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{{#Wiki_filter:}} | {{#Wiki_filter:SECTION 6 PLANT ENGINEERED SAFEGUARDS TABLE OF CONTENTS 6.1 Summary Description ................................................................. 1 6.2 Emergency Core Cooling System (ECCS) ................................. 1 | ||
6.3 Main Steam Line Flow Restrictions ........................................... 1 6.4 Control Rod Velocity Limiters .................................................... 1 6.5 Control Rod Drive Housing Supports ........................................ 1 6.6 Standby Liquid Control System ................................................. 1 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability ....................................... 1 6.8 References ................................................................................... 1 6.FIGURES ................................................................................................................. 1 | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.1 Summary Description 6.1.1 Introduction | |||
6.1.2 Containment Systems 6.1.3 Emergency Core Cooling System (ECCS) | |||
6.1.4 Other Systems or Features | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.2 Emergency Core Cooling System (ECCS) 6.2.1 Introduction | |||
6.2.2 Core Spray System | |||
6.2.3 Residual Heat Removal System (RHR) | |||
6.2.4 High Pressure Coolant Injection System (HPCI) | |||
6.2.5 Automatic Depressurization System (ADS) | |||
6.2.6 ECCS Performance Evaluation | |||
6.2.7 Additional Analysis | |||
* Note: See Section 14.7.2 for the flow assumed by the plant safety analysis. | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.3 Main Steam Line Flow Restrictions 6.3.1 Design Basis | |||
6.3.2 Description | |||
6.3.3 Performance Analysis | |||
6.3.4 Inspection and Testing | |||
Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 2SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.4Control Rod Velocity LimitersA topical report describing and analyzing the control rod velocity limiter wassubmitted separately to the Atomic Energy Commission as APED-5446, Control Rod Velocity Limiter, (Reference 32) and is incorporated herein by reference.6.4.1Design BasisThe purpose of the control rod velocity limiter is to reduce the consequences in the event a high-worth control rod became detached from its rod drive and dropped out of the reactor core. To accomplish this purpose the velocity limiterwas designed using the following basis:a.The control rod free fall velocity is less than 5 ft per second.b.A minimum impedance of the control rod scram time or positioning abilityis maintained.c.The velocity limiter is integrally attached to the control rod structure.6.4.2DescriptionThe velocity limiter assembly consists of a single Type 304 stainless steelcasting in the shape of two nearly-mated conical elements. These elements are separated from one another by four radial spacers. The separated surfaces ofthe upper and lower conical elements differ by 15°, with the peripheralseparation less than the central separation.The velocity limiter assembly, shown in Figure 6.4-1 with its associatedcomponents, acts within a cylindrical guide tube. The annulus between the guide and the velocity limiter assembly permits the free passage of water over the smooth surface of the cone when the control rod is scrammed in the upward direction. In the opposite direction, however, water is trapped by the lower coneand discharged through the interface between the two conical sections. Becausethis water is jetted in a partially reversed direction into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod and limiter assembly.The guide tubes are 10-inch, schedule 10, Type 304 stainless steel pipe. Eachguide tube has a back-seat on the lower end which rests on the control rod drive thimble. This seat restricts water flow out of the tube during a velocity limiterfree-fall; the seat also restricts water flow into the interior of the guide tubeduring normal reactor operation to prevent coolant bypass of the fuel elements.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.04Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 2I/mab6.4.3Performance AnalysisDuring the development of the velocity limiter, sensitivity tests were performed toassess the effect of manufacturing tolerances in the following items on thevelocity limiter performance: Limiter and guide tube diametral tolerance; Nozzle (interfacial gap between cones) gap; Top cone thickness; Limiter/guide tube eccentricity; and Surface finish. These tests and the optimization of the velocity limiter design are described in detail in APED- 5446, Control Rod Velocity Limiter(Reference 32). The results of these tests are summarized as follows:Dropout VelocitiesCold reactor - 2.46 ft/secHot reactor - 2.86 ft/secScram Times10% of full insertion- 0.55 sec90% of full insertion - 5.0 sec6.4.4Inspection and TestingTesting and inspecting of the control rod velocity limiter is not required followinginstallation of the control rod assembly. In addition to close surveillance duringthe fabrication of the rod velocity limiter and control rod assembly manufacture,random control rod assemblies were shop tested which included rod drop tests. | |||
Each velocity limiter was visually inspected and gauged prior to assembly. The operation of the individual control rod assemblies for normal operation and scram conditions was confirmed during preoperational testing. | |||
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 3SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.5Control Rod Drive Housing Supports6.5.1Design BasisThe control rod drive housing supports protect against additional damage to thenuclear system process barrier or damage to the fuel barrier by preventing any significant nuclear transient in the event a drive housing breaks or separatesfrom the bottom of the reactor vessel. To accomplish this the control rod drivehousing supports were designed in accordance with the following:Design Basisa.Control rod downward motion shall be limited, following a postulatedcontrol rod drive (CRD) housing failure, so that any resulting nucleartransient could not be sufficient to cause fuel damage.b.Clearance shall be provided between the housings and the supports toprevent vertical contact stresses due to their respective thermal expansion during plant operation.6.5.2DescriptionThe control rod housing supports are illustrated in Figure 6.5-1. Horizontal beams are installed immediately below the bottom head of the reactor vessel, between the rows of control rod housings and are bolted to brackets which are welded to the steel liner of the drive room in the reactor support pedestal.Hanger rods, about 10 feet long by 1-3/4 inches in diameter, are supported fromthe beams on stacks of disc springs which compress about 2 inches under design load.The support bars are bolted between the bottom ends of the hanger rods. Thespring pivots at the top and the beveled loose-fitting ends on the support barsprevent substantial bending moment in the hanger rods if the support bars are overloaded.Individual grids rest on support bars between adjacent beams. Because a singlepiece grid would be difficult to handle in the limited work space and because it isnecessary that control rod drives, position indicators, and incore instrumentation components are accessible for inspection and maintenance, each grid is designed to be assembled or disassembled in place. Each grid assembly is made from two grid plates, a clamp and a bolt. The top part of the clamp acts asa guide to assure that each grid is correctly positioned directly below therespective CRD housing which it would support in the postulated accident.