY ST. LUO( PLANT UMIT I REACTOR AUX. BUILDING SPRAY PUMPS* DISCH. PIPING P.W. REST R'S FlGlJU: ). 6-33
- / CO"IT ISOM. on ! .. ,. -0 ° v :r:-tm -to * )>)> * )>->tj -() H :::!?Z-t r--0 2-03::0 co :zS2::o Q :E tx1 GI -t )> rn c w -OV'lx -o po . :e"* ,.. °' . ;:oOJ > r I ..,.,l> % -w .-v-<c -t GI .i::-m -V'l-ur c: :r: c--tc2 :z -t cs ;:o 3:: z -() "ly, * * -u GI -t It'/' vr, o f' *s *o .... ('It )>-. 'n C'o z 0* u 0. -< '4'c
- 2[)'-0" N i'll.,..i)t, '°&..i,,o '*e s.o-s> -"I y +Y +Z +X
- 1-12-CS-S P.W. RESTR. PIPE WHIP RESTRAINT
- """'+.-,, .... ,.,Q\. ,, +V +Z +X EL. 12.0 EL.4IMI MS * ,,,E WHIP RESTRAINT EL.113". I 1/32 J.74' u.oo l(l " 4.25' 11.724 * §':T2 . 311** .. RE I 38" MS-62 _ a: "'..J ffi !; E "' ' ,. --'"'"" "
- Amendment No. 16, (1/98) FLORIDA POWER & LIGHT COMPANY ST. LUOI! PLANT UHIT 1 REACTOR BUILDING MAIN STEAM PIPING RESTRAINTS 11Qlll! ). 6-3S
- +Y "'>l +Z +X MS
- PIPE WHIP RESTRAINT * .. ;:/ ,,,ro 11.138 EL. 62-0 EL. 49-6 11.828 4.25 Amendment No. 16, (1/98) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT l REACTOR BUILDING MAIN STEAM PIPE WHIP RESTRAINTS FIGURE 3.6-36
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE UMIT STEAM TRESTLE MAIN STEAM PIPE WHIP RESTRAINTS FIGURE 3.6-37
- +Y +'>l +X +Z MSH IS A ""E WHICH ALSO SERVES AS A Pll't WHIP RESTRAINT
- ii: " :;) 0 a:!:! wO: uW zl'. WO. = :g "'" ot-t-:!:
- FLORIDA POWER & LIGHT* COMPANY ST. LUCIE PLAlfT UMIT 1 STHM TRESTLE MAIN STEAM PIPE WHIP RESTRAINTS FIGURE 3.6-38
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT REACTOR BUILDING FEEDWATER PIPE WHIP RESTRAINTS FIGURE 3.6-39
- * ......... --< I
- x + N + \ .... z < a: .... Cl) w a: CJ) 0... u.. ::t: co ;;:: w 0... ii: LI.. co FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UMIT REACTOR BUILDING FEEDWATER PIPING PIPE WHIP RESTRAINTS FIGURE 3.6-40
- \ I \
- x + > + I N + * \ ... z :cc a: ti w a: c. i w c. :r u all u a: In N h R c: . gj w w ... a: z c. C) * -a: Cl) w w c. Q 0 00 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UHIT 1 REACTOR BUILDING CHEMICAL VOLUME CONTROL P.W. RESTR'S FIGURE 3.6-41
- ./ * \ \ \ x + > + I N +
- 0 !!? :;:! M o. *N I--0 ::c: uu \ \ z 0 "7 .N I-.. z C')I 0 ::c: uu oe I ::2 I .,. CL? M N \o C! ;;; FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT l eves PIPE WHIP RESTRAINTS FIGURE 3.6-42
- 1 Ill .. +/- I.I \
- x + > + \ I N +
- I/) .... z < a: .... I/) w a: Q. i w Q. A: ::c u 0 !:'.: z 0 c;< " .('< .... -z c;< 0 J: uu I 0 !'l M I II ':4 ...: w w .... a: z Q. C) ri ;:; ... w Q. 0 0 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UHIT REACTOR BUILDING eves PIPE WHIP RESTRAINTS FIGURE 3.6-43
- *
- N + ::c CJ :c u FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING eves PIPE WHIP RESTRAINTS FIGURE 3.6-44
- 1\1 +Z ,, r 0 () ;o < "' 0 () :-t )> I (./'I ;o '"J:j -Um r--u H :0> co mn n :E -m :E-1 m ;o VJ -UOJ ;oC I mj= -t C> .i::-U'lo :x: °' c -I ;oz )> (;) -t 0 z .... -I (./'I )> z -< +Y '
- 136°-0* 2"-CH-109 --, CH-67 ,___t .* .,.,
- REGENERATIVE HEAT EXCHANGER Cl-t-65
- NO PIPE HANGERS
- PIPE WHIP REST. CH PIPE WHIP RESTRAINT
- * * +Y +Z DESIGN TEMP.= 550°F OPER. PRESS." 2200 PSI +X 9" CH-86 :i (.) x w ...: w :c ...J z N N w 0 c:i z w a: N 1" DRAIN CH -PIPE WHIP RESTRAINT FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UMIT 1
- REACTOR BUILDING eves PIPE WHIP RESTRAINTS FIGURE 3. 6-47
- *
- REFER TO DRAWING 8770-G-799, Sheets 5 & 8 Amendment No. 16, (1/98) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING PIPE RESTRAINTS (SG lA BLOWDOWN PIPING) Fimm:E 3.6-48
- *
- REFER TO DRAWING 8770-G-799, Sheets 10, 11 & 15 Amendment No. 16, (1/98) FLORIDA & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING PIPE RESTRAINTS (SG lB BLOWDOWN PIPING) FIGURE 3.6-49
- * .lllJI. 33 136 I 11 I 34 .,,,,, 35; .,/,(,//1. 13 6 7 (MS-5) 8 C) " C"i -(MS-6) 9 (MS-7) 11 12 REF. ISOMETRIC: MS -147*1 10 MATERIAL -CARBON STEEL PRESSURE -985 PSI TEMP. -520 °F EH = 26.26 x 106 PSI YLD STR. = 27 ,800 PSI (MS-X) refers to restraint location in Fig.3.6-36 restraint location
- fictional node locations required for pipe whip analysis Amendment No. 16, (1/98) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 MAIN STEAM LINE INSIDE CONTAINMENT GUILLOTINE BREAK AT NODE #12 FIGURE 3.6*52 I * * +Y +ZA+x 23 19 22 '1f1' 18 '717T ..JJJ.JJ. 37 38 .,,,,. "" 13 i "t;..' ' .C2 ...,,,,.. REF. ISOMETRIC: MS-1.0-1 5 6 7 8 9 10 1¥ 0 ... -< ... 1¥ IU IU _, :c "'1::;) MATERIAL -C.t.RBOM STEEL PRESSURE -985 PSI TEMP. -520°F EH = 26.26 x 106 PSI YLD. STR. = 27,800 PSI Amendment No. 16, (1/98) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PL.t.MT UNIT 1 MAIN STEAM LINE INSIDE CONTAINMENT GUILLOTINE BREAK AT NODE #16 FIGURE 3.6-53
- *
- J.UL 10 9 11 .,,.,,. REF. ISOMETRIC: BF -149-l 15 13 ..,,,,. .JJJ.J 18 17 .,.,,,.. MATERIAL -CARBON STEEL PRESSURE -1100 PSI TEMP. -440°F EH = 26.76 x 1o6 PSI YLD. STR. 29,320 PSI. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 BOILER FEEDWATER LINE OUTSIDE CONTAINMENT GUILLOTINE BREAK AT NODE #7 FIGURE
- +Y +z +X
- 6 10 REF. ISOMETRIC: BF-147-1
- MATERIAL : CAR BON STL. PRESSURE: 1100 PSI TEMP: 440 Of EH = 26.76 x 106 PSI YLO. STR. = 29,320 PSI FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 BOILER FE EDWATER LINE INSIDE CONTAINMENT GUILLOTINE BREAK -NODE #7 FIGURE
- *
- lo' -DATA USED 1M 'PLAST ..; .. = ---DAT A OllTPUT Of 'RELAP' .. ... .. 0 H \\ ,l ,, FEEOWATER 0.100 a1so (1.b50 (SEC; FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT I SLOWDOWN DATA FOR MAIN STEAM AND FE EOWA HR LINES FIGURE 1.6-'>6
- 2.0 2KPA = 2.07 x6 LBS 1.6 I \ 1.2 I \ \ 0.8
- I vi 0.4 0 0.1 0.15
- I ....... \ I\ : \ \ I \ I LEGEND FOR MODE II AND REACTION DIRECTION. i'tODE 9 ( + Z) ----NODE 9 (+ X) ---HOOE 13 IN DIRECTIOt-1 OF UNIT VECTOR. v = 0.67 x . 0.7 42 z 0.2 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAl-IT UNIT l MAIN STEAM LINE INSIDE CONTAINMENT (BREAK AT NODE 16) REACTIONS AT PIPE WHIP RESTRAINTS (NODES 13 & 9) VERSUS TIME FIGURE 3.6-57
- ..£"' 0 >< ,,,., al °' w 0 0 z
- t-< ,,,., z 0 j:: u < w Cl::
- 1.5 -:-l. -0.5 -2.72 x 105 :: 2.34 x 106 :* -=-2 _KP_A_=_2._0_7 !I LEGEND FOR DIRECTION spring reaction : I I I ! I: I ; ! l I I ; i I'.! : 1 :i 11 I i 1. I' i.I * !'* ! ' . :!! I II l I j lit J \ o. . i .j1 I :l I .01 TIME (SEC.) .1 F'....ORIDA POWER & LIGHT COMPANY ST. LL1CI E PLANT UNIT 1 //, i'i STEA'.' INSIDE CONTAINMENT 1*.:o:_µx AT 12, REACTION AT PIPE \'.r,.r RESTR!--'r\T AT tWDE 9 IN+ X & + Z DIRECl!OtiS VERSUS TIME Fi GU RE J. 6-58
- *
- 1.6 -1.2 .... iii'"" (2 KPA) c ... )( Cl) al ..J 0.8 ,.... a: f2 z 0 j::: (,J <t'. 0.4 .... 0. 0.01 I TIME (SEC) I 0.1 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAl'IT UNIT 1 FW LINE INSIDE CONTAINMENT REACTION AT PIPE WHIP RESTRAINT AT NODE 6 VS. TIME FIGURE 3. 6-59
- Vi' II.I :c u .... :z: w w
- u 4( ..J Q. V) Ci
- 8. 6. 4. 0.01 TIME (SEC.) 0.1 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PlAHT UHIT 1 MAIN STEAM LINE INSIDE CONTAINMENT BREAK AT NODE 16 DISPLACEMENT OF NODE 16 VERSUS TIME MAX. PLASTIC STRAIN =O. FIGURE 3.6-60
- v;-w ::c u * :z: 0 t= u w ..J I.I. w 0
- 6. 4
- 0.01 0.1 TIME (SEC.) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 34u MAIN STEAM LINE INSIDE CONTAINMENT (BREAK AT NODE 12) DEFLECTION OF NODE 12 VERSUS TIME MAX. PLASTIC STRAlN = .00762 FIGURE 3. 6-61
- ,.... w 0 0 z z <( Ci <( a: N 0 ...
- Cl) x <( > I-::i 0 a:l <l'. z 0 ;::: <( I-0 a:
- 5. 2.5" GAP -1.031" WALL MAX PLASTIC STRAIN 4. IN PIPE= 0.00259 3. 2. 1. TIME (SEC) FLORIDA POWER & LIGHT COMPAr<'r ST. LUCIE PLANT UNIT l 20" FW LINE INSIDE CONT A!NMENT ROTATION OF NODE 7 VS. TIME FIGURE 3. {>-62
- 0.8 -4" GAPS -1.5" WALL TH. MAX. FORCE= 0.7806 x 106# (2 KPA = 1. x 106 LBS) **-*-@ 0.04904 SEC 0.6 .... f.0-Q ... )( Cl) a:i ..J w u a: 0 0.4 ......
- u.. z ' 0 i= u <( w a: 0.2 ..... ' ' 0. 11 0.01 0.1 TIME (SEC) FLORIDA POWER & LIGHT COMPANY
- ST. LUCIE PLANT UNIT 1 20" FW LINE OUTSIDE CONT REACTIONS AT PIPE WHIP RESTRAINT AT NODE 5 VS. TIME FIGURE 3.6-63
""' w Q 0 z < u; -10. -S. -6. u z z 0 (.) w u:: -4. w 0 ..J <t (,J . a:: w > -2. INITIAL YIELD @NODE #5 (0.04904 SEC) u w Cl) C). dj 9.255" .. -.. ----*-.. --9.074" TIME (SEC) (4" GAP -1.5" WALL THICKNESS) MAX PLASTIC STRAIN IN PIPE= .00148 Amendment No. 18, 04/01 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 20" FW LINE OUTSIDE CONTAINMENT DISPLACEMENT AT NODE 7 VS. TIME FIGURE 3.6-64
- 1. -.8 -ID-0 ... x .6 -_, w (,) a: 0 u..
- z 0 .4 j:: lo-(,) <( w a: .2 -0. 0.01 * (2" GAP -1" WALL THICKNESS) 2 KPA = 1 x 106 LBS. 0.8721 x 106 # \ \ TIME (SEC) I ' a 0.1 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 20" FW LINE OUTSIDE CONTAINMENT REACTIONS AT PIPE WHIP RESTRAINT AT NODE 5 VERSUS TIME FIGURE 3.6-65
- 8. ;; 6. (.) z """ z w (.) < 4 . ..I Q. "' 6 ..I < (,.) t:= 2. > (2" GAP -1" WALL THICKNESS) MAX. PLASTIC STRAIN IN PIPE. 0.0065 TIME (SEC) rJ w Cl) en , --: I ' 6.20T Amendment No. 18, 04/01 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 20" FW LINE OUTSIDE CONTAINMENT DISPLACEMENT OF NODE 7 VS. TIME FIGURE 3.6-66
- .8 -.6 ..... 0 ... )( Cl) al ..J
- w u a: .4 -0 LI. z 0 (.) <( w a: .2 -0. 0.01 * (2" GAPS -1.5" WALL THICKNESS) (2 KPA = 1. >< 106 LBS) I I I I I I I I TIME (SEC) .1 FLORIDA POWER & LIGHT (')MPANY ST. LUCIE PLANT UNIT 1 20" FW LINE OUTSIDE CONT Al NM ENT REACTIONS AT PIPE WHIP RESTRAINT AT NODE 5 VS. TIME FIGURE 3. 6-6 7
- -5. -4. en w ::c: (.) z -3.
- z 0 j:: (.) w ..I u. w c ..I -2. <t (.) j:: a: w > -1. 0 0.01
- c ..I w > ..I <t j:: I cJ w en -.... CD 0 0 5.446" (2" GAP -1.5" WAll THICKNESS) MAX PLASTIC STRAIN IN PIPE= .0072 0.1 TIME (SEC) FLORIDA POWER & LIGHT COMPAt<Y ST. LUCIE PLANT UNIT 1 20" FW LINE OUTSIDE DISPLACEMENT OF NODE 7 VS. T :lt.E FIGURE 3. 6-68
TABLE 3.7-8COMPARISON OF STRUCTURAL RESPONSESFOR RESPONSE SPECTRA AND TIME HISTORY SEISMICANALYSIS METHODSN-S Direction (0.1 G Horizontal Earthquake)ResponseReact. Aux. Bldg.Response SpectraTime History Elevation Method Method 82.00.382g0.40g 62.00.345g0.28g 43.00.292g0.23g 19.50.214g0.23g - 0.50.134g0.30gE-W DIRECTION (0.1 G HORIZONTAL EARTHQUAKE)ResponseReact. Aux. Bldg. Response SpectraTime History Elevation Method Method 82.00.360g0.52g 62.00.325g0.40g 43.00.279g0.36g 19.50.219g0.34g - 0.50.166g0.34g3.7-57 TABLE 3.7-9REACTOR BUILDING NATURAL PERIODS OF VIBRATION (SEC)HORIZONTALMODE E=48000 psi E=44000 psiE=40000 psi E=36000 psi E=32000 psi 10.67660.70460.73730.77400.8180 20.27250.28400.29710.32210.3302 30.13650.13700.13750.13790.1383 40.08790.08810.08840.08860.0888 50.07160.07160.07160.07160.0716 60.04920.04930.04930.04940.0494 70.04210.04210.04210.04210.0421 80.03830.03830.03830.03830.0383 90.02830.02830.02830.02830.0283 100.02320.02320.02320.02320.0232VERTICAL MODEE=48000 psi E=44000 psiE=40000 psi E=36000 psi E=32000 psi 10.44470.46470.48760.51440.5461 2 0.08280.08280.08280.08290.0829 30.04340.04340.04340.04350.0435 40.03790.03790.03790.03790.0379 50.02270.02270.02270.02270.0227 60.01700.01700.01700.01700.0170 70.01640.01640.01640.01640.0164 80.01550.01550.01550.01550.0155 90.01290.01290.01290.01290.0129100.01120.01120.01120.01120.01123.7-58
TABLE 3.7-15FUEL HANDLING BUILDINGNATURAL PERIODS OF VIBRATION (SEC)HORIZONTALModeE=42000 psiE=38500 psiE=35000 psiE=31500 psiE=2800 psiN-S E-WN-S E-WN-S E-WN-S E-WN-S E-W 1.4649 .5303.4854 .5533.5090 .5804.5363 .6111.5687 .6475 2.2450 .1931.2351 .2015.2673 .2112.2812 .2222.2977 .2355 3.0592 .0665.0593 .0666.0593 .0666.0594 .0666.0594 .0667 4.0303 .0313.0304 .0314.0304 .0314.0305 .0315.0305 .0315 5.0166 .0178.0166 .0178.0166 .0178.0166 .0178.0166 .0178VERTICALModeE=42000 psiE=38500 psiE=35000 psiE=31500 psiE=2800 psi10.24400.25470.2672 0.28130.298220.09500.09510.09510.09510.095130.04730.04730.04730.04730.047340.02670.02670.02670.02670.026750.01410.01410.01410.01410.014160.00750.00750.00750.00750.00753.7-64
- *
- TABLE 3. 7-16 FUEL HANDLING BUILDING HORIZONTAL STRUCTURAL RESPONSES N-S Mass E = 42000 psi E = 38500 psi E = 35000 psi E = 31500 psi E = 28000 psi Envelop No.
- A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) l .1637 .0158 .1621 .0168 .1602 .0180 .1582 .0194 .1561 .0211 .1637 .0211 2 .1719 .0122 .1716 .0130 .1711 .0141 .1706 .0153 .1700 .0169 .1719 .0169 w 3 .1790 .0111 .1790 .0119 .1789 .0130 .1787 .0142 .1786 .0157 .1790 .0157 . ...... Base .1942 .0097 .1952 .0106 .1962 .0117 .1972 .0130 .1983 .0146 .1983 .0146 I °' \.Jl E-W Mass E = 42000 psi E = 39500 psi E = 35000 psi E = 31500 psi E = 28000 psi Envelop No. A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) 1 .2003 .0457 .1947 .0484 .1886 .0515 .1822 .0552 .1756 .0597 .2003 .0597 2 .1423 .0286 .1397 .0303 .1364 .0323 .1331 .0347 .1295 .0375 .1423 .0375 3 .1404 .0218 .1396 .0231 .1378 .0247 .1360 .0265 .1340 .0287 .140/i .0287 Base .1832 .0075 .1860 .0081 .1867 .0083 .1873 .0097 .1880 .0107 .1880 .0107
- *
- TABLE 3. 7-17 FUEL HANDLING BUILDING VERTICAL STRUCTURAL RESPONSES Mass E = 42000 psi E = 38500 psi E = 35000 psi E = 31500 psi E = 28000 psi Envelop No. A (g) D (ft) A (g) D (ft} A (g) D (ft) A (g} D (ft) A (g) D (ft} A (g) D (ft) 1 .1486 .0072 .1485 .0079 .1483 .0086 .1481 .0096 .1479 .0107 .1486 .0107 2 .1468 .0071 .1468 .0078 .1467 .0085 .1467 .0095 .1467 .0106 .1468 .0106 3 .1464 .0071 .1464 .0077 .1464 .0085 .1464 .0094 .1464 .0106 .1464 .0106 4 .1522 .0074 .1517 .0080 .1512 .0088 .1507
- OOCJ 7 .1503 .0109 .1522 .0109 5 w .1738 .0084 .1711 .0090 .1685 .0098 .1660 .0107 .1636 .0118 .1738 .0118 -...J Base .1443 .0070 .1445 .0076 .1447 .0084 .1448 .0093 .1450 .0105 .1450 .0105 I Q'\ "'
- TABLE 3. 7-18 INTAKE STRUCTURE NATURAL PERIODS OF VIBRATION (SEC} HORIZONTAL Mode E=48000 psi E=44000 psi E=40000 psi E=36000 psi E=32000 psi N-S E-W N-S E-W N-S E-W N-S E-W N-S E-W 1 .2782 .2824 .2903 .2946 .3043 .3086 .3248 .3392 2 .1025 .0999 .1069 .1042 .1120 .1092 .1150 .1245 3 .0294 .0190 .0295 .0191 .0296 .0191 .0192 .0297 4 .0253 .0163 .0255 .0164 .0257 .0164 .0165 .0262 5 .0206 .0124 .0206 .0125 .0206 .0125 .0125 .0207 VERTICAL Mode E=48000 psi E=44000 psi E=40000 psi E=36000 psi E=32000 psi 1 0.2550 0.2662 0.2791 0.2941 o. 3119 2 0.0165 0.0165 0.0165 0.0165 0.0165 3 0.0090 0.0090 0.0090 0.0090 0.0090
- 4 0.0070 0.0070 0.0070 0.0070 0.0070 5 0.0057 0.0057 0.0057 0.0057 0.0057
- 3.7-67
- *
- TABLE 3. 7-19 INTAKE STRUCTURE STRUCTURAL RESPONSES HORIZONTAL N-S Mass E = 48000 psi E = 44000 psi E = 40000 psi E = 36000 psi E = 32000 psi Envelop No. A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) 1 .2987 .0186 .2987 .0202 .2976 .0221 --.2904 .0267 . 2987 .0267 2 .2085 .0131 .2087 .0143 .2081 .0157 --.2030 .0190 .2087 .0190 3 .1509 .0087 .1519 .0096 .1525 .0106 --.1522 .0129 .1525 .0129 Base .1231 .0038 .1262 .0042 .1295 .0046 --.1360 .0058 .1360 .0058 HORIZONTAL E-W \..V . Mass E = 48000 psi E = 44000 psi E = 40000 psi E = 36000 psi E = 32000 psi Envelop ........ I No. A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) (j\ CXl 1 .2934 .0188 .2934 .0205
- 29 ll .0223 .2879 .0244 --.2934 .0244 2 .2223 .0145 .2223 .0157 .2209 .0172 .2183 .0188 --.2223 .0188 3 .1720 .0109 .1725 .0119 .1721 .0130 .1707 .0143 --.1725 .0143 Base .1303 .0068 .1320 .0075 .1334 .0082 .1347 .0091 --.1347 .0091 VERTICAL Mass E = 48000 psi E = 44000 psi E = LOOOO psi E = 36000 psi E = 32000 psi Envelop No. A (g) D (ft) A (g) D (ft} A (g) D (ft} A (g) D (ft) A (g) D (ft) A (g) D (ft) 1 .1477 .0078 .1476 .0085 .1475 .0094 .1474 .0104 .1460 .0116 .1477 .0116 2 .1473 .0078 .1473 .0085 .1472 .0094 .1471 .0104 .1457 .0116 .1473 .0116 3 .1466 .0077 .1466 .0085 .1466 .0093 .1466 .0103 .1453 .0115 .1466 .OllS 4 .1477 .0078 .1476 .0085
- 1476 .0094 .1475 .0104 .1460 .0116 .1477 .0116 Base .1456 .0077 .1457 .0084 .1458 .0093 .1459 .0103 .1446 .0115 .1459 .0115 TABLE 3. 7-20
- DIESEL GENERATOR BUILDING NATURAL PERIODS OF VIBRATION (SEC) HORIZONTAL E=36000 psi E=33000 psi E=30000 psi E=27000 psi E=24000 psi Mode N-S E-W N-S E-W N-S E-W N-S E-W N-S E-\:' 1 .1915 .1969 .1997 .2053 .2095 .2150 .2205 .2262 .2336 . 2396 2 .0928 .0915 .0969 .0952 .1016 .0997 .1070 .1049 .1134 .1111 3 .0282 .0377 .0282 .0378 .0282 .0379 .0283 .0380 .0283 .0381 4 .0150 .0311 .0150 .0311 .0150 .0311 .0150 .0311 .0150 .0311 5 .0141 .0223 .0141 .0223 .0141 .0223 .0141 .0223 .0141 .0224 6 .0101 .0171 .0101 .0171 .0101 .0171 .0101 .0171 .0101 .0172 VERTICAL Mode E=36000 psi E=33000 psi E=30000 psi E=27000 psi E=24000 psi 1 0.1363 0.1492 0.1572 0. 1666 2 0.0458 0.0459 0.0459 0.0459 3 0.0115 0.0115 0.0115 0. 0115 4 0.0057 0.0057 0.0057 0.0057 5 0,0053 0.0053 0.0053 0.0053
- 6 0.0046 0. 0046 0. 0046 0.0046
- 3.7-69
- *
- TABLE 3. 7-21 DIESEL GENERATOR BUILDING STRUCTURAL RESPONSES Mass E = 36000 psi E = 33000 psi E = 30000 psi E = 27000 psi E = 24000 psi Envelop U} No. A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) A (g) D (ft) I z 1 .2646 .0079 .2678 .0087 .2673 . 000 5 .2668 .0105 .266L .0118 .2678 .Oll8 2 .2336 .0070 .2365 .0077 .2362 .0084 .2359 .0093 .2355 .0105 .2365 .0105 3 .2024 .0060 .2052 .0067 .2054 .0073 .2054 .0081 .205L .0091 .2054 .0091 4 .1983 .0059 .2015 .0065 .2020 .0072 .2023 .0080 .2026 .0090 .2026 .0090 5 .1983 .0059 .2015 .0065 .2020 .0072 .2023 .0080 .2026 .0090 . 2026--'
- 0090 w Base .1756 .0052 .1784 .0057 .1791 . .0063 .1796 .0070 .1801 .0079 .1801 .0079 ......, I ......, 1 .2873 .0090 .2880 .0098 .2873 .0107 .2865 .0118 . 2859 .0132 .2880 .0132 0 I 2 .2429 .0077 .2437 .0084 .2432 .OC92 .2427 .0101 .2421 .0113 .2437 .0113 3 .1983 .0063 .1994 .0068 .1995 .OC75 .1996 .0083 .1997 .0093 .1997 .0093 4 .1941 .0061 .1956 .0067 .1960 .OC74 .1965 .0082 .1969 .0092 .1969 .0092 5 .1941 .0061 .1956 .0067 .1960 .0074 .1965 .0082 .1969 .0092 .1969 .0092 Base .1608 .0050 .1626 .0054 .1634 .0060 .1645 .0067 .1655 .0095 .1655 .0075 1 .1251 .0019 --.1334 .0024 .1359 .0027 .1385 .0031 .1385 .0031 2 .1247 .0019 --.1330 .0024 .1356 .0027 .1382 .0031 .1382 ' . 0031 .-4 tll .1377'*'0.0034 () 3 .1239 .0019 --.1323 .0024 .1350 .0027 .1377 .0034 *..-! µ ,... 4 .1421 .0021 --.1481 .0027 .1492 .0030 .1505 .0031
- 1502 .0031 <l) :> 5 .1233 .0019 --.1317 .0024 .1344 .0027 .1372 .0031 .1372 .0031 6 .1233 .0019 --.1317 .0024 .1344 .0027 .1372 .0031 .1372 .0031 Base .1230 .0019 --.1315 .0024 .1342 .0027 .1370 .0031 .1370 .0031 TABLE 3.7-21A
- NODE FREQUENCY DISPL.'\CE!'.E:lT AT T = !. 98 s=O OR EBASCO DYNA:llC EBASCO DYNA.i.'1IC HASS PT. ANALYSIS PROG. STARDYNE ANALYSIS PROG. STARDY:JC 1 2.4373 2.4346 0.01631 0.01635 2 5.0158 5.0149 0.01568 0.01572 3 8.3874 8.3599 0.01154 0.01458 4 8. 9140 8.9068 0.01368 0.01372 5 11. 3003 11.2466 0.01319 0.01323 6 13.1218 11.1031 0.01227 0.01231 7 15.5296 15.4381 0.01091 0.01093 8 17.7919 17.7128 0.00979 0.00982 9 18.4370 18. 3135 0.00885 0.00887 10 22.1156 22.0018 0.00781 0.00782 11 23.9480 23. 8722 0. 01484 0.01488 12 26.0547 25.9860 . 0.01448 0.01452 13 28.0716 27 .9681 0.01412 0.01417 14 30.7635 30.5644 0.01398 0.01402 15 33.2434 33.1839 0.01366 0. 01370 16 36.2133 36.0596 0.01319 0.01323 17 37.0848 36.8921 0.01262 0.01265 18 38.5026 38.2458 0.01207 0.01210 19 43.8437 43.6126 0.01153 0. 0115(1 I
- RPfer to Figure 3.7-llA for the representative model
- 3. 7-71 TABLE 3. 7-21B
- COMPARISON BET\'EFN RFRM OTHER Rf.FF.RD\CF:S A. Sample Problem No. 1: Reference -Structural Theory Sutherland & Bowman Page 264 6K 3 I 1'o" -r I I 21 20' 6K )Ii 2 I ....., ---5 moment sign convention: I 21 301 1 6 ;, .. ' . ' ', ,, ,,.
- f 201 l 1 RFRM RESULTS REFERENCE RESULTS FOR THE MEMBER 3 4 FX( 3 ) = 2.61907 KIPS FY( 3 ) ... -4.22005 KIPS M ( 3 ) = 40.04487 FT-K M(3)
- 39.7 FT-K FX( 4 ) = -2.61907 KIPS FY( 4 ) = 4.22005 KIPS M ( 4 ) = 44.35622 FT-K M(4)
- 43.8 FT-K AXIAL FORCE == -2.619 KIPS FOR THE MEMBER 2 5 FX( 2 ) . 4.90394 KIPS FY( 2 ) = -9.02556 KIPS M ( 2 ) = 85.26272 FT-K M(2)
- 84.5 FT-K FX( 5 ) --4.90394 KIPS FY( 5 ) . 9.02656 KIPS M ( 5 ) . 95.26841 FT-K M(5)
- 94.7 FT-K AXIAL FORCE * -4.904 KIPS
- 3.7-72 TABLE3.7-21B (Cont'd)
- RFR..111 RFSVLTS REFERENCE RESULT FOR TilE MENBER 3 2 FX( 3 ) = 3.38093 KIPS FY( 3 ) = 4.22005 KIPS M ( 3 ) -40.04487 FT-K M(3) = 39.7 FT-K FX( 2 ) -3.38093 KIPS FY( 2 ) = -4.22005 KIPS M ( 2 ) = -27 .57370 FT-K M(2) = -26.7 FT-K AXIAL FORCE = 4.220 KIPS FOR TIIB MEMBER 4 5 FX( 4 ) = 2.61987 KIPS FY( 4 ) -4.88005 KIPS M ( 4 ) -44.38622 F1'-K M(4) = -43.8 FTM K FX( 5 ) -2.61907 KIPS FY( 5 ) = -.22005 KIPS M ( 5 ) = -5.02521 FT-K M(5) = -5.7 FT-K AXIAL FORCE = -4.220 KIPS
- FOR TIIE MI:MDER 2 1 FX( 2 ) = 4 .-4 7689 KIPS FY( 2 ) = 13.24661 KIPS M ( 2 ) -57.58-02 FT-K M(2) --57.8 FT-K FX( 1 ) = -4.47683 KIPS FY( 1 ) = -13.24661 KIPS M ( 1 ) = -76.62087 FT-K M(l) = -75.0 FT-K AXIAL FORCE = -13.247 KIPS FOR TIIB MEMBER 5 6 FX( 5 ) = 7.52301 KIPS FY( 5 ) = -13.24661 KIPS M ( 5 ) = -87.24319 FT-K M(S) = -87 .O FT-K FX( 6 ) --7.52301 KIPS FY( 6 ) = .13.24661 KIPS M ( 6 ) ... -138. 44721 FI'-K M(6) =-136.4 FT-K AXIAL FORCE = -13.247 KIPS
- 3.7-73
- *
- TABLE 3.7-21B (Cont'd) B. Sample Problem No. 2: Reference -Reinforced Concrete Structure (CASE A) Peabody 73 I ! l14' I I 2 16' 28' 1 ; FOR TI!E MEMBER FOR TI!E MfilffiER 5 6 RFRM RESULTS 3 4 FX( 3 ) = FY( 3 ) = M ( 3 ) = FX( 4 ) = FY( 4 ) = M ( 4 ) AXIAL FORCE = 2 5 FX( 2 ) = FY( 2 ) = 20' 7 8 12.09692 27. 87728 -96.73661 -12.09692 28.12272 100.17281 -12.097 -7.05437 39.74361 M ( 2 ) = -125.91879 FX( 5 ) 7.05437 FY( 5 ) = 44.25639 M ( 5 ) = 189.09767 AXIAL FORCE = 7.054 3.7-74 Pages 388 to 393 moment sign convention: ?+ REFERENCE RESULTS KIPS KIPS FT-K M(3) = -97.0 FT-K KIPS KIPS FT-K M(4) = 100.8 FT-K KIPS KIPS KIPS FT-K M(2) =-122.1 FT-K KIPS KIPS FT-K M(5) = 197.1 FT-K KIPS TABLE 3.7-21B (Cont'd)
- RFRM RESULTS REFF.RENCE RESULT FOR THE MEMBER' 5 7 FX( 5 ) = 2.75273 KIPS FY( 5 ) = 23.28177 KIPS M ( 5 ) = -95.30735 FT-K M(5) =-103.9 FT-K FX( 7 ) = -2.75273 KIPS FY( 7 ) = 16. 71823 KIPS M ( 7 ) = 29.67202 FT-K M(7) = 24.5 FT-K AXIAL FORCE = -2.753 KIPS FOR THE MEMBER 3 2 FX( 3 ) = -12.09692 KIPS FY( 3 ) = -27. 87728 KIPS M ( 3 ) = 96.73661 FT-K M(3) = 97.0 FT-K FX( 2 ) = 12. 09692 KIPS FY( 2 ) = 27.87728 KIPS M ( 2 ) = 72.62025 FT-K M(2) = 73.5 FT-K
- AXIAL FORCE = -27. 877 KIPS FOR THE MEMBER 2 1 FX( 2 ) --5.04255 KIPS FY( 2 ) = -67.62089 KIPS M ( 2 ) ... 53.29854 FT-K M(2) = 48.7 FT-K FX( 1 ) = 5.04255 KIPS FY( 1 ) = 67.62089 KIPS M ( 1 ) = 27.38224 FT-K M(l) = 24.3 FT-K AXIAL FORCE = -67.621 KIPS FOR THE MEMBER 4 5 FX( 4 ) ... 12.09692 KIPS FY( 4 ) ... -28.12272 KIPS M ( 4 ) ... -100.17281 FT-K M(4) =-100. 8 FT-K FX( 5 ) ... -12.09692 KirS FY( 5 ) .... 28.12272 KIPS M ( 5 ) = -69.18404 FT-K M(5) = -69.l FT-K AXIAL FORCE = 28.123 KIPS
- 3.7-75
- TABLE 3.7-21B (Cont'd) RFR'1 RFSULTS REFERENCE RESULT FOR THE MEMBER 5 6 FX( 5 ) == 2.28982 KIPS FY( 5 ) -95.66088 KIPS M ( 5 ) = -24.60627 FT-K M(S) D -24.1 FT-K FX( 6 ) -2.28982 KIPS FY( 6 ) = 95.66088 KIPS M ( 5 ) = -12.03081 FT-K M(6) = -12.1 FT-K AXIAL FORCE = -95.661 KIPS FOR THE MEMBER 7 8 FX( 7 ) = 2.75273 Kl PS FY( 7 ) ,.. -16. 71823 KIPS M ( 7 ) = -29.67202 FT-K M(7) = -24 .5 FT-K FX( 8 ) = -2.75273 KIPS FY( 8 ) = 16.71823 KIPS M ( 8 ) = -14,37167 FT-K M(8) == -12.3 FT-K AXIAL FORCE = -16. 718 KIPS
- c. Sample Problem 2 (CASE B) 2K/£t
- 3.7-76
- TABLE 3.7-21B (Cont'd) RFP-.:! RESULTS REFEr..r::::cE RESULTS FOR THE HEHBER 3 4 FX( 3 ) = 6. 38496 KIPS FY( 3 ) = -2.43141 KIPS M ( 3 ) 28.26689 FT-K M(3) = 28.4 FT-K FX( 4 ) = -6.38496 KIPS FY( 4 ) 2.43141 KIPS }! ( 4 ) = 39.81273 FT-K M(4) = 39.8 FT-K AXIAL FORCE = -6.385 KIPS FOR TIIB MIDIBER 2 5 FX( 2 ) 21. 74075 KIPS FY( 2 ) = -9.70459 KIPS M ( 2 ) = 147.89689 FT-K M(2) = 147 .17 FT-K FX( 5 ) = -21. 74075 KIPS FY( 5 ) = 9.70459 KIPS M ( 5 ) 123.83150 Fr-K M(5)
- 124 .18 FT-K
- AXIAL FORCE = -21. 741 KIPS FOR THE MEMBER 5 7 FX( 5 ) = 10.35020 KIPS FY( 5 ) = -6.54840 KIPS M ( 5 ) = 59.19128 FT-K M(5) = 59.14 FT-K FX( 7 ) = -10.35020 KIPS FY( 7 ) == 6.54840 KIPS M ( 7 ) = 71.77667 FT-K M(7) = 72.76 FT-K AXIAL FORCE = -10.350 KIPS FOR TilE MEMBER 3 2 FX( 3 ) -6.38496 KIPS FY( 3 ) == 2.43141 KIPS M ( 3 ) = -28.26689 FT-K M(3) = -28.42 FT-K FX( 2 ) = -21. 61504 KIPS FY( 2 ) = -2.43141 KIPS M ( 2 ) = -78.34363 FT-K M(2) = -78.62 FT-K AXIAL FORCE = 2.431 KIPS
- 3. 7-77 TABLE 3.7-21B (Cont'd)
- RFRM RESULTS REFE.RENCE RESULTS FOR THE MEMBER 2 1 FX( 2 ) = -0.12571 KIPS FY( 2 ) = 12.13600 KIPS M ( 2 ) = -69.55325 FT-K M(2) =-68.55 FT-K FX( 1 ) = -31. 87429 KIPS FY( 1 ) -12.13600 KIPS M ( 1 ) -184.43536 FT-K M(l) =-182. 24 :n -K AXIAL FORCE = 12.136 KIPS FOR THE MEMBER 4 5 FX( 4 ) = 6.11496 KTPS FY( 4 ) = -2.43141 KIPS M ( 4 ) = -39.81273 FT-K M(4) = -39.84 FT-K FX( 5 ) -6.38496 KIPS FY( 5 ) = 2.43141 KIPS M ( 5 ) -49.57675 FT-K M(5) * -49.12 FT-K AXIAL FORCE = -2.431 KIPS
- FOR THE MEMBER 5 6 FX( 5 ) = 17. 77551 KIPS FY( 5 ) --5.58760 KIPS M ( 5 ) = -133.44603 FT-K M(5)=-134.30 FT-K FX( 6 ) = -17. 77551 KIPS FY( 6 ) = 5.58760 KIPS M ( 6 ) = -150.96212 FT-K M( 6)=-151. 07 FT-K AXIAL FORCE = -5.588 KIPS FOR THE MEMBER 7 8 FX{ 7 ) = 10.35020 KIPS FY( 7 ) = -6.54840 KIPS M ( 7 ) "" -71. 77667 FT-K M(7)= -72.76 FT-K FX( 8 ) = -10.35020 KIPS FY( 8 ) ... 6.54840 KIPS M ( 8 ) .. -93.82657 FT-K M(8)= -95 .16 FT-K AXIAL FORCE * -6.548 KIPS
- 3.7-78
- *
- TA.BLE 3.7-21B (Cont'd) D. Sample Problem No. 3: Reference -Moment Distribution J.M. Gere Page 102 & 366 -r* 12K 2 B 4 D moment sign convention RFRM RESULTS REFERENCE RESULTS FOR THE MEMBER 2 1 FX( 2 ) = -2.41838 KIPS FY( 2 ) = -9.37467 KIPS M ( 2 ) = 36.27563 FT-K M(2) 36.5 FT-K FX( 1 ) = 2.41838 KIPS FY( 1 ) = 9.37467 KIPS M ( 1 ) 0.00000 FT-K AXIAL FORCE = -9.375 KIPS FOR THE MEMBER 2 3 FX( 2 ) = 2.41838 KIPS FY( 2 ) = 9.37467 KIPS M ( 2 ) ... -36.27563 FT-K FX( 3 ) ... -2.41838 KIPS FY( 3 ) = 2.62533 KIPS M ( 3 ) = -7.03039 FT-K AXIAL FORCE = -6.356 KIPS 3.7-79 TABLE 3.7-21B (Cont'd)
- RFR'1 RESULTS REFERENCE RESULTS FOR THE MEMBER 3 4 FX( 3 ) = 2.41838 KIPS FY( 3 ) = -2.62533 KIPS M ( 3 ) = 7.03039 FT-K FX( 4 ) = -2.41838 KIPS FY( 4 ) 2.62533 KIPS M ( 4 ) = 21. 25242 FT-K AXIAL FORCE = -3.337 KIPS FOR THE MEMBER 4 5 FX( 4 ) = 1. 95960 KIPS FY( 4 ) ... -2.25075 KIPS M ( 4 ) ... -29.39406 FT-K M(4) * -29.4 FT-R FX( 5 ) = -1. 95960 KIPS FY( 5 ) = 2.25075 KIPS M ( 5 ) = 0.00000 FT-K AXIAL FORCE = -2.251 KIPS
- FOR THE MEMBER 4 6 FX( 4 ) = 0.45877 KIPS FY( 4 ) = -0.37458 KIPS M ( 4 ) = 8.10163 FT-K FX( 6 ) = -0.45877 KIPS FY( 6 ) = 0.37458 KIPS M ( 6 ) = 3. 97759 FT-K AXIAL FORCE = -0.243 KIPS FOR THE ME:t-ffiER 6 7 FX( 6 ) = 0.45877 KIPS FY( 6 ) = -0.37458 KIPS M ( 6 ) ... -3.97769 FT-K FX( 7 ) --0.45877 KIPS FY( 7 ) -0.37458 KIPS M ( 7 ) ... 6.88158 FT-K AXIAL FORCF. * -0.578 KIPS
- 3.7-80
- TABLE 3.7-21B (Cont'd) RESULTS REFERENCE RESULTS FOR THE MEMBER t 7 8 FX( 7 ) = o. 45877 KIPS FY( 7 ) = -0.37458 KIPS M ( 7 ) = -6.88158 FT-K M(7) = -7.0 FT-K FX( 8 ) = -0.45877 KIPS FY( 8 ) = 0.37485 KIPS M ( 8 ) = -0.00000 FT-K AXIAL FORCE = -0.375 KIPS *
- 3.7-81
- *
- TABLE 3.7-22 NATURAL FREQUENCIES-AND DOMINANT DEGREES OF FREEDOM (Historical Data Only) SHEET 1 Mode Number 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 ?.4 25 26 27 29 29 30 31 32 33 34 35 36 37 38 39 Note: Frequency .i£:etl 3.32 3.32 4.67 4.69 5.09 5.11 8.01 8.03 8.87 8.89 10.51 10.51 10.58 10.70 11. 01 11.12 12.14 12.16 19.84 19.86 23.06 23.10 23.65 24.52 24.56 27.39 27.68 27.80 30.27 37.19 39.22 39.24 39. 71 41.01 42.25 42.84 44.96 49.90 50.75 Ril Ril M61 M52 M66 M43 M66 M43 M61 M52 SGSA & B SG5A & B M61 M52 M66 M43 RI2 RI2 SG9A & B SG9A & B M65 M42 Ril M60 M51 SG9A & B SGSA & B SGSA & R SG9A & B Vl SGSA & B SGSA & B M60 M51 M65 M42 Vl Vl V4 Dominant Degrees of Freedom Direction 7. x X, Y, Z X, Y, Z x, y x, y Z, X z, x Z, X Z, X x x x, y x, y X, Y, Z X, Y, Z z x *Z z z z y z z x y y x z z z x x x x x z x Reactor Internals Reactor Internals Pump 1A2 Pump lBl Pump lAl Pump 1B2 Pump lAl Pump 1B2 Pump 1A2 Pump lBl Steam Generators lA & lB Steam Generators lA & lB Pump 1A2 Pump lBl Pump lAl Pump 1B2 Reactor internals Reactor Internals Steam Generators lA & lB Steam Generators IA & lB Pump 1A1 Pump 1B2 Reactor Internals Pump 1A2 Pump lBl Steam Generators lA & lB Steam Generators lA & lB Steam Generators lA & lB Steam Generators IA & lB Reactor Vessel Steam Generators lA & lB Steam Generators lA & lB Pump 1A2 Pump 1Bl Pump lAl l:'i.u:np 1B2 Reactor Vessel Reactor Vessel Reactor Vessel The steam generator modes (Modes 11, 12, 19, 20, 26, 27, 28, 29, 31, and 32 of Sheet 1) have been evaluated tor a conservative decrease of two percent in frequency to account for the additional mass of the replacement steam generator. All frequency changes for other components are insignificant . 3.7-82 Amendment No. 16, (1/98)
- Mode Frequency Number (cps) l 14. 2 61.14 3 62.60 4 167.82 5 209.94 -... .. -* 6 363.3? .. ------*-.. *-l '5. 6'5 " 2 ] -:<..8? **--*--3 16.42 4 20.94 5 24.48 6 30.65 7 -:cs.2-=< 8 l 451.21 9 69.70 10 74.15
- 84.56 . 11 12 lC::7,6'.<. 13 162.21 14 179.B3 15 206.49 16 210.qo 17 2s4.14 18 293.40 ---19 348.27 20 403.80 2J c:;15 .1] 22 62-:<..61 ___ gJ_ _____ hhR "ih 24 tS85.G4 --.. -25 785.8 --883.67 _21 ___ ***---*-,...._ ------* ------...
