ML17212A074

Final Safety Analysis Report, Rev. 30, Chapter 5, Reactor Coolant System and Connected Systems

ML17212A074
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Site:	Millstone
Issue date:	06/29/2017
From:	Dominion Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation
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MPS-3 FSARMillstone Power Station Unit 3 Safety Analysis Report Chapter 5 MPS-3 FSAR 5-i Rev. 30CHAPTER 5 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Table of ContentsSection Title Page5.1

SUMMARY

DESCRIPTION..............................................................................5.1-15.1.1Schematic Flow Diagram............................................................................5.1-65.1.2Piping and Instrumentation Diagram..........................................................5.1-65.1.3Elevation Drawing......................................................................................5.1-65.2INTEGRITY OF REACTOR COOL ANT PRESSURE BOUNDARY..............5.2-15.2.1Compliance with Codes and Code Cases...................................................5.2-15.2.1.1Compliance with 10 CFR 50.55a................................................................5.2-15.2.1.2Applicable Code Cases...............................................................................5.2-15.2.2Overpressure Pr otection..............................................................................5.2-25.2.2.1Design Bases...............................................................................................5.2-35.2.2.2Design Evaluation.......................................................................................5.2-35.2.2.3Piping and Instrument ation Diagrams........................................................5.2-45.2.2.4Equipment and Componen t Description.....................................................5.2-45.2.2.5Mounting of Pressure-R elief Devices.........................................................5.2-45.2.2.6Applicable Codes and Classification..........................................................5.2-55.2.2.7Material Specifi cations...............................................................................5.2-55.2.2.8Process Instrume ntation..............................................................................5.2-55.2.2.9System Reliab ility.......................................................................................5.2-65.2.2.10Testing and Insp ection................................................................................5.2-65.2.2.11RCS Pressure Control during Lo w Temperature Operation.......................5.2-65.2.2.11.1System Oper ation........................................................................................5.2-65.2.2.11.2Evaluation of Low Temperatur e Overpressure Transients.........................5.2-75.2.2.11.3Operating Basis Eart hquake Evaluation.....................................................5.2-75.2.2.11.4Administrative Procedures..........................................................................5.2-75.2.3Materials Selection, Fabric ation, and Processing.....................................5.2-105.2.3.1Material Specifi cations.............................................................................5.2-105.2.3.2Compatibility With Re actor Coolant........................................................5.2-115.2.3.2.1Chemistry of React or Coolant..................................................................5.2-115.2.3.2.2Compatibility of C onstruction Materials with Reactor Coolant...............5.2-115.2.3.2.3Compatibility with Ex ternal Insulation and Envi ronmental Atmosphere 5.2-125.2.3.3Fabrication and Processing of Ferritic Materials......................................5.2-125.2.3.3.1Fracture T oughness...................................................................................5.2-125.2.3.3.2Control of Wel ding...................................................................................5.2-135.2.3.3.3Pressurized Ther mal Shock......................................................................5.2-135.2.3.4Fabrication and Processing of Austenitic Stainless Steel.........................5.2-145.2.3.4.1Cleaning and Contaminati on Protection Procedures................................5.2-145.2.3.4.2Solution Heat Trea tment Requirements

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5.2-14 MPS-3 FSARCHAPTER 5 -REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Table of Contents (Continued)

Section Title Page 5-ii Rev. 305.2.3.4.3Material Inspection Program....................................................................5.2-155.2.3.4.4Prevention of Intergranular Attack of Unstabilized Austenitic Stainless Steels.

5.2-155.2.3.4.5Retesting Unstabilized Austenitic Stainless Steel Exposed to Sensitization Temperatures............................................................................................5.2-175.2.3.4.6Control of Wel ding...................................................................................5.2-185.2.4Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.2-195.2.4.1System Boundary Subject to Inspection...................................................5.2-195.2.4.2Accessibili ty..............................................................................................5.2-205.2.4.3Examination Techniques and Procedures.................................................5.2-225.2.4.4Inspection Inte rvals...................................................................................5.2-235.2.4.5Examination Categories and Requirements..............................................5.2-23 5.2.4.6Evaluation of Examination Results...........................................................5.2-235.2.4.7System Leakage and Hydrostatic Pressure Tests......................................5.2-235.2.4.8Relief Reque sts.........................................................................................5.2-235.2.5Detection of Leakage Through React or Coolant Pressure Boundary.......5.2-245.2.5.1Identified Leakage....................................................................................5.2-245.2.5.1.1Definition of Identified Leakage...............................................................5.2-245.2.5.1.2Collection and Monitoring of Identified Leakage....................................5.2-245.2.5.1.3Controlled L eakage...................................................................................5.2-255.2.5.2Unidentified Leakage................................................................................5.2-265.2.5.2.1Definition of Unidentified Leakage..........................................................5.2-265.2.5.2.2Collection of Uniden tified Leakage..........................................................5.2-265.2.5.2.3Detection of Unidentified Leakage...........................................................5.2-265.2.5.2.4Leakage Detection Method Sensitivity and Response Times...................5.2-275.2.5.2.5Leakage Detection Method Indicators and Alarms..................................5.2-275.2.5.2.6Seismic Capability of L eakage Detection Methods..................................5.2-285.2.5.2.7Testing and Cali bration.............................................................................5.2-285.2.5.3Intersystem Leakage.................................................................................5.2-28 5.2.5.4Technical Specifi cations...........................................................................5.2-315.2.5.5Primary Coolant Sources Ou tside Containment.......................................5.2-315.2.6References for Section 5.2........................................................................5.2-325.3REACTOR VESSEL...........................................................................................5.3-15.3.1Reactor Vessel Materials............................................................................5.3-15.3.1.1Material Specifi cations...............................................................................5.3-15.3.1.2Special Process Used for Manuf acturing and Fabrication..........................5.3-15.3.1.3Special Methods for Nondestructive Examination.....................................5.3-25.3.1.3.1Ultrasonic Exa mination..............................................................................5.3-25.3.1.3.2Penetrant Ex aminations

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5.3-2 MPS-3 FSARCHAPTER 5 -REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Table of Contents (Continued)

Section Title Page 5-iii Rev. 305.3.1.3.3Magnetic Particle Examination...................................................................5.3-35.3.1.4Special Controls for Ferritic and Austenitic Stainless Steels.....................5.3-35.3.1.5Fracture Toughness.....................................................................................5.3-45.3.1.6Material Surve illance..................................................................................5.3-45.3.1.6.1Measurement of Integr ated Fast Neutron (E

> 1.0 MeV) Flux at the Irradiation Samples.......................................................................................................5.3-75.3.1.6.2Calculation of Integrated Fast Neut ron (E. 1.0 MeV) Flux at the Irradiation Samples.....................................................................................................5.3-105.3.1.7Reactor Vessel Fasteners..........................................................................5.3-12 5.3.2Pressure-Temperature Limits....................................................................5.3-125.3.2.1Limit Curves.............................................................................................5.3-125.3.2.2End-of-Life RT PTS Projections.................................................................5.3-135.3.2.3Operating Proc edures................................................................................5.3-145.3.3Reactor Vessel In tegrity............................................................................5.3-145.3.3.1Design.......................................................................................................5.3-14 5.3.3.2Materials of C onstruction.........................................................................5.3-155.3.3.3Fabrication Me thods.................................................................................5.3-155.3.3.4Inspection Requirements...........................................................................5.3-15 5.3.3.5Shipment and Installation.........................................................................

5.3-155.3.3.6Operating Cond itions................................................................................5.3-165.3.3.7Inservice Surveillance...............................................................................5.3-17 5.3.4References for Section 5.3........................................................................5.3-195.4COMPONENT AND SUBS YSTEM DESIGN...................................................5.4-15.4.1Reactor Coolant Pumps..............................................................................5.4-15.4.1.1Pump Flywheel Integrity............................................................................5.4-1 5.4.1.1.1Design Bases...............................................................................................5.4-15.4.1.1.2Fabrication and Inspection..........................................................................5.4-15.4.1.1.3Material Acceptance Criteria......................................................................5.4-15.4.1.2Reactor Coolant Pu mp Assembly...............................................................5.4-25.4.1.2.1Design Bases...............................................................................................5.4-2 5.4.1.2.2Pump Assembly De scription......................................................................5.4-25.4.1.3Design Evaluation.......................................................................................5.4-45.4.1.3.1Pump Performa nce......................................................................................5.4-45.4.1.3.2Coastdown Capa bility.................................................................................5.4-65.4.1.3.3Bearing Integr ity.........................................................................................5.4-65.4.1.3.4Locked Roto r..............................................................................................5.4-75.4.1.3.5Critical Sp eed..............................................................................................5.4-75.4.1.3.6Missile Gene ration......................................................................................5.4-75.4.1.3.7Pump Cav itation.........................................................................................

5.4-7 MPS-3 FSARCHAPTER 5 -REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Table of Contents (Continued)

Section Title Page 5-iv Rev. 305.4.1.3.8Pump Overspeed Considerations................................................................5.4-85.4.1.3.9Anti-Reverse Rotation Device....................................................................5.4-85.4.1.3.10Shaft Seal L eakage......................................................................................5.4-85.4.1.3.11Seal Discharge Piping.................................................................................5.4-95.4.1.4Tests and Inspections..................................................................................5.4-95.4.2Steam Generators........................................................................................5.4-9 5.4.2.1Steam Generator Materials.........................................................................5.4-95.4.2.1.1Selection and Fabricati on of Materials.......................................................5.4-95.4.2.1.2Steam Generator Design Effects on Material...........................................5.4-105.4.2.1.3Compatibility of Stea m Generator Tubing with Primary and Secondary Cool

-ants............................................................................................................5.4-115.4.2.1.4Cleanup of Secondary Side Materials.......................................................5.4-125.4.2.2Steam Generator Inservice Inspection......................................................5.4-125.4.2.3Design Basis.............................................................................................5.4-135.4.2.4Design Description...................................................................................5.4-14 5.4.2.5Design Evaluation.....................................................................................5.4-155.4.2.6Quality Assurance.....................................................................................5.4-175.4.3Reactor Coolant Piping.............................................................................5.4-185.4.3.1Design Bases.............................................................................................5.4-185.4.3.2Design Description...................................................................................5.4-185.4.3.3Design Evaluation.....................................................................................5.4-21 5.4.3.3.1Material Corrosion/Er osion Evaluation....................................................5.4-215.4.3.3.2Sensitized Stai nless Steel..........................................................................5.4-215.4.3.3.3Contaminant C ontrol.................................................................................5.4-215.4.3.4Tests and Inspections................................................................................5.4-225.4.4Main Steam Line Flow Restrictor.............................................................5.4-225.4.4.1Design Basis.............................................................................................5.4-225.4.4.2Design Description...................................................................................5.4-225.4.4.3Design Evaluation.....................................................................................5.4-225.4.4.4Tests and Inspections................................................................................5.4-23 5.4.5Main Steam Isolation System...................................................................5.4-235.4.6Reactor Core Isolati on Cooling System....................................................5.4-235.4.7Residual Heat Removal System................................................................5.4-235.4.7.1Design Bases.............................................................................................5.4-235.4.7.2System Design..........................................................................................5.4-255.4.7.2.1Schematic Piping and Inst rumentation Diagrams.....................................5.4-255.4.7.2.2Equipment and Componen t Descriptions.................................................5.4-275.4.7.2.3System Operat ion......................................................................................5.4-295.4.7.2.4Control......................................................................................................5.4-345.4.7.2.5Applicable Codes and Classifications.......................................................5.4-355.4.7.2.6System Reliability Considerations

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5.4-35 MPS-3 FSARCHAPTER 5 -REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Table of Contents (Continued)

Section Title Page 5-v Rev. 305.4.7.2.7Manual Actions.........................................................................................5.4-375.4.7.3Performance Evaluation............................................................................5.4-385.4.7.4Preoperational Te sting..............................................................................5.4-385.4.8Reactor Water Cleanup System................................................................5.4-385.4.9Main Steamlines and Feedwater Piping....................................................5.4-385.4.10Pressurize r.................................................................................................5.4-385.4.10.1Design Bases.............................................................................................5.4-385.4.10.1.1Pressurizer Su rge Line..............................................................................5.4-395.4.10.1.2Pressurize r.................................................................................................5.4-395.4.10.2Design Description...................................................................................5.4-39 5.4.10.2.1Pressurizer Su rge Line..............................................................................5.4-395.4.10.2.2Pressurize r.................................................................................................5.4-405.4.10.3Design Evaluation.....................................................................................5.4-41 5.4.10.3.1System Pre ssure........................................................................................5.4-415.4.10.3.2Pressurizer Pe rformance...........................................................................5.4-415.4.10.3.3Pressure Setp oints.....................................................................................5.4-415.4.10.3.4Pressurizer Spray......................................................................................5.4-415.4.10.3.5Pressurizer Desi gn Analysis.....................................................................5.4-425.4.10.3.6Natural Circulation Following Loss of Off Site Power............................5.4-435.4.10.4Inspection and Testing Requirements.......................................................5.4-435.4.10.5Instrumentation Requirements..................................................................5.4-44 5.4.11Pressurizer Relief Di scharge System........................................................5.4-445.4.11.1Design Basis.............................................................................................5.4-44 5.4.11.2System Description...................................................................................5.4-44 5.4.11.2.1Pressurizer Reli ef Tank.............................................................................5.4-455.4.11.3Safety Evaluation......................................................................................5.4-45 5.4.11.4Instrumentation Requirements..................................................................5.4-46 5.4.11.5Inspection and Testing Requirements.......................................................5.4-465.4.12Valves.......................................................................................................5.4-465.4.12.1Design Bases.............................................................................................5.4-46 5.4.12.2Design Description...................................................................................5.4-475.4.12.3Design Evaluation.....................................................................................5.4-475.4.12.4Tests and Inspections................................................................................5.4-48 5.4.13Safety and Relief Valves...........................................................................5.4-485.4.13.1Design Bases.............................................................................................5.4-48 5.4.13.2Design Description...................................................................................5.4-485.4.13.3Design Evaluation.....................................................................................5.4-495.4.13.4Inspection and Testing Requirements.......................................................5.4-49 5.4.14Component S upports.................................................................................5.4-505.4.14.1Description................................................................................................5.4-50 5.4.14.1.1Reactor Vessel Stru ctural Support (RVSS)

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5.4-50 MPS-3 FSARCHAPTER 5 -REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Table of Contents (Continued)

Section Title Page 5-vi Rev. 305.4.14.1.2Steam Generator Supports........................................................................5.4-505.4.14.1.3Reactor Coolant Pump Supports...............................................................5.4-515.4.14.1.4Pressurizer Support...................................................................................5.4-515.4.14.1.5Pressurizer Safety Valve Supports............................................................5.4-515.4.14.2Design Basis.............................................................................................5.4-525.4.14.3Evaluation.................................................................................................5.4-52 5.4.14.4Tests and Inspections................................................................................5.4-535.4.15Reactor Vessel Head Vent System...........................................................5.4-535.4.15.1Design Basis.............................................................................................5.4-535.4.15.2System Description...................................................................................5.4-545.4.15.3Safety Evaluation......................................................................................5.4-555.4.15.4Inspection and Testing Requirements.......................................................5.4-55 5.4.15.5Instrumentation Requirements..................................................................5.4-555.4.16References for Section 5.4........................................................................

5.4-55 MPS-3 FSAR 5-vii Rev. 30CHAPTER 5-REACTOR COOLANT SY STEM AND CONNECTED SYSTEMS List of Tables Number Title5.1-1System Design and Operating Parameters5.2-1Applicable Code Addenda for Class 1 Reactor Coolant System Components5.2-2Primary and Auxiliary Compone nts Material Specifications5.2-3Reactor Vessels Internal Material Specifications 5.2-4Reactor Coolant Water Chemistry Specification5.2-5Safety Valve Support Bracket Loads 5.2-6Relief Valves Referenced to Code Case N-242 5.2-7Millstone Unit No. 3 rt pts Values (°F)5.3-1Reactor Vessel Non-De structive Examination5.3-2Reactor Vessel Fract ure Toughness Properties5.3-3Reactor Vessel Beltline Region Materi al Chemical Compos ition (wt Percent)5.3-4Adjusted Referenced Temperature (ART) Values (°F)5.3-5Reactor Vessel Design Parameters 5.4-1Reactor Coolant Pump Design Parameters 5.4-2Reactor Coolant Pump Non-Destructive Examination Program 5.4-3Steam Generator Design Data 5.4-4Steam Generator Nondestructive Examination Program 5.4-5Reactor Coolant Piping Design Parameters 5.4-6Reactor Coolant Piping Quality Assurance Program 5.4-7Design Bases for Residual Heat Removal System Operation5.4-8Residual Heat Removal System Component Data 5.4-9Residual Heat Removal System - Cold Shutdown Operations-Failure Modes and Effects Analysis5.4-10Pressurizer Design Data 5.4-11Reactor Coolant System Design Pressure Settings5.4-12Pressurizer Quality Assurance Program 5.4-13Pressure Relief Tank Design Data MPS-3 FSAR List of Tables (Continued)

Number Title 5-viii Rev. 305.4-14Relief Valve Discharge to the Pressurizer Relief Tank5.4-15Reactor Coolant Syst em Design Parameters5.4-16Non-Destructive Examination Prog ram Reactor Coolant System Valves5.4-17Pressurizer Valves Design Parameters 5.4-18Equipment Supports, Loading Combinat ions, and Design Allowable Stresses5.4-19Reactor Vessel Head Vent Sy stem Equipment Design Parameters MPS-3 FSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

5-1 Rev. 30CHAPTER 5 - REACTOR COOLANT SY STEM AND CONNECTED SYSTEMS List of Figures Number Title5.1-1(Sheets 1-63) P&IDs Reactor Coolant System5.1-2Reactor Coolant System Process Flow Design Notes to Figure 5.1-2: Mode A Steady State Full Power Operation5.2-1Reactor Vessel Inspection Area5.2-2Model F Steam Generator Inspection Area5.2-3Pressurizer Inspection Areas5.3-1Identification and Location of Beltline Region Material for the Reactor Vessel5.3-2Reactor Vessel5.4-1Reactor Coolant Pump5.4-2Reactor Coolant Pump Estimate d Performance Characteristics5.4-3Model F Steam Generator5.4-4Quatrefoil Tube Support Plates5.4-55 (Sheets 1-3) P&IDs Low Pressure Safe ty Injection / Contai nment Recirculation5.4-6Residual Heat Removal System Process Flow Diagram (Mode A)

Notes to Figure 5.4-6: Mode A Initiati on of Residual Heat Removal System Operation5.4-7Pressurizer Relief Tank 5.4-8Pressurizer5.4-9RPV Support System5.4-10Leveling Device (Typical) RPV Support System 5.4-11Vertical Supports (Typical) Reactor Coolant Pumps and Steam Generator5.4-12Lateral Support (Typi cal) Steam Generator5.4-13Lateral Support (Typical) Reactor Coolant Pump5.4-14Pressurizer Support 5.4-15Pressurizer Safety Valve Support System MPS3 UFSAR5.1-1Rev. 30CHAPTER 5 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1

SUMMARY

DESCRIPTIONThe reactor coolant system (RCS) (Figure 5.1-

1) consists of four similar heat transfer loops connected in parallel to the reactor pressure vessel. Each loop contains a reactor coolant pump and steam generator. In addition, the system includes a pressurizer, a pressurizer relief tank, interconnecting piping, valves, and instrumentation necessary for operational control. All these components are located in the containment building.During operation, the RCS transfers the heat generated in the core to the steam generators where steam is produced to drive the turbine generator. Borated demineraliz ed water is circulated in the RCS at a flow rate and temperature consistent with achieving the reactor core thermal-hydraulic performance. The water also acts as a neutron moderator and reflector, and as a solvent for the neutron absorber used in chemical shim control.

The RCS pressure boundary provides a barrier agai nst the release of radioactivity generated within the reactor and is designed to ensure a hi gh degree of integrity th roughout the life of the plant.

RCS pressure is controlled by the use of the pressurizer where wa ter and steam are maintained in equilibrium by electrical heaters and water sprays. Steam can be formed (by the heaters) or condensed (by the pressurizer spray) to minimize pressure variations due to contraction and expansion of the reactor coolant. Spring-loaded safety valves and power-operated relief valves are mounted on the pressurizer and discharge to th e pressurizer relief ta nk, where the steam is condensed and cooled by mixing with water.

The extent of the RCS is defined as:1.The reactor vessel including control rod drive mechanism housings2.The reactor coolant side of the steam generator3.The reactor coolant pumps 4.A pressurizer attached by a surge line to one of the reactor coolant loops5.The pressurizer relief tank6.The safety and relief valves7.The loop isolation valves 8.The interconnecting piping, valves, and fi ttings between the principal components listed above MPS3 UFSAR5.1-2Rev. 309. The piping, fittings, and valves leading to connecting auxiliar y or support systems Reactor Coolant System Components 1.Reactor Vessel (Section 5.3

)The reactor vessel is cylindr ical, with a welded hemispherical bottom head and a removable, flanged and gasketed, hemispherical upper head. The vessel contains the core, core support structures, contro l rods, and other components directly as sociated with the core.

The vessel has inlet and outlet nozzles lo cated in a horizontal plane below the reactor vessel flange but above the top of the core. Coolan t enters the vessel through the inlet nozzles, flows down the co re barrel-vessel wall annulus, and is then redirected at the bottom to flow up through the core and out the outlet nozzles.2.Steam Generators (Section 5.4.2

)The steam generators are vertical shell and U-tube evaporators with integral moisture separating equipment. The react or coolant flows through the inverted U-tubes, entering and leaving through the nozzles located in the hemispherical bottom head of the steam generator. Steam is generated on the shell side and flows upward through the moisture separators to the outlet nozzle at th e top of the vessel.3.Reactor Coolant Pumps (Section 5.4.1

)The reactor coolant pumps are single speed centrifugal units driven by air-cooled, three phase induction motors. The shaft is vertical with the motor mounted above the pumps. A flywheel on the shaft above the motor provides additional inertia to extend pump coastdown. The inlet is at the bottom of the pump; discharge is on the side.4.Piping (Section 5.4.3

)The reactor coolant loop piping is specified in sizes consistent with system requirements.

The hot leg inside diameter is 29 inches and the inside diameter of the cold leg return line to the react or vessel is 27.5 inches. The piping between the steam generator and the pump suction is increased to 31 inch inside diameter in order to reduce pressure drop and improve flow conditions facil itating pump suction.5.Pressurizer (Section 5.4.10

)The pressurizer is a vertical, cylindrical vessel with hemispherical top and bottom heads. Electrical heaters are installed th rough the bottom head of the vessel, while MPS3 UFSAR5.1-3Rev. 30 the spray nozzle, relief, and safety valve c onnections are located in the top head of the vess el. 6.Loop Isolation Valves (Section 5.4.3

)The reactor coolant loop isolation valves are remote controlled motor operated double disk gate valves. The hot and cold leg valves are identical except for the valve nozzles which are sized to ma tch the corresponding piping. The steam generator and RCP in each loop may be isolated from the reactor vessel by closing the isolation valves.7.Safety and Relief Valves (Section 5.4.13

)The pressurizer safety valves are of the totally enclosed pop-type. The valves are spring-loaded, self-actuated with back-pressure compensation. The power

operated relief valves are solenoid ope rated valves, which are operated automatically or by remote manual control. Remotely ope rated valves are provided to isolate the inlet to the power operated re lief valves if excessive leakage occurs.