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.05Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 3I/mabWhen the support bars and grids are installed, a gap of less than 1 inch at roomtemperature (approximately 70°F) is provided between the grid and the bottomcontact surface of the control rod drive flange. During system heatup, this gap is reduced by a net downward expansion of the housings with respect to the supports. In the hot operating condition, the gap is approximately 1/4 inch.In the postulated CRD housing failure, the CRD housing supports are loadedwhen the lower contact surface of the CRD flange contacts the grid. The resulting load is then carried by two grid plates, two support bars, four hangerrods, their disk springs, and two adjacent beams.The American Institute of Steel Construction (AISC) Specification for the Design,Fabrication, and Erection of Structural Steel for Building was used in the design of the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses were taken as 90% of yield, and the shear stress as 60% ofyield. These are 1.5 times the corresponding AISC allowable stresses of 60%and 40% of yield. This stress criterion is considered desirable for this application and adequate for the once in a lifetime loading condition.For mechanical design purposes, the postulated failure resulting in the highestforces is an instantaneous circumferential separation of the CRD housing fromthe reactor vessel, with an internal pressure of 1250 psig (reactor vessel designpressure) acting on the area of the separated housing. The weight of the separated housing, control rod drive, and blade, plus the force of 1250 psig pressure acting on the area of the separated housing gives a force of approximately 35,000 lbs. This force is multiplied by a factor of 3 for impact,conservatively assuming the housing travels through a 1-inch gap beforecontacting the supports. The total force (105 lbs) is then treated as a static loadin design formulas.6.5.3Performance AnalysisDownward travel of CRD housing and its control rod following the postulatedhousing failure is the sum of the compression of the disk springs under dynamicloading and the initial gap between the grid and the bottom contact surface of the CRD flange. If the reactor were cold and pressurized, the downward motion of the control rod would be limited to the approximate 2 inch spring compressionplus a gap of less than one inch. If the reactor were hot and pressurized, thegap would be approximately 1/4 inch and the spring compression slightly less than in the cold condition. In either case, the control rod movement following a housing failure is limited substantially below one drive notch movement (6 inches). The nuclear transient from sudden withdrawal of any control rodthrough a distance of one drive notch at any position in the core does not resultin a transient sufficient to cause damage to any radioactive material barrier. This meets the fuel damage prevention criteria of design basis 6.5.1-a. | |||
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 3I/mabThe control rod drive housing supports are in place any time the reactor is to beoperated. The housing supports may be removed when the reactor is in theshutdown condition even when the reactor is pressurized, because all controlrods are then inserted. Even if a control rod is ejected under the shutdown condition, the reactor remains subcritical, because it is designed to remain subcritical with any one control rod fully withdrawn at any time.At plant operating temperature a gap of approximately 1/4 inch is maintainedbetween the CRD housing and the supports, at lower temperatures the gap isgreater. Because the supports do not come in contact with any of the CRD housings, except during the postulated accident condition, vertical contact stresses are prevented as required by safety design basis 6.5.1-b.6.5.4Inspection and TestingWhen the reactor is in the shutdown mode, the control rod drive housing supports may be removed to permit inspection and maintenance of the control rod drives. When the support structure is reinstalled, it is inspected for proper assembly, particular attention being given to assure that the correct gap between the CRD flange lower contact surface and the grid is maintained. Since thestructure is not stressed until an accident occurs, testing is unnecessary. If anaccident should occur any deformed parts would be replaced during repair. | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.6 Standby Liquid Control System 6.6.1 Design Basis | |||
6.6.2 Description | |||
6.6.3 Performance Analysis | |||
6.6.4 Inspection and Testing | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability 6.7.1 Design Basis | |||
6.7.2 Description | |||
6.7.3 Performance Analysis | |||
6.7.4 Inspection and Testing | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.8 References | |||
Revision 29USAR-06.FIGMONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 10I/arb Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 10USAR-06.FIGI/arbFigure 6.2-2 Typical Core Spray Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 10USAR-06.FIGI/arbFigure 6.2-4 Typical RHR Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 4 of 10USAR-06.FIGI/arbFigure 6.2-5 RHR - Simplified P&ID - LPCI ModeSelecting Specified LoopTORUS RING HEADERPUMPSJETHxRHRPUMPSRHRCONDENSATESTORAGETANK Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 5 of 10USAR-06.FIGI/arbFigure 6.2-6 RHR - Simplified P&ID - Containment Spray/Cooling ModeRHRHxPUMPSRHRTORUS RING HEADERRECIRC PUMPSTORUSPUMPSJETHxRHRPUMPSRHR Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 6 of 10USAR-06.FIGI/arbFigure 6.2-7a LPCI System - Logic Control System ArrangementREACTORVESSELLPCISPUMPPUMPP1SBPWWPPPPPPPPPPPAAAAAAAAABBBBBBBBIISD42DPPBPAPAB1, 2, 3, 4= P P(Riser Differential Pressure)WHERE:A/B indicates recirc loop D = Ricirc Pump Discharge ValveI = RHR/LPCI Injection ValvesP = Pressure S = Recirc Pump Suction Valve W = LPCI Injection Water3 Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 7 of 10USAR-06.FIGI/arbFigure 6.2-7b LPCI System - Loop Selection/Break DetectionFunctional Block DiagramYESReactor Pressure > 420 psigpermissiveReactor pressure< 900 psigPermissiveInject LPCI in Loop AClose A Recirc Pump Discharge ValveOpen A LPCI Injection ValveBlock Closed B LPCI Injection ValvesClose Containment Spray/Cooling ValvesInject LPCI in Loop BClose B Recirc Pump Discharge ValveOpen B LPCI Injection Valve Block Closed A LPCI Injection ValvesClose Containment Spray/Cooling ValvesReactor Pressure > 420 psigpermissiveLoop A SelectedRecirc Riser PPA> PBPA= PBPB> PALoop B Selected2 Sec Time DelayRecircPump A runningNTSP of > 3.4 psid( >87.2 inWC)High DrywellPressureLow-Low Reactor Water LevelTrip Recirc PumpDrive Motor BreakerYESNONORecircPump B runningNTSP of > 3.4 psid( > 87.2 inWC) | |||
Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 8 of 10USAR-06.FIGI/arbFigure 6.3-1 Main Steamline Flow Restrictor Nozzle Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 9 of 10USAR-06.FIGI/arbFigure 6.4-1 Control Rod Velocity Limiter IsometricHANDLE MODELD120 EXTENDEDHANDLEBLADESHEATHRODSNEUTRON ABSORBERVELOCITY LIMITERHANDLECOUPLING RELEASECOUPLING SOCKET143"6.5" Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 10 of 10USAR-06.FIGI/arbFigure 6.5-1 Control Rod Drive Housing Support Isometric01294216 SECTION 6 PLANT ENGINEERED SAFEGUARDS TABLE OF CONTENTS 6.1 Summary Description ................................................................. 1 6.2 Emergency Core Cooling System (ECCS) ................................. 1 | |||
6.3 Main Steam Line Flow Restrictions ........................................... 1 6.4 Control Rod Velocity Limiters .................................................... 1 6.5 Control Rod Drive Housing Supports ........................................ 1 6.6 Standby Liquid Control System ................................................. 1 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability ....................................... 1 6.8 References ................................................................................... 1 6.FIGURES ................................................................................................................. 1 | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.1 Summary Description 6.1.1 Introduction | |||