- Names M"l M5 M5 -*-*" MS TABLE 3.7-22 (Cont'd) SHEET 2 Dominant Degrees Direction x y -x x _.--11,L __ --x 1'.f? x -'---of Freedom Locations .Pressurizer " II II II " 1r.c; v Suri:re Line Hl x. _z __ " II M6, R3-.. !I !I x Hl y " " H3. M7 z II " 1,rr::: M4 y II " .v, r,f7. M:8 x II II M4. 11'5 x II II M'=I x. y II I! H3 y II II H3, M8 x II " M'5. M4 x z II II MS. M4 y II If v " II -M7 y II II M7 x II II -M7 y II ti M7 y II II ---_ .. M8, M7 y II II Miuvr8. MS x. Y. z II II M6. M4 z II II MS. M7 y II II .. M2 y 7 " " . M5_,_ M4 7 II II x. z " II ---.M'L z " II M2 y II II ---.... ----* ----------...... ----------------3.7-83 -* -----*--
- *
- TABLE 3.7-23 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (Historical Data Only) Seismic Load Seismic Component and Design Component Original RSG Excitation Location Calculated Maximum Maximum Combined Reactor Fx (kips) 60.4 62.2 X and Y Outlet Fy (kips} 15.2 15.7 Nozzle Fz (kips} 1. 0 1. 0 and Reactor Mx (in-kips} 15.8 16.3 Vessel Outlet My {in-kips) 95.3 98.2 Piping Mz {in-kips) 1267.5 1,305.5 Reactor Fx (kips) 50.8 52.3 Vessel Inlet Fy (kips) 17.9 18.4 Nozzle Fz {kips) 47.4 48.8 Mx {in-kips) 2262.8 2,330.7 My {in-kips) 1517.0 1,562.5 Mz (in-kips) 1876.8 l,933.1 Combined Reactor Fx {kips) 16.9 17.4 Zand Y Vessel Outlet Fy (kips} 2.4 2.5 Nozzle Fz (kips) 6.4 6.6 and Reactor Mx (in-kips} 150.0 154.5 Vessel Outlet My (in-kips) 800.0 824.0 Piping Mz (in-kips) 244.4 251. 7 Reactor Fx (kips) 29.7 30.6 Vessel Inlet Fy (kips) 12.4 12.8 Nozzle Fz (kip:;) 29.8 30.7 Mx (in-kips) 1874.8 1,931.0 My (in-kips) 731.1 753.0 Mz (in-kips) 1548.9 1,595.4 Specified for Design 241.0 253.0 10.0 1042.0 1041. 0 .:13733.0 190.0 146.0 57.0 9638.0 5981. 0 12770.0 30.0 172.0 40.0 1170.0 7521. 0 37446.0 61. 0 105.0 108.0 22392.0 10085.0 14087.0 All original calculated maximum load::;; ho.v<:: Lt:en evaluated for o. conservative increase of three percent, except for the steam generator upper support which has been evaluated for a conservative increase of eight percent, to account for the increase in mass and center of gravity of the replacement steam generator. These conservatively increased loads are given in the RSG Maximum column of this table. These values show that the seismic loads remained below the bounding specified for design loads with the replacement steam generators . 3.7-84 Amendment No. 16, (1/98)
- Component Seismic and Excitation Design Location Combined Steam X and Y Generator Inlet Nozzle and Steam Generator Inlet Piping Steam Generator Outlet Nozzle
- Combined Steam Z and Y Generator Inlet Nozzle and St:Parn Generator Inlet Piping Steam Generator Outlet Nozzle -*****
- TABLE 3.7-23 (Cont'd) SHEET 2 Component Original Calculated Maximum F:x (kip:::;) 60.9 Fy (kips) 52.7 Fz (kips) 1.0 Mx (in-kips) 73.8 My (in-kips) 85.8 Mz ( in-J<;:ips) 695.8 Fx (kips) 26.4 Fy (kips) 31. 7 Fz (kips) 26.4 Mx (in-kips) 3018.3 My (in-kips) 2708.2 Mz (in-kips) 3018.3 Fx (kips) 17.0 Fy (kips) 15.2 Fz (kips) 6.4 Mx (in-kips) 356.5 My (in-kips) 381.6 Mz (in-kips) 63.4 Fx (kips) 27.8 Fy (kips) 22.0 Fz (kips) 27.8 Mx (in-kips) 1866.0 My (in-kips) 1586.4 Mz (in-kips) 1866.0 Seismic Load RSG Specified Maximum for Design 62.7 119.0 54.3 68.0 1.0 4.0 76.0 743.0 88.4 708.0 716. 7 16828.0 27.2 30.0 32.7 134.0 27.2 33.0 3,108.8 1:2955.0 2,789.4 3161.0 3,108.8 6686.0 17.5 27.0 15.7 149.0 6.6 7.0 367.2 575.0 393.0 3793.0 65.3 11206.0 28.6 30.0 22.7 93.0 28.6 119. 0 1,922.0 6467.0 1,634.0 4251.0 1,922.0 7140.0 3.7-85 Amendment No. 16, (1/98)
- Component Seismic and Excitation Design Location Combined Reactor X and Y Coolant Pump Inlet Nozzle Reactor Coolant Pump Outlet Nozzle
- Com.biniad Reactor Z and Y Coolant Pump Inlet Nozzle Reactor Coolant Pump Outlet Nozzle
- TABLE 3.7-23 (Cont'd) SHEET 3 Component Original Calculated Maximum Fx (kips) 10.7 Fy (kips) 31. 7 Fz (kips) 4.5 Mx (in-kips) 1181.1 My (in-kips) 707.6 Mz (in-kips) 3357.1 Fx (kips) 51.0 Fy (kips) 17.2 l"Z (kips) 37.8 Mx (in-kips) 1347.2 My (in kips) 2180.3 Mz (in-kips) 2300.8 F.IC (kips) 15.3 Fy (kips) 21.9 Fz (kips) 11. 6 Mx (in-kips) 938.6 My (in-kips) 768.0 Mz (in-kips) 2451.3 Fx (kips) 29.8 Fy (kips) 12.4 Fz (kips) 19.7 Mx 839.2 My (in-kips) 1031.1 Mz (in-kips) 1262.0 Seismic Load RSG Specified Maximum for Design 11.0 78.2 32.7 137.0 4.6 37.9 1,216.5 5713 .2 728.8 4976.0 3,457.8 16461.0 52.5 199.3 17.7 145.9 38.9 40.0 1,387.6 3989.0 2,215.7 85S3. 9 2,369.8 16658.0 15.8 39.1 22.6 94.9 11. 9 109.9 966.8 11173.0 791.0 11862.0 2,524.8 8334.1 30.7 125.1 12.8 91.4 20 .3 94.9 864.4 17532.0 1,062.0 1100. 0 1,299.9 63:n. u 3.7-86 Amendment No. 16, (1/98)
- TABLE 3.7-23 (Cont'd) Sheet 4 Component Seismic Load Seismic and Component Original Excitation Design RSG Specified Location Calculated Maximum for Maximum Design Corn:bim::iu St.ed.l:u M {in-kips) 3323.6 3,423.3 12000.0 X and Y Generator Outlet Piping Steam M (in-kips) 2235.1 2,302.2 12000.0 Generator Outlet Piping Pump M (in-kips) 1794.0 1,847.8 12000.0 Inlet Piping Pump M (in-kips) 3628.1 3,736.9 12000.0 Inlet Piping Pump M (in-kips) 2955.1 3,043.8 12000.0
- Outlet Piping R.V. M (in-kips) 2636.1 2 I 715
- 2 12000.0 Inlet Piping Combined Steam M (in-kips) 1974.1 2,033.3 12000.0 Z and Y Generator Outlet Piping Steam M (in-kips) 1537.8 1,583.9 12000.0 Generator Outlet Piping Pump M (in-kips) 1193.5 1, 229 .3 12000 n Inlet Piping Pump M (in-kips) 2627.3 2,706.1 12000.0 Inlet Piping Pump M (in-kip!":) nm?. 1,'iS7.7 12000.0 Outlet Piping R.V. M ( 1952.8 2,011.4 12000.0 Inlet Piping
- 3.7-87 Amendment No. 16, (1/98)
- Component Seismic and Excitation Design Location Combined Reactor X and Y Vessel Outlet Support Reactor Vessel Inlet Support Steam Generator Lower Support Steam Generator
- Upper Support Pressur-izer Support Reactor Coolant Pump Vertical Support Reactor Coolant Pump Horiz. Support
- TABLE 3.7-23 (Cont'd) Sheet 5 Component Original Calculated Maximum H (kips) 5.9 v (kips) 214.7 H (kips) 261.3 v (kips) 122.7 Fy (kips} 1S7 .5 Fz (kips) 44.8 M.x (in-kips) 0.1 My (in-kips) 156.5 Mz (in-kips) 2700.2 Fx (kips) 123.4 Fx (kips) 22.2 Fy (kips) 24.6 Mz (in-kips) 5681. 8 Fy (kips) 1.1 Fa (kips) 5.6 Seismic Load RSG Specified Maximum for Design 6.1 22.0 221.l 47.3.0 269.1 1052.0 126.4 469.0 193.1 62.:l. O 46.1 54.0 0.1 21.0 161.2 455.0 2 781.2 24383.0 -* 133.3 140.0 22.9 82.5 2'.i.3 80.7 5,852.3 17207.4 1.1 ,a,_ 6 5.8 25.0 3.7-88 Amendment No. 16, (1/98)
- Cumponent Seismic and Excitation Design Location Combined Reactor z and Y Vessel Outlet Support Reactor Vessel Inlet Support Steam Generator Lower Support Steam Generator
- Upper Support Pressur-izer Support Reactor Coolant Pump Vertical Support Reactor Coolant Pump Horiz. Support
- TABLE 3.7-23 (Cont'd) Sheet 6 Component Original Calculated Maximum H (kips) 257.5 v (kips) 37.7 H (kips) 134.4 v (kips) 177 .3 Fy (kips) 172.9 Fz (kips) 81.6 Mx (in-kips) 5019.2 My (in-kips) 722.4 Mz (in-kips) 0.3 --Fz (kips) 64.6 Fx (kips) 24.6 Fy (kips) 22.2 Mx (in-kips) 5681.8 Fy. (kips) 0 7 Fa (kips) 18.9 Seismic Load RSG Specified Maximum for Design 265.2 663.0 38.8 392.0 138.4 304.0 182.6 692.0 17!L1 40S.O 84.0 397.0 5,169.8 24422.0 744.1 9772. 0 0.3 4132.0 69.8 240.0 25.3 80.6 22.9 82.9 5,852.3 17101. 5 0 7 9.2 19.5 25.0 3.7-89 Amendment No. 16, (1/98)
- THIS PAGE INTENTIALLY LEFT BLANK *
- 3.7-90 Amendment No. 16, {1/98) l,.o.) ...... I '° ...... '
- TABLE 3. 7-24 REACTOR INTERNALS AND CORE: LOAD AND STRESS CRITERIA PLANT CONDITION Upset Emergency Faulted LOAD COMBINATION Normal Operating + Operating Basis Normal Operating + Design Basis Earttquake Normal Operating + Safe Design Basis + Loss of Coolant Accident Pm = general primary membrane stress PB = primary bending stress S = maxinum allowable stress as defined by the ASME Code. m APPLICABLE COMPONENT Internals Core Internals Core Internals
- STRESS LIMITS Figure NG 3221.1 including notes p < s m -m P + PM< 1.5 S B -m Figure NG 3224.1 including notes p < 1.5 s m-m PB+P <2,25S m-m Appendix F Rules for Evaluating Faulted Conditions The maximum allowable stresses of components composed of materials not covered by the code with the exception of zirconium based alloys, shall be calc-Jlated as directed by Section III, ASME Boiler and Pressure Vessel Code for materials of similar properties. The rnaximun allowable stresses, Sm, of nium based alloys shall not exceed two-thirds of tie unirradiated minimum yield strength at temperature.
- * * *
- TABLE 3.7-25 NATURAL FREQUENCIES FOR VERTICAL SEISMIC ANALYSIS MATHEMATICAL MODEL (Historical Data Only) Mode No. 1 2 3 Frequency, cps Sub-Model I 21.60 67.75 124.59 Sub-Mode 1 II 72.98 404.09 3.7-92 Am. 3-7 /85
- * (Historical Data Only) Structural Component Core Support Barrel Lower Core Support Structure Grid Beam CEA Shrouds Single & Dual Upper Grid Beams Upper Guide Structure Flange *
- TABLE 3.7-26 SEISMIC STRESSES IN CRITICAL REACTOR INTERNALS COMPONENTS FOR THE DESIGN BASIS EARTHQUAKE Location Upper Section of Barrel Beam Flange End of Shroud Center of Beam Junction of Flange & Barrel Cylinder Stress Tension & Bending Bending Tension & Bending Bending Tension & Bending 3. 7-93 Design Load Stress 1,129 psi 5,278 psi 3,548 psi 2,762 psi 1,652 psi 2,823 psi
- Dynamic Analysis Stress 907 psi 686 psi 1, 771 psi 1,729 psi 222 psi 161 psi Am. 3-7/85
- *
- TABLE 3.7-27 SUMMARY OF FLOOR RESPONSE SPECTRA DATA Design Periods of Piping 1st Mode 2nd Mode Period Period Structure J)irection (Sec) (Sec) Reactor Hor. 0.76 0.29 Building Vert. 0.50 Reactor Hor. E-W 0.45 0.24 Auxiliary Hor. N-S 0.76 0.22 Building Vert. El. 82. 0 0.32 0.25 Vert. -Others 0. 57 0.33 70% of Minimum 2nd Mode Period (Sec) 0.20 o. 15 Design Accelerations Based on Design Periods (0.20 & 0.15 Resp.) 1. 5 x Design Structure Direction Max. Acc.* Max. Acc.** Acc. For (g) (g) OBE Reactor Hor. 68.5 0.26 0.39 0.39 Hor. 60.0 0.25 0.37 0.37 Building Hor. 44.0 0.27 0.40 0.40 Hor. 24.0 0.28 0.43 0.43 Hor. 18.0 0.32 0.49 0.49 Vert. All 0.13 0.20 0.20 *Reactor Hor. E-W 82.0 0.31 0.46 0.46 Hor. N-S 82.*o 0.27 0.46 Auxiliary Hor. E-W 62.0 0.23 0.35 0.35 Building Hor. N-S 62.0 0.15 0.35 Hor. E-W 43.0 0.20 0.31 0.31 Hor. N-S 43.0 0.13 0.31 Hor. E-W 19.5 0.21 0.35 Hor. N-S 19.5 0.23 0.35 0.35 Hor. E-W -0.5 0.23 0.52 Hor. N-S -0.5 0.34 0.52 0.52 Vert. 82.0 0.,0 01* 75 0.75 Vert. Others 0.40 0.60 0.60 *For Periods from 0 to Design Period of Piping **Selecting the Higher of E-W or N-S Horizontal Acceleration for Each Elevation 3.7-94 Design Period (Sec) 0.20 0. 20 0.15 0.15 0.15 0.15 Design Acc . For DBE 0. 78 0. 74 0.80 0.86 0.98 0.40 0.92 0.92 0. 70 0. 70 0.62 0.62 0. 70 0. 70 1. 04 1. 04 l. so 1. 20
' e TABLE 3.7-28 RESULTS OF SAMPLE PIPING SYSTEMS ANALYSIS Examples of Locating Restraints for a Preset Design Period Preset Sample Pipe Design Actual Problem Size Material Period Period (sec) No. 1 2" Stn. St. 0 .. 20 0.174 No. 2 12" Stn. St. 0.15 0.142 No. 3 8" Carb. St. 0.15 O. ll5 w Examples of Conservatism in the 1. 5 Participation Factor \D \JI Static analysis with Full Re.!r.onse Loads Sample Pipe No. of Period Acc. Problem Size Material Modes Per No. 4 8" Carb. St. 6 0.173 to 0. 098 1.0 No. 5 14" Carb. St. 5 0.155 to 0.091 1.0 No. 6 18" Carb. St. 6 0.195 to 0.085 1.0 *Factor of Conservatism
- Max. Stress by Direct Static Analysis Max. Stress by Full Response Analysis Max. S{resJ ps1 1639 292 4409 4403 2190 5296 Direct Static Anal!!!_s __ Factor of Max eonserv-Dir. Dir. athm*-X-Y 1. 5 8372 X-Y 5.11 Y-Z 1968 Y-Z 6.74 X-Y 1. 5 8775 X-Y 1. 99 x-z 6624 Y-Z 1. 50 X-Y l. 5 5880 X-Y 2.68 Y-Z 9107 Y-Z l. 72
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT S EARTHQUAKE OPERATING SPECTRUM DESIGN RESPONSE FIGURE 3.7-l
- -v ;..:J *I> ...... -->-:-0 0 ' .....I > *
- 10 0 .. a . e
- 6 ... . a *I .2 /4 *" PERIOD (SECS) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT DESIGN BASIS DESIGN RESPONSE
- *
- RELATIVE INTENSITY OF ACCELEROGRAM 0 INPUT PARAMETERS I. EXPECTED DURATION OF EARTHQUAKE MOTION ASSUMED TO BE 20 SECONDS IN EACH TRIAL. 2. EXPECTED MAXIMUM GROUND ACCELERATION RANGED FROM 0.04 g ro o. 10 g DEPENDING UPON THE COMPUTER RUN. 3. SHAPE OF THE POWER SPECTRAL DENSITY FUNCTION: TRIAL# I WAS BASED ON THE TAJIMI FUNCTION WITH JENNING'$ CONSTANTS. TRIAL# 2 WAS A SLIGHT MODIFICATION OF THE TAJIMI FUNCTION. 4. SHAPE OF THE INTENSITY FUNCTION WAS THE SAME IN ALL TRIALS. 2 7 TIME (SECONDS) 20 FLORIDA POWER 8: LIGHT COMPANY ST. LUCIE PLANT SYNTHETIC EARTHQUAKE INTENSITY FUNCTION FIGURE 3.7*3
- *
- u 0 0) > -_g Q) =r: NOTE: THIS SPECTRA IS FOR QBE.DOUBLE ACCELERATION VALUES FOR DBE. 0.3 0.5 0.7 Period (sec.) HOUSNER FLORIOA POWE:.R COMPANY ST. LUCIE PLA1'T UNIT 1 COMPARISON OF TIME HISTORY AND HOUSNER RESPONSE SPECTRA FIGU
- *
- MASS I ... 10 SH I E-LD STRUCTUIU* 11 ... ?O STH*L 21 ... 24 5TRUCTUI° DIAMHER F-Dl-.1 r FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR BUILDING HORIZONTAL MATHEMATICAL MODEL FIGURE 3.7*5
- T I '9 i * ! NI "'1" I I + I o' -' i 0 "'1 ! +* 0 I 0 '<::S + 0 -. 0 'i::t +
- 01 . ' I i '9' -* ! ti\ I "1 ! _j__ __
- T I I 2"1:()1 U'I + 7 3 o! .. ' 0 "' I 8 t I 4 01 -* ! 01 v I + 9 *J S 01 -* ! "-.9 01 0 "'1 I 10 "Q 9-t-*1.9 MASS I "' ; Sl-llELD <D ... I 0 5TE-EL. COt.JTAINMHJT II FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT RE ACTOR BUILDING VERTICAL MATHEMATICAL MODEL FIGURE 3.7*6
- oT--* c (\,a '9 I -* I c:Q I I -: I I +-I o: * -' NI I t '-? I ('a I * , £:L.-O*Co -d
- 2 3 4 FLORIDA POWER 8: LIGHT COMPANY ST. LUCIE PLANT REACTOR AUXILIARY BUILDING HORIZONTAL MATHEMATICAL MODEL FIGURE 3.7-7
- *
- s T o: 8 7 CD -t I o: C\J i 2 0 I '9 . "" IS 14 13 ('4 4 + I 0 -01 EL. -o'-co NI J_ -----...... ,, NOTc: HC-AVY l..HJE-Mf'MBE-12. FLORIDA POWER 8: LIGHT COMPANY ST. LUCIE PLANT REACTOR AUXILIARY BUILDING VERTICAL MATHEMATICAL MODEL FIGURE
- *
- 1271-0 I--...j N ---,... ---H. -I (-.. () . HORllOt-JTAL MODEL 2 4 FLORIDA POWER !lt LIGHT COMPANY ST. LUCIE PLANT FUEL HANDLING BUILDING MATHEMATICAL MODELS FIGURE 3.7*9
- *
- _J + ! '-'I + '-'I ' I ! . -1 H.* !>1-0 I c* -.n _I__ . .__ ..... _ ........ AL MODEL VE-RTICAL MODEL f.lf:AVY LIWf' R IGI 0 FLORIDA POWER 8: LIGHT COMPANY ST. LUCIE PLANT INTAKE STRUCTURE MATHEMATICAL MODELS FIGURE 3.7-10
- T-* I N"I +-2 I i I () i ' I -..(): -! I 5 3 4 * +-' 0 -* ' 9 I I i-*.n MOOcL
- H -* --** .. --j. 5 VcRTICAL MOD&L 4 HEAVY LINE: FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT DIESEL GENERATOR BUILDING MATHEMATICAL MODELS FIGURE 3.7* 11
- *
- MATHEMATICAL MODEL 11 12 13 14 15 16 17 18 19 20 2i 22 23 24 25 26 27 28 29 2 3 4 5 6 7 8 9 10 30 31 32 33 34 35 36 37 38 39 40 41 42 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UHIT 1 COMPARISON BETWEEN FRQA AND ST ARDYNE FIGURE 3. 7-llA 11.so II .1&0 ll. 20 11.00 3.80 3.60 3. llO 3.20 3.00 2.ao 2.60 2.llO 2.20 2.00 1. 8(J 1. 60 t. llO t. 20 t. 00 o.BO 0.60 0.1.w 0.20 * * * .--CJ 1..-.J z D ........ I-a: a: w _J w u u a: 1 I 0 .. 10 0.20 a.Jo 0.40 0.50 o.oo o.7o a.so a.go 1.oJ 1.10 1.20 1.30 NOTE: THIS SPECTRA IS FOR DESIGN EARTHQUAKE. DOlJllLE ACCELERATION FOR MAXIMUM HYPHOTHETICAL EARTHQUAKE. PERIODCSECONOJ I I I I 1.qo 1.so 1.50 t.7o 1.so 1.90 2.00 FLORIOA PO\!IEP 6 LIGHT COMPANY ST. LUCIE PLANT REACTOR BUILDING EL IS.0 HORIZONT Al FLOOR RESPONSE SPECTRf\ FIGURE 3.7*12
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- 11. 00 J.80 J.60 J. 40 J.20 ,--.. J.oo CJ 2-80 ..__, z: 2.60 0 ,__ 2.40 I-a: 2.20 a: w 2.00 -_J w t. 80 u u 1. 60 a: t. ttO t. 20 1. 00 0.80 0.60 O. ttO 0.20 I 0.10 0.20 0.30 0.40 a.so 0.50 0.10 J.Bo 0.90 i.oo 1.10 1.20 1.30 t.l!O 1.so t.&o 1.10 i.so 1.90 2.00 PER I DOC SECOND J FLORIDA POWER 6 LIGHT COMPANY NOTE: ST. LUCIE PLANT THIS SPECTRA IS FOR DESIGN EARTHQUAr.E. DOUILE ACCELERATION VALUES FOR MAXIMUM HYPHOTHETICAL EARTHQUA[f. REACTOR BUILDING EL 14.0 HORIZONTAL FLOOR RESPONSE SPECTRA 3.7-13
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- I i I o. 10 0.20 o. 30 0.40 0:50 a.Bo 0.10 a.BO 0.90 1.00 1. 10 I .20 J.30 1.so I.BO J.70 t.80 1.90 2.00 NOTE: THIS SPECTRA S FOR OESIGN EARTHQUAKE. OOlJBLE ACCELERATION VALUES FOR MAXIMUM HY*HOTHETICAL EARTHQU,KE. PERIODCSECONOJ FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANf REACTOR BlJILOING EL 44.0 HORIZONTAL FLOOR RESPONSE SPECTRA FIGURE l.7*14
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- 2.&o 2.40 2.20 2. 00 -lt"""""\ t..'.:> \.....J t. 00 _Jz 0 ........ 1. 60 _JI-a: a: 1.40 _Jw _J w 1.20 _Ju u 1. 00 ,a: Q.80 0.60 Q.40 0.20 I I I i I I i I I I i i I I I 1 0.10 a.2a a.Jo 0.40 a.so o.oo o.7o o.so a.go 1.00 1.10 i.20 1.Jo 1.1.,0 1.so 1.00 1.10 1.Ao 1.90 2.00 PERIODCSECONOJ FLORIOA POWER & LIGHT COMPANY NOTE: ST. LUCIE PLANT THIS SPECTRA I!. FOR EARTHQUAKE. DOUBLE ACCELERATION VALUES FOR MAXIMUM HYPHOTHE TICAl EARTHQUAKE. REACTOR BUILDING EL 60.0 HORIZONTAL FLOOR RESPONSE SPECTRA FIGURE l.7-15
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- J.oo 2.00 2.60 2.llO I ,....., 2. 20 -le:> '-' z 2.00 _Jo t. BO er: 1.60 _Jw _J 1. llo u 1.20 la: t. 00 o.ao Q.60 Q.llO 0.20 ----------D=O. s-i. i o. to 0.20 a.Jo o.llo a.so o.&o 0.10 a.so a.go 1.uo 1. 10 1.20 1.Jo 1.llo 1.so 1.&o 1.10 1:00 1.90 2.00 NOTE: THIS SPECTRA IS FOR DESIGN EARTHQUAKE. DOUBLE ACCELERATION VALUES FOR MHIMUM HYPHOTHE TICAL EARTHQUAKE. PERIODCSECONDJ FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR BUILDING EL 68.5 HORIZONTAL FLOOR RESPONSE SPECTR-. FIGURE 3.7-16
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- J.80 3.60 3.110 3.20 3.00 r-"\ 2.ao Q ......... 2.so z D 2.qo ......... I--2.20 a: a: 2.00 w _J 1. 80 w u 1. 60 u a: 1 .110 t. 20 1. 00 o.ao o.so o.qo 0.20 -------D=O. 5% I l I l I I I r--i I I I t o.as 0.10o.15 0.20 0.25 0.30 0.35 0.40 0.45 a.so 0.55 0.60 0.55 0.10 o.1s o.so o.s5 o.go 0.95 1.00 PER I DOC SEC CJ NO J FLORIOA POWE<> !Ir LIGHT COMPANY NOTE: ST. LUCIE PLANT TH!S SPECTRA IS FOR DESIGN DOJBLE ACCELERATION VJLUES FOR MAXIMUM TICAL EARTHQUAKE. REACTOR BUILDING VERTICAL FLOOR RESPONSE SPECTRA FIGURE l. 7-17
- 1.1.00 3.90 3. E-0 3.140 3.20 ,-,, 3.(!Q 0 2.S)Q '-' z 2.00 0 ,._ 2.1.10 f-a: 2.20 a:: l1.J 2.00 _j w 1. 5-10 Ll Ll 1. EiC a: 1 .1.10 1. 20 1. 00 Q.80 o.&c 0.1.10 0.20 NOTE: * * ----C=0.5"1. I I I o.bs o. 10o.1s 0.20 o.2s 0.10 o.1s a.so o.ss a.Loo.Ls 0.10 o.1s a.BJ o.gc: a.gs 1 .c1c THll SPECTRA 15 fOR EARTHQUA(E. DOUBLE ACCELERATION VALUES FOR MAXIMUM HYPHOTHE TICAL EARTHQUAKE. PERIOOCSELONOJ FLORIOA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR AUK. BLDG. EL-05 HORIZONTAL FLOOR RESPONSE SPECTRA FIGURE 3.7-18
- 2.1.10 2.20 2.00 1.&Q J;; ..__, z 1.so _Jo -. 1.1.10 -1a: a: w t. 20 _j-1 w u *l.OO :c. 80 *C.60 0.40 0.20 * * // ---------...____ O=O. 5% *-........ o.ns o. 10J.1s 0.20 o.zs o. 30 o.3s o.4c1 0.45 a.so o.ss o.so o.ss 0.10 o.1s a.Boo 85 o.so a.gs 1.00 PERIODCSECONDJ FLORIDA POWER & LIGHT COMPANY NOTE: ST. LUCIE PLANT THIS SPECTRA IS FOR OE SIGN EARTHQUAKE. DOUBLE ACCELERATION VALUES FOR HYPHOTHE TICAL EARTHQUAKE. REACTOR AUXILIARY BUILDING El 19.S HORIZONTAL FLOOR RESPONSE SPECTRA FIGURE 3.7-19
- 3.llO 3.20 3.00* 2.ao 2.so j.-. 2.40 z 2.20 I-2. 00 ..Ja: a: t.BO _Jw _J
- t .60 ..JW LJ LJ t .40 la: t. 20 t. 00 o.ao a.so o.1w 0.20 * -----------/ //
- D=O. S'Y.. T I 1--r-I ----,--1 T .l o.os o. 10 o. is 0.20 o.2s 0.30 0.35 0.40 o.4s a.so o.ss a.so o.6S 0.10 o.7s 0.80 o.85 o.go o.95 1 .oo PERIODCSECONOJ FLORIDA POWER & LIGHT COMIPANY NOTE: ST. LUCIE PLANT THIS SPECTRA IS FOR OESlvN EARTHQU .. KE. DOUBLE ACCELERATION VALUES FOR MAXIMUM HYPHOTHE TICAL EARTHQU .. KE. REACTOR AU)(. BLOG. El 43.0 HORIZONTAL FLOOR RESPONSE SPECTRA FIGURE 3.7*20
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- lj. 40 ll. 20 ,ll.OQ J.ao .3.60 3.40 /""'"\ 3.20 c.!.> <.....) J.oo z 0 2.ao ....... I-2.60 a: 2.40 a: w 2.20 _J w 2.00 u 1.ao u a: 1.so 1. 40 1. 20 1. 00 o.ao o.so 0.1.10 ll=O. St. 0.20 J __/' I I I I I I I I I I 1 I I I I o.os o. 10o.1s 0.20 o.2s o.Jo 0.35 o.40.0.45 o.so o.s5 0.60 o.Gs o. 10 0.15 o.ao c.as o.so o.ss i.oo PER I DOC SECOND J FLORIDA POWER & LIGHT COMPANY NOTE: ST. LUCIE PLANT THIS SPECTRA IS FOR OES!GI' EARTHQUAKE. DOUSLE ACCELERATION VALUES FOR MAXIMUM HYPHOTHETICAL EARTHQUAKE. REACTOR AUX. BLDG. EL 62.0 HORIZONTAL FLOOR RESPONSE SPECTR,l. FIGURE 3.7-21
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- s.so s.zs 5.00 q.75 1.1.so IL25 I.I. 00 3.75 3.50 J.25 3.00 2.15 2.so .2.2s 2.00 l. 7S 1 .. 50 1. 25 1. 00 o.1s o.so o.2s ,....... c.!> <....J z 0 ,.._ I-a: a: w _J w (_) (_) a: o.os o. 10 o. is c.20 a.2s 0.10 a.1s 0.40 o.qs o.so o.ss o.&o o.ss 0.10o.1s a.so a.as 0.90 a.gs i.oo PERIUOCSECONOJ FLORIDA II LIGHT COMPANY NOTE: ST. LUCIE PLANT THIS SPECTRA IS FOR DESIGN EARTHQUAKE. DOUBLE ACCELERATION VALUES FOR M'XIMUM HYPHOTHE TICAL EARTHQUAKE. REACTOR AUX. BLDG. EL 82.0 HORIZONTAL FLOOR RESPONSE SPECTRA FIGURE J.7-22 4.20 4.oo 3.ao 3.60 3.40 3,20 3.00 2.ao 2-60 2. 40 2.20 2.00 1. 80 1° 60 t. 40 1. 20 t. 00 o.ao a.so 0.40 0.20 * *
- l.'.:> '-J z 0 ........ I-a: a: w _J w u u a: ---. ____ o=o. sf., r r T l I I I I I I I I . o.os o. 10o.15 0.20 0.25 0.30 0.35 0.40 o.45 o.so o.55 0.50 o.55 c.7o o.75 o.ao o.85 0.90 o.95 1.00 PERIODCSECONDJ NOTE: FLORIDA POWER & LIGHT COMPANY THIS SPECTRA IS FOR OE SIGN EARTHQUAKE. DOUBLE ACCELERATION VALUES FOR MAXIMUM EARTHQUAKE. ST. LUCIE PLANT TOR AUX. El 62.0, 43.0, 19.5, -0.5, -10.0 VERTICAL FLOOR RESPONSE SPECTRA FIGURE 3.7*2l 9 so 9.(lQ a.so 9.QQ 1.so 1.00 6.50 c.oo s.so s.oo 4.50 4.00 3.so 3.(*0 2.su 2 .(JO 1. so. 1 . 00 O.S(! * *
- r-'\ c..j z D ,._ f-er a: w -' L:..J LJ u CI "'--------------------.---D"O. 5 f., I l I HTH -, I I O.OS Q.j(l O.iS Q.2(1 D.2S LJ.3C' D.1'.i O.c!C: O.c!S C.'::iU 0.55 CJ.LO o.r:,r_; D.70 0.7S O.'W O.Si'..i Q.3(' 0.9S t.(lC PER I OCJC SECOND J NOTE: THIS SPECTRA IS FOR DESIGN EARTHQUAKE. OOLBLE ACCELERATION VALUES FOR MAXIMUM HYPHOTHETICA" EARTHQUAKE. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANI' REACTOR AUX. BLDG. El 82.0 VERTICAL FLOOR RESPONSE SPECTRA FIGURE 3.7-24
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- A23A SB1 STEAM GENERATOR 1A1 SNA2 PUMP 1A2 EN LARG ED VIEW f-OF 11 REACTOR INTERNALS -COMPONENT MASS POINT NAME NAME V4, Rl1, Rl2 REACTOR V1, V3 SG5A,SG58 STEAM GENERATORS SG9A,SG98 SG 10A, SG 108 M43, M52 REACTOR M61, M66 COOLANT PUMPS M42, M51 M60, M65 SG9B R12 A23 PUMP 1A1 MASS POINT DEGREES OF FREEDOM X,Y,Z x,z X,Y,Z X,Z X, Y,Z x,z y t SNA1 M66 0 MASS POINT §---SUPPORT POINT E-MEMBER RE LEASES STEAM GENERATOR 181 COMPONENT SUPPORT NAME NAME RA1 REACTOR RA2, RA3 SG1A,SG18 STEAM SA1, S81 GENERATORS A23, A23A A39, A39A (VERTICAL REACTOR SUPPORTS 4 COOLANT PLACES ON EACH PUMPS PUMP) SNA1, A2, 81, 82 SNB1 PUMP 181 SUPPORT RESTRAINT DIRECTIONS Fy, Fz Fy, FH* Fy, Fz, My Fx Fz Fy Fx, Fz *FH RESTRAINTS ARE PERPENDICULAR TO THE REACTOR VESSEL RADIAL DIRECTION IN THE HORIZONTAL PLANE. FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 FOR HISTORICAL INFORMATION ONl Y TYPICAL REACTOR COOLANT SYSTEM SEISMIC ANALYSIS MODEL Am. 3-7/85 Figure 3.7-25
- MASS POINT RJ3, 15, 18 1, RJl, RJ4, RJ5 2, 4, 9, 10 16, 17, 19, 20 * /', * ' " CORE SHROUD *
- FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 DEGREE OF FREEDOM X, Y,Z x,z FLANGE bi. MASS POINT 0 STRUCTURAL JOINT SUPPORT POINT E-MEMBER RELEASES f RJ4 REACTOR COOLANT SYSTEM "-......._SEISMIC ANALYSIS MODEL ....... ......... ....... ......... ,.... __ .._ ..... ,_...,...... ................ ooll!Q1 CORE SUPPORT BARREL 2 RJ3 le 1, 20 I/ REACTOR VESSEL I I I I 9 I I I THERMAL I SHIELD I I y 4 (Removed) I 18 I 10 I I I x 17 I RJ1 16 I z I I I 15 RJ5 Am. 3-7/85 FOR HISTORICAL INFORMATION ONLY Figure (RV14) REACTOR AND INTERNALS SEISMIC ANALYSIS MODEL 3.7*26 y I z PRESSURIZER
- MASS DEGREES OF POINT FREEDOM 2-8, Hl, H3 X, Y, Z x 0 STRUCTURALJOINT MASS POINT }-SUPPORT POINT E-MEMBER RELEASES HOT LEG SUPPORT RESTRAINT NAME DIRECTION Hl, H2, H3 y "Figure FLORIDA POWER & LIGHT CO. St. Lucie Plant Surge Line Seismic Analysis Model 3. 7-27
Florida Power & Light CompanySt. Lucie Plant Unit 1Original PressurizerSeismic Analysis ModelFigure 3.7-28aAmendment No. 21 (12/05) Florida Power & Light CompanySt. Lucie Plant Unit 1Replacement PressurizerSeismic Analysis ModelFigure 3.7-28bAmendment No. 21 (12/05)
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- o MASS NODE 0 MASSLESS NODE I L ' ' " ' ft** I ..._ A -6141 IS _/' \, 6411 I I Ill** I; II J. lo 141 n 4l 0 "10 14 41 I IS ..... 6**
- 40 6SO 1 * *-10 I I .. emoved) -W I 611 7 II .. -, I I .e.ae (R .. , _ IT
- IT I I 11 6H II I .. 6U I I .. 614 4 II I 6U 14 .. °'IZ . ..
- I ' 'I lj UC 111 2\ 'ii ii 1" 'i1 Ii a ..l ... I I ' FOR HISTORICAL INFORMATION Am. 3-7 /85 FLORIDA POWER & LIGHT CO. St. Lucie Plant Representative Node Locations Horizontal Mathematical Model Figure 3. 7-29
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- MASS NODE Q MASSLESS NODE -=== RIGID CONNECTING LINK 'T' HINGED CONNECTION 47 LL{fU R.V. FLANGE 48 13 46 UPPER --GUIDE 12 45 44 43 STRUCTURE CORE ..... --SUPPORT 42 33 CORE SHROUD LOWER SUPPORT STRUCTURE FUEL ASSY'S 32 31 30 29 28 27 26 25 FOR HISTORICAL INFORMATION FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 MATHEMATICAL MODEL HORIZONTAL SEISMIC ANALYSIS BARREL 17 SHIELD 16 (Removed) 15 14 Am. 3-7/85 Figure 3.7-30
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- 6 MASS NODE Q MASSLESS NODE c=::::::::J RIGID CONNECTING LINK UPPER GUIDE STRUCTURE MODEL I LOWER SUPPORT STRUCTURE FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 37 33 28 24 20 18 14 9 39 52 FUEL ALIGNMENT PLATE FUEL MASS AND FUEL ALIGNMENT PLATE FOR HISTORICAL INFORMATION ONLY MATHEMATICAL MODEL VERTICAL SEISMIC ANALYSIS CORE SUPPORT BARREL 30 27 23 THERMAL 17 SHIELD (Removed) 13 8 Am.3-7/85 Figure 3. 7-31
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- FLORIDA Core Support Barrel Figure POWER & LIGHT co. Upper Flange st. Lucie Plant Finite Element Model 3. 7-32
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- MODE 1 FREQUENCY 3.065 CPS FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 I ¢ I 9 I 9 I ¢ I 9 I 9 I ¢ FOR HISTORICAL INFORMATION ONLY LATERAL SEISMIC MODEL -MODE 1, 3.065 CPS ----------Am. 3-7/85 Figure 3.7-33
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- MODE 2 FREQUENCY 5.118 CPS I ? I ? I 9 I ? I 9 I a_ _____ _ FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 I 9 I Q I I I I I I 9 I I 9 I I I I 9 9 I I I I 9 9 I I ?--t=-=-==-=\. I 0 I 9 I I ? I I ¢ L --t:::..=-=-=-=-=-=-=-=-=-=-=---==----.......... "'"" FOR HISTORICAL INFORMATION ONLY LATERAL SEISMIC MODEL -MODE 2, 6.118 CPS Am. 3-7 /85 Figure 3.7-34
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- MODE 3 FREQUENCY 8.166 CPS er------I ? I ? I ? I -----I ? 9 I ¢ I I I -----------I 9 I 9 I 9 ¢ I 9 I ¢ I 9 L-----------. FOR HISTORICAL INFORMATION ONLY FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 LATERAL SEISMIC MODEL -MODE 3, 8.166 CPS ------------Am. 3-7/85 Figure
- 0 Lt"\ 0 ("f"\ N ...... . a:) . d 0 0 6 'NOI!VH313JJV c::> ...... . . 0 0 0 0 ...... 0 0 ...... d V'l 0.. u u z L..LJ :::::> 0 L..LJ 0::: u...
- POWER & LIGHT co. Vertical Response Spectrum forlhe Operating 3. 7-36 st. Lucie Plant Basis Earthquake (1% DampingJ
- -I \ ':! '7\ * /? < * \ I I I I >-z < a.. '; u c:: "' co I-...... I-.-1 V> : i3 0... ::i c:J (:; o6 "?'. w Q IX IX :l 0: => W,_J !:? :s: L.L. 0 . Q a.. ..., ::> <I:(/) :5, Cl Ci:: 0 ..J u...
- * * --**--**** ... . ... ..,. ;, _;_ __ ... ., . . ll I / I
- BRIDGE GIRDER EARTHQUAKE RESTRAINT * ,, , 11 'i 111 11 11 BRIDGE GIRDER * --..... _::",::-...:-,::---_.. + + + + + + + + + + + + TROLLEY STRUCTURE EARTHQUAKE RESTRAINT FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING CRANE BRIDGE EART;QUAKE RESTRAINT FIGURE 3.7-39
Florida Power & Light CompanySt. Lucie Plant Unit 1Seismic Monitor ArrangementFigure 3.7-40Amendment No. 21 (12/05)
!!!!
TABLE 3.8-12COATING - INSIDE CONTAINMENTDATA INTENTIONALLY DELETED
TABLE 3.8-13RESTRAINT ANALYSIS - LINES ANALYZED(Historical Information)Fluid In PipeSystem & Line NumberFluid in PipeSystem & Line NumberFluid in PipeSystem & Line NumberSteamMain Steam:WaterReactor Coolant:WaterReactor Coolant:"MS-1,3"RC-114,123"RC-834"MS-2,4"RC-112,115,121,124"RC-885,836"MS-10,11,13"RC-108Chemical & VolumeWaterControl System:"MS-28,29"RC-147,151 to 154,162"CH-106,107,110,111"MS-50,51"RC-102,103"CH-104,109,112,125"MS-52,53"RC-109,141"CH-146,147,148,149"MS-63 to 78"RC-113,116,122,125"CH-126,127,135,136,137 "M-79,80"RC-142,145,148.149WaterFeedwater:"RC-150"CH-113,117,121"BF-28,30"RC-105,117,118.119"CH-115,118,123,134"BF-42,43"RC-120,130,137,138"CH-101,102"BF-32"RC-139,140"CH-100,103,128"BF-33,35"RC-101"CH-142,143"Deleted"RC-156,157"CH-300,301,312"BF-13,18,55,56"RC-104,107"CH-309,310"BF-14,19"RC-822"CH-304,307"BF-29,34,31,36"RC-824,827,828,829"CH-305,308"BF-51,52"RC-825,8263.8-126Amendment No. 17 (10/99) TABLE 3.8-13 (Cont'd)Fluid in PipeSystem & Line NumberFluid in PipeSystem & Line NumberContainment SprayWater& Safety Injection:WaterBlowdown System"CS-8,9,10,11"I-B-42,43,61,62"CS-14,15,18,19"B-52,54"SI-406,407,412,414"SI-408,410,814"SI-472,474"SI-430"SI-415,416"CS-58,59"CS-36,37"SI-222,224"CS-38,39"SI-110,111,112,113"SI-137,138,139,140"SI-457,458,459,460"SI-101,102,103,100, 148,149,150,151SI 1" lines at Safety In- jection Tank3.8-127Amendment No. 17 (10/99)
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- Refer to drawing 8770-G-793 Sheet 1 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONT A1NMENT VESSEL SH.1 FIGURE 3.8-1 Amendment No. 15 (1/97)
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- Refer to drawing 8770-G-793 Sheet 4 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONTAINMENT VESSEL SH.4 FIGURE 3.8-1a Amendment No. 15 (1/97)
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- I ---r. -------------+------r-------r------;--------____ _ _ ...;....;,__ ............ +-....., ...... __ _;.._;-_....., _______ _,,_---------j;---------1--219.0 0. 280 1 i---------+----------+------* ... 1---i---------,,>---------'"---.------+-------+--------T*7 0.276 ...._ ___ ;__ ___ +---------+------.,1--...-------------<--------208.0 1 1 188.0 ,' -+-----** -_;__:_ __ ....;_-+_;_;.....;_,___..;._-+--.;....+.-...... _+-------il-----IJ-0. 256 -------+----. ----..o---....;..+--------+-----11r.__ -+-------. I I --*-------+--...._ _ _... __ ..,_ ______ 168.0 ! I I ' ! I !-.. o.235 E-* = t---*--I 148.0 I !--I-----------+::-=-:-..::=-.=--: 1--.;___----+------7--f-0.213 -+------1--. ------_.., __ ..,_;_;.. __ ______ -_ _.-. .:_-_ -------. =-:_ :J _ *-----= :-_ I ' ' I I ' ' _______ ..._.._._+---.----:--0. 193 ,, +--------+---------.-.----------***-*-----o. --------T I ------------+..,_;.._;.....;....;._ _ _;.....;....;... _ _;.. ___ 1-#---.----------ll *---------------I 1 I --------------i-.:.-.;.... __ ,_;.._,_,.__... ____ ..--t-.;....---1--.;.....+..;.._------<-1 ---------------,.-___ ;._ _____ _, __ ...... _.__._...;...J....;4 ________ ,.....+ ____ J-#*-* I ______ _. _______ ---.. ---... T 0.159 ------*-1---------+----#-,---...-----.. __,_...;._ _____ _. __ *----*-------------------IP-. 0,136 , . -r---------------------------t------..-,4-'-..._ __ ....,.._.;._ __ I ' I I i ,
- I -.....;'l....,...;._..;'--'-'-...;._-+-.;.....,_;..---...;._-J....;4_._, I I _____ _,._,._.._ __ ---......-+-;.....;.:-+:..;'...;..'...;._' -+-.;....;-'--T-'.l...j ..... 0.115
- I ' I ' 1 ' I I I l l I l I ! j I I l j I I I T i---,---;----7' ' I I 0.1 ' I 0.2 I I I ' 0.3 I I I ' I I ' I I I I I I : ' ! I ; ! j ! J IT! ' ! i I I j j I ! I l j I ; ; ! ! I l I I ! I ! -n:; ...... , _.1-+' -i,-+-'i-:'-.j...;:....:.'...;..' 1 1 J 1J...L.L1 , ** , 1 1 , ' , , 1 -t-.... , -;,-H-,-!1-1 ...._-+ ....... '-1 +, -i,-;-..,1 ....... ._..11-+-.._, ...:.,_;...., -, ..;....;.., MAXIMUM ACC El.. ERA T IOH IH *c;* UHIT5 -,jr--, ..,,-: "!'1 ..,1r--; ..,:-1r+..,1-+-+-;: .... :-;.....;..::-i:---11 ..... ;+.;...: ...;.1--'1 -1....;.1 0 E SI CH EARTH QUA K E (0 .0 5 C) I I FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONTAINMENT VESSEL OBE SEISMIC LOADING
- SHEET 1 FIGURE 3.8-2 Amendment No. 15 (1/97)
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- i i -. -'.-0 1 _;.l _,l..;..I .... ,_...., .... , ....... , -,'-+-i1-+--'l-Hi-+-+l-<i-+-l-+-,_,.l-+-o.1 0.2 rro.3 -I O,.C -++0,5 I < 0.6 _-,....., __ -"'-,--'1_.__..__1 ""1-+-1f-'-1-+-+1-.;-,11,_,.1_,_,11-.,.....1f-'-1-1 ""1-I I !'TT"'I l 1--rrt I I I 1-r-r"I I I * * ' 1 I I I I I I I I I I I I 1 I I I 1 I I 1 I I I I I I I I I 1 I ' ' I I I ! I I i ' I I I I ! ! I I I I I I I I I I l I I I I I I I i ' I I I ' I I I l I I ! l l I I I I f I I i I I ! I _ ...... 1, ... 1_,H-1 +1 +MA XI MUM AC CELE RAT I OH I H 'G' UH ITS -<'-;..! .._1 -'-;..' -+-<'---;...' _.,_..11-11-...._..11-:,-.i...1, ..... :-'.-+-+1 ...,,_,_ . .._. 11,,..1'-......,. 1' ,,..1-+-*, -i-'-1 *1-11-+-, HY PO THE TIC AL EAR TH OU AK E (0. 1 OG) -:1-.,.....1---,. -+-.-.1 .... 1 '"'1-----t FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONTAINMENT VESSEL OBE SEISMIC LOADING -SHEET 2 FIGURE 3.8-3 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-493 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING CONTAINMENT VESSEL SUPPORT PLAN & SECTIONS M & R FIGURE 3.8-3a Amendment No. 15 (1/97)
- BOUNDARY COMDITIONS ( Q = 0 AT END N0= _f!. =39.6 (840.95) = 16652 LS/IM 2 2 M0 = 0 AT START l W = 0 u0 = o 80 0
- 4" THICK DOWH CHEM CO. ETHAFOAM 220 ASSUMING 103 COMPRESSION E = 35 fSI = 3.5 PSI STRESS .10 STRAIN k 35 = 8.75 PSI/IN 4 SPRING LOAD PER UNIT WIDTH OF CIRCUM, F = 3.75 i8.75) (l) kh F 32.8 L.B/IN ELEV 23.0' ELEV 19.0'
- END 32.8 lb/in "'I 52.5 ..;I 43.8 48.1 ::;1 56.4 ..; -64.S :t 64.9 57.0 N0 R 840.95" \. 1.903" CYL p '"'39 6 PSIG
- 111 ll..RL © © ss.s0 vi] 870 830 810 80.25°
- FLORIDA POWER & LIGHT COMPANY St. Lucie Plant CONTAINMENT VESSEL EMBEDMENT STRESS MODEL FIGURE 3.8-4
- SHELL TEMPERATURES UNDER ACCIDENT CONDITIONS (l20°F AMBIENT BEFORE ACCIDENT) CASE 1 4 7 TIME 5 MIN 20MIN 35 MIN 205°F---241°F --247°F--205 241 247 205 ---241 247 156 ---183 --197 --125 151 ---163 --120 129 135 120 --121 --125 --120 ---120 122 120 ---120 --121 PART 10 SO MIN © I :::1 I 249 "' 249 ..;H.L. 0 90° .,; © 205 88.5° .,; 171 87° .... M ,.: 141 85° ,.: 128 83° "' .. ,.: 123 122 "' I \"'.'\ 810 , .. CD so.1s0
- FLORIDA POWER & LIGHT COMPANY St. Lucie Plant EMBEDMENT LOCA TEMPERATURE GRADIENT FIGURE 3.8*5
- 9\ * : 1*n-r-:;-: i -, * ,--T I i -----*m* 'j -++* -!-** t*** JI "_ j: __ ! I: i I l[ *** l \. f *. l *,!.: t! 1' * * *\1 * ; * * ** j , r ' . . ' I I I I ' I I ' , l;i.:1.j; .. l,:1*-n*;.j:i, _,-_ , *-* . -:...:..._[ i , , *r1 i--: \. , ii, I . . : =--! _:" . /I ! : i : ! t I : : 1 l -*---* /-. . i N -r -I -: I .. ' +1 ---, . . --!---+--+-+-0.4 . I . I ; l I I\ ; i ;ii,1 111111 ' I-'-'--+---+-* I I I : I ! I I ' t : 1 I ! ' t -; I : '0 I I I \ ' I I -o.3 J *' . . !1 : : I I I I I I I *r i . ,_ I ' ' I ' I I _,. ......... ,I . : j : ; ' t---+-+-w -,, ' L-1-i-. -I-+ .. l -'-+-+-8 I / i . I ! i I
- I !1' \ > , I ! w-1 I ' 'I, i : ' L i0.2 J1 / i ; . ",, \ Q(1 ; . T-iHI . 1
- 1 l!:S,-l,ii i-L+---. "'---+1 i ---r* -1: +-+-, ! ! ' ' . I I I , ' "-. . ; l _j ! I ' I . \I i **-+-1-1-+-' I II ! 0.1. ; ,. "'--11 1"\. I I : I 't' ' I t-i--t-t: -l I t j __ . . I I r-0ii r-rl I . i I I . . . * .*. ,. , . ti 1: l. Jn FLORIDA POWER I!. LIGHT COMPANY St. Lucie Plant AIR LOCK SEISMIC RESPONSE SPECTRA FIGURE 3.8-6
- *
- CASE I +---r--CASE II + CASE III ... . . DIRECTION Of SEISMIC ACCELERATION FLORIDA POWER !Ir LIGHT COMPANY ST. LUCIE PLANT AIRLOCK SEISMIC ANALYSIS MODEL FIGURE 3.8-7
... FLORIDA POWER & LIGHT COMPANY St. Lucie Plant ELECTRICAL PENETRATIONS FIGURE 3.8-8 Ll!G!MD * * . c'J'*" ,..,,, -, ...... . :1" d' .... 1 .. d'"' :\ .... :\ .. ,, \...... t.\... ' d' ,1 .-. .,.t: :t. .,. .. , ..... ..,t.\\t.'¥ t. ... ,,... :\.. \ .... T" :\ ,,, '* .. t; .>).t.+-...... .. .... *sEIMC HSTIAINT 0... -...,,,,.._,_,,'?# -.. ".,. PIPE IUPTUIE llSTIAINT " 9'>c. ,, t. .;fl' \t."' _.JI,.'\ ,,, _,, o-* .,,,P ...,.. .. ,,,, 4'-t,. p"*i;;' '*bo-t l((J *o..-.,,.,, . ./ .. "" d c. .. .. ..,\ .... .-. .. l .. d+ .... t,\ ..... P" .,,.,, _,,,t. 1 .. 'o, .... 11 "< '(J . MOT TO SCALI FLORIDA POWER & LIGHT COMPANY ST. LUOE PUNT UHIT I LETDOWN PENETRATION 76 ADJACENT RESTRAINTS FIGURE 3.8-8A
- *
- Refer to drawing 8770-G-814 Sheet 9 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 MISCELLANEOUS STRUCTURAL STEEL ALL AREAS SH.9 FIGURE 3.8-Saa Amendment No. 15 (1/97)
/ ,...,,..,. .. ,.t.,, Ll!GEMD
- SEIOllC IHllAIMT *' ................. .