Position indicating lights for these valv es are provided in the control room.8.Reactor Head Vent Piping (Section 5.4.15

)Reactor Coolant System Performance Characteristics Tabulations of important desi gn and performance characteristi cs of the RCS are provided in Table 5.1-1.1.Reactor Coolant FlowThe reactor coolant flow, a major paramete r in the design of the system and its components, is established with a detailed design procedure supported by operating plant performance data, by pu mp model tests and analysis, and by pressure drop tests and analyses of the reactor vessel and fuel assemblies. Data from all operating plants have indicated th at the actual flow has been well above the flow specified for the thermal desi gn of the plant. By applying the design procedure described below , it is possibl e to specify the expected operating flow with reasonable accuracy.

Three reactor coolant flow rates are identified for the various plant design considerations. The definitions of thes e flows are presented in the following paragraphs.2.Best Estimate Flow The best estimate flow is c onsidered to be the most likely value for the actual plant operating condition. This flow is based on th e best estimate of the reactor vessel, MPS3 UFSAR5.1-4Rev. 30 steam generator and piping flow resistance, and on the best estima te of the reactor coolant pump head flow capability with no uncertainties assigned to either the system flow resistance or the pump hea

d. System pressure drops, based on best estimate flow, are presented in Table 5.1-1. Although the best estimate flow is the most likely value to be expected in operation, more conservative flow rates are applied in the thermal and mechanical designs.3.Thermal Design Flow Thermal design flow is the basis for the reactor core thermal performance, the steam generator thermal performance, and th e nominal plant pa rameters used throughout the design. T o provide the required margin, the thermal design flow accounts for the uncertainties in reactor vessel, steam generator and piping flow resistances, reactor coolant pump head, and the methods used to measure flow rate.

The thermal design flow is approximately

6.7 percent

less than the best estimate flow and includes10 percent equivale nt steam generator tube plugging. The thermal design flow is confirmed when the plant is placed in operation. Tabulations of important design and performance characteristics of the RCS, based on thermal design flow, are provided in Table 5.1-1

.4.Mechanical Design Flow Mechanical design flow assumes 0 per cent equivalent steam generator tube plugging, and is the conservative ly high flow used in the mechanical design of the reactor vessel internals and fuel assemblies. To assure that a conservatively high flow is specified, the mechanical desi gn flow is based on a reduced system

resistance and on increased pump head capability. The mechanical design flow is approximately 3.9 percent greater than the best estimate flow.

Pump overspeed, due to a turbine generator overspeed of 20 percent, results in a peak reactor coolant flow of 120 perc ent of the normal operating flow. The overspeed condition is applicable only to operating conditions when the reactor and turbine generator are at power.

Interrelated Performance and Safety Functions The interrelated performance and safety functions of the RCS a nd its major com ponents are listed below:1.The RCS provides sufficient heat transfer capability to transfer the heat produced during power operation and when the reactor is subcritical, including the initial phase of plant cooldown, to the st eam and power conversion system.2.The system provides sufficient heat tr ansfer capability to transfer the heat produced during the subsequent phase of plant cooldown and cold shutdown to the residual heat removal system.

MPS3 UFSAR5.1-5Rev. 303.The system heat removal capability under power operation a nd normal operational transients, including the transition from fo rced to natural circulation, assures no fuel damage within the operating bounds permitted by the reactor control and protection systems.4.The RCS provides the water used as the core neutron moderator and reflector and as a solvent for chemical shim control.5.The system maintains the homogeneity of soluble neutron poison concentration and rate of change of coolant temperat ure such that u ncontrolled reactivity changes do not occur. Interlocks are provi ded on the loop stop isolation valves to prevent the addition of cold or diluted water at excessive rates.6.The reactor vessel is an in tegral part of the RCS pressure boundary and is capable of accommod ating the temperatures and pressures associated with the operational transients. The reactor vessel functions to support the reactor-core and control rod drive mechanisms.7.The pressurizer maintains the system pre ssure during operation a nd limits pressure transients. During reduction or increase of plant load, reactor coolant volume changes are accommodated in the pressurizer via the surge line.8.The reactor pumps supply th e coolant flow necessary to remove heat from the reactor core and transfer it to the steam generators.9.The steam generator tube and tubeshee t boundary are designed to prevent or control to acceptable levels th e transfer of activity generated within the core to the secondary system.10.The RCS piping serves as a boundary for containing the coolant under operating temperature and pressure conditions and fo r limiting leakage (and activity release) to the containment atmosphere. The RCS piping contains demineralized borated water which is circulated at the flow rate and temperatur e consistent with achieving the reactor core th ermal and hydraulic performance.11.The components of the RCS are surrounde d by concrete structures which provide support, radiation shieldi ng and missile protection. RCS shielding permits limited access to the containment during power opera tion. The reactor vessel is installed in a thick concrete cavity formed by the prim ary shield. The entire RCS is enclosed by the secondary shield.12.Portions of the RCS are relied upon to function in conjunction with other systems of the cold shutdown design during a safety grade cold shutdown (Section 5.4.7

). It is expected that the systems normally us ed for cold shutdown would be available anytime the operator chooses to perform a reactor cooldown. Should only safety

grade equipment be available, the RCS pr ovides safety grade letdown capability MPS3 UFSAR5.1-6Rev. 30 via the reactor vessel head vent system to the pressurizer relief tank, and safety grade depressurization capability by venting the pressurizer through the pressurizer power-operated relief valves. Refer to Section 5.4.7.2.3.5 for a detailed description of safety grade cold shutdown.The reactor vessel head letdown line and associat ed piping and valves also provide the capability to mitigate a possible condition of inadequate co re cooling or inadequate natural circulation.

5.1.1 SCHEMATIC

FLOW DIAGRAM The RCS, shown schematically on Figure 5.1-2, include s typical values for principal parameters of the system under normal steady state full power operating conditions. These values are based on the best estimate flow. RCS volume under these conditions is presented in Table 5.1-1.

5.1.2 PIPING

AND INSTRUMENTATION DIAGRAM A piping and instrumentation diagram of the RCS (Figure 5.1-

1) shows the extent of the systems located within the containment, and the points of separation be tween the RCS and the secondary (heat utilization) system.

5.1.3 ELEVATION

DRAWING Reactor coolant system compone nts are shown on Figures 3.8-59 a nd 3.8-60. These figures detail the component relationships with the surrounding concrete structures.

The concrete structures provide support, radiation shielding, and mis sile protection for the reactor coolant system components. The concrete shielding permits limited acces s to the containment during power operation.

Primary shielding for the reactor vessel is provided by the neutron shield tank and a thick concrete wall which surrounds the vessel.

All reactor coolant system comp onents are enclosed by the crane wall, which serves as a secondary shie ld within the containment structure.

MPS3 UFSAR5.1-7Rev. 30TABLE 5.1-1 SYSTEM DESIGN AND OPERATING PARAMETERSPlant Design Life (years) 60Nominal Operating Pressure (psig) 2,235Total System Volume Including Pressurizer and Surge Line (ft

3) 11,750 Pressurizer Spray Rate (m aximum gpm) 1,800Note: Total Heater Capacity may be less due to heater unavailability)Pressurizer Relief Tank Volume (ft

3) 1,800 System Thermal and Hydraulic Data 4 Pumps Running(a)(b)NSSS Power (MWt)3,6663,666 Reactor Power (MWt)3,6503,650 Thermal Design Flows (gpm)

Active Loop (c)90,80090,800Reactor363,200363,200Total Reactor Flow (10 6 lb/hr)135.3136.9Temperatures (°F)Reactor Vessel Outlet622.6615.1Reactor Vessel Inlet556.4547.9Steam Generator Outlet556.0547.6Steam Generator Steam537.4531.2 Feedwater445.3445.3Steam Pressure (psia)942894Total Steam Flow (10 6 lb/hr)16.2916.25 Best Estimate Flows (gpm)Active Loop97,30099,700Reactor389,200398,800 MPS3 UFSAR5.1-8Rev. 30a) Parameter based on full pow er operation with 10% equiva lent steam generator tube plugging and reactor vessel average temperature of 589.5

°F.(b) Parameters based on full power operation w ith 0% equivalent steam generator tube plugging and reactor vessel average temperature of 581.5

°F.(c) The 4% reduction in thermal design fl ow was addressed by a 10 CFR 50.59 safety evaluation in 1993.(d) System pressure drops are based on a best estimate flow at full power.System Pressure Drops (d):Reactor Vessel P (psi)44.445.9Steam Generator P (psi)44.039.1 Hot Leg Piping P (psi)2.12.2 Pump Suction Piping P (psi)3.03.2 Cold Leg Piping P (psi)4.04.3Pump Head (ft)303291 System Thermal and Hydraulic Data4 Pumps Running(a)(b)

MPS3 UFSAR5.1-9Rev. 30 FIGURE 5.1-1 (SHEETS 1-63) P&

IDS REACTOR COOLANT SYSTEM The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS-3 FSAR November 1987Rev. 20.2 FIGURE 5.1-2 REACTOR COOLANT SYSTEM PROCESS FLOW DIAGRAM MPS-3 FSARPage 1 of 2Rev. 16NOTES TO FIGURE 5.1-2Mode A Steady State Full Power Operation Flow (1) Location FluidPressure psig Temperature

°F GPM (2)lb/hr (3)Volume 1 R.C.2246.7 617.2111,211 37.1052 - 2 R.C.2244.1 617.2111,211 37.1052- 3 R.C.2206.0 556.8 99,692 37.1052- 4 R.C.2202.8 556.8 99,692 37.1052- 5 R.C.2297.0 557.0 99,600 37.1052- 6 R.C.2294.6 557.0 99,599 37.1048- 10-18 R.C.See Loop No. 1 Specifications 19-27 R.C.See Loop No. 1 Specifications 28-36 R.C.See Loop No. 1 Specifications 37 R.C.2297.0 556.8 1.0 0.0004- 38 R.C.2297.0 556.8 1.0 0.0004- 39 R.C.2235.3 652.7 2.0 0.0008- 40Steam 2235.3 652.7- - 720 41 R.C.2244.0 652.7- - 1080 42 R.C.2244.0 652.7 2.5 0.0008- 43 R.C.2246.2 617.2 2.5 0.0008- 44Steam 2235.3 652.7 0 0 - 45 R.C.2235.3< 652.7 0 0 Minimize 46 N 2 3.0 120.0 0 0 -

MPS-3 FSAR Page 2 of 2Rev. 16 47 R.C.2235.3< 652.7 0 0 Minimize 48 N 2 3.0 120.0 0 0 - 49 N 2 3.0 120.0 0 0 - 50 N 2 3.0 120.0- - 450 51PRT Water 3.0 120.0- - 1350NOTES TO FIGURE 5.1-2Mode A Steady State Full Power Operation Flow (1) LocationFluidPressure psig Temperature

°FGPM (2)lb/hr (3)Volume (1) Flows measured at 130

°F and 2300 psia.

(2) At the conditions specified (3) x 10 6 MPS3 UFSAR5.2-1Rev. 30

5.2 INTEGRITY

OF REACTOR COOLANT PRESSURE BOUNDARYPer Regulatory Guide 1.70, Revision 3, this section discusses the measures employed to provide and maintain the integrity of the reactor cool ant pressure boundary (RCPB) for the plant design lifetime. The RCPB, as defined in 10 CFR 50.2, ex tends to the outermost containment isolation valve in system piping which penetrates the cont ainment and is connected to the reactor coolant system (RCS) (Section 5.1). Since other sections of this FSAR al ready describe the components of these auxiliary fluid systems in detail, the discussions in this section are limited to the components of the RCS as defined in Section 5.1, unless otherwise noted.

Additional information on the RCS and the components which are part of the RCPB (as defined in 10 CFR 50) is given in the following sections:

Section 6.3 - The RCPB components which are part of emer gency core cooling system Section 9.3.4 - The RCPB components which are part of the chemical and volume control system Section 3.9N.1 - The design loadings, stress limits, and analyses applied to the RCS and American Society of Mechanical Engineers (ASME) Code Class 1

components Section 3.9N.3 - The design loadings, stress limits, and analyses applied to ASME Code Class 2 and 3 components The phrase, RCS, as used in this section is as defined in Section 5.1. When the term RCPB is used in this section, its definition is that of Section 50.2 of 10 CFR 50.

5.2.1 COMPLIANCE

WITH CODES AND CODE CASES 5.2.1.1 Compliance with 10 CFR 50.55aRCS components are designed and fabricated in accordance wi th 10 CFR 50.55a. The actual addenda of the ASME Code applied in the origin al design of each component are listed in Table 5.2-1.5.2.1.2 Applicable Code CasesRegulatory Guides 1.84 and 1.85 are discussed in Section 1.8. The following discussion addresses only unapproved or conditionally approved code cases (per Regulatory Guides 1.84 and 1.85) used on Class 1 primary components and component supports.

Code Case 1528 (SA 508 Class 2a) ma terial was used in the manufac ture of the Millstone 3 steam generators and pressurizers.

Purchase orders for this equipment were placed prior to the original issue of Regulatory Guide 1.85 (June 1974). Regulatory Guide 1.85 Revisi on 6 (May 1976) reflected conditional NRC Approval of Code Case 1528. Th e Westinghouse test program demons trates the adequacy of Code MPS3 UFSAR5.2-2Rev. 30 Case 1528 material. Its results are documented in Eicheldinger' s letter (10/4/76) and WCAP-9292. WCAP 9292 and a request for approval of the us e of Code Case 1528 were submitted to the NRC Eicheldinger's letter (3/17/78).

Code Cases N-242 (Paragraphs 5.4, 5.5 and 5.6) and N-242-1 (Paragraphs 5.3, 5.4, 5.5 and 5.6) material was used in the manufacture of the Millstone 3 mechanical shock arrestors. Code Case N-242-1 (Paragraphs 1.0 through 4.0) material wa s used in welding operations on Millstone 3 reactor plant component cooli ng check valve 3CCP*V3, S/N C61870. Code Case N-242 was also used on J. E. Lonergan Relief Valves which are listed in Table 5.2-6. Regulatory Guide 1.85, Rev.

18, allows the use of these code cases.

Code Case N-71 (1644-6) ma terial, A-500-74a Grade B, was used in the fabrication of cable tray supports attached to the CRDM Seismic Support Platform. Regulatory Guide 1.85, Rev. 20, allows the use of the code case.

Code Case N-275, Repair of Welds, was used to waive LP exam ination requirements in the repair of welds where the back side of the weld joint assembly is not accessible for removal of the examination material.

Code Case N-407 was invoked for li mited repair welds of A-487 Cla ss 10Q steel castings without post weld heat treatment. The castings are for pa rts of the steam genera tor and reactor coolant pump supports (FSAR Section 5.4.14).

Material listed in Code Case N- 249-4, specifically A-668 Class M, was used for pins in the steam genera tor and reactor coolant pump pressurizer supports.

These Code Cases have not been endorsed by the NRC in Regulatory Guides 1.84 or 1.85. A request for approval of Code Cases N-407 and N-249-4 was submitted to the NRC in Counsil's letter (6/8/84) with an attached report, 12179-J(B)-131. Code Case N-407 was approved by the NRC in Youngblood's letter (2/12/85) based on the test program results attached in Counsil's letter (6/8/84). Code Case N-249-4 was approved by the NRC in Youngblood's letter (9/24/85).ASME Code Case N-640 in conjunction with ASME Code Section XI, Appendix G has been used to develop the reactor vessel beltline P/T limits.

This Code Case permits the use of an alternate fracture toughness curve (K Ic) in lieu of the lower bound K Ia curve. Use of this Code Case was provided by the NRC as documented in letter dated January 9, 2002.

5.2.2 OVERPRESSURE

PROTECTION RCS overpressure protection is provided by the pressurizer and st eam generator safety valves along with the reactor protection system and associated equipm ent. Combinations of these systems assure compliance with the overpressure requirements of the AS ME Code,Section III, paragraphs NB-7300 and NC-7300, for pressurized water reactor systems.

The only portion of an auxiliary sy stem used for overpressure prot ection of the RCS is the liquid relief valves of the heat removal (RHR) system. These valves protect the RCS at low temperatures when the RHR system is on operation. They are located inside containment and discharge to the pressurizer relief tank.

MPS3 UFSAR5.2-3Rev. 30 5.2.2.1 Design Bases Overpressure protection is provided for the RCS by the pressuri zer safety valves. This protection is afforded for the following events which envel op those credible events which could lead to overpressure of the RCS if adequate ove rpressure protection were not provided: 1.Loss of electrical lo ad and/or turbine trip2.Uncontrolled rod withdrawal at power3.Loss of reactor coolant flow4.Loss of normal feedwater5.Loss of offsite power to the station auxiliaries The sizing of the pressurizer safety valves is based on an alysis of a complete loss of steam flow to the turbine with the reactor operating at 102 per cent of engineered safe guards design power. In this analysis, feedwater flow is also assumed to be lost, and no credit is taken for operation of pressurizer power operated relief valves, pressu rizer level control system, pressurizer spray system, rod control system, steam dump system, or steam line power operated relief valves. The reactor is maintained at full power (no credit for direct reactor trip on turbine trip), and steam relief through the steam generator safety valves is considered. The total pressurizer safety valve capacity is required to be at least as large as the maximum surge rate into the pressurizer during this transient.

This sizing procedure results in a safety valve capacity well in excess of the capacity required to prevent exceeding 110 percent of system design pressure for the events listed in this section.

Overpressure protection for the steam system is provided by steam generator safety valves. The steam system safety valve capaci ty is based on providing enough relief capacity to remove the engineered safeguards design steam flow. This must be done while li miting the maximum steam system pressure to less than 110 percent of the steam generator shell side design pressure.

Blowdown and heat dissipation syst ems of the nuclear steam supply system (NSSS) connected to the discharge of these pressure relieving devices are discussed in Section 5.4.11.Steam generator blowdown systems for the bala nce of plant are disc ussed in Section 10.4.8.

5.2.2.2 Design EvaluationA description of the pressurizer safety valves performance characteristics along with the design description of the incidents, a ssumptions made, method of analysis , and conclusions are discussed in Chapter 15.

MPS3 UFSAR5.2-4Rev. 30The relief capacities of the pressurizer and steam generator safety valves are determined from the postulated overpressure transient conditions in conjunction with the action of the reactor protection system. WCAP-7769 (Cooper et al., 1972), and Eicheldinger's letter (1975) evaluate the functional design of the overpressure protec tion system and analyze the capability of the system to perform its function for a typical plan

t. WCAP-7769 describes in detail the types and number of pressure relief devices employed, relief device description, locations in the systems, reliability history, and the details of the methods used for relief device sizing based on typical worst condition. An overpressure protection report specifically for Millstone 3 is prepared in accordance with Article NB-7300 of Section III of the ASME C ode. WCAP-7907 (Burnett, et al., 1972) describes the analytical model used in the analysis of the overpressure protection system and the basis for its validity.

5.2.2.3 Piping and Instrumentation Diagrams Overpressure protection for th e RCS is provided by the pressurizer safety valves shown on Figure 5.1-1. These valves discharge to the pressurizer relief tank through a common header.

The steam generator safety valves are discus sed in Section 10.3 and are shown on Figure 10.3-1.

5.2.2.4 Equipment and Component Description The operation, significant design parameters , number and types of operating cycles, and environmental conditions of the pressurizer safety valves are discussed in Sections 5.4.13, 3.9N.1, and 3.1 1N.A discussion of the equipment and components of the steam system overpressure protection features is included in Section 10.3.

5.2.2.5 Mounting of Pressu re-Relief DevicesThe pressurizer safety valve suppor t is designed to withstand seismi c, thermal, pipe rupture, and deadweight forces in addition to the valve dischar ge reactions. The supports consist of: 1.A circumferential box girder supported of f four vertical columns2.Radial support arms from each valve to the box girder 3.Pinned column connections at the pr essurizer safety valve support brackets The supports are welded in place.

Three safety valves are supporte

d. The three valves are assume d operating simultaneously. The discharge load from each valve, in combination w ith the seismic, thermal, pipe rupture, piping, and deadweight load, is applied to the valve s upports at the valve inlet flange. These loads are taken by the radial support arms which then transmit thrust, bending, and torsional loads into the box girder ring. These are distributed to each of the four columns and down to pin connections at MPS3 UFSAR5.2-5Rev. 30 the pressurizer safety valve support brackets. Included in this load at the bracket is the effect of seismic excitation of the support steel itself.

Section 3.9B.3.3 gives the particular loading combinations analyzed: 1.The normal condition includes: deadwe ight + 1/2 SSE + occasional (valve operation) loads2.The upset condition includes: deadweight

+ 1/2 SSE + thermal + occasional (valve operation) loads3.The faulted condition includes: deadweight + SSE + pipe rupture + occasional (valve operation) loadsTable 5.2-5 lists the loads at the pr essurizer safety valve support br ackets for each combination of design loads. Included within the parentheses are Westinghouse allowable loads for the same design load combinations.

Review of this table s hows that faulted loads ex ceed other load conditions by a factor greater than 2 except for those loads that have insignificant effect on stresses such as F x and M y. The ratio of allowable faulted stress to the allowable for normal, upset, or emergency is less than 2. Therefore, the faulted load condition is the limiting condition.

5.2.2.6 Applicable Codes and Classification The requirements of the ASME Boiler and Pr essure Vessel Code,Section III, NB-7300 (Overpressure Protection Report) and NC-7300 (O verpressure Protection Analysis), are met.

Piping, valves, and associated equipment used for overpressure protection are classified in accordance with ANSI-N18.2, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants." These safe ty class designations are delineated in Table 3.2-2 and shown on Figure 5.1-1.

5.2.2.7 Material Specifications Section 5.2.3 describes mate rial specifications.

5.2.2.8 Process Instrumentation Each pressurizer safety valve di scharge line incorporates a c ontrol board mounted temperature indicator and an alarm to notify the operator of steam discharge due to either leakage or actual valve operation. Chapter 7 discu sses process instrumentation as sociated with the system.

MPS3 UFSAR5.2-6Rev. 30 5.2.2.9 System Reliability The reliability of the pressure relieving device s is discussed in Sect ion 4 of WCAP-7769 (Cooper et al., 1972) and Eicheldi nger's letter (1975).

5.2.2.10 Testing and InspectionTesting and inspection of the ov erpressure protection componen ts are discussed in Section 5.4.13.4 and Chapter 14.5.2.2.11 RCS Pressure Control during Low Temperature Operation Administrative procedures are av ailable to aid the operator in controlling RCS pressure during low temperature operation. However, to provide a backup to the operator and to minimize the frequency of RCS overpressurizati on, an automatic system is provide d to mitigate any inadvertent excursion.

Protection against an overpressurization event is provided through the use of two PORVs, two RHR suction relief valves, or one PORV and on e RHR suction relief valve to mitigate any potential pressure transients. Analyses have shown that one relief valve is sufficient to prevent violation of these limits due to anticipated mass and heat input transients. The mitigation system is required only during low temperature operation; it is manually placed in service and automatically actuated.5.2.2.11.1 System OperationTwo pressurizer power-operated reli ef valves are each supplied with actuation logic to en sure that an automatic and independent RCS pressure control backup featur e is available for the operator during low temperature operations. This system has the capability for RCS inventory letdown, thereby maintaining RCS pressure within allowable limits. Sect ions 5.4.7, 5.4.10, 5.4.13, 7.7 and 9.3.4 give additional information on RCS pressure and inventory control during other modes of operation.

The basic function of the system logic is to c ontinuously monitor RCS te mperature and pressure conditions whenever plant operation is at low temperatures. An auctioneered system temperature is continuously converted to an al lowable pressure and then compar ed to the actual RCS pressure.

The system logic first annunciates a main contro l board alarm whenever the measured pressure approaches within a predetermined amount of th e allowable pressure, thereby indicating that a pressure transient is occurring and on a further increase in measur ed pressure, an actuation signal is transmitted to the PORVs when required to mitigate the pressure transient.