6.1.2 Containment Systems 6.1.3 Emergency Core Cooling System (ECCS) | |||
6.1.4 Other Systems or Features | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.2 Emergency Core Cooling System (ECCS) 6.2.1 Introduction | |||
6.2.2 Core Spray System | |||
6.2.3 Residual Heat Removal System (RHR) | |||
6.2.4 High Pressure Coolant Injection System (HPCI) | |||
6.2.5 Automatic Depressurization System (ADS) | |||
6.2.6 ECCS Performance Evaluation | |||
6.2.7 Additional Analysis | |||
* Note: See Section 14.7.2 for the flow assumed by the plant safety analysis. | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.3 Main Steam Line Flow Restrictions 6.3.1 Design Basis | |||
6.3.2 Description | |||
6.3.3 Performance Analysis | |||
6.3.4 Inspection and Testing | |||
Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 2SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.4Control Rod Velocity LimitersA topical report describing and analyzing the control rod velocity limiter wassubmitted separately to the Atomic Energy Commission as APED-5446, Control Rod Velocity Limiter, (Reference 32) and is incorporated herein by reference.6.4.1Design BasisThe purpose of the control rod velocity limiter is to reduce the consequences in the event a high-worth control rod became detached from its rod drive and dropped out of the reactor core. To accomplish this purpose the velocity limiterwas designed using the following basis:a.The control rod free fall velocity is less than 5 ft per second.b.A minimum impedance of the control rod scram time or positioning abilityis maintained.c.The velocity limiter is integrally attached to the control rod structure.6.4.2DescriptionThe velocity limiter assembly consists of a single Type 304 stainless steelcasting in the shape of two nearly-mated conical elements. These elements are separated from one another by four radial spacers. The separated surfaces ofthe upper and lower conical elements differ by 15°, with the peripheralseparation less than the central separation.The velocity limiter assembly, shown in Figure 6.4-1 with its associatedcomponents, acts within a cylindrical guide tube. The annulus between the guide and the velocity limiter assembly permits the free passage of water over the smooth surface of the cone when the control rod is scrammed in the upward direction. In the opposite direction, however, water is trapped by the lower coneand discharged through the interface between the two conical sections. Becausethis water is jetted in a partially reversed direction into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod and limiter assembly.The guide tubes are 10-inch, schedule 10, Type 304 stainless steel pipe. Eachguide tube has a back-seat on the lower end which rests on the control rod drive thimble. This seat restricts water flow out of the tube during a velocity limiterfree-fall; the seat also restricts water flow into the interior of the guide tubeduring normal reactor operation to prevent coolant bypass of the fuel elements.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.04Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 2I/mab6.4.3Performance AnalysisDuring the development of the velocity limiter, sensitivity tests were performed toassess the effect of manufacturing tolerances in the following items on thevelocity limiter performance: Limiter and guide tube diametral tolerance; Nozzle (interfacial gap between cones) gap; Top cone thickness; Limiter/guide tube eccentricity; and Surface finish. These tests and the optimization of the velocity limiter design are described in detail in APED- 5446, Control Rod Velocity Limiter(Reference 32). The results of these tests are summarized as follows:Dropout VelocitiesCold reactor - 2.46 ft/secHot reactor - 2.86 ft/secScram Times10% of full insertion- 0.55 sec90% of full insertion - 5.0 sec6.4.4Inspection and TestingTesting and inspecting of the control rod velocity limiter is not required followinginstallation of the control rod assembly. In addition to close surveillance duringthe fabrication of the rod velocity limiter and control rod assembly manufacture,random control rod assemblies were shop tested which included rod drop tests. | |||
Each velocity limiter was visually inspected and gauged prior to assembly. The operation of the individual control rod assemblies for normal operation and scram conditions was confirmed during preoperational testing. | |||
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 3SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.5Control Rod Drive Housing Supports6.5.1Design BasisThe control rod drive housing supports protect against additional damage to thenuclear system process barrier or damage to the fuel barrier by preventing any significant nuclear transient in the event a drive housing breaks or separatesfrom the bottom of the reactor vessel. To accomplish this the control rod drivehousing supports were designed in accordance with the following:Design Basisa.Control rod downward motion shall be limited, following a postulatedcontrol rod drive (CRD) housing failure, so that any resulting nucleartransient could not be sufficient to cause fuel damage.b.Clearance shall be provided between the housings and the supports toprevent vertical contact stresses due to their respective thermal expansion during plant operation.6.5.2DescriptionThe control rod housing supports are illustrated in Figure 6.5-1. Horizontal beams are installed immediately below the bottom head of the reactor vessel, between the rows of control rod housings and are bolted to brackets which are welded to the steel liner of the drive room in the reactor support pedestal.Hanger rods, about 10 feet long by 1-3/4 inches in diameter, are supported fromthe beams on stacks of disc springs which compress about 2 inches under design load.The support bars are bolted between the bottom ends of the hanger rods. Thespring pivots at the top and the beveled loose-fitting ends on the support barsprevent substantial bending moment in the hanger rods if the support bars are overloaded.Individual grids rest on support bars between adjacent beams. Because a singlepiece grid would be difficult to handle in the limited work space and because it isnecessary that control rod drives, position indicators, and incore instrumentation components are accessible for inspection and maintenance, each grid is designed to be assembled or disassembled in place. Each grid assembly is made from two grid plates, a clamp and a bolt. The top part of the clamp acts asa guide to assure that each grid is correctly positioned directly below therespective CRD housing which it would support in the postulated accident.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.05Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 3I/mabWhen the support bars and grids are installed, a gap of less than 1 inch at roomtemperature (approximately 70°F) is provided between the grid and the bottomcontact surface of the control rod drive flange. During system heatup, this gap is reduced by a net downward expansion of the housings with respect to the supports. In the hot operating condition, the gap is approximately 1/4 inch.In the postulated CRD housing failure, the CRD housing supports are loadedwhen the lower contact surface of the CRD flange contacts the grid. The resulting load is then carried by two grid plates, two support bars, four hangerrods, their disk springs, and two adjacent beams.The American Institute of Steel Construction (AISC) Specification for the Design,Fabrication, and Erection of Structural Steel for Building was used in the design of the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses were taken as 90% of yield, and the shear stress as 60% ofyield. These are 1.5 times the corresponding AISC allowable stresses of 60%and 40% of yield. This stress criterion is considered desirable for