- COllllNED ..... C, IUPJUIE IES11AIMT .. ..... :i' t,'ii' """' ,..-">>t.'"'-.. ,_t.,, 4':- ,,.,t. ..... t,'i>' 4':- "I" / 9t." ... ..... cP+ / .\ro 4 ., \"' r:i" '\ 1."'-t,S * 'f.t,c,'\"'-.old!!\')' "'o "'""'ii t"\)ti NOT TO SCALE FLORIDA POWER & LIGHT COMPAN l ST. LUCIE PUNT UNIT 1 MAIN STEAM PENETRATION 1 ADJACENT RESTRAINTS FIGURE 3.S-8B
- *
- Refer to drawing 8770-G-213 Sheet 4 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR CONTAINMENT BUILDING PIPING PENETRATIONS FIGURE 3.8-9a Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-793 Sheet 2 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONTAINMENT VESSEL SH.2 FIGURE 3.8-12 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-793 Sheet 3 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 CONTAINMENT VESSEL SH.3 FIGURE 3.8-13 Amendment No. 15 (1/97)
- *
- FLORIDA POWER &: LIGHT COMPANY ST. LUCIE PLANT SHI ELD BUILDING COMPUTING MODEL LAYOUT FIGURE 3.8-14
- w.,..i *
- Wi w* \. FLORIDA POWER 8: LIGHT COMPANY ST. LUCIE PLANT SHIELD BUILDING DEAD LOAD CONDITION FIGURE 3.8*15
- *
- FLORIDA POWER Br LIGHT COMPANY ST. LUCIE PLANT SHIELD BUILDING LIVE LOAD CONDITION FIGURE 3.8-16
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT SHIELD BUILDING TORNADO INTERNAL PRESSURE LOAD CONDITION FIGURE 3.8*17
- 1.oi \.o Ct> ...
- 0.4i lo..,1')1U<!.\ to..ioq10\>!. 4 I) \.o.._t> 0.4-q, /)....
- FLORIDA POWER & LIGHT COMPANY St. Lucie Plant SHIELD BUILDING TORNADO AND HURRICANE WIND LOAD CONDITION FIGURE 3.8-18
- * * "f, ,, "to "(g 't 1\ *1o,. \. , , \' " / ca...i;'=' ' . \J..,,, '2. * \.)..:)\, FLORIDA POWER St LIGHT COMPANY ST. LUCIE PLANT SHIELD BUILDING TEMPERATURE LOAD CONDITION FIGURE 3.8-19
- *
- Lo"-0\t.J°" <:io1l .._uo \*W'-\\: .. \Ot:\.'?.\1 I I W"-"H:1'_ ,0 11::\. \o \:\.. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT SHIELD BUILDING SOIL AND GROUND WATER LOAD CONDITIONS FIGURE 3.8*20
- (e) * * / / . c;,.,,.. 1.. ,, 't <de°a *Oi g** Cl\CI\ I 6C'.e.O ,, $it.. \ FLORIDA POWER 6 LIGHT COMPANY ST. LUCIE PLANT SHIELD BUILDING WATER AND EARTH PRESSURE PROFILE FIGURE 3.8*21
- J' J '! 0 ci - ,,, '-' 6 *
- *
- Refer to drawing 8770-G-488 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING BASE SLAB-PLAN-MAS FIGURE 3.8-23 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-490 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING BASE SLAB-PLAN SH. N0.1 FIGURE 3.8-24 Amendment No. 15 (1197)
- *
- Refer to drawing 8770-G-501 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING CYLINDER WALL-PLAN & SECT-MAS FIGURE 3.8-25 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-503 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUIL:.DING CYLINDER DEV. MAS . FIGURE 3.8-26 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-511 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING DOME-PLAN & MAS FIGURE 3.8-27 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-512 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING DOME-REINF SH 1 FIGURE 3.8-28 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-513 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING DOME-REINF. SH 2 FIGURE 3.8-29 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-518 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING INTERNAL CONC. PLANS & SECTS.-MAS.-SH. N0.1 FIGURE 3.8-30 Amendment No. 15 (1/97)
- *
- Ref er to drawing 8770-G-519 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 REACTOR BUILDING INTERNAL CONC. PLANS & SECTS.-MAS.-SH. N0.2 FIGURE 3.8-31 Amendment No. 15 (1/97)
- *
- cL'2oo.o' 'E:l*IO.O et-r..o LOAD 1so*/c:.J:: u *. i 140 ,419 1'20 100 ea (-) "° 2.0& 40 .'>54 _1Q_ 1'.&/ I .014 9.o ...____. ll(U . IS.fl . ___ J o ff MEIC:IOIOIJ..l.L N\OM!:NT MOOP MOMENT MUIOIOt.1'-L l"Oil?l:i HOOP IJOIZIMI. DEl'l.EC.TIOll llUIOIOIJ.i.L SMElo.ll (I(* Fl/Lfo) (IC*F-1/LP.) ("/Lp.) 0:/Lfo) (11J) (t:/LF-) IJOT£:: OIJ TEIJ5101.J (*) llJOlc ..... n.c:. n..is101J (.*) llJOICA."Tl':> COl<'\?lli!'JSIOIJ FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RESULTS OF STRUCTURAL ANALYSIS DEAD LOAD FIGURE 3.8-32
- b.l 'Zoo.o' Ji& \<io 149_ 110 100 e>o 5LQ_ AQ_ E:L **O o* 11 ..'.!Q._ E;L-f> o' .e. _,40 1.AE.IZIOIOIJ.1.l MOIAE.IH ( -FT/LF-) *°"
- LIVE LOAD 30*/sF .IS .01 (*) i.n \JDOP 1.4:>MEl.JT f.OfCE. MOOP F-Oi!C'-IJOl21Al>.L OEF-li:.CTIOIJ (t::*F-T/LF) ( t::/1.F) (1::/LF) ( 11.1.\ UOMl:tJ15 SHOWIJ oi.J Tli:l..l510tJ 51 OE:. (*) llJDICb.Tl:.5 n.1.1s101J (-) llJDIC ID 1...1 .015 SHEA&: { I::/ LF-)
- FLORIDA POWER 11. LIGHT COMPANY St. Lucie Plant RESULTS OF SHIELD BUILDING STRUCTURAL ANALYSIS LIVE LOAD FIGURE 3.8-33
- *
- EL iioo.o' *L*10.o' 14o 110 100 eo jQ_ 1Q__ IkJTERf.JA.L PRESSLlRE. PSl 4-."6 I .V?, ].:"\."\. o::-F-T/I."') .01 lo'\ (+) (+) ?.,;"\ . clC.."\. . 'tt.C.. 1.&o 11,.oi;i___ "1. ."\"!.. 'U o liOO? MOf.M.IJl ME.elDtOlll Foi!CI: llORMAL DcFi.tCTl.JM McJ2\Dl()U.l ... e (K*FT/1..P:. (1::/LF) (t:./l.F-) (l>J.) (K/l..F-) UOTlih MOMlitJTS $MOWIJ Ol>i "\'li.1.J51oM g11';)&. (*) 1'1i:M510N (-) IMOIC.6.TE.':> CoMPQ.&ass laN FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RE SUL TS OF SHIELD BUILDING STRUCTURAL ANALYSIS TORNADO INTERNAL PRESSURE LOAD FIGURE 3.8-34
- *
- t. o ( 3(00 e
- 0° ll. 'l.oo.o' l&o 100 140 (*)' 14,& tl_Q_ IOC *) 6o a.o *a.I'> 40 ... 10.0' '20 '*"' a*!>.O' LI _-1Q11L-1 19& L__J -iw"."ll I . -_Vo '"""< I 1.iili?.IDINAJ. MOMENT llOOP ll\011\EllT F-oa::e MOOP P.otiCE .aotWL Dt:F-LECTIOM MUIDIOM.a.1.. SMc*I? <"*FT/LF) (k:*l"T/l.Jll) (t:/f&T) o:./,.T) (IM) UOTE: 01ll r.N510M (*l IMO IC>. 'Tf:t.i!llOt.I (-) FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RE5Ul TS OF SHIELD BUILDING STRUCTURAL ANALYSIS TORNADO WIND LOAD -0° FIGURE 3.8-35
- 1eo IC.O 140 1'10 JOO &o coo 40 !L .. 10.0 ,,__'!Q_ .JL*r..o* i.JOTE.i
- t.o T0k21JA.DO (3GO e :90° 13.1 --* MOME:t.iT ( 11::* FT /Fl) fl.& (k-1'-l/F-l) MOME.'-ITS 51-!0WIJ OIJ 'T&.l.ISIOIJ 510E:. .. 0.7 10.4 7."1 '20 *.IS TWl5Ti..iC. MOMEIJT (l('*fT/1'-T) ( .. ) lt.IOICA.'Tli&.S TEt.ISION (*) llJC\CA.'TE:S. * (t) (;.& 585 MER l()IOIJAL FOIZCE. (r./H) FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RESULTS OF SHIELD BUILDING STRUCTURAL ANALYSIS TORNADOWIND LOAD -906 SHEE1 I FIGURE 3.8-34
- LO MPl-I) e .. 0.0341-ELZOO.O' **** -ir-43.4 o.9 JW 1'10 1.15 100 +) &o a.o !Q__ I&.& _ ____ _ o ,._'!Q_ E:l+IO.o IWOP F-OiCE DEl'*Lt.CTIOtJ SllEA.li? 400P SMEU 0::/,.T) {llJ) (.le: /F-T) (K/F-1) OU iE.1115101J (+) llllOIC.A.i'-!> 'Tl&U':>IOl.J (*) COMPRl..SSIOIJ 10.6 ---\ 11.'l u_
- 114 llJ -Pl.At.It. 'SME.-'lil (K/F-T) FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RESULTS OF SHIELD SUILOING STRUCTURAL ANALYSIS TORNADO WINO LOAD -90* SHEET 2 FIGURE 3.a.37
- JL1oo.o' ll*IO.O 11. *S.o'
- 1.0 TOl21JADO (3"0 MPH) 9*1fi0° 1'20 100 (> &o GO 40 10 1\4.o ,,, 10.G._ ME21DIOl.llJ. MOIAUIT (tc:-ntm IJOTE* 1*.01 I I \ l'l.li 1.& -HOOP Mll!IDQjA.l l'OtCE. (tr::-F-T/FT) ( K/,..T) MOMlN"TS 5MOWLI OIJ 1'11!>1011 '010t. Ol 1u01CA1.'fU (*) tl.IOIC.i."TlS )l.f) ua 1.4c;. HOOP FOllC.E llOŽt.L OEt-LECTtOl.I (k./H) ( tN\ .s, .i&rf .19 .59 * ( 11:/,.."T) FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RESULTS OF SHIELD BUILDING STRUCTURAL ANALYSIS TORNADO WIND LOAD -1eoo FIGURE 3.S-38
- 160 lt;,o 14o \110 100 &O t;,o 4o *l*IO.O II '20 M.*t.O \.4
- UWIT TELIPERATURE -IWCRE..A.SE o.1o1 014 I -=-1.'Lll!> (-) 0.111 (<) O.OO'!I ,:11 LJ O.l'l' I (*) s 1!..I(, 0.08 0.015 1.91 MOMt.NT (li'*M/l>) HOOP h'\OMEllT MIRIOIOl4.t.L J:ollCli 1-lOOP l'<*E: OE:i:.lli:C:TIOIJ ( k* Fl/LI) (k' /L>) (k'/LF-) ( l(/LF-) * '-lO'T&.: Mo...ialolT'& s .. owN 01.1 s1oli. (*> lll,IOICAio.'Tlir.S Tli:lo.ISIOW (-) INOIC.\'fr*5 FLORIDA POWER & UCiltT a..Mf't St. Lucie Plant RESULTS OF SHIELD BUILDING STRUCTURAL ANALYSIS UNIT TEMPERATURE -UMffOltll INCREASE LOADING FIG4JIE :U-39
- al.*IQO l&o IGO 140 110 100 80 "'° A.a 'lO
- TEMPERATURE -Gl<A.DIEUT . 1.84' M!QIOIOIJAI. MOMlilJ'T ( W:.-F-T/LF-) '1..0 I
- 1.4 o.1b1 o.<>>G. (*) 0.611 (-) o.9!:lf> I L 0.131 1-!00P MOM&l.JT t.li:JllOIOIJb.l F0'2ca F-o12C.E: MEk?1DIOl.IA.l S\.!E,._I< (-t*F-l/l.1-) (K./LF') ("K/1.F-) (1'/1..F-)
- MOl.AE.IJ1S SHOWl.J 01.J "TC.IJSIOl.J ':>'Dt:: (1) 11.JO\C.a..,laS TalJSIO"" (*) 11.JOICA..Ta:, FLORIDA POWER IL LIGHT COMPANY St. Lucie Plant RESULTS OF SHIE[D8Ul STRUCTURAL ANALYSIS UNIT TEMPERATURE -GRADIENT LOAD FIGURE 3.8-<<>
- *
- WATER
- GRADE e E:l *le>' 0.000'2.1 trio 0.00.) _l!Q_ _140 .!1Q_ '!!"_ ,., "*"" 1.iz O,'l '°-o.ici. 1.1 ... , M1, _. lt\Ot.11lNT l<OOP 1.4ilt1DICN..i_ f-OICCE: MOOP FORCI: '-IORM4l MUIDIOlli.l .!145""8 (ll:*f'T/LF-) :l(*f:.'T/LF-) (1(./1..f'-) (K/Lf-) (V../1-,.) IJOTE.* SMOWIJ 510*. (*) lt.JOICAU.S ili.lll!;ION (*) IMOIC4'TI:.,!:, COMPl?E'!t!tlOW FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RESULTS OF SHIELD BUILDING STRUCTURAL ANALYSIS WATER tA&lE AT EL t 3' LOADING FIGURE tB-41
- i\.'\oo.o l\.*\o.o' 1.§9 1&0 '"° _lQ. "lO
- WATER TAE:>LE <= EL12.l1 G,R.ADE*@ El+l81 IS.1 Q\""\ r=:: I "'1ll210101J.ltol.. (I<:* F-1(1..f-) .ooe.'\ IAOlllE.111 Mcl?IOIG>J"L F<>l.'CE. (l'*F-T/L..F-) (. l(/Lf-) *0000\to \."I <>o**u. RJl!CE:. ( t::/LF-) OE:H.E.CTI0'-1 l.IE.1?10\?'l.t.L (11-1) ( k:/ LF-)
- t.JOTE: ON 51DE:. [*) llJOICA.il:.S fE.IJ::>IOIJ (-) COVIP1.d'..:i_,1c*1.J FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RESULTS OF SHIELD BUILDINC STRUCTURAL ANALYSIS WATER TABLE AT EL +21' LOADING FIGURE 3.8 ... 2
- *
- 1.0 E .b..RTl-I (0.1 %'). 9* 0° 140 100 g.o '1.:::> 1iL+10.o It-. 1 . . **41. ME:iZIDIOIJt.l (IC* i::T/1.f.) 1.c.1 _'11_.41 I I -IQA4 MOOP IAOMEIJT Mli.RIO\OWAl, Foli!CE ( 1(-FT/Lf.) ( IC/Lf.) MOMl;NTS !:>l-IOW\.J 01\j 'TE."15101-.1 51oe. (-) \.!OOP F<l2'-E ( '(/Lf.) (+) Tr*us1ot-.1 <*) tl.IOICA.Tlir= COMPilE:f)!>IOM O.Cil 1.w!!'. IJOli?MAL MU!\OIOIJ'-1. 5!.IUll C11.1) FLORIDA POWER & LIGHT COMPANY St. Lucie Plant RESULTS OF SHIELD BUILDING STRUCTURAL ANALYSIS SEISMIC LOADING FIGURE 3.8-'3 FLORIDA POWER & LIGHT COMP.ANY ST. LllCIE PLANT UNIT l TYPICAL CONCRETE SUPPORT FOIJNDATION FIGURE 3.8-44
Lines Associated with Steam Generator (For Steam Generator transients see Ref. 31) Operating Condition Lifetime Category Condition Occurrences N a Plant Heatup 100 F/hr 500 N b Plant Cooldown 100 F/hr 500 N c Plant Loading, 5% /min 15,000 N d Plant Unloading, 5% /min 15,000 N e 0% Step Load Increase 2,000 N f 10% Step Load Decrease 2,000 U g Reactor Trip 400 T h Primary side hydrostatic test, 10 3125 psia, 100 - 400 F T i Secondary side hydrostatic test, 10 1250 psia T j Primary side leak test, 2250 psia, 200 100 - 400 F T k Secondary side leak test, 1000 psia 200 N l Cold Feed Following Hot Standby 15,000 N m Normal Plant Variations, +/- 6 F primary, 106 +/- 40 psi secondary N n Primary Coolant Pump Starting and 4,000 Stopping U o Loss of Primary Flow 40 U p Loss of Turbine Generator Load 40 E q Loss of Secondary Pressure 5 E r Loss of Feed Flow 8 Reactor Vessel Head Vent Line (For Reactor Vessel Head transients see Ref. 31) Operating Condition Lifetime Category Condition Occurrences N a Heatup, 100 F/hr 500 N b Cooldown, 100 F/hr 500 N c Loading, 5%/min 15,000 N d Unloading, 5%/min 15,000 N e Step Load Increase, 10% 2,000 N f Step Load Decrease, 10% 2,000 U g Reactor Trip, Loss of Load 400 T h Hydrostatic Test, 3125 psia, 10 100 - 400 F T i Leak Test, 2250 psia, 200 100 - 400 F N j Normal Plant Variations, 106 +/-100 psi, +/-6 F U k Loss of Primary System Flow 40 U l Abnormal Loss of Load 40 E m Loss of Secondary Pressure 5 UNIT 1 3.9-19 Amendment No. 27 (04/15)
- SUPPORT Reodor Vessel Coolant Pump Lower St Gen Upper St Gen Steam line FLORIDA Ste:mline ICH Steam Generator 12 Pump 1 B TRANSLATIONAL FLEXIBILITY K(DN) K{UP) y y On/lb) On/lb) 1 x 10-14 1 x 10-4 Kz (in/lb) 1x10-12 1 x 10-7 1 x 10-12 5. 38 x 10-8 2 x 10-10 3. 96 x 10-9 Pump 2A KH (in/I b) 1 x 10-14 Figure POWER & LIGHT co. Reactor Coolant System Supports Mathematical Model St. Lucie Plant 3.9-1
- * '"ti 0 (/) :E ,... m :0 ,, cc: l'YI r :::> 0 ic.-0 ;:;: a;* r ::o ...... --J:!G'>O Q) :r: )> a -i (') 9 ::i:: CJ) 0 -n z -I m m r m s: m z -I s:: 0 0 m r 0 ,, -I ::r: m (') !/) OJ ---cri !/) -< !/) -I m s:: .,, 0 ::0 :::c: 0 ::0 C"') )> .... z .,, 0 ::0 :s:: )> :::! 0 z 0 z .... -< t -n I " -* IQ -s co CD \J'1 20 Ort>Z, I I IN. I 15 20 ..!---, I I 40 I 74.0---,i [J' ....... I { I S5.S 1'-DETAIL A R,IN. 60 I 25 I 6.125 I 74.0--10 12 ** 9**11 85.5-8* 1* 6* 5* 4* 3* 2* I .1 I DETAIL A UPPER FLANGE 80 I CSB * *
- 100 120 140 160 180 200 220 240 260 280 300 320 328.5 I I I I I I I I I I I I I 30 35 40 50 60 70 80 90 100 110 115 120 ,-, I I I I I I I I I I I I II I \ . r . ..
- I \ I '\DETAILS SUPPORT _J \ ..... \\ I l SNUBBER LUGS ELEVATION UPPER JACKSCREWS .....1 _J L -LOWER NODE JACKSCREWS NUMBERS c::-=; I 69.o
- 128 *-. 75.75 I 321.125 328.5 DETAIL B LOWER FLANGE
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- i 1 TENSION ONLY i ONLY FUEL ALIGNMENT PLATE UPPER GUIDE SUPPORT STRUCTURE PLATE FLANGE CONTROL ---ELEMENT SHROUD ASSEMBLIES CORE CORE SUPPORT PLATE FUEL GUIDE TUBES ..__LOWER SUPPORT STRUCTURE /1111/ll EXPANSION COMPENSATING RING CSB ...,._UPPER FLANGE CORE SUPPORT .,..,. __ BARREL SUPPORT LUGS ..,.__THERMAL SHIELD (Removed) FOR HISTORICAL INFORMATION ONLY Am. 3-7 /85 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 VERTICAL SHOCK MODEL Figure 3.9-3 I * * * * \_ UPPER GUIDE SUPPORT STRUCTURE PLATE ASSY CONTROL ELEMENT SHROUD ASSEMBLIES FUEL ALIGNMENT PLATE LEGEND: CORE SHROUD MASS POINT CONNECTED BY LATERAL SPRINGS FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 CORE SUPPORT PLATE LOWER SUPPORT STRUCTURE FOR HISTORICAL INFORMATION ONLY LATERAL SHOCK MODEL CORE SUPPORT BARREL THERMAL SHIELD (Removed) Am. 3-7/85 Figure* 3.9*4
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- Upper Barrel Core Region Lower Barrel FOR HISTORICAL INFORMATION Am. 3-7 /85 SAMMS OR/ D YNAS OR Figure FLORIDA POWER & LIGHT CO. Finite Element Model of Core Support Barrei 3.9-5 St. Lucie Plant
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- 345 LBF/IN LEVEL OF AXIS THROUGH HOT LEGS *-*i*-*-*-*-+ + CSB 65.2 IN. L_ 103.7 LBF/IN 90° . *-LEVEL5 213.4 IN. ' + + + 12.7 LBF/IN 90" * -*+ -*-LEVEL3 2700 I. 199.60 37.8 LBF/IN + + 35.6 LBF/IN TIN= 500 °F; Q = 130 % OF ODES Am. 3-7/85 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 NORMAL OPERATING LATERAL HYDRAULIC LOADS ON THE CORE SUPPORT BARREL Figure 3.9-6
A.4.16 kv SWITCHGEARThe following is a summary of Westinghouse test report G.O. NYMI18414-Yl dated February 1973.The report was issued as seismic qualification for the 4.16 kv switchgear.PCM's 89350 and 89351 modified the 4.16 Kv switchgear to accept PIP as well as Monolithic stylebreakers. While the PCM states that the seismic qualification of the switchgear remains valid, NLIreport R-037088-2, revision 1 documents in-cabinet response spectra which reflects the modifiedswitchgear configuration. 1.SynopsisA representative 50DHP350 metal-clad switchgear unit was subjected to an intensive test programper the intent of IEEE Standard #344 to qualify it for use in nuclear power stations with seismic vibration requirements. The seismic qualification of the 50DHP250 metal-clad switchgear unit is based on the test data of this 50DHP350 switchgear unit, a comparison of physical features, and static and dynamic analyses. The results of this test and analytical program have been analyzed in relation to the specific requirements of St. Lucie Unit No. 1 of The Florida Power and Light Company, per Spec. FLO-8770-284 with the floor response curves submitted by EBASCO Services Incorporated. The conclusion is made that the test program and the subsequent correlation and additionalanalysis of special features of the subject switchgear has verified the ability of both the 50DHP250 and the 50DHP350 metal-clad switchgear furnished for this station to operate satisfactorily in the specified seismic environment. 2.Description of Equipment TestedThe 50DHP350 circuit breaker and a representative cell, complete with relays and auxiliary equipment required on most orders, were selected for testing. This breaker, which has an interrupting rating of 41,000 amperes at 4.76 kv, is one of the most common ratings found in nuclear power stations. The unit tested was a standard design or "off-the-line" unit with no attempt made to improve its seismic capability. The relays were selected to represent the various types of relay movements and were located on the front door by the same computer program normally used for this purpose on regular commercial orders. The switchgear equipment was supported on three structural channels to duplicate normal servicemounting. The structural channels, in turn, were welded to steel plates which were bolted to the vibration machine. 3.Test ResultsThe switchgear was tested independently in three directions: front-to-back, side-to-side, and vertical. In each direction, tests were made at the resonant frequencies found during the initial frequency sweep as well as at selected off-resonant frequencies. The resonant natural frequencies discovered in the test3B-2Amendment No. 19 (10/02) equipment are 7 Hz side-to-side and 9.5, 13, and 16 Hz front-to-back. In addition, the vibrationmagnification that occurs at resonance of the equipment during the continuous sweep frequency search permitted the determination of the critical damping factors in the switchgear.The following damping factors at the various accelerometer locations are retabulated below: SideFront -to--to- Accelerometer LocationSideRear Arc Chute5%7% Door-5% Cell6.5%13% Side Panel6%- Potential Transformer Support-10%The lowest damping factor found in the switchgear was 5 percent. Therefore, 5 percent is a conservative figure for the qualification of switchgear equipment.In all, over 75 tests were run at various peak amplitudes; each test consisted of 5 sine beats,simulating one earthquake; each sine beat contained 5 cycles of the specified vibration frequency. Tests were made with the breaker closed, with the breaker open, and with the breaker opening and closing during a sine beat test. In all cases, a successful test required the breaker to retain its status quo unless signaled to respond. Failure to maintain its status or failure to respond would be considered a failure.To compare a floor response spectra curve to the test values, the value from the response spectracurve is divided by the Q-factor 5.5. EBASCO Services Incorporated has furnished the floor response curves at both the 19.5 elevation and the 43.0 elevation, for equipment, such as switchgear, having a 5 percent damping factor. Each of these curves shows that the peak responses occur only at periods greater than 0.2 sec. (or frequencies lower than 5 Hz) in any direction. As stated previously, the lowest natural frequency associated with the switchgear is 7 Hz and the switchgear was tested at this frequency or the greatest fundamental period. Because of the 5 percent damping factor, the maximum DBE response would then be as determined from the curves:3B-3 DirectionElevationElevation of Motion19.5 43.0 _________________________________ E-W0.44g0.39g N-S0.32g0.23g Vertical0.36g0.36g The required peak input for each response is determined by dividing the "Q" value (Q = 5.5 forequipment with 5 percent damping) into the peak response specified. In the above table, the greatest input required is for the E-W motion at the 19.5 elevation, since The actual test input to the switchgear was 0.8g a factor of ten (10) times the requirement of 0.08g. Qualification is established on the basis that the maximum expected response as shown in the EBASCO curves is smaller than the responses actually withstood by the equipment during the tests. 4.Test Summary 1.During this series of tests, the equipment was subjected to many more earthquakes than itwould ever experience during its economic life. 2.Though the equipment had been subjected to an excessive number of tests, there was nophysical equipment failure. 3.At no time during the tests did the breaker trip or close, unless called upon to do so. When signaled to operate, it did so relialy every time. 4.During some of the tests, a slight bounce of a normally closed contact of an SG-relay (with the coil de-energized) was observed. The switchgear required at this station includes 50DHP250 switchgear which has the same basiccell design as the 50DHP350 equipment, but the 50DHP250 circuit breaker has one basic difference from that of the 50DHP350 circuit breaker - namely, a smaller arc chute. The qualification of the 50DHP250 breaker is based on the test data for the 50DHP350, comparison of physical features, and dynamic analysis. Under conditions of seismic vibration, the dynamic response in the 50DHP250 breaker will be less severe than in the 50DHP350 breaker because the pole units are identical and the3B-4 effective mass of the 50DHP250 arc chute at the support is less than 80 percent of the effectivemass of the 50DHP350. Thus, the 50DHP250 arc chute with its (pole unit) support is well qualified for use in any seismic environment where the 50DHP350 arc chute and its supports have been previously qualified. Since the 50DHP350 breaker has been qualified by testing for service in the seismic environmentpostulated at the St. Lucie Unit No. 1 Nuclear Power Station as specified by EBASCO Services Incorporated, and since all components of the 50DHP250 breaker are as well qualified for the same seismic environment postulated as the corresponding parts for the 50DHP350 breaker, the 50DHP250 breaker is also qualified for service per the EBASCO Services specification. 5.Conclusion The sine beat method of testing is conservative because: 1.The response from a sine beat input is more severe than from random inputs. 2.Testing is performed at all natural frequencies determined in the equipment. 3.Additional tests are performed at other selected frequencies.From the tests and analysis presented in this report, it is concluded that the 50DHP350 metal-cladswitchgear will satisfactorily withstand the maximum seismic requirements of St. Lucie Unit No. 1 Nuclear Power Station. The qualification of the 50DHP250 switchgear equipment for service at St. Lucie is established by the seismic tests performed on the 50DHP350 switchgear equipment along with analytical comparisons based upon the static loading tests and other simple measurements. The conclusion is made that the test program and the subsequent correlation and additional analysis of special features of the subject switchgear unit has verified the ability of the 50DHP250 metal-clad switchgear furnished for this station to operate satisfactorily during the specified seismic environment.Refer to NLI report R-037088-2, revision 1 for in-cabinet response spectra which reflects the modified switchgear configuration as documented in PCM's 89350 and 89351.3B-5Amendment No. 19 (10/02)
MonitoredResonantTransmissibility 1 PointsAxisFreg. (cps)(output/input) ___________________________________________________________________________ 2Front to back19.0 29.0 3Side to Side12.0 3.5 4Side to Side10.0 1.75 5Front to back16.0 35.0 6Front to back22.0 60.0 There were no appreciable resonant conditions of Points 1 or 7, nor the structure frame, in the frequency range scanned.In 1972, an additional vibration scan was performed along the front to back axis at position number 2 onthe outside of the left front panel 10.5 inches from the top. The data is listed below:
FrequencyTransmissibility c.p.s. (ratio of output/input) 51.061.0 71.081.094.0 102.0 112.0 122.0 132.0 142.0 152.0 163.0 174.0 188.5 1916.0 208.0 215.0 226.25 236.0 243.33 252.33 261.0 270.66 280.35 290.75 301.0 311.0 321.5 331.0 1)Transmissibility is defined as a ratio of the output amplitude divided by the input amplitude.3B-7
TABLE 3B-2480 VOLT SWITCHGEARRESULTS OF SHOCK TEST 1 (AVERAGE ACCELERATION)I Six Cycle Shock TestsBLOWINPUTINPUTOUTPUTNO.HORIZ. ACCEL. gRESULT. ACCEL. gHORIZ. ACCEL. g10.60.70.621.11.41.131.62.01.641.92.22.052.12.52.362.32.72.472.63.22.982.73.23.092.73.22.8II Eleven Cycle Shock TestsBLOWINPUTINPUTOUTPUTNO.HORIZ. ACCEL. gRESULT. ACCEL. gHORIZ. ACCEL. g10.50.60.8520.81.11.431.41.72.442.02.53.452.22.73.862.32.83.91The output horizontal acceleration for blows above were interpreted from oscillograph film an thebasis of average acceleration and differ from values listed on Table 3B-1, which were peakvalues.3B-11 4.Mounted EquipmentFloor response spectra curves provided for usage on this equipment give a maximum DBEacceleration of 0.25g input to devices mounted on the equipment for 19 Hz resonance. The following devices require evaluation for this acceleration input.B/MPc. No.DeviceWithstandabilityRemarks50/51IAV> 0.25gCertified59EOB> 0.25gCertifiedCrkt. Bkr.33C77> 0.25gCertified5.Station Service TransformerThe following is a summary of the BBC Brown Boveri seismic qualification of the 480 voltstation service dry transformer.The transformer was mounted to a triaxial seismic test table. The test commenced with aseries of three sine wave resonance exploratory scans, one in each orthoganal axis. Thescanning rate was one octave per minute between 1 and 50 Hz at 0.2g. acceleration. Thenext seismic test was the first of five operating basis earthquake tests. For the seismic teststhe transformer secondary was energized so that primary voltage appeared at the primaryterminals. This test enveloped the Required Response Spectrum. However, it was noticedthat the low and high voltage air terminal chambers had a tendency to flap up and down onthe table since they had not been separately welded down. At this point they were welded byone inch welds between the lifting eyes on the bases of the air terminal chambers. It was alsonoticed that the displacement measuring linear variable differential transformer mounting wasin motion. The mounting was strengthened.The next four operating basis earthquake tests were performed without event. It wasobserved, however, that the differential transformer mounting continued to be in motion,invalidating the displacement measurements taken.Following the five operating basis earthquake tests a design basis earthquake test wasperformed. No anomalies were noted.Subsequent to the required tests reported above an optional test at the discretion of BBCBrown Boveri (the manufacturer) was performed at table limits in the region of the maximumof the required response. Levels above 10g.'s were recorded for the table motion. No damagewas sustained by the transformer, the mountings, or the transformer enclosures.At the conclusion of the qualification seismic tests, the transformer was returned to the Bland,VA transformer manufacturing plant where it was retested for all performance requirements.All the readings were within the limits of the ANSI requirements for dry type transformers.3B-12Am. 8-7/89 I Eba*eo Se.rvi£a.* Incorporat:<!ld Tw<i Reeter New York1 New York 10000 ITE Imperial .;o """-,."' " ... ,.,. '"""' Att:entim:t; Mr-. L . .J. Mulligan Thit ta to eerttfv that: 1-T-£ Type K L4lw 'Voltage Circuit fu:'aakera have LMn eei.lllmil) teated and li!lre c.art:tff.ed tor use at tlu.\c Florida Power & Ltsht, St. Lucie station. During taet* the eireuit breakers 1ueeftl$fully electrically op;rinad and elaetrically cloeed, I-T-E Il!Pi;:RlAL CORPORAT TEF/cr JB-13 -** """"'""' *-* C. SEISMIC QUALIFICATION OF 480 VOLT MOTOR CONTROL CENTERSThe following is a summary, of General Electric's Report No. 70ICS100, dated September 3,1970,which qualifies the 480 volt motor control centers used on St. Lucie Unit 1.1.AbstractThese tests indicate that the 7700 line motor control center is suitable up to at least 0.5g baseinput accelerations through a frequency band width from 5 to 500 Hz.2.Description of TestsThe test article was attached to the vibration table using conventional bolting methods typical of actual installation techniques. The vibration fixture (table) was non-resonant within the frequency band width of interest.Accelerometers monitored input acceleration at the base of the equipment and resultingresponse accelerations at significant points within the test article. Each response monitor point was instrumented to detect vibration acceleration response in the direction of the input forcing (base) vibration.The test article was swept in frequency from 5 to 500 Hz at a one-half octave per minutesweep rate, at a constant 0.5g input acceleration.The equipment was vibrated in each of its three orthogonal axes; vertical, horizontal, in-breadth, and horizontal fore-and-aft.All vibration sweeps were made with a 480 volt ac 60 Hz source connected to the controlcenter main bus.Vibration sweeps were made in two modes of functional status: a)with each starter unit disconnect in the ON position, but starter not energized. b)with each starter energized.The equipment was vibrated with all doors removed from the enclosure to facilitateobservation of component behavior during tests. The doors add little to the rigidity of the structures, and their removal did not significantly alter test results.The input acceleration levels were held at a constant 0.1 inch double amplitude displacementfrom 5 Hz to 10 Hz instead of a constant 0.5g level due to machine limitations. The date was normalized and this deviation does not affect the accuracy of the data acquired.3B-14 3.Test Results and DiscussionPlots of input and response acceleration levels for the three orthogonal axes of vibration weremade for various accelerometer locations. The plots reflect data recorded during the de-energized mode of starters, and represent worst case conditions insofar as response acceleration levels are concerned.No changes in functional status were observed during the vibration sweeps: i.e., starters inthe de-energized mode did not close, and starters in the energized mode did not oven.An examination of the plots shows a prominent resonant point, at 5-6 Hz, when the equipmentwas vibrated in both horizontal directions. This resonance at all monitored points represents the structure resonant frequency. Resonances at various higher frequencies represent individual component resonant points.Table 3B-3 summarizes maximum acceleration levels between 5 and 100 Hz measured ateach response point, as taken from plots. An examination of the data of Table 3B-3 indicates the following:a)highest response accelerations were generallv found, as would be expected, atbrackets attached to starter unit frames (Pushbutton brackets, cantilever mounted,are particular examines) .b)no discernible pattern is evident to relate the vertical position of a starter unit withinthe structure to the response acceleration of the unit.c)disregarding the relatively high acceleration responses of pushbutton brackets, allmonitored points display reasonably similar maximum response levels in any given direction of input acceleration (maximum recorded difference indicates an approximate 3:1 ratio of response level).4. Conclusionsa)The tests conducted indicate the equipment as tested is suitable for applications up toat least 0.5g base input accelerations through a frequency band width from 5 to 500 Hz.b)The resonant level of the equipment structure appears to be at 5-6 Hz, whilecomponent resonant levels are at higher frequencies.c)The devices can be satisfactory operated when subjected to the seismic testingdescribed above. (Table 3B-2B is a letter certifying this).3B-15 TAIL! lB--21 GENERALlllJ ELECTRIC '""' ...... .. of Com;ilia<< .M r<>r< ,_" ' Upt ec.""1' lh.>tchi!i*ml 1.!11-11, J'lorid:a INDUSTlY DOPJ.RTM*NT Thi§ 1* tc. =nify that ek C:.U. l f,.,,,,i*h*d ..., tilt* N.,.tat*i.,.. _,,, tlt!I ,,f. Eb&!!co .... }11).-t.t U ttt-riJill<f Ju.fy <(l, J.tit. tutl1:11 ** dta<:.;il>tl ta T<<llh:liW kpll'tl t&. <R. g.,...... t<) @d t\lbM:rtli<Je U;foCll * .. , ' ' )(llr., t!>i*2Ld.v pf 077 ll74. TABLE 3B-2B (Cont'd)NOTE ON SPECIFICATION 210-69Ebasco specification 210-69 (July 20, 1971) specified the following:"If a seismic design is specified in Part One, Seller shall provide test data to demonstrate theadequacy of his product to withstand the effects of the specified seismic forces. A product which complies with the intent of this requirement shall after exposure to specified seismic forces:a)Exhibit no undue deflection which would prevent any component specified in thisspecification from performing normal uninterrupted operation.b)Have no components dislocated, which would prevent uninterrupted normal operation(i.e. fuse thrown out of fuse holder, bolt used to mount control transformer sheared, etc.).c)Maintain all components in the same operating position during the disturbance asthey were prior to it.d)Permit operation of all components during the disturbance, i.e. if starter is energizedor deenergized during the disturbance, it will react accordingly.In testing, the horizontal and vertical accelerations shall be applied simultaneously. The application of the combination of these two forces shall be repeated to simulate horizontal acceleration of the gear, in as many directions as necessary to demonstrate the adequacy of the gear due to its asymmetry. Response spectra curves for the ground motion due to earthquake are attached and form part of this specification."3B-17 TABLE 3B-3MOTOR CONTROL CENTERS MAXIMUM RESPONSE ACCELERATIONS TO 0.5g BASE INPUT ACCELERATION (BETWEEN 5 & 100 Hz) VERTICALHORIZONTALHORIZONTAL IN-BREADTH FORE-AND-AFT (g)(g)(g) ACCEL. LOCATION SZ. 4 FVNR UNIT FRAME1.71.22.0 SZ. 4 RVNR UNIT1.3 .951.3 SZ. 4 RVNR STARTER BASE1.4 .801.5 SZ. 4 RVNR PUSHBUTTON BRACKET5.5 .801.7 SZ. 3 FVNR DISCONNECT BRACKET2.2 .90 .75 SZ. 3 FVNR UNIT FRAME2.1 .90 .60 SZ. 1 FVNR DISCONNECT BRACKET2.4 .85 .85SZ. 3 FVNR PUSHBUTTON BRACKET8.41.48.1 SZ. 4 RVNR UNIT FRAME1.5 .721.1 SZ. 4 RVNR DISCONNECT BRACKET2.8 .85 .85 SZ. 2 FVNR UNIT FRAME1.9 .951.1 3B-18
ESFAS Relays - Cabinet Amplification Summary 3The NRC in its supplement to the SER of May 1975 at Section 8.3.3 questioned the validity of thetests performed in reference 2 with respect to the possible amplification of input g values for relays mounted in the ESFAS cabinets.The relay panel was analyzed (reference 3) to determine its natural frequency. The calculatedfundamental frequency of the panel was found to be 16.7 Hz. The cabinet with the relays was also subjected to a shake table test (reference 1). The test results indicated that the overall cabinet framing natural frequency was 20Hz or over. The cabinet relay resonance frequency of 20 Hz compares favorably with the calculated value of 16.7 Hz.The floor response spectra for elevation 6.2 of the reactor auxiliary building are provided by Figures3.7-21 and 23. From these figures it is evident that components with natural frequencies greater than 6.7 Hz will not experience resonances. Therefore, the cabinet and all of its components will respond as a rigid body to a maximum floor acceleration of about 0.2g, i.e., the cabinet will not amplify the floor response. In view of the 10 g accelerations used in the tests of reference 2, it is clear that the relays have been successfully qualified at seismic levels well above the design levels for St. Lucie 1.3B-28 References1)Consolidated Controls Corporation, Engineering Report No. 824, "Seismic Test Report forEngineered Safeguards Panels - St. Lucie Nuclear Power Station Unit #1," December 5,1973.2)Consolidated Control Corporation Engineering Report No. 862, "Seismic Test Report of4CP36-AF Relay Mfg. by S. H. Couch Div - ESB Safety Features Actuation System" May 7, 1974.3)Consolidated Control Corporation, Engineering Report No. 912, "Second Addendum toSeismic Test Report for Engineered Safeguards Panels - St. Lucie Nuclear Power Station Unit #1," May 23, 1975.3B-29 3B-30 CERTIFICATIONEngineered Safeguard Logic PanelsSt. Lucie Nuclear Power StationUnit #1Februrary 14, 1974IEngineered Safeguard Logic modules 6N81 through 6N87 similar to 6N88, 6N89, 6N90, 6N91and 6N92 were temperature qualified to 130 degrees Fahrenheit and 94% RH. The modulesmet all specification during the seven day test. Consolidated Controls Corporation Engineering Report # 803 dated 2/21/73 incorporates the procedure and data of this qualification.IIEngineered Safeguard Logic actuation relays S. H. Couch type 4CP36-AF were tested both inthe energized and de-energized state during seismic qualification of the St. Lucie I equipment. Proper operations of the relays were verified prior to and subsequent to the application of the seismic forces. This data is contained in Consolidated Controls Corporation Engineering Report # 824 dated 12/5/73.In addition to the above qualification testing, Consolidated Controls Corporation haspreviously qualified actuation relays built by S. H. Couch Company. Consolidated Controls Corporation Engineering Report #771, Confidential Restricted Data dated 12/5/69, documents the operation of S. H. Couch 4AP37-AF relay during a MIL-STD-167 vibration test. After three hours of vibration, two relays were switched due to the application3B-31 CERTIFICATIONPage Twoof trip imput driving signals. The equipment tested was Reactor-Protective Equipment being providedto the U. S. Navy. The 4AP37-AF relay is of the same construction as the 4CP36-AF relay being used in the St. Lucie I equipment. The only differences are the "A-P" which changes contact current capacity and the "37-36" which changes the coil impedence. The relay operated properly during the vibration and met all specified performance requirements./lmp3B-32 F. SEISMIC QUALIFICATION OF BATTERY CHARGERThe following is a summary of tests performed by TII Testing Laboratories entitled "Report of SeismicShock Test on One (1) Battery Charger for C & D Batteries Division of Electra Corporation - Plymouth Meeting, Pennsylvania," dated September 29, 1970.1. Vibration SurveyPrior to the seismic shock tests, the battery charger was subjected to a vibration scan in each of thethree mutually perpendicular axes, in a frequency range from 5 to 55 cps. The following resonant and/or natural frequencies were noted and recorded:Axis Resonant Frequency TransmissibilityVerticalNo appreciable resonance notedFront to Back27 cps3.75 - 1Side to Side27 cps5 - 12. Test ProcedureOne battery charger, bolted to a one-inch steel plate, was mounted to the table of the Seismic ShockMachine.Accelerometers were attached to the shock machine platform to monitor the vertical and horizontalaccelerations. One accelerometer was attached to the unit under test at the top front of the charger.Once satisfactory operating conditions were established, the unit was subjected to a series of seismicshocks applied through the front to back direction of the unit at an angle of 33 +/- 2 from the horizontal.The magnitude of shock acceleration was increased to a maximum of 1.96 gravity units horizontal simultaneously and linearly combined with 1.28 gravity units vertical. After each blow, the unit was visually examined for evidence of physical or operational damage.At the completion of this portion of the test, the battery charger, with base plate, was reoriented 90about its vertical axis and again secured to the shock machine. The charger was subjected to at leastone seismic shock in the side to side direction at the maximum acceleration loadings as described above.The fundamental frequency of the seismic shock was 10 cps.3B-33
- 3. Results of Seismic Shock TestsThe battery charger was energized prior to and after shock along the front to back direction. Therewas no apparent physical or operational damage.The following accelerations were recorded during these shocks:
Shock No.InputInputOutput at Top ofHorizontalVerticalBattery Charger(g's)(g's)(g's)10.50.41.120.90.91.231.21.02.041.51.41.851.71.52.562.01.62.6The battery charger was energized at no load during the tests in side to side direction. There was no apparent physical or operational damage to the unit.The following accelerations were recorded during these shocks: Shock No.InputInputOutput at Top ofHorizontalVerticalBattery Charger(g's)(g's)(g's)10.60.40.920.90.71.431.21.01.641.91.62.252.01.72.44. ConclusionsThere was no apparent physical or operational damage to the battery charger as a result of simultaneous accelerations of at least 1.96 g's horizontally combined with 1.28 g's vertically during the shock tests at a fundamental frequency of 10 cps.3B-34
3B-36
it is 55 psi and 7 psi respectively. These resultant stresses are of insufficient magnitude to have any deleterious effects on the pump.