The isolation valves between the RCS and the RHR suction relief valves must be open to make the RHR suction relief valves operable for RC S overpressure mitigation. When the RHR system is operated for decay heat removal or low pressure letdown control, the isolation valves between the RCS and the RHR suction relief valves are open, and the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.

MPS3 UFSAR5.2-7Rev. 305.2.2.11.2 Evaluation of Low Temperature Overpressure TransientsPressure Transient Analyses ASME Section III, Appendix G, establishes guidelines and upper lim its for RCS pressure primarily for low temperature conditions ( 350°F). The mitigation system (Section 5.2.2.11) satisfies these conditions at temperatures 226°F, which is the enabling temperature required to protect the RCS against non-ductile failure.Transient analyses determined the maximum pressu re for the postulated ma ss input and heat input events.

The limiting mass input transient which would occur duri ng RCS low temperat ure operation is the injection of a charging pump at a run-out flow of 570 gpm with letdown isolated.

The heat input transient analysis is performed over the entire RCS shutdown temperature range.

This analysis assumes a reactor coolant pump startup with a 50

°F mismatch between the RCS and the temperature of the ho tter secondary side of the steam gene rators. Inadvertent RCP starts are not considered credible during low temperature operation since two separate operator actions are required to start an RCP. In addition, restrictions on the allowable mismatch are required to limit relief flow to values within th e capacity of the RHR relief valves.

Both heat input and mass input an alyses take into account the single failure criteria and, therefore, only one relief valve is assumed to be available for pressure relief. These events have been evaluated considering the allowa ble isothermal beltline pressure/temperature limits. The evaluation of the transient results concludes that th e vessel integrity and plant safety will not be impaired.5.2.2.11.3 Operating Basis Earthquake Evaluation A fluid systems evaluation has been performed considering th e potential for overpressure transients following an operating basis earthquake (OBE).

The Millstone 3 power-operated relief valves have been designed in accordance with the ASME code to provide the integrity required for the reactor coolant pressure boundary and qualified in accordance with the valve operability program wh ich is described in de tail in Section 3.9N.3.2.2.

Based on this evaluation, hypothesized overpressu re transients following an OBE are not a concern.5.2.2.11.4 Administrative ProceduresAlthough the system described in Section 5.2.2.11.1 is designed to maintain RCS pressure within the allowable pressure limits, administrative pr ocedures have been provided for minimizing the potential for any transient th at could actuate the overpressure relief system. The following MPS3 UFSAR5.2-8Rev. 30 discussion highlights these procedur al controls, listed in hierarc hy of their function in mitigating RCS cold overpressurization transients.

Of primary importance is the ba sic method of operation of the plant. Normal plant operating procedures maximizes the use of a pressurize r cushion (steam bubble) during periods of low pressure, low temperature operation. This cushion dampens the plant's response to potential transient generating inputs, providing easier pre ssure control with the slower response rates.An adequate cushion substantially reduces the severity of some potential pressure transients such as reactor coolant pump induced heat input and slows the rate of pressure rise for others. In conjunction with the previously discussed alarms, this provides reasonable assurance that most potential transients can be terminated by operator action before the overpressure relief system actuates.However, for those modes of ope ration when water solid operatio n may still be possible, the following procedures further high light precautions that minimize the potential for developing an overpressurization transient. The followi ng specific recommendations are made: 1.Prior to removing the RHR letdown fr om service, alternate provisions for maintaining an RCS mass inventory balance sh all be established to ensure that the cold overpressure protection syst em (COPPS) is not challenged.2.Whenever the plant is water solid and the reactor coolant pressure is being maintained by the low pressure letdown c ontrol valve, letdown flow must bypass the normal letdown orifices, and the valve in the bypass line must be in the full open position. During this mode of operation, all three let down orifices must also remain open.3.If all reactor coolant pumps have been stopped for more than 5 minutes, and the reactor coolant temperature is greater than the charging and seal injection water temperature, no attempt shall be made to start the first reactor coolant pump when the RCS is water-solid until administrative , procedural guidelines which limit the temperature difference between the RCS and the charging and seal injection water are met. This will minimize the pressure transient when the pump is started and the cold water previously injected by the charging pumps is circulated through the warmer reactor coolant components.4.If all reactor coolant pumps are stopped and the RCS is further cooled down by the residual heat exchangers, a nonuniform te mperature distribution may occur in the reactor coolant loops. For th is case, the Technical Specifications provide restrictions for starting the first reacto r coolant pump that bound the most limiting heat injection transients, thereby ensuri ng that the RCS pressure is maintained within the allowable pressure limits. No attempt shall be made to start the first reactor coolant pump when the RCS is water-solid until administ rative, procedural guidelines which limit the temperature difference between the RCS and the steam generator secondary side fluid are satisf ied. These administrative limits provide MPS3 UFSAR5.2-9Rev. 30 reasonable assurance that pot ential transients can be te rminated by operator action before the overpressure relief system actuates.5.During plant cooldown using the main c ondenser , all steam generators shall be connected to the steam header to assure a uniform cooldown of the reactor coolant loops.6.During normal cooldown, at le ast one reactor coolant pump shall be maintained in service until the reactor coolant temperature is reduced to 160

°F.These special precautions back up the normal oper ational mode of maximi zing periods of steam bubble operation so that cold overpressure transient prevention is conti nued during periods of transitional operations.

The specific plant configurations of ECCS test ing and alignment also highlight procedures required to prevent developing cold overpressurization transients. Du ring these lim ited periods of plant operation, the following actions minimize the probability of system overpressurization: 1.To preclude inadvertent ECCS actuati on during cooldown, blocking of the low pressurizer pressure and low steam line pr essure safety injection signal actuation logic occurs between the P-11 setpoint pressure of 1985 psig and the SI signal actuation pressure of 1877.3 psig. During heatup, the low pressurizer pressure and low steam line pressure safety injection signal actuation logic remains blocked until the pressure exceeds the P-11 setpoint.2.During further cooldown, closure and power lockout of the accumulator isolation valves is required at a pressure of less than or equal to 1,000 psig and no sooner than two and one-half hours following reactor shutdown, but before the RCS pressure reaches the accumulator pressure , providing additional backup to item 1.

above. Prior to placing the cold overpressu rization protection syst em in service, all but one charging pump and all SI pumps are rendered incapable of injecting into the RCS.3.Periodic ECCS pump performa nce testing requires the te sting of the pumps during normal power operation or at hot shutdown conditions. This precludes any potential for developing a cold overpressurization transient.

If cold shutdown testing with the vessel closed is necessa ry , the procedures require ECCS valve closure and RHS alignment to both isolate potential ECCS pump

input and to provide backup bene fit of the RHS relief valves.The safety injection pump can be run to fill the accumulators or for testing during cold shutdown with the vessel closed provided the safety injection pump is rendered incapable of injecting into the RCS by at least two independent means.

The following are examples of acceptable actions which meet th is requirement: 1) closing the pump discharge valve(s) to th e injection line and either removing the MPS3 UFSAR5.2-10Rev. 30 power from the valve operator(s) or locking manual valve(s) closed and 2) closing the valve(s) from the injection source and either removing the power from the valve operator(s) or locking manual valve(s) closed.4.Safety Injection signal circuitry te sting, if done during cold shutdown, also requires RHS plus SI pump alignments and non operating char ging pump power lockout to preclude inadvertent SI discharge to the RCS.These procedural recommendations covering normal operations with a steam bubble, and transitional operations where potentially wate r solid, when followed by specific testing operations, provide in-depth cold overpressure preventions or reductions, augmenting the installed overpressure relief system.

5.2.3 MATERIALS

SELECTION, FABR ICATION, AND PROCESSING 5.2.3.1 Material SpecificationsTypical material specifications used for the princi pal pressure retaining a pplications in Class 1 primary components and for Class 1 and 2 auxiliary components in systems required for reactor shutdown and for emer gency core cooling are listed in Table 5.2-2. T ypical material specifications used for the reactor internals required for emergency core cooling, for any mode of normal operation or under postulated accident conditions, a nd for core structural load bearing members are listed in Table 5.2-3.In some cases, Tables 5.2-2 and 5.2-3 may not be to tally inclusive of the ma terial specifications used in the listed applications. However, the listed specifications are representative of those materials used. All of the materials used were procured in accordance with ASME Code requirements.The welding materials used for joining the ferritic base materials of the RCPB conform to or are equivalent to ASME Material Specifications SFA 5.1, 5.2, 5.5, 5.17, 5.18, and 5.20. They are tested and qualified to the require ments of ASME C ode,Section III.The welding materials used for joining the austenitic stainless steel base materials of the RCPB conform to ASME Material Specifications SF A 5.4 and 5.9. They are tested and qualified according to the requirements of ASME Code,Section III.The welding materials used for joining nickel-chromium-iron alloy in similar base material combination and in dissimilar ferritic or austenitic base mate rial combination conform to ASME Material specifications SFA 5.11 and 5.14. They ar e tested and qualified to the requirements of ASME Code,Section III.

MPS3 UFSAR5.2-11Rev. 30 5.2.3.2 Compatibility With Reactor Coolant 5.2.3.2.1 Chemistry of Reactor Coolant The RCS chemistry specificati ons are given in Table 5.2-4.The RCS water chemistry was selected to minimize corrosion. A periodic analysis of the coolant chemical composition is performed to verify that the reactor coolant quality meets the specifications.

The chemical and volume control system provides a means fo r adding chemicals to the RCS during all power operations subsequent to startup. Table 5.2-4 gives the oxygen content and pH limits for power operations.

The pH control chemical employ ed is lithium-7 hydroxide. This chemical was chosen for its compatibility with the materials and water chemistry of borated water/stainless steel/zirconium/Inconel systems. In addition, lithium is produced in solution from the neutron irradiation of the dissolved boron in the coolant.

During reactor startup from th e cold condition, hydrazine is employed as an oxygen scavenging agent. The hydrazine solution is introduced into the RCS from the chem ical and volume control system. Dissolved hydrogen controls and scavenges oxygen produced due to radiolysis of water in the core region. Sufficient partial pr essure of hydrogen is maintained in the volume cont rol tank such that the specified equilibrium concentration of hydroge n is maintained in the reactor coolant.

5.2.3.2.2 Compatibility of Construction Ma terials with Reactor Coolant All of the ferritic low alloy and carbon steels used in principal pr essure retaining applications have corrosion resistance cladding on all surfaces that are exposed to the reactor coolant. This cladding material has a chemical analysis which is at least equivalent to the corrosion resistance of Types 304 and 316 austenitic stai nless steel alloys or nickel-chromium-iron alloy, martensitic stainless steel, and precipitation hardened stai nless steel. The cladding on ferritic type base materials receives a post weld heat tr eatment, as required by the ASME Code.

Ferritic low alloy and carbon st eel nozzles are safe ended with either stainless steel wrought materials, stainless steel weld metal analysis A-7 (designate d A-8 in the 1974 Edition of the ASME Code), or nickel-chromium-iron alloy weld metal F-Number 43. The latter buttering material requires further safe ending with austenitic stainless steel base material or stainless steel weld metal analysis A-8 after completion of the post weld heat treatment when the nozzle is larger

that a 4 inch nominal inside diameter and/or the wall thickness is greater than 0.531 inches.All of the austenitic stainless steel and nickel-chromium-iron alloy base materials with primary pressure retaining applications are used in the solution anneal heat treat condition. These heat treatments are as required by the material specifications.

MPS3 UFSAR5.2-12Rev. 30 During subsequent fabrication, these materials are not heated above 800

°F other than locally by welding operations. The solution annealed surge line material is subse quently formed by hot bending followed by a resoluti on annealing heat treatment.Components with stainless steel sensitized in the manner expected during component fabrication and installation operate satisfactorily under norma l plant chemistry conditions in pressurized water reactor systems because chlorides, fluorides , and oxygen are controlled to very low levels.

5.2.3.2.3 Compatibility with External Insulation and Environmental AtmosphereIn general, all of the materials listed in Table 5.2-2 which are used in principal pressure retaining app lications and which are subject to elevated temperature during system operation are in contact with thermal insulation that covers their outer surfaces.

The thermal insulation used on the RCPB is either reflective stainless st eel type or made of compounded materials which yield low leachable ch loride and/or fluoride concentrations. The compounded materials in the form of blocks, boards, cloths, tapes, adhesives, cements, etc, are silicated to provide protection of austenitic stainless steels against stress corrosion which may result from accidental wetting of the insulation by spillage, minor leakage, or other contamination from the environmental atmosphe re. Section 1.8 includes a discus sion which indicates the degree of conformance with Regulatory Guide 1.36, "Nonm etallic Thermal Insulation for Austenitic Stainless Steel."In the event of coolant leakage, the ferritic materials will show in creased general corrosion rates.

Where minor leakage is anticipate d from service experience, such as valve packing, pump seals, etc., only materials which are compatible with the coolant (Table 5.2-2) are used. Ferritic materials exposed to coolant leakag e can be readily observed as part of the inservice visual and/or nondestructive inspection program to assure the integrity of the com ponent for subsequent service.5.2.3.3 Fabrication and Processing of Ferritic Materials 5.2.3.3.1 Fracture Toughness The fracture toughness propertie s of the RCPB components meet the requirements of ASME Section III, Paragraphs NB-2300, NC-2300, and ND-2300 as appropriate.

Limiting steam generator a nd pressurizer RT temperat ures are guaranteed at 60

°F for the base materials and the weldments. These materials meet the 50 ft-lb absorbed energy and 35 mils lateral expansion requirements of the ASME Code,Section III at 120

°F. The actual results of these tests are provided in the AS ME material data reports which are supplied for each component and are submitted to the licensee at th e time of shipment of the component.

Calibration of temperature instru ments and of Charpy impact test machines is performed to meet the requirements of the ASME Code,Section III, Paragraph NB-2360.

MPS3 UFSAR5.2-13Rev. 30Westinghouse has conducted a test program to determine the fr acture toughness of low alloy ferritic materials with specified minimum yield strengths gr eater than 50,000 psi to demonstrate compliance with Appendix G of th e ASME Code,Section III. In this program, fracture toughness properties were determined and shown to be adequate for base metal plates and forgings, weld metal, and heat affected zone metal for higher st rength ferritic materials used for components of the reactor coolant pressure boundary. The resu lts of the program are documented in WCAP-9292 (1978), which has been submitted to the NRC for review via Eicheldinger's letter (1978).

5.2.3.3.2 Control of Welding All welding is conducted using procedures qualified according to the rules of Sections III and IX of the ASME Code. Control of welding variable s, as well as examination and testing, during procedure qualification and production welding is performed in accordance with ASME Co de requirements.

Section 1.8 includes discussions which indicate the degree of conformance of the ferritic materials components of the RCPB with Regul atory Guides 1.34, "Control of Electroslag Properties," 1.43, "Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components," 1.50, "Control of Preheat Temperature for Welding of Low-Alloy Steel," 1.66, "Nondestructive Examination of Tubular Products," and 1.71, "W elder Qualification fo r Areas of Limited Accessibility."Westinghouse practices for storag e and handling of welding electrodes and fluxes comply with ASME Code,Section III, Paragraph NB-2400.

5.2.3.3.3 Pressurized Thermal ShockIn accordance with 10 CFR 50.61, reactor pressure vessel materials have been reviewed to establish a reference temperature for pressurized thermal shock (RT PTS). This review evaluated core loading patterns and the act ual amount of copper and nickel in the vessel materials. It also compared the vessel material co mposition and properties to surv eillance capsule materials from which tests and measurements were taken. A summary of this review is as follows: 1.Copper/Nickel Content:*Plates - full chemistry results available.*Welds - full chemistry results available. 2.Core Configuration:

The maximum fluence level of 2.70 x 10 19 n/cm 2, as determined by Westinghouse, was conservatively applied to all vessel locations to determine the end-of-life RT PTS. This value is based on the results of the updated neutron fluence analysis for 54 EFPY considering 3,650 MWt rated thermal power conditions. The third MPS3 UFSAR5.2-14Rev. 30 surveillance capsule analys is is documented in WCAP-16629-NP , Revision 0, "Analysis of Capsule W from the Domini on Nuclear Connecticut Millstone Unit 3 Reactor Vessel Radiation Surveillance Pr ogram," September 2006.

This represents the most current information regarding neutron flux and associated material degradation. This analysis considered core loading patterns and past power levels to predict the peak surface fluence. WCAP-11878, "Analysis of Capsule U from the Northeast Utilities Service Company Millstone 3 Reactor Vessel Radiation Surveillance Program," June 1988 provides the evaluation of the first surveillance capsule removed. WCAP-15405, Revision 0, "Analysis of Ca psule X from the Northeast Nuclear Energy Company Millstone Unit 3 Reactor Vessel Radiation Surveillance Program," May 2000 pr ovides the evaluation of the second surveillance capsule removed.Calculated RT PTS values have been obtained using the above assumptions. Table 5.2-7 provides the results of the calculations. This table will be updated whenever changes in core loadings, surveillance measurements, or other information indicate a significant change in the RT PTS projected values, as required by 10 CFR 50.61(b)(1).

The values that were calculated do not exceed the RT PTS screening criteria of 270

°F for plates, forgings, and axial weld materials, and 300°F for circumferential weld materials.

5.2.3.4 Fabrication and Processing of Austenitic Stainless Steel Sections 5.2.3.4.1 through 5.2.3.4.5 address Regulator y Guide 1.44, "Control of the Use of Sensitized Stainless Steel," and present the methods and controls used by Westinghouse to avoid sensitization and prevent intergra nular attack of austenitic st ainless steel components. Also, Section 1.8 discusses conforma nce with Regulatory Guide 1.44.

5.2.3.4.1 Cleaning and Contamination Protection Procedures It is required that all austenitic stainless steel materials used in the fabrication, installation, and testing of nuclear steam supply components and systems are to be handled, protected, stored, and cleaned according to recognized and accepted methods which are designed to minimize contamination which could lead to stress corros ion cracking. The rules covering these controls are stipulated in the Westinghouse Electric Corporation process specifications. These process specifications are also given to the A/E and to the owner of the plant for recommended use within their scope of supply.

5.2.3.4.2 Solution Heat Treatment Requirements The austenitic stainless steels listed in Tables 5.2-2 and 5.2-3 are used in the final heat treated condition required by the respective ASME Code,Section II materials specification for the particular type or grade of alloy.

MPS3 UFSAR5.2-15Rev. 30 5.2.3.4.3 Material Inspection ProgramWestinghouse practice is that aust enitic stainless steel materials of product forms with simple shapes need not be corrosion tested provided that the solution heat treatment is followed by water quenching. Simple shapes are defined as all plates , sheets, bars, pipe and tubes, as well as forgings, fittings, and other shaped products whic h do not have inaccessibl e cavities or chambers that would preclude rapid cooling when water quenched. When testing is required, the tests are performed in accordance with ASTM-A-262, Pract ices A or E, as amended by Westinghouse Process Specification 84201 MW.

5.2.3.4.4 Prevention of Intergranular Attack of Unstabilized Austenitic Stainless SteelsUnstabilized austenitic stainle ss steels are subject to intergranular attack provided that three conditions are present simultaneously. These are: 1.An aggressive environment, e.g., an ac idic aqueous medium containing chlorides or oxygen2.A sensitized steel 3.A high temperature If any one of the three conditions described above is not present, inter granular attack will not occur. Since high temperatures cannot be avoided in all components in the NSSS, Westinghouse relies on the elimination of conditions 1 and 2 to prevent intergranular att ack on wrought stainless steel components.The water chemistry in the RCS of a Westi nghouse pressurized water reactor is rigorously controlled to prevent the intrusion of aggressive species. In particular, the maximum permissible oxygen and chloride concentrations are limite d to those in table 5.2-4. WCAP-7735 (Hazelton 1971) describes the precautions taken to prevent th e intrusion of chlorides into the system during fabrication, shipping, and storage. The use of hydrogen overpressure precludes the presence of oxygen during operation. The effectiv eness of these controls ha s been demonstrated by both laboratory tests and operating experience. The long time exposure of severely sensitized stainless in early plants to pressurized wa ter reactor coolant environments has not resulted in any sign of intergranular attack. WCAP-7735 describes th e laboratory experiment al findings and the Westinghouse operating experience. The additional years of operations since the issuing of WCAP-7735 have provided further confirmation of the earlier conclusions. Severely sensitized stainless steels do not undergo any intergranular attack in We stinghouse pressurized water reactor coolant environments.Although there never has been any evidence that pressurized water reactor coolant water attacks sensitized stainless steels, Westinghouse considers it good metallurgical practice to avoid the use of sensitized stainless steels in the NSSS components. Accordingly, measures are taken to prohibit the purchase of sensitized stainless steel s and to prevent sensitization during component fabrication. Wrought austenitic stainless steel stock used for co mponents that are part of: the MPS3 UFSAR5.2-16Rev. 30 RCPB, systems required for reactor shutdown, systems required for emergenc y core cooling, and reactor vessel internals that are relied upon to permit adequate core cooling for normal operation or under postulated accident c onditions is used in one of the following conditions: 1.Solution annealed and water quenched2.Solution annealed and cool ed through the sensitization temperature range within less than approximately 5 minutes It is generally accepted that these practices pr event sensitization. Westin ghouse has verified this by performing corrosion tests (ASTM-393) on as-received wrought material.Westinghouse recognizes that the heat affected zones of welded component must, of necessity, be heated into the sensitization temperature range, 800

°F to 1500°F. However, severe sensitization, i.e., continuous grain boundary precipitates of chromium carbide, with adjacent chromium depletion, can still be avoided by control of weld ing parameters and welding processes. The heat input (Equation 5.2-1) and associated cooling rate through the carbide pr ecipitation range are of primary importance. Westinghouse has demonstrat ed this by corrosion testing a number of weldments.

Heat input based on expression given in Arc Welding Handbook is calculated as follows: H = (E)(I)(60)

÷ S (5.2-1)where: H = joules/inch E = volts I = amperes S = travel speed (inches/minute)

Of 25 production and qualification weldments teste d, representing all major welding processes, and a variety of components, and incorporatin g base metal thicknesses from 0.10 to 4.0 inches, only portions of two were severely sensitized.

Of these, one involved a heat input of 120,000 joules and the other involved a hea vy socket weld in relatively thin walled material. In both cases, sensitization was caused primaril y by high heat inputs relative to the section thickness. However, in only the socket weld did the sensitized condi tion exist at the surface, where the material is exposed to the environment. The component has been redesigned a nd a material change has been made to eliminate this condition.Westinghouse controls the heat input in all austenitic pressure boundary weldments by: 1.Prohibiting the use of block welding2.Limiting the maximum inte rpass temperature to 350

°F MPS3 UFSAR5.2-17Rev. 303.Exercising approval rights on all welding proceduresTo summarize, Westinghouse has a f our point program designed to pr event inter granular attack of austenitic stainless steel components: 1.Control of primary water chemistr y to ensure a benign environment2.Utilization of materials in the final heat treated condition and the prohibition of subsequent heat treatments in the 800

°F to 1,500

°F temperature range3.Control of welding proce sses and procedures to avoid heat af fected zone sensitization4.Confirmation that the welding procedures used for the manufacture of components in the primary pressure boundary and of reactor internals do not result in the sensitization of heat af fected zones Both operating experience and laboratory experiments in prim ary water have conclusively demonstrated that this program is 100 percent ef fective in preventing inter granular attack in Westinghouse NSSSs using unstabilize d austenitic stainless steel.