this application and adequate for the once in a lifetime loading condition.For mechanical design purposes, the postulated failure resulting in the highestforces is an instantaneous circumferential separation of the CRD housing fromthe reactor vessel, with an internal pressure of 1250 psig (reactor vessel designpressure) acting on the area of the separated housing. The weight of the separated housing, control rod drive, and blade, plus the force of 1250 psig pressure acting on the area of the separated housing gives a force of approximately 35,000 lbs. This force is multiplied by a factor of 3 for impact,conservatively assuming the housing travels through a 1-inch gap beforecontacting the supports. The total force (105 lbs) is then treated as a static loadin design formulas.6.5.3Performance AnalysisDownward travel of CRD housing and its control rod following the postulatedhousing failure is the sum of the compression of the disk springs under dynamicloading and the initial gap between the grid and the bottom contact surface of the CRD flange. If the reactor were cold and pressurized, the downward motion of the control rod would be limited to the approximate 2 inch spring compressionplus a gap of less than one inch. If the reactor were hot and pressurized, thegap would be approximately 1/4 inch and the spring compression slightly less than in the cold condition. In either case, the control rod movement following a housing failure is limited substantially below one drive notch movement (6 inches). The nuclear transient from sudden withdrawal of any control rodthrough a distance of one drive notch at any position in the core does not resultin a transient sufficient to cause damage to any radioactive material barrier. This meets the fuel damage prevention criteria of design basis 6.5.1-a. | |||
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 3I/mabThe control rod drive housing supports are in place any time the reactor is to beoperated. The housing supports may be removed when the reactor is in theshutdown condition even when the reactor is pressurized, because all controlrods are then inserted. Even if a control rod is ejected under the shutdown condition, the reactor remains subcritical, because it is designed to remain subcritical with any one control rod fully withdrawn at any time.At plant operating temperature a gap of approximately 1/4 inch is maintainedbetween the CRD housing and the supports, at lower temperatures the gap isgreater. Because the supports do not come in contact with any of the CRD housings, except during the postulated accident condition, vertical contact stresses are prevented as required by safety design basis 6.5.1-b.6.5.4Inspection and TestingWhen the reactor is in the shutdown mode, the control rod drive housing supports may be removed to permit inspection and maintenance of the control rod drives. When the support structure is reinstalled, it is inspected for proper assembly, particular attention being given to assure that the correct gap between the CRD flange lower contact surface and the grid is maintained. Since thestructure is not stressed until an accident occurs, testing is unnecessary. If anaccident should occur any deformed parts would be replaced during repair. | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.6 Standby Liquid Control System 6.6.1 Design Basis | |||
6.6.2 Description | |||
6.6.3 Performance Analysis | |||
6.6.4 Inspection and Testing | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability 6.7.1 Design Basis | |||
6.7.2 Description | |||
6.7.3 Performance Analysis | |||
6.7.4 Inspection and Testing | |||
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.8 References | |||
Revision 29USAR-06.FIGMONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 10I/arb Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 10USAR-06.FIGI/arbFigure 6.2-2 Typical Core Spray Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 10USAR-06.FIGI/arbFigure 6.2-4 Typical RHR Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 4 of 10USAR-06.FIGI/arbFigure 6.2-5 RHR - Simplified P&ID - LPCI ModeSelecting Specified LoopTORUS RING HEADERPUMPSJETHxRHRPUMPSRHRCONDENSATESTORAGETANK Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 5 of 10USAR-06.FIGI/arbFigure 6.2-6 RHR - Simplified P&ID - Containment Spray/Cooling ModeRHRHxPUMPSRHRTORUS RING HEADERRECIRC PUMPSTORUSPUMPSJETHxRHRPUMPSRHR Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 6 of 10USAR-06.FIGI/arbFigure 6.2-7a LPCI System - Logic Control System ArrangementREACTORVESSELLPCISPUMPPUMPP1SBPWWPPPPPPPPPPPAAAAAAAAABBBBBBBBIISD42DPPBPAPAB1, 2, 3, 4= P P(Riser Differential Pressure)WHERE:A/B indicates recirc loop D = Ricirc Pump Discharge ValveI = RHR/LPCI Injection ValvesP = Pressure S = Recirc Pump Suction Valve W = LPCI Injection Water3 Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 7 of 10USAR-06.FIGI/arbFigure 6.2-7b LPCI System - Loop Selection/Break DetectionFunctional Block DiagramYESReactor Pressure > 420 psigpermissiveReactor pressure< 900 psigPermissiveInject LPCI in Loop AClose A Recirc Pump Discharge ValveOpen A LPCI Injection ValveBlock Closed B LPCI Injection ValvesClose Containment Spray/Cooling ValvesInject LPCI in Loop BClose B Recirc Pump Discharge ValveOpen B LPCI Injection Valve Block Closed A LPCI Injection ValvesClose Containment Spray/Cooling ValvesReactor Pressure > 420 psigpermissiveLoop A SelectedRecirc Riser PPA> PBPA= PBPB> PALoop B Selected2 Sec Time DelayRecircPump A runningNTSP of > 3.4 psid( >87.2 inWC)High DrywellPressureLow-Low Reactor Water LevelTrip Recirc PumpDrive Motor BreakerYESNONORecircPump B runningNTSP of > 3.4 psid( > 87.2 inWC) | |||
Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 8 of 10USAR-06.FIGI/arbFigure 6.3-1 Main Steamline Flow Restrictor Nozzle Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 9 of 10USAR-06.FIGI/arbFigure 6.4-1 Control Rod Velocity Limiter IsometricHANDLE MODELD120 EXTENDEDHANDLEBLADESHEATHRODSNEUTRON ABSORBERVELOCITY LIMITERHANDLECOUPLING RELEASECOUPLING SOCKET143"6.5" Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 10 of 10USAR-06.FIGI/arbFigure 6.5-1 Control Rod Drive Housing Support Isometric01294216}} | |||
Revision as of 21:59, 28 May 2018
| ML16054A420 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/26/2016 |
| From: | Northern States Power Co, Xcel Energy |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16054A376 | List:
|
| References | |
| L-MT-16-004 | |
| Download: ML16054A420 (83) | |
Text
SECTION 6 PLANT ENGINEERED SAFEGUARDS TABLE OF CONTENTS 6.1 Summary Description ................................................................. 1 6.2 Emergency Core Cooling System (ECCS) ................................. 1
6.3 Main Steam Line Flow Restrictions ........................................... 1 6.4 Control Rod Velocity Limiters .................................................... 1 6.5 Control Rod Drive Housing Supports ........................................ 1 6.6 Standby Liquid Control System ................................................. 1 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability ....................................... 1 6.8 References ................................................................................... 1 6.FIGURES ................................................................................................................. 1
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.1 Summary Description 6.1.1 Introduction
6.1.2 Containment Systems 6.1.3 Emergency Core Cooling System (ECCS)
6.1.4 Other Systems or Features
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.2 Emergency Core Cooling System (ECCS) 6.2.1 Introduction
6.2.2 Core Spray System
6.2.3 Residual Heat Removal System (RHR)
6.2.4 High Pressure Coolant Injection System (HPCI)
6.2.5 Automatic Depressurization System (ADS)
6.2.6 ECCS Performance Evaluation
6.2.7 Additional Analysis
- Note: See Section 14.7.2 for the flow assumed by the plant safety analysis.