Assuming a rupture in a main steam line and assuming an adiabatic expansion of the escaping steam, the temperature of the steam will decrease to approximately 320F upon release from the steam line. This situation can be assumed to exist for a total of from 60 to 95 seconds (depending on initial power level) during which time the effected steam generator blows dry. (We assume a loss of normal feedwater since this is the only condition which would require the use of the auxiliary feedwater system). The maximum temperatures will only be experienced by the pump towards which the jet is directed.
The equipment manufacturers for the pumps and pump motors have stated that their equipment can function in the ensuing environment described above with the only possible ill effects being the failure of pump seals due to the temperature. This type of failure could result in the loss of a maximum of 5 to 10 gpm but no loss of function. There is also no danger that a rupture of a steam line or feedwater line could cause a loss of function of more than one auxiliary feedwater pump due to flooding. Each of the three pumps are provided with a flood wall around them to elevation +22 ft. with an access opening that would preclude any water buildup. Under normal conditions accumulation within the enclosure is impossible since the condensed steam will run out over plant grade. There is no other credible postulation of interaction between a ruptured main steam line and any connected branch line that could lead to a more detrimental condition than that described above or otherwise affect the plant capability for safe shutdown.
- 6. Stresses were analyzed in the steam trestle using the working stress method. The load combinations and allowable stresses are as follows: Loading Stress Dead Load + Live Load + Thermal - Allowable stress per Load + Seismic Load (Operating AISC Code (A-36 steel) Basis Earthquake)
Dead Load + Pipe Break Load - 1.5 x allowable stress per AISC code (A-36 steel) Dead Load + Thermal Load - 1.5 x allowable stress + Seismic Load (Design per AISC Code Basis Earthquake) (A-36 steel)
UNIT 1 3C-4 Amendment No. 27 (04/15)
one auxiliary feedwater pump becomes inoperable; there will be no loss of system function if the motor driven steam stop valve (I-MV-08-3) becomes inoperable; and, there will be no loss of isolation function on the intact main steam line since the valve is designed to fail closed on loss of power. Safety related features of the plant, other than those identified above, will not be affected by a main steam or feedwater line break since the propensity for damage is attenuated with distance and no additional safety related equipment is located in the area.
- 14. Drawing 8770-G-149, sheets 1 and 2, and drawing 8770-G-408, sheets 2A & 2B show the routing of the MS and FW lines from the containment to the turbine building, the locations of the auxiliary feedwater pumps and their respective piping, and the electrical cable routing in the trestle area. As stated in item 12 of this report, the control room wall and one of its 100 percent capacity ventilation intakes are approximately 85 ft south of the closest FW line and 100 ft south of the closest MS line. A second 100 percent capacity ventilation intake is located on the south side of the reactor auxiliary building. Both ventilation intakes are located at elevation +78' 9". 15. As stated in section 5 of this report, there is no potential for a flooding event caused by a main steam or feedwater line to prohibit the auxiliary feedwater system from performing its function.
Each of the three pumps are surrounded by a flood wall to elevation +22 ft with an access opening which would preclude any water buildup.
- 16. The main steam and feedwater piping are designed as seismic Class I and nuclear safety class 2 (quality group B) from the steam generators up to and including the isolation valves outside the containment on the trestle. For fabrication and installation, these lines have complied with (fabrication) and will comply with (installation) the quality assurance requirements of class 2 piping as specified in ANSI B31.7. This entails 100 percent radiography of all welds as well as the non-destructive testing requirements stated in the material specifications. The St. Lucie Plant has an inservice inspection program for quality group B components.
- 17. Not applicable.
UNIT 1 3C-9 Amendment No. 27 (04/15)
APPENDIX 3DANALYSIS OF HIGH ENERGY LINE RUPTUREOUTSIDE CONTAINMENT*INTRODUCTIONThe systems analyzed herein are the lines used for shutdown cooling which includes portions of thelow pressure safety injection system, chemical volume and control system (letdown and charging lines), steam generator blowdown system and auxiliary steam system.In analyzing the effects of rupture in these high energy lines on systems or components required forsafe shutdown, except as noted hereinafter, the following general criteria were considered:a)For those concrete structures protecting systems and components essential to safeshutdown, the load combination of pipe rupture and design basis earthquake (DBE) isassumedb)Single active failure in addition to the pipe rupture is assumedc)Other than normal shutdown systems (e.g., ECCS) are considered acceptable to achieve safeshutdownd)Piping which is pressurized only during testing is not considered e)The criterion used to demonstrate structural adequacy is to show no loss of function.*Refer to Appendix 3C for analysis of main steam and feedwater lines. Pursuant to therequirements of IE Bulletin 79-01B, a reevaluation of the environmental qualification of electricalequipment installed in the plant was performed. This updates a portion of the information provided in this appendix. See Section 3.11 for referencing to FPL responses to the bulletin.Information provided in this appendix is historical and shall not be updated; however, it may still be similar to the re-evaluation documentation if no changes have taken place. 3D-1Amendment No. 18, (04/01) In addition to the MS and FW lines as presented in Appendix 3C, there are four pipe lines which havebeen identified as high energy pipelines outside the containment. These are the steam generator blowdown lines, auxiliary steam lines, letdown and charging lines, and the shutdown cooling system (including portions of the LPSI System). All these lines are located in or on the roof of the reactor auxiliary building which adjoins the containment structure.These lines with the exception of the auxiliary steam lines are designed with steel pipe restraintswhich are supported from embedded plates in the concrete floors, walls or ceilings. The pipe restraints are only loaded in the event of a pipe break; separate pipe hangers or supports carry the normal pipe loads and separate seismic restraints provide seismic resistance.Ultimately, all loads are resisted by the affected concrete walls, floors and ceilings. To analyze for theloads imposed on concrete elements, each affected pipe system has been traced so that all pipeloads imposed on a particular concrete element are identified and located. For pipe restraints, the piperestraint which imposed the most critical load (in terms of magnitude and point of application) is applied to the concrete element. The concrete element is then analyzed in terms of all imposed loads in accordance with loading combinations indicated in Section 3.8.6.4.It should be noted that the largest pipe break load for these lines imposed on a pipe restraintsupported by concrete is in the order of 7,000 lbs. Normal pipe anchor loads are of similar magnitude. Most of the structural concrete walls and slabs affected by this investigation are 2 ft. thick withmedium to heavy reinforcing. The loads imposed on the elements by piping, either normal or accident, are relatively small in comparison to the capacity of the element.An analysis is presented herein of wall Number 5. This wall was identified as a critical element. As canbe seen from the calculations, the flexural capacity of the wall is twice the imposed load. See Appendix 3E.The effects of high energy pipe breaks have been reviewed for their effect on structural concreteelements in accordance with the criteria set in Section 3.8.6.4. No deficiencies in the structural strength of concrete elements have been determined; therefore, it is concluded that the effects of high energy pipe breaks outside the containment are within the capacity of the structural elements of the building.3D-2Amendment No. 17, (10/99) ANALYSISThe following analysis is based on the AEC issued "General Information Required for Consideration ofthe Effects of a Piping System Break Outside Containment."'1)Protection against pipe whip is provided outside containment based on the following criteria:a)Shutdown Cooling System (including portions of the LPSI System)(1)Maximum operating pressure-and temperature during shutdown cooling is450 psig and 300 F, respectively.(2)Although restrained, a ruptured shutdown cooling system line, were it notrestrained, could impact a component cooling water line and a containmentspray line. Of the two, the containment spray line is not necessary for safe shutdown.NOTE:The shutdown Cooling System entry temperature has been increased to 325F byrevision to the plant Technical Specifications. This increase in temperature does not increase the potential for loss of structural function as detailed in this analysis.b)Chemical and Volume Control (CVCS) System - letdown and charging lines(1)Maximum operating pressure and temperature for the letdown line duringnormal operation is 2219 psia and 450F, respectively. Downstream of theletdown heat exchanger and after the pressure has been lowered by theletdown control valves, a pressure reducing valve lowers the pressure to 200psig; temperature at this point is 140F.Maximum operating pressure and temperature for the charging line duringnormal operation is 2300 psig and 120F, respectively. However, any rupture in the line reduces the pressure to essentially zero since the charging pumps are of the reciprocating type.(2)Although restrained, a ruptured letdown line, were it not restrained upstreamof the pressure reducing valve, could impact a component cooling water line which is necessary for safe shutdown.c)Steam Generator Blowdown System(1)Maximum operating pressure and temperature during normal operation is900 psig and 532F, respectively.(2)Although restrained, a ruptured blowdown line were it not restrained couldimpact a component cooling water line which is necessary for safe shutdown.3D-3Amendment No. 17 (10/99) d)Auxiliary Steam System(1)Maximum operating pressure and temperature during normal operation is 40psig and 350 F for one main section of the auxiliary steam supply system,and 75 psig and 350 F for the other. Various branch lines operate at lower pressures and temperatures and will be discussed separately.(2)One branch of the auxiliary steam system supplies steam to the boric acidand waste concentrators and passes through portions of the RAB. Another branch supplies steam to the decontamination facility but does not enter the RAB except for the decon area itself.2)The design criteria employed throughout the plant for the design of restraints and theirspacing are predicated on restraining the individual pipe regardless of the location or orientation of all postulated ruptures. To achieve this, tables and charts were developed which indicate maximum spans allowable between restraints to prevent development of a plastic hinge and concomitant pipe whip. The tables and charts were prepared based on calculated load combinations which can be expected to result from either circumferential or longitudinal ruptures occurring along straight runs of piping, after elbows, etc. The tables and charts take into account variables such as size and configuration of the lines and the contained energy. Lines 1 inch in diameter or smaller are not considered.3)Addressed in section 2 above.4)A static analysis which combined the design loads of the piping at operating conditions, theloads calculated to exist after a pipe rupture, and DBE forces was utilized in the design of thepipe whip restraints and their spacing.Both the circumferential and the longitudinal breaks considered are based on a flow areaequivalent to the cross-sectional area of the pipe. Refer to Section 3.6 of the Safety Analysis Report for a complete description of pipe whip analysis method.5)A description of the measures employed to protect against pipe whip, blowdown jet andreactive forces for the systems under consideration are presented in detail in Section 3.6 of the SAR. The criteria presented in Section 3.6 are extended to apply equally well to non-seismically designed portions of high energy piping. A summary of the criteria is presented below:Pipe whip restraint locations for this plant are chosen based on maximum pipe spans requiredto develop ultimate moment and torque capabilities of the pipe cross section. Break locations in piping are assumed to occur at any location along the piping in all systems with normal operating pressures above 125 psig. The design loading combinations and stress criteria used for pipe restraint design are stated in SAR Section 3.6.3.3D-4 Actual pipe whip restraint location and spacing are shown on SAR Figures 3.6-10 and 3.6-19through 3.6-32 for the shutdown cooling and low pressure safety injection system and on Figure 3D-1 for the letdown line. The auxiliary steam system is not restrained.6)The evaluation of the structural adequacy of Category I structures as well as the designcriteria used for these structures is discussed in Section 3.8.6.7)The structural design loads, including the pressure and temperature transients, the dead, liveand equipment loads, and the pipe and equipment static, thermal, and dynamic reactions are discussed in Section 3.8-6.8)Seismic Category I structural elements such as floors, interior walls, exterior walls, buildingpenetrations and the buildings as a whole have been analyzed for eventual reversal of loads due to the postulated accident. The accident forces are relatively low; design changes are not necessary, based on an analysis of all high energy lines.Ruptures of auxiliary steam lines will not adversely affect Category I structures,9) Not applicable. 10)An evaluation of the consequences of a high energy line accident including failures caused bythe accident is presented below: a)Shutdown Cooling System (including portions of LPSI system)The limiting or worst case shutdown cooling line rupture is the one which allows thegreatest amount of reactor coolant to escape into the reactor auxiliary building, i.e., a rupture of a 12 inch line just at the onset of shutdown cooling. The temperature and pressure conditions at this time are 300 F and 450 psig, respectively.
- The resultant temperature and pressure rise, 50 F and less than 1 psig respectively, have an insignificant effect on surrounding structures. Refer to Section 3.8.6.See note in paragraph 1.a) for system max operating temperature revision.3D-5 b)Letdown LineThe limiting or worst case letdown line rupture occurs between the containment penetration andthe pressure reducing valve where the internal energy is maximized. A guillotine rupture of the line is considered which results in a critical two phase blowdown into the reactor auxiliary building pipe tunnel area. The steam and water discharge at near sonic velocities through the 2 inch pipe.A high temperature alarm downstream of the regenerative heat exchanger alerts the controlroom and also initiates closure of the upstream letdown isolation valves. Approximately 1000 lbs of letdown water will blowdown before valve closure. The water is assumed to be at 600 F since no cooling in the regenerative heat exchanger is expected at such a high flow rate. The resultant temperature and pressure rise, 50 F and less than 1 psig respectively, have an insignificant effect on surrounding structures. Refer to Section 3.8.6.c)Steam Generator Blowdown LineThe letdown line rupture discussed in (b) above is limiting, i.e., the blowdown line rupture results in a less severe transient. The blowdown lines are isolated by one of two containment isolation valves (one inside and one outside of containment). Redundant primary devices detect a line rupture, alert the control room and initiate closure of the isolation valves.d)Auxiliary Steam SystemThere will be no structural failures as a result of auxiliary steam line ruptures.Auxiliary steam lines 3-AS-13, 14 and 16 (refer to Figure 3D-2) have been terminated (flangedoff) prior to entrance into the RAB and as close to the main header as possible. Auxiliary steam lines 12-AS-1 and 3/4-AS-31 are the only auxiliary steam lines to enter the RAB, although only for a short distance and not near safety related equipment.
Nonetheless, redundant heat sensors have been added along the run to alert the control room of a rupture and to automatically terminate blowdown by closing valves PCV-16-1 and MV-08-12 for 12-AS-1, and valves PCV-16-6 and TCV-08-5 for 3/4-AS-31. (Refer to Figure 3D-2).In addition, heat sensors have been added to the control room north outside air intake duct, which will automatically close the control room isolation valves in the event of a rupture of 12-AS-1 which is routed approximately 12 feet away. Since the line pressure at this point is only = 27 psig, impingement is not a concern. Calculations have shown that the maximum temperature rise in the control room is within 20 F of ambient assuming 750 cfm (intake fan capacity) of steam is drawn directly into the control room until closure of the isolation valves.3D-6Amendment No. 20 (4/04) 11)The analysis presented below is an evaluation of the effect of high energy line ruptures onsafety-related systems necessary to mitigate the consequences of those ruptures and place thereactor in a cold shutdown condition.Safety related systems needed to mitigate the effects of any particular high energy line break may vary depending on the particular circumstances and assumptions. These are discussed ona case by case basis below. Safety related systems needed to bring the reactor to a cold shutdown generally include the auxiliary feedwater system (assuming main feedwater is not available) and the shutdown cooling system (including portions of the LPSI system).Further, the analysis investigates cases where high energy line ruptures may result inenvironmentally induced failures in other protection systems which are not required to function to mitigate the rupture or bring the plant to cold shutdown. For these cases loss of redundancy but no loss of function is permitted.a)Shutdown Cooling System (including portions of the LPSI system)A shutdown cooling line rupture does not initiate an automatic protective system function. Since the shutdown process is a carefully controlled administrative process, the operator is always aware of critical system parameters as pointed out in Section 20 of this appendix.During the shutdown cooling mode of operation, the shutdown cooling system can accept arupture of one shutdown cooling line and a single active failure. However, it is not a design basis of the system to accept any passive failure although selective passive failures could be tolerated.Although not designed for passive failures, a single pipe rupture in the shutdown coolingsystem can not affect other portions of the shutdown cooling system which are redundant. This is accomplished by physical separation. However, certain portions of the shutdown cooling system which are not redundant could disrupt the shutdown cooling process were they to rupture.The calculated temperature rise to 170F will not have any effect on cables in the area sincethese cables are qualified to 270F, 44 psig and 100 percent humidity for at least 15 minutes, for longer than the duration of the blowdown effects. All electric pump motors and motor operators for valves are designed for a 40C temperature rise.b)Letdown LineA letdown line rupture could initiate an ESFAS signal (SIAS) if reactor coolant pressuredropped below3D-7Amendment No. 17 (10/99) the SIAS actuation setpoint. If an SIAS were generated, it would automatically close the letdown loopisolation valves thereby terminating blowdown. Blowdown may be terminated before the SIAS by thetemperature element downstream of the regenerative heat exchanger which also automatically closes the isolation valve (on high temperature). Since the isolation valve is located inside containment, any rupture outside containment, will not effect them.A letdown line rupture can not affect the operation of the auxiliary feedwater system or the shutdown cooling system due to physical separation of the components.The calculated temperature rise to 175 F will not have any detrimental effect oncables in the area since these cables are qualified to 270 F, 44 psig and 100 percent humidity for at least 15 minutes, far longer than the duration of the blowdown effects.c)Steam Generator Blowdown LineA blowdown line rupture would not initiate an automatic protective system function. Flow through a ruptured blowdown line will be automatically terminated by primary devices located downstream of the isolation valves. The steam air environment resulting from line rupture will not adversely affect the sensors or isolation valve located outside of containment.The effect of blowdown line rupture and subsequent temperature rise in the pipetunnel is less severe than that calculated for the letdown line, therefore, no detrimental effect on cables in the area is expected.d)Auxiliary Steam SystemThe only auxiliary steam lines entering the RAB are 12-AS-1 and 3/4-AS-31. Neither of the lines are routed near safety related equipment and by virtue of the relatively low line pressures, = 27 psig and 75 psig respectively, ruptures will have no effect on safety related structures. In any event, ruptures will be detected by local redundant heat sensors and automatically terminated.12) Control room habitability is not affected by any high energy line break since redundant heatsensors automatically isolate the control room if unusually high temperatures are sensed.3D-8Amendment No. 17 (10/99) 13)The necessity for electrical equipment to operate as a result of high energy line breaks isdiscussed below:All safety related electric pump motors and motor operators for valves are designed for a 40 Ctemperature rise as stated in Section 11. All electrical cable to safety related equipment outside containment is qualified for 270 F, 44 psig and 100 percent humidity for 15 minutes as stated in Section 11.Since no protection system functions are necessary to mitigate the effects of ruptures in thehigh energy lines considered, no equipment within the sphere of influence of the rupture need operate. Note, however, that all safety related equipment in proximity to the ruptures considered are designed for environmental conditions more severe than those calculated to exist because of the ruptures. This assures no loss of redundancy even in protection systems not required to function as a result of the rupture.Sufficient physical separation and pipe whip restraints preclude damage to protection systemequipment from either jet impingement or pipe whip. No barriers are thus required.As stated in Section 12, the control room and its equipment will not be affected by any highenergy line rupture.14)Design diagrams for the shutdown cooling and LPSI systems are provided in the SAR asFigures 3.6-10 and 3.6-19 through 3.6-32. The letdown line is shown on Figure 3D-1. A drawing showing the routing of the auxiliary steam system is shown on Figure 3D-2.15)The potential for flooding of safety relates equipment in tile event of a high energy line ruptureis discussed below:a)Shutdown Cooling System (including portions of the LPSI system)A rupture in any portion of the shutdown cooling or LPSI piping could directly oreventually drain to the ECCS pump rooms which house safety related equipment. However, these rooms are equipped with sumps which alarm in the control room on high level which allows the operator to isolate all drain lines to that room or to shut off the source of the leak or both.b)Letdown, Blowdown and Auxiliary Steam SystemsThere exists no potential for flooding safety related equipment since the volume of water involved in any case is limited. In any event, all leakage to the ECCS pump rooms is alarmed and isolable.3D-9 16)The quality control and inspection programs followed for piping systems outside thecontainment is as described below:a)Shutdown Cooling System (including portions of the LPSI system)For St. Lucie Unit 1, the system is designed as Quality Group A and B outsidecontainment (St. Lucie Unit 2 system are Quality Group B). For fabrication and installation on Unit 1, these lines will comply with the quality assurance requirements of code class 1 and 2 piping as specified by ANSI B31.7 (for Unit 2 the ASME Section III code class 2 is used). This entails 100 percent radiography of all welds as well as the non-destructive testing requirements stated in the material specifications. These components are covered by the plant inservice inspection program.b)Letdown SystemOutside of containment, this line is Quality Group B up to the letdown control valves designed to code class 2 of ANSI B31.7 for Unit 1, and designed to code class 2 of ASME Section III for Unit 2. The quality control and inspection specified for class 2 applies.c)Blowdown LineOutside of containment up to and including the isolation valve, the blowdown line is Quality Group B designed and inspected as in item (b) above. In the original plant design, beyond the isolation valves up to and including the primary (pressure) sensors, the blowdown line was Quality Group C and seismic Category I. Beyond the sensors the line is designed to ANSI B31.1 standards. The NDT requirements are consistent with that required by the code. Since the line is at high pressure, at least 10 percent of the welds in the line are radiographed.d)Auxiliary Steam LinesThese lines are designed to ANSI B31.1 standards and they do not have any special NDT beyond that required by the code.17)Not applicable.18)Since ruptures in the letdown, blowdown and auxiliary steam systems have no effect on theplant's ability to shut down, the emergency procedures described in SAR Section 9.3.6 for theshutdown cooling system apply for all cases.19)High energy lines that pass near structures, systems or components important to safety aredesigned as seismic Class I and either Quality Group A, B or C, or the analysis of ruptures provided herein has shown no degradation in the plant's ability to either shutdown or mitigate an accident20)The assumptions and methods used in the analyses follow:3D-10Amendment No. 18, (04/01) a)Shutdown Cooling System (including portions of the LPSI system)The system operates only during shutdown and the operator monitors the cooldown rate toassure compliance with the Technical Specifications by observing flows, pressures and temperatures. In addition a number of flow, temperature and pressure instrumentation is provided to indicate and/or alarm abnormal conditions in the control room. Typically these include:1)FIC 3306 will indicate loss of flow and open FCV-33062)PI 3303 - will indicate loss of pressure on inlet to SDHX 3)TR 3351 - will indicate low temperature 4)PI 3307 - will indicate loss of pressure 5)PI 1102 - will alarm low level in pressurizer 6)Changing pump flow rate will increase The ECCS rooms and pipe tunnel communicate with other parts of the reactor auxiliarybuilding, i.e., it is not isolated. Certain equipment and structures in the ECCS room are available as heat sinks.With the above assumptions, the calculated pressure rise from a rupture of a line in the pipetunnel or ECCS room is below 1 psig. The temperature rise is approximately 50 F.In order to terminate blowdown following a rupture, the operator has to close the shutdowncooling line isolation valves located inside containment. With two valves in each line, a single active failure will not preclude isolation.b)Letdown LineThe assumptions used in calculating a pressure and temperature rise in the pipe tunnel for a letdown line rupture are:1)Guillotine break with 146.7 lbs/sec blowdown2)Closure of letdown line isolation valves by temperature element downstream ofregenerative heat exchanger3)Subsequent to the rupture, no credit is taken for temperature reduction through theregenerative heat exchanger.4)Blowdown is terminated in approximately 6 seconds.3D-11 5)No significant venting area from the pipe tunnelWith the above assumptions, the calculated pressure rise from a rupture of a line inthe pipe tunnel is below 1 psig. The temperature rises to approximately 175 F.c)Blowdown LineThe assumptions used in calculating a pressure and temperature rise in the pipe tunnel are similar to item (b) above with the following exceptions:1)Blowdown is at 900 psig and 532 F 2)Blowdown is terminated in approximately 15 seconds The results are less severe than those indicated in item (b) above.d)Auxiliary Steam SystemAuxiliary steam lines in the RAB are not routed near safety related equipment. Inaddition 12-AS-1 does not pass through any enclosed areas so that structurally significant pressure buildup is not possible. In any event blowdown is terminated automatically if there is an abnormal increase in ambient temperature.Outside of the RAB where 12-AS-1 passes approximately 12 feet from the controlroom north outside air intake, a conservative calculation assumed that 750 cfm of500F steam was drawn into the control room. In actuality the steam temperature isapproximately 350F in the line and the control room isolation valve would close wellbefore 750 cfm could be drawn in. Even for this conservative set of assumptions, thecontrol room temperature rose only 12F above ambient.Auxiliary steam line 3/4-AS-31 is not routed in any portion of the RAB except the decontamination area itself. The decontamination area by definition is a controlled atmosphere area exhausted through HEPA filters at a rate of approximately 1200 cfm, when decon is not in progress and 10,000 cfm when decon is in progress. A line rupture would trigger the heat sensors and initiate termination of steam blowdown in approximately 5 seconds. Assuming critical flow through the 3/4 inch line and a 1200 cfm atmosphere exhaust rate, calculations predict an insignificant pressure increase, i.e., less than 1 psig.21)All penetrations through the primary and secondary containment are designed toaccommodate pipe rupture loads. No significant loads from the rupture are transmitted to the secondary containment by virtue of the design of the penetration assemblies.3D-12 fg 0 c;') (I) <: 9 1-i 0 G1 H"'d * '"Tl r 0 Cit .... 0 . > r-'1J c: 0 Q =e mm :::0 ;!? 11" ,.. r z -.... G'> :i: c: .... 2:n .... 0 !':: '1J > z -< . .('( :> * "'.3 -'.S-a, --.......... <1 *so, 4( "< o. >p,, <o >>o. 2 k> Cl. 34-' c.o.JT \ "1\JC: t> o" 7 Cl4*PIPE WHIP RESTRA\NS * , N' +z
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- Appendix 3FTABLE 2D. O. STORAGE TANKMISSILE PENETRATION RESISTANCEMissile Missile Velocity (fps)2" x 4" x 10' Plank1614" x 12" x 12' Plank1021" Steel Rod1366" Sch 40 x 15' Pipe 9212 " Sch 40 x 15' Pipe 92Utility Pole 68TABLE 3MAIN STEAM AND FEEDWATER PIPINGMISSILE PENETRATION RESISTANCEMissile Missile Velocity (fps)2" x 4" x 10' Plank5834" x 12" x 12' Plank3621" Steel Rod4796" Sch 40 x 15' Pipe33312" Sch 40 x 15' Pipe336Utility Pole2483F-18 Appendix 3FTABLE 4CCW HEAT EXCHANGERMISSILE PENETRATION RESISTANCEMissileMissile Velocity (fps)2" x 4" x 10' Plank2654" x 12" x 12' Plank1711" Steel Rod2206" Sch 40 x 15' Pipe15412" Sch 40 x 15' Pipe154Utility Pole114TABLE 5CONDENSATE STORAGE TANK (1)MISSILE PENETRATION RESISTANCEMissileMissile Velocity (fps)2" x 4" x 10' Plank336 (2)4" x 12" x 12' Plank3361" Steel Rod1986" Sch 40 x 15' Pipe13012" Sch 40 x 15' Pipe130(1) Assumes 14 feet of water in tank.(2) Wooden missiles are stopped by water in tank (see reference 11).3F-19
this facility. The discussion supra demonstrates that separation provides an acceptable level ofprotection. It is fortuitous that these heat exchangers have considerable inherent capability to accommodate these missiles. This is demonstrated in the paragraphs that follow.6.3.1 OVERALL IMPACTIVE LOADING To assess the magnitude of loading to which the heat exchanger and its supports would be subjectedas a result of tornado missile impact an analysis was performed in which the spectrum of missiles evaluated for penetration effects were postulated to impact the heat exchanger at the maximum acceptable penetration velocities as given in Table 4. In addition, the impact of a 4000 lb automobile travelling at 73 fps was evaluated. The velocity of 73 fps was chosen for the automobile since this velocity was found by NRC to be an acceptable design value for this missile for a recent plant (Reference 13). Detailed missile studies and tornado analyses have been performed for St. Lucie. These establish the propriety of selecting various missile velocities. The value of 73 fps is assumed here for analysis of capability only.The equivalent static forces resulting from impact of the various missiles at the velocities assumedwere calculated using Reference 12, conservatively assuming no penetration (the velocities are those that would analytically penetrate) and using the natural period of vibration of the heat exchanger(0.088 sec). A ductility ratio (µ) of 10 was used although as discussed in Section 6.15 a value of µ=20could be justifiably used for the materials involved from a purely structural point of view.The resulting forces and accelerations are shown on Table 9 and compared with the equivalent statichorizontal seismic force and acceleration for which the heat exchanger were designed. As indicated, the missile impactive loads are less than the DBE seismic design load for which the heat exchanger and its supports must have been designed to maintain functional integrity. The results strongly suggest that the impactive loading effects of the missiles will not be the limiting damage mechanism. The large impactive loadings associated with the large missiles, particularly the automobile, indicate further analysis is warranted. However, it must be noted that these heat exchangers accommodate these types of missiles thru separation and redundancy. The additional studies are discussed below.6.3.2 SHELL AND PIPING RESPONSE ANALYSIS The dynamic response of the heat exchanger shell and interconnected ICW and CCW piping wasanalyzed using the PLAST computer code to determine resulting stresses and displacements following missile impact. The PLAST code is an elasto-plastic dynamic analysis method of evaluating the response of equivalent piping systems to impactive loading. The program has been approved for use in pipe rupture analysis and is described fully in Section 3.6.4.3.Figure 3F-11 shows the model used in the analysis. The heat exchanger is modelled as an equivalenthorizontal pipe and is represented by nodes 1 through 9 on Figure 3F-11. Each of the elements of the heat exchanger3F-21Amendment No. 19 (10/02) (9), (18), (26) and (33) are of length equal to the average shell radius and are selected in order todetermine the stresses at the junction of the nozzles and shell (nodes 2, 3, 7 and 8). The connected piping is similarly modelled out to the first seismic restraint on each piping run. The physical and material properties assumed for the heat exchanger were obtained from the manufacturer's data.A total of six loading cases were considered including both the 4000 lb auto travelling at 73 fps andthe 1500 lb utility pole travelling at its penetration velocity of 114 fps at each of three locations. The three locations chosen were (refer to Figure 3F-11):a)Impact at the center of the heat exchanger to its axis (node 5 in the -x direction). b)Impact at the extreme end of the heat exchanger normal to its axis (node 9 in the -x direction). c)Impact on the end of the heat exchanger along its axis (node 9 in the z direction). The cases considered represent the most severe spectrum of modes of impact. The missilesconsidered are (due to mass and size) the most limiting missiles for impactive effects. Each missile was modelled as an elasto-plastic spring using conservative values for equivalent spring constants.The 4000 lb auto was considered to have 2 stiffening channels of 1.5 square inch total cross-sectionand 100 inch in length. Then the elastic spring constant is:Since energy absorption occurs very early in the bumpers and chassis, a yield displacement of 0.001"and a strain hardening slope of the plastic region of the bilinear force deflection curve of 0.04 k were used.The auto was treated as a uni-axial elasto-plastic strain hardening spring, one end of which wasconnected to a node point on the heat exchanger. The auto mass was concentrated at the other end of the spring. This mass was given an initial velocity of 73 fps towards the heat exchanger and the system consisting of the piping, heat exchanger and auto was allowed to respond to the resulting impulse.The maximum kinetic energy imported by the auto to the system occurs just after the auto velocitytowards the heat exchanger reaches zero and begins to change duration. Peak stresses in the pipes connecting to the nozzles occur at about this time.For the utility pole (1.5" in diameter x 35' long weighing 1490 lbs) an equivalent spring was chosenwith the following physical properties;3F-22 Am. 7-7/88 E = 1.83 x 106 psip = 1830 psi (proportional limit)Then, the elastic spring constant is:The slope of the plastic region of the bilinear force-deflection curve isS = 0.05 KThe wood pole, on impact, releases part of its energy in splinters with 'captured' kinetic energy. Thisenergy loss is modeled as plastic behavior with an effective yield displacement of 0.415".The pole was treated as a uni-axial elasto-plastic spring with its mass concentrated at one end as inthe auto impact model and was given an initial velocity of 114 fps towards the heat exchanger.The resulting stresses and time of occurrence for the most highly stressed locations are given onTable 10. In all cases the stresses are within yield and less than the ASME III permitted stress of 27,000 psi for emergency loading conditions for Class 2 and 3 components.The results indicate that the heat exchanger and connected piping are capable of withstandingsubstantial impactive loading without loss of integrity and that the dynamic effects of impactive loading will not be the limiting damage mechanism when compared to penetration effects.6.3.3 LOCAL DEFORMATION ANALYSIS Having evaluated the overall support loading conditions and "far field" dynamic response effects ofmissile impact and found the results acceptable, the local or "near field" effects were evaluated to determine the degree of local deformation in the shell in the zone of impact. This analysis was performed to determine whether rupture of the shell could occur locally and to determine whether the degree of deformation could result in internal damage to heat exchanger tubing due to shell impingement. It should be noted that the velocities selected are those that analytically predict penetration. The discussion that follows demonstrates that the less rigorous approach utilized in the penetration analysis is conservative.The case considered involved impact of the 4000 lb automobile at 73 fps normal to the shell at thecenter of the heat exchanger, midway between the supports. This missile was considered to be the most severe with respect to local deformation and the point of impact was chosen since it is the most critical with respect to tube damage and is the point of impact where greatest local deformation would occur.3F-23 The analysis was performed using the ANSYS computer program (Reference 14). Triangular plasticshell elements were used to model the shell and three dimensional plastic isoparametric elements were used to model the automobile. The geometric models are shown on Figures 3F-12, 3F-13 and 3F-14. An initial velocity of 73 fps was given to the auto and it was allowed to impact upon the heat exchanger as shown in Figures 3F-13 and 3F-14. The time dependent analysis included the effects of large deformation and material non-linearities. There were a total of 282 elements and 1083 degrees of freedom included in the model.Structural damping was assumed to be 3 percent and appropriate factors were applied to the systemmass and stiffness materials to form the damping matrix for the entire structure.The analysis was carried out to a time of 0.01 seconds. At this time the missile velocity has beenreduced to 16.5 fps as shown in Figure 3F-15. At this point the velocity gradient is nearly linear and by extrapolation, a zero velocity condition would be expected at about 0.015 second.The results at 0.01 seconds indicate that the heat exchanger is still sound from a structural standpointeven though deformations have occurred in the zone of impact. Figure 3F-16 shows the radial displacement at various elements in the impact zone vs. time. At 0.01 seconds a maximum displacement of 4 inches has occurred. At 0.01 seconds the displacement curve is approaching an asymptotic value and a total maximum displacement of less than 4.5 inches would be expected. This maximum deflection occurs in the shell directly under the impact zone. This is a local effect which does out away from the impact zone. The peak generalized strain occurs in element 149 which is along the.border of the impact zone. At 0.01 seconds the strain in element 149 is 6.2 percent which compares to an ultimate strain of 22 percent for mild steel. According it is concluded that the shell will not rupture due to impact.Some yielding also occurs in the area of the supports. In this region the peak generalized strain at0.01 seconds is 0.94 percent, well below ultimate of 22 percent. All other areas remain elastic with stresses well below yield. Although it has been demonstrated that shell rupture will not occur, the internal tube arrangement is such that impingement of the tube bundle will occur in the impact area. For a 4.5 inch radial displacement, the maximum number of tubes which could be affected is 38 which is approximately 2 percent of the total number of tubes (1950).The tube damage resulting from shell impingement may in many cases be limited to bending orflattening without rupture due to the ductility of the tubes. Adequate heat transfer capability exists even if all of the 38 tubes affected by the missile strike are completely blocked. The ability of the heat exchanger to perform its heat removal function in the present of tube damage was further investigated and this evaluation is discussed in the following section.The foregoing analysis also indicates, based on energy absorption considerations, that the heatexchanger can withstand the 4000 lb automobile at a velocity of 30 fps without tube damage.3F-24 6.3.4OPERATION WITH DAMAGED TUBESThe following analysis was performed to demonstrate that it is possible to cooldown the plant with asingle, damaged CCW heat exchanger. It is not intended that this analysis be directly incorporated inplant procedures. Post-event operator actions in response to a tornado would be dictated by existingplant conditions and available equipment. Specific