5.2.3.4.5 Retesting Unstabilized Austenitic Stai nless Steel Exposed to Sensitization TemperaturesIt is not normal Westinghouse practice to expose unstabilized austen itic stainless steels to the sensitization range of 800

°F to 1,500

°F during fabrication into components. If, during the course of fabrication, the steel is in advertently exposed to the sens itization temperature range, 800

°F to 1,500°F , the material may be tested in accord ance with ASME-A-393 or A-262 as amended by Westinghouse Process Specification 8 4201 MW to verify that it is not susceptible to intergranular attack, except that test ing is not required for: 1.Cast metal or weld metal with a ferrite content of 5 percent or more2.Material with a carbon content of 0.03 percent or less that is subjected to temperatures in the range of 800

°F to 1,500

°F for less than 1 hour3.Material exposed to special processing provided the processing is properly controlled to develop a uniform product and provided that ad equate documentation exists of service experience an d/or test data to demonstrat e that the processing will not result in increased susceptibility to intergranular stress corrosion If it is not verified that such material is not susceptible to inte rgranular attack, the material is re-solution annealed and wa ter quenched or rejected.

MPS3 UFSAR5.2-18Rev. 30 5.2.3.4.6 Control of Welding The following paragraphs address Regulatory Gu ide 1.31, "Control of Stainless Steel Welding,"

and present the methods used, and the verification of these methods, for austenitic stainless steel welding.The welding of austenitic stainless steel is contro lled to mitigate the occurrence of microfissuring or hot cracking in the weld. Although published data and experience have not confirmed that fissuring is detrimental to the quality of the weld, it is recognized that such fissuring is undesirable in a general sense. Also, it has been well documented in the te chnical litera ture that the presence of delta ferrite is one of the mechan isms for reducing the susc eptibility of stainless steel welds to hot cracking. However, there is insufficient data to specify a minimum delta ferrite level below which the material is prone to hot cr acking. It is assumed that such a minimum lies somewhere between 0 and 3 percent delta ferrite.

The scope of these controls discussed herein encompasses welding processes used to join stainless steel parts in compone nts designed, fabricated, or stam ped in accordance with ASME Code,Section III, Class 1 and 2, and core support co mponents. Delta ferrite control is appropriate for the above welding requirements except where no filler metal is used for other reasons such control is not applicable. Th ese exceptions include electron beam welding, autogenous gas shielded tungsten arc welding, explosive weldi ng, and welding using full y austenitic welding materials.

The fabrication and installation specifications re quire welding procedure and welder qualification in accordance with the ASME Code,Section III, and include welding materials that are used for welding qualification testing and for production processing. Specifically, the "starting" welding materials are required to contain a minimum of 5 percent delta ferrite as determined by magnetic methods on an undiluted weld deposit. As an a lternative, delta ferr ite determination for consumable inserts, bare weld rods, and wire filler metal used with the gas tungsten arc welding process can be predicted from their chemical composition using the appropriate weld metal

constitution diagrams in the ASME Code, Secti on III. (The equivalent ferrite number may be substituted for percent delta ferrite.) When new welding procedure qualification tests are evaluated for these applications, including repair welding of raw materials, they are performed in accordance with the requireme nts of Section III and Sect ion IX of the ASME Code.The "starting" welding materials used for fabricat ion and installation welds of austenitic stainless steel materials and components meet the requirements of the ASME Code,Section III. The austenitic stainless steel welding material conforms to ASME weld metal analysis A-7, (designated A-8 in the 1974 Edition of the ASME Code). Bare weld filler metal, including consumable inserts, used in inert gas welding processes conform to ASME SFA-5.9, and are procured to contain not less that 5 percent delt a ferrite according to the ASME Code,Section III. Weld filler materials used in flux shielded processes conform to ASME SFA-5.4 or SFA-5.9 and are procured in a wire-flux comb ination to be capable of providing not less than 5 percent delta ferrite in the deposit according to the ASME Code,Section III. Welding ma terials are tested using the welding energy inputs to be employed in production welding.

MPS3 UFSAR5.2-19Rev. 30 Combinations of approved heat and lots of "sta rting" welding material s are used for welding processes. The welding quality assurance program includes identification and control of welding material by lots and heats as appropriate. All of the weld proce ssing is monitored according to approved inspection programs which include review of "starting" materials, qualification records, and welding parameters. Welding systems are also subject to quality assu rance audit including calibration of gages and instruments: identification of "s tarting" and complete d materials; welder and procedure qualifications; availability and use of appr oved welding and heat treating procedures; and documentary evidence of compliance with materials; we lding parameters and inspection requirements. Fabrication and instal lation welds are inspected using nondestructive examination methods according to the ASME Code,Section III rules.To assure the reliability of these controls, Westinghouse has co mpleted a delta ferrite verification program, described in WCAP-8324-A (Enriett o 1975) which has been approved as a valid approach to verify the Westinghouse hypothesis and is considered an acceptable alternative for conformance with the NRC In terim Position on Regulatory Guide 1.31. The NRC acceptance letter and topical report eval uation were received on December 30, 1974. The program results, which support the hypothesis presented in WCAP-8324-A, are summarized in WCAP-8693 (Enrietto 1976).

Section 1.8 includes discussions wh ich indicate the degree of c onformance of the austenitic stainless steel components of the RCPB with Regulatory Guides 1.34, "Control of Electroslag Properties," 1.66, "Nondestructive Examination of Tubular Products," and 1.71, "Welder Qualification for Areas of Limited Accessibility."

5.2.4 INSERVICE

INSPECTION AND TESTING OF REACTOR COOLANT PRESSURE BOUNDARY An inservice inspection program for Safety Class 1 (reactor cool ant pressure boundary) components has been developed to en sure the structural integrity of all applicable vessels, piping, valves, pumps, and appurtenances throughout th e plant service lifetim

e. The program was developed to meet the requirements of AS ME Code,Section XI, 1983 Edition, Summer 1983 Addenda, Subsections IW A, IWB, and IWF. All of the detailed ex aminations listed in the Code were performed as a preservice examination pr ior to plant startup to demonstrate access, inspection equipment and techniques, and to esta blish a baseline for inservice examinations.

Subsequent inservice insp ections will be performed as specifie d in the edition of the ASME Code,Section XI, which is in effect for the inspection period in which the inspection is being performed.

Any exceptions to the ASME Code will be documented and approved in accordance with 10 CFR 50.552.5.2.4.1 System Boundary Subject to Inspection In addition to the reactor pre ssure vessel (RPV), components and supports within the ASME Code,Section III, Class 1 boundaries are subject to the requirements of inservice inspection per ASME Code,Section XI, Subsection IWB.

MPS3 UFSAR5.2-20Rev. 30 The following components and their supports within the reactor coolant pressure boundary (defined as Class 1 per ASME Code,Section III) regularly have inservice inspections. They are shown schematically on Figure 5.1-1, Reactor Coolant System Boundary DiagramSteam Generators (primary side)

Pressurizer Reactor Coolant Pump Reactor Coolant PipingChemical and Volume Control - to isolation valves

Residual Heat Removal - to isolation valves High Pressure Safety Injection - to isolation valves Low Pressure Safety Injection - to isolation valves 5.2.4.2 AccessibilityTo meet the accessibility requirements necessary for inservice inspection to ASME XI, sufficient space is provided around each inspection area to permit access by the inspector and his equipment. Space allowance for assembly and di sassembly of tooling and equipment, such as scaffolding, lighting, and insulation, has been provided.

In establishing the physical layouts of the piping systems within the inspection boundaries as defined by the Code, the following general accessibility criteria were followed:1.The surfaces of pipe welds were held at a minimum of 6 inches from an adjacent flat surface such as a wall.2.Where the adjacent surface is curved as in the case of pipes arranged parallel to each other , the minimum clearance may have been reduced to 4 inches. In providing these clearances, allowance was made for insulation which may be on the adjacent pipes.3.Space is provided on both si des of any pipe weld such that an operator has complete access to the pipe inspection area.4.The ultrasonic examination of welds requires that, in addition to the weld, a length of pipe on each side of the weld can be completely accessible to the operator. Insulation has been designed to be removable over applicable pipe lengths.

Field experience and developmen t of new ultrasonic techniques have shown that pipe welds ground smooth and flat rather than crowne d produce satisfactory inspection results.

All piping welds in Class 1 system s were evaluated, and abrupt or sharp edges eliminated to merge smoothly with the adjacent pipe or component surface. Any grinding was restricted to the MPS3 UFSAR5.2-21Rev. 30 weld proper and not permitted to stray into the adjacent base ma terial. Care was taken not to undercut the weld or violate the mi nimum weld or pipe wall thickness.

Pipe welds prepared in the manner described a bove can be successfully examined ultrasonically by approaching the welds from both sides.Where the weld can only be examined from one side with little or no access from the opposite side as with pipe to valve welds or pipe to fitting welds, grindi ng the weld flush permits necessary transducer contact over the weld area to fully comply with the Code requirements.To assist in the provision of adequate inspection access, the follo wing information was considered in establishing the plant layout:

1.Reactor vessel - The reactor vessel closure head is examined at the head laydown area. The closure studs receive both a volumetric and a surface examination.

A special tool is available specifically for examining the reactor vessel from the inside when full of water. This vessel in spection tool performs remote examination of all the required inspection areas in the vessel apart from the bottom head disc to ring weld and incore instrumentation penetration nozzles. Lower head welds require manual examination from the outs ide of the vessel. The building design allows for free access to the bottom of the reactor cavity for easy passage of personnel and equipment. The inclusion of a bottom head disc to ring weld is a feature of all reactor vessels and the inherent inability to examine this internally is likely to be a limitation of all reactor vessel inspection tool designs. Reactor nozzle safe end welds are required to have volumetric, visual, and surface examinations.

Access to these locatio ns has been provided.

Figure 5.2-1 shows the reactor vessel inspection areas.

2.Steam generator - Requirements for this vessel, on the primary side , are volumetric examination of the channel head to t ube sheet weld, visu al examinations on pressure retaining bolting, and surface a nd volumetric examination of nozzle to safe end welds and volumetric examination of the nozzl e inner radius sections. Adequate clearance for personnel access was provided in these areas between any adjacent missile shieldi ng or support structures.

Figure 5.2-2 shows the steam generator inspection areas.

3.Main coolant pump - The examination requirements for pumps include visual and volumetric inspections of integrally we lded supports and pressure retaining bolting. Additionally, relief has been granted to allow the internal surface of a disassembled pump to be visually examin ed during maintenance activities. Access required for disassembly to permit these inspections is provided in maintenance

considerations.

MPS3 UFSAR5.2-22Rev. 30 4.Pressurizer - The pressurizer is constructed of several courses, each of which is formed in cylinders and containing one longitudinal weld. All the vessel circumferential a nd longitudinal welds together with all the instrumentation, surge, spray, and relief nozzle welds were 10 0 percent volumetrically inspected during the preservice inspection. Subsequent inserv ice inspections require that 1 foot of

each longitudinal shell weld that intersect s the circumferential shell-to-head welds and 100 percent of each circumferentia l shell-to-head weld are inspected. Additionally, the spray, surge, relief, and safety nozzle inner radius sections are volumetrically examined.

Figure 5.2-3 shows the pressurizer inspection areas.

5.Valves - Class 1 valves, unless exempted by the exclusion criteria of the Code, require volumetric examination of pressure retaining welds and pressure retaining bolting 2 inches and larger in diameter.

Also included are visual examination of internal pressure boundary su rfaces on selected valves ex ceeding 4 inches nominal pipe size, pressure retaini ng bolting smaller than 2 inches in diameter, and surface or volumetric examinations, as applic able, on integrally welded support attachments. Adequate space is provided for personnel and equipment access to perform required inspections.

6.Piping - Piping, safe end, and branch connection welds 4 inches and greater require both volumetric and surface examinations, while those welds less than 4 inches require surface examinations only. Pressure retaining bolting exceeding 2 inches in diameter requires a volumetric examination, whereas bolting 2 inches and less requires visual examinations. Integral attachment welds for vessels, piping, pumps, and valves require a surf ace or volumetric examination, as applicable. Support components outside the IWB boundary require visual examinations in accordance with subsection IWF. Sufficient space has been provided for personnel and equipment access.

5.2.4.3 Examination Techniques and ProceduresVisual examinations are conducted in accord ance with the guidelines of Paragraph IW A-2210,Section XI, ASME Code.

Surface examinations are conducted in accordance with the guidelines of Paragraph IWA-2220,Section XI, ASME Code.

Volumetric examinations are conducted in a ccordance with the guide lines of Paragraph IWA-2230,Section XI, ASME Code.

Remote ultrasonic scanning equipment is used at Millstone for the reactor vessel nozzle, flange, and shell weld examinations for both the preo perational baseline and the later inservice inspections. The remote scanning equipment is supported from a fixture which is positioned on the reactor vessel internal's support flange. Each time the fixture is placed on the support flange, it MPS3 UFSAR5.2-23Rev. 30 is located in the same orientation as the previous inspection, thus giving an accurate reference for each inspection.

The fixture acts as the main support and positioning mechanism for the various inspection attachments (i.e., nozzle scanner, flange scanner, and vessel-s hell scanner). The various scanners have multiple transducers to accommodate varying vessel geometrics and weld configurations. Appropriate drives provide the required movements of the transducers. The scanners can be indexed to assure accurate repr oducibility for later inspections.

Manual inspection techniques ar e used on the steam generators, pressurizer, and piping.

5.2.4.4 Inspection IntervalsAs defined in subarticle IWA-2400 and IWA-2420 (Inspection Program B) of ASME Code,Section XI, the inspection interval is 10 years. Th e interval may be extende d by as much as 1 year to permit inspections to be concurrent with plant outages.

The inspection schedule is in accordance wi th IWB-2420. It is intended that inservice examinations be performed during normal plan t outages, such as refueling shutdowns or maintenance shutdowns occurring during the inspection interval.

5.2.4.5 Examination Categories and RequirementsThe extent of examinations performed is in accordance with ASME C ode,Section XI, Table IWB-2500-1.

In addition, preservice inspections complied with IWB-2200.

5.2.4.6 Evaluation of Examination ResultsEvaluation of examination results is conducted in accordance with IWB-3000, with flaw evaluation in accordance with Table IWB-3410-1. Criteria for determining the need for repair is in accordance with IWB-3000; necessa ry repairs comply with IWB-4000.

5.2.4.7 System Leakage and Hydrostatic Pressure Tests System leakage and hydrostati c tests are conducted in accor dance with IW A-5000 and IWB-5000.

5.2.4.8 Relief Requests The Class 1 portion of the PSI program was deve loped using the criteria of the ASME Code,Section XI, 1980 Edition, Winter 1980 Addenda alo ng with existing construction drawings as they were issued. An ISI program was finalized using the criteria of the ASME Code,Section XI, 1983 Edition, Summer 1983 Addenda known to be appl icable and submitted to the NRC pursuant to 10 CFR Part 50. At that time relief requests were identified.

MPS3 UFSAR5.2-24Rev. 30

5.2.5 DETECTION

OF LEAKAGE THROUGH REACTOR COOLANT PRESSURE BOUNDARY Methods are provided for detect ion of leakage through the react or coolant pressure boundary (RCPB). These methods meet the requirements of General Design Crit erion 30 (Section 3.1.2) and the guidelines of Regulatory Guide 1.45 (Section 1.8).

5.2.5.1 Identified Leakage 5.2.5.1.1 Definition of Identified LeakageIdentified Leakage is comprised of: 1.Leakage (except Controlled Leakage) into closed systems, such as pump seal or valve packing leaks, that is collected and diverted to a collecting tank2.Leakage into the containment atmosphere from sources that are specifically located and are known not to interfere with the operation of the leakage detection systems or are known not to be reactor coolant pressure boundary leakage3.Reactor coolant system leakage thr ough a steam generator to the secondary coolant system 5.2.5.1.2 Collection and Monitoring of Identified Leakage1.Valve stem leakageValve stem leakoffs for the following valv es are piped to the valve stem leakof f header in the reactor plant gaseous drains system (Section 9.3.3

): reactor coolant system loop isolation valves and loop bypass valves, and the pressurizer spray line isolation valves. The leakof f header drains to the containment drains transfer tank. Excessive stem leakage results in an increase in the rate of drainage collection in this tank. Tank level is monitored and alar med in the control room. Inspection of flow glasses, located at several points in the common drain header, permits the source of leakage to be na rrowed to a smaller group of valves. Determination of the leaking valve(s) is made by sequentia lly changing individual valve positions and observing changes in leakage rate.2.Leakage from pressurizer safety va lves or power operated relief valves Leakage is indicated by high temperature or mass flow in a safety valve dischar ge line or by high temperature in the combined discharge line from the power operated relief valves. High temperature or flow actuates an alarm in the control room. These valves discharge to the pressurizer relief tank (Section 9.3.3

). Level indication and high level alarm are provided in the control room.

MPS3 UFSAR5.2-25Rev. 303.Reactor vessel flange leakageTemperature in the leakoff line from the reactor vessel flange O-ring seal leakage monitor connection is indicated and annunciated in the control room.An increase in temperature of the leakoff line above ambient is an indication of O-ring seal leakage. High temp erature actuates an alarm in the Control Room. This leakage is collected in the containment drains transfer tank.4.Post Accident Sampling System Flow Flow from the Post Accident Sampling System may be di rected to the containment drains sump during system line pu r ging, sample acquisition, and flushing.5.All leakage (liquid or vapor) into th e containment atmosphere, which is not collected in the containment drains sump, is collected in the unidentified leakage sump. Some of this leakage is identified leakage from sources that are specifically located and are known not to interfere with the operation of the leakage detection systems or are known not to be reactor c oolant pressure boundary leakage. This identified leakage, from eith er the reactor coolant or a uxiliary systems, is normally monitored as unidentified leakage, along with the rest of the leakage to the unidentified leakage sump. However, to improve the effectiveness of the unidentified leakage sump level monitoring system alarm (the alarm alerts operators to the possibility of RCPB leakag e), the alarm set point may be adjusted to account for identified leakage.

5.2.5.1.3 Controlled Leakage1.Controlled leakage consists of seal wa ter flow supplied to the reactor coolant pump seals.2.Reactor coolant pump shaft seal leakag e Leakage may be identified by one, or a co mbination, of the following indications and/or alarms:

MPS3 UFSAR5.2-26Rev. 30a.High flow rate of CVC seal return (CBO (controlled bleed of f)): indication and alarm in control room. b.High temperature of CVC seal re turn (CBO): indication and alarm provided by the computer in control room. c.High CVC seal return (CBO) temperatur e u pstream seal water filter: local indicator. d.Increasing level in the containment drains transfer tank: indication and alarm in control room.

5.2.5.2 Unidentified Leakage 5.2.5.2.1 Definition of Unidentified Leakage Unidentified leakage is all leakage which is not identified leakage.

5.2.5.2.2 Collection of Unidentified Leakage All reactor coolant leakage in the containment structure, which is not collected in the containment drains transfer tank, in th e pressurizer relief tank, or in the cont ainment drains sump is collected in the unidentified leakage sump (Sect ion 9.3.3). A drain trench in th e containment floor is provided for this purpose.

5.2.5.2.3 Detection of Unidentified Leakage The following methods are used to detect unidentified leakage: 1.Containment (unidentifie d or drains) sump level or sump pump run time monitoring2.Containment airborne partic ulate radioactiv ity monitoring3.Containment airborne gaseous radioactivity monitoring4.Containment pressure, temperature, and humidity monitoring (backup method)5.Operator actions: a.Check makeup rate to the reactor coolant system for abnormal increase.

Instrumentation is provided to meas ure the amount of reactor coolant diverted to the boron recovery system. Taking diverted letdown flow into consideration, net level changes in th e pressurizer and volume control tank are all means for identifying system leakage.

MPS3 UFSAR5.2-27Rev. 30b.Review logs for maintenance actions which may have resulted in dischar ge of water into the containment structure.

5.2.5.2.4 Leakage Detection Method Sensitivity and Response Times Sensitivity and response times for leakage detection methods 1 through 4, Subsection 5.2.5.2.3, are as follows: 1.Unidentified leakage sump leve l and sump pump instrumentation Sump level change and sump pump run time are utilized to determine the rate of flow of unidentified leakage into the su mp. This detection method is capable of detecting a 1 gpm change in the leakage rate into the sump within one hour

.2.Containment airborne particulate and gaseous radioactivity monitors These monitors respond to the increase in airborne radioact ivity resulting from RCPB leakage, provided there is limite d ambient airborne concentration from previous leakage into the containment. The actual time required to detect reactor coolant leakage depends upon the rate and location of leakage, reactor coolant gaseous activity level, and the containment ambient background activity. A 1 gpm RCPB leak can be detected in less than one hour with the particulate monitoring system and the gaseous monitoring syst em provided that the reactor coolant activity is sufficiently high and the containment activity is below a level that would mask the change in activity corresponding to this leak rate. To ensure adequate response to a coolant leak with lower coolant and higher containment activity, the monitor setpoints are set as low as possible without ca using an excessive number of spurious alarms.3.Containment pressure, temperature, and humidityRCPB leakage causes an increase in containment pressure, temperature, and humidity. Humidity, temperature or pr essure monitoring of the containment atmosphere are considered as alarms or indirect indication of leakage to the containment.

5.2.5.2.5 Leakage Detection Method Indicators and Alarms The following indicators and/or al arms are provided in the Control Room as a means for alerting the operator to RCPB leakage: 1.Unidentified leakage sump and sump pump Unidentified leakage sump pump operation for greater than a preset time period results in an alarm in the control room.

The plant computer monitors unidentified leakage sump level and sump pump runni ng time (this information is also MPS3 UFSAR5.2-28Rev. 30 available at the liquid waste panel in the waste disposal building). The plant computer normally provides an alarm to alert operators if leakage to the unidentified leakage sump exceeds 1 gpm in any given hour. The alarm set point may be adjusted (not to exceed 2 gpm) to account for identified leakage from

reactor coolant or auxiliary systems which goes to the unidentified leakage sump. Additionally, the level instrumentation in the "Containment Drains Sump" (Sump #3) can be monitored if the unidentified leakage sump system is determined to be inoperable. Procedures are provided to th e operator for the determination of this leakage rate should the plant computer be unavailable.2.Containment airborne particul ate and gaseous radioactivity Indicators and alarms are pr ovided in the control room.3.Containment pressure, temperature, and humidityIndication and alarm are provided for pressure. Indication is provided for temperature and humidity

.5.2.5.2.6 Seismic Capability of Leakage Detection Methods The containment airborne particulate and gaseous radioactivity monitors are qualified to remain functional when subjected to the Safe Shutdown Earthquake (SSE).

5.2.5.2.7 Testing and Calibration All equipment and instrumentati on used for RCPB leak detection are in continuous operation. The provisions for testing and calibrati on of each method are described in the specific section for that system, as follows:

5.2.5.3 Intersystem Leakage Potential intersystem leakage path s with associated instrumentat ion and monitoring methods used to detect s uch leakage are as follows: 1.Secondary side of steam generators One or a combination of the following methods are used to identify steam generator tube and tube sheet leaks:

Method Section Unidentified Leakage Sump

9.3.3 Containment

Radiat ion Monitoring 12.3.4Containment Pressure, Temperature and Humidity7.5 MPS3 UFSAR5.2-29Rev. 30a.High activity as monitored and alarme d in the condenser air ejector vent line.b.Steam generator secondary side radi oactivity , as determined by sampling (Section 9.3.2

)c.Radioactivity, boric acid, or conducti vity in condensate, or blowdown e.g., from main steam line drain traps, as indicated by laboratory analysis.2.Secondary side of reactor coolant pump thermal barrier Rupture of the thermal barrie r results in an increase in flow in the reactor plant component cooling water system retu rn line from the thermal barrier (Section 9.2.2.1). At a predetermined setpoint of in creasing flow, an air-operated valve in the return line closes; this, in conjuncti on with a check valve in the supply line, isolates the thermal barrier. The position of the air-operated valve is monitored in the control room. Additionally, the tw o main headers in the reactor plant component cooling water system are continuously monitored for radioactivity.3.Low Pressure System Accumulators Leakage of reactor coolant pa st the check valves in the accumulator discharge line results in an increased level in the accumulator. High level is alarmed in the control room.4.Secondary side of letdown heat excha nger , excess letdown heat exchanger, RHR heat exchanger, and reactor coolant pump seal water heat exchanger These heat exchangers are cooled by th e reactor plant component cooling water system. Leakage into this system would be detected by the radiation monitors in the reactor plant component cooling water system.5.Safety injection systems (high and low pressure)

Potential leakage paths that exist in the ECCS are the accumulator c heck valve bypass leakage to the RCS and piping and mechanical equipment leakage outside the containment. Accumulator leakage is detected by leve l and pressure inst rumentation provided for each accumulator. This instrumentati on is continuously monitored during plant operation. Flow from each accumulator can be directed at any time through a test line to determine check valve leakage.