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.3 Main Steam Line Flow Restrictions 6.3.1 Design Basis
6.3.2 Description
6.3.3 Performance Analysis
6.3.4 Inspection and Testing
Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 2SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.4Control Rod Velocity LimitersA topical report describing and analyzing the control rod velocity limiter wassubmitted separately to the Atomic Energy Commission as APED-5446, Control Rod Velocity Limiter, (Reference 32) and is incorporated herein by reference.6.4.1Design BasisThe purpose of the control rod velocity limiter is to reduce the consequences in the event a high-worth control rod became detached from its rod drive and dropped out of the reactor core. To accomplish this purpose the velocity limiterwas designed using the following basis:a.The control rod free fall velocity is less than 5 ft per second.b.A minimum impedance of the control rod scram time or positioning abilityis maintained.c.The velocity limiter is integrally attached to the control rod structure.6.4.2DescriptionThe velocity limiter assembly consists of a single Type 304 stainless steelcasting in the shape of two nearly-mated conical elements. These elements are separated from one another by four radial spacers. The separated surfaces ofthe upper and lower conical elements differ by 15°, with the peripheralseparation less than the central separation.The velocity limiter assembly, shown in Figure 6.4-1 with its associatedcomponents, acts within a cylindrical guide tube. The annulus between the guide and the velocity limiter assembly permits the free passage of water over the smooth surface of the cone when the control rod is scrammed in the upward direction. In the opposite direction, however, water is trapped by the lower coneand discharged through the interface between the two conical sections. Becausethis water is jetted in a partially reversed direction into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod and limiter assembly.The guide tubes are 10-inch, schedule 10, Type 304 stainless steel pipe. Eachguide tube has a back-seat on the lower end which rests on the control rod drive thimble. This seat restricts water flow out of the tube during a velocity limiterfree-fall; the seat also restricts water flow into the interior of the guide tubeduring normal reactor operation to prevent coolant bypass of the fuel elements.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.04Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 2I/mab6.4.3Performance AnalysisDuring the development of the velocity limiter, sensitivity tests were performed toassess the effect of manufacturing tolerances in the following items on thevelocity limiter performance: Limiter and guide tube diametral tolerance; Nozzle (interfacial gap between cones) gap; Top cone thickness; Limiter/guide tube eccentricity; and Surface finish. These tests and the optimization of the velocity limiter design are described in detail in APED- 5446, Control Rod Velocity Limiter(Reference 32). The results of these tests are summarized as follows:Dropout VelocitiesCold reactor - 2.46 ft/secHot reactor - 2.86 ft/secScram Times10% of full insertion- 0.55 sec90% of full insertion - 5.0 sec6.4.4Inspection and TestingTesting and inspecting of the control rod velocity limiter is not required followinginstallation of the control rod assembly. In addition to close surveillance duringthe fabrication of the rod velocity limiter and control rod assembly manufacture,random control rod assemblies were shop tested which included rod drop tests.
Each velocity limiter was visually inspected and gauged prior to assembly. The operation of the individual control rod assemblies for normal operation and scram conditions was confirmed during preoperational testing.
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 3SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.5Control Rod Drive Housing Supports6.5.1Design BasisThe control rod drive housing supports protect against additional damage to thenuclear system process barrier or damage to the fuel barrier by preventing any significant nuclear transient in the event a drive housing breaks or separatesfrom the bottom of the reactor vessel. To accomplish this the control rod drivehousing supports were designed in accordance with the following:Design Basisa.Control rod downward motion shall be limited, following a postulatedcontrol rod drive (CRD) housing failure, so that any resulting nucleartransient could not be sufficient to cause fuel damage.b.Clearance shall be provided between the housings and the supports toprevent vertical contact stresses due to their respective thermal expansion during plant operation.6.5.2DescriptionThe control rod housing supports are illustrated in Figure 6.5-1. Horizontal beams are installed immediately below the bottom head of the reactor vessel, between the rows of control rod housings and are bolted to brackets which are welded to the steel liner of the drive room in the reactor support pedestal.Hanger rods, about 10 feet long by 1-3/4 inches in diameter, are supported fromthe beams on stacks of disc springs which compress about 2 inches under design load.The support bars are bolted between the bottom ends of the hanger rods. Thespring pivots at the top and the beveled loose-fitting ends on the support barsprevent substantial bending moment in the hanger rods if the support bars are overloaded.Individual grids rest on support bars between adjacent beams. Because a singlepiece grid would be difficult to handle in the limited work space and because it isnecessary that control rod drives, position indicators, and incore instrumentation components are accessible for inspection and maintenance, each grid is designed to be assembled or disassembled in place. Each grid assembly is made from two grid plates, a clamp and a bolt. The top part of the clamp acts asa guide to assure that each grid is correctly positioned directly below therespective CRD housing which it would support in the postulated accident.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.05Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 3I/mabWhen the support bars and grids are installed, a gap of less than 1 inch at roomtemperature (approximately 70°F) is provided between the grid and the bottomcontact surface of the control rod drive flange. During system heatup, this gap is reduced by a net downward expansion of the housings with respect to the supports. In the hot operating condition, the gap is approximately 1/4 inch.In the postulated CRD housing failure, the CRD housing supports are loadedwhen the lower contact surface of the CRD flange contacts the grid. The resulting load is then carried by two grid plates, two support bars, four hangerrods, their disk springs, and two adjacent beams.The American Institute of Steel Construction (AISC) Specification for the Design,Fabrication, and Erection of Structural Steel for Building was used in the design of the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses were taken as 90% of yield, and the shear stress as 60% ofyield. These are 1.5 times the corresponding AISC allowable stresses of 60%and 40% of yield. This stress criterion is considered desirable for this application and adequate for the once in a lifetime loading condition.For mechanical design purposes, the postulated failure resulting in the highestforces is an instantaneous