recovery actions beyond that provided in the plantEmergency Operating Procedures (EOPs) and the Emergency Plant (i.e., E-Plan), including operationwith one damaged CCW heat exchanger, are typically provided by the Technical Support Center Staff.In the event of missile impact on the heat exchanger which results in the tube damage, the resultwould be outleakage of water from the shell side (CCW System) to the tube side (ICW System) since there is normally a pressure differential of approximately 35 psi from the CCW side to the ICW side of the heat exchanger.Depending on the number of tubes ruptured and the availability of CCW makeup sources, theintelligence to identify and assess the magnitude of the leak will be provided by CCW surge tank level and makeup flaw instrumentation. Emergency CCW makeup water is available from the demineralized makeup water system and the city water tanks (1,000,000 gallons capacity) via a normally closed cross-connection from the fire protection system. (See Figure 9.2-2 for CCW System P & ID and Figure 9.2-5 for the makeup water system P & ID). Although these tanks and the fire pumps are not protected against tornado missiles and are not in the final analysis relied upon for demonstrating continued functioning of the CCW system following heat exchanger tube damage, a discussion of their effect on the action to be taken in response to the incident is pertinent. In addition, since the tanks and fire pumps are widely separated from the CCW area on site by over 500 feet (see Figure 1.2-2), there is a high probability that emergency makeup sources will be available following tornado missile impact at the CCW area since the tornado would have to cross the plant in a narrowly defined path in order to cause damage in both areas.The leakage flow rate out of the CCW system in the event of tube damage would be approximately 50gpm per tube assuming complete tube severance and double ended flow. The capacity of the emergency makeup line to the surge tank is approximately 600 gpm. Therefore, if the makeup source were available it would be possible to continue to maintain inventory in the CCW system for up to 28 hours with tube damage equivalent to 12 tubes completely ruptured using the water stored onsite. This time could be extended indefinitely since additional water would be available if the 12" supply line from the underground city water supply to the tanks on site remains intact and operable.In the event of tube damage concurrent with the absence of the normal or emergency makeup, theoperator would be alerted to the occurrence of a CCW anomaly by a low surge tank level alarm in the affected redundant subsystem (A or B).Due to the physical separation between redundant heat exchangers and the inherent capability of theheat exchangers to resist tornadic debris, it would be highly improbable that both heat exchangers would sustain internal damage as a result of the same event. (This is demonstrated quantitatively in Section 3.0 supra.) However, the capability to maintain the plant in a safe condition with internal heat exchanger tube damage has been evaluated assuming that only one heat exchanger (the damaged one) is available following occurrence of the tornado. The procedure that follows is independent of the number of tubes damaged.If the surge tank low level alarm does not clear within a short period, the operator would, from thecontrol room, isolate the non-essential3F-25Amendment No. 19 (10/02) header and stop the pump in the affected heat exchanger loop. The plant is designed such that thosesystems and components required to maintain the plant in a safe hot standby condition are notdependent on CCW system operation, e.g., the diesel generators are cooled by missile protectedclosed loop water-forced air heat exchangers; the auxiliary feedwater pumps are cooled from theirown discharge; and control room cooling is provided by air conditioners.The plant can be maintained at safe hot standby conditions without CCW system operation as long asthere is auxiliary feedwater available in the condensate storage tank to provide for decay-heat removal and cooldown to 325F. Assuming the Unit 1 condensate storage tank contains the minimum inventory permitted by plant Technical Specifications, (normally the level is maintained well above the tech spec limit) at the onset of the tornado incident, the operators will have 4 hours to effect the procedures required to place the damaged heat exchanger in service. If the condensate storage tank is full or near full, this time will be as long as 16 hours. The missile protected intertie with the Unit 2 condensate storage tank, when effected (see Section 6.10 herein), will extend these times even further.There will be substantial time, therefore to investigate the cause of the CCW anomaly, identify heatexchanger tube leakage as the cause and place the system in the requisite operating mode which simply requires balancing of shell and tube side pressures.As stated previously, the initial consequence of heat exchanger leakage will be a non-clearing surgetank low level alarm which will prompt the operator to isolate the non-essential header and to secure the affected pump. Once the pump is shut down, the level will rise again in the affected CCW loop due to inleakage of intake cooling water until the surge tank level is restored. A high surge tank level will alert the operator to restoration of level. Restart of the CCW pump will again cause the surge tank level to drop. A visual inspection of the heat exchanger shell will confirm that missile impact has occurred. Thus the necessary steps can then be taken to place the heat exchanger in the tornado contingency mode of operation.Since, with the CCW pump running, the pressure is normally higher on the CCW side of the heatexchanger, it is necessary to locally reposition the CCW pump discharge valve (or alternatively the CCW heat exchanger inlet valve I-SB-14160A or B and the ICW heat exchanger outlet valve TCV 4 A or B) to balance the pressure across the heat exchanger tubing. See Figure 9.2-l. The valves involved are located below grade.The system head and flow characteristics have been evaluated for operation in the above describedmode. Figure 3F-17 shows the head in the CCW heat exchanger versus flow for both the shell side (CCW system) and tube side (ICW system) for various conditions of valve throttling in either system. The optimum condition for operation in the emergency mode is with the heat exchanger pressure balanced at about 120 feet head. Under this condition the CCW flow would be approximately 75003F-26Amendment No. 17 (10/99) gpm and the ICW flow approximately 7000 gpm. These flows are more than adequate for long termdecay heat removal.To achieve a pressure balance at 120 feet head, the CCW pump discharge valve would be throttled tobetween 250 and 300 from the closed position. The ICW valve is normally in throttling service fornormal operations. It is placed in a normal operating position. However, it must be noted that it is not necessary to operate with pressure balanced precisely at 120 feet. This is the operating condition at which the surge tank level would be at its normal level. The system can operate satisfactorily over a wide range of CCW system level from a condition with the surge tank full and overflowing to one with the level considerably below the surge tank. Once in the operating mode the inertia of the system is such that rapid changes in CCW level will not occur. Thus the operation can allow the surge tank level to hunt between upper and lower limits.The initial system operation would involve re-positioning the throttle valves to their approximatepositions required to maintain 120 feet head at the heat exchanger. The CCW pump would then bere-started and the surge tank level observed by the operator in the control room. The control room operator would then direct the local valve operator to increase or decrease valve openings depending on whether the surge tank level were rising or falling. This communication can be achieved by direct visual or voice signal. Since there is a wide operating range under which decay heat removal can be assured, such a mode of control is considered entirely adequate for this highly unlikely event.Once the pressures are balanced or near the balance point, changes in surge tank level will occurslowly. The control room operator can also monitor CCW flow through the shutdown heat exchanger.The ability of the system pumps and valves to operate in the required mode has been evaluated. TheCCW pump characteristic curve is shown on Figure 3F-18. With the CCW head balanced at 120 feet, the pump will be operating very close to the design flow conditions for normal operation. The ICW pump characteristic curve is shown on Figure 3F-19 and with 120 feet at the CCW heat exchanger, the pump flow will be near the post-LOCA design flow conditions. Thus the pumps will not be subjected to extreme or unusual flow conditions.The CCW pump discharge valve is a 20" butterfly valve and the ICW heat exchanger outlet valve(TCV-14-4 A or B) is a 30" butterfly valve. These valves are provided by the same manufacturer and are suitable for throttling service. The limiting factor for valve throttling application is cavitation which can occur if the pressure drop and flow across the valve are excessive. Reference 15 provides a method for evaluating whether cavitation will occur for a given throttled condition of a butterfly valve. The method involves calculation of a cavitation constant which, if less than 1.0, would indicate seriouscavitation. With the 20" CCW valve throttled to 270, the required approximate closure position, thecalculated cavitation constant is 1.6, well above the point at which serious cavitation can occur. Thus the 20" CCW pump discharge valve is suitable for the throttling service envisioned by this procedure. The manufacturer's valve characteristic data for the 20" CCW valve is given in Table 11.3F-27Amendment No. 17 (10/99)
Appendix 3F TABLE 6 ICW or CCW PIPING(1) MISSILE PENETRATION RESISTANCE MissileMissile Velocity (fps) 2" x 4" x 10' Plank218 4" x 12" X 12' Plank138 1" Steel Rod180 6" Sch 40 x 15' Pipe124 12" Sch 40 x 15' Pipe 97 Utility Pole 66 (1) Based on 3/8" pipe thickness with correction made for pipe curvature (30" diameter)3F-36 Appendix 3FTABLE 7MAIN STEAM LINE BRANCH PIPING WALL THICKNESS ________________________________________________ Nominal Line Wall Size Thickness (in) 1"0.179 1 1/2"0.200 2 1/2"0.276 4"0.337 10"0.593 3F-37 Appendix 3F TABLE 8IMPACTIVE LOADING CAPABILITYOF PROTECTIVE STRUCTURE Maximum Tolerable Velocity (fps) MissileICW PumpCCW PumpAux F.W. HousingHousingArea Shield 4" x 12" Plank>1000>1000>1000 1" Steel Rod>1000>1000>1000 6" Sch 40 Pipe 995 712 592 12" Sch 40 Pipe 380 273 226 Utility Pole 189 134 141 4000 lb Automobile 69 50 75 3F-38 Appendix 3F TABLE 9 COMPARISON OF EQUIVALENT STATICLOADINGS ON CCW HEAT EXCHANGER AND ITS SUPPORTSKipsAcceleration(1) DBE Seismic Loading 155.20.8g Missile Impactive Loading(2) 2" x 4" Plank - 265 fps3.74 0.02g 4" x 12" Plank - 171 fps17.4 0.09g 1" Steel Rod - 220 fps0.89 0.005g 6" Sch 40 Pipe - 154 fps22.4 0.11g 12" Sch 40 Pipe - 154 fps62.7 0.32g Utility Pole - 114 fps87.0 0.45g 4000 lb Auto - 73 fps148.5 0.76g Notes: (1)Based on heat exchanger flooded weight of 194,000 lbs. (2)Except for the automobile, the velocities are those that would result in penetration (aspredicted by conservative analytical techniques). The value of 73 fps for the auto (there is nopenetration) was arbitrarily selected such that it complies with the staff's guidance for current facilities.3F-39
Appendix 3FTABLE 11CHARACTERISTICS OF CCW SYSTEM BUTTERFLY VALVESDegreesOpenK (see below)51562510 386015 93520 33725 14530 71.835 39.640 21.645 12.750 7.4255 4.4160 2.6465 1.5970 0.95275 0.62080 0.49685 0.46090 0.447Head Loss = (KV2)/(2g) V = Fluid Velocity (ft/sec) g = 32.2 ft/sec 23F-41 REFERENCES FOR APPENDIX 3F1)Thom, H.C.S., 1963, Tornado probabilities, Monthly Weather Review, Volume 91, p. 730-736.2)Howe, G.M., April 1974, Tornado oath sizes, Applied Meteorology, Volume 13, No. 3, p. 343-347.3)Wilson, J.W., and Morgan, G.M., October 1971, Long track tornadoes and their significance,Preprints, Seventh Conference on Severe Local Storms, p. 183-186.4)Fawbush, E.J., and Miller, R.C., April, 1954, The types of air masses in which Americantornadoes form, Bulletin of American Meteorological Society, Volume 35, No. 4, p. 145-165.5)Beebe, R.G., April 1958, Tornado proximity soundings, Bulletin of American MeteorologicalSociety, Volume 39, No. 4, p. 195-201.6)U.S.A.E.C., May 1974, Technical basis for interim regional tornado criteria, WASH-1300 (UC-11).7)Appendix 3E to St. Lucie Unit #2 PSAR (docket 50-389), Amendment 21, dated October 1,1974.8)Amirikian "Design of Protective Structure Structures," Report NP 37,16, Bureau of Yards andDocks, Dept. of Navy, August 1950.9)Gwaltney, "Missile Generation in Light Water Cooled Reactors," ORNLNSIC-22, March 1,1967.10)DR DeBoisblanc, et. al., "RID Report 1003," "The Effect of Small Equipment Missiles onIntegrity of Vital Piping," July 1971.11)"Design of Structures for Missile Impact," Bechtel Power Corporation, BTOP-9 Revision 1,July 1973.12)RA Williamson and RR Alvy, "Impact Effects of Fragments Striking Structural Elements" NP-6515, Holmes and Narren, 1957.13)Safety Evaluation Report for Greenwood Energy Center (dockets 50-452, 50-453), pages 3-8,July 17, 1974.14)Swanson Analysis Systems, Inc. Rev. 2, Update 155. 15)Power, July 1974, pp. 42-44.3F-42
- *
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- I{,,;.. ..*< ' ', . ;,, -.......-... AERIAL VIEW LOOKING SOUTH "* FIGURE 3F-5
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- Refer to drawing 8770-G-836 Sheet 1 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 MISSILE BARRIER INTAKE PUMP AREA FIGURE 3F-6 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-703 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 TORNADO MISSILE PROTECTION DIESEL GEN BLDG-M&R SH 1 FIGURE 3F-7b Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-705 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 TORNADO MISSILE PROTECTION DIESEL GEN BLDG-REINF SH 3 FIGURE 3F-7d Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-836 Sheet 4 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 MISSILE BARRIER AUX. F.W. PUMPS FIGURE 3F-9 Amendment No. 15 (1/97)
- *
- Refer to drawing 8770-G-836 Sheet 2 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 MISSILE BARRIER COMPONENT COOLING AREA FIGURE 3F-10 Amendment No. 15 (1/97)
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- 1 " / f'--, __ ,' l,J..C) J.I I li 1)11 f, "' -\ .. :1 :l--';,,,.1 v1J ).'t: 1.q11 I i ,,:,/{ ;.,;tt .. : :;, ,)f'-1.01 .:c.--3c-v:i.;1 ..i,:'.) l '11 '""" 2.' ,; + / ... t' 1'LAST MOM L 'I i . "-.. Vt' /t' C..C.W. HE.AT EXCHANGER AND INLf:IS .:!:. OUTL.E. T.S )(x 'tz NOT To SCALE. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 PLAST MODEL OF CCW HEAT EXCHANGER AND CONNECTED PIPING FIGURE JF-11
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT l FINITE ELEMENT OF CCW HEAT EXCHANGER SH ELL FIGURE 3F-12 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 FINITE ELEMENT ISOMETRIC MODEL OF CCW HEAT EXCHANGER AND AUTOMOBILE FIGURE 3F-13 * *
- V> w :::c: u 2 >-..J ..J < u j:: Cl: w > w u z < I-VI i5 2l'+ l'i?3 ll'.l ig;:o i'EiS: I + 2 I I..:i I 15 150 DISTANCE HORIZONTALLY (INCHES) FLORIDA POllER & LIGHT COMPANY ST. LUCIE PLANT UNIT I CCW HEAT EXCHANGER -AUTOMOBILE MODEL VERTICAL PLANE FIGURE 3F-14
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- C> IJ') U") N 0 0 0 = co 0 " 0 ..,0 C> I.I") 0 '<::t ---V'.I D z 0 u LU if) LU FLORIDA POWER & LIGHT COMrANY ST. LUCIE PLANT UNIT 1 VELOCITY TIME HISTORY OF FRONT OF AUTOMOBILE FIGURE 3F-15
! ! I / / / //' / ! I I f / I / I x 7 f . I ! I ; ; I I J I / / / / [J "-x 7 / l:H D"1 ,-H X'< -;; I, I I /I // /1 !/ I -I J I I // :" / / I I, /,/ / I // ;/ >-z < ll. a ... u_ I-% "'5...1 D::: 1:1.. UJ ... so o=> c. ...I < ii'..,, 0 ...J u. Q___ _ __ .Q_ __ ___J..._ ____ .o__ __ ___o ____ ___u ____ _o__ __ _o__ __ _u_ __ --l..,, LI' U-: C: l/ C lf' C: l!' :i r*i * ;tj f"i ' (S3HJNI) lN3W3JV1dSIO w z 0 u.N 01-1-U z< wa. u:. UJ-uz ..... <-...JV> UJ C.1-Q:: "'Z :::::> z:sw 1..::1 u:: ...JUJ <...J CiUJ <...1 D:::...1 UJ :i: "'
- c-* c::> c -:ii! Q. E c w 0 a: 3: g I I I LL g ij I I I I I c 0 8 s U) .... ... ... ("J.:I) J.'13H J.'1 0'13H FLORIDA *POWER & LIGHT COMPANY ST. LUCIE PLAMT UMIT 1 FLOW AND HEAD CHARACTERISTICS AT CCW HEAT EXCHANGER FOR TORNADO CONTINGENCY OPERATING MOOE FIGURE 3F-17
- > u z w 0 z -.... I-z Z-c( -z I-a: c( 0 a: I-:! 0 :::> * :::!: z ii: 0 0 0 a: c( 0 a: I-LI. w c( I-Cl. a: z 0 w -_, 0.. 0 < 0 Cl. :::!: a: 0 z * (' J.::U HOJ. dWnd
- 8 o. N .... 0 e o, 0 .... 0 8 co 'i a.. 0 0 w i t-< a: 3: 0 _, LL g FLORIDA *POWER & LIGHT COMPANY ST. LUCIE 1 COMPONENT COOLING WATER PUMPS PERFORMANCE CURVES FIGURE 3F-18 250 200.,... ............ __ -150 ...: u. -::c 100 0 .... a.. 60 a.. 0 0 4000 OPTIMUM TORNADO CONTINGENCY OPERATING POINT OPERATING POINT FOR SHUTDOWN OPERATING POINT FOR NORMAL OPERATION 8000 12,000 16,000 FLOWRATE (GPM) *
- 20,000 FLORIDA POWER & LIGHT COfM'ANY ST. LUCIE PLANT UNIT 1
- INTAKE COOLING WATER PUMPS PERFORMANCE CURVES FIGURE 3F-19
w G") I N * (M:)TUU 1'UIMS ........... ......,_ ....... ,... . _,,_,. !ENGINEERING DEPARTMENT I.!17R!iil!CIIOJI nnD HEAD CAU:ULATI<JIS mt flV'mCDIS<JI I:SIJJID Ml'Oll"I' NO. 7].91D-l nie pw11oet or thls nport. 18 to t.h* *tr.*H* 1n the 'l'1Pe I Md 'l'1Pe III fl.wad heildlt under a eocbinat.11111 of P4ie rvpt.UN md t.h.,...i -.neton leadlng u ep.citl*d br !bueo, Tho *tro** .. te b9 pnnntod VII on a prialry stresa l:iui.e const.atent With th* requ1..-nte of' AS.H! Section Ill fer faulted conditions, Th* bade equations ued &re drnloped in Tube Turn. l'!nt;i.neerinfi Depa.rta.it report no. 5.1.14 A, 1tt.ached, Soi1111 &dd!tional lqll&tions &re nHded for 'l'1Pe I hew, Thee* .,.. clneloped u needed in the tat or t.hls report, A.'.ISU!!PTIQNS 1. Load combinatiou to be conlid*red tor 'l'1Pe III .,.., A, e.ntung and .ll1al B, Bend.1ng and Trarutnn* (Shur) C, Ton1on and Tnnsnree (Shear) 2, illovahl.e *tl"llae** .,.. iuen u 11111t*rW )'ield at. p!'OHH pip. operating t.empenit.;11'9, '!'ha cpecit'ic numerical nl.1111 will be taken ll'lldl the 197l edit.ion of Sect.ion III. ), Th* 1t"HH in thl prcceH plpa tapeNd lnabe nHd 11ot IMI naluatAd, B.r Yir.u* ot a ta?'INd hub, additional thidcnH1 ia proYided "'1ich obvlou.sl7 reault1 in low.r prl.Ju.rr than th* attached pipl.n<<. 4. For frpe III hill&da, the loada &re resilt*d t,,. the oont.atn.nt nosll.e hub, S. For ':')-pt I heed*, the load* are re.t.sted u f'oll-t A. W&l and Ton!.on t,,. TM111111ona B. Shur and Bend:.ni t,,. t.vo *hear platH locat*d 7 Ctet apart. C, T!>e trunnione U'ill &111UJMd loaded at a point 4 in. outward fl"Oll t.h* nued head, t, For T7Pt I IHl&da, load cOlllbin&tiona &re u epecit1e* t,,. tbuoo, Intol'llllll -datec Decllllber 22, 1971. t-'iU
- m s 6 26 27 28 J6 )7 J8 ... -*,..._,,
- mxc DPPSICIW. WA rsm m m If.ADS 71993-1 o>> R* T' ... T +.s-t.,: t ... ., i)eJ' i>..., = O!>a. -'t,. J"-.4 ob
- V'*P* o.Z>. !: Htt 1>"'6 ' Soc.wfT o. R(in) ...QBilltl .),062 2.0Jl l0.7So .illnl BIWiM. .Sl9 2 ilOS GrII .),062 2,0Jl l0.7So .519 2 ll05 Gril .J ,Jl2 2.1S6 12, 750 1.148 ll,602 ) Al.82 F304 J.Jl2 2.lS6 10, 750 l,125 9.625 2 l/2 Al.82 1)04 .1315 , 9JS 6.625 ,IJ2 6.193 1 l/2 il82 FJOI. 6.625 J.812 16,ooo 1,250 ' i;2 Al.82 nor. .),812 16,000 1.250 14. 750 .) l/2 il82 F304 6.61.S J,812 16,000 l,250 14. J l/2 il82 P'JOI. 6.b2S J,812 16,000 14. J l/2 il82 F)QI, Proc. Pipe .,,!9(.F) S2C S2C 450 l20 40 10.7W "875 24,000
- 76S 2.) ,2.) s ) il82 F)CX, 5SCJ JJO J)O JJO JJO JOO JOO 100 JOO u 44 48 64 3.062 2,0Jl e.62S 1.500 l.25 6,625 ., '. 6.62S 10.7,0 5.87S 24.000 .us .628 .628 .765 (l,l 9J 2 Al.82 T.JOJ, S, m l l/2* Al82 T.JQI, s.m l l/2* , ' W2 "°' * !1tilllated du* to. incoqliete load v*nrus dilllens ion data
- 7 I :nt!YJ. i;. ()O!;i:c: 60l!J. SLC'n: 1!9691t t6S( <:11:'1 uru. Ot( .,9 'll'l11(J t"IJOl CUGl:llO PUV 8'1 ()O!;jz pwO'l *'l*'[CillooUI O'l llllCJ l>Wl1'11'lrdao:l ON .,, J1 ooien "91.T 096( 9ecrt 066 an 61..bt '19 rt ()O!;Z?; 6CltL SLC'n: t6S( (%9 Ml O"[( O'r ocmz snn: 69/,Jl ?:Ot6 S!;SS O'r( "[f!('f Olt 6( ocmz srm 69/,Jl ?:016 ssss 0'1( "(9£"1' Olt ac *'I OO'TIZ snn: 69/,Jl iot6 ssss O'I( tat"l Olt LC al ocmz srm 691.8 ioth ssss O'I( "C8t"1 Olt 9( e. OOWt 1.9( 'I()!; m '[ii"[ 9t't m S9
- 91.6:e S06S 'Tt(*J an S'l6?: Z"tZ a OSoc:Ji: 'r6'r('; l!ll:6'r 'rSCS UZt .,..,., a;: 9l: OO'WZ O'lf:il: 9VC9 t9'rt Z:Z( ?:Ch<: t9t 9 0'16'l: 9VC9 t9'rt tt( ?:£6?: t9t s (ICd) 4 (Ic.l)S (ISd)S (ISd)S (I!:.f)9?J_ (Is.I) 721..:t (ISd) 'Z!!..2. GI AO"['['f *.m,t + lrUS.ZJ. 1>11116 + llW.lJ. *uaa + rr UOJC.ZOJ. "CU'll.lJ. .JU'TI>UGg trtXY sau.xsra.ua ss:mts Hl'lfllYM SO'fO'l 'TYllllAicnn lfOd S3ss:ilLtS U!OQ CIYaH lll :i:lll *
- 11
- s> ' .. ... :-g 1 I !i 8 § !-0 t :;J '5 a l a § .. i i 0
- rp i;. !i .. .e ... g § ... s .... 9 B ! i .... 1 -;a ol .. I foe 1.-j !.f 1 Jo I §§§§!* .. '§.! ii' $ . .,, a § i: u. E .. .. . . "' 8 .t' '!i s :; a ... 9 ... ;j ' 0 .. f """ """ .... Ill .... i * ] ... ii * *' 0 t" . . ,J -*1 '5 i i "O ""., i :!': 8 8 8 i j;* Fm* 8 8 § a 0 a 'i 1 3 :i ::;, . :a ... .... ... .... 1t *l I :! ;j... ... ... i 0 .... :I ;!I 8 8 8 8 .. i 8 8 § a 0
- G_;ij l 3 3 3 !l Ji; :a ... .... ... .... it 3 ... i' .. I>. 0 .... +> ** '"'4 '3 .::f e la'! -oi ...
- Q "' ..., 00-=.
- fl. I .... t. ..., """ g :a :I i'" .!l .. * .... "" "" ...... i 3G-3 v-D(
- 1 i I I
- Tt1'E m catrA.IllHDIT ICllZU! IK1B S1'RJ:sSIS :massES Pat DillTil1IAL LOUIS Ml..OlUC S'nal:.'IS Dm'JISI'l'IIS 0 *'JO* .u:1.a1. Trana_ .... .lz + a.. Trana + !leis Trina+ Tor. .1.1.l.ow la <;"* !FBI2 °1iBlPSI2 SlJ'SI2 S'Jll!II2 S'PSI2 s 246 1781 4912 891 :mr:l'1 1781 2766 28100 6 246 1781 4912 8'l :tt1Z'1 1781 71M 29.100 26 1,)6 886 272 w lQZZ 886 l..UO 2IOO!il) 27 2D lJlJ !il)6 6S7 lJlJ 2;J26 28(XlO -{ a.a 11 Tl 142 .58 148 284 .i,Q:> lll!JOO r2 J6 247 lZ7'7 441. USS .JJ24 :J2"/7 'IR'M ZllilOO lu 37 247 J277 49'. USS JS24 J277 9298 21400 ,..., *le! J8 247 J277 49l USS )S24 3277 9298 21400 i J9 247 3277 494 USS )S24 3277 92911 21400 40 &.14 7202 U28 )6W. 71U6 96.58 22.500 r 41 us 2222 290 llll 2222 2802 22.500 44 llO OCXAJ'f.l"nmS tm: TO Llc.t <:Jr LOlD DAU 27.500 4.8 761 S64 1S22 282 1325 Jar.. )608 64 614 r.m 1.228 3601 7816 72l2 96S4 22.500 T.t.BLB 2 ;:I ! ,.. _J I 1 l -* l ........ iJ + + _.. N Tr !<J CQ:-l .... .-1 ! p !>> .. "' *@ '2-1 0 i "' "' '>* " i i ra i' .r l r:: I tJ I I ,m "'lo '"\2 ..., 0 c Ii + ... I 7ryf .G " : i t ... .... e ..... 1l ' . .... \\ _'r;;t .. en L.1-lo e ';:! " 1 a e: :... .... ,..
- w (j) I V'I t-166 * !ADI TW! ASl'.S ..
- LOAD DATA 1'0R T'fPE I Pn!ETRATIOO 7198)-1 AXIAL tQmlQJf. )41>,000 lb. -Lb l ,200,000 ID lb S,400,000 inlb l -191,:lOO 49,2lX>,OCO S,400,IXX> 2 80,000 28,900,000 47,400,000 ) 2)1>,<XXl -28,800,000 5,400,000 4 19,000 ro,ooo 2,61.C,000 lao,ooo l -86,COO 5,0l.G,000 1.20,000 2 A.SES --ro,coo 1.20,000 ) --66,000 2,640,000 2,520,000 4 ll*re 111;&.U! th* above COl:lbln*d pipe rupture and thermal *:q>ansicn load lat& vu ebt.ained frCAl the 11111mo or DecOlllber 22, 1971. IEVELOMllT OF ADDITIQIAL FOR T!'PE I Th* eq11at1c.ns d***loped ln Ntpert S.114 A for head bodJ' etre11H are atil lu.rectly applicable to T)'pe l heads. The nonle or hub etr .. a eq.i&Hons are lalao atW valid, However, some of the load input. tot.he nonle equationa ill required, Thia will be pertorned 1n the tolloving, In addition, the equatima for CClll?Utinii the lead* and 1Ubaequentl7 th* wld at.tachm nt. *t.re111ea for th* trumions vUl 'oe below, l. Nonle L<;w Par assumptions at.ated aboTll, th out.tl'lllOllt hubs ot the rlued head l"laiat the applied bendllll( lllOllMll1t. (M) a.nd tr.n.nne load (V), Ille to sr:-try, each rub l'Hht.e one half of th* bendir.g and elao ona half th* applied trananrH load. Therefore, ming th* eqlU.tiona of at.aUca, tie hub loads becOIDlll L
- 1.2 in, ,...p+fi-} vu*t!. 2L 2 Uei.ng Mil and V11 above, th* *treas int.*naitiH ln th* nonl* hube can be det.emlned Sn a abilar mnner to th* T:1Jle III heads, 2, '!'Jounnion Lo&da l'wr atated 11."°"8 ttie two trunnion& reaitt. th* u1al load \f) and t.orsiooel load (Mt), Therefore, th* loada t.ran111dtted to th' tl'UM1ons becomt1 PA* f' Pt* _m.._ O>z + 8 'a* 1fi"T+ Fta * . The load.a t.nwllit.t.ed t.o ihe t"111111<111 Vlllda beo011111 Vo* Pa Ho
- IJ'R 7198)-1 Since the trumion wall t.llicltn*H 111 greater than the total vald thrtat., the wld *I.re**** will gowrn, Therefore, only they will be nal.1111.ted bdov1 Vo z. ... 6 -'701Kh(Dl. + D.2) r;--He> V1> 'fTC Where I : 11R3t for thin ring D:i.
- l"U'E OD + .5b o2
- PIPE ID * ,5b '!"here tor\ .. 11 1
- 707h {l :t + { J} c .. D.i./2 Th* abow bending and shoar stress are eanbined to for111 a *Maldnlm Streu Intansityw (aa outlimd in Nport 5.ll.I. A) and tc the yield strengt.h at. operatlng tampant.ure rcr the lowr of head materilll er trur.nicn r.iaterial, STRESS Fal TYPE I llEAOO The head body st..re11s relllllte are tabulated 1n Table J 'lllill* the nosil* streaae11 are 11hOM1 in Table 4. The trunnion veld 11tre11eeu are ehoim belows A. MAil! SM (Pl 1111d P2) By inspection case ) 1a 11.llliting FA* O Ft
- 47 LOO 000 *
- 592,500 72 + 8 Fa.. 592, 500 lb V0 .. 592,500 lb Mo .. u )( -2,)70,000 ln. lb. .,.. :. -592 500 ,. ll2 PSI '""' ,'{O"/T>. x .b25 (20.)l + l6.l>25) SJ. I
- 11 X
- 707 X .1>25 { (20231 )' + {r126?J.) '/-. aJ.04 in.3 c ... 20,ll -10,16 2 *
(ISi!) Xs ll"lBVllO' m ; &'1£( oC6C 91.SS Z"IO[" SL98 (6(()"[ 9(.L('[ irm: (ISil)C ll.I!KUJrl S9SIUS XYK 9LC 9LC (I!id) *!l MJIC!m 3G-6 'IS'n: 'l!;'n: 9LlZ 'ISTI Zt0'1 n1" CS89 L9t t'ISdF2 OO:QJGB * * *
- * * , .* t-7'6 @!!!l..!!!!!5--**-* .-mut.-71983-1 71983-1 nrsrussI.ON It can be ... u, obe*rved t.h&t the etr .. e** calculated h-111 are ror t.he 111011t pll't. quite lmr 1n compar1Jlon to .UowablH, Thia condition results CJUI th* heada to produce Code acceptable streoa "' i ! i l valu** Cor nomal operatl.J\g loada ao wll u pi)ll rupture low. Under nonnal operatl.J\g conditiom th* dee1gnar must c011111der all stre11H1 ill th* head 1nelud1nr. the errecta or 1truct.ural and *treH conemtl'a1 en which are not required for faulted I a ! "' f'\ N 5! ; i 1i n H .. w i i Cl ..Cf"\f'\f'4f'4f'4f'4 I """ I ... *§U!I "§ *.; ;! f'\f'\::;! ::;!,.. ,... i § § § § § ... N"' PATEi Z9 /:E(j 1Z Wlll'l"l'D Bf I f,;[,/(f )nttt1v \llllllll s. Haben:ian I ... APl'llOVED Bft D t/,IJ,10.r{ ft f'\ ... ... ft f"I ... Kiehul V. 1'allr:me ... ft C'f ... ... ., ... ! 1 1 1 1 1 1 1 ... ... ... f'\ f'\ "' l"'I
- APPENDIX 3G2 REPORT NO. 71983-3 *
- 3G-8 w cp "' * ** llM_., * * .-. .. ......-* M.Ni I RN.& I ----+-----------i---I Dealg.-: Calcula:lona for Florida f-:>wer w Light Company Hwclllnson Islar:d Plant llti'ORT NO, 7196)-) Type llt Cor.t.alnmer.i Ptpln& Penetration Alaemblles P.O. No. NY-422264 Ir.c!* .. -:!".! are: Per.etratlon No. * " ,, " " " * .It * .. .. .. " .. .. .. " .. ff " "
- 5, Slowdown (SG-lA), I-2B-*1 6,. Slowdown (S3--1B), I-28-2 26, Letd:>wn Llr.e, I-ZCH-10.3 27, Charpng Llne, l-2CH-109 36, Safety lNe.n Loop J.A2, I-661-U.3 37, . " lo
- 1A1, I-66 l-112 *36, . " " " lBl, 1-661-111 39. .. .. II 182. I-6SI-110 t.O, Shutd:>wn CooUng, I-1061-422 41, Safety InJect:on Tank Test, I-251-479 t.1., P.eac!Or Coolanl P:.mp11 Bleed Off, l-3/I+ CH-12 64, Shut4>wn Cooling, I-1051-420 Prepare:! s:1: "vi, ;_;, Hac.i:rr.:an Bellows Engineering Cate: May e, 1972 .i!\ (Jct ll, ff1Z Appro,ed Sj*: 1n. v. ?no Date: May e, 1q72 M. V. t!c.lk:nua :-'.i;r. Bello-..,s Er.(.neerlng ..._.,-. 71983-3 TABLE OF Do 1. Summary of Report 3 2. Bellows Calcl.llattona I+ ). ProceH Pipe Thlckneaa Check 13 1+. Noszle Streseea Due to Pipe Rupture 15 5. Jet Force Calculations 18 6. Nozzle Streaaes Due llO Jet Forces 20 7, Penetration Seismic Anal7sts 25 8. Penetration As1embly Drawings w C') I ...... 0 ""* u;:)!!!,* .. !!!..N?.._ .. -REPORT NO. 71983*-3 . 1. S11111mary or P.eport
- As sutctant1ated by the calculations herein, all components in this group or penetration assemblies have been demonstra ted to comply with the design inrormation supplied. The only exception are the rlued heads which are evaluated as group us1nr, rour heads. This is treated in Report !lo. 71983-). In ari errort to eva:uate the penetration assemblies as a whole, we have consJdered the following major areas or concern: l. Pipe pressure Carrying integrity. 2. Pipe rupture integrity 3. 11-!llows lntegrity "* Jet Impingement integrity 5. Sel:lm1c ear:ib111 tf. In evaluating the we have utilized the allowable rrom USAS 831.7-1969 ror Class II piping ror all but raulted conditions where we have used minimum pipe 7leld at process p1pP. operatlng temperature rrom ASME :ection III (1971} due to a lack or adequate guidance rrom 831.7
- LOUllYIU.I lo &DITllCllT ltlfl'OllT NO. 719S3-3 z. Pe1'0ws !:'es\*'"'."! Cel,.ull'<'1o:ia f ** The i:::e-.: ui >Us ae.::llon la."; derr.onatrate o; t'-e structural adec;ua:-v the seco:-::!1::-!f seal bellows for the dHlp eondldons 1tat< ! ln Paragraph 8 be:Jo.... rr.11, coniisur11:;:i;: of the tello"'* I; deplcied o:: Drawlr.gs 71983 03.1 and 71983 CJ.2 elto:hed. B. Dealm Cor..titlona: c. .o. Etessure
- t 5 PSIG Te:nperature
- Z61t *F Material
- A21,0 TP J16L Movemen*ta: Axlal E:l.ler.slo:i Lateral 1. 1.'!::3AS BJ1. ?-1'1.;_1 It Cr :les z.; In. .t t .5 IZ (Ot'C C.*.:h!S .5 ! .125 z. St.ar.dards of l!-.e Jolr.t l-'.:inufacb.trers' Aaan. * "'Inc. Third Edi::on flellow& Hoop Stress: 'The bellows ho::-p or clr.::u:::!era!'ll!d meir.brane stres* Is deterr.:lr.ed by uat* of the Eulow Forir.ala ll::>r atralght pipe* suitably r.:od:.\eJ to accoi....,: ror* a convoluted :llstrlbu!\o:i oi rr.e:.al. A.a a;:plled ID a l:ellowa i.:.e equation takes the follow\ng forr:i: lcl+w)'I* Pd -$ __ ... ;.._...--..
- n, (. 571 ,. ) tll WHERE: Sb
- Stress \ t"SI) p * 'Deelgn Press-.:re ( c:I
- Bellows Root Ola:::eter (ln. I ., t. ** Bellows Wall Thk-1 Plr (tn.I v
- Convol"-U:>n Helg::t (In.) q
- Convolutlo:t P\tch Un .* ) r:p
- No, Eellows Plies t + f d
- I I-' I-' * @!!'!*.!!!..NL.-. _,,,_., lllKl'ORT NO. 7198.)-J E. Meridional Bending Pressure Stress The bellows meridional stress (stress tending to crush or collapse the convolutions In an axial direction) Is determined by the fellowing theoretical formula, with empirical tions to :orrelate with test data. Sn Pwz
- 10 2/J t2 np F. Bellows Capacity: The movement capability of a bellows :s evaluated on an "equivalent axial movement" per convolutbn basis. For the secondary seal bellows, the allowable movements have been limited to the following: Allo"'1able compressive movement (e comp.) * .eS'(Q/2 -t np) Allo11<able extentlon movement (eext,) * .,.3(Q/2 -t npl The allowable for compression Is based primarily on cal llmlt.!.tlon from the geometry. For extension, the tion Is based on empirical data to prevent dimpling of the convolution crest and to prevent excessively high stresses that would slgnlftcantly reduce cyclic life, The axial and late:-a.l movements tabulated In Paragraph B a.re converted to equivalent axial movement per convolution by the following formulas: (Ref.: Standards of the Expansl.on Joint Ma.:iufacturera* Association): ex x -* i'i Axial Traverse per Conv, {
- Is compressive) JDy Equivalent Axial Traverse 8y
- n(L-1')
- per Conv, (caused by Lateral Movement) e
- Y.pc
- Axial Traverse due to ft-ecompresslon pc n ( + Is compressive) eext * -ex -ep= + ey
- Total Axial Ext. per Convolution * -* @!!'!.* .. !!!!i-".:r'-** -1111tl'01rT NC. 719e J-3 WHERE: :X *Axial YJovement (In.) y
- Lateral Movement (In,) Xpc * &eco:npresslon (In,) D
- Conv, Crest Dia,* d+2w (In,) n
- No, Convolutions L
- Convoluted Length * -Xpc (in,) G, Bellows Spring Rates: The bellows axial spring rate !las been proven experimentally to be predlcted by the follcwing equation: .a fx *2n:i:(d+w) E "r/Cu WHERE: fx
- Bellows axial spring force per convolution (L':>/In.) -6 E
- Modulus of Elasticity x10 (PS!)
- 27.4 ! *
- Thickness to height ratio. Cu
- Spring rat'! denominator -an empirical facto:-which Is a !unction of convolution pitch, height and dia.:::eter np
- Number of bellows plies. The axial spring rate of the total jolr,t can be expressed as: WHERE: *kx
- Fx
- fx ex ex
- 1L n x
- 1 In. Therefore by substitution we obtain: kx. L rl (Lb/In,) The lateral spring rate can be expressed as: k *ll y y WHERE F
- ftQ.a y 2L e * .....1.£L.. y n(L-,:) y
- l In.
? .... N tlll:l'OtrTNG. 71983-3 Therebre b7 eubatltutlon we obtain:
- ky * .A...Q... 2L kJ * ..1.JL.12! 2Ln lL -4') 30 n (L-.1) (t.b*fin.) H. Cn;Jlc Lift Determlll1!:19n: It U. beea pro*en .....,,. ..._... eln te*lh!C that the CJcftC Ufe of a U-ebeped Hiio** caa be predicted bJ t111e of the lollowlnC formula: (, s.,)4 N:: 4 s.* WN'q: t'AIUJUllTEI> Nfll/tll/J£11 Ill" rted'S
- IA1'/111Art IY MttM/1$ 1419Tcl11U (fSI) P. ).& = (14 :6,f '15"t£e./aJ:: nAllEU& STW"SS Ml:AHlTEK. (1'$1) ... Sp ="I.!' c:'P/(1t P/11rlh:'h, :Plf£s5JJl!E PllRllN£n::.e s.,': s;.+.sp: T?J1111. STrESS MRA1-1E"11!f ("31) p I It. s, .. f: /2 L :r !!E.! = Ti/l(l'PESS 7l) Rll1'10 w _, E: 27..tf = 1101>11.us OF' x10 (!'SI) e.-= -= En:ter111c-neAv*1l5E" ffe Ct)NV()l.tJTlbAJ (1#1.) * 'IUll YUINS _..__.......__ -**-.,._NG. 719e.3-.3 llJ
- CONIJOl.rJ1'11AI 11£"11T (11/.) Sc
- 17S'DO ,,,.,
- c.aD ,tlLtowAl&.f' S,,
- IH06 PSI * #fff1' 1Nt.Oa11t*tC *Etc..ellllt .Sfte'$S QH ,_ tr #*
- E#'.uwE' QI/ti. Ill {111) t: OJI' llf'W.S Pl.IE$ (J * * *T/Clll.I> snrnc l!lllb* * **1 e c.
- fSO i = Eus11c 111w,esc Wk coN11Dlt1'r10AJ ('11J*) For purpoaes of cyclic llfe determination, the non-cycllc presslve movement tnetalled st the factory la not considered as part of the total eq11\valent a.xlal movement per convolution. Therefore, we obtain: *eye * -*x + e7 I, RESULTS: 1. The static capacity evaluation of the various secondary seal bellows from both a pressure stress and movement atar.dpol::.t Is tabulated In Table 1, page !Q.. The allowable stress has bee:: taken from 831, 7 Appendix A, Table A,8 while the allowable bellows equivalent axial traverse Is computed per Paragraph F of Llils report. For completeness pertlr.ent bellows dimensions have also been cluded In the table. 2, Computation of bellows axial and lateral spring rates are Included In Table 2 *page 11* ). Computation of bellows cyclic life Iii Included In Table 3 page ll... Since the 600 cycle design mov*ements ( N1) exceed 7000 calculated cycles It was unnecessary to calculate the life of the lesser condition ( N2),
- Vol G') I w * @!P!* .. !!!,!!L....., * -*.111111'11111!'1 Ra:l'ORT NO. 71983-.J J, f;is-::ussl*rn: As can be easily observed In the results presented In Tables 1 and 3, all bellows have more than ade-quate capacity to meet design objectives, Finally, In summary It Is worth mentioning 11 few words relatlve to the formulas used In bellows design at Tube Turne. Due to the vast com-plexity In a rigorous exacting analytic approach to the problem which still must be correlated with test data to be use:Ul, we have adopted the procedure of using empirical r::oCiflcations lo a somewhat more basic theoretical approach, J: hese empirical correlations have been obtained through hundreds of tests on ac:Ual produ:tlon bellows. The net result Is a procedure with good experimental correlation at a large expenditure ol effort. For this reason we have treated this lnforrr.atlon In a proprle:ary manner, for review In our offices. However, It Is avallable * -!J) ..l ::> !J) w Cl! >< H u u u H <t: Ul 0 ...J ..J w ID "": {,.., ..... ; ,.'i£ ... ,. ..... I;) .... ....... <:) ,.... " \!:!' "" l\i () -:. "' ") '-"" ,... ;:;; ""' ,... 0-. 'ti ... ..... t I.) * .. !!!!!-_,,_,. Ra:l'ORT NO. 71983-3 It ... ...... ...... <:itl fV'I ": ..... °"' :l I'!\ Q t:: ..... ' "": ..... . t; <:(I " "": 't t. i::: t ..... .... ... ...... -,* .. r 1' 1* ,. I () <) 0 0 ... .. ,. "'l CY) "' ;::! " 0 Ill'\ ..J ... " ...... ...... ...... "' ..... .:i < t-< "' \\ "' "' ..... ...... ' ..... () (} 0 '-l "' 0 <:\ -:. "-' " .... .....: ..:. 0 "" "' "" -...: ... ..... -..: ...,: ..... .... <:> 0-°'" t'4 .... ..... .... ..... 0-C). <:.-) ()-.. "" s .... ..... ""' " " " '? """' f\!
7!-Df AXI.-\.L AND LATER.->,.L BELLOWS SR":IN3 RATES *
- Pew a t: .A) r rz. ' c L J-x. #,. NO. (1/'I.) (1#) (IN.) (tM) (11.J.) "' (..,.) (ii') (iJJ/1/'i) 14/11'J.) t: l/..r.1<11.tw l1.m1 1.0 I 1s l2.11sl z.1SI 2.ssl 1100 112.'111.,.,.11 410 I 9'0 6 11;:191-"JI l,1.1.731 1.0 I 15" llS".111.a.Jt l/.JZ>I /.o I 16 27 IJ.1.111 .a!l l/.3751 /.0 I IS" 13'*J11/'1-21 *b.S'D1 ,?.Cl / . .5"'1 /2. 40,IA .?.l> l*S1 1.3 .. us1r.S's-I 710011.?.,3114'.'-'1 4'7C I 960 1900 "'9D I II/JO 2.zS"l z.s.51 1100 I 12.'1IH.9JI 410 I 2.so 12..591 t'5.t.J IJ7.1JI 1200 I .J4DO Z.SD Z.6'1 17.13 !JtJ >ID" 41 .0$11/.31511.0 I 'rl I 400 I 'so 44 '1.941.o.11 1..11.si 1.0 14 r.2.s-lz.41 S'fitJO 11." 11.911 .t/l'O I 'Zo TABLE 2 SEAL BELLOWS CYCLIC I..IFE RESULTS PEN I d I t I w I 1 NO. (1/11.) (11'/.) (1N.) (IN.) n .. pt. e,. ..... ,;;,e I v I 1X**2'*31y::*1'1e I .L' /1AJ) c II (IN-) (1N.) t't#) (P.111 s'* I se' I Al I N (,.,, l'l"Jlf) _, ..,J, S" 1J.1'i I .0J1 lt.J7s1 /. o I 1S"l2.37.51.?.s.>1-.15JI. 1n>I. 3/JI 1111'0 10 I 81fDl11#0d 1.t>G I 1J.1<tl.csi 11..nsl 1.0 I 1.5"' 2(. I 15.191. eu-1 lt.1151 1. o I 16 Z7 I 1J'.t'1 l.tu1 I 1.J75 I 1. o I 15' 3'*391 n24 .oSol ?.o I ;. s-112. z.o /.S-/.3 -41 11194 l.AJI I 1.3ZS1 1.0 I 15 44 I 9. 94 I. AJI I / . .17.S'I /. 0 I 14 r.dl-'* 3111 tn.n '10 I lllHO 2.'D 1-.H*I. /JVI. JQJ I 8+f1 1.,0 I 8f:JO 2.s5 l-.1.f.3l.1,ol..1IJI s7t&I 9D I 440 2.s<Jl-./9Zl...?.ffl..f.1'l'.V-OI ao f 9#0 l-:1771.VBI. -?ci *01100 , .,,_ ?.471-.1.0l ./<fB'I . 111116180 ID lil<MC .?.41 -1"4 .1471. J// I &90 I ao I 8710 'l'ABLZ::.: .3 I J ;:! b '° e I I J ...;J f!li. "" I "" *@ l!i :-II l1il -@ l!i 11 w ? .... VI -* * @!!!!.' .. !!!!!!....--1,-* -Rl:l'Oln' NO. 7198 3-3 3, Ftocess Wall Thickness Cileck A. The process pipe minimum wall requirements for design pressure and temperature as specified on the "Ebasco" drawings has been computed per the following from B31.7:
- I:..Q__ t,.. 2 csE + .i.P> The SE values In the above equation were taken from 831. 7 J'..ppendlx A, Table AB. The required nominal thickness was taken as: trJOM
- t,,./ .875 8, 1. USAS B31.7 -1969 2, Ebasco Dwg, No. 8770-6-213 C, f'.esul\s: The results of this evaluation are presented In 'I able 4 below. tlote that all process pipe possess sufftclent wall thickness to resist Internal pressure. * @TUii TURNS .. ...._ .................... -1, l:8ftVClfY
- Rll:l'OlllT NO. 7195)-3 FROCESS FIFE REQUIRED WALL THICKNESS PcrJ p T 1-wrl :SE Do CAL'" /(Cf/] p,T. fM t_ "'>if" /J(J. (psro) 'f*F) (PSI) 6"*) fol.} (111.) (11.) s qst' 5SO "101..1,tB ?.375' .:J7" .087 6 SS'O 11/Dl (,/! 8 l.5'000 ?..375' .()7(, .oB7 .218 2(, l'IB! IWZTP304 14450 l-37£ . /flZ . Zl9 .344 27 l735 /IJIZ,,, 3()4-Uo>e:> 2.315"' ./90 .Zl7 .344 3t-3'f 248( r.s-o 11112TP36tf /-f 300 &.r.25' .5'40 .6/7
- 711 300 3S"() 1'1JIZTP3v4 1n.!"o 10.1.r ./()!' .120 .2!0 ""' Joo 3$'0 AJ!ZT/>JOf 15250 ?.31t' .01J3 .6Z7 ./Sf 44 2-18( l!iO llJl2TPJIJ4 "0)0 /.ot' .()7'1 .otf .21f TABLE 4
? ..... °' * @TUIE TURNS ......... ....., ............ . -**-lttrl'Oln' NO. 71983-3 1+. Nozzle ( 8'.111rd Pipe I Stresses Due to Pipe Ruoture: A. General: For purposes of analysis, consider the following *tree body diagram" of a typical penetration assembly. It wUI be assumed that the bellows has little If any effect to this lar problem. ......,.. Ell1> NIJ. tfSSY ,.,... .. -tt----/: ('ollT* '/ESSltL "'l.. Nf"llD -* L In the above M, F, V C.. Mt each represent a combined effect of thermal expansion and pipe rupture loading, This Is arrived at ty assuming one proce11'! hub :o be undergoing Its normal piping flexltlllty thermal exp!lnslon reactions while the other hub Is loaded pipe rupture reactions, For conservatism these are assume/! additive, For analysis purposes, the following load combinations will be considered: 1, Bending and Lateral ( M, V) 2, Torsion and Lateral (Mt, V) 3, Bending and P.xlal (M, F)
- By Inspection of the loads and dimensions lt becomes obvious that the maximum stress Intensity occurs at the end of the tration assembly and Is due to either Case 1 or 2 above, To determine the stressu the following formulas will be used: CASE I fJ;* ¥ v; 2 !:!L (* Ci;., VL/i-$::z1tvtr r ... = "i YM> 11/ f.rom Inspection ot tt.e above Case 2 can only govern where *'1 >!JI, @!!!.l..!!!!!L ..... -'* llllftRlllY *
- 1ttr_,. NO. 719S3-3 B.
References:
l, USAS BJl.7 -1969 2. Ebasco lnrormal memo dated December 22, 1971 L.J. Sas to w. S, Haberman. 3. "Formulas for Stress and St!'aln" by R, J. Roark
- C, Reaults: The results of the above computatlons are presented 1;-:r.:t;le 5 attached ( p, 17 )
- Note that all atreues are well within the allowables also tabulated. These allowal:les are based on minlmum yield at process pipe operating te:r.perature as speclfted on Ebasco Drawing No, 8770-G-213, The wse of pipe yield as an allowable la based on an assur.:ed faulted condition and consistent with Section III Is Invoked due to lack of deftnlttve allow ables fer E ,3:. 7 Class II piping under faulted conditions. It should l:e ::c:ec'!, however, tha: a 1, SS allowable for Class II piping wou!d l::e met In this cue anyway. *
- * * @!!!' .. !!!!..N!....-* -'* DllllCll:1' _,,,_ * -14 RIEl'OftT NO. 7198 J-.3 ltl:l'OllT NO. 719!!'.:-3 0 0 0 Ii) 5. Lateral Jet Force Calculation Q 0 0 <41, 0 c:i. 0 <;) 0 A, General: The jet forces for a long\tudlnal break are deter-...... ,.. " ,... .... .. '3-mined per the 11Ebasco 11 spec, as !ollo1'1s: "' N Ill "' t=j "" Fj
- KPA ....., () c 0 0 g 0 0 \/! (} t: () WHERE: FJ *Jet Force (Lb,) " ., .... "I' 0) K
- Phase Factor (pg, 9 of spec,) .... ..... --i (I p
- Operating pressure "' g A
- Pipe Flow Area (In. ) ... . ,. * . f' "' : Per the "Ebasco" Specification, . " Co\ "' ;.:; 0 § 0 § Break Length { 1 ) z 2 x l?\pe Dia. { D) ::: <> a Break Area
- A4 "' .. () :, ...... <31) r11' .... ...... .. Ii) .. Therefore, Break Width (cl *
- A4/J. ....., N' ""' ,, '""' 0 0 i 0 0 0 ....t !* 8 0 g 'O 0 UJ ). 't M B,
References:
"" 1, Ebasco Speclflcattcn No. FL0-877C, 124 !.. ..... .,. -e 0-"' ,. ' .... ... 2, Ebasco Dwg, No, 8770-G-21.3 ... () (:) () 0 " 0 8 "' 0 ... e
- 1 Ill C, The results of this computation using operating ,,J r-. <:> " " ...I pressures and plplr.g dimensions from Ebasco Dwg. No. ;; ... "' I -..... "' ID 8770-G-21.3 are presented In Table 6 below
- The forces ..... 0 0 8 g <( w!ll be used In the subsequent section, .... () <l ..... 'I:) & I t< " .. "" ',1 ... ... tiJ s 0 .. r;i () El g 'i] () () "' l"j I re " <> .... <:> " ... ,, --,:;; ,_ N ...... N .... -.l ...I t.... t "' "' Cl() 'Jl t " c:i ..... .,. ...... ....., V) "" 0--.., \_, N ...I t "' ao -.::; :; 't 'f:. "" t::-.... 'j ...... a i2 ,,.. 0 & z "iJ ... () ' "' c:. ;:::. -.... " -..:. ,,... g tl \,. \A <i ... ... -. .. '::! ... "' " 0-"'.§ "' ... .... ""
"'
- I .. c:: "' CID 'ii.= * "' * ... i; ... id .. .a r--c:Z f '5 I :s =
- 15. 5 .. .M :s g: * =.c:: 0 * : -r-;n .s * .. u "ti -s u &i: ... -'-: .D
- 11"tl ... 0 '3"' l o.! s .. 3 0 _j_ Q. ... ........ "' * * ).. I *iii .. ... i .... . ... c::-.. I } ti c:: '°
- 0 lo
- 0 I u . ..._ f r:: II ; -; i ,,. _, '.,,' ... =::a t " .a "' .. \S .. t ... t< l.(" .a
- Iii e
- r:: ... ii '4 "ti,;! e .... 53 't 0 -; ., .. Q .. .. r:: * .. "' ..., r:: ll"tl.fi"" i "' l8 II C @} BE..,:: .a "3 J 'It .. .c:: * \ l:.' l; "' "" .. -e o-; .si ' ta0,S !!I .. .! :s J * -* ... E :! .c .s * "ti. ;c.!li.r:: c .....