With respect to piping and mechanical equipment outside the containment, considering the provisions for visual insp ection (if access is available) and leak detection, leaks are detected before they propagate to major proportions. A review MPS3 UFSAR5.2-30Rev. 30 of the equipment in these systems indicate s that the largest sudden leak potential would be the sudden failur e of a pump shaft seal. Ev aluation of leakage rate, showed that flows of less than 50 gpm wo uld result. Piping leaks, valve packing leaks, or flange gasket leaks are consider ed less severe than the pump seal failure.Based on this review, the auxiliary and e ngineered safety features buildings and related equipment are designed to be ca pable of handling leaks up to a maximum of 50 gpm. Means are also provided to de tect and isolate s uch leaks in the emergency core cooling flow path with in approximately 30 minutes in the ESF Building and within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for leaks in the Auxiliary Building (Sections 6.3 , 7.3).Larger leaks in the ECCS ar e prevented by the following: a.The piping is classified ANS Safety Class 2 and, therefore, must comply with the corresponding quality assuranc e program associated with this safety class.b.The piping, equipment, and supports are designed to ANS Safety Class 2 seismic classification permitting no lo ss of function resulting from the design basis earthquake.c.The system piping is located within a controlled area on the plant site.d.The piping system receives periodic pressure tests and is accessible for periodic visual inspection.e.The piping is austenitic stainless stee l which is not susceptible to brittle fracture during operating conditions.6.Residual heat removal system (inlet and discharge)Each suction and discharge line in the RH RS is equipped with a pressure relief valve. Each suction side relief valve is sized to relieve the flow of one char ging pump at the relief valve set pressure. The discharge side relief valves relieve the maximum possible back leakage through the valves separating the RHRS from the RCS. Their relief flow capacity is 20 gpm at a set pressure of 600 psig (Section 5.4.7).The fluid discharged by the su ction side relief valves is collected in the pressurizer relief tank. The fluid discharged by the discharge side relief valve is collected in the recycle holdup tank of th e boron recovery system (Section 9.3.5

).

MPS3 UFSAR5.2-31Rev. 30 5.2.5.4 Technical SpecificationsRefer to Millstone Unit 3 for Technical Specifi cations for app licable RCPB leakage detection methods.5.2.5.5 Primary Coolant Sources Outside Containment Subsection 50.55a of 10 CFR 50 describes the codes and standards which must be implemented in the design, construction, testing and inservice insp ection of fluid systems subject to the ASME Boiler and Pressure Vessel Code. Preservice and inservice inspection program and leakage acceptance criteria is based in part on the appl icable section of the ASME Code,Section XI.

Appendix J to 10 CFR 50 addresses leak rate te sting which must be performed not only on the containment structure but also on the systems which penetrate th e containment barrier and are open to containment subsequent to an accident 10 CFR 50, Appe ndix J. III A.I(d). Appendix J requires these leak tests be perf ormed periodically throughout the life of the plant and that the results be reported to the NRC.

Millstone 3 has a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids in a pos t-accident situation.

The program includes the following: 1.System design and construc tion were reviewed to ensure that the potential for inadvertent releases of radio active fluids is eliminated.2.The implementation of all practical leak reduction measures for all systems that could carry radioactive fluid outside containment3.The measurement of actual leak rates 4.A leak reduction program of preventive maintenance to reduce leakage to as-low-as-practical levels. Pressure testing at system operatin g pressure and integrated leak tests at intervals not to exceed each refueling cycl e are typical demonstrations of system integrity.Since the letdown and charging system are used in the determination of reactor coolant system leakage (inventory balance) the integr ity of these systems is maintained.

Surveillance of the leak tightness of other system s which routinely contain radioactive fluids or gases is assured by routine surveillance of the auxiliary and waste disposal buildings and airborne radiation monitors in these buildings. The leak tightness of these systems is determined by the objectives of keeping occupational and routine releases as low as reasonably achievable.Some plant systems are excluded because the containment isolation system s prevent significant releases to these systems and the design of the pl ant does not require operation of these systems to mitigate an accident.

MPS3 UFSAR5.2-32Rev. 30Since the containment building always has the largest inventor y of radioactive materials, increased surveillance on a component containi ng a small fraction of th e containment building inventory cannot reduce the risk of a release significantly. Therefor e, upgrading the leak testing of the components described above, beyond the requi rements of Appendix J and the inservice inspection required by Section XI of the ASME Code is not contemplated.

Systems outside containment which are maintained under this program incl ude the recirculation spray, safety injection, charging portion of chemical and volume control, and hydrogen recombiners, in accordance with Millstone 3 Technical Specification 6.8.4a.

5.

2.6 REFERENCES

FOR SECTION 5.25.2-1Eicheldinger C., "Fracture Toughness Prope rties of SA533 Class 2 and SA508 Class 2a Steels." Letter NS-CE-1228 (10/4/76) to J. F. Stolz of NRC, Office of Nuclear Reactor Regulation, Westinghouse Nuclear Safety Dept., Westinghouse Corp., Pittsburgh, Penn.5.2-2Eicheldinger C., Transmittal Letter fo r Westinghouse Topical Re port WCAP-9292, Letter NS-CE-1730 (3/17/78) to J. F. Stolz of NRC Office of Nuclea r Reactor Regulation, Westinghouse Nuclear Safety Dept., Westinghouse Corp., Pittsburgh, Penn.5.2-3WCAP-7477-L (Proprietary), March 1970, Golik, M.A. and WCAP-7735 (Non-proprietary), August, 1971, Hazelton, W

.S. "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," Westinghouse Corp., Pittsburgh, Penn.5.2-4WCAP-7769, Rev. 1, June 1972, Cooper, K., et al., "Overpressure Protection for Westinghouse Pressurized Water Reactors," Westinghouse Corp., Pittsburgh, Penn.5.2-5Eichelding, C., Transmittal of additional data requested by NRC for review of WCAP-7769, Rev. 1, Letter NS-CE-622 (4/16/75) to D. B. Vassallo of NRC, Directorate of Licensing, Westinghouse Nuclear Safety Dept., Westinghouse Corp., Pittsburgh, Penn.5.2-6WCAP-7907, October 1972, Burnett, T.W.T., et al., "LOFTRAN Code Description,"

Westinghouse Corp., Pittsburgh, Penn.5.2-7WCAP-8324-A, June 1975, Enriett o, J. F., "Control of Delta Ferrite in Austenitic Stainless Steel Weldments," Westinghouse Corp., Pittsburgh, Penn.5.2-8WCAP-8693, January 1976, Enrietto, J. F., "D elta Ferrite in Production Austenitic Stainless Steel Weldments," Westinghouse Corp., Pittsburgh, Penn.5.2-9WCAP-9292, March 1978, Logsdon, W.A., et al., "Dynamic Fracture Toughness of ASME3 SA508 Class 2a and ASME SA53 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals," Westinghouse Corp., Pittsburgh, Penn.

MPS3 UFSAR5.2-33Rev. 305.2-10WCAP-11878, "Analysis of Capsule U from the Northeast Utilities Service Company Millstone Unit 3 Reactor Vessel Ra diation Surveillance Program."5.2-11Counsil, W.G., "Millstone Nuclear Station, Unit No. 3 Request for Acceptance of a New Code Case and a Revised Code Case," Le tter B11216 (6/8/84) to B. J. Youngblood of NRC Division of Licensing, Nuclear Regulatory Commission, Washington, D.C., Northeast Utilities Energy Company, (With attached Report 12179-J(B)-131, 1983, Banic, M., et al, "The Effect of Carbon Content on the Need to Postweld Heat Treated ASTM A 487 Class 10Q Material," Stone and Webs ter Engineering Corporation, Boston, MA.)5.2-12Youngblood, B.J., "Use of AS ME Code Case N-407 for Millstone Nuclear Power Station, Unit 3," Letter dated 2/12/85 for Docket No. 50-423 to W. G. Counsil of Northeast Nuclear Energy Company, Nuclear Regulatory Commission, Washington, D.C.5.2-13Youngblood, B.J., "Use of ASME Code Ca se N-249-4 for Millstone Nuclear Power Station, Unit 3," Letter dated 9/24/85 for Docket No. 50-423 to J.F. Opeka of Northeast Nuclear Energy Company, Nuclear Regulatory Commission, Washington, D.C.5.2-14The Procedure Handbook of Arc Welding, 12th Edition, Lincoln Electric Company , June 1973.5.2-15WCAP-15405, Revision 0, May 2002, "Analysis of Capsule X from the Northeast Nuclear Ener gy Company Millstone Unit 3 Reactor Vessel Radiation Surveillance Program."5.2-16WCAP-16629-NP, Revision 0, September 2006. "Analysis of Capsule W from the Dominion Nuclear Con necticut Millstone Unit 3 Reactor Vessel Radiation Surveillance Program."

MPS3 UFSAR5.2-34Rev. 30TABLE 5.2-1 APPLICABLE CODE ADDENDA FOR CLASS 1 REACTOR COOLANT SYSTEM COMPONENTS Reactor vessel ASME III, 1971 Edition through Summer 73 CRDM head adapterASME III, 1971 Edition through Summer 73 HJTC Pressure Boundary ASME II I, 1974 Edition through Summer 74Steam generatorASME III, 1971 Edition through Summer 73 Core Exit Thermocouple Nozzle AssemblyASME III, 1980 Edition through Winter 80 PressurizerASME III, 1971 Edition through Summer 73 CRDM housing Full lengthASME III, 1974 Edition through Summer 74 Reactor coolant pump ASME III, 1974 Edition through Summer 74 Reactor coolant pipe ASME III, 1971 Edition through Summer 73Surge line ASME III, 1971 Edition through Summer 73 NSSS valves Pressurizer safetyASME III, 1971 Edition through Winter 72Power-operated relief ASME III, 1977 Edition through Summer 79 Pressurizer sprayASME III, 1971 Edition through Summer 73 ControlASME III, 1971 Edition through Winter 1972 addenda to 1977 Edition through Summer 1979 AddendaMotor-operated Loop isolationASME III, 1971 Edition through Winter 73 Loop bypass ASME III, 1971 Edition through Summer 72 Head vent isolationASME III, 1977 Edition through Summer 79BOP valves in in terconnecting linesDresser forged stainless steel 2 inchesASME III, 1974 EditionVelan forged stainless steel 2 inches ASME III, 1977 Edition through Summer 79Cast stainless steel 2 1/2 inchesAS ME III, 1971 Edition through Summer 73Forged stainless steel 2 1/2 inches ASME III, 1971 Edition through Summer 73 Control valves ASME III, 1971 Edition through Summer 73 Interconnecting piping ASME III, 1971 Edition through Summer 73 MPS3 UFSAR5.2-35Rev. 30TABLE 5.2-2 PRIMARY AND AUXILIARY COMPONEN TS MATERIAL SPECIFICATIONSReactor Vessel Components Shell and head plates (other than core region)SA-533, Gr. A, B, or C, Class 1 (vacuum treated)Shell plates (core region)SA-533, Gr. A or B, Class 1 (vacuum treated)Shell, flange, and nozzle forgingsSA-508, Class 2 or 3Nozzle safe endsSA-182, Type F304 or F316CRDM head adaptor and upper headSB-166 or 167 and SA-182, Grade F304 F304L, or F316 Heated Junction Thermocouple SystemSA-479, 213, 479, Type 304; SA 182, F3 Instrumentation tube appurtenances, lower headSB-166 or 167 and SA-182, Type F304, F304L, or

F316 Closure studs, nuts, washers, inserts, and adaptorsSA-540, Class 3 Gr. B24Core support padsSB-166 with carbon less than 0.10%Monitor tubes and vent pipeSA-312 or 376, Type 304, 316, SB-166 or SB-167 or SA-182 Type 316Vessel supports, seal ledge and head lifting lugsSA-516, Gr. 70, quenched and tempered or SA-533, Gr. A, B, C, Class 1 or 2 (vessel supports may be of

weld metal buildup of equivalent strength)Cladding and butteringStainless steel weld metal analysis A-7 and Ni-Cr-Fe weld metal F-Number 43 Steam Generator Components Pressure platesSA-533, Gr. A,B, or C, Class 1 or 2Pressure forgings (including nozzles and tubesheet)

SA-508, Class 1,2,2a, or 3Nozzle safe endsStainless steel weld metal analysis A7Channel headsSA-533, Gr. A,B, or C, Class 1 or 2 or SA-216, Gr. WCCTubesSB-163, Ni-Cr-Fe annealedCladding and butteringStainless steel weld metal analysis A-7 and Ni-Cr-Fe weld metal F-Number 43Closure boltingSA-193, Gr. B7 MPS3 UFSAR5.2-36Rev. 30 Pressurizer Components Pressure platesSA-533, Gr. A,B, or C, Class 1 or 2Pressure forgingsSA-508, Class 2 or 2aNozzle safe endsSA-182, Type 316 or 316L and Ni-Cr-Fe weld metal F-Number 43Cladding and butteringStainless steel we ld metal analysis A-7 or A-8 for Code dates later than 1974 and Ni-Cr-Fe weld metal F-Number 43Closure boltingSA-193, Gr. B7Reactor Coolant Pump Pressure forgingsSA-182, Type F304, F316, F347, or F348Pressure castingSA-351, Gr. CF8, CF8A, or CF8MTube and pipeSA-213, 376, or 312, seamless Type 304 or 316Pressure platesSA-240, Type 304 or 316Bar materialSA-479, Type 304 or 316Closure boltingSA-193, 540, or 453, Gr. 660, SB-637 Gr. NO771BFlywheelSA-533, Gr. B, Class 1 Piping Reactor coolant loop pipeSA-351, Gr. CF8A centrifugal casting Reactor coolant fittings, branch nozzlesSA-351, Gr. CF8A static casting, and SA-182, Code Case 1423-2, Gr. 316NSurge lineSA-376, Gr. TP304Loop bypassSA-376, Gr. TP304 Auxiliary pipingSA-312 and SA-376 Grades TP304 and TP316 to ANSI B36.10 or B36.19Socket weld fittingsANSI B16.11 Butt weld fittingsANSI B16.9Piping flangesANSI B16.5 Full Length CRDM Latch housingSA-182 Grade 304, SA-336 Class F8, or SA-351, Gr. CF8TABLE 5.2-2 PRIMARY AND AUXILIARY COMPONEN TS MATERIAL SPECIFICATIONS MPS3 UFSAR5.2-37Rev. 30Rod travel housingSA-182, Gr. F304 or SA-336, Gr. F8CapSA-479, Type 304Welding materialsAnalysis A-8, Type 308, or 308L Valves BodiesSA-182, Type F316 or SA-351, Gr. CF8 or CF8M BonnetsSA-182, Type F316 or SA-351, Gr. CF8 or CF8M or SA-479 Type 316 DiscsSA-182, Type F316 or SA-564, Gr. 630, or SA-351 Gr. CF8 or CF8M or SA-479 Type 316StemsSA-182, Type F316 or SA-564, Gr. 630 Pressure retaining boltingSA-453, Gr. 660 Pressure retaining nutsSA-453, Gr. 660 or SA-194, Gr. 6 Auxiliary Heat Exchangers HeadsSA-240, Type 304 Nozzle necksSA-182, Gr. F304; SA-240 and SA-312, Type 304TubesSA-213 and SA-249, Type 304TubesheetsSA-182, Gr. F304; SA-240, Type 304 and 515, Gr. 70 with Type 304 SS weld overlay ShellsSA-240 and 312, Type 304 Auxiliary Pressure Vessels, Tanks, Filters, etc.Shells and headsSA-240, Type 304 and Type 316; SA-351 Gr. CF8M or SA-264 consisting of SA-537, Gr. C11 with stainless steel weld meta l analysis A-8 cladding Flanges and nozzlesSA-182, Gr. F304 and SA-105 or 350, Gr. LF2 and LF3 with stainless steel weld metal analysis A-8

cladding PipingSA-312 and 240, Type 304 or 316 seamless Pipe fittingsSA-403, Type 304 seamless Closure bolting and nutsSA-193, Gr. B7 and SA-194, Gr. 2H Auxiliary Pumps Pump casing and headsSA-351, Gr. CF8 or CF8M and SA-182, Gr. F304 or F316TABLE 5.2-2 PRIMARY AND AUXILIARY COMPONEN TS MATERIAL SPECIFICATIONS MPS3 UFSAR5.2-38Rev. 30Flanges and nozzlesSA-182, Gr. F304 or F316 and SA-403, Gr. WP316L seamlessPipingSA-312, Type 304 or 316 seamlessStuffing or packing box coverSA-351, Gr. CF8 or CF8M and SA-240, Type 304 or 316Pipe fittingsSA-403, Gr. WP316L seamless Closure bolting and nutsSA-193, Gr. B6, B7, or B8M and SA-194, Gr. 2H or 8M, SA-193, Gr. B6, B7, or B8M, SA-453, Gr. 660, and nuts, SA-194, Gr. 2H, 8M, and 6TABLE 5.2-2 PRIMARY AND AUXILIARY COMPONEN TS MATERIAL SPECIFICATIONS MPS3 UFSAR5.2-39Rev. 30TABLE 5.2-3 REACTOR VESSELS INTERNAL MATERIAL SPECIFICATIONSForgingsSA-182, Type F304 and F304H, or Type 403 per Westinghouse Procedure 80280NL PlatesSA-240, Type 304 PipesASTM A-358, Grade 304, Class 1, SA358 Grade 304 Class 1TubesSA-213, Type 304; SA249 Grade TP304; ASTM A-511, MT 304; and ASTM A-554, MT 304 BarsSA-479, Type 304 and 316

ASTM A-276, 304 and SB-166 CastingsSA-351, Gr. CF8 BoltingSA-193, Gr. B8M Code Case 1618, Inconel 750 SA-637, Gr. 688 Type 2; SA-479, Type 316, Strain Hardened (Code Case 1618)

NutsSA-194, Gr. 8 or 8A; SA-479, Type 304 and SA-637, Grade 688, Type 2Locking devicesSA-479, Type 304, 304L or Type 316, SA-240, Type 304, ASTM A-240, Type 304; and ASTM B-166 Weld butteringER 308, ER 308L, E308-15, E308L-15, E308T-3 MPS3 UFSAR5.2-40Rev. 30TABLE 5.2-4 REACTOR COOLANT WATE R CHEMISTRY SPECIFICATION Electrical conductivity De termined by the concentration of boric acid and alkali present, expected range is

< 5to 60 µS/cm at 25°C.Solution pH Determined by the concentration of boric acid and alkali present, expected values range between 4.5 (high boric acid concentrat ion) and 1 1.0 (low boric acid concentration) at 25

°C; value will be 6.9 or greater at normal operating temperatures when the reactor is critical.

Oxygen, maximum (ppm) 0.1 Chloride, maximum (ppm) 0.15 Fluoride, maximum (ppm) 0.15 Hydrogen (cc(STP)/Kg H 2O) 25 to 50Total suspended solids, maximum (ppm) 0.05pH control agent (Li7OH) (ppm) 0.3 to 6.0 as Li

Boric acid (ppm B) Variable from 0 to approximately 4,000NOTES:1.Oxygen concentration must be controlled to less than 0.1 ppm in the reactor coolant at temperatures above 250

°F by scavenging with hydrazine. During power operation with the specified hydrogen concentr ation maintained in the coolant, the residual oxygen concentration control value becomes0.005 ppm.2.Halogen concentrations must be maintained below the specified values at all times regardless of system temperature.3.Hydrogen must be maintained 15 cc (STP)/kg H 2 O whenever the reactor is critical. The normal operating range should be 25 to 50 cc (STP)/kg H 2 O.4.Solids concentration determined by filtra tion through filter ha ving 0.45 micron pore size.

Suspended solids concentrations as high as 0.35 ppm may be observed during startups and shutdowns. However , sustained plant operation with suspended solids > 0.05 ppm should be investigated, and crud mitigation measures taken as necessary.5.Limits for lithium hydroxide established for normal full power operation in conjunction with the fuel vendor. Prior to reactor criticality, sufficient lithium hydroxide is added to ensure a minimum at-temperature pH of a least 6.9. Lithium may be removed shortly before plant shutdown to aid in th e clean up of RCS corrosion products.

