ML16054A420

Revision 33 to the Updated Final Safety Analysis Report, Section 6, Plant Engineered Safeguards

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: L-MT-16-004, Revision 33 to Updated Final Safety Analysis Report, Drawings NF-36058 Through NH-36243-1, Rev. 82 (Public) ML16054A388 ML16054A413 ML16054A414 ML16054A415 ML16054A416 ML16054A417 ML16054A418 ML16054A419 ML16054A420 ML16054A421 ML16054A422 ML16054A423 ML16054A424 ML16054A425 ML16054A426 ML16054A427 ML16054A428 ML16054A429 ML16054A432 ML16054A434 ML16054A435 ML16054A436 ML16054A438 ML16054A439 ML16054A440 ML16054A441 ML16054A442 ML16054A443 ML16054A444 ML16054A445 ML16054A446 ML16054A447 ML16054A448 ML16155A258 ML16155A259
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 Syst em

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

6.3.3 Performance

Analysis

6.3.4 Inspection

and Testing

Revision 22 USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 2SECTION 6PLANT ENGINEERED SAFEGUARDS I/mab6.4Control Rod Velocity Limiters A topical report describing and analyzing the control rod velocity limiter was submitted separately to the Atomic Energy Commission as APED-5446, Control Rod Velocity Limiter, (Reference 32) and is incorporated herein by reference.6.4.1Design Basis The 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 limiter was 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 ability is 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 steel casting 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 peripheral separation less than the central separation.The velocity limiter assembly, shown in Figure 6.4-1 with its associated components, 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 cone and discharged through the interface between the two conical sections. Because this 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. Each guide 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 limiter free-fall; the seat also restricts water flow into the interior of the guide tube during normal reactor operation to prevent coolant bypass of the fuel elements.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTP Assoc Ref:

SR:2yrs N Freq: USAR-MANARMS:USAR-06.04Doc Type:Admin Initials:Date:

9703 Revision 22 USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 2 of 2 I/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 the velocity 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 Velocities Cold reactor - 2.46 ft/sec Hot 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 during the 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 22 USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 3SECTION 6PLANT ENGINEERED SAFEGUARDS I/mab6.5Control Rod Drive Housing Supports6.5.1Design Basis The control rod drive housing supports protect against additional damage to the nuclear 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 drive housing supports were designed in accordance with the following:

Design Basisa.Control rod downward motion shall be limited, following a postulated control 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 to prevent 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 from the 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. The spring pivots at the top and the beveled loose-fitting ends on the support bars prevent 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 is necessary 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 as a guide to assure that each grid is correctly positioned directly below the respective CRD housing which it would support in the postulated accident.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTP Assoc Ref:

SR:2yrs N Freq: USAR-MANARMS:USAR-06.05Doc Type:Admin Initials:Date:

9703 Revision 22 USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 2 of 3 I/mabWhen the support bars and grids are installed, a gap of less than 1 inch at room temperature (approximately 70

°F) is provided between the grid and the bottom contact 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 loaded when 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 hanger rods, 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% of yield. 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 highest forces is an instantaneous circumferential separation of the CRD housing from the reactor vessel, with an internal pressure of 1250 psig (reactor vessel design pressure) 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 before contacting the supports. The total force (10 5 lbs) is then treated as a static load in design formulas.6.5.3Performance Analysis Downward travel of CRD housing and its control rod following the postulated housing failure is the sum of the compression of the disk springs under dynamic loading 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 compression plus a gap of less than one inch. If the reactor were hot and pressurized, the gap 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 rod through 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 22 USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 3 of 3 I/mabThe control rod drive housing supports are in place any time the reactor is to be operated. The housing supports may be removed when the reactor is in the shutdown condition even when the reactor is pressurized, because all control rods 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 maintained between 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 Testing When 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