- o I .. =.Jr :: 0"' : t< c:: ** E 8 * ,.8 -=1L..J I *-=r; II 5 * ... "tl"tl b"'"' 0 \;-= l Q,, "'c:: c: CJ); b = 5 "'"' c * .a ...: ! l! Cz4-. ij. 0 .... ::: g. .. z i: '() 0
- z< .D '* * "' I 9 3'"l8V.L 0--... r--i i /'FP. IH' 01*z (.I,,. QS/ *?1/t. "'l' .,so*/ t+ I ..,St8 "<tt.* ..SL..,, L£* ;J!!Ufh'? osr ;.sr S"'t? ,, ,11 1'1't: ,S'"IZ L.6. ;, a.J.lf/(7/ OOE' o.sz.-""" 6S .. I szr1 "'. ;,attfllY .snz 11l' /,IL' ft.1'7 'ME II.: iii 080.G .SI...;. Lb' :8.Lll"'/ ..S£EZ 01-Z'Z vK* Lt @} (},t,; Jt*; ;,;,1.W'? OO'U Ol>Zt? +K' 't: Ofr11 "ZZ?' E1" ,. .. .Slid $tl"* ' 0#1/ "Z7,. ,S'tf.*J. £1' 1-/ll:iU!T ..$88 CS6'-Z e1z* .Sl.'i z 5 , .. .,, \'ft'.) (1'1') tl#l::tH C)1W) (,..Nl] ('Jtl) (*fi/) 'a-I !:J ':;) .., " f*".:I d td :/-iz. rod S3:::>?JO.i'I .I.:3 l'
- I 3G-18 w (j) I f--' \0 * @!!!!.* .. * -1._, llll!l'OllT MO, 71963-.3 2, The second area of concern le the hoop streH caused In the nozzle or guard pipe due to the jet force striking a finite area. The area ls developed as follows consistent with the "Ebll.Sco" spec: 1-ac .. _..--r TJ,,/z. 1o*rvP. Jo.c...I,.. D.h, DI--'*___,, *. T'VP 1---L --f C,,, a C + 2 TAAi 10° (1:iel/% -: c -t-{1>.1-Do) '""' 10* l11 = J... + (A,-D.) TRN 10° C11 I..,.; , -= [G + (o,..-b.) 71MJ ,t:;J f L1-(A*l>.) 1#110*] P= £.. V't:: ;l-i; WHERE: ON
- Nozzle I.D. (In,) 00
- Pr:icess Plp11 (O.D.) t
- Nozzle Thickness (In,) Fj
- Jet Force (Lb.) c
- Slot Width (In.) L *Slot Length (In.) *
- r.:i:)YUll MlNS .-**-ltlll'OllT NO. 71SS.3-3 B.
References:
1, Ebasco Spec, No, FL0-6770.124 2, "Formulas for Stress and Strain* by R, J. Roark. .3. USAS 831.7-1969 C. Results: The tabulated results of the above computa.tior.s are presented In Tables 7 and 6 attached (pages 23 11.r.d 2;.). This condition has been assumed as faulted therel::y just!=>1r.g a speclaed minimum yield allowable stress, See Parag:-aph 4C for discussion on use of this allowable. OZ-DE * .... ... .:& "" .... &i .... -...... IS' .... °' s ---"'\ " !;:.-; ::?1 z C> 0 0 ....... 0 0 N N !:! .... ....... C N c; "" ,..,. \.)' ...:.., t'l *@ 1-3 ... , \ ..,, Cl ). ..... 0 l"l Cl (> (; ..... c.. z I* l' I;) -l"I ':'--:--':"'-! ..... t'J 0 , ... t'l t' ,... N :0 ,.., " .... ... :-Ii z -.J "! G> ;::, ... '.,, 2 CJ) 1*1 -" 8 "' 0 *-"" w 4) "' "' -...;: FA I N w .... c... r-. en .... en ... Qt 'Q 0 C) d 0 <> i:::. FA I ..., "" N !::S I ...., SJ\ ij' .... i c;;; J 0 0 0 >.."!.. -..:i ... .... I ....
- NOZZLE HOOP STRESSES DUE TO JET FORCES ' 'FDl D.i t u. G L. Fi,, II QC /¥). {, .. , ,, />J. \ ( ,...,,, l'11J.\ {11/.\ It.a (111*) l'f'51\ (psi\ .59.:J 2. 375' 11.40 .J/I. / JHO 21/fJO ' .5'3 2. 37$" 475' /l.40 10* I 13,0 27/DO z' /Q.Jjl)O 1.u.s:-?. ns-' . .flJ. 475" "1170 11*8 lflO lJ/00 *@ 27 8.500 /. rz.5" ?375' * 'f7Z,, ,57)80 9.04' 2.,10 3/.,00 Ui X*J't /3.5'1.4
- 1.218 t.t,2(' /.5'1 /j.1$' 40.7 _,SIO $./DO* :-11 111 -1(),t.1 21.00 /.0 /d.15b 38"'-'21.i 2-fOOO 131. ;ozo ,JJllO ,,,., l-'25 .soo ;?.37) ,7-z,'1 815' '/.'fo 7oS' .so;r;o 4* 5.lt'/ .71? /. 05" * /-II 2.10 7.Z. I Jf,./ 2.4, 3/'f(X) I J -..:i ... Tl\BLE: e I ....
- l.i G"l I N I-' * . -,. -llEl'ORT NO, 7198)-.J 7, Seismic Anahsls A, Consider the following free body diagram of a ty;>lcal Type III penetration less the bellows: L11 B ortt UY .. L, Wal
- VESSEL L CONT NOPfl..C ( Gr.t...,A'l> -r 'io .. The general bending monent equation can be written as follows: Har= !;. x+ "1r ,. .,H*'RE: F;.
- Fp H='s +Fe_ + -G) l1r.., Hp-"18 -F;.L, *.ta/..* -(&) F, = V.8. £ P1P1A16 1.fJllJD T11ERMA1. E'IP LMD (t8. Mfl' ** .. .. 611.u . F. c = Ta, r WCl.i> (ND SE"IS/1/1/C. iMD (lB;) F.. = Ta. .. Flf/£D /llil9/) J..c4'i) (1.8.) fS z 8FllbWS END (1.B.) #a= .. (111. 1.s.) W,, = 1a. * ('oAJf. AID 'I-lie .J£ASMIC. IOI}/) ('8)41.
- Report No. 7198.3-.3 p. 26 of .34 We. = llJEt.l> """ r. ( t.B) w',, = f:L. HT>. wt.T Ct.a) W., = PO?lU ?1J't 4161"/wir (JS/,.,.) a. = (FT,/st.-:C."') q:: uAJl?:Wi ( n/SEG ")
- s-A.'\ 12 7ilE #Al'//'1.IH -4f1Nt:N7' lllMvt.'l> OCC<Jll! ;#"T ')!-= x, M/e #19t1E' M' z :r f;. 4 + #; ,&.. 7t 4 L, -@ '7)11$ M"llt..'i> ;Je /I .sr.eA/6:#T FM!M4,ei) St:JtuntJ,J EKLPT ?'>>'* ,/Je/.L6'WS EN!> .l!eAC;?cAJS '9..t'.!= A FCIAICfi&,) cF ?#c ,/)ISJ't:Ac.E,11£?NTS IMPtJScj). l"NSd)( .. ?e, /Ne feu..u.dJJ( F.e.e-£ :ilooy .2)1llt:A!AH. "91/[) i!esVG71AJ6 i'(;!fiflt1 . OF AND : i :ti ) "'* FB '4-9 '+(& r::-_ Ws ,!!. _ 3 {.,J> t {,'If 8L) _ /'!"\ rs -z 3-z "a t 2. ,,. iA -0/al E:. _ Jf,,Di {t.1 + 2§) r10 -IT (}-41... (!* .3 -!§)
I N N Report No, 7198.3-.3 p, 27 of .31+ ltlfll S91f.'7. Wiit/!*: Wa: ae:c.\ewS w&r (LS) h' 8£&.UwC Vl'r (Ll/111/0AJV*) 'j)
- 8E;U.IW$ C,Ufi.T 'J)1llHF:1'EI! (111 ) l> = 0:-t.LOW (AMI. (J1rcH (,.., *) ,_: BELUwS Cot11J0ttJTt:-O Le!#JITH (1u.) Q*-= BelLAIJS eAJJ) 'bEFt. (1u-) 9 = 8E',U.DtJ' &i b ez,;,, 1111.J ( tl'lf).) ,,. /fYAJr Ne" "1HI£ .sv/J$'>>7'lJ?'e!) "Ill? Pl/1.t! ('Ji"' M,) ,q,e 11 .re<11110 se; fj11 "'-(9 ). r.eumous iJE1"'w£8J #a ""(} AtN/J lt-.1 H1i WAI
- 7h'esE HE : 9=-r L a. * .a rr " Mrla1 !Mi a. 2iiI' -ET -123 r: I I 3 \.. I 4 Fr W #r/N w .... ¢ I.,, .J r:r r 2EI .,. iijTI' -({) {111 = de= d" "L1B -@ SV,t:'FfC16J7 ctf)vltT!tJNS 1-Mt/c BEN 7Z> /K#/Et1t: "" SdLvTNAI. SEUE"el9t. ,111Er;1.1:.os-,,;Re l'?t/19/UJ!Lcj 7'/c t:WF BE t;.r£i:J NC.('$ Nll.t,, /JE ,If/,/ /4.f'Rd/90/. A.J AJl.uAl.S /. -,... .!!.. Lil a. , ,., ....,,(}Ht:; l"a. : 2.. U rtS ::--;;;; "'!-/. C/,14AJre I), AuTJ MT t er;. f' CD) .J. 2 (/a A#l> eJ (eq. <P,@,f)) ""* 11!1JJG .a tiMltJrc JS ANb Ha @, ©) *
- Report No, 7198.3-.3 p. 28 of .31+ p,;q.' Sh-:t"7' ... S. CoMPARe ?'7> .J Ir A SltfNIF/e,111 r /)IF.Ft:ReWCE l.S P.fl!!.ScXJT; ASStJ/VIE #EH1 l/ALllCS tJI=' J=',i /9vl> .Ma I. Rt:Pc4r 2 7N.ev .S"' JS /19CN1t:Vl?-/) Ar 7'>>/S AJ/Alr SllrfiC/t::A/r IS "'44/ P,.?t=SedT 71> '7JIE Mt9NNv>? ;lfluEAJT /Al P1A:: ("eip. 3)A11D /1*1./Cc 7Ne S!ZESS. r-:--1411, ""' -i! EA.lb /)EFtCCnoAi /J/RRMA7J"J.i /U"l.A?"wC:-10 /He fi'Et{.(J"UIS /.S "9VA/1.A/.3tfF CIM/.JtJn: /';t/C ctPt1/vA1e:-11T l'l>>At.. 'J"Zl'J!/ER.!C: p.?,e CJPvo.(.vr16 AJ RY 7llt= F()1..c.11w1AJt; : e = 'Qi ( 8 +-3t*) +-W,. La. '-3.(,, D} B* J. R.eHJus Foe AJlu " h' Ji>.J. KM.Ct: J. EllASt'o fl.IF°-ot:H"-he-Mo ,M;-C-!J 111 ,,71 L .J. S,tS° ?b ,.Y,f&fiP;tdA,/ 3. £/J/Jrco sPt"C. A.lo. FLO* J'7l<L11. .. + 4. ()SIJS 33/. 7
- 19'9
- I N "" * * * @!!!.l .. !!!..NL..-* -. 1. -@!!!.* .. !!!!I-. -,. -I __ ,.!_J I --11.!.,I C. Sample C11lculatlon: For the purpose of Illustrating the seismic capabl!lty of a complete penetration, f!j6 wlll be used u an example. First consider the head with app:led piping reactions: FllJE1J HEIJO °ltH .. Vue+ v,,,
- Mos11 + IArH. On the basis of the a.bove, there would be no shear r.iltted to the containment nozzle through the head, Therefore, for conser-ratlsm the shear on the le-H-. side wlll be neglected. Hence we have: F,.
- VosE + V"' : '1000 +7000 = llJDf!Jt> 1.8, M,.
- Z (Mou+-Mr,,)= Z(zooooo lf/2.)rtooooo AitL8 To solve the problem the following additional Input Is needed, w,, = soo tB d,, * '",. t.,
- l.218" ) ... w., I .C '3fil = / 1,Lf!l./ttJ I. tSf&. 1114
- lie zr !'14. > ,,.,.s J, .. 20 ,,,., ) . w. = .S:.;. llJ<#./ :r .tfJ.3 1.8 *
- c 12. t,:: *SOO L8
- 8.75111. Le :: '*1S /No L, II 11a."' 4 .. I., .,. 'I = 11'1.1' /No D: 23.ZS" 1AI. J j:: 1.S-11.J. f., = 2 1t 7100 :: Ii/ZOO Wo
- 110 LB BIFLUWS J>llTll L "' IS:'" t#J a.= >" +? a.w11r. > i = +(..Jg)'" : 3.'13-F%Ec:."" Fi..: :u,1we. =3.M,,.,tf3.3 ,, 1s,1.s f;, r 3.6/ W,,:: 3.,/ >e S'OO.: /BIO I.II At this point we are ready to begin the lteratlve part of the solutlor._ /Jr n>Y /. 1.ET I;"' 20() 1.8 11a : SZI '"" 1.8 z. F,. = fi. + Fc. /. r; +Fe } <011s1oet:s (/* *B .. o '" eq @ ""0 =-/8<<>0 +15{, f'/$10 f-ZOO = zozoo Mr = Mp -F.. Le. --fi La
"" 't I t .,. .. ..,.. i r-;;-. ti "'()-i ... .... a Ir QQ e. .. :: ... Ci Clo ,.* l ... .... ... + i It ., 0 ... "" ..... 0 0 ... ,.. i ai \! ? "' () . -0,.
- ao "? () .... ... ... :::: -4.. + -:.. ... ,, + 11 ' ..... "" 0-.. 0--() J ... .. '" ft' * "' I IQ ;::? i"> ... e; "' "' .. iii .... I . II co I -d l ... 0 II It e " ... c.:.... 0 ... C) ._.. ...... i.:::.:.....i -.l () O" N @} i .. ... r \,., .... ._. ... ... * .... ...... I § = I I ' () "" ) ...... 0 .. i I 0 ... 't. .. a 0 -... " It () .... fi. f It .., "" c::. <) t\ 4,,* i () t )( i . . "' N .... "I) "" -II
- II " II I ....
- ti u * . II * .... .. II " I II l Iii> Q) . tw; -.: \.:, N ...;
- f 1, * "" I 't ' I .. .i .. .. t t >( '!: .,,_ .. "' ....... i '" -a !! .* : .. -':::, ... "" .. ... I ... ...... It ll Q) .... 'O \It " ... i ! * .... .... .::: " "" "' ' --:t "'Iii lit . l I ... "' 1' ::t. I ... ){ I .. + "' .... ... I $ -....:.. 0 * """'-"' () ..... !.::! .... . "' () ..;, .., II ... t; ::: ,. II ) if{ ... 4)( () !:: () 't * " .. + ...-"" .. I \.., () ti " :;l I "' .... ... co r-: l .... II QQ Ill) ' )r II Ill """ Iii ... i.., \.:} .... .... I I It 0 )i Ci 'll () 'I( () " ... .. .., ..a-.. 0-"' I @} I 0 ... .... .... ..... 1t ... -.... I ... () 0 1t"' I 0 0 Iii ... ' ' ' ,, 0 8 i' .., "' 0 * .. 0 ..... ti C) "" Cl 0 -..r & .. s ,... i-a .... "' . . ..... .. II * . ,, II II II *
- ll I& ** r It " i
- II II II i: <D l.f \(' "i ..... '-i ,.... !
- I 3G-24 I r-..:> U'I * **-... t, lllNIUCXY
- IA Rll'Oln' N0°7l';->l)-J (js s. 3"1,0 1-(-. OOS'S'l"/>rf-1d o:: * .J"f78 IN s. /HI$ '?'1' 81! CHJ$C F,.::1 18800 ii! Mr : B ll OIJO ,., LS Sr*, ()tJS'SI i'A'/) !e. ::: . 348 '"' 8EAJ'b11J6 Ma.,3::: F;. l, +µr + 4 L,,,.... /(,)f,1.&/ .. ::: 1IP00"110.-f + 821()()() * --:z->tlltJ*f ::: 3, ZS"O, oOO 1AJ *. LB. r-M..,, .J Z £0. 1>()0 v., : -s:"' = ' ' = 111()() PSI .... -,,,,,,,. G" = .ZZS-00 PH (Fltllll B-'l* 1 '"II') BELLOl.'\t.S' E:fVf/111. llX1IOL. MOtJEIAE'AIT e .. :u.zr,,,1.5' (-.oD>Slf ]Y. H')#-l/OKtr-43JtJ.'1 l'S:&J /S:&,J, J>tl.fUJOJfZJ*ZS' = . 13( +. ()()' = . 1-11-w.jcoul/. e .. yc =. " Pl!tJi>tJe.es A/> 70tJO <:.ye.u:s 7#* 1"18tJtJI: CAl.WlA-r*/) I/Al.tic IS wEU. 8£/.(J()} rH.t: C.YGLIC VALVE uscl> /A) J.IFc CALCt/J.A71Pl\JS. ?h'B'EFM"e:; 711* ORt&Al/11. VALVES "4£* cSSc4J 7"/Al..t.. y ,/vsr1F1*b 19S CtJ#SE1!"11A7/vE * .. _ 9-'* -RCl'Oln'-. 719BJ-J D. Discussion: As can be observed in tho sample calculation atove, the resultant bending stress in t.he penetration is quite lov it eompari!on to alloi.able s. Likewise, the belloi.'ll equiv a.lent axial mover.ient per convolution is low bee Section 2). Therefore, ti'.e penetration aasew:cy-is trell able t.o 5upport tl:e seis::ic lo.id use as defined by "Ebasco". Furthermore. sl.r.ce the stNsses rrodi.;ced. by pipe rupture case are low for all poin<tratior.s (see 5 J ..: * .! cohsidering pipe rt.pture loads exceod the seis:r.ic loai!s, it wo:W..:i not appear that. furt.her seisr.iic calculations are necessary. :.o.,,_ ever, ir. the interest of conservatisn scilar calculatior.s were made on Penetratior.s 5 a.nd 40. 'rhe results as quite lov. Those as well ae .those fro:n the s11.1ple calculation (F36) are presented in the table below: Pt!;_ Ob 1.5 Sb _e_ !Sl'.S.t. 5 9610 PSI 2::?500 PSI .074 :L\ * * :;c,3 ..... 36 16700 !'SI 22500 PSl .142 n;, .4)6 il. 40 8020 PSI 22500 PS! .052 :il., ,455 11. Therefor&, on t.he basis of the above results which cover Ua complete size range of penetrations as well as rei:reser.t the higl'est. loaded ones, it is now safe to conclude that all o: Type Ill penetrations a.re suitable !or the seisr..ic C:esign lo:lds defined by "Eba.sco'. *
- APPENDIX 3G3 REPORT NO. 71983-4 *
- 3G-26
\.>) 'P N ...... !*1'6 * .. __,,,_ REPORT NO. DESIGN CALCULATIONS FOR FLORIDA PCWER & LIGHT COMPANY ST. LUCIE PLANT (FOP.MEP.LY HUTCHINSON ISLAND) TYPE I COllTAiflMEtlT PIPING PENETRATION ASSEMBLIES P.O. NO. NY-422264 are: !lo. 1, Main Steam (SG-lA), I-34MS-2B No. 2, Main Steam (SG-18), I-34MS-29 !lo. 3, Feedwater, (SG-iA), I-208F-14 No. 4, Feedwater, (SG-18), I-20BF-19 by: lLw. a erman Date: .J..I. Al'l. lf."IJ. , Bellcws Engineering ?.;.r-r-:ived by:/Jrl* (/, "/1/,,..tf_ ..... _,, M. ;t. il!a!kmus Date: 7/¢3 Manager, Bellows Engineering *
- f-166 IUINS _______ ._....... -.. -TABLE OF CONTENTS 1. Summary or Report 2. Bellows Calculations 3, Process Plpe Wall thickness Check 4. Nozzle Stresses Due to Pipe Rupture Loads 5. Lateral Jet Force Calculation 6. Guard Pipe Stresses Due to Jet Impingement 7. Penetration Seismic Analysis B. Penetration Assembly Drawing w (j) I N CD t-116 9-.. -1. SIJll'.mary or Report
- A: sutstant1ated by the calculations herein, all major components have been demonstrated to comply with the 1nrormat1on supplied. The only exception are the hea1s which are discussed in Reports 71983-1 and 71?8J-6. t-116 * @!!!'-!!!.."!..-. -.. --2. Bellows Design Calculations A. The intent or this section is to demonstrate by suitable C)mputation, the structural adequacy or the primary and secondary bellows ror the design conditions stated in paragraph B below. The conrigurations or the bellows are depicted on Drawings 71983-Dl.l, Cl.2, and Cl.3, attached. B. Ccnditlons: (1) Primary Bellows Condition 11(600 Cycles) Axial Movement J.68" comp. l.4J ext. Horizontal Movement t 1.78" Vertical Movement t 1.5* Material -A 240 TP 316L Pressure -44 PSIG Temperature -Factory Precompression -3/4" (2) Secondary Bellows Condition 11 ( 600 Cycles) Axial movement t1.43" Horizontal movement+/-1.78H Vertical movement +/-1.5" Material -A 240 TP 316L Pressure -+/- 5 PSI -264°F Factory Precompression 3/8" C.
References:
- l. UsAs B3l. 7-1969 12(7000 Cycles' .5" t .2 '2(700C C\'Cles .5" +/- .2n 2. Standards of the Expansion loint Manufacturers' Assn;, Inc. Third Edition 1969. D. Bell*)Ws Hoop Stress: The bellows hoop or circumrerent1al membrane stress ls determined by.use of the Barlo* for S:raight pipe suitably modified to tor a distribution metal. As applied to a bellows the equation takes the ;ollow1ng form: ?cl .,, = Ztn;. ( S7f 'f" 1"°1)
- I N \0 t-766
- WllEP.E: E. F. Sb ? d t 'II q n p 9--'* lllNMlft' * :::tress (P:::r:
- Pressure (PSIG) *Bellows Root Diameter (in.)
- Bellows Wal: Thk-1 Ply (in.) *Convolution Height (1n.)
- Convolution Pitch (in.) * !lo. Bellows Plies Bendinv l'ressure Stress '.r.<: :..<:llc,;c m:riolonal stress (stress tending to crush o convolutton:i in an axial directlor) is d".:ter!'.:lr.ed by the following theoretical formula, with empirical nod1ficatlons to correlate with test data.
- Pw2 10 273 t'-' np srn &: llows :i:0verr.<:nt Ca:iacity: The movement capabl 11 ty of a Eellows Is evaluate:i on an "equivalent axial movement" per convolution basls. For the secondary seal bellows, the allowaLle movements have been limited to the following: Allowable movement ce comp,) .85 (Q/2 -t np) Allowable extention novement (e ext.) * .45 (q/2 -t np) 7he allowable for compression is based primarily on limitation from the geometry. For extension, tt.<: 111'.'litation is on empirical data to prevent cf the convolution crest and to prevent exc".::31v<:ly high stresses that would signiflcaitly reduce cyclic Ji fe. 7he axial and lateral movements tabulated ln Paragraph B ar" converted to equivalent axial movement per convalut1or by the folllwing formulas: (Ref: Standards of the Ex*anslon Joint Manufacturers' Assoclatlon):
- t-T66 WHERE: e x ey epc x *2n ,. lu'l!!!,* .. \!!J_..., t, lllMl'llCll\f
- Axial Traverse per Conv. ( + is compressive) eKDt 2n L-C-x/2)
- Equivalent Axial Traverse per Conv. (caused by Lateral Movement)
- Axial Traverse due to (+ is compressive)
- eext * -ex -epc + ey
- Total Axial Ext. ecomp.
- ex + epc + ey x y x PC 0
- Total axial compression per convolution
- Axial movement (in.)
- movement (in.)
- Precompression (in.)
- crest dia. (in.)
- d + 2w n
- of convolutions in one bellows L
- Jverall bellows length 2C
- 2nq K
- Lateral orfset constant from Standards based on L/2C ratio G. Bellows Spring Rates: The bellows axial rate has been proven experimentally to be predicted tte following equation: WHERE:f
- x E
- T
- fx
- 2 n (d + w) E 'r3/c:.i p Bellows axial spring force per convolutio!'.(L!'/I:-.. ) Modulus of Elasticity xlo-6 (PSI)
- 27.4 lOOt W---* to height ratio Cu
- Spring rate denominator -an empirical factor which is a function of convolution pitch, diameter. n
- Number of bellows plies. p The axial spring rate of tP,e tot'.ll joint can expressed as:
w Q I w 0 f*TM I 9--** -k 'Fx . T x WHE.P.E: Fx
- rx e x e
- x x n x
- l 1n. 'n1erefore t.y substitution "e obtdn: kx
- fx 7ri (Lb/In.) The lateral spr1ne rate can be expressed as: Icy y WH:Sf:E: ry . rx:i ey 2L *y n .L -CT 1
- 1 1n. 7herefcre bY we obtain: ky
- fx D K D 7"r.
- K rx o2 (LB./IN.) ky liLn(L -CJ :!. Cyclic Life !Jelermlnation: It has been proven through extensive testing that the cyclic life of a U-shaped bellows can be predicted by use of the following forr.tula: ll *1!(6f,:*t t*TM *
- 1.TJ-.. -WHERE: N
- Calculated Number of Cycles Su
- Ultimate Tensile Strength of Bellows (PSI) S0 * (1 + p/np p1)2 675fEe./llJ f
- I )" Sp
- 4. p T (I+ P/"" s0'
- S. + s,. s 7ilw/ Sir us ff,,,.,..& lcr (psi) p I e .s. 't /z i.cJe. ""' -* 1
- 1* ' * -;;:;-= iil.c.lr.ttu ..4 "'-':Jll ,..r1* E
- 21.4 ... 1-Ao,,/.,/o1s o I El*s /.1i:.:fy (Psi) <!,
- ec1c -::: Elfccl,;,t fn.vtnc per (ilf.) w
- Convolution Height (in.) Sc* 171500 PSI *Cold Allowable Bellows Stress SH
- 16.l!-O PSI
- Hot Allowable Bellows S!ress 'We
- W -f/3
- Effective Conv. Ht. (In.) n
- Number cf Bellows Plies p Q
- Sc + SH
- Hot/Cold Stress Range Patio e.
- We2
- Elastic Traverse per Convolution (in.) ""' 1f51lT For purposes of cyclic life determination, the non-cyclic precompress1ve movement installed at the factory ls r.ct considered as part of the total equivalent axial per convolution. Therefore. we obtain: e *(e +e) eye x y max. *
? w .....
- r-1" I. I _.. .. ......,, P.esults: l. The pertinent bellows static pressure stresses are presented 1n Table 1. page to. The allowable stresses have teen taken from B31.7 Appendix A. Table A.8. 2. The bellows movement capacity evaluation is shown in Tab le 2 page II
- The allowable equivalent axial tra-verse per convolution is computed per paragraph F of this report. 3. Computation of axial and lateral spring rates are inclu ded in 'Iatle 3, page 12. II, of bellows cyclic life is presented in Tablr 11, pa11;e I 3, tor the 600 cycle cond1t ion. It is obvious that similar cor.iputations for the 7000 cycle condition would produce results for in excess of 7000 cycleo for all bellows. these herein. There fore* 1t is redundant to reproduce * -PE/II #0. I, z. 1, z JEC. Put. 3,4 .er,. I I I * @TUii TURNS ....... ..-..... _,.,,,,_.,
- lllll"ORT-. d t w s n. s, s ... s..,, * .,, (JN} (HI.I (11!) (1A1) (ls1) (PSI) (PSI) rf.3 .os ;>.O IS" 8*8 4'3'0 3300 11r't>O I I (f.3 I 10*10 ;'o J7S-: j ..,4,3 B*B Jl'IO .JJoO I i I I I i I 1 ! .fS'O .J1S-f$,J 10-10 TAlll.c J IJFuovvs .S71iESS R£S(.llr$ )
iE-!J (
- Ta.I d w Tl t. YfC. c:r. e., e, c::..,. e.,.. e,, e, e .. 1 e ...... (1A1) (t/IJ) (111J f1A1J (JAJI """' ,,.,, '"") ,,,.. l (, .. 1 '"*' , .. , {1'1\ '"" (,,,, '* z 5?.' . cs-<!'.O IS' I* I S" 1 .... ,, .047 .2JO .zu .rfo .s.sJ -.Gff .l48 .l'IO . l. '1.3 P SPRING RATli.S .P.e-N. c( t: (..(.) 1 Yt K T CIA. '-i., '1 NtJ. l'w'I OAI) ('flJ.J t1N.) l'.w.) (1N) (1.V1111.) t'llV1MJ /, 2. 59J .as?> t?O /.S" e-8 /.Z.J L.;o 217 .37100 S7 .,r 2170 U.80 /, Z ,.,.J JO *IC / . .J+ r.so i'.79 4Jloo So JS" Zl?O .,soo *@ SEC. l!I i,llf 4"1.$ a-a / * .l.8 .!. so ?. 7"1 Z81DO .,ll ,, 1110 IS',0 .J,4 S-S:J I 10-10 /.JS .ii.SO UI> JS",00 <IS" .so 1780 .JZoo 11 .PFC.. TABLE 3 * .. I J
- VJ C'l I VJ VJ
- i ' " "I \) ... ... \I ). \I () "" \l: @!!:'.!.' .. !!!!!L--1.lllNIUCXY
- 0 8 0 g 0 () .. '41 .. () l 0 j }; ... " ... I\ ' () 0 .... 1>1-.,: ""' ..... :::: Q;\ ' 0 °"' ..i .... ... Vi-.,: () 0 ... ..... .... a 0 0 0 0 ell () ::::: .. ... "" t ..... It-.... ... "' ... .;:: ":' .... i .. J "" °" ... " l\,j l\,j () .... * " .... """ "' '" ... .;:::. 0 I\,--*-------------...... "' <> ""' ..... ... "'I "' e:. ... ... \ll '-... "' \,\ "'v .... ... "'v "l' '.!\ * * """ 1'URNS .. ..._ ___ _... .......... IHl'OllT-. 3. Process Pi2e Wall Thickness Check A. General: The process pipe minimum wall requirements f;r design pressure and temperature as specified on the drawings has been computed per the following equation B3l. 7: tm
- PD 2 (SE t .4P) The sE values in the above equation were taken tron B31.7 Appendix A, Table AB. The requried nominal thickness was taken as: tNOM
- trr(. 675 (For SMLS Pipe) B.
References:
- l. USAs 831.7 -1969 2. Ebasco Drawing No. 6770-6-213 c. Results: ,,. The results of this evaluation are presented in ratle 5 telow. Note that all process pipe possess "' wall thickness to resist the internal pressures sreci-"' fled. Act. ..L _£_ _g_ --1,2 985 34 17500 . 936 -1.125 :.'.in. 3,4 1100 20 15000 . 712 .Blll 1.031 !@1Ll. I I I t-7" w Q I w "'"
- 4. . __,.,_ Secondary Seal llozzle Stresses Due to Pipe Rupture Loading Cc.ndition: A. For purposes of analysis, consider the following "free body diagram" of a typical type I penetration assembly. -Ht:lfO "* ,--) 'e11rec .... h 1-10lur L --tL R, ...._-+--.... r ) H M>lft' In tt1e aliove M,F ,V and Mt each represent a combined effect of thermal expansion and pipe rupture loading. This 1s arrlvec at b7 assuming on<: pr:>cess pipe hub to undergoing its normal flex11i111t7 thermal expansion reactions while the other t.ut 1s loadEd ty pipe rupture reactions. For conservatism, thete ar<: assuned additive. For analysis purposes, the following comb1natlons should be considered as to their relative '= rrects: 1. 2. 3, Bending and Lateral (M,V) Torsion and Lateral(Mt* V) and Axial (M,F) the load transmitting structural attachments consist or loose rlr,gs at the ends of the nozzles and trunnions on the flued heads, it becomes obvious the worst condltlon from a nozzle standpoint ls the Bending and Lateral case. This ls due to the trunnion resistance er the axial and torsional loads. There fort, fr?m statics we have: R + R
- v 1 * -RlJ.. + R2...(
- M Solving we obtain: R2 . M + v 1l:"" .... " Rl * -M + v !:(" 1 t-TU
- TUINS --------... --Converting to a max nozzle bending moment we dbtain: (Mb).."'** = b 1<2. Therefore, the ma1lmum primary bending stress becomes: Jb Where: CTt:
- Mb
- z . _&___ ::' 'i!:. Maxlmum bending stress (PSI) Maximum bending moment (in. lb.) Nozzle section Modulus (1n.3) The above a maximum prlnary bending stress anj is therefore comparable to an assumed allowable or plpe yield per Section III. However, since the loads are the nozzles to loose ring flanges, : t becomes nece:ctsary :o determine the bending errects 1r.:pcsed (stress causing nozzle to deform at the ends). This effect result in an ovallzation of the ring end sufficient tc or yield a cylinder or 1nsurr1c1ent thickness. 'l'o arr!\*e a: a suitable stress equation to evaluate this effect, it was . necessary to strain energy techniques. In tt:e ir.terest or brlevity only the highlights of the developr.:ent will te presented here. Basically, it was assumed the nozzle can be represented as a cylinder fixed at one end and free at the other. A lcs.:! ls applied near the free end distributed over half the ference as a cesirle function. The potential energy (V) of the applied shear distribution is equated to the strain etergy (U) or the cylinder using the "First Minimal Principle" or the minimum of total potential: ...illL + d.ti_ 'dV =0 L llhere,:.. is any arbitrary variable.or parameter.
- I w VI * , __ {ff;)!!!!'..!!!.N!.._ 9-.. -R:>r our deflection equation it was assumed: lot.I : L (" (1-(os -fr) Cos 119 q, The strain e._nefl' of the cylinder is represented by: { / H(t*.AA)Xxe"} a.el9olt, *
- lihere: Xx *
- meridional bending curvature ..L x8 .a.> 419a.
- tangential bending curvature XxlJ torsional bending curvature The applied shear load is as: ,, ZR (. o.7T the potential ot the applied load becomes: (IT'/>' -v = 2. Jo % & w . Q d. e Ler1vatives ot and V are taken and equated as follows; . dJd.. t :i!:!_ = 0 ac.... ciC.:.....__ * """ @!!!!* .. 9-.. INllCllY The above equation is then solved for the unknown Cm which represents the mth term Of the series E(,.(1*(11'¥._J<.tslf'9 and hence an expression defining all terms of the series 1s now available. The final deflection equation becomes: * -Po.' f '!l!J {!-c-o"'" (*-,,,.. J "a w: TO (1-(ol lL (lr-1)+ Lh'k.-t'kJ 11-I' ., C1f2. l I"' f'"I Where: i:t' D *---12(1-.u.") The circumferential bending stress equation is: *Et ( v 9'
- 2r,:;;Jj '){ c9 + ,)J ){" ) Therefore, assuming /J* .3 and X
- L, we obtain by substitution: ... . ,., ... If* St TY!J \ fl (*I (oSll 19 V.1>
- 7 (i-( *l rr:1 L .-,-. 11:2 ... ".. -;:r.:-Examination of the above leads to the conclusion that the mum stress occurs at Q
- 90°. This is due to the fact that a:: terms or the series become positive. Therefore, the final equation becomes: oO I)" 1nR n,), Ft . , --ra-ft*tosri:;L + (%) ..... .. Where: R
- Maximun transverse load*(lb.} a *Radius of Nozzle (in.) t
- Thickness or nozzle (1n.l L
- Length of nozzle (in.) x1
- Distance of load plane rrom fixed end (1n.)
t-7" .-.. -t-T&6 I w °' B. i. 031.1 -1,cs 2. Ebasco informal memo dated December 22. 1971 L.J. Sas to W.S.
- 3. "Advanced Strengtt or Materials" by J.D. Dan Hartog -* "Theory or Plates and Shells" by S. Timoshenko C. ons l. l'.ain 3team P'!netratlons (Pl + P2) *49,l'JO,OOO H LB b
- 38.25" 'I.
- l 'J l 'J'l'J LB a
- 33.8 in. *"'
- 44 lN t
- 2.5 in. L
- 44.S Ill X1 * <fo.75 P.
- 4')?1J'.lll'.l0 + . 655000 LB ' x 44 < 1\
- bR2
- 3L25 x 655000
- 25,lOO,OOOJ IN LB B*.m11ng ::tres:;: Vb _ fh -? -7T,i.3J K'"x 2.', = 2 71.fu Psi Local Dlscontlnuity a/L *
- 7f.O l*/b
- 4o. 75 * ,91e .!Cl.! *. 916 x 90
- 82 .4° i! L. V-t * ..... 0-(*582.11°) x .0221 : 30100 PSI *
- t'M:l!!!.l.!!!!5-\!!J-.. _.,
- 2. Feedwater (P3 & P<fl M
- 5,040,000 IN LB. v,, 86,000 LB ..-<
- 114 rn. L
- 44 111. b
- 39. 5 a *27.6IN t * .1 IN x1
- 41. 5 ,,,. R2
- 5040000 2T1ni +86000
- 100,000 LB 2 Mb
- b x R2
- 39.5 x 100000
- 3,950,000 Primary Bending Stress fJi
- Mb
- 3950000
- 1650 PSI r 7r xt7. 6" x .1 27.6 a/L * --n;-* .627 11/L
- Ill. 5 *
- 943 -irr Local D1scont1nu1t) 7fti-. ,91!3 x 90
- 84.9* <Tei.* 15.3 x lOOCOO 0-84.9°) x .0224
- 31200 PSI D *. R_esults: As can easily be observed all values are within taken as l.5 sm* ror the primary stresses(O£) and 3 the seconcary stresses, (<78b). Sm
- 23300 PSI e 10C°F rcr A 515 Gr. 70
- I w ......
- r-746 5. 9-** -Lateral Jet Force Calculation A. The Jet forces ror a longitudinal break are determined per the "Ebasco" spec. as follows: FJ
- lPA WHERE: Fj
- Jet Force (LB.) K
- Phase Factor (pg. 9 of spec.) P
- Operating pressure (psig) A
- Pipe Flow Area (in.2) Per the "Ebasco* Specification. Break Length (,l)
- 2 x Pipe Dia. (D) Break Area
- Ar Therefore, Breat. Width (c)
- Ar/2D
- Ar; J. b. Peferences: 1. Ebasco Spec:rication No. FL0-8770.124 2. EbilSCO Dwg. No, 8770-G-213 c. P.e:ult>: The results of this computation using operat1 pressures and piping dimensions from Ebasco Dwg. No. 8770-G-213 are presented in Table 6 below. The forces will be used in the subsequent section. !'Ell. :n (IU) r,7 3,11 20 t (I!l.) 1.031 At P (IN2) K ITiJ ) .'63 253 1050 ,97 TABLE 6 L C (INl ilfil 11.5 40 6. 33 i! 516000 Per the Ebasco specification; the above K double in the case of fluid rebound. Hence, the forces comr-uted above reflect this effect. * 'l-1'6 6. --------_. .. __. Guard Pipe Stresses due to Jet Im?ingement A. General: To determine the et ect of a slot break within the penetration assembly, a conservative approach will be taken as follows:
- 1. The maximum bending stress that could occur in a guard pipe with a hinge would be at the center cf the assembly as shown below: f' 011r. 6<1ACll -z--:1 *-: :5' ";.-:: .. *
- r i.----t, ---I M_ .. !fi (/. -L/z) = .q. ( l.J
- 4f I 'l'T;.. .. #..-* /'l: wHt!1£ if r 1f" ('D11 ";_ t ( f; Note: The values or FJ are taken rrom Section 5 of this report. 2. The second area or concern is the hoop stress caused in the nozzle or guard pipe due to the !et force striking a finite area. The area is develcre as follows consistent with the "Ebasco" spec.: 1---C.i--I --...t --a U D11/Z . L io*r.,p ' 10* -r TYP I (DN/2 _ Do/2) D./Z C
- C + 2 TAN 10° N
- c + (DN -Do) TAN 10° LN
- L + (DN -Do) TAN 10° A * * +(DN -Do> TAN10°) [L + (DN-Do) TAN 10°]
t-1'6 '5 I U> o:> p
- FJ -,;, 'T'-.. _. Vt
- P(DN +t) <'.t
- WHEP.E: DN
- G.P. I.D. (in.) Do
- Process Pipe (O.D.) t
- Nozzle thickness (in.) FJ
- Jet (LB.) c
- Slot Width (in.) L *Slot length (in.I E. 1. EtaSC') Spec. :lo. FL0-8770-124 2. "F0!"'.:!ula: for Stress and Strain" by R.J. Roark 3. u::.1,::. eJL 7-1'.JS? '=* The tabulated results of the above computatiorts are pre:ented in Tables 7 and 8 below. This condition t.a::. t'!'!n a::umed a:; faulted thereby Justifying a yield allowable stress. ,,,... Fj L L S i:;:*:* IL&.) Cill) (!YI) crn3J (PSI) (PSI) I 868006 l8 166 2440 11746 29400 2 68 168 11920 29400 3 40 188 92B 23400 30600 4 5llvuo 110 177 926 21800 30600 TABLE 7 r.;i t D:> C A2 t Sy r;',. WI) (W.) (Irl) (HI) (IN ) (PSI) (PSI) r-44 1.5 34 11.5 925 14200 29400 2 114 1.5 . 311 11.5 925 111200 29400 3 2'.J.5 1.25 20 6.33 3311 19000 30600 II 29.5 1.25 20 6.33 334 19000 30600 TABLE 8 if the hinge were not present the bending stresses in the guard pipe adjacent to the flued head would be by the following formula: t-1"
- tulNS ------._... .. _ Vbx
- FJ (Lop -1/2 L)] 'i! .... 1 L,, *I * .. .* . FJ.1160 f:J t ..... l . .L.U '?. IS..,,uo ,.,#'# NEAi> The resu:ts are shown in the table below thus demonstrating the imposition of atrt'sses in excess of yield if a hinge is not employed. PEN. Lc;p Vbz Vi (PSI) NO. 1/2 L l 195 311 57200 27600 2 200 311 $9000 27800 3 225 20 1111000 29100 4 213 20 107000 29100 8. Penetration Seismic A. General: Consider t e following "free body of a type I pEnetration undergoing a seismic d1sturtanc R.t-----.... _..J.._ __ i""*-***-* 1 Ri r.. (ti+ -L* *--*---1--L -l,-<tl fu £E_ ---** ' ..... , jF. J I F&\i 1'z t .. ...: ,. -----L, ------------\ F. ') "" a' i.w. Where: l="s, * -+ <: cs,) cbellows seismic force (lb.) Vlle1* wgt. of one bellows element (LB.) wgt. of bellows center spool (LB.) i:-52.
- a/g (uln 1.t Wco l
- bellows seismic force (lb.' \r.IB
- wgt. of one bellows element (LB.) Wcu= wgt. of bellows center spool (LB.) Fc.1* F,1
- 112 a/g "ap 'ltl,p. Wgt, of gurad pipe (LB.).
- a/g w11 Wt1
- wgt . flued head (LB.)
- *
- t-$ 'IUINS ---------.. -t-TU tiil!!!.'..!!!.."!-'TI-.. -____ ,.._ __________ "" ? "" '° Ws
- a/g Ws' *See. Seal nozzle seismic load (LB/in) Ws' *Pipe wgt. (LB/in} "11
- a/g WN'
- Cont. nozzle seismic load (LB/in) WN' *Cont. noizle pipe wgt. (LB/1n) R1. R2
- Anchor structure load bearing reactions (IB) Fp
- D.B.E. + Thermal E1pansion f1ping Load (LB)
- D.B.E.+thermal expan11on piping load (IN LB) U:ilng the principles of statics we obtain: P.1 + R2
- F'J' -Pp 44 (R2-R1) * -Mp -MT WHEP.E: MT * -*2 Ws + WsL4 (L5 + 1/2 L4) +WN L3 (L6 + l/2L3) + P81 (L4 + L5) + Pa2 (LJ L6) + Fa1L1 + Fa2L2 PT
- PR +Ws (42 +L4) + WNL3 + Pei + F92+Fa1 + Fa2 Zolvlng we obtain: FT -Fp R *---i 1 2 R2
- FT -Fp . -Mp + -SS M + fl'T p 615 The maxlmum nozz:e bending moment on the bellows side becomes: H., : i, (44 -1.,)-1i1 I.,,
- W.. L//z. The maxlmum nozzle bending moment on the outboard end becomes: M,,
- R,, ( 44-1.,) -88.2 W$ The maximum primary bending stress becomes: tr: {Pf .. )_., .. The maximum secondary stress at the end of the nozzle is determined by using the equation developed in section of this report. B. Rererences-1. Ebasco Specification No. FL0-8770-124 2. Ebasco Infonnal memo dated December 22, 1971 L.J. Sas to W.S. Haberman 3. ASME Section III 4. USAS BJl.7-1969 C. Computations: 1. Main Steam (P2) For the purposes of this computation the is assumed: WBl
- 320 LB. WGP
- 13600 Ll
- WCSl
- 730 LB. WFH
- LB L2
- l i" W92
- 2.20 LB
- Ws'
- 160 LB/IN L3
- 1C2" L4 ** 42.25" Wcs2
- 950 LB. WN'
- 28 LB/IN Ls
- 5.75" L6
- 6.75" FP *VT!!.+ VDBE
- 120000 LB. Mp
- 1(MTH + MDBEJ
- 2 x 30000000*60,000,000 VJ G"l I """ 0 r-1'6 fiil!!!.l .. !!!!!L-9-t.--' The above assunes the moments from both inboard and outboard are a11itive while shear from both sides negate each other. Therefore, for conservatism, the shear on the left side will be neglected. a :: J. "I-P81* 3.61 c320 + 112 x r3o)
- 2470 lb. Fa2* 3.61 {220 + 1/2 x 950)
- 2510 lb. F01* P
- 1/2 x 3.61 x 13600
- 24500 lb. 02 F
- 3.61 x 8500
- 30700 lb. H Ws
- 3,61 x 160
- 578 :.b/in. WN
- 3,61 x 28
- 101 lb./1n. !t.T * -112 x 5 78 ( 5. 75 + 21) + 57 3 x 112. 25 ( 5. 75 + 21.12) + 101x102 (6.75 + 51) + 2470 (42.25 + 5.75) + 2510 (102 + 6.75) + 24500 (208 + 17)* 6,500,000 lb. FT
- 30700 + 578 (42 + 42.25) + 101 x 102 + 2470 + 2510 + 2 l 24500
- 1411,000 lb. We must now consider two cases. One where Mp & Pp are pos1t1ve and a second where they are negative. & FP Positive) pl
- 144000 -120000 2 + 60.000,000 + 6,500.000 88
- 12000 + 756000
- 768000 lb. P.2
- 144000 -120000
- 60,000,000 + 6,500,000 BB
- 12000 -756000 * -742000 lb.
- t-1'6
- 1UINS ....----_..,_ (M + F beth negative) p p R1
- 144000 + 120000 + -60,000,000 + 6,500,000 BB
- 132,ooo -6081000 * -476.ooo LB. R
- 1114000 + 120000 2 -60,000,000 + 6,500,000 BB
- 132,000 + 608,000
- 740,000 LB. Inspection of the above reveals the maximWll value of
- 768,ooo lb. while the maximwn value of R2 * -742,000 Ll!. & Mbl
- 768000 -2470 °X 112.25 -578 x
- 28,800,000 IN LB Mb2 *-742000 (44-5.75) -882 x 578 * -28,900,000 IN LB. Therefore the maximum primary bending stress becomes:
- r -z 8, 'oo,:oo "
- 3UO PSI " .,,..,, n.a .,z.S' The maximun secondary stress becomes: r:-7610110 ttJOtOO
- JS'"JOO PSI V* ..
- HS-Orf) l . ,, .... Ctcf,.* 4 2. Feedwater (P3) Assume the following: VBl
- 250 LB. WGP
- 8600 LB. wcsi
- 460 LB. WFH
- 3500 LB . w B2
- 160 LB. Ws'
- 52 LB/Ill w .* 520 lb: cs2 WN I
- 21 LB/IR LI
- 230 JN. Lz.*16I:l. L!>
- 130. !N. Lt
- 55.4 IN. L5
- 4.5 IN.
- 5.5 :N * *
,.. w ? ....