circumferential separation of the CRD housing fromthe reactor vessel, with an internal pressure of 1250 psig (reactor vessel designpressure) acting on the area of the separated housing. The weight of the separated housing, control rod drive, and blade, plus the force of 1250 psig pressure acting on the area of the separated housing gives a force of approximately 35,000 lbs. This force is multiplied by a factor of 3 for impact,conservatively assuming the housing travels through a 1-inch gap beforecontacting the supports. The total force (105 lbs) is then treated as a static loadin design formulas.6.5.3Performance AnalysisDownward travel of CRD housing and its control rod following the postulatedhousing failure is the sum of the compression of the disk springs under dynamicloading and the initial gap between the grid and the bottom contact surface of the CRD flange. If the reactor were cold and pressurized, the downward motion of the control rod would be limited to the approximate 2 inch spring compressionplus a gap of less than one inch. If the reactor were hot and pressurized, thegap would be approximately 1/4 inch and the spring compression slightly less than in the cold condition. In either case, the control rod movement following a housing failure is limited substantially below one drive notch movement (6 inches). The nuclear transient from sudden withdrawal of any control rodthrough a distance of one drive notch at any position in the core does not resultin a transient sufficient to cause damage to any radioactive material barrier. This meets the fuel damage prevention criteria of design basis 6.5.1-a.
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 3I/mabThe control rod drive housing supports are in place any time the reactor is to beoperated. The housing supports may be removed when the reactor is in theshutdown condition even when the reactor is pressurized, because all controlrods are then inserted. Even if a control rod is ejected under the shutdown condition, the reactor remains subcritical, because it is designed to remain subcritical with any one control rod fully withdrawn at any time.At plant operating temperature a gap of approximately 1/4 inch is maintainedbetween the CRD housing and the supports, at lower temperatures the gap isgreater. Because the supports do not come in contact with any of the CRD housings, except during the postulated accident condition, vertical contact stresses are prevented as required by safety design basis 6.5.1-b.6.5.4Inspection and TestingWhen the reactor is in the shutdown mode, the control rod drive housing supports may be removed to permit inspection and maintenance of the control rod drives. When the support structure is reinstalled, it is inspected for proper assembly, particular attention being given to assure that the correct gap between the CRD flange lower contact surface and the grid is maintained. Since thestructure is not stressed until an accident occurs, testing is unnecessary. If anaccident should occur any deformed parts would be replaced during repair.
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.6 Standby Liquid Control System 6.6.1 Design Basis
6.6.2 Description
6.6.3 Performance Analysis
6.6.4 Inspection and Testing
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability 6.7.1 Design Basis
6.7.2 Description
6.7.3 Performance Analysis
6.7.4 Inspection and Testing
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.8 References
Revision 29USAR-06.FIGMONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 10I/arb Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 10USAR-06.FIGI/arbFigure 6.2-2 Typical Core Spray Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 10USAR-06.FIGI/arbFigure 6.2-4 Typical RHR Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 4 of 10USAR-06.FIGI/arbFigure 6.2-5 RHR - Simplified P&ID - LPCI ModeSelecting Specified LoopTORUS RING HEADERPUMPSJETHxRHRPUMPSRHRCONDENSATESTORAGETANK Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 5 of 10USAR-06.FIGI/arbFigure 6.2-6 RHR - Simplified P&ID - Containment Spray/Cooling ModeRHRHxPUMPSRHRTORUS RING HEADERRECIRC PUMPSTORUSPUMPSJETHxRHRPUMPSRHR Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 6 of 10USAR-06.FIGI/arbFigure 6.2-7a LPCI System - Logic Control System ArrangementREACTORVESSELLPCISPUMPPUMPP1SBPWWPPPPPPPPPPPAAAAAAAAABBBBBBBBIISD42DPPBPAPAB1, 2, 3, 4= P P(Riser Differential Pressure)WHERE:A/B indicates recirc loop D = Ricirc Pump Discharge ValveI = RHR/LPCI Injection ValvesP = Pressure S = Recirc Pump Suction Valve W = LPCI Injection Water3 Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 7 of 10USAR-06.FIGI/arbFigure 6.2-7b LPCI System - Loop Selection/Break DetectionFunctional Block DiagramYESReactor Pressure > 420 psigpermissiveReactor pressure< 900 psigPermissiveInject LPCI in Loop AClose A Recirc Pump Discharge ValveOpen A LPCI Injection ValveBlock Closed B LPCI Injection ValvesClose Containment Spray/Cooling ValvesInject LPCI in Loop BClose B Recirc Pump Discharge ValveOpen B LPCI Injection Valve Block Closed A LPCI Injection ValvesClose Containment Spray/Cooling ValvesReactor Pressure > 420 psigpermissiveLoop A SelectedRecirc Riser PPA> PBPA= PBPB> PALoop B Selected2 Sec Time DelayRecircPump A runningNTSP of > 3.4 psid( >87.2 inWC)High DrywellPressureLow-Low Reactor Water LevelTrip Recirc PumpDrive Motor BreakerYESNONORecircPump B runningNTSP of > 3.4 psid( > 87.2 inWC)
Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 8 of 10USAR-06.FIGI/arbFigure 6.3-1 Main Steamline Flow Restrictor Nozzle Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 9 of 10USAR-06.FIGI/arbFigure 6.4-1 Control Rod Velocity Limiter IsometricHANDLE MODELD120 EXTENDEDHANDLEBLADESHEATHRODSNEUTRON ABSORBERVELOCITY LIMITERHANDLECOUPLING RELEASECOUPLING SOCKET143"6.5" Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 10 of 10USAR-06.FIGI/arbFigure 6.5-1 Control Rod Drive Housing Support Isometric01294216 SECTION 6 PLANT ENGINEERED SAFEGUARDS TABLE OF CONTENTS 6.1 Summary Description ................................................................. 1 6.2 Emergency Core Cooling System (ECCS) ................................. 1
6.3 Main Steam Line Flow Restrictions ........................................... 1 6.4 Control Rod Velocity Limiters .................................................... 1 6.5 Control Rod Drive Housing Supports ........................................ 1 6.6 Standby Liquid Control System ................................................. 1 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability ....................................... 1 6.8 References ................................................................................... 1 6.FIGURES ................................................................................................................. 1
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.1 Summary Description 6.1.1 Introduction
6.1.2 Containment Systems 6.1.3 Emergency Core Cooling System (ECCS)
6.1.4 Other Systems or Features
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.2 Emergency Core Cooling System (ECCS) 6.2.1 Introduction
6.2.2 Core Spray System
6.2.3 Residual Heat Removal System (RHR)
6.2.4 High Pressure Coolant Injection System (HPCI)
6.2.5 Automatic Depressurization System (ADS)
6.2.6 ECCS Performance Evaluation
6.2.7 Additional Analysis
- Note: See Section 14.7.2 for the flow assumed by the plant safety analysis.