6.6.3 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

6.7.3 Performance

Analysis

6.7.4 Inspection

and Testing

SECTION 6 PLANT ENGINEERED SAFEGUARDS

6.8 References

Revision 29USAR-06.FIGMONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 10I/arb Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 2 of 10USAR-06.FIG I/arbFigure 6.2-2 Typical Core Spray Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 3 of 10USAR-06.FIG I/arbFigure 6.2-4 Typical RHR Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 4 of 10USAR-06.FIG I/arbFigure 6.2-5 RHR - Simplified P&ID - LPCI Mode Selecting Specified LoopTORUS RING HEADER PUMPS JET Hx RHR PUMPS RHRCONDENSATESTORAGETANK Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 5 of 10USAR-06.FIG I/arbFigure 6.2-6 RHR - Simplified P&ID - Containment Spray/Cooling Mode RHR Hx PUMPS RHRTORUS RING HEADER RECIRC PUMPSTORUS PUMPS JET Hx RHR PUMPS RHR Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 6 of 10USAR-06.FIG I/arbFigure 6.2-7a LPCI System - Logic Control System ArrangementREACTOR VESSEL LPCIS PUMP PUMP P 1 S B P W W P P P P P P P P P P P A A A A A A A A A B B B B B B B B I I S D 4 2 D P P B P A P A B 1, 2, 3, 4= P P(Riser Differential Pressure)

WHERE:A/B indicates recirc loop D = Ricirc Pump Discharge ValveI = RHR/LPCI Injection Valves P = Pressure S = Recirc Pump Suction Valve W = LPCI Injection Water 3

Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 7 of 10USAR-06.FIG I/arbFigure 6.2-7b LPCI System - Loop Selection/Break Detection Functional Block Diagram YES Reactor Pressure

> 420 psig permissive Reactor pressure

< 900 psig Permissive Inject LPCI in Loop AClose A Recirc Pump Discharge ValveOpen A LPCI Injection ValveBlock Closed B LPCI Injection ValvesClose Containment Spray/Cooling Valves Inject LPCI in Loop BClose B Recirc Pump Discharge ValveOpen B LPCI Injection Valve Block Closed A LPCI Injection ValvesClose Containment Spray/Cooling Valves Reactor Pressure

> 420 psig permissive Loop A Selected Recirc Riser P P A> P B P A= P B P B> P ALoop B Selected2 Sec Time Delay Recirc Pump A running NTSP of >

3.4 psid( >87.2 inWC)

High Drywell Pressure Low-Low Reactor Water LevelTrip Recirc Pump Drive Motor Breaker YES NO NO Recirc Pump B running NTSP of > 3.4 psid ( > 87.2 inWC)

Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 8 of 10USAR-06.FIG I/arbFigure 6.3-1 Main Steamline Flow Restrictor Nozzle Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 9 of 10USAR-06.FIG I/arbFigure 6.4-1 Control Rod Velocity Limiter Isometric HANDLE MODEL D120 EXTENDED HANDLE BLADESHEATH RODS NEUTRON ABSORBER VELOCITY LIMITER HANDLE COUPLING RELEASE COUPLING SOCKET 143"6.5" Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 10 of 10USAR-06.FIG I/arbFigure 6.5-1 Control Rod Drive Housing Support Isometric 01294216 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 Syst em

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

6.3.3 Performance

Analysis

6.3.4 Inspection

and Testing

Revision 22 USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 2SECTION 6PLANT ENGINEERED SAFEGUARDS I/mab6.4Control Rod Velocity Limiters A topical report describing and analyzing the control rod velocity limiter was submitted separately to the Atomic Energy Commission as APED-5446, Control Rod Velocity Limiter, (Reference 32) and is incorporated herein by reference.6.4.1Design Basis The 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 limiter was 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 ability is 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 steel casting 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 peripheral separation less than the central separation.The velocity limiter assembly, shown in Figure 6.4-1 with its associated components, 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 cone and discharged through the interface between the two conical sections. Because this 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. Each guide 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 limiter free-fall; the seat also restricts water flow into the interior of the guide tube during normal reactor operation to prevent coolant bypass of the fuel elements.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTP Assoc Ref:

SR:2yrs N Freq: USAR-MANARMS:USAR-06.04Doc Type:Admin Initials:Date:

9703 Revision 22 USAR 6.4MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 2 of 2 I/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 the velocity 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 Velocities Cold reactor - 2.46 ft/sec Hot 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 during the 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 22 USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 3SECTION 6PLANT ENGINEERED SAFEGUARDS I/mab6.5Control Rod Drive Housing Supports6.5.1Design Basis The control rod drive housing supports protect against additional damage to the nuclear 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 drive housing supports were designed in accordance with the following:

Design Basisa.Control rod downward motion shall be limited, following a postulated control 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 to prevent 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 from the 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. The spring pivots at the top and the beveled loose-fitting ends on the support bars prevent 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 is necessary 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 as a guide to assure that each grid is correctly positioned directly below the respective CRD housing which it would support in the postulated accident.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTP Assoc Ref:

SR:2yrs N Freq: USAR-MANARMS:USAR-06.05Doc Type:Admin Initials:Date:

9703 Revision 22 USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 2 of 3 I/mabWhen the support bars and grids are installed, a gap of less than 1 inch at room temperature (approximately 70

°F) is provided between the grid and the bottom contact 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 loaded when 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 hanger rods, 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% of yield. 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 highest forces is an instantaneous circumferential separation of the CRD housing from the reactor vessel, with an internal pressure of 1250 psig (reactor vessel design pressure) 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 before contacting the supports. The total force (10 5 lbs) is then treated as a static load in design formulas.6.5.3Performance Analysis Downward travel of CRD housing and its control rod following the postulated housing failure is the sum of the compression of the disk springs under dynamic loading 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 compression plus a gap of less than one inch. If the reactor were hot and pressurized, the gap 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 rod through 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 22 USAR 6.5MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 3 of 3 I/mabThe control rod drive housing supports are in place any time the reactor is to be operated. The housing supports may be removed when the reactor is in the shutdown condition even when the reactor is pressurized, because all control rods 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 maintained between 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 Testing When 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

6.6.3 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

6.7.3 Performance

Analysis

6.7.4 Inspection

and Testing

SECTION 6 PLANT ENGINEERED SAFEGUARDS

6.8 References

Revision 29USAR-06.FIGMONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 10I/arb Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 2 of 10USAR-06.FIG I/arbFigure 6.2-2 Typical Core Spray Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 3 of 10USAR-06.FIG I/arbFigure 6.2-4 Typical RHR Pump Characteristics Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 4 of 10USAR-06.FIG I/arbFigure 6.2-5 RHR - Simplified P&ID - LPCI Mode Selecting Specified LoopTORUS RING HEADER PUMPS JET Hx RHR PUMPS RHRCONDENSATESTORAGETANK Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 5 of 10USAR-06.FIG I/arbFigure 6.2-6 RHR - Simplified P&ID - Containment Spray/Cooling Mode RHR Hx PUMPS RHRTORUS RING HEADER RECIRC PUMPSTORUS PUMPS JET Hx RHR PUMPS RHR Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 6 of 10USAR-06.FIG I/arbFigure 6.2-7a LPCI System - Logic Control System ArrangementREACTOR VESSEL LPCIS PUMP PUMP P 1 S B P W W P P P P P P P P P P P A A A A A A A A A B B B B B B B B I I S D 4 2 D P P B P A P A B 1, 2, 3, 4= P P(Riser Differential Pressure)

WHERE:A/B indicates recirc loop D = Ricirc Pump Discharge ValveI = RHR/LPCI Injection Valves P = Pressure S = Recirc Pump Suction Valve W = LPCI Injection Water 3

Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 7 of 10USAR-06.FIG I/arbFigure 6.2-7b LPCI System - Loop Selection/Break Detection Functional Block Diagram YES Reactor Pressure

> 420 psig permissive Reactor pressure

< 900 psig Permissive Inject LPCI in Loop AClose A Recirc Pump Discharge ValveOpen A LPCI Injection ValveBlock Closed B LPCI Injection ValvesClose Containment Spray/Cooling Valves Inject LPCI in Loop BClose B Recirc Pump Discharge ValveOpen B LPCI Injection Valve Block Closed A LPCI Injection ValvesClose Containment Spray/Cooling Valves Reactor Pressure

> 420 psig permissive Loop A Selected Recirc Riser P P A> P B P A= P B P B> P ALoop B Selected2 Sec Time Delay Recirc Pump A running NTSP of >

3.4 psid( >87.2 inWC)

High Drywell Pressure Low-Low Reactor Water LevelTrip Recirc Pump Drive Motor Breaker YES NO NO Recirc Pump B running NTSP of > 3.4 psid ( > 87.2 inWC)

Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 8 of 10USAR-06.FIG I/arbFigure 6.3-1 Main Steamline Flow Restrictor Nozzle Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 9 of 10USAR-06.FIG I/arbFigure 6.4-1 Control Rod Velocity Limiter Isometric HANDLE MODEL D120 EXTENDED HANDLE BLADESHEATH RODS NEUTRON ABSORBER VELOCITY LIMITER HANDLE COUPLING RELEASE COUPLING SOCKET 143"6.5" Revision 29MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 10 of 10USAR-06.FIG I/arbFigure 6.5-1 Control Rod Drive Housing Support Isometric 01294216