- IUINS --------.. -Pp
- VTH + VDBE
- 98000 LB. Mp
- 2 (MTH + MOOE)
- l_0,100,000 IN LB a
- 3.61t FBl
- 3.61 (250 + 1/2 x
- 1730 LB JB2
- 3.61 (160 + 1/2 x 520)
- 1520 LB. FGl
- FG2
- 1/2 l 3.61 x 3600 r 15500 LB. Fl'!
- 3.61 x 3500
- 12600 L.B. Ws
- J.61 x 52
- 188 LB/CN '1:1 * ;.61 x 21
- 11i LB/IN * -42 x1188 (4.5 + 21) + 188 I 55.4 (4,5 + 27.7l + 76 x 130 (5.5 + 65) + 1730 (55.4 + 4.5) + 1520 {130 + 5.5) + 15500 {230 +16)
- 4, 950I1)00 [fl LB. FT
- 12GOO + 188 (42 + 55.4) + 76 x 130 + 15?0 + 1730 + 2 x 15500
- 75)00 LB. (1".'p + Fp Pos1t1ve) P. * -98000 + 10,100,000 + 4,950,000 1 88 * -11500 + 171, 000 * !§.Q.JJQO LB. * -/P,Ql)I) -:0,100,000 + 4,950,000 ( 18 *--11500 -171,000. 183,000 1). * ,.. 1UINS .-.. .. -(Mp + Pp negative) R,
- 75000 + 98000 + -10,100,000 + 4,950,000 88
- 86500 -58500
- 28000 LB. Rz
- 75000 + 98000 --10,100,000 + 4,950,000 2 88
- 86500 + 58500
- 145,000 LB. *
'-* t..> ? ,I:-"' * -----.. -Inspection of the above values indicate the maxlmum moment w11 be produced by M * *183000 (44 -4.5) -882 x 188 bt *-7,390,000 IN LB. Therefore the maximum primary stress becomes:
- Mtu I:' -7, *
- 3110 PSI
- i! 1rlf Z'H ,.., The maximum secondary stress becomes: r--/ FS ooo Jt JIZ.OO
- _ $ 71"0 PSI V9i,
- 100.ono D. Results: It is easily seen that the above stress values fall within allowable taken as 1.5 Sm for primary stresses and 3 S,,. taken for secondary stresses. Sm
- 233000 PSI for A5lb Gr. 70 plate at l00°P. * *
- APPEND IX REPORT NO. 71983-5 *
- 3G-43 t-TM w t;') I ;:-. ;:-. 9--'* DIMllV REPORT NO.: 71983-5 FLUED HEAD STRESS REPORT FOR FLORlDA POWER & LIGHT COMPANY HUTCHINSON ISLAND PLANT CONTAINMENT PIPING PENETRATION ASSEMBLIES P.O. NO. NY-422264 Included are: Penetration No. 5 -Blowdown (SG-lA), I-2B-l Penetration No. 26-Letdown Line, I-2CH-l03 Penetration No, 36-Safety Injection LOop 1A2, I-6 SI-113 Penetration No. 40-Shutdown Coolinq, I-10 SI-422 PP.EPARED BY: /{!]{ff ,,.,. *-*---DATE: 1.1 l. *t.J w.s. Haerman, PE :;; Y.'{ 7'/J2 APPROVED BY: 4/;. (/, '/J1 t> /" i,, ,.._ .-DATE: 1{_11{_?-:J M.v. Mailtlnus I > this report is believed to be accurate, nothing c::.r1t.air,.:1J t.ereln st.all Le eonstrued as establishing any l)r 1mpll<,d. l
- MM
- 9--... .. -Section I Certification 1. sUITlllU'7 or IUlport 2. Progra.111 Antlyticiea J, Program Veriticetion Report 4, frpe Ill 1'9netrat.ion b1911d>l,f Draving 5, P11netration 26 AxiaJ. l'henaal Gradient. Ccliput.er Reaulta :;ect.ion II -Penetration 5 COB¥>ut.er Results 1. 2. ). 4. 5. fl. 7. I!, 9. 10. 11. 12. 1). 14. 15. 16. 17. 1'1. Material Oau Thermal Model Plot. Thermal iiesul ts Stress Model Plot Stress "n4ais Input Table Internal Pressure St."ese1 Thel"IMl Gridient. 3tre*ee* .lx.1al Load :itresses Torsion&l Load Stresses Transverse Shear Stresses Bending Monent Streesea Bending Morient (at. Snd) St.re11e1 Normal Operating (illcial & Bending) Normal Operating (Tr:uirreree & Bending) Normal (Transveree & Torsion) Preaaure, Thermal, a.rill Olk: (Axial & Bending) Pressure, Thermal, and Dee; (Transverse Ci Bending) Pressure, Th<1r1:111.l, and DBI:: ('l'ransverse & Torsion) Section Ill -Penetntion ;6 COftiluter Result.11 1. 2. J. 4. s. 6. 7. 8. 9. 10. ll. 12. lJ. 14. 15. 16. 17. Material Property Oat.a Thermal Model Plot. Ther1:111.l liesults S tru s llodel Plot. Stress Anal(sis Input. Table lntem.i.l rrusure Stressee Therlllal Grailient Stresses Axial Load :itreaaes Torsional Load Stroeses Transverse Shear Stresses Bonding Molllllnt 5tresse9 Bending Mom&nt (lit. End) Stres111ta Normal Operating (Axial & Bending) Normal Operating (1'ranm1ree & Bending) Norlll.ll Operoting (Transverse & Torsion Pressure, Thermal, and DBE (Axial & Bending) Pressure. Thermal. a.nd DBE (TransverBO a.nd *
- *
- t-7'6 .. * -** llllmlCXY t*TM @!!!E..!!!!!!-. _.....,,_, ....., C1 I Vt ::ect1on Ill -h!net.ntion *6 Co"°"ut.er ltesults 1'!. l-'ressur.,, lher"1al, v.c LIU.:: ('l'rdlleverae & Torsion) :;.,ct.ion I\' er.etnt.ion 3E Computer heeult.a 1. 2. ). I.. 5. 6. 7. 9. 10. 11. 12. 13. ;i.. lS. ii. 17. 19. <O. .21. 22. 23
- l'rC>p'lrty D.t.ta 7her::ial Model Hot St.ead;r Suto *1 her::u.1 .:at.a 7r;,.nsi.,nt. v:it.a Hodel l'lot. ::itr" i.nalysis lnput. Table l'!"lss*.ire ;t,ress!IS Th"r::i.il Gradi<?nt :::t.re!ses Tr;inai<?r.t l!'.er"lit.l :itresses Lo.:id ror,ior.'ll l.ovl :,tMslMs '!'rUlsverse . .itre!ses il'!r.din;r .tr.,sses ber.Ji."I: ;.:o""'r*t. (.1t. ::rd) ::itresSS! r,pentir.e (.vcltl r, B.,ndins) :.or"41 CiP"rsting (Tr'lllsverse and Bending_) !.or.,-al (lr..r1sveree 3Jld Torsion] :T'!SSUl!. :r ... r'llal' ;.r.d il:oL ( w.al and Sending) f'MSS*.IM, iher,,-.,.l, ;,.rid Oili; (Tra.nsverse and Bending) IT"ssur,, Tt<'ir* .. l, 'illJ uo;:; and Toren) fnnsient Co'lbina.Lion 1 (. 5 x Trans, Therm, Stress = Alt.. Stress) fnnsillnt. Cot'lbination 2 ( .5 x Trillls, Therm. Stress -,5 x S * .:i. Thermal Gra.dient. ::itress) Transient Co!:lbin>tion J ( .5 x **h*rl!l.ll Gradient :;tress) :;*ct.ion V -l*er.etr;.t.ior, 4fJ kesults 1. '.at.er Li) h'operty J,.h 2. iher:r.al ::o'.'.!el f'lot 1. iherm;,.l rtesults I.. :;tr'Ss Hodel Hot 5. ;i.roas .\ria.lysis JJlput Table 6. Int*rn;il hessur* St.ressea 7. fhormal C.:-dier.t :>t.reaees 11, Axiil Load :;tr,,sses 9. forlional Lo.id ::itresses 10. :ior.1il1;i: (Ht. in:!) Streseee ll. Tra.">nere* 5hear 3tr'!ssee 12. 3en11ng 13. >or:ral 0P"n.ting (Axial & Bendl.ng) 14. Operlting (7r.i.nsverse and !lending) 15. Openting (Tr.weverse and Torsion) lb. f'r"ssure, TMr:nal and DBE (Axial dlld Bending) 17. and lJ!lE (Transverse and Bending) };i. Pressure, Ther::ia.l and DI£ (Transverse and Torsion) Introduction The following, when combined with the balance of items listed in the index, constitutes the stress report on four (4) flued head forgings to be supplied as part of the containment tration piping assemblies for the Hutchinson Island Plant. All calculations for stress distributions were made using finiti element computer techniques. Descripticn data relative to the program and its verification is attached in Section I-3 and I-4. Basic Approach For analysis purposes the flued head and attached piping model was assumed as shown below: ' j.----1'" Yi.t f1.IJt!l> ¥' I < '-) 0:: F;cE f:"/KlrY po£ Y1,.t END ,,_./ E¥T. LOADS ---. ', c ,-----. .., .... i..-2* -,...--.,,* --1 ' I t=;**----:-* ---Li.-PEN l/2t r-26 1-1/2 36 1-3/4 40 1-1/2 t ! 3 3-1/2 3 L2 I9 19-1/2 19-3/ 4 19-1/2 L1 .,-7-1/2 7-3/4 7-1/2 I I OPERATING PRESS. 885 PSIG 520°F 2200 Psrn 4so°F 2235 PSIG 330°F 300 PSIG 300°F THK. J'.\St:L, -r--1-1.'2 2 2 u.> ? .i::-0\ f*TM @!!!.* .. !!!!!!..__ .. ---The table belov clll!lb1.n .. , Thenu.l Expansion and ()peraUll'l&l baa11 *arth* qualc* loads per Ebuco Mso, December 22, 1'172. (F) (M) (V) (Mt.) PEN WA!.. !.!!!!§. TOOSION 5 16SO lb. 16500 in. lb. 1650 lb. 16500 in. lb. 26 JOO 2250 JOO 2250 )6 1)500 200000 l.3500 )00000 40 9150 1.14000 9150 474000 It .. coulder Thtnal Bxpan1ion pl.Ill d**ian buil IU't.h411.!ake .. hit.ft the rollMd.nc: !Y s )6 40 6W1i 2200 lb. 400 18000 12600 I!!6l§ 22000 111. lb. 2200 lb. )000 )00 400000 18000 6 )2000 12600 l'!!!§.!Q!! 22000 in. lb. )000 l.00000 6)2000 Th* ratio bet.wen the ab4ml tvo .. u or load* be-* 1,))) 111111.dering operating bub *arthqualce "be -haJ.1' or the deai4P1 bHU YU...1, Furt.h*r, t.h* ab°" Yalue1 act at the eenter or the nuec1 hlad and compri" pipe loads *PPllctbl* to bot.t. the inbovd and outboard 11d* 1a the worst clllllbillation troa 111 onrill tea.cl etn11 1ta.ndpoint. For the actual 1tre .. runu the abcmt load1 wre Coi:Ytrted to stat.icall.7 equinlent. loacl11 applied t.o t.he ende of the et.res* lllOdel 8ho1<<1 lA t.he above sketch. Combi.na.tions in the clll!lbined stl"llH reault.s rune include th* follovJ.n& tor both the nol'lllal operat.ina condit. lals (include a pNHllMI, thermal, illll1 operating b&aia *IU'\hqllakll) llld t.hll ..,.gen1:7 conditions (inclldes preaaure, thermal, and d**l&n bub earthquake) Co!!!bination l l, pipe pr9HllMI etreeaa1 2. Head Th'Jl'llal. gradient et.re9111 )
- Uhl thermal load et.re11" 4, Bending th*l'll&l. load etret1aa1 S. Arl&l .. 111111c Load et.re*** 6. Bending .. 11111c 10&4 lt.NHH f-tU * (ii)!!!,* ... ........... Combination 2 1. Process pipe pressure stresses 2. Head thermal stresses 3. Transverse thermal lotd stresses 4. Bending ther1Ul load stresses 5. Transverse seismic load stresses 6. Bending seismic load stresses Combination 3 1. Process pipe pressure stresses 2. Head ther1111 gradient stresses J. Transverse therwul load stresses 4. Torsion*! thenul load stresses 5. Tr1nsverse seismic load stresses 6. Torsional se.fs111lc load stresses In addttton to the above ftntte ele*ent stress analysts for the f1ued heads, the 1Ki1l therm11 gradients were tnvesti91ted on Penetration 26 to denonstrate tn Imposed concrete shield vessel temperature of 1so*r or less. For this analysts. a slightly different model was used as shown helot: U,&,,,,,J 6.u111l**1 )\ CtUC. THk. Finally the pipe rupture loading contions have been treated In Report No. 71983-1 previously submitted. the balance of penetration major components have been evalu1ted tn Report 71983-3.
t-7'6 w (j) I .!:'-'"' * ?r::uvr::; SUM?'.ARY
- 7t.e results for the four (4) heads analyzed are presented below: PE!IETRATION NUMBER 5 Max Stress Stress Allowable t;or.-.b1nat1on Intens1ty(PSI) Location Locatio #1 l1':1rmal OP. (Al+BE!l) 23776 OD Outboard Process 57729 Pipe Hub ( #64) n OP. (Tram;+eEU) 19845 OD Inboard Process Pipe Hub ( 1177) 57124 #3 OP. (Trans+Tor) 24575 OD Outboard Process Pipe Hub ( 164) 57729 #4 (Ax-BEil) 31158 OD Outboard Process Pipe Hub ( #611) 57729 EXER. (Trans+BE!I) 26940 CD Inboard Process F1pe Hub ( lt77) 57124 N? E!'.ER (Trans +':'or) 35689 CD Outboard Process fipe Hub ( N64) 57729 PEllET!iATION NOOBER 26 ll,ax :;tress Stress Allowabl* !'"°"" Intensity(PSI; Location Stress #1 li:irr..al OP. (Ax+BEll) 25588 Inboard Process Pipe 54188 Fillet RAD(ll83) #2 :lormal OP. (Trans+BEN) 25608 " " N":, !lormal OP.(Trans+Tor) 25616 " " ,4 E!-:ER. 25581 " " E!'.ER. (Trans+BE!l) 25608 " " #6 E:t.ER. (Trans+T:>r) 25620 II "
- Y.."6 Combination 1UINS . -.. _, PENETRATION NUMBER 36 Max Stress Intensity Stress Location 11 Normal OP.(Ax+Ben) 19036 ID Head(Nl25) #2 Normal Op.(Trans+BEN) 19118 ID Head(U25)
- Allcwat: l 57t52 57t52 13 Normal OP.(Trans+Tor) 20402 Outboatd rrocess Pipe Hub OD Inboard Process #4 EMER. (Ax+BEN) 23569 Pipe Hub (#91) 57508 OD Inboard Process 15 EMER.(Trans+BEN) 22365 Pipe Hub(#91) 575ca Outboard Process N6 EMER.(Trans+Tor) 26775 Pipe Hub (1¥23) 57744 Max Transient Stress
- 25430 (ALT) @505 cycles /Usage Factor <.Cl Stress located at I.D. or Outboard Process Pipe Hub PENETRATION NUMBER 4C Max Stress Stress Allowable Combination Intensit;i'. Location 11 Normal OP.(Ax+BEN) 22542 ID Inbcard Process Pipe Hub ( #86) 56100 12 Normal OP.(Trans+BEN) 22300 OD Outboard Process " Pipe Hub (#22) 13 Normal OP.(Trans+Tor) 22439 If If #4 EMER. (AxtBEN) 259115 " " 15 EMER. (Trans+BEN) 25918 " " #6 EMER. (Trans+Tor) 29889 " " PENETRATION 26 AXIAL THERMAL GRADIENT The results cf the axial thermal gradient analysis produced a temperature adjacent to the concrete of 145°F theret:y compliance with an imposed 150°F limit. The* above*results are well within Code allowable stresses as 3 Sm due to the inclusion of discontinuity or seccndary effects. Therefore, the flued heads have been der.cnstra ted to comply with the design information supplied.
., P[N?6CM00lr1EO) Ot:TAIL ) 26 MATEalAL. SD..ECTIOHS IV TT'E NUMtea HllO I l'P()(ESS ,.Ip£ , GUARD ,tn * ..OlZLP: '1'1:
- SuPPOlllT aa.--* UBL[* l .. ALLOWA8L[ STPFSS*JNTrNSITYCSMt SOUCJCE
- USlS Bll.7 fl9f*Q> C*lOHJ) CPC\U . NllCLOR POwER PJPtNG lAOLf A.l MATUUAf.j NR. )
- 5TAINLfS5 ST[EL f'OPGl,..0$ TYP(
- r3n4 JU'* era 10.1 20 *** ,oo.o uo.e soo.o HO.I 650 .. t ., .... '"'*' Allel 11'RfSS ($"') ?0.0000 20.0000 IQ.MOO 17.6000 16.4000 ir..6000 lS.3000 1-..1000 lt..9000 \4ell000 3G-48 * * *
- *
- TABL[ 2
- MODULUS Of* ELASTJclTY<F.> C*lOH6j CPSU TYP£ .. lfMP en 10.0 200.0 30('1.0 40(1.0 c;oo.o 600.0 100.0 f'OO.O TABLE 3 ... COEFf"ICtn'T Of' TtifiJMAI EXPANS.I C*lt)**-!'-)
- USAS 8Jl.7 NUCLEAR P.OW[R PIPJNO ** f ** 2A.34>00 n.1000 ?.6.t,noo ?6.lOOO zi ** eooo 21 ** 1000 'TA8Lf. A.6 FOllGJNGS SOUHCf -USAS 831.7 N**C1.(l.R PtPlNG ll*BLl. Ae '> MAT£AlAL* l
- STA1Nl.ESS STEE.l TYPE -TEMP ff') 10.0 101'1.D 1*,.0.0 i?OCl.ll ?51).0 300.0 1">0. 0 400.0 450.0 ... oo.u S"iO.O 1.00.0 100.0 Hoo.o 3G-49 f' lot. 9.1100 '*e l**OO *1.3400 Q.4100 l).4700 9e5300 *l.6c;oo q. 7000 c;.7600 9.8?.00 <<il.S7(IO 9.9Vi() 9.9900 10.0 .... 00 TABLE 4 -CON{)UCTIVJTVCY) 18TU/11Re*fT e*flJ SOUP.CE
- 811.7 PIPING TARLr A.4 MR. l
- STAINLESS STE(L FOl-IGINGS yt"tlP If) 10.0 100.0 1 "i:l .o 21>0.o 2*:0.11 ioo.o )l;(j. 0 "0(1.0 .o 1:.on.o ssn.n 1.nn.o 6'*o.c 100.0 1so.n TABLE 5
- THERMAL Olf'f'USIYITY (K/CP) C(fT**?)/Hltet f'J04 ** I( ** fle'."'-00 f'.4000 A0'170(1 ... 9000 c:..1 i'*)() n
- 1',IJl'I Q.:"000 10.0000 16."100 10.4500 to. 7001) I0.<<000 11.lOOil 11. '.'!50i) u. 5500 SOUPCf
- USAS B31. 7 Cl 4f' 'H NllCi.f_l\R PO*ER PtPlNt; l *ALF. A ... MATfRIAL NR *. 1
- STAINLESS ST[El f'OHGtNGS Tn.tP fF') 'FO.O 100.0 1*-0.0 100.0 Z'*I)
- o 300.0 401).0 41\t'etl ..,oo.o '551). 0 MlO. 0 6"10.0 701'.0 1*;t:1.o 100.0 3G-50 f'304
- l'/CP * .11* *5 .157.S
- ls*. 'l .l .. 6*> .1;;. olMl .. .lf.**r.
- l"*"" .17.;7 .11z1 .l lli.. ,, , , ... ., .1178 * * *
- *
- TAAL.£* b
- STHESS lMPLITUOESc SOURCE
- USAS 831.7 Cl9o9> SA) CPSU NUCt.EflA PO't1ER PJPJNft FIG 1-1os J.3CA).C8) MATfPUL.* NR. l .. SUtNL[SS STEEL TYPE -AlP2 STf.'fSS<'iA) 2hOOO. '.\7"ir.O. 5*1500. 109000. 2'-0ooo. 6">0000
- TASL!-. l -ALLO-ABLf. STl*ESS-lNTr::NSJTY(SM> (*};)*it.")) jL*'i}) f'J04 tHC'YCLF:S) 1000000.0 iooooo.o 10000.0 1000.0 ioo.o 10.0 sou:cr -USAS All.7 t.:l-"Cl.f PO,.f.iol Pll'P**G TABLF. A.l MATFRtAL HR* 7 -STEEL S[llMLESS PIPE hf't': .. A312 10.0 200.0 300.0 .i.oo .o c;oo.o 600.n 6':>0.0 701.'.0 1so.n l'Oo.o 3G-51 SSIS*') zo.oooo i'0.0000 1q .... ooo 17 .6000 l'i.4000 1*:.600(1 lC:..JOOO 1* .* 1000 l**eQOOO lt141R000 TABLE 7.
- HOf'IVLllS Of' EL11!>TlCITY1r> c*1r**n <P:-n TYPE -Al12 10.0 200.0 JOO.I> 400,0 -:oo.o 600.0 100.0 r:oo.o TP3(14
- USAS PO\!f.P PIPING T AEIL f. A* b SEl'.141.ESS PIPE ** r ** 18.3000 n.1000 z1.1000 7.6,6000 ?'i.t.000
- A-000 zi..1000 l -COEFFICIENT or Ori 1Alf't1AJ c*1 o**-1.J S0U"tf -USllS D31.7 nq, .. <>) t'1CL[**N PO,..fR PJrJt.iG TABLf NR. 7 -STAINLESS STf[L TYPr
- A31l 1(1.(1 ioo.o .c 100.0 zso.o :tno.o ;\<;0,(1 "001u l.SO.r. "*00 .o S'lO *(I **r.o.t1 6':1).0 *10(1.0 7"D*O roo.o TP304 3G-52 0.1100 0.1"'00 Q *. l400 Q,4l:>U ci.4100 o.s 100 9.591)(1 *.1000 c;.76on 9
- Clf"<(l(j tt.'JQOO u.o **00 SEAMLESS PJPE * * *
- *
- 4 -f PTU1H1.*.-fT .-r; SOU!*'<r -US.\S n:i1. 7 ( \q1*.Cl) t:IJCl.Gk Pil,l'!t; Tl.BU' A.'-MATEPJAL NR* 7
- STAINLESS STEEL SEAMLESS PIP( TYPE -Alli? 7t>.O ioo.o l'>O *fl ZOfl.O ?.1i(l. 0 301l. r. 3*;c. * 40000 45(*. 0 50o.<> sso.o I 0(\
- 0 6'il).0 700.0 -(Y/CP> f (fl**2) TPJOft. .. " .. R.3r,oo ;:. .4000 fl..6700 I'* 9:)00 C*ol/{IQ '.'l'i'i(; Cj o .-.(,Oil 9. "C*OO 10.0000 10.23ou 10 * .c.*:*t10 10.7000 lo.*;ooo 1) .1()00 11. 11.S!>OO Sr.UFICE -AH.7 Cl'ff.'1) N11c.t f I"();.'[ R PIP t NC. TARU: A.4 NR. 7
- STEEL SEAMLESS PIPE TYPE'
- 7r. .o tno.n 1 "*{\ .6 zoc.o 2'-*0 .o 300.0 'h001l 41)0.0 4SO.o 1*00, 0 6'*0,0 700.0 7c.('.,.I) 800.fl iP304 3G-53
- l .. 98 .1 ****S
- 1 c;zi; *
- l "','*9
- Jb(ll .. 1f,1n .101<.1 , l6<;9 .1i:,,,,.. .17*'7 .1721 .17*',7 ol 778 TABLE 6
- ALTEPNATlNi. S,\) CPSI> SOUP<'£
- llS 11\S B:H. 7 C1 C>'*.9) t.lllCL f r.R po *. f n p Ip HlC. f JG J.JCa).(BJ MATfPIAL NHe 7
- STAINLESS STEEL SE/tHLESS PIPE TYf'E -Alli? TP304 STr.rss 'SI., ?l'>OOOe s*,c::oo. 1or1000. z*.Qooo. 6"'0000e TABLE l -ALLOliABlE C*H:*e*"H fp*;J) rHCYCl.fS) 1000000.0 100000.0 10000.0 1000.0 JOo.o 10.0 SOU1;Cf -lJS1'.5 fDleT e1c:.i1-'il M1n.Et1R Ptlo'lt*G TA0LF. A. l NR. 9
- ChRBOU SEllMLESS PIPE 10.c 4.0o.o c.,oo.o 600.0 100.0 3G-54 SS C!\l 20.0000 ?o.oooo 1".901)0 17. "\000 11.0000 11 ** aooo * * *
- *
- TJBLE l -MODULUS or ELASTirtTYc[) t) MATEPIAL NRe -STEEL CiPeB -USAS 831.7 (}Q*Qj Mfl".:1.F'\i:I Pn,;rR PJPJNC, TARl f-. A."' SEfl"4Lf:SS PIPt T'ff'f:
- A 1 C 6 fFl ** F: ** 10 .o ioo.o 30(1.0 400.0 soo.o 600.0 100.0 TABL£ l -COEFFICIENT OF THFPMll EKPANSt MATFRIAL 9 * ?.I .9000 27.7000 z*r .4000 u.oooo ?6.4000 z*,. 1000 l'* .sooo SOURCE -usas "]1.7 tHlt:l.f'.hll POritR Plf'JNG TAUl.f A.S SEAMLESS PIPE hPF -t. l n!'I r,I) .e TfMP If) *At.r'HA* 10.0 l* .* 0100 100.0 f,.l,<'O l '*O
- 0 Ii.I' Of; 200.0 , ..* J>i(l 'I o &.*.-<<OO 300.0 6.6000 :t". (i "* 7101'1 4t.l(i.O "'*('l)CI '*'>O *(I 6,c.';>(11) C,l)O. 0 7.CIZOO c:,..:o.o 7, V(IO **OQ .o 7.?.lO:> 6"iO.O 7.3'101) 700.(1 7.4400 71\1). (I 1.s*.09 eon.o 7.6 00 3G-55 TABLE 4
- CflTUIH*-'e*f'T e*f l sou;;*c.E -US*\S R:n .7 (1q.,<;) Nl.'ClPI" P(),.tR PJP\l:G r .l9*.' A,4 MATF.RJAL NR* 9 -SF.t.14L£SS Pl PE TYPF.'
- It l rib TEMP ff I 7o.o ioo.o l'ln.o 2'0 (I. 0 ?<;(J. 0 300.0 J>>o.n *or1 .o 4So.o *.;or.o , OQ
- 0 61>0,o 700.0 1**0.0 TABLE -1HEPM*L OIFFUSIVITY ( ( ft**? ) / ... R , ) .. "' .. )l*,0000 '.'lh.6000 '.l'>. 7fJOO Jlt, 11&:.(; !) 33.?000 3?.3(100 ?9.!'*000 .t,000 i?P.3000 'Z7 ,c:.ooo 1.'*.AOOO lh.0000 sour*cr ... usr-s Bll. 7 09***,; POiiC.R PlPH.('. T Al\Lt: A.4 "ATERtAl NR. 9 -CtPBON STEEL SEAfolLF.:SS PIPE 10.0 tno.o 150.0 200.0 ?1\1),() 300,0 0 1.no.o 450 *(I !:.f'IO.O sso.o "oo.o f>So.o 7r>o.o 1c.o.o i:100.o 3G-56
- l\/(fl *
- 6c;f' !\ .65S'> .6 ... 4")
- 6 :1" .!' .. ,,9
- C:,f.??. "i7
- s ** ,,!;.) .47';'.l * .:.56') .411\7 .417:) ,3'H1'i .31a1
- 36!16 * * *
- *
- TABLE 6
- ALTERN4TH11* STRFSS S.\) sourr;E -us.r.s o:n. 1 119.-.Q, PO-¥lR PlPl!*.G MATfPTAL NP. 9 -CftPBON Sff!MLESS P:PE TYP( -Al06 STf.'fSS(5/lt I 7e O'l, ?ooe:o. )"000. *i:'lOoO. '-rOOOO
- NCCYCLfS) 1000000.0 100000.0 10000.0 1000.0 100.0 10.0 ***** T E P M A L! A N A L Y S I S ***** STEADY STATE CASES TO BE ANALYZEO 0 CASES TO ANALYZED JArK[T DATA THJ('."t.:fss or It-.'.uLntoN ., t.51lOO 1 Hr--tS'>IVITY or ouTF1*' su"F'ACJ" .. .:H)ooooooE*OO IWtmf:11 OF LONC>ITUOIN.AL CAVJT1 A TlON SfliMC:rns r: Tf"'Pf.flATUl'lF.: (f') 41).000 :l00.000 '* 00. t'OO 100.000 BOUN1>flllY CClNOITIONc; CONDUCTIVITY
- 30000000f *00 *llG .s1oooooor.oo *"*000000f *CIO INJTf/I.!. l(t-!PfQAlU1?£ or MOl"\f"L (FJ
- 200.000 11'-l['IrtlT Tf't*r*f'IMTUPf (f., * *1'"*000 LINF fLUID ClrP( lJ or c ,
- VELOCITY OF' llflll.li O"T/SEC.)
- 6.;.oo 3G-57 STATF ANALYSIS Pt<t'.'*':E'"*S PIH' F'ttJIO TE't-'F'f.'IHTllRE <f't Cf) * <*U* PIPE lEU:*fflllTUPE LEN;.rH ! IN, "' ... "IM'" Tt t4f.i'!\TlJRf" tF) * "LL LrNGTlt Of" NOZ71..r (INl .. Olftt!fT* JCAL r>t*:fAl\IC[ or '<ALL {JNJ * "10UUTINt; Rlt'G TEf!PERt..TUkE (f)s RECSUTOR DATA
- rUNC:itON ? 3G-58 .t,SO."C *0.000 '1".000 }i* .00(1 -o.ooo * * *
- *
- STEAOY STATE CASE 1 STEM>Y STATE (",**AflIE*iT er> P1Jnr:rc;s
- 450.0000 M'PIF=NT A pi
- 4*;.0000 C.*YITY IW Ut A *mriF.: I Cl* 35.53)
- 23&.5qq7 PT£RlOQ ,,. tf'7. 7"71. cm1r1;fTE "':ILL * **"*.0000 INTli<N*L NOf*E (Jl'I i **9.7151\ z 449.ASl'Jo 3 449.4;..94 .. 449.7?% s 44;..1t4J6 6 443, 7471. 1 1+4 ... 8 44*'.2177 I'.) lo 441;. 'fl-'71 11 .. )'*.lJ'..7 p 4J7 13 .9499 l4 15 4;1-l.!'1141\ 16 41l.7d74 17 416.7'.JJS 18 l 449 0A"lf*4 2') 77<i4 21 449,$lAl.4 7-2 449. 71 *4 ZJ "'*'i.81)1>4 ?4 449. 7 ii*4 ?-5 44'-i.PA..,4 26 44Q. 771*4-?7 44'1.f.\l:\i'-4 ?8 44<>. 77*'4 ?9 . )I\ 441). 77,-,4 31 449.f!A"4 32 449.71*4 33 44<<;;. ]4 44<f.77'" \5 44'). ,.q.,4 )'.> 44?.77'14 ,1 44*1. (lfl11 4 ;:18 44Q.77r4 39 449.f!ll' 4 40 44*.7704 VJ 41 44q.11,:i.,4 42 44'1.774 4.l 44'l.f'i;.4 44 44'./. 77! 4 45 44'*0A::l.o.7 GJ I 46 44'". 77no 47 48 449.7212 41) 441l.:>llt.1 51) 441'>. c;I' nc; V1 51 44S0Q1lS s ... 4410 CJl+J5 SJ 54 444. 94 55 44"'. ;>?1<.; "° 56 444.411'1 5., 447
- 6:fli 'I 58 445.44 *l S9 4'4*) .f;7i'8 l>(J 7 34 J "' 4:i-.. fl) l 8 '*?. 442. 7437 f.3 415. l'>J'l b4 4'.l*1.1r.1z '>':> 4?1.1"13 4J7. )11) ,,7 l 66 4J4.7l'lb b9 4 \:"." "l l 70 441.1111 71 4 )l\
- f)"'B'-' 1'l 4'.\'*.1)217 7l 74 71_, lo?6. )(*0 1t> 425.U!B 71 1tr1 .?**f15 'TB 4;:c.i
- 1 J2t> 7r, 1+n. n1*1 80 4)'l.'41 J\ ,, l *is ... ll2 4J9.7J/5 aJ 4]1.9**95 f\4 417.7l'HJ Ac; i.oo.1no fl6 1\7 .199.A799 86 4'(',.2Q) I 09 41?.><17S q,1 42i'. 7l l" 91 414. C)") ls 9l. 4,*,,a4fP8 93 JG1> .* 94 )<;*11.o'TJrJ Q<:, 371-*._" <>e, 37J.(1:14J '*7 377.6211'> 'JS 3FI)
- 33 liJ 99 ..,r;7.i:4i*A 100 417, *1-;n 101 *07.91.or; lr.? 1Q6.66Jlt lnl J*'-,.}F*61 }1;4 't14. 71'>1 L; 11)5 lel6 ln7 J.,4. 7Jj1) lf.f.l 37,*.J'i1.,9 1!19 J*;1 .. ll:l 31.'** _,., 16 111 .1?:....11>>7 )I?. J?'->. '*f-41 lP 3*4.1711 114 J. "'* fl4,'l 1J5 )Pl)* f*,;>J u-, 21' A* 117 4 118 zn .r; ul 11-i< 21-1n.1'37l I ?IJ 27,_. 77l.?. lZl Z76.!-716 122 277.J4'1'j 123 21h *. J;>r19 124 27'\.f1'Hl 12'> z11.r.2ri7 126 Zf." *'le; 1 127 Z'*". rr. Z 1 !2B 2n.q,..,5 129 ?.,r-.Q.44f17 , ,,.. 2'*"*' "17.< 131 2"J.46Jt! )12 Z63.'>'*"7 }33 134 l7".;;r;sr 2'*'). ?'* ;, 2 l'.:\6 2till.4':Jf,l 1.*7 l 91. y? 1:1!1 1 *,4 1'19 19?..;. ?l l4U PlJr;'5 141 193.9)0:.*' 142 lqJ.I' .\'-9 143 1"'**3"7-.1 144 l' i:.. ;>"i5':> 14'> 14'> 199.)')7"' J47 2'lJ.91.l9 148 2: J.'l?'.;l }49 2*,?.1*)4.'.t l ':>0 20*1." i'7t I'll 217.0'l'jf) 1i..z 21 7 *I)'>;":; ??5.1)6:111 1">4 7iS.15n4 2J:i * '* r, .,a; 156 212.1or;'>1 J!i7 ?. H.4414 1S8 21'l.4 I <;6 }51.J 2 4':). l "'" ?4'i. ,, t) jh 1ft1 21>11. J.<')7'1 lh2 ?64.3?47 } hJ 2:.7. ;.<.lt:i lM 2$1"' .* ,, .?'*? 11.s 106 2"4.5151 ?;4.42?.9 }68 2* 609}EIR pn 171 z11.1z,.z l"/2 14) .':i9 .. 6 173 174 l1*t l l4'.*"4'3'l 111'> hZ.t-154 177 147..401><) 178 14 } *);>2 l1'J 141.1917 l8'J 14.1.Qf.)9 1f!1 au.1n1.1. )tl2 144.11771+ 183 144. 7;174 l'"** \ 4S *' r; l '> \4'i.41l(\ }A6 J!i7 1'*f .* 1"56 lM 141'>.88?5 )89 }t.6. 71>\fl 1<H* l "h." q2c;
- APPENDIX 3GS REPORT NO. 71983-6 *
- 3G-60 w (j) I °' ..... * **-. -,.-ENGINEERING DEPARTMENT 11/l:l'Oln' -7198 3-6 fLUED HEAV S1RESS REPCRT FOR FLORIDA POHR AND LIGHT COMPANY ST. LUCIE PLANT (FORMERLY HUTCHINSON ISLAND) CONTAIRMENT PIPJNG PENETRATION ASSEMBLIES P.O. NO. NY-422264 Included Are: Penetration No. 1, 'afn Steam (SG-1A). I-34MS-28 No. 3, feedwater (SG-1A), 1*20BF*14 Prepared by: /Uf 11.s. Haerman /.!. /.I * , Approved by:*?,/' . .: * 'f M. V. Mal mus Date: 7.lv& 13 Date: "iJ/' ;7 J ; Although stress report fs believed to be accurate, nothing contained herein shall be construed as ing a warranty, express or implied.
- lfil!!!!.' .. !!!..HL-9-1. -ENGINEERING DEPARTMENT Section I 1. Report 2. Type I Assembly Drawing l'll:l'OllT NO. Section II-Main Steam (Pl, P2) Computer Results I. Thermal Model Plot 2. Stress Model Plot 3. Material Property Data 4. Thermal Results 5. Steady State Thermal Stress Distribution 6. Internal Pressure Stress Dfstrfbutfon 7. Axial Unit Load Case 8. Torsional Unit Load Case 9. Shear Unit Load Case JO. Bending Moment Unit Load Case -left End 11. Bending Moment Unit Load Case
- Right End 12. Nozzle Shear Unit Load Case 13. Combined Stress Case 11 ( o. 8. E.) 14. Combined Stress Case IZ (O.il.E.) 15. Combined Stress Case 13 (0.8.E.) 16. Combined Stress Case 14 (O.B.E.) 17. Combined Stress Case #5 (D.B.E.) 18, Combined Stress Case #6 (O.B.E.) 19. Combined Stress Case 17 (O.B.E.J Section III -Feedwater (P3,P4) Computer Results 1. Thermal Model Plot 2. Stress Model Plot 3. Material Property 4. Thermal Results 5. Steady State Thermsl Stress Distribution 6, lnteral Pressure Stress Distribution 7. Axial Unit Load Ca1e 8. Torsional Unit Load Case 9. Shear Unit Load Ca1e 10. Bending Moment Unit Load Case -Left End 11. Bending Moment Unit Load Case -Right End 12. Nozzle Unit Case 13. Combined Stress Ca1e #1 (O.B.E.) 14, Combined Stress Ca1e #2 (O.B.E.) 15, Combined Stress Ca1e #3 (D.B.E.) 16. Combined Stress Ca1e #4 (0.8.E.) 17. Combined Stress Ca1e 15 ( D.B.E.) 18. Combined Stress Ca1e #6 (O.B.E.)
- Vo> ? O' N .._ (ii)!!!!.E_!!!!S_ 9 ....... 1.anuarr ENGINEElllNG DEPARTMENT tlltll"ORT NI), 71983-6 Report Summary: Introduction The following. when combined with the balance of ltems listed In the Index, constitutes the stress report on the two (2) flued head configurations utilized In the Type I penetration p1pln9 assemblies for St. Lucie No. 1 (for*erly Hutchinson Island). All c1lculatlons for stress distributions, herein, were made using f1i1te element computer techniques. Descriptive data relative to the program and its verification is found ln Sectlon 1-3 1-4 of Report No. previously submitted. Basic Approach For analysis purposes the flued head and attached pfp1ng model was assumed 1s shown below: l **AO fTI/. strC NO! c=::i:" l-----,.,. ... tlloy I 1NSflLlf'f'*IJ ,,,./ Cl : rw) 1'8" t.,rrR " The external load combinations considered are shown below. These were specified by Ebasco tn a memo dated December 22, 1972 are assumed to act at the flued head centerline and can be produced by either the right or left hand process pipe lnthe combination stresswise: oa he rm a (D. B.E.
- 60 60 120 1300 100 lOO 450 800 450 2 3 * .. _ @1Ull 1WNS ....... _....__ _,,_,,
- ENGINEERING DEPARTMENT --71983-6 MAIN STEAM oa MOMEN1 (FT.KIPS) (FT .KIPS) 2560 q5o o 120 100 1150 7
- w c;") 0\ w * * (H;)!!!.' .. !!!!!L-9-ENGINEERING DEPARTMENT H:PORT NO. 71983-6 FEEDVATER Load Combinations Axial Trans Bending Torsion Case (KIPS) (KIPS) Moment (FT.KIPS) NO, 'thermal & "elsmlc 44 66 (FT.KIPS) 226 1o 1 (O.B.E.) 19 79 320 10 2 19 220 110 3 'Ihermai & S'elsmlc 69 6 220 10 4 (IJ.B.E.) 19 98 420 10 5 19 98 220 210 6 For the actual stress runs, the above loads were converted to statically eiuivalent loads applied to the ends of the cal model shown pre*riously. The actual loading data used is shc.wn be 101o1: Load ! KAIN STEAN (Pl, P2) 2 l !!. 2. 6 7 FR -120000 -40000 -40000 -200000 -4000C -40000 -FL -120000 -40000 -40000 -200000 -40000 -40000 -VL -60000 60000 -120000 120000 --1200000 -156'l0001 -120000< --1200000 -3000000! 1200 -1200000 -12950001 1411000( -408000C -21172000! 1200 'IR -5400000 -5400000 -960000( -51100000 -51100000 -1380 TL -5400000 -540)000 -9600001 -5400000 -5400000 -1380 VII -27 ,273 -384500 -57280 --87273 -741800 -2727'* PRESS 885PSIG 885PSIG 885PSic 885PSIG 885PSIC 885PSIG 885PS 1 520°F s2o*F 52o*p 520°p 520°F 520°F 520°Fi I * @!!!.'..!!!!5-9-'* -ENGINEERING DEPARTMENT ltCPOWr NO. 7 ::_ 3-t PEEDWATER (P3, P4) LOAD ! g_ 1 4 2. §. . --* -*. *-*. --*--*---------**--** ... *--. ---* --------FR -19000 -19000 -59000 -19000 -19000 FL -19000 -19000 -59000 -19000 *190CO VL 6C,OOO 79000 79000 96000 9eocc MR -2640000 -3840000 -2640000 2540000 -5040000 -2640000 ML 0 -364000 836000 0 -728000 16720CO TR -120000 -120000 -1320000 -120000 -120000 -252CC20 TL -120000 -120000 -1320000 -120000 -120000 VN -soooo -126770 -99500 -163550 -109000 PRESS 1C50PSIG 1050PSIG 1050PSIG 1050PSIG lCSOFS:il S.S. 440°F 44o0p Thermal 4ll0°F 440°F The nomenclature and sign convention for the above is as follows:
w ? .c::--9-ENGINEERING DEPARTMENT ltal'Oln'NO. 71963-6 PR
- Axial load rt. end of process pipe (lb.) PL
- Axial load 1eft end of process pipe (lb.) VL
- Transverse load left end of process pipe (lb.) MR
- Bending moment end of process pipe (in.lb.)