MPS3 UFSAR5.2-41Rev. 30aValue within ( ) equals Westinghouse allowable loads.bWhen combined with other loads in norm/upset

/test conditions, total value is less than Westinghouse allowable. Westinghouse has no allowable loads for pipe rupture.TABLE 5.2-5 SAFETY VALVE SUPPORT BRACKET LOADSLOADS (KIPS, INCH)

CONDITION F x F y F z M x M y M zDEAD WEIGHT0 70000 (5) a(20)(5)(100)(1)(32)THERMAL1833101350 243 b(20)(75)(20)(523)(5)(135)SEISMIC 1/2 SSE21211150027(20)(30)(20)(390)(2)(134)SEISMIC SSE22714190027(35)(50)(35)(645)(2)(235)VALVE OPER. (OCCASIONAL)114570014(30)(100)(30)(560)(2)(204)FAULTED CONDITION182779212460244(140)(325)(140)(1715)(2)(963)SAFETY LINE PIPE RUPTURE15229739860203 MPS3 UFSAR5.2-42Rev. 30TABLE 5.2-6 RELIEF VALVES REFERENCED TO CODE CASE N-242 Mark Number ServiceLocation3CCE*RV40A&BCharging Pump Cooler Relief ValvesAuxiliary Building3CCE*RV43A-CCooler Relief Valves Auxiliary Building3CCI*RV31A&BSafety Inject ion Pump Cooler Relief ValvesESF Building3CCI*RV36A&BCooler Relief Valves ESF Building3CCP*RV39 Excess Letdown Heat Exchanger Relief Valve Containment3CCP*RV54A-DReactor Coolant Pump Thermal Barrier Relief Valves Containment3CCP*RV59A&BFuel Pool Cooler Relief ValvesAuxiliary Building3CCP*RV64A&BResidual Heat Exchanger Relief ValvesAuxiliary Building3CCP*RV82Letdown Heat Exchanger Relief ValveAuxiliary Building3CCP*RV85Seal Water Heat Exchanger Relief ValveAuxiliary Building3CCP*RV239A&BRHR Pump Cooler Relief ValvesESF Building3CCP*RV258A-DReactor Coolant Pump Upper Bearing Relief Valves Containment3CCP*RV275A&BContainment Penetration Relief ValvesAuxiliary Building3CDS*RV105A&BContainment Penetration Relief ValvesAuxiliary Building3CDS*RV106A&BContainment Penetration Relief ValvesAuxiliary Building3CHS*RV7006Letdown Reheat Heat Exchanger Relief Valve Auxiliary Building3CHS*RV8119Letdown to Low Pressure Demineralizer Relief Valve Auxiliary Building3CHS*RV8120Volume Control Tank Relief ValveAuxiliary Building 3CHS*RV8121Seal Water Return Relief ValveAuxiliary Building 3CHS*RV8123RCP Seal Water Return HeaderAuxiliary Building3CHS*RV8124Charging Pump Suction Header Relief ValveESF Building3DAS*RV87 Containment Penetration Relief ValveAuxiliary Building3DGS*RV51 Containment Penetration Relief ValveAuxiliary Building3FWA*RV45Turbine Pump Relief ValveESF Building MPS3 UFSAR5.2-43Rev. 303FWS*RV47A-D3FWS*CTV41A-D Bonnet Relief ValvesMS Valve Building3GWS-RV35Degasifier Relief ValveAuxiliary Building3GWS-RV77GWS Relief ValveAuxiliary Building3PGS*RV77Containment Penetration Relief ValveAuxiliary Building3RHS*RV8708A&BRHR Pump Suction Relief ValvesContainment3SFC*RV52A&BFuel Pool Cooler Relief ValvesFuel Building3SIH*RV8925A&B*P1A&B Suction ReliefsESF Building3SIH*RV8851Cold Leg Injection Relief ValvesESF Building3SIH*RV8853A&BSIS Pump Discharge Relief ValvesESF Building3SIH*RV8858SIS Pump Suction Header Relief ValveESF Building3SIL*RV8842Hot Leg Injection Relief ValveESF Building 3SIL*RV8855A-DAccumulator Tank 1 Relief ValvesContainment 3SIL*RV8856A&BRHR Pumps Safety Injection Line Relief ValvesESF Building3SIL*RV8857Accumulator Nitrogen Supply LineContainment 3FWA*RV64A&B*P1A&B Suction ReliefESF Building3FWA*RV65*P2 Suction ReliefESF Building3CHS*RV8501A-C*P1A-C Suction ReliefESF Building 3FPW*RV87Containment Penetration Relief ValveAuxiliary Building3SWP*RV95B3CCP E1 Relief ValveAuxiliary BuildingTABLE 5.2-6 RELIEF VALVES REFERENCED TO CODE CASE N-242 Mark Number ServiceLocation MPS3 UFSAR5.2-44Rev. 30TABLE 5.2-7 MILLSTONE UNIT NO. 3 RTPTS VALUES (°F)1."M" is the margin term added to cover uncertainties in the values of initial RTNDT , copper and nickel content, fluence and calculational procedures.Location Chemical Wt.% Cu Content Wt.% Ni"I" Initial RTNDT"M" 1 Error Term Base Plate (CF/LF)0.050.636034Weld0.050.05-5040.23Vessel Inside Surface Fluence (E 1 MeV)10 19 n/cm 2 RTPTS (°F)Location54 EFPY54 EFPY Base Plate (CF/LF)2.70133Weld2.7030 MPS-3 FSAR September 1997 Rev. 20.2FIGURE 5.2-1 REACTOR VESSEL INSPECTION AREA FLANGE TO UPPER SHELL CIRCULAR WELD INLET NOZZLE TO SHELL WELD (TYP. OF 4)

UPPER TO INTERMEDIATE SHELL

CIRCULAR WELD INTERMEDIATE SHELL

LONGITUDINAL WELD (TYP. OF 3)

LOWER TO INTERMEDIATE SHELL

CIRCULAR WELD LOWER SHELL LONGITUDINAL

WELDS (TYP. OF 3)

LOWER HEAD TO SHELL WELD LOWER HEAD DISC TO LOWER RING

CIRCULAR WELD LOWER HEAD MERIDIONAL

WELDS (TYP. OF 4)

UPPER SHELL

LONGITUDINAL WELD (TYP. OF 3)

OUTLET NOZZLE TO

SHELL WELD (TYP. OF 4)

MPS-3 FSAR September 1997 Rev. 20.2FIGURE 5.2-2 MODEL F STEAM GENERATOR INSPECTION AREA MPS-3 FSAR December 1997 Rev. 20.2FIGURE 5.2-3 PRESSURIZER INSPECTION AREAS MPS3 UFSAR5.3-1Rev. 30

5.3 REACTOR

VESSEL

5.3.1 REACTOR

VESSEL MATERIALS 5.3.1.1 Material SpecificationsMaterial specifications are in accordance with the American Society of Mechanical Engineers (ASME) Code requirements and are given in Section 5.2.3.In addition, the ferritic materials of the reactor vessel beltline were rest ricted to the following maximum limits of copper, phosphorous, and vanadium to reduce sensitivity to irradiation embrittlement in service.

5.3.1.2 Special Process Used for Manufacturing and FabricationThe vessel is Safety Class 1. Desi gn and fabrication of the reactor vessel was carried out in strict accordance with ASME Code,Section III, Class 1 requirements. The head flanges and nozzles were manufactured as for gings. The cylindrical portion of the vessel is made up of several shells, each consisting of formed plates joined by full penetration longit udinal weld seams. The hemispherical heads were made with dished plates. The integral parts of the vessel and closure head subassemblies were joined by welding, pr imarily using the single or multiple wire submerged arc.

The use of severely sensitized stainless steel as a pressure boundary materi al has been prohibited and has been eliminated by either a select c hoice of material or by programming the method of assembly.The control rod drive mechanism adaptor threads and surfaces of the guide studs are chrome plated to prevent possible galling of the mated parts.

At all locations in the reactor ve ssel where stainless steel and Inconel are joined, the final joining beads are Inconel weld metal in order to prevent cracking.

Core region shells fabricated of plate material have longit udinal welds which are angularly located away from the peak neutron exposur e experienced in the vessel, where possible.ElementBase Metal (%) As Deposited Weld Metal (%)

Copper 0.10 0.10 0.12 (check)

Phosphorous 0.012 (ladle) 0.015 0.017 (check)Vanadium 0.05 (check) 0.05 (as residual)

MPS3 UFSAR5.3-2Rev. 30 The location of full penetration weld seams in the upper closure head and vessel bottom head are restricted to areas that permit accessibility during inservice inspection.The stainless steel clad surfaces were sampled to assure that compos ition and delta ferrite requirements are met.

The procedure for cladding low alloy steel (SA-508, Class 2) is qualified in accordance with the recommendations of Regulat ory Guide 1.43 (Section 1.8).

Minimum preheat requirements have been establ ished for pressure boundary welds using low alloy material. The preheat is maintained either until (at least) an inte rmediate post weld heat treatment is completed or until the completion of welding. In the latter case, upon completion of welding, a low temperature (400

°F minimum) post weld heat trea tment is applied for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by allowing the weldment to cool to am bient temperature. For primary nozzle to shell welds, preheat is maintained until an intermediate or full post weld heat treatment is completed.

5.3.1.3 Special Methods for Nondestructive Examination The nondestructive examination of the reactor vessel and its appurtenances is conducted in accordance with the ASME Code,Section III re quirements; also numerous examinations are performed in addition to ASME Code,Section III requirements.

Nondestructive examination of the vessel is discussed in the follow ing sections and shown in Table 5.3-1.

5.3.1.3.1 Ultrasonic Examination In addition to the ASME Code st raight beam ultrasoni c test, angle beam inspection of 100 percent of plate material was performed during fabrication to detect discontinuities that may be

undetected by longitudinal wave examination.

In addition to the ASME Code,Section III, nondestructive examinati on, all full penetration ferritic pressure boundary welds and heat af fected zones in the reactor vesse l were ultrasonically examined during fabrication. This test is performed upon completion of the welding and intermediate heat treatment but prior to th e final post weld heat treatment. Section 5.3.3.7 discusses this examination.

In addition to ASME Code,Section III, nondestru ctive examination, all full penetration ferritic pressure boundary welds in the reactor vessel were ultrasonically inspected after hydrostatic testing to establish additional assurance that th e vessel would pass the ASME Code,Section XI, preservice inspection requirements.

5.3.1.3.2 Penetrant ExaminationsThe partial penetration welds for the control rod drive mechanism head adaptors and the bottom instrumentation tubes were inspected by dye penetrant after the root pass in addition to code

requirements. Core support block a ttachment welds were inspected by dye penetrant after the first MPS3 UFSAR5.3-3Rev. 30layer of weld metal and after each one-half inch thickness of weld metal. All clad surfaces and other vessel and head internal surfaces were inspected by dye penetrant after the hydrostatic test.

5.3.1.3.3 Magnetic Particle Examination The magnetic particle examination requirements be low are in addition to the magnetic particle examination requirements of S ection III of the ASME Code.

All magnetic particle examinations of materials and we lds were performed in accordance with the following: 1.Prior to the final post weld heat trea tment - only by the Prod, Coil, or Direct Contact Method.2.After the final post weld treatment - only by the Yoke Method.

The following surfaces and welds were examined by magnetic particle methods. The acceptance standards are in accordance with Section III of the ASME Code.

5.3.1.3.3.1 Surface ExaminationsThere are three surface examinations: 1.All exterior vessel and head surfaces are magnetic particle examined after the hydrostatic test.2.All exterior closure stud surfaces a nd all nut surfaces ar e magnetic particle examined after final machining or roll ing. Continuous circul ar and longitudinal magnetization are used.3.All inside diameter surfaces of carbon and low alloy steel products that have their properties enhanced by acceler ated cooling are magnetic particle examined. This inspection is performed after forming and machining (if performed) and prior to cladding.5.3.1.3.3.2 Weld ExaminationTable 5.3-1 shows the non-destructive examinations for the Reactor Vessel.

5.3.1.4 Special Controls for Ferritic and Austenitic Stainless SteelsWelding of ferritic steels and austenitic stainles s steels is discussed in Section 5.2.3. Section 5.2.3 includes discussions which indicate the degree of conformance with Regulatory Guides 1.31 and 1.44. Section 1.8 discusses the degree of conf ormance with Regulat ory Guides 1.34, 1.43, 1.50, 1.71, and 1.99.

MPS3 UFSAR5.3-4Rev. 30 5.3.1.5 Fracture Toughness Assurance of adequate fracture t oughness of ferritic materials in the reactor vessel (ASME Code,Section III, Class 1 component) is provided by co mpliance with the requi rements for fracture toughness testing included in NB-2300 of Section III of the ASME Code and Appendix G of 10 CFR 50.The initial Charpy V-notch minimum upper shelf fracture energy levels for the reactor vessel beltline (including welds) ar e 75 foot-pounds as required pe r Appendix G of 10 CFR 50. The fracture toughness data for the reactor vessel are given in Table 5.3-2. Reactor vessel beltline region material composition is given in Table 5.3-3. The predicte d end-of-life beltline region material information is given in Table 5.3-

4. Plate locations are shown on Figure 5.3-1. The reactor vessel closure head stud, nut, and washer material information is given in Table 5.3-5.

5.3.1.6 Material Surveillance In the surveillance program, the evaluation of th e radiation damage is based on pre-irradiation testing of Charpy V-notch and tensile specimens and post-irradiation testing of Charpy V-notch, tensile and one-half T (thickness) compact tension (CT) fracture mechanics test specimens. The program is directed toward evaluation of the ef fect of radiation on th e fracture toughness of reactor vessel steels based on the transition temperature approach and the fracture mechanics approach. The program conforms with ASTM-E-185-82, "Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels," and 10 CFR 50, Appendix H.The reactor vessel surveillance program uses six specimen capsules. The specimens are oriented as required by NB-2300 of Section III of the ASME Code. The ca psules are located in guide baskets welded to the outside of the neutron sh ield pads and are positioned directly opposite the center portion of the core. The capsules can be removed when th e vessel head a nd upper internals are removed and can be replaced when the lower internals are removed. The six capsules contain reactor vessel steel specimens, oriented both parallel and normal (longitudinal and transverse) to the principal rolling direction of the limiting base material located in the core region of the reactor vessel associated weld metal and weld heat affected zone metal. The six capsules contain 54 tensile specimens, 360 Charpy V-notch specimens (which include weld metal and weld heat affected zone material), and 72 CT specimens. Archive material sufficient for two additional capsules is retained.

Dosimeters, including nickel (Ni), copper (Cu), iron (Fe), cobalt-aluminum (Co-Al), cadmium (Cd) shielded Co-Al, Cd shielded neptuni um-237 (Np-237), and Cd shielded uranium-238 (U-238), are placed in filler blocks drilled to contain them. The dosimeters permit evaluation of the flux seen by the specimens and the vessel wall. In addition, thermal monitors made of low melting point alloys are included to monitor the maximum temperature of the specimens. The specimens are enclosed in a tight fitting stainl ess steel sheath to prevent corrosion and ensure good thermal conductivity. The complete capsule was helium leak tested.

MPS3 UFSAR5.3-5Rev. 30 Each of the six capsules contains the following specimens: NOTES:*Specimens oriented in the majo r rolling or working direction.**Specimens oriented normal to th e major rolling working direction.***Weld metal to be selected per ASTM-E-185.

The following dosimeters and thermal monitors are included in each of the six capsules.

Dosimeters Iron Copper Nickel Cobalt-aluminum (0.15 percent cobalt)

Cobalt-aluminum (cadmium shielded)

Uranium-238 (cadmium shielded)

Neptunium-237 (cadmium shieldedThermal monitors 97.5 percent lead (Pb), 2.5 percent silver (Ag) (579

°F melting point).

97.5 percent lead (Pb), 1.75 percent silv er (Ag), 0.75 percent tin (Sn) (590

°F melting point).The fast neutron exposure of the specimens occurs at a faster rate than that experienced by the vessel wall, with the specimens being located between the core and the vessel. Since these specimens experience accelerated e xposure and are actual samples fr om the materials used in the vessel, the transition temperature shift measuremen ts are representative of the vessel at a later time in life. Data fro m CT fracture toughness specimens ar e expected to provide additional information for use in determining allowable stresses for irradiated material.

Correlations between the calculations and the measurements of the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, are described in Section 5.3.1.6.1.

Material Number of Charpys Number of Tensiles Number of Compact Tensions Limiting base material

• 15 3 4 Limiting base material

• • 15 3 4Weld metal

• • • 15 3 4Heat affected zone 15--

MPS3 UFSAR5.3-6Rev. 30 They have indicated good agreement. The anticipat ed degree to which th e specimens will perturb the fast neutron flux and energy distribution will be considered in the evaluation of the surveillance specimen data. Verification and possi ble readjustment of the calculated wall exposure will be made by use of data on all caps ules withdrawn. The sche dule for removal of the capsules and the measured or expect ed neutron fluence is as follows: (a)Updated in Capsule W dosimetry analysis.(b)Effective Full Power Years (EFPY) from plant startup.(c)Plant specific evaluation.

(d)This fluence is not less than once or greater than twice the pe ak end of license fluence, and is approximately equal to the peak vessel fluence at 63 EFPY.(e)Capsules Y and V were withdrawn after 13.80 EFPY (EOC 10) and placed into storage after accruing 2.98 x 10 19 n/cm 2 fluence.(f)Capsule Y was reinserted into location 61 ° at EOC 17 (approximately 23.4 EFPY).

(g)Capsule Z was withdrawn at approxima tely 23.4 EFPY (EOC 17) after accruing approximately 5.37 x 10 19 n/cm 2 fluence. Dosimetry analysis was performed and the test specimens placed into vendor storage for future testing.(h)This projected fluence is greater than once and less than twice the projected 72 EFPY and 90 EFPY peak vessel fluence.(i)Capsule Y is installed for fluence monitoring during the operating license in accordance with ASTM E 185-82.

This schedule meets the requirements of ASTM E 185-82. The Millstone 3 intermediate shell plate B9805-1 is the most limiting surveillance materi al based upon predicted ad justments of reference temperature, RTNDT in accordance with Regulatory Guid e 1.99. All materials are predicted to exhibit an EOL RTNDT of less than 100

°F, ASTM E185-82 requires that the program contain a CapsuleLocation Lead Factor (a)Removal Time (EFPY)(b)Fluence(n/cm 2 E>1.0MeV)(a)U58.5°4.06 1.3 4.00 x 10 18 (c)X238.5°4.35 8.0 1.98 x 10 19 (c)W121.5°4.22 13.8 3.16 x 10 19 (c)(d)Y (e)241°3.98 13.8--Y (f)61°3.98--Footnote (i)V (e)61°3.98Storage--Z (g)301.5°4.22 23.4 5.37 x 10 19 (h)

MPS3 UFSAR5.3-7Rev. 30 minimum of three capsules which ar e to be removed at three different times during the plant's life.

Six capsules are contained in the Millstone 3 program. The three primary surveillance capsules removed to date satisfy the ASTM E185-82 survei llance capsule withdrawal requirements for a design life of 54 EFPY. Additional standby capsules remain in the reactor or in the spent fuel pool. The standby capsules are managed to pr ovide surveillance data should subsequent extensions of the plant' s design life be desired.

5.3.1.6.1 Measurement of Integrated Fast Neutron (E

> 1.0 MeV) Flux at the Irradiation SamplesTo effect a correlation between fast neutron (E

> 1.0 MeV) exposure and the radiation induced properties changes observed in the test specimen s, a number of fast ne utron flux monitors are included as an integral part of the reactor vessel surveillance program. In particular , the surveillance capsules contain detect ors employing the following reaction.Fe-54(n,p) Mn-54Ni-58(n,p) Co-58Cu-63(n,) Co-60Np-237(n,f) Cs-137U-238(n,f) Cs-137 In addition, thermal neutron flux monitors, in the form of bare and Cd shielded Co-Al wire, are included within the capsules to enable an asse ssment of the ef fects of isotopic burnup on the response of the fast neutron detectors.

The use of passive neutron sensors such as included in the internal surveillance capsule dosimetry sets does not yield a direct measure of the energy dependent ne utron flux level at the measurement location. Rather, the activation or fi ssion process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the cour se of the irradiation period. An accurate assessment of the average flux level and, hence, time integrated exposure (fluence) experienced by the sensors may be deve loped from the measuremen ts only if the sensor characteristics and the parameters of the irradiation are well known. In particular, the following variables are of interest:1.The measured specific activity of each sensor 2.The physical characteristics of each sensor3.The operating history of the reactor 4.The energy response of each sensor5.The neutron energy spectrum at the sensor location MPS3 UFSAR5.3-8Rev. 30In this section the procedures used to determine sensor specifi c activities, to develop reaction rates for individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described.

5.3.1.6.1.1 Determination of Se nsor Reaction RatesThe specific activity of each of the radiometric sensors is determined using established ASTM procedures. Following sample preparation and weig hing, the specific activity of each sensor is determined by means of a lithium drifted germanium, Ge(Li), gamma spectrometer. In the case of the surveillance capsule multiple foil sensor sets, these analyses are perf ormed by direct counting of each of the individual wires; or, as in the case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution and chemical separation of cesium from the sensor.

The irradiation history of the reactor over its operating lifetime is obtained from NUREG-0020, "Licensed Operating Reactors Stat us Summary Report" or from other plant records. In particular, operating data are extracted on a monthly basis from reactor start up to the end of the capsule irradiation period. For the sensor sets utilized in the surveillan ce capsule irradiations, the half-lives of the product isotopes are long enough that a monthly histogr am describing reactor operation has proven to be an adequate representa tion for use in radioactiv e decay corrections for the reactions of interest in the exposure evaluations.

Having the measured specific activ ities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation are determined from the following equation:

where:A =measured specific activity (dps/gm)R =reaction rate averaged over th e irradiation period and referenced to operation at a core power level of P ref (rps/nucleus)

N 0 =number of target elemen t atoms per gram of sensorF =weight fractions of the target isotope in the sensor material Y =number of product atoms produced per reaction P j =average core power level during irradiation period j (MW)

P ref =maximum or reference core power level of the reactor (MW)

R A N 0 FYj---P j P ref---------C j 1et j-]et d--[-----------------------------------------------------------------------

=

MPS3 UFSAR5.3-9Rev. 30 C j =calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period = decay constant of the product isotope (sec

-1)t j = length of irradi ation period j (sec) t d =decay time following irradiation period j (sec) and the summation is carried out over the to tal number of monthly intervals comprising the total irradiation period.

In the above equation, the ratio P j/P ref accounts for month by month variation of power level within a given fuel cycle. The ratio C j is calculated for each fuel cycle and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel cycle. For a single cycle irradiation C j = 1.0. However, for multiple cycle irradiat ions, particularly those employing low leakage fuel management the additional C j correction must be utilized.

5.3.1.6.1.2 Corrections to Reaction Rate Data Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 5.3.1.6.1.3, additional corrections are made to the U-238 measurements to account for the presence of U-235 impurities in the sensors as we ll as to adjust for the build-in of plutonium isotopes over the course of the irradiation.In addition to the corrections made for the presence of U-235 in the U-238 fission sensors, corrections are also made to both the U-238 an d Np-237 sensor reaction rates to account for gamma ray induced fission reactions occu rring over the course of the irradiation.

5.3.1.6.1.3 Least Squares Adjustment ProcedureValues of key fast neutron exposure parameters are derived from the measured reaction rates

using the FERRET least squares adjustme nt code (SCHMITTROTH, 1979). The FERRET

approach uses the measured reacti on rate data, sensor reaction cro ss-sections, and a trial spectrum as input and proceeds to adjust the group fluxes from the spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data. The "best es timate" exposure parameters along with the associated uncertainties are then obtained from the best estimate spectrum.

In the least squares adjustment, the continuous qua ntities (i.e., neutron spectra and cross-sections) are approximated in a multi-group format consisting of 53 energy groups. The trial spectrum is converted to the FERRET 53 group structure us ing the SAND-II code (McELROY et. al., 1967).

This procedure is carried out by first expandi ng the trial spectrum into the SAND-II 620 group structure using a SPLINE inte rpolation procedure in regi ons where group boundaries do not coincide. The 620 point spectrum is then re-collapsed into the group structure used in FERRET.The sensor set reaction cross-sections, contained within FERRET, are also collapsed into the 53 energy group structure using the SAND-II code. In this instance, the trial spectrum, as expanded MPS3 UFSAR5.3-10Rev. 30 to 620 groups, is employed as a weighting function in the cross-section collapsing procedure.

Reaction cross-section uncer tainties in the form of a 53 x 53 covariance matrix for each sensor reaction are also constructed from the information contained on the ENDF/B-VI data files. These matrices include energy group to energy group uncerta inty correlations for each of the individual reactions.

Due to the importance of providi ng a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neut ron spectrum input to the FERRET evaluation is obtained from calculations for each dosimetry location (Section 5.3.1.6.2.1).

5.3.1.6.2 Calculation of Integrated Fa st Neutron (E. 1.0 MeV) Flux at the Irradiation Samples Fast neutron exposure calculations for the reactor geometry are carried out using both forward and adjoint discrete ordinates transport techniques. A single forward calculation provides the relative ener gy distribution of neut rons for use as input to neutr on dosimetry evaluations as well as for use in relating measurement results to the actual exposure at key locations in the pressure vessel wall. A series of adjoin t calculations, on the other hand, establish the means to compute absolute exposure rate values us ing fuel cycle specific core pow er distributions; thus, providing a direct comparison with all dosim etry results obtained over the operating history of the reactor.In combination, the absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra distributions from the forward calculation provided the means to:1.Evaluate neutron dosimetry fro m surveillance capsule locations.2.Enable a direct comparison of anal ytical prediction with measurement.3.Determine plant specific bias factors to be used in the evaluation of the best estimate exposure of the reactor pressure vessel.4.Establish a mechanism for projection of pr essure vessel exposure as the design of each new fuel cycle evolves.