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.3 Main Steam Line Flow Restrictions 6.3.1 Design Basis
6.3.2 Description
6.3.3 Performance Analysis
6.3.4 Inspection and Testing
Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 2SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.4Control Rod Velocity LimitersA topical report describing and analyzing the control rod velocity limiter wassubmitted separately to the Atomic Energy Commission as APED-5446, Control Rod Velocity Limiter, (Reference 32) and is incorporated herein by reference.6.4.1Design BasisThe purpose of the control rod velocity limiter is to reduce the consequences in the event a high-worth control rod became detached from its rod drive and dropped out of the reactor core. To accomplish this purpose the velocity limiterwas designed using the following basis:a.The control rod free fall velocity is less than 5 ft per second.b.A minimum impedance of the control rod scram time or positioning abilityis maintained.c.The velocity limiter is integrally attached to the control rod structure.6.4.2DescriptionThe velocity limiter assembly consists of a single Type 304 stainless steelcasting in the shape of two nearly-mated conical elements. These elements are separated from one another by four radial spacers. The separated surfaces ofthe upper and lower conical elements differ by 15°, with the peripheralseparation less than the central separation.The velocity limiter assembly, shown in Figure 6.4-1 with its associatedcomponents, acts within a cylindrical guide tube. The annulus between the guide and the velocity limiter assembly permits the free passage of water over the smooth surface of the cone when the control rod is scrammed in the upward direction. In the opposite direction, however, water is trapped by the lower coneand discharged through the interface between the two conical sections. Becausethis water is jetted in a partially reversed direction into water flowing upward in the annulus, a severe turbulence is created, thereby slowing the descent of the control rod and limiter assembly.The guide tubes are 10-inch, schedule 10, Type 304 stainless steel pipe. Eachguide tube has a back-seat on the lower end which rests on the control rod drive thimble. This seat restricts water flow out of the tube during a velocity limiterfree-fall; the seat also restricts water flow into the interior of the guide tubeduring normal reactor operation to prevent coolant bypass of the fuel elements.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.04Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 2I/mab6.4.3Performance AnalysisDuring the development of the velocity limiter, sensitivity tests were performed toassess the effect of manufacturing tolerances in the following items on thevelocity limiter performance: Limiter and guide tube diametral tolerance; Nozzle (interfacial gap between cones) gap; Top cone thickness; Limiter/guide tube eccentricity; and Surface finish. These tests and the optimization of the velocity limiter design are described in detail in APED- 5446, Control Rod Velocity Limiter(Reference 32). The results of these tests are summarized as follows:Dropout VelocitiesCold reactor - 2.46 ft/secHot reactor - 2.86 ft/secScram Times10% of full insertion- 0.55 sec90% of full insertion - 5.0 sec6.4.4Inspection and TestingTesting and inspecting of the control rod velocity limiter is not required followinginstallation of the control rod assembly. In addition to close surveillance duringthe fabrication of the rod velocity limiter and control rod assembly manufacture,random control rod assemblies were shop tested which included rod drop tests.
Each velocity limiter was visually inspected and gauged prior to assembly. The operation of the individual control rod assemblies for normal operation and scram conditions was confirmed during preoperational testing.
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 3SECTION 6PLANT ENGINEERED SAFEGUARDSI/mab6.5Control Rod Drive Housing Supports6.5.1Design BasisThe control rod drive housing supports protect against additional damage to thenuclear system process barrier or damage to the fuel barrier by preventing any significant nuclear transient in the event a drive housing breaks or separatesfrom the bottom of the reactor vessel. To accomplish this the control rod drivehousing supports were designed in accordance with the following:Design Basisa.Control rod downward motion shall be limited, following a postulatedcontrol rod drive (CRD) housing failure, so that any resulting nucleartransient could not be sufficient to cause fuel damage.b.Clearance shall be provided between the housings and the supports toprevent vertical contact stresses due to their respective thermal expansion during plant operation.6.5.2DescriptionThe control rod housing supports are illustrated in Figure 6.5-1. Horizontal beams are installed immediately below the bottom head of the reactor vessel, between the rows of control rod housings and are bolted to brackets which are welded to the steel liner of the drive room in the reactor support pedestal.Hanger rods, about 10 feet long by 1-3/4 inches in diameter, are supported fromthe beams on stacks of disc springs which compress about 2 inches under design load.The support bars are bolted between the bottom ends of the hanger rods. Thespring pivots at the top and the beveled loose-fitting ends on the support barsprevent substantial bending moment in the hanger rods if the support bars are overloaded.Individual grids rest on support bars between adjacent beams. Because a singlepiece grid would be difficult to handle in the limited work space and because it isnecessary that control rod drives, position indicators, and incore instrumentation components are accessible for inspection and maintenance, each grid is designed to be assembled or disassembled in place. Each grid assembly is made from two grid plates, a clamp and a bolt. The top part of the clamp acts asa guide to assure that each grid is correctly positioned directly below therespective CRD housing which it would support in the postulated accident.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-06.05Doc Type:Admin Initials:Date:9703 Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 3I/mabWhen the support bars and grids are installed, a gap of less than 1 inch at roomtemperature (approximately 70°F) is provided between the grid and the bottomcontact surface of the control rod drive flange. During system heatup, this gap is reduced by a net downward expansion of the housings with respect to the supports. In the hot operating condition, the gap is approximately 1/4 inch.In the postulated CRD housing failure, the CRD housing supports are loadedwhen the lower contact surface of the CRD flange contacts the grid. The resulting load is then carried by two grid plates, two support bars, four hangerrods, their disk springs, and two adjacent beams.The American Institute of Steel Construction (AISC) Specification for the Design,Fabrication, and Erection of Structural Steel for Building was used in the design of the CRD housing support system. However, to provide a structure that absorbs as much energy as practical without yielding, the allowable tension and bending stresses were taken as 90% of yield, and the shear stress as 60% ofyield. These are 1.5 times the corresponding AISC allowable stresses of 60%and 40% of yield. This stress criterion is considered desirable for this application and adequate for the once in a lifetime loading condition.For mechanical design purposes, the postulated failure resulting in the highestforces is an instantaneous circumferential separation of the CRD housing fromthe reactor vessel, with an internal pressure of 1250 psig (reactor vessel designpressure) acting on the area of the separated housing. The weight of the separated housing, control rod drive, and blade, plus the force of 1250 psig pressure acting on the area of the separated housing gives a force of approximately 35,000 lbs. This force is multiplied by a factor of 3 for impact,conservatively assuming the housing travels through a 1-inch gap beforecontacting the supports. The total force (105 lbs) is then treated as a static loadin design formulas.6.5.3Performance AnalysisDownward travel of CRD housing and its control rod following the postulatedhousing failure is the sum of the compression of the disk springs under dynamicloading and the initial gap between the grid and the bottom contact surface of the CRD flange. If the reactor were cold and pressurized, the downward motion of the control rod would be limited to the approximate 2 inch spring compressionplus a gap of less than one inch. If the reactor were hot and pressurized, thegap would be approximately 1/4 inch and the spring compression slightly less than in the cold condition. In either case, the control rod movement following a housing failure is limited substantially below one drive notch movement (6 inches). The nuclear transient from sudden withdrawal of any control rodthrough a distance of one drive notch at any position in the core does not resultin a transient sufficient to cause damage to any radioactive material barrier. This meets the fuel damage prevention criteria of design basis 6.5.1-a.
Revision 22USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 3I/mabThe control rod drive housing supports are in place any time the reactor is to beoperated. The housing supports may be removed when the reactor is in theshutdown condition even when the reactor is pressurized, because all controlrods are then inserted. Even if a control rod is ejected under the shutdown condition, the reactor remains subcritical, because it is designed to remain subcritical with any one control rod fully withdrawn at any time.At plant operating temperature a gap of approximately 1/4 inch is maintainedbetween the CRD housing and the supports, at lower temperatures the gap isgreater. Because the supports do not come in contact with any of the CRD housings, except during the postulated accident condition, vertical contact stresses are prevented as required by safety design basis 6.5.1-b.6.5.4Inspection and TestingWhen the reactor is in the shutdown mode, the control rod drive housing supports may be removed to permit inspection and maintenance of the control rod drives. When the support structure is reinstalled, it is inspected for proper assembly, particular attention being given to assure that the correct gap between the CRD flange lower contact surface and the grid is maintained. Since thestructure is not stressed until an accident occurs, testing is unnecessary. If anaccident should occur any deformed parts would be replaced during repair.
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.6 Standby Liquid Control System 6.6.1 Design Basis
6.6.2 Description
6.6.3 Performance Analysis
6.6.4 Inspection and Testing
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.7 Main Control Room, Emergency Filtration Train Building and Technical Support Center Habitability 6.7.1 Design Basis
6.7.2 Description
6.7.3 Performance Analysis
6.7.4 Inspection and Testing
SECTION 6 PLANT ENGINEERED SAFEGUARDS 6.8 References
Revision 29USAR-06.FIGMONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 10I/arb Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 10USAR-06.FIGI/arbFigure 6.2-2 Typical Core Spray Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 10USAR-06.FIGI/arbFigure 6.2-4 Typical RHR Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 4 of 10USAR-06.FIGI/arbFigure 6.2-5 RHR - Simplified P&ID - LPCI ModeSelecting Specified LoopTORUS RING HEADERPUMPSJETHxRHRPUMPSRHRCONDENSATESTORAGETANK Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 5 of 10USAR-06.FIGI/arbFigure 6.2-6 RHR - Simplified P&ID - Containment Spray/Cooling ModeRHRHxPUMPSRHRTORUS RING HEADERRECIRC PUMPSTORUSPUMPSJETHxRHRPUMPSRHR Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 6 of 10USAR-06.FIGI/arbFigure 6.2-7a LPCI System - Logic Control System ArrangementREACTORVESSELLPCISPUMPPUMPP1SBPWWPPPPPPPPPPPAAAAAAAAABBBBBBBBIISD42DPPBPAPAB1, 2, 3, 4= P P(Riser Differential Pressure)WHERE:A/B indicates recirc loop D = Ricirc Pump Discharge ValveI = RHR/LPCI Injection ValvesP = Pressure S = Recirc Pump Suction Valve W = LPCI Injection Water3 Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 7 of 10USAR-06.FIGI/arbFigure 6.2-7b LPCI System - Loop Selection/Break DetectionFunctional Block DiagramYESReactor Pressure > 420 psigpermissiveReactor pressure< 900 psigPermissiveInject LPCI in Loop AClose A Recirc Pump Discharge ValveOpen A LPCI Injection ValveBlock Closed B LPCI Injection ValvesClose Containment Spray/Cooling ValvesInject LPCI in Loop BClose B Recirc Pump Discharge ValveOpen B LPCI Injection Valve Block Closed A LPCI Injection ValvesClose Containment Spray/Cooling ValvesReactor Pressure > 420 psigpermissiveLoop A SelectedRecirc Riser PPA> PBPA= PBPB> PALoop B Selected2 Sec Time DelayRecircPump A runningNTSP of > 3.4 psid( >87.2 inWC)High DrywellPressureLow-Low Reactor Water LevelTrip Recirc PumpDrive Motor BreakerYESNONORecircPump B runningNTSP of > 3.4 psid( > 87.2 inWC)
Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 8 of 10USAR-06.FIGI/arbFigure 6.3-1 Main Steamline Flow Restrictor Nozzle Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 9 of 10USAR-06.FIGI/arbFigure 6.4-1 Control Rod Velocity Limiter IsometricHANDLE MODELD120 EXTENDEDHANDLEBLADESHEATHRODSNEUTRON ABSORBERVELOCITY LIMITERHANDLECOUPLING RELEASECOUPLING SOCKET143"6.5" Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 10 of 10USAR-06.FIGI/arbFigure 6.5-1 Control Rod Drive Housing Support Isometric01294216