- Bending moment left end of process pipe (in.lb.) TR
- Torsional moment right end of process pipe (in.lb.) TL
- Torsional moment left of process pipe (in.lb.) v11
- Transverse load outer nozzle (lb.) To arrlve at the final product a set of unit run cases was followed by a set of combination routines using appropriate multjpllers. It be noted pipe rupture ccnditions and the balance cf major penetration colll()onents have been evaluated in Reports 71983-1 and 71983-4. ' Swr.r.tarz of Results The results or the two (<)analyses are summarized below: Mair. Ste811 (Pl, P2) Combination Max.Stress Stress AllowablE Intensity( PSI) Location l 26203 E1'90-Head Body 611U I.D. Left Side 2 36732 Ell62 O.D. Left 61060 Process Pipe Hub 3 25806 EU90 Head Body 61141 I.D. Left Side 4 26110 E1190 Head Body 6UIH I.D. Left Side * * .. _ TUINI --------ENGINEERING DEPARTMENT *--719S3-MAIN STEAM (Pl1P2) Combination Mu. Stress Stress Allowabl* Lo:ation Stress ( p;j!) 5 25950 E1190 Head Body 61141 I.D. Left Side 6 57626 Ellll4 O.D. Left 61442 Process Pipe Hub 7 25338 EH90 Head Body 611U ID Left Side FEEDl(ATER (PJ,P4) 1 32205 ElllOO Head Body 60520 I.D. Left Side 2 34763 ElllOl Head Body 60473 I. D. Le ft Side 3 31736 El#lOO Head Body 6C520 I.D. Left Side 4 32964 El#LOl Head Body 60473 I.D. Left Side 5 36166 ElllOl Head Body 60473 I.D. Left Side 6 31944 El#lOO Head Body 60520 I.D. Left Side *
- : : : } : : M >> "' ' '-'
l 1 ! 1 l l : ; l l -----,-,.,*, :.;.-* ':..f' '" 0"' "' => .. .J '"' ;:;V) t; ' g -""11 iIIID J!l1 l '1lID r l:!I """' 1 .. --cilj iii = lllt '"' '1JJ 'N ;lJ '"i r-*d "' *--o
- l. r g "' "' ' "
-<:::}-------:!ffihlJ r* i. \* jc I \' I j. i* j. t ' g
l
l
l
E. CONCLUSIONSThe reactor support system will withstand the combination of loads postulated in Section C andsupport deformations are within allowable limits. The overstressing condition identified in the application of load combination 'g' would result in additional concrete deformations. These secondary deformations would reduce the thermally induced loads caused by the heating up of the steel beam-column assembly and would result in movement of the reactor supports of less than 3/32 inches. This would not result in an adverse effect on the primary piping. This condition is considered acceptable from a safety standpoint. Therefore, the reactor support system can accept all design loading conditions postulated.F. THERMAL SHIELD REMOVAL AND CORE SUPPORT BARREL REPAIR EFFECTSThe removal in 1984 of the reactor thermal shield from the St. Lucie 1 reactor vessel will produce anincrease in the radiation level in the reactor cavity area, and consequently in the temperature. Temperature increases in both the primary shield wall concrete and the steel reactor support structure will be experienced.Extensive analysis have been performed to determine the effects of various thermal loadingconditions in combination with the other loads imposed on the reactor support structure. It was recognized that the removal of the reactor thermal shield necessitated a reinvestigation of some of these previously evaluated cases in order to demonstrate that the earlier conclusions as presented in Sections "A" through "E" are not altered by the present modification.The earlier analysis had indicated that two loading cases were particularly critical from the structuralpoint of view. The LOCA condition, presented as combination C.1.a(d), when later modified by the addition of the North Anna Syndrome loads developed by Combustion Engineering (CE) represented the worst loading case for the concrete primary shield wall structure. The appropriate temperature distribution for this condition is the normal operating steady state condition. This loading case also resulted in some local yielding of the steel reactor support girder, however, this was considered minor and not warranting re-evaluation for the present investigation.What was considered critical for the steel girder-column assembly was the vertical deformation, whichcould potentially impact on the CE stress analysis for the reactor coolant piping. This parameter was clearly sensitive to changes in the thermal distribution. The appropriate loading combination was the loss of fans condition, presented as combination C.2.a(d). The minimum required operating fansconsist of a sufficient number of containment fan coolers to maintain containment atmospheretemperature 120oF, 1 of 2 reactor cavity fans, 1 of 2 CEDM cooling fans and 1 of 2 reactor supportcooling fans. That condition was later modified and re-evaluated by allowing the time for operatoraction following fan failure to increase from 15 minutes to 45 minutes. If the minimum requiredoperating fans can not be restored within 45 minutes, operator action to trip the reactor is required. Itis this latter case that was considered in the present analysis.This analysis presents the evaluation of two cases: (1) the combination of the new thermal loads withthe previously evaluated North Anna Syndrome loads and load, equipment, subcompartment pressure and seismic loads; (2) the vertical growth of the steel girder-column assembly under thermal conditions generated by a loss of fans in combination with the new radiation heating conditions. 3H-19 Amendment No. 16, (1/98) F.1 ENERGY DEPOSITION RATERemoval of the reactor thermal shield increases the radiation level in the reactor cavity area,and consequently, the temperature. CE calculated the new energy deposition rate radially outward from the reactor vessel in the reactor cavity, and a partial distance into the primary shield wall concrete at the elevation of the reactor core centerline. A deposition rate in the reactor vessel support column was also determined. Ebasco estimated the vertical energy deposition rate profile in the support column based onmeasured Unit 1 neutron flux data in the cavity. Assuming the energy deposition rate will fall off in the same manner as the measured fast neutron flux moving up or down from the core centerline, the approximation is made that the deposition rate will decrease from 1.3x 10-3w/cm3 in the support column at the centerline, to 2.9 x 10-4 w/cm3 at the top and bottom of thecore (a factor of 4.5 decrease). These energy deposition rates became source terms for the thermal analysis portion of the study.F.2THERMAL ANALYSISThe effect of the enhanced radiation heat load on temperatures in the reactor vessel supports and the primary shield wall is examined using several simplified heat transfer models. The results show a peak temperature in the primary shield wall concrete of 1480F at the elevationof the core midplane and about one-half foot into the wall, reflecting the influence of the input radiation heat load. Similarly the vertical reactor vessel support leg shows a peak temperature of 1410F at the elevation of the core midplane, falling to a steady 1240F below thecore. Above the core midplane a similar temperature gradient is seen until the elevations where the influence of the reactor leg is felt. As the top of the horizontal vessel support is just above the top of the core the temperature over the previous detailed transient heat transfer studies was used as the basis of the temperature distribution for the transient conditions examined in the structural analysis.Two models were constructed for analysis by the heat transfer code HEATING5. The heatload derived for the CE data was placed in: (1) a two dimensional model of a section of the primary shield wall; and, (2) a model of the vertical support leg and concrete below the leg. It was seen in the previous heat transfer studies that the influence of the RV leg on the support was restricted to the area immediately around the interface between the support and the shoe. The elevation dependence of the heat load will then impose the greatest loads below this elevation, within the vertical support leg. To study this effect a detailed model of the vertical support leg was constructed and the heatload imposed as a function of elevation. Convective heat transfer to 1200F air was assumedon exposed surfaces with a coefficient of 0.85 BTU/hr-ft2-0F. Radiation to a 1500F surface(either the RV insulation or the surface of the now-heated primary shield wall) was assumed on the appropriate surfaces. Steady-state runs were made with these models for these conditions.3H-19aAm. 3-7/85 For the vertical support leg model, Figure 3H-30 shows the centerline temperature as afunction of distance from the top to the bottom of the support leg. The distribution shows the effect of the heat load, with the peak temperature occurring at the centerline elevation of the core, and failing off symmetrically in both directions from the centerline to a temperature of 124OF below the core and in the base concrete. This result suggests that the influence of theadditional heat load on the horizontal support will be to raise the temperature in the exposed steel from 5 to 10OF in the high temperature region below the interface with the support foot.These results were used as the basis for the study of the 45 min loss-of fans transient case. The increase in temperature seen in the supports and shield wall between the present study and the previous analysis as applied as an increase on top of the transient conditions, as the heat load from the core will not change during most of the transient. The change of the thermal gradient will change the rate of temperature increase in the leg and horizontal support by limiting heat flow from the support foot; on the other hand, temperatures will slightly increase when forced convection ceases and turbulent natural convection from the exposed surfaces begins. Increases in temperature over the original analysis are shown in Figure 3H-31.F-3 DEVELOPMENT OF STRUCTURE MODELThe temperature distribution in the steel girder-column assembly and the concrete primary shield wall were developed as described above for the normal operating steady state condition. The temperature gradient in the mass concrete was then determined assuming a straight line gradient between the primary shield wall temperatures and the exterior concrete surfaces (assumed at 70OF).Vertical and horizontal models of the reactor cavity as previously described were used for analysis. The EBS/NASTRAN cracked element program, developed by Ebasco and linked to the commercially available NASTRAN program, was used.Using the developed temperature distribution in the vertical model (Figure 3H-32), therestraining effect due to structural stiffness and vertical growth of the lower mass concrete on the two dimensional horizontal model (Figure 3H-33) was obtained. This effect was represented in the horizontal model by a temperature drop in the fictitious radial beams located around the outer periphery of the primary shield wall. (The effective "spring constants" for these radial beams had been determined by the application of a range of horizontal loads to the vertical model at the elevation of the steel girder.)The temperatures and LOCA loads (including pressure loading on the cavity wall) wereapplied to the horizontal model. The resulting total radial horizontal forces in each sector were obtained and proportioned for a 17O arc, later to be used for the 17O arc vertical model. Aniterative process was used whereby these resulting radial forces from the horizontal model run were then used to redetermine the radial stiffness or "spring 3H-19bAm. 3-7/85
Finally, the vertical displacements of the mass concrete and concrete pedestal under the steamgenerator sliding base support were subtracted to obtain the desired differential vertical displacement.F.6 CONCLUSIONThe maximum horizontal load on the primary shield wall is carried by sector III of the horizontal model. The maximum tensile stress in the vertical reinforcing steel is 27.6 ksi, which is within the allowable of 0.9 x 40ksi.The differential vertical displacement of the reactor support girder relative to the top of the steamgenerator pedestal is 0.184", which is within the allowable of 3/16."The foregoing analysis therefore demonstrates that the removal of the reactor vessel thermal shielddoes not alter the conclusions presented in Sections "A" through "E".REFERENCES TO APPENDIX 3H1.Letter R. E. Uhrig (FPL) to D. K. Davis (NRC) Re: St. Lucie Unit 1, Docket 50-335, ReactorPressure Vessel Support System, L-77-265 dated 8/30/77. 3H-20Am. 3-7/85
- ;;o m .,, r )> 0 n ;;o -I 0 0 ;;o )> n "' \J -4 0 )> rm -n -I c: ;;o Cl -< C! Ii"> c n mr ::0 c .,, -m -I r C> )> >I w :*. z -I :I: )> ...... n .!... -< 0 ;;; 0 \J )> m z -I -< ;;o -n w I ....., ....... co VI * * * *
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY PLAN AND SECTIONS FIGURE 3H*2 A. NORMAL OPERATIOH WITH OPERATING BASIS EARTHQUAKE W = D + L 1 +A + OBE + T 1 NOTE: LOADS SHOWH ARE LOADS TRANSMITTED FROM REACTOR TO REACTOR SUP PORTS. -.... & .... f , ______ _
- 4: -4 . . "' . PLA...i FLORIDA POWER & 'LIGHT COMPANY ST. LUCIE PLANT REACTOR SUPPORT °LOADING DIAGRAM -SHEET 1 FIGURE' 3H-3
- B. LOCA CASES: W = D + L 1 + A+ P + Q + DBE + T 1 1. COLD LEG SLOT 2. NORTH OR SOUTH HOT LEG GUILLOTINE NOTE: LOADS SHOWN ARE LOADS TRANSMITTED FROM REACTOR TO REACTOR SUPPORTS. {). . b --------------p A FLORID.A POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR SUPPORT LOADING DIAGRAM -SHEET 2 3H-4 N ----C. LOSS OF FAHS: W=D+Ll+A+T4 D. POST LOCA CASE: W=D+Ll+A+T3 LOADS SAME AS LOSS OF FAHS. HOTE: LOADS SHOWH ARE LOADS TRAMSMITTED FROM REACTOR TO REACTOR SUPPORTS. LEGEND': D DEAD LOAD L 1 EQUIPMENT DEAD LOAD p Q LOSS OF COOLANT ACCIDENT PRESSURE LOAD LOSS OF COOLANT ACCIDENT PIPE OR EQUIPMENT LOAD A NORMAL EQUIPMENT OR PIPE ANCHOR LOAD T 1 NORMAL OPERATING THERMAL LOAD* OBE OPERATING BASIS EARTHQUAKE LOADS DBE DESIGN BASIS EARTHQUAKE LOADS T 3 POST LOCA THERMAL LOADS T 4 LOSS OF FAHS THERMAL LOADS 4 *A A b PLAN FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR SUPPORT LOADING DIAGRAM -SHEET 3 FIGURE 3H-5 *
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR BUILDING INTERNAL CONCRETE -PLANS AND SECTIONS MASONRY -SHEET l FIGURE 3H-6 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR BUILDING INTERNAL CONCRETE -PLANS AND SECTIONS MASONRY -SHEET 2 FIGURE 3H-7 *
- *
- 11,400 CFM FROM CONTAINMENT G.D
- G.D. J>,000 CFM, *o So F FROM MAIN RING REACTOR SUPPORT COOLING SYSTEM (ONE FAN STANDBY) HEADER SS,200 CFM = t-TO CONTAINMENT ffi A TMOSPH ERE 55,200 CFM ...,...__2'.w ! ffi REACTOR CAVITY COOLING SYSTEM (ONE FAN STANDBY) EL. 18' REACTOR CAVITY SUMP PUMP SHAFT LI. u Q 8 <:z:: e:; CONTROL ROD DRIVE MECH COOLING SYSTEM \ (ONE FAN STANDBY) '\\I MISSILE SHIELD u.. ' \ \).. 0 ti) ...J !? .J )( < 0 ;a: a:: Q. 0 Q. ...J < !!:! ::c Vl CONTROL ROD DRIVE MECH ,-COOLING SHROUD I -----EL. 36' CONTROL R DRIVE MECH I I I : I \ ...... "' ' \ ,...._ ____ _...... \ 't OTHER SUPPORT STRUCTURES REAC COOL.PIPE(TYPl --_______ u APPROX 120 °F R >-E v Ci::: E < A ;! c a::: T a.. 0 R s s E L OTHER SUPPORT STRUCTURES CEDM cooling based on 14 x 14 fuel assembly requirements. Compatibility with 16 x 16 design will be evaluated. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT Al R FLOW DIAGRAM FOR REACTOR CAVITY, SUPPORT LEG AND CONTROL ROD
- DRIVE MECHANISM COOLING SYSTEMS FIGURE 3H..S FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY REINFORCING SHEET 1 FIGURE JH.9 * *
- *
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY REINFORCING SHEET 2 FIGURE 3H-10
° 19 *
- SECTION 8-B t----l-----Kl-+--ST!:El. £1.JO Pl. .i. TE EL1&.oo' l FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY REINFORCING SHEET 3 * *
- S'Z1,o)
- UNITS PSI
- FROM $Pl...\T TENSIU: MOHR. t!MVEL.OPE FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT MOHR'S FAILURE ENVELOPE PLAIN CONCRETE FIGURE 3H-12
'SHlELO BUILOING WA.LL-.-I j-='":.:, t=====r.:' =:::::L=r--..J I STEEL \/ \L L 0 WE R R \ N Ga --"'"'t---------MOOE.L) REACTOR C,\VITY -.... UPPtR Rlf\lG (M .. IN MODEL) 'J E'&<.1' \CAL CSL ICE MOOt:l) MA':>5 cot.Jc RE'TE SECT\ON . /,!:.. \ 'j:,' .. *. .0. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAl-IT REACTOR INTERNAL STRUCTURE ISOMETRIC SECTION FIGURE 3H-13 *
- N -* 1 -"' " ' I ... * -+ -*
- PRIMARY SHIELD WALL
- REFUELIMG CAMAL ' . \ \ '\ \_ ,,.,.---SUPPORT BEAM I 1 I I T. r T T 1 T
- 1 11. .. FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAl-4T REACTOR SUPPORT MAIN HORIZONTAL MODEL FIGURE 3H-14 11.0 113 101.s IOI 88 02.5 76 TO 110 110 IZO %%%110 l'2.Q 110 l'Z.O 110 110 %%% 1tt.I 111 110 l'lO %%% ; 111 ltO '20 It % I '2'5 1'1,t I" l'ZO 1.0 I _%" % I t'5 ... /t ;(, 0 I % .. ltO I 110 l'lO '% '" 120 l'Z'5 I I/ IH 110 no t 7,:\ llS ll"T 11. '5 11 14 114 11 'Z. 17.S I'? I 110 108 11.l.9 114 °F 106 104 1'2.0 no ('T'fP) 113 IOG q9 97 99 q4 89 86 88 es 8'1. 81 7S 14 74 10 10 70 10 VERTICAL FINITE ELEMENT MODEL WITH NORMAL TEMPERATURE DISTRIBUTION 1'2.0 l'ZO llS 109 1013 IOI qs S7 61 13 10 iO 1'20 1'2.0 .ltO 114 113 112 101 104 102 103 IOI 98 99 97 q5 9.3 91 89 84 83 79 19 78 73 13 13 70 10 10 110 120 1'20 110 109 108 89 99 97 94 81 93 91 ao qz qo 86 78 87 8<0 84 76 82 81 80 7'5 77 17 13 71 72 71. 11 "T 0 10 10 FLORIDA POWER & LIG'1T COMPANY ST. LUCIE PLANT VERTICAL FINITE ELEMENT MOJEL FIGURE 3H-15 "TO iO iO 70 10 70 70 10 70
- *
- LL b -d' RING SPRING (T'f?.) SOIL SPRING C ) l I U\ '1.. -U\ 0 0 ,.... 0 -4 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT LOWER MASS CONCRETE BOUNDARY CONDITIONS 3H-16 171 i;,.' Til' Tir* i:* 82 +. 6t r r '\O ' FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAl'-11 PLAN-HORIZONTAL FINITE ELEMENT MODEL FOR LOWER RING FIGURE 3H-17 * * *
- I -20.01 31 * ->c1 -4.62 3.08 10.77 NORMAL OPERATING-STEADY STATE U ... 1.5 (D + L' +A+ Tl) uo.llTS: l'-i l"EET FOli'C& 11-l W::IPS ' -t -COMPli'E:';,')IOIJ FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY WALL DISPLACEMENTS AND FORCES -SH. l FIGURE 3H-18 20.01 N ,,,,,..--== -12.31 -4.62 3.08 NORMAL OPERATING WITH OPERATING BASIS U-1.S(D+L'+A+Tl) + 1.5 (QBE) CONCRETE TEMP .. 12oof 1.J1-.1rrs, 0tSP1..Aa:l.1$.li 11.1 FEcT FOl?Cc 11J' ICIPS I.JOT!:';;.: + lclJ':>IOM -COMPi'E:S')IO!o.! 1 0 . 7 7 -xi 18.47 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY WALL DISPLACEMENT AND FORCES SHEET 6 FIGURE JH-19 * *
- *
- t t .. *i>H '\ -!llM 1 -4. U = Tl + l 15 (D + L' + A + P + Q + LOCA -SOUTH HOT LEG GUlllOTIME BRUK (CRITICAL lOCA LOADIMG) FOl?CtS l>J onPS N FU.T + IE'-1<;\0t..!
- COMP!C'IE.'i.<;10>J ';/-<;! -* _ *-/-* *---{ '><. * .... Xy* ooot* ->c1 <11 .....,
- I 10.77 l 8 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY WALL DISPLACEMENT AND FORCES -SHEET 5 F 1 'W \,H"\. ii..
-20. N 4. POST U)CA -STEADY ST ATE U = D + L' + A + T3 wrn. : i:-02ci;,o;. ,,. .:11>$ D1S<"\.i..C\Mt,\,ll'!> ".l Ff:E.T i.IO'Ttii s. : .. .,..,.. .. ,Ol.l -COMl'"C:&.. '5.S.10 I.I ,/ I I / ' -.,.5 -,,,, --11 . ,. ---... , _ -2oH -1u ... 1 0 . 1p
- w
- FLORIDA POWER & LIGHT COMPANY ST. LUCIE REACTOR CAVITY WALL DISPLACEMENTS AND FORCES FIGURE JH-21 * * *
- * -20.01
- N-=====-----12-31 i ,) -xi ;-0 /l ..... .': ------7f>b _::._7 ____ --{)(,*. oo(')o X7* -1& -4.62 3.08 10.77 LOSS OF FAN -STEADY ST ATE U==D+L'+A+T4 Ft.(T ... T'£.l>il$\Qt.,l -COMPftSSIOlol FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY WALL DISPLACEMENT AND FORCES -SHEET 3 FIGURE 3H*22 l I:'" ... , -t"' * -20.01 -12.31 -4.62 /
- 11111 * ><, ....... -x. ** ooote
- 3.08 10.77 START-UP TRANSIENT U = D + L' + A + T2 : ... "'P!. OISPl..AC.cME:lol"T'!. l"J nE."T uati:.'\. : + 'T!.lolS\00.J -COMP"1!$;,.:> .. 1 8 . 4 7 *
- FLORIDA POWER & LIGHT COMPANY
- ST. LUCIE PLANT ,____ __ _............. REACTOR CAVITY WALL DISPLACEMENT AND FORCES -SHEET 2 FIGURE 3H-23
- N .....::::;;...__ ___ _ CONTOUR LEVELS A -50.0 KSF = -347 PSI B -40.0 KSF = -278 PSI C -30.0 KSF = -208 PSI D -20.0 KSF = -139 PSI E -10.0 KSF = -69 PSI F 0.0 G 10.0 KSF = H 20.0 KSF = I 30.0 KSF = J 40.0 KSF = K 50.0 KSF = 69 PSI 139 PSI 208 PSI 278 PSI 347 PSI NORMAL OPERATING-STEADY STATE U = 1.5 (0 + L' + A + Tl) FLORIDA POWER & LIGHT COMPANY ST. LUCl-E PLANT REACTOR CAVITY WALL PRINCIPAL STRESS PLOT -SHEET 1 FIGURE 3H-24 CONTOUR LEVELS ----A -40.0 KSF = -278 .PSI B .... 20.0 KSF = -139 PSI LOCA -STEADY ST ATE c 0.0 D 20.0 KSF = 139 PSI U = Tl + 1. 15 (O + L' + A + P + Q + DBE) E 40.0 KSF = 278 PSI F 60.0 KSF = 417 PSI G 80.0 KSF = 556 PSI H 100. KSF = 694 PSI I 120. KSF = 833 PSI J 140. KSF = 972 PSI LOCA -SOUTH HOT LEG GUILLOTINE BREAK (CRITICAL LOCA LOADING) FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY WALL PRINCIPAL STRESS PLOT -SHEET 2 FIGURE 3H-25 * *
- CONTOUR LEVELS A -60.0 KSF = -417 PSI
- B -40.0 KSF = -278 PSI C -20.0 KSF = -139 PSI D 0.0 E 20.0 KSF = 139 PSI F 40.0 KSF = 278 PSI G 60.0KSF= 417PSI H 80.0 KSF = 556 PSI. LOSS OF FAN -STEADY ST ATE U = D + L' + A + T 4 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT REACTOR CAVITY WALL PRINCIPAL STRESS PLOT -SHEET 3 FIGURE 3H-26 CONTOUR LEVELS A -60.0 KSF = -417 PSI B -40.0 KSF = -278 PSI C -20.0 KSF = -139 PSI D 0.0 E 20.0 KSF = 139 PSI F 40.0 KSF = 278 PSI G 60.0 KSF = 417 PSI H 80.0 KSF = 556 PSI I 100. KSF -694 PSI J 120. KSF = 833 PSI POST LOCA -STEADY ST ATE U = D + L' + A + T3 FLORIDA POWER & LIGHT COMPANY ST. q*CI E PLAMT REACTOR CAVITY WALL PRINCIPAL STRESS PLOT -SHEET 4 FIGURE 3H-27 * * *
- *
- N I / CONTOUR LEVELS A -10.0 KSF .. 69 PSI B 0.0 C 10.0 KSF "' 69 PSI D 20.0 KSF "' 139 PSI E 30.0 KSF .. 208 PSI F 40.0 KSF -278 PSI G 50.0 KSF "' 347 PSI H 60.0 KSF .. 417 PSI I 70.0 KSF "' 486 PSI J 80.0 KSF .. 556 PSI 5 TART -UP TRANSIENT U = D + L' + A + T2 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAHT REACTOR CAVITY WALL PRINCIPAL STRESS PLOT -SHEET 5 FIGURE 3H-28
-y . t 2EACT02 VE 5$EL 1 MOZZ Lt: I L SYM A!>T 4. -NOTES:
- DARK LINES REPRESENT STIFFENERS AND FLANGES PERPENDICULAR TO XY PLANE.
- MODEL CONSISTS OF 476 QUADRILATERAL AND 64 TRIANGULAR PLATE ELEMENTS
- ELEVATION . FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT BEAM -COLUMN REACTOR SUPPORT FINITE ELEMENT MODEL FIGURE JH-29 * * *
- * * *
- 145 u.. <.:1 140 w e w a: :::> I-<t 135 a: w 0.. :E w I-0: 0.. :::> 130 en w 2 ...I a: w I-2 w 125 () <t_ OF CORE BOTTOM OF CORE 120 0 2030 40 60 80 100110120 140 160 180 200 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 DISTANCE DOWN VERTICAL LEG (INCHES) VERTICAL RV SUPPORT LEG TEMPERATURES 300 Am. 3-7 /85 Figure 3H-30 CE -EBASCO INTERFACE TEMPERATURE (230°F) REACTOR SUPPORT I FOOT t 295° T/GIRDER EL. 25.29' 121 t 136° 132 ' 146° GIRDER EL 22.76' 162 ' 173° EL -1.17' BOTTOM COL EL -267' FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 205 233--.243° 205 ' 216° EL 18.25' EL. 15.18' -----!-+. 221 EL 9. 73' 221 ----1.J. EL 4.28' 221 ' 216° 226° ---. 243 231° 223° 162 + 173° REACTOR SUPPORT TEMPERATURE DISTRIBUTION LONG TERM LOSS OF FANS TIME = 4.5 HOURS AIR TEMPERATURE= 295°F (45 MIN) 132 + 146° Am.3-7/85 Figure 3H-31 * * * * *
- Ii 124 117 110.7 105 99.1 93 87.4 79 70
- 121 120 120 120 120 121 121 1 1 121 :iR' 122 122 1 122 125 23 123 127 1211 124 133 126 124 138 131 127 125 '30 142 141 128 126 '30 144 I .. l;i.::-128 126 120 s: 124 138 1 1 1 28 s: 130 ,. 32 2$ 26 22 1 125 117 123 124 121 116 122 1216F 114 tTYPI 124 121 117 115 108 105 104 98 93 92 88 79 78 77 70 70 70 70 VERTICAL FINITE ELEMENT "10DEL WITH NORMAL TEMPERATURE DISTRIBUTION
- 120 120 120 120 118 116 115 113 112 113 110 108 105 102 110 107 104 102 99 106 104 101 98 96 100 98 96 93 91 92 90 S9 87 86 85 84 82 81 80 76 75 75 74 74 70 70 70 70
- 120 120 111 109 100 97 97 94 93 91 89 81 84 82 79 78 74 73 70 70
- 120 120 70 108 89 70 94 82 70 91 80 70 88 78 70 84 77 70 81 75 70 77 73 70 73 71 70 70 70 Am. 3-7 /85 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 VERTICAL FINITE ELEMENT MODEL FIGURE 3H-32
- 150 145
- u.. w 0 w 140 cc ::::> I-<( a: w c.. ::?! w 135 I-0.:
- c.. ::::> ('J) w z ..J 130 a: w I-z w (.,) 125 <t_ OF CORE BOTTOM OF CORE
- 0 20 40 60 80 100 120 140 160 180 200 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 DISTANCE DOWN VERTICAL LEG (INCHES) VERTICAL RV SUPPORT LEG TEMPERATURES 300 Am. 3-7/85
- Figure 3H-33
- NOTE: NO CRACKS ON THIS SIDE. *
- SECTOR I
- NOTE: CRACKED ELEMENTS ARE INDICATED BY SHADING. Am. 3-7 /85 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 HORIZONTAL MODEL CRACKING PATTERN FOR SECTOR I FIGURE 3H-34 r
- SECTOR II NOTE: SHADING SHOWS CRACKED ELEMENTS. *
- Am. 3-7/85 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 HORIZONTAL MODEL CRACKING PATTERN FOR SECTOR II FIGURE 3H-35 *
- *
- SECTOR Ill NOTE: SHADING SHOWS CRACKED ELEMENTS. /i *
- Am.3-7/85 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 1 HORIZONTAL MODEL CRACKING PATTERN FOR SECTOR Ill FIGURE 3H-36
-25'-6 592K -22*.91 ........ -18'-0 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 TOTAL LOAD= -592K (SECTION 1) (REF. -COMPUTER RUN# VKGAJDS, DT. 3/16/84) 154., 155u 156 71 72 LOAD SUBCASE 1 NOTE: THERE ARE NO CRACKS FOR THIS CASE. Am. 3-7 /85 85 73 Figure 3H-37 * * * * *
- * * * * -25'-6 451K -22' .g,_....,. ____ _ -18'-0 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 TOTAL LOAD= -451K (SECTOR 1-2) (REF. -COMPUTER RUN# VKGAJDS, OT. 3/16/84) 15'\., 155u 156 NOCO<l'Q(.) 71 72 LOAD SUBCASE 4 NOTE: THERE ARE NO CRACKS FOR THIS CASE . Am. 3-7 /85 98 85 73 Figure 3H-38
-25'-6 -22'-9 -18'-0 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 TOTAL LOAD = 390K (SECTOR 2) I) (REF. -COMPUTER RUN# VKGAJDS, OT. 3/16/84) 15ll, 155c., 156 71 72 LOAD SUBCASE 2 NOTE: THERE ARE NO CRACKS FOR THIS CASE. Am..3-7/85 85 73 Figure 3H-39 * * * * *
- * * * * -25'-6 1254K -22'-9 .. -18'-0 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 a: Ill TOTAL LOAD= 1254K (SECTOR 2-3) (REF. -COMPUTER RUN# VKGAJDS, DT. 3/16/84) 154, 155c,, 156 ... .... a: Ill () .-or-.col-U 84 71 72 LOAD SUBCASE 5 0 I-0 en NOTE: SHADING DENOTES CRACKED ELEMENT . I-() () I-C') Oen ... () 98 I-I-() llO o" ... llO!lO 0 ... Am. 3-7 /85 85 73 Figure 3H40
-25'-6 1637K -22'-9 * -18'-0 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 TOTAL LOAD = 1637K (SECTOR 3)) (REF. -COMPUTER RUN# VKGAJDS, OT. 3/16/84) 15.i\_, 155u 156 71 72 LOAD SUBCASE 3 NOTE: SHADING DENOTES CRACKED ELEMENT . Am. 3-7/85 85 73 Figure 3H-41 * * * * *
- * * * * -25'-6 1083K -22'-9 * -18'-0 FLORIDA POWER & LIGHT CO. St. Lucie Plant Unit 1 TOTAL LOAD= 1083K (SECTOR 3-1) (REF. -COMPUTER RUN# VKGAJDS, OT. 3/16/84) 154:,, 155o 156 71 72 LOAD SUBCASE 6 NOTE: SHADING DENOTES CRACKED ELEMENT . 98 85 NOCOIOOU Am. 3-7/85 73 Figure 3H-42
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- LL' 12.. w a: :::::> I-<( a: w Q.. :::!: w I-140 130 128 126 127° @88" FROM CORE ct_ = 4° AT THIS DISTANCE 23.2' 124 EL. 9.73' *-* 1240 122 120 0 TO TOP OF CORE ct_ OF CORE 10 20 30 I 40 50 60 m r .. CIO ;.,, CJ? ct_ CORE @ EL. 18.25' EL. 4.28' BOTTOM OF CORE 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 DISTANCE DOWN THE VERTICAL LEG (INCHES) ct_ SUPP "A"@ EL. 22.76' -AVG TEMP= 135°F GIRD * *
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- 120 121 REACTOR SUPPORT TEMPERATURE DISTRIBUTION STEADY STATE NORMAL OPERATION CONDITIONS AIR TEMPERATURE = 12QOF INTERFACE TEMPERATURE (260°F) 124 124 REACTOR SUPPORT I FOOT I ....,.....___,r"""I I I ...,--""' 125 124 123 122 3H-A12 124 124 121 120 FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 Figure l SCALE: 1/4 IN. = 1 FT * * * *
- *
- CONCRETE SHIELD WALL ELEVATION 22.5 FEET STEADY STATE TEMPERATURE DISTRIBUTION NORMAL OPERATION AIR TEMPERATURE. 120°F 'LORIDA POWER & LIGHT COMPANY .t. Lucie Unit No. 1 'igure 2 3H-Al3 SCALE: 3/16 IN. == 1 FT.
LEG STEADY ST A TE NOZZLE AND SUPPORT FOOT TEMPERATURE DISTRIBUTION FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 ?igure 3 375 350 325 300 275 595 575 ------_L. INTERFACE 2so------230 ------220 3H-Al4 * * *
- 73 *
- REACTOR SUPPORT TEMPERATURE DISTRIBUTION START .UP TIME = 18.0 HOURS INTERFACE TEMPERATURE (260°".) 92 AIR TEMPERATURE == 12QOF INITIAL TEMPERATµRE -7rfF 115 REACTOR SUPPORT I FOOT I n----inl 117 114 110 3H-Al5 115 107 92 73 FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 Figure 4 SCALE: 1/4 IN. = 1 FT
- CONCRETE SHIELD WALL ELEVATION 22.S FEET TEMPERATURE DISTRIBUTION 120
- JRIDA POWER & LIGHT COMPANY St. Lucie Unit No. l SEVERAL DAYS AFTER ST ART -UP 70 Figure 5 SCALE: 3/16 IN. = 1 FT. 3H-A16 * * *
- -u.. 0 -w 0::: :::> I-* < 0::: w c.. :e w I-* -500 400 300 200 100 MEAN AIR TEMPERATURE OF REACTOR CAVIT:Y FOLLOWING LOSS OF FANS ACCIDENT WJTHOUT SHUTDOWN 0 1 2 3 TIME (HOURS) 3H-A17 4 s FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 Figure 6 6 VJ t::C I > ..... (X)
- FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 Figure 7 750 I OPERATION ACTION -u. 600h/ 0 -w 0:: :> t-< 450 0:: w 0.. :e w t-t-300 % j 0 0 u 150 REACTOR COOLANT TEMPERATURE VERSUS TIME FOLLOWING OPERATOR ACTION FOR LOSS OF FANS CASE BORATION CYCLE 0 1 2 3 4 5 6 TIME (HOURS) * *
- 121 *
- REACTOR SUPPORT TEMPERATURE DISTRIBUTION LONG TERM LOSS OF FANS TIME .. 4.5 HOURS AIR TEMPERATURE .. 29QOF INTERFACE* TEMPERATURE (2910F) 131 161 202 228 202 161 217 209 131 121 FLORIDA POWER & LIGHT COMPANY SL. Lucie UniL No. 1 Figure 8 200 SCALE: 1/4 IN. = 1 FT
- 3H-Al9 CONCRETE SHIELD WALL ELEVATION 22.S FEET TEMPERATURE DISTRI BUTIOH 4.5 HOURS AFTER LOSS OF FAHS OCCURRENCE DURING NORMAL OPERATIONS 1LORIDA POWER & LIGHT COMPANY it. Lucie Unit No. 1 9 3H-A20 SCALE: 3/16 IN. = 1 FT. * * *
- *
- CONCRETE SHIELD WALL ELEVATION 22.S FEET TEMPERATURE DISTRIBUTION 4.5 HOURS AFTER LOSS OF FANS OCCURRENCE DURING ST ART -UP FLORIDA POWER & LIGHT COMPANY ST. Lucie Unit No. 1 Figure 10 SCALE: 3/16 IN. = 1 FT. 3H-A21 120 REACTOR SUPPORT TEMPERATURE DISTRIBUTION LOCA TIME
- 1.0 SECOND CE-EBASCO INTERFACE TEMPERATURE (2600F) 121 124 AIR TEMPERATURE
- 12QOf 124 125 124 123 122 3H-AZ2 124 124 121 120 FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 Figure 11 SCALE: 1/4 IN. "' 1 FT * * * *
- CONCRETE SHIELD WALL ELEVATION 22.5 FEET LOCA TEMPERATURE DISTRIBUTION TIME ... 1.0 SECOND
- FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 SCALE: 3/16 IN. = 1 FT. Figure 12 3J:i-A23 400 300 -LI.. 0 -w :::::> 200 I-< D::: w D.. :::E w I-100 0 .1 30.0 25.0 -20.0 I-w w LI.. 15.0 -z 0 i'.'= 19.0 < > w ..J w 5.0 0.0 -2.92 -5.0 .1 POST. LOCA BOUNDARY CONDITIONS FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No._l Figure 13 STEAM-AIR-WATER WATER 1 10 100 1,000 10,000 100,000 TIME (SECONDS) REFLOOD WATER LEVEL VERSUS TIME FLORIDA POWER & LIGHT CO}fPANY Sc. Lucie Unit No. l Figure 14 1 10 100 1,000 10,000 100,000 TIME (SECONDS) 3H-A24 * *
- CONCRETE SHIELD 22.S FEET
- 130
- POWER & LIGHr COMPANY St. Lucie Unit No. 1 Figure 15 POST .. LOCA TEMPERATURE DISTRIBUTION WITH OPERATION AS INITIAL CONDITION TIME "" 1.0 HOUR 120 SCALE: 3/16 IN. = l FT.
121 128 REACTOR SUPPORT TEMPERATURE DISTRIBUTION POST -LOCA TIME -1.0 HOUR INTERFACE TEMPERATURE (2000F) 205 205 200 200 200 3H-A26 205 205 128 121 FLORIDA POWER & LIGHT COMPANY St. Lucie Unit No. 1 Figure 16 SCALE: 1/4 IN. = 1 FT * * *
- APPENDIX 3I PERFORMANCE QUALIFICATION TEST PROGRAM INFORMATION FOR REPRESENTATIVE BALANCE OF PLANT CLASS 1E EQUIPMENT*A.Switchgear - 4.16 KVB.Motor Control Centers - 1A7, 1B7C.Valve Operators (in containment) - Solenoid D.Motors - Component Cooling Water Pump MotorE.Logic Equipment - Engineered Safety Features Actuation SignalsF.Cable - 2/C #14 Shielded Twisted Pair G.Diesel Generator Control Panel - 1A_______________*The information contained in this appendix is an abstract of equipment qualificationdocumentation contained in FPL Quality Assurance files at the plant site. These files areavailable for auditing by NRC.The test report summaries were included in response to NRC letters dated July 23, 1975and October 10, 1975. It is not the intent to maintain the information in this appendixcurrent. Subsequent designs will ensure that equipment utilized is equivalent or betterthan the existing and qualified for its application.Pursuant to the requirements of IE Bulletin 79-01B a reevaluation of the environmentalqualification of electrical equipment installed in the plant was performed. This updatesthe information provided in this appendix. See Section 3.11 for referencing to FPLresponses to the bulletin.In cases where the information in this section overlaps with 10CFR50.49, the EquipmentQualification List (EQ) and Documentation Packages (Doc Pacs) supersede theinformation provided here. Therefore, the EQ List and appropriate Doc Pacs should beconsulted prior to use of the information in this section.Information provided in this appendix is historical and shall not be updated; however, itmay still be similar to the re-evaluation documentation if no changes have taken place.3I-iAmendment No. 18, (04/01)
A. 4.16KV SWITCHGEAREbasco specification FLO-8770-284C, initially issued June 30, 1969 for the subject equipmentrequires the following:1)Equipment to be built in accordance with recognized standards*2)Equipment to be tested in accordance with recognized standards*3)Equipment to satisfy special requirements (i.e. environmental, etc.)*Standards listed in specification are: IEEEANSI NEMA ASTM ASME NBFUIPCEAThese standards may or may not be directly applicable to equipment purchased, ie, ASTM andASME are applicable to oil circuit breakers, but not to oiless circuit breakers actually purchased. IPCEA standards apply only to internal wiring, IEEE standards are duplicates of ANSI and NEMA.Thus, two groups of standards are directly bearing on subject equipment's design and testing,namely ANSI and NEMA. Standards dealing with testing of medium voltage switchgear, are ANSI C37.09-1964 - "American Standard Test Procedure for AC High-Voltage Circuit Breakers" and NEMA Publication No. SG 4-1963 "High-Voltage Power Circuit Breakers". Both of those above two standards separate the tests required into 2 general categories: a) Design tests and b) Production testsCategory a) deals with equipment of a new design to ascertain its compliance with recognizedstandards. A vendors certificate, of compliance is furnished to demonstrate vendor's satisfactionof design testing to include environmental, short circuit, voltage and heat run, etc. The specifiedservice conditions of 104F ambient (max occasional short time temperature of 49C), 35-95% relative humidity, etc, is within the design testing range.Category b) is directly related to the production tests performed on subject equipment. Table Ilists the specific tests and results achieved.Seismic qualification data is available in Chapter 3, Appendix 3B. 3I-1Amendment No. 17, (10/99) Appendix 3I TABLE I PRODUCTION TESTS (APPLICABLE) TESTS PERFORMED BY VENDOR__________________________________________________________________________ __________________________SPECNEMA(SG-4)ANSI (C37.09) 1. Hi potential TestSG4-4.22,23 Various1500V Secondary & Control Wiring (1 min) (1) 19KV Primary (1 min) (2) 2. Operating TestsSG4-4.21 09-5.11Proper operation at normal, max & min control voltage (3) 3. Operating - Interlock 5.9 4. Correctness of WiringSG4-4.1309-5.9 a) Wire check/continuity against wiring diagrams (4) 5. Operation (Devices) 5.9 b) Meters checked against laboratory standards c) Relays operated (4)(Number) Refers to notes to Table I.3I-2
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TEST PROCEDURES AND RESULTS (Cont'd)b)Full level testing Steady state sine dwell testing was performed at frequencies of 7,8,11, 15,23 and 33 Hz. The input during these dwell tests were: Horizontal Input 0.3g peak Vertical Input0.2g peak In addition, at 15 Hz the inputs were raised to .37g horizontal and .25g vertical to insure the integrity of the lockout relay. All monitored circuits performed as required during the sine dwell test. c) Circuit monitoring Following the reinforcement and bolting of the rear door no further relay chatter occurred. d) Insulation resistance The insulation resistance measurements were taken from a terminal point to ground (readings in megohms): Circuit Tested Before Test After Test1002017.510221
17.0CONCLUSION
To insure the reliability of the structure addition formed gussets and/or channels were added toeach compartment with emphasis on directing the stress toward the base of the cabinit. This elevated the natural frequency a greater amount over the floor spectra.3I-33 3J-i Amendment No. 12, (12/93) A.BACKGROUNDThis position on pipe rupture postulation is intended to comply with the requirements of General DesignCriterion 4, of Appendix A to 10 CFR Part 50 for the design of nuclear power plant structures and components. It is recognized that pipe rupture is a rare event which may only occur under unanticipated conditions, such as those which might be caused by possible design, construction, or operation errors; unanticipated loads or unanticipated corrosive environments. Our observation of actual piping failures has indicated that they generally occur at high stress and fatigue locations, such as at the terminal ends of a piping system at its connection to the nozzles of a component. The rules of this position are intended to utilize the available piping design information by postulating pipe ruptures at locations having relatively higher potential for failure, such that an adequate and practical level of protection may be achieved.B.DESIGN CRITERIA FOR THE DETERMINATION OF POSTULATED PIPING BREAKS1.High-Energy Fluid Systems Pipinga.Fluid Systems Separated From Essential Systems and ComponentsFor the purpose of satisfying the separation provisions of plant arrangement asspecified in B.1.a of Branch Technical Position (BTP) ASB 3-1, a review of the pipinglayout and plant arrangement drawings should clearly show the effects of postulated piping breaks at any location are isolated or physically remote from essential systemsand components1. At the designer's option, break locations as determined from B.1.c.of this criteria may be assumed for this purpose.b.Fluid System Piping in Containment Penetration AreasBreaks and cracks need not be postulated in those portions of piping from containment wall to and including the inboard or outboard isolation valves provided they meet the requirement of the ASME Code, Section III, Subarticle NE-1120 and the followingadditional design requirements:_______________1Systems and components required to shut down the reactor and mitigate the consequences of apostulated pipe rupture without offsite power. 3J-1 Amendment No. 16, (1/98) B.1.b.(1)The following design stress and fatigue limits should not be exceeded:For ASME Code, Section III, Class 1 Piping(a), (b), (c) THESE SECTIONS ARE INTENTIONALLY OMITTED For ASME Code, Section III, Class 2 Piping(d)The maximum stress as calculated by the sum of Eqs. (9) and (10) inParagraph NC-3652, ASME Code, Section III, considering those loads and conditions thereof for which level A and level B stress limits have been specified ln the system's Design Specification (i.e., sustained loads, occasional loads, and thermal expansion) including an OBE event should not exceed 0.8(1.8 Sh + SA). The Sh and SA are allowable stresses at maximum(hot) temperature and allowable stress range for thermal expansion, respectively, as defined in Article NC-3600 of the ASME Code, Section III.(e)The maximum stress, as calculated by Eq. (9) in NC-3653 under the loadings resulting from a postulated piping failure of fluid system piping beyond these portions of piping should not exceed the lesser of 2.25 Sh and 1.8 SY.Primary loads include those which are deflection limited by whip restraints. Higher Stresses may be allowed following a failure when the following conditions exist:i)The piping is between the outboard isolation valve and the first restraintii)A plastic hinge is not formediii)operability of the valves with the higher stress is assured inaccordance with the requirements specified in SRP Section 3.9.3iv)piping is constructed in accordance with ANSI B31.1v)The piping shall either be of seamless construction with full radiography of all circumferential welds, or all longitudinal and circumferential welds shall be fully radiographed.(2)Welded attachments, for pipe supports or other purposes, to these portions of piping should be avoided except where detailed stress analyses, or tests, are performed to demonstrate compliance with the limits of B.1.b.(1).(3)The number of circumferential and longitudinal piping welds and branch connections should be minimized. Where guard pipes are used, the enclosed portion of fluid systempiping should be seamless construction and without circumferential welds unless specific access provisions are made to permit inservice volumetric examination of the longitudinal and circumferential welds.(4)The length of these portions of piping should be reduced to the minimum length practical.3J-2 Amendment No. 12, (12/93 B.1.b(5)The design of pipe anchors or restraints (e.g., connections to containment penetrationsand pipe whip restraints) should not require welding directly to the outer surface of the piping (e.g., flued integrally forged pipe fittings may be used) except where such welds are 100 percent volumetrically examinable in service and a detailed stress analysis is performed to demonstrate compliance with the limits of B.1.b.(1).(6)Guard pipes provided for those portions of piping in the containment penetration areas should be constructed in accordance with the rules of Class MC, Subsection NE of the ASME Code, Section III, where the guard pipe is part of the containment boundary. In addition, the entire guard pipe assembly should be designed to meet the following requirements and tests:(a)The design pressure and temperature should not be less than the maximumoperating pressure and temperature of the enclosed pipe under normal plantconditions.(b)The Level C stress limits in NE-3220, ASME Code, Section III, should not beexceeded under the loadings associated with containment design pressure and temperature in combination with the safe shutdown earthquake.(c)Guard pipe assemblies should be subjected to a single pressure test at a pressure not less than its design pressure.(d)Guard pipe assemblies should not prevent the access required to conduct the inservice examination specified in B.1.b.(7). Inspection ports, if used, should not be located in that portion of the guard pipe through the annulus of dual barrier containment structures.(7)A 100% volumetric inservice examination of all pipe welds should be conducted during each inspection interval as defined in IWA-2400, ASME Code, Section XI.B.1.c.Postulation of Pipe Breaks In Areas Other Than Containment Penetration(1)THIS SECTION INTENTIONALLY OMITED(2)With the exceptions of those portions of piping identified in B.1.b., breaks in Class 2 and 3 piping (ASME Code, Section III) should be postulated at the following locations in those portions of each piping and branch run:(a)At terminal ends.(b)At intermediate locations selected by one of the following criteria:(i)At each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandardfitting), welded attachment, and valve. Where the piping contains nofittings, welded attachments, or valves, at one location at each extreme of the piping run adjacent to the protective structure. 3J-3 Amendment No. 17 (10/99 B.1.c(2)(b)(ii)At each location where stresses calculated2 by the sum of Eqs. (9)and (10) in NC/ND-3653, ASME Code, Section III, exceed 0.8 timesthe sum of the stress limits given in NC/ND-3653.As a result of piping reanalysis due to differences between the designconfiguration and the as-built configuration, the highest stress locations may be shifted; however, the initially determined intermediate break locations may be used unless a redesign of the piping resulting in a change in pipe parameters (diameter, wall thickness, routing) is required, or the dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe whip restraints and jet shields.(3)Breaks in seismically analyzed non-ASME Class piping are postulated according to the same requirements for ASME Class 2 and 3 piping above3.(4)Applicable to (1), (2) and (3) above:If a structure separates a high energy line from an essential component, that separatingstructure should be designed to withstand the consequences of the pipe break in the high energy line which produces the greatest effect at the structure irrespective of the fact that the above criteria might not require such a break location to be postulated.(5) Safety-related equipment must be environmentally qualified in accordance with 10 CFR 50.49. Required pipe ruptures and leakage cracks (whichever controls) must be included in the design bases for environmental qualification of electrical and mechanical equipment both inside and outside the containment.B.1.d.The designer should identify each piping run he has considered to postulate the break locations required by B.1.c. above. In complex systems such as those containing arrangements of headers and parallel piping running between headers, the designer should identify and include all such piping within a designated run in order to postulate the number of breaks required by these criteria.B.1.e.With the exception of those portions of piping identified in B.1.b, leakage cracks should bepostulated as follows:(1)THIS SECTION INTENTIONALLY OMITTED(2)For ASME Code, Section III Class 2 and 3 or nonsafety class (not ASME Class 1, 2 or3) piping, at axial locations where the calculated stress2 by the sum of Eqs. (9) and(10) in NC/ND-3653 exceeds 0.4 times the sum of the stress limits given tn NC/ND-3653._______________2For those loads and conditions in which Level A and Level B stress limits have been specified in theDesign Specification (including the operating basis earthquake).3Note that in addition, breaks in non-seismic, that is, non-Category I piping, are to be taken into account as described in Section II.2.k. "Interaction of Other Piping with Category I Piping" of SRP 3.9.2.3J-4Amendment No. 17 (10/99)
B.3.b.Longitudinal Pipe BreaksThe following longitudinal breaks should be postulated in high-energy fluid system piping at thelocations of the circumferential breaks specified in B.3.a:(1) Longitudinal breaks in fluid system piping and branch runs should be postulated innominal pipe sizes 4-inch and larger, except where the maximum stress rangeexceeds the limits specified in B.1.c.(1) and B.1.c.(2) but the axial stress range is at least 1.5 times the circumferential stress range.(2) Longitudinal breaks need not be postulated at terminal ends.(3) Longitudinal breaks should be assumed to result in an axial split without pipe severance. Splits should be oriented (but not concurrently) at two diametrically opposed points on the piping circumference such that the jet reactions cause out-of-plane bending of the piping configuration. Alternatively, a single split may be assumedat the section of highest tensile stress as determined by detailed stress analysis (e.g.,finite element analysis).(4)The dynamic force of the fluid jet discharge should be based on a circular or elliptical (2D x 1/2D) break area equal to the effective cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified byan analytically or experimentally determined thrust coefficient as determined for a circumferential break at the same location. Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account as applicable, in the reduction of jet discharge.(5)Piping movement should be assumed to occur in the direction of the jet reaction unlesslimited by structural members, piping restraints, or piping stiffness as demonstrated by inelastic limit analysis.B.3.c.Leakage CrackLeakage cracks should be postulated at those axial locations specified in B.1.e for high-energy fluid system piping and in those piping systems not exempted in B.2.c.(1) for moderate-energy fluid system piping.(1)Leakage cracks need not be postulated in 1 inch and smaller piping.(2)For high-energy fluid system piping, the leakage cracks should be postulated to be inthose circumferential locations that result in the most severe environmental consequences. For moderate-energy fluid system piping, see B.2.c.(2).(3)Fluid flow from a leakage crack should be based on a circular opening of area equal to that of a rectangle one-half pipe diameter in length and one-half pipe wall thickness in width. 3J-6 Amendment No. 17 (10/99)}}