5.3.1.6.2.1 Reference Forward Calculation The forward transport ca lculation for the reactor is carried out in r

, geometry using the DORT two dimensional discrete ordinates code (Version 3.1) and the BUGLE-96 cross-section library (ORNL). The BUGLE-96 library is a 47 neutr on group, ENDF/B-VI base d, data set produced specifically for light water reactor applications. In these analys es, anisotropic scattering is treated with a P 3 expansion of the scattering cr oss-sections and the angular di scretization is modeled with an S 8 order of angular quadrature.

The spatial core power distribution utilized in th e reference forward calculation is derived from statistical studies of long-term operation of Westinghouse four loop plants. Inherent in the development of this reference core power distri bution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a MPS3 UFSAR5.3-11Rev. 30 2 uncertainty derived from the statistical eval uation of plant to plant and cycle to cycle variations in peripheral power is used.

Due to the use of this bounding sp atial power distribution, the resu lts from the reference forward calculation establish conservative exposure projections for reactors of this design. Since it is unlikely that actual reactor operation would result in the implementation of a power distribution at the nominal +2 level for a large number of fuel cycles and, further, because of the widespread implementation of low leakage fuel management strategies, the fuel cycle specific calculations for this reactor will result in exposure rates well below these conservative predictions.

5.3.1.6.2.2 Cycle Specific Adjoint CalculationsAll adjoint analyses are al so carried out using an S 8 order of angular quadrature and the P 3 cross-section approximation from the BUGLE-96 library. Adjoint source lo cations are chosen at several key azimuths on the pressure vessel inner radius. In addition, adjoint calculations were carried out for sources positioned at the geometric center of all surveillance capsules. Again, these calculations are run in r, geometry to provide neutron sour ce distribution importance functions for the exposure parameter of interest; in this case, (E > 1.0 MeV).

The importance functions generated from these i ndividual adjoint analyses provide the basis for all absolute projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yield absolute predictions of neutron exposure at the locations of inte rest for each of the operating fuel cycles; and, establish the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles.

Having the importance functions and appropriate core source di stributions, the response of interest can be calculated as:

where:(R 0,0) = Neutron flux (E

> 1.0 MeV) at radius R 0 and azimuthal angle 0I(r,,E) = Adjoint importance function at radius r, azimuthal angle , and neutron source energy E.S(r,,E) =Neutron source strength at core location r, and energy E.It is important to note that the cycle specific neutron source distributions, S(r

,,E), utilized with the adjoint importance functions, I(r,,E), permit the use not only of fuel cycle specific spatial variations of fission rates within the reactor core; but, also allow for the inclusion of the effects of the differing neutron yield per fission and the variation in fission sp ectrum introduced by the build-in of plutonium isotopes as the burnup of individual fuel assemblies increases.R 00),(rE)SrE)r ,, (,, (rddE d Er=

MPS3 UFSAR5.3-12Rev. 30 5.3.1.6.2.3 Calculated Neutron Flux Distribut ions Within the Reactor Geometry The design basis core power dist ribution used in the transpor t analysis was derived from statistical studies of long-te rm operation of Westinghouse four-loop plants. Inherent in the development of the design basis co re power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2 uncertainty derived from the statistical eval uation of plant-to-plant and cycle-to-cycle variations in peripheral power was used. Since it is unlikely that a singl e reactor would have a power distribution at the nominal + 2 level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results. This is especially true in cases where low leakage fuel manageme nt has been employed. Having the calculated neutron flux distributions within the reactor geometry, the exposure of the capsule as well as the lead factor between the capsule and the vessel may be determined.

5.3.1.7 Reactor Vessel Fasteners The reactor vessel closure studs, nuts, and washers are designed and fabricated in accordance with the requirements of the ASME Code,Section III.

The closure studs are fabricated of SA-540, Class 3, Grade B24. The closure stud material m eets the fracture toughness requirements of the ASME Code,Section III and 10 CFR 50, Appendix G. Compliance with Regulatory Guide 1.65 is discussed in Section 1.8. Nondest ructive examinations are perfor med in accordance with the ASME Code,Section III. Fracture toughness data for bolting materials are presented in Table 5.3-5.Millstone Nuclear Power Station re fueling procedures require the studs, nuts, and washers to be removed from the reactor vessel closure and be placed in stor age racks during preparation for refueling. The storage racks are th en removed from the refueling cavity and stored in convenient locations on the containment operati ng deck prior to removal of th e reactor vessel closure head and refueling cavity flooding. Alternatively, the st uds, nuts and washers may be lifted out of the refueling cavity with the reactor vessel closure head. Therefore, the reactor vessel closure studs are never exposed to the borated refueling cavity water. Additional protection against the possibility of incurring corrosion effects is assured by the use of an initial manganese base phosphate surfacing treatment plus the use of an approved lu bricant. An alternate surface treatment to manganese base phosphate is a vapor phase plating process.

The stud holes in the reactor vessel flange are s ealed with special plugs before removing the reactor vessel closure head thus preventing leakage of the borated refueling water into the stud holes.5.3.2 PRESSURE-TEMPERATURE LIMITS 5.3.2.1 Limit CurvesStartup and shutdown operating limitations ar e based on the properties of the core region materials of the reactor pressure vessel. Actual material property test da ta is used. The methods outlined in Appendix G to Section XI of the AS ME Code as modified by ASME Code Case N-MPS3 UFSAR5.3-13Rev. 30 640 are employed for the shell regions in the analysis for protection against non-ductile failure. The initial operating curves are calculated assu ming a period of reactor operation such that the beltline material will be limiting. The heatup and cooldown curves are given in the Technical Specifications. Beltline ma terial properties degrade with radia tion exposure, and this degradation is measured in terms of the adju sted reference nil-ductility temp erature which includes a reference nil-ductility temperature shift (RTNDT).The limiting RTNDT used to establish the pressure/tempe rature limit curves, is periodically updated to incorporate the effects of irradiation exposure using the methodology described in Regulatory Guide 1.99, Revision 2. This methodology calculates the increase in RTNDT based on each material's copper c ontent and nickel content and also ba sed on the neutron fluence to which the material is expected to be exposed during the period of applicability of the pressure-temperature limit curves. RTNDT values are calculated for the 1/4t and 3/4t locations (i.e., tips of the ASME Code reference flaw when the flaw is assumed at the inside diameter and outside diameter locations), respectively. For the selected period of operation, this shift is of sufficient magnitude so that no unirradiated ferritic materials in other components of the reactor coolant

system (RCS) will be limiting in the analysis.

The operational curves (P/T limits) have been established for the ferritic materials of the RCS considering ASME Boiler and Pressure Vessel C ode Section XI, Appendix G as modified by ASME Code Case N-640, and the additional requirements of 10 CFR 50 Appendix G.

Implementation of the specific requirements provide adequate margin to brittle fracture of ferritic materials during normal operation, anticipated operational occurr ences, and system leak and hydrostatic tests. Changes in fracture toughness of the core region plates, weldments, and associated heat affected zones due to radiati on damage will be monitored by the surveillance program discussed in Section 5.3.1.6.

The results of the radiation surveillance program will be used to verify that the RTNDT predicted from the effects of the fluence, copper cont ent, and nickel content, using the methodology described in Regulatory Guide 1.99, Revision 2, is appropriate and to make any changes necessary to correct the chemistry factors as desc ribed in paragraph 2.1 of the Regulatory Guide if RTNDT determined from the surveillance program is greater than the predicted RTNDT. Temperature limits for inservice leak and hydrotests along with core criticality limits are included in the Technical Specifications. Note that the core criticality limits provide margins associated with brittle fracture and do not consider core physics.

5.3.2.2 End-of-Life RT PTS ProjectionsTo protect the reactor vessel ag ains t pressurized ther mal shock events, the NRC promulgated the PTS rule. This rule established end-of-life screening limits based on affects of neutron irradiation damage at the reactor vessel surface which woul d provide acceptable level of risk due to PTS events. This calculation is performed by predicting the shift in the reference transition temperature (RTNDT). The shift in the reference transition temperature (RTNDT) is calculated using the methodology provided by 10 CFR 50.61. The value of RT PTS can be calculated by the following expression:

MPS3 UFSAR5.3-14Rev. 30 RT PTS = Initial RTNDT + RTNDT + Margin This calculation provides an end-of-life value of R T PTS at the vessel clad/base metal interface based upon the limiting projected surface fluence of 2.70 x 10 19 n/cm 2 (E > 1MeV). Table 5.3-4 provides the results of the calculation for the limiting base and weld material.

5.3.2.3 Operating Procedures The transient conditions that are considered in the design of the reactor vessel are presented in Section 3.9N.1.1. These transients are representative of the operating conditions that should prudently be considered to occur during plan t operation. The transients selected form a conservative basis for evaluation of the RCS to ensure the integrity of the RCS equipment.

Those transients listed as upset condition transi ents are listed in tabl e 3.9N-1. None of these transients will result in pressure-temperatu re changes which exceed the heatup and cooldown limitations as described in Section 5.3.2.1 and in the Technical Specifications.

5.3.3 REACTOR

VESSEL INTEGRITY 5.3.3.1 DesignThe reactor vessel is cylindrical with a welded hemispherical bottom head and a removable, bolted, flanged, and gasketed, hemi spherical upper head (Figure 5.3-2). The rector vessel flange and head are sealed by two hollow metallic O-ri ngs. Seal leakage is detected by means of two leakoff connections: one between the inner and outer ring and one outside the outer O-ring. The vessel contains the core, core su pport structures, control rods, and other parts directly associated with the core. The reactor vessel closure head c ontains head adaptors. These head adaptors are tubular members, attached by part ial penetration welds to the unde rside of the closure head. The upper end of these adaptors contains acme thr eads for the assembly of control rod drive mechanisms, head adaptor plugs (s pares), or instrumentation adaptors. The seal arrangement at the upper end of these adaptors consists of a we lded flexible canopy seal, except for some space head adaptors with plugs that have mechanical Canopy Seal Clamp Assemb lies installed over the existing canopy seal welds to prevent possible l eakages. Inlet and outlet nozzles are located symmetrically around the vessel. Outlet nozzles ar e arranged on the vessel to facilitate optimum layout of the RCS equipment. The inlet nozzles are tapered from th e coolant loop vessel interfaces to the vessel inside wall to reduce loop pressure drop.

The bottom head of the vessel contains penetration nozzles fo r connection and entry of the nuclear incore instrumentation. Each nozzle consists of a tubular member made of Inconel. Each tube is attached to the inside of the bottom head by a partial penetration weld.Internal surfaces of the vessel which are in cont act with primary coolant are weld overlay with 0.125 inch minimum of stainl ess steel or Inconel. The exterior of the reactor vessel closure head is insulated with canned stainless st eel reflective insulation. The reactor vessel assembly is insulated with canned stainless steel panels of fibrous, powdered and reflective insulation. The insulating MPS3 UFSAR5.3-15Rev. 30 panels are contoured to enclose the entire vessel. Top and bottom head and top dome insulation panels are designed to be easily removed for inspection, maintenance and refueling.

Principal design parameters of the reactor vessel are given in Table 5.3-6.There are no special design features which would prohibit the in situ annealing of the vessel. Various modes of heating could be used depending on the desired temperature.The reactor vessel materials surv eillance program is adequate to accommodate the annealing of the reactor vessel. Sufficient specimens are available to evaluate the effects of the annealing treatment.

Cyclic loads are introduced by normal power ch anges, reactor trip, startup, and shutdown operations. These design base cycles are selected for fatigue evaluation and constitute a conservative design envelope for the projected plant life. Vessel analyses result in a usage factor that is less than one.

The design specifications require analysis to prove that the vessel is in compliance with the fatigue and stress limits of the AS ME Code,Section III. The loadi ngs and transients specified for the analysis are based on the most severe c onditions expected during service. The maximum heatup and cooldown rate consistent with plant operating limits is 100

°F per hour for normal operating conditions. These rates are reflected in the ve ssel design specifications.

5.3.3.2 Materials of Construction The materials in the fabrication of the r eactor vessel are discussed in Section 5.2.3

.5.3.3.3 Fabrication Methods The Millstone Unit 3 reactor vessel manufactur er is Combustion Engineering Incorporated.

Combustion Engineering Incorporated is the lar gest reactor vessel fabricator in the United States and their experience is demonstrated by the fact th at they have fabricated over 40 reactor vessels for Westinghouse designed NSSS's as well as additional vessels for other reactor vendors.

The fabrication methods used in the construction of the reactor vessel are discussed in Section 5.3.1.2.

5.3.3.4 Inspection Requirements The nondestructive examinations performed on the reactor vessel are described in Section 5.3.1.3.

5.3.3.5 Shipment and Installation The reactor vessel is shipped in a horizontal position on a shipping sled with a vessel lifting truss assembly. All vessel opening are sealed to prevent the entrance of moisture and an adequate quantity of desiccant bags are placed inside the vessel. These are usually placed in a wire mesh MPS3 UFSAR5.3-16Rev. 30basket attached to the vessel cover. All carbon steel surfaces are pa inted with a heat resistant paint before shipment except for the vessel support surfaces and the top surface of the external seal ring.The closure head is also shipped with a shippi ng cover and skid. An enclosure attached to the ventilation shroud support ring pr otects the control rod mechanism housings. All head openings are sealed to prevent the entrance of moisture and an adequate quantity of desiccant bags are placed inside the head. These are placed in a wire mesh basket attached to the head cover. All carbon steel surfaces are painted with heat resistant paint before shi pping. A lifting frame is provided for handling the vessel head.

5.3.3.6 Operating Conditions Operating limitations are presented in Secti on 5.3.2 and in the T echnical Specifications. The procedures and methods used to ensure the inte grity of the reactor vessel under the most severe postulated conditions are de scribed in Section 3.9N.1.4.

In addition to the analysis of primary com ponents discussed in Se ction 3.9N.1.4, the reactor vessel is further qualified to ensure agains t unstable crack growth under faulted conditions. Actuation of emergency core cooling system (E CCS) following a loss-of-coolant or steam line break accident procedures relatively high thermal stresses in re gions of the reactor vessel which come into contact with ECCS water. Primary consideration is gi ven to these areas, including the reactor vessel beltline region and the reactor vessel primary coolant nozzle, to ensure the integrity of the reactor vessel under these severe postulated transients.

The principles and procedures of linear elastic fracture mechanics (LEFM) are used to evaluate thermal effects in the regions of interest.

The LEFM approach to the design against failure is basically a stress inte nsity consideration in which criteria are established for fracture instability in the presence of a crack. Consequently, a basic assumption employed in LEFM is that a crack or crack-like defect exists in the structure. The essence of the approach is to relate the stress field developed in the vicinity of the crack tip to the applied stress on the structure, the material properties, and the size of defect necessary to cause failure.

The elastic stress field at the crack tip in any cracked body can be described by a single parameter designated as the stress intensity factor, K. The ma gnitude of the stress intensity factor K is a function of the geometry of the body containing the crack, the size and location of the crack, and the magnitude and distribution of the stress.

The criterion for failure in the presence of a crack is that failure will occur whenever the stress intensity factor exceeds some critical value. For the opening mode of loading (stresses perpendicular to the major plane of the crack) the stress intensity fact or is designated as K I and the critical stress intensit y factor is designated K IC. Commonly called the fracture toughness, K IC is an inherent material property which is a function of temperature and strain rate. Any MPS3 UFSAR5.3-17Rev. 30 combination of applied lo ad, structural configuration, crack geometry, and size which yields a stress intensity f actor greater than or equal to K IC for the material will result in crack instability.

The criterion of the applicabilit y of LEFM is based on plasticity considerations at the postulated crack tip. Strict applicability (as defined by ASTM) of LEFM to large structures where plane strain conditions prevail requires that the plastic zone devel oped at the tip of the crack does not exceed 2.25 percent of the crack depth. However, LEFM has been su ccessfully used to provide conservative brittle fract ure prevention evaluations, even in cases where strict applicability of the theory is not permitted due to excessive plasticity. Recently, experiment al results from Heavy Section Steel Technology Program in termediate pressure vessel te sts have shown that LEFM can be applied conservatively as l ong as the pressure component of the stress does not exceed the yield strength of the material. The addition of the thermal stresses, calculated elastically, which results in total stresses in excess of the yield strength does not affect the conservatism of the results, provided that these thermal stresses are in cluded in the evaluation of the stress intensity factors. Therefore, for faulted condition analyses, LEFM is considered applicable for the evaluation of the vessel inlet nozzle and beltline region.

In addition, it has been well esta blished that the crack propagation of existing flaws in a structure subjected to cyclic loading can be defined in terms of fracture mechanics parameters. Thus, the principles of LEFM are also applic able to fatigue growth of a postu lated flaw at the vessel inlet nozzle and beltline region.

An example of a faulted conditi on evaluation carried out accordi ng to the procedure discussed above is given in WCAP-8099, 1973). This report discusses the evaluation procedure in detail as applied to a severe fau lted condition (a post ulated loss-of-coolant acciden t) and concludes that the integrity of the reactor coolant pressure boundary would be maintained in the event of such an accident.5.3.3.7 Inservice Surveillance The internal surface of the reacto r vessel is capable of inspection pe riodically using visual and/or nondestructive techniques over the accessible area

s. During refueling, th e vessel cladding is capable of being inspected in certain areas betw een the closure flange and the primary coolant inlet nozzles, and, if deemed necessary , the core barrel is capable of being removed, making the entire inside vessel surface accessible.

The closure head is examined visually per the applicable ASME Edition and Addenda of Section XI, "Rules for Inservice Inspection of Nuclear Components." Optical devi ces permit a selective inspection of the cladding, contro l rod drive mechanism nozzles, a nd the gasket seating surface.

The knuckle transition piece, which is the area of highest stress of the closure head, is accessible on the outer surface for visual inspection, dye pe netrant or magnetic particle, and ultrasonic testing. The closure studs can be inspected peri odically using visual, ma gnetic particle, and/or ultrasonic techniques.

MPS3 UFSAR5.3-18Rev. 30The closure studs, nuts, washers, and the vessel flange seal surface, as well as the full penetration welds in the following areas of the installed irradiated reactor vessel are available for visual and/

or nondestructive inspection.1.Vessel shell - from the inside surface2.Primary coolant nozzles - from the inside surface 3.Closure head - from the insi de and outside surfaces Bottom head - from the outside surface4.Field welds between the reactor vesse l nozzles and the main coolant piping The design considerations which have been incor porated into the system design to permit the above inspection are as follows: 1.All reactor internals are completely removable. The tool s and storage space required to permit these inspections are provided.2.The closure head is stored dry on the ve ssel head storage stand during refueling to facilitate direct visual in spection.3.All reactor vessel studs, nuts and washer s can be removed to dry storage during refueling.4.Removable plugs are provided in the pr imary shield. Insula tion around the nozzles may be removed for inspection requirements.The reactor vessel presents access problems be cause of the radiation levels and remote underwater accessibility to this component. Beca use of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufacturing procedures in preparation for the periodic nondestructive test s which are required by the ASME inservice inspection code. These are: 1.Shop ultrasonic examinations are performe d on all internally cl ad surfaces to an acceptance and repair standard to assure an adequate cladding bond to allow later ultrasonic testing of the base metal from the inside surface. The size of cladding bonding defect allowed is 0.25 in ch by 0.75 inch with the gr eater direction parallel to the weld in the region bounded by 2 T (T = wall thickne ss) on both sides of each full penetration pressure boundary we ld. Unbounded areas exceeding 0.442 square inch (0.75 inch diameter) in all other region s are rejected.2.The design of the reactor vessel shell is a clean, uncluttered cylindrical surface to permit future positioning of the test equipment without obstruction.

MPS3 UFSAR5.3-19Rev. 303.The weld deposited clad surface on both sides of the welds to be inspected is specifically prepared to assure meaningful ultrasonic examinations.4.During fabrication, all fu ll penetration ferritic pr essure boundary welds are ultrasonically examined in addition to ASME Code, Sectio n III, requirements.5.After the shop hydrostatic testing, all full penetration ferritic pressure boundary welds are ultrasonically examined in addition to ASME Code,Section III, requirements.

The vessel design and construction enables insp ection in accordance with the ASME Code,Section XI.5.

3.4 REFERENCES

FOR SECTION 5.35.3-1WCAP-8099 1973, Buchalet, C. and Mager, T.

R., "A Summary Analysis of the April 30 Incident at the San Onofre Nuclear Generating Station Unit 1," Westinghouse Corp., Pittsburgh, Penn.5.3-2Soltesz, R. G. et al., 1970, "Nuclear Rocket Shielding Methods, Modification, Updating, and Input Data Preparation, Volume 5 - Two- Dimensional Discrete Ordinates Technique. WANL-PR-(LL)-034."5.3-3Schmittroth, E.A., "FERRETT Data Anal ysis Code," HEDL-TME-79-40, Hanford Engineering Development Laboratory

, Richland, Washington, September 1979.5.3-4McElroy, W. N., et. al., "A Computer-Automated Iterative Me thod of Neutron Flux Spectra Determined by Foil Activation,"

AFWL-TR-67-41, Volumes I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.5.3-5"ORNL RSCI Data Library Collection DLC-76, SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray , P 3 , Cross Section Library for Light Water Reactors."5.3-6RSICC Compute Code Collection CCC-650, DOORS 3.1, One, Two, and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System, August 1996.5.3-7RSIC Data Library Collection DLC-185, BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetr y Applications, March 1996.

MPS3 UFSAR5.3-20Rev. 30NOTES:*RT = Radiographic**UT = Ultrasonic***PT = Dye penetrant

• • • • MT = Magnetic particleTABLE 5.3-1 REACTOR VESSEL NON-DESTRUCTIVE EXAMINATIONRT*UT**PT***MT****ForgingsFlangesyesyesStuds and nutsyesyesHead adapter flangesyesyes Head adapter tubesyesyesInstrumentation tubesyesyesMain nozzlesyesyes Nozzle safe endsyesyesPlatesyesyes WeldmentsMain seamyesyesyesControl rod drive head adapter connectionyesInstrumentation tube connectionyes Main nozzleyesyesyesCladdingyesyesNozzle safe endsyesyesyes Head adapter forging to head adapter tubeyesyesAll ferritic welds accessible after hydrotestyesyes All non-ferritic weld s accessible after hydrotestyesyesSeal ledgeyes Head lift lugsyes Core pad weldsyesVessel Support Weld Buildupyesyes*****

MPS3 UFSAR5.3-21Rev. 30*****= Required inspection; pr ogressive MT or Final UT MPS3 UFSARMPS3 UFSAR5.3-22Rev. 30TABLE 5.3-2 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIESAvg. Upper ShelfComponentCode No.GradeCu (%)N (%)T (°F)RT (°F)NMWD (ft-lb)MMWD (ft-lb)Closure Head DomeB9812-1A533B, CL. 10.08-40096.0---Closure Head TorusB9813-1A533B, CL. 10.11-4010107.5---

Closure Head FlangeB9803-1A508, CL. 2---3030121.0---Vessel FlangeB9801-1A508, CL. 20.11-40-40116.5---Inlet NozzleB9806-3A508, CL.20.091010162.0---

Inlet NozzleB9806-4A508, CL. 20.0900158.0---Inlet NozzleR5-3A508, CL. 20.07-10-10130.0---Inlet NozzleR5-4A508, CL. 20.0800136.0---

Outlet NozzleR6-1A508, CL. 2----40-40128.0---Outlet NozzleR6-2A508, CL. 2----30-30127.0---Outlet NozzleB9807-1A508, CL. 2----30-30121.0---

Outlet NozzleB9807-2A508, CL. 2----30-30126.0---Nozzle ShellB9804-1A533B, CL. 10.05-404085.5---Nozzle ShellB9804-2A533B, CL. 10.08-4040104.5---

Nozzle ShellB9804-3A533B, CL. 10.05-500103.5---Inter, ShellB9805-1A533B, CL. 10.050.64-4060113.389.0Inter, ShellB9805-2A533B, CL. 10.050.64-606.290.070.7 Inter, ShellB9805-3A533B, CL. 10.050.65-40-3.3106.3136.5 MPS3 UFSARMPS3 UFSAR5.3-23Rev. 30NOTES: NMWD = normal to major working direction MWD = major working direction Lower ShellB9820-1A533B, CL. 10.080.63-507.076.7124.5 Lower ShellB9820-2A533B, CL. 10.070.60-3038.875.7114.5 Lower ShellB9820-3A533B, CL. 10.060.61-3018.679.3124.0Bottom Head TorusB9816-1A533B, CL. 10.13-50-4091.5---Bottom Head DomeB9817-1A533B, CL. 10.15-30-30161.0---TABLE 5.3-2 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIESAvg. Upper ShelfComponentCode No.GradeCu (%)N (%)T (°F)RT (°F)NMWD (ft-lb)MMWD (ft-lb)

MPS3 UFSAR5.3-24Rev. 30 ND = not detectedNOTE: Applicable for all beltline region weld seams.TABLE 5.3-3 REACTOR VESSEL BELTLIN E REGION MATER IAL CHEMICAL COMPOSITION (WT PERCENT)

ElementPlate B9805-1 Plate B9805-2Plate B9805-3Plate B9820-1Plate B9820-2 Plate B9820-3Weld Control 4P6052C0.230.230.220.220.240.210.14Mn1.321.321.391.371.421.381.25 P0.0100.0140.0090.0060.0080.0070.011 S0.0100.0120.0100.0190.0180.0230.009 Si0.210.220.220.220.240.220.12 Ni0.640.620.650.630.600.610.05 Cr0.030.030.030.050.030.030.03 Mo0.570.590.580.600.600.570.48 Cu0.050.050.050.080.070.060.05 Cb< 0.01< 0.01< 0.01< 0.01< 0.01< 0.01---

PbNDNDNDNDNDND---

W< 0.01< 0.01< 0.01< 0.01< 0.01< 0.01---

As0.0050.0060.0070.0060.0050.004---

Sn0.0030.0050.0050.0030.0020.001---

Co0.0120.0130.0120.0110.0110.011---

N20.0070.0060.0100.0080.0090.008---

Al0.0240.0240.0250.0200.0320.033---

V0.0060.0060.0040.0030.0050.0050.004 B< 0.001< 0.001< 0.001< 0.001< 0.001< 0.001---

Ti< 0.01< 0.01< 0.01< 0.01< 0.01< 0.01---

Zr< 0.001< 0.001< 0.001< 0.001< 0.001< 0.001---

MPS3 UFSAR5.3-25Rev. 30TABLE 5.3-4 ADJUSTED REFERENCED TEMPERATURE (ART) VALUES (°F)Location Chemical Wt. % Cu Content Wt. % Ni Initial RTNDT Margin at 54 EFPYBase Plate (B9805-1)0.050.636034Weld0.050.05-5040.23 Location F=Fluence (E 1 MeV)10 19 n/cm 2 at 54 EFPYSurface ART Expiration Date at 54 EFPYBase Plate (B9805-1)2.70133Weld2.7030 MPS3 UFSAR5.3-26Rev. 30TABLE 5.3-5 REACTOR VESS EL DESIGN PARAMETERS Design/operating pressure (psig)2485/2317 Design temperature (°F)650 Overall height of vessel and closure h ead, bottom head outside diameter to top of control rod mechanism adapter (foot-inch) 43-10 Thickness of canned stainless steel insulation 3 Reflective and fibrous insulation (inch) Powdered insulation (inch)1 Number of reactor closure head studs 54 Diameter of reactor closure head

/studs, minimum shank (inch) 6-13/16 Inside diameter of flange (inch) 167 Outside diameter at flange (inch) 205 Inside diameter at shell (inch) 173 Inlet nozzle inside diameter (inch) 27.5 Outlet nozzle inside diameter (inch) 29 Clad thickness, minimum (inch) 1/8 Lower head thickness, minimum (inch) 5-3/8Vessel beltline thickness, minimum (inch)

8.5 Closure

head thickness (inch) 7 MPS-3 FSAR Rev. 20.2FIGURE 5.3-1 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE REACTOR VESSEL 90°90°270°270°0°0°180°180°101-124A B9805-2 101-124B B9805-3 101-142A B9820-1 101-142B B9820-2 101-142C B9820-3 101-171 101-124C B9805-1 LOWER SHELL INTERMEDIATE SHELL CORE MPS-3 FSAR June 1997 Rev. 20.2FIGURE 5.3-2 REACTOR VESSEL CONTROL ROD MECHANISM HOUSING COOLING DUCT

SUPPORT RING CLOSURE STUD, NUT & WASHER OUTLET NOZZLE INSTRUMENTATION TUBE INLET NOZZLE MONITOR TUBE LIFTING LUG VENT PIPE CORE SUPPORT LUG MPS3 UFSAR5.4-1Rev. 30

5.4 COMPONENT

AND SUBSYSTEM DESIGN

5.4.1 REACTOR

COOLANT PUMPS 5.4.1.1 Pump Flywheel IntegrityThe integrity of the reactor coolant pump flywheel is assured on the basis of the following design and quality assurance procedures.

5.4.1.1.1 Design Bases The calculated stresses at operating speed are ba sed on stresses due to centrifugal forces. The stress resulting from the interfer ence fit of the flywheel on the sh aft is less than 2,000 psi at zero speed, but this stress becomes zero at approximately 600 rpm because of radial expansion of the hub. The reactor coolant pumps run at approximately 1,190 rpm and may operate briefly at

overspeeds up to 109 percent (1,295 rpm) during loss of off site electrical power. For conservatism, however, 125 percent of operating sp eed was selected as the design speed of the reactor coolant pumps. The flywheels are given a preoperational test of 125 percent of the maximum synchronous speed of the motor.

5.4.1.1.2 Fabrication and Inspection The flywheel consists of two thic k plates bolted together. The flyw heel material is produced by a process that minimizes flaws in the material and impr oves its fracture toughness properties (i.e., an electric furnace with vacuum degassing). Each plate is fa bricated from SA-5 33, Grade B, Class 1 steel. Supplier certi fication reports are available for all plates and demonstrate the acceptability of the flywheel material on the basis of the requirements of NRC Regulatory Guide 1.14.

Flywheel blanks are flam e-cut from the SA-533, Grade B, Class 1 plates, with at least 0.5 inch of stock left on the outer surface and bore surface fo r machining to final dimensions. The finished machined bores, keyways, and dr illed holes are subjected to magne tic particle or liquid penetrant examinations in accordance with the requirements of Section III of the ASME Code. The finished flywheels, as well as the flywhe el material (rolled plate), are subjected to 100 percent volumetric ultrasonic inspection using procedures and acceptance standards specified in Sect ion III of the ASME Code.The reactor coolant pump motors are designed such that, by removing the cover to provide access or by removing the flywheel from the pump motor shaft, the flywheel is available to allow an inservice inspection program. For a description of inservice inspecti on of the flywheels, refer to the MP3 ISI Program.

5.4.1.1.3 Material Acceptance Criteria The reactor coolant pump motor fl ywheel conforms to the following material acceptance criteria:

MPS3 UFSAR5.4-2Rev. 301.The nil-ductility transition temperature (NDTT) of the flyw heel material is obtained by two drop weight tests (DWT) which exhibit "no-break" performance at 20°F in accordance with ASTM E-208. The above drop weight tests demonstrate that the NDTT of the flyw heel material is no higher than 10

°F.2.A minimum of three Charpy V-notch (C) im pact specimens from each plate shall be tested at ambient (70

°F) temperature in accordan ce with the specification ASME SA-370. The Charpy V-notch (C) energy in both the parallel and normal orientation with respect to the final rolling direction of the flyw heel plate material is at least 50 foot pounds and 35 mils lateral expansion at 70

°F and, therefore, the flywheel material has a reference ni l-ductility temperature (R T) of 10°F. An evaluation of flywheel overspeed has been performed which concludes that flywheel integrity will be maintained (WCAP-8163, 1973).

Thus, it is concluded that flywheel plate materi als are suitable for use and can meet Re gulatory Guide 1.14 acceptance criteria on the basis of suppliers certification data. The degree of compliance with Regulatory Guide 1.14 is further discussed in Section 1.8.

5.4.1.2 Reactor Coolant Pump Assembly 5.4.1.2.1 Design Bases The reactor coolant pump assembly ensures an adequa te core cooling flow ra te for sufficient heat transfer to maintain a departure from nucleate boiling ratio (DNBR) greater than 1.30 within the parameters of operation. The required net positive suction head is, by conservative pump design, always less than that availabl e by system design and operation.Sufficient pump assembly rotational inertia is provided by a motor flywheel, motor rotor, and pump rotating parts which provide adequate flow during coastdown conditions. This forced flow following an assumed loss of off site electrical power and the subsequent natural circulation effect provides the core with adequate cooling. The reactor coolant pump motor is tested, without mechanical damage, at overspeed s up to and including 125 percent of normal speed. The integrity of the flywheel during a loss-of-coolant accide nt (LOCA) has been demonstrated and is undergoing generic review by the NRC (WCAP-8163, 1973).The reactor coolant pump is shown on Figure 5.4-1. The reactor coolant pump design parameters are given in Table 5.4-1.

Code and material requirement s are provided in Section 5.2.

5.4.1.2.2 Pump Assembly Description Design Description The reactor coolant pump is a vertical, single-stage, controlled leakage, centrifugal pump designed to pump large volumes of reactor coolant at high temperat ures and pressures.

MPS3 UFSAR5.4-3Rev. 30 The pump assembly consists of three major sections: the hydraulics, the seals, and the motor.1.The hydraulic section consists of the casi ng, impeller, turning vane-diffuser, and diffuser adapter.

2.The seal section consists of three identi cal mechanical face-typ e sealing stages in series, assembled as a single piece cartri dge. The seal system provides a pressure breakdown from the reactor coolant system (RCS) pressure to ambient conditions.3.The motor section consists of a drip pr oof, squirrel cage, induction motor with a vertical solid shaft, an oil-lubricated double-acting Kingsbury type thrust bearing, upper and lower oil lubricated radial guide bearings, and a flywheel.

Additional components of the pump are the shaft, pump radial bearing, thermal barrier heat exchanger assembly , coupling, spool piece, and motor stand.

Description of Operation The reactor coolant enters the suction nozzle, is pumped by the impeller through the diffuser, and exits through the discharge nozzle. The diffuser adapter limits the leakage of reactor coolant back to the suction.

Seal injection flow, under slightly higher pressure than the reactor coolant, enters the pump through a connection on the thermal barrier flange and is directed into the plenum between the thermal barrier housing and the shaft. The flow splits with the major portion flowing down the shaft through the radial bearing a nd into the reactor coolant system. The remaining seal injection flow passes up the shaft through the seals.

Component cooling water (Section 9.2.2.1) is provided to the thermal barrier heat exchanger.

During normal operation, the thermal ba rrier limits th e heat transfer from hot reactor coolant to the radial bearing and to the seals. In addition, if a loss of seal injection flow should occur, the thermal barrier heat exchanger cool s the reactor coolant to an acceptable level before it enters the bearing and seal area.

The reactor coolant pump motor oil lubricated bearings are of conventional design. The radial bearings are the segmented pad type, and the th rust bearing is a double-acting Kingsbury type.

Component cooling water is supplied to the external upper bearing oil cooler and to the integral lower bearing oil cooler. Each RCP motor is equipped with an oil co llection system to mitigate the consequences of oil leaks. Section 9.5.1 1 describes this system in detail.

MPS3 UFSAR5.4-4Rev. 30 The motor is a drip-proof, squirrel-cage, i nduction motor with Class B thermalastic epoxy insulation, and fitted with extern al water/air coolers. The roto r and stator are of standard construction and are cooled by air. Six resistance temperature detectors are embedded in the stator windings to sense stator temperatur

e. A flywheel and an anti-reverse rotation device are located at the top of the motor.

The internal parts of the motor are cooled by air. Integral vanes on each end of the rotor draw air in through cooling slots in the motor frame. This air passes through the motor with particular emphasis on the stator end turns. It is then routed to the external water/ai r heat exchangers, which are supplied with chilled wate r (Section 9.2.2.2). Each motor has two such coolers, mounted diametrically opposed to each other. Coolers are sized to maintain optimum motor operating temperature. The air is finally exha usted to the contai nment environment.

Each of the reactor coolant pump assemblies is equipped for continuous monitoring of reactor coolant pump shaft and frame vibration levels. Shaft vibration is measured by two relative motion shaft probes mounted on top of the pump seal housing; the probes are located 90 degrees apart in the same horizontal plane and mounted near the pump shaft. Frame vibration is measured by two velocity seismoprobes located 90 degrees apart in the same horizontal plane and mounted at the top of the motor support stand. Proximeters and converters linearize the probe output which is displayed on monitor meters in the control room. The monitor meters automatically indicate the highest output from the relative probes and seism oprobes; manual selection allows monitoring of individual probes. Indicator lights display caution and danger l imits of vibration.

The spool piece, a removable shaf t segment, is located between the motor c oupling flange and the pump coupling flange. The spool pi ece allows removal of the pump se als with the motor in place. The pump internals, motor, and motor stand can be removed from the casing without disturbing the reactor coolant piping. The flywheel is available for inspection by removing the cover.

All parts of the pump in contact with the reactor coolant are austenitic stainless steel except for seals, bearings, and special parts.

5.4.1.3 Design Evaluation 5.4.1.3.1 Pump Performance The reactor coolant pumps are sized to deliver flow at rates which equal or exceed the required flow rates. Initial RCS tests confirm the total delivery capability. Thus, assurance of adequate forced circulation coolant flow is provided prior to initial plant operation.

The estimated performance characteristic s are shown on Figure 5.4-2. The "knee", at approximately 25 percent design flow, introduces no operational restric tions, since the pumps only operate at a speed which corresponds to full flow.

The reactor trip system ensures that pump ope ration is within the assumptions used for loss-of-coolant flow analyses, which also assures th at adequate core cooling is provided to permit an orderly reduction in power if flow from a reactor coolan t pump is lost during operation.

MPS3 UFSAR5.4-5Rev. 30

M ajor parameters influencing the seal environment which can effect seal life include axial and radial shaft motions, radial shaf t vibrations, temperature, pressure, oxidizing water chemistry, the presence of particulates, and pump start/stop cycles. The sealing syst em has demonstrated through design, testing, and field operation to be capable of withstanding all specified operating conditions.

The seal cartridge uses three identical seal stages based on hydrodynamic operating principles.

The critical parts of each seal stage are the rotating face ring, stationary face ring, and secondary seals. A secondary seal O-ring is used to isolat e stage pressures and provides a sliding secondary seal between the stationary ring and the bala nce sleeve. This arrangement eliminates the requirement for a flat surface to support the stationary ring. The stationary face subassembly is mounted to the pressure breakdown device with springs. By flexibly mounting the stationary face subassembly, the stationary face can accommodate axial and radial displacement of the rotating face subassembly with minimum disruption to the lubricating film. In addition, the backing springs provide the seal closure force when sealing pressure is low and aid the hydraulic force balance when sealing pressure is low. The optimiz ed deflection control of the seal design results in repeatable and predictable be havior with greater operating margin to tolerate transients.

During normal operation, each seal stage will be subjected to a differential pressure of approximately one-third of reactor coolant system (RCS) pressure. Each of the three individual sealing stages is designed to with stand full RCS pressure indefinitely with the RCP idle, and for a limited period of time with the pump running at a nominal speed of 1200 rpm, to allow for a controlled shutdown. The seal is designed to operate with a thin fluid film gap. As a result, design allowances must be made for short-term contact of the seal face ring materials, part icularly during low pressure pump starts. Therefore, the stationary seal face ring material is resin-impregnated graphite. Rotating face ring materials for this application include chro mium carbide, silicon carbide, silicon nitride, and tungsten carbide. Tungsten carbide is used in the seal because of its good fracture resistance and thermal conductivity along with favorable tr ibologic properties. All of the elastomers performing static sealing functions in th e seal cartridge are ethylene propylene.

The normal operating mode of the sealing system, with one-third of RCS pressure across each stage is created by tubular seal staging flow coils. The coil is part of a subasse mbly designated the pressure breakdown device (PBD). Th ere is a separate staging coil for each sealing stage, located in the pressure retaining housing for that stage. T hus, each coil acts as an orifice to reduce the pressure available at each seal stage, resulting in equal pressu re distribution amongst the stages (unless there is significant leak age through one or more of the seal stages). A second function of MPS3 UFSAR5.4-6Rev. 30the flow, aside from developing seal system pre ssure distribution, is to provide cooling flow through the sealing system to carry away fricti onal heat generated by the rotating seal parts.

Maintaining stable seal temperat ures is important to limit th ermal gradients during transient conditions. The existing cooling systems - thermal barrier and injection - have been maintained without change for the RCP. The effect of loss of off site power on the pump itself is to cause a temporary stoppage in the supply of injection flow to th e pump seals and also of the com ponent cooling water for seal and bearing cooling. The emergency generators are started automatically due to loss of off site electrical power so that component cooling flow and seal injection flow are automatically restored.5.4.1.3.2 Coastdown CapabilityIt is important to reactor protection that the reactor coolant continue s to flow for a short time after reactor trip. In order to provide this flow following loss of outside electrical power, each reactor coolant pump is provided with a flywheel. Thus, the rotating inertia of the pump, motor, and flywheel is employed during the coastdown period to continue the reactor coolant flow. The coastdown flow transients are provided on the figures in Secti on 15.3. The pump/motor system is designed for the Safe Shutdown Earthquake (SSE) at the site. Hence, it is concluded that the coastdown capability of the pumps is maintained even under the most adverse case of loss of off site electrical power coincident with the SSE. Core flow transients and figures are provided in Section 15.3.1.

5.4.1.3.3 Bearing Integrity The design requirements for the reactor coolant pum p bearings are primarily aimed at ensuring a long life with negligible wear , so as to give accurate alignment and smooth operation over long periods of time. The surf ace bearing stresses are held at very low values, and even under the most severe seismic transients do not begin to approach loads which cannot be adequately carried for short periods of time.Because there are no established criteria for short ti me, stress related failures in such bearings, it is not possible to make a meanin gful quantification of such parameters as margins to failure, safety factors, etc. A qualitativ e analysis of the bearing design , embodying such considerations, gives assurance of the adequacy of th e bearing to operate without failure.

Low lube oil levels in the motor lube oil sumps signal an alarm in the control room. Each motor bearing containing embedded temperat ure detectors, and so initiation of failure is monitored as a high bearing temperature on the control room computer. Upon contro l room receipt of a low level alarm, bearing temperature is monitored and once the manufacturer's recommended maximum temperature is reached, the reactor is tripped followed by RCP trip. If bearing temperature indications are ignored, and the bearing proceeded to failure, th e low melting point of Babbitt metal on the pad surfaces ensures that sudden seizure of the shaft will not occur. In this event, the motor continues to operate, as it has sufficient reserve capacity to drive the pump under such MPS3 UFSAR5.4-7Rev. 30conditions. However, the high torque required to drive the pump will require high current which will lead to the motor being shutdow n by the electrical protection systems.

5.4.1.3.4 Locked RotorIt may be hypothesized that the pump impeller might severely rub on a stationary member and then seize. Analysis has shown that under such conditions, assumi ng instantaneous seizure of the impeller , the pump shaft fails in torsion just below the coupling to the motor, disengaging the flywheel and motor from the shaft. This constitute s a loss of coolant flow in the loop. Following such a postulated seizure, the motor continues to run without any overs peed, and the flywheel maintains its integrity, as it is still supported on a shaft with two bearings. Flow transients are provided in Section 15.3.3 for the assumed locked rotor.There are no other credible sources of shaft seizure other than impeller rubs. A sudden seizure of the pump bearing is precluded by graphite in the bearing. Any seizure in the seals results in a shearing of the anti rotation pin in the seal ring. The motor has ad equate power to continue pump operation even after the above occurrences.

Indications of pump malfunction in these conditi ons are initially given by high temperature signals from the bearing water temperature detector, and by excessive CVC seal return (CBO) indications, respectively.

5.4.1.3.5 Critical Speed The reactor coolant pump shaft is designed so that its operating speed is below its first critical speed. This shaft design, even under the most seve re postulated transient, gives low values of actual stress.

5.4.1.3.6 Missile Generation Precautionary measures taken to preclude missile formation from reactor coolant pump components assure that the pum ps do not produce missiles unde r any anticipated accident condition. Appropriate components of the reactor coolant pump have been analyzed for missile generation. Any fragments of the motor rotor would be contained by the heavy stator frame. The same conclusion applies to the pump impeller because the small fr agments that might be ejected would be contained by the heavy casing. Further discussion a nd analysis of missi le generation are contained in WCAP-8163.

5.4.1.3.7 Pump Cavitation The minimum net positive suction head required by the reactor coolant pump at best estimate flow is approximately a 300 foot head (approximately 133 psi). In or der for the controlled leakage seal to operate correctly, it is necessary to require a minimum differential pressure of approximately 200 psi across the seal. This corresponds to a primary loop pressure at which the minimum net positive suction head is exceeded and no limitation on pump operation occurs from this source.

MPS3 UFSAR5.4-8Rev. 30 5.4.1.3.8 Pump Overspeed ConsiderationsFor turbine trips actuated by either the reactor tr ip system or the turbine protection system, the generator and reactor coolant pumps remain connect ed to the external network for 30 seconds to prevent any pump overspeed condition.An electrical fault requiring immediate trip of the generator (with resulting turbine trip) could result in an overspeed condition. However , the tu rbine control system and the turbine intercept valves limit the overspeed to less than 120 perc ent. As additional backup, the turbine protection system has a mechanical overspeed protection trip, usually set at about 110 percent (of turbine speed). In case a generator trip deenergizes the pump buses, the reactor c oolant pump motors will be transferred to off site power within 6 to 10 cycles. Overspeed of the pump, due to a discharge side pipe rupture, is prevented by the motor which when connected to the elec trical system acts as an induction generator. The electrical connect ion box is located 180 degrees from the pump discharge side so that both the electrical lead s and the connection box are protected by the motor from a jet impingement of the r eactor coolant. This protection is required for 5 seconds so the motor can prevent overspeed due to the desc ribed condition. Further discussion of pump overspeed considerations is contained in WCAP-8163.

5.4.1.3.9 Anti-Reverse Rotation Device Each of the reactor coolant pumps is provided wi th an anti-reverse rotation device in the motor. This anti-reverse mechanism cons ists of pawls mounted on the outside diameter of the flywheel, a serrated ratchet plate mounted on the motor frame, a spring return for the ratchet plate, and two shock absorbers.At an approximate forward speed of 70 rpm, the pawls drop and bounce across the ratchet plate; as the motor continues to slow, the pawls drag across the ratchet plate. After the motor has slowed and come to a stop, the dropped pawls engage the ra tchet plate and, as the mo tor tends to rotate in the opposite direction, the ratchet plate also rotate s until it is stopped by the shock absorbers. The rotor remains in this position until the motor is energized again. When the motor is started, the ratchet plate is returned to its original position by the spring return.

As the motor begins to rotate, the pawls drag over the ratchet plate. When the motor reaches sufficient speed, the pawls are bounced into an elevated position a nd are held in that position by friction resulting from centrifugal forces acti ng upon the pawls. While the motor is running at speed, there is no contact betwee n the pawls and ratchet plate.

Considerable plant experience with the design of the anti-reverse rota tion device has shown high reliability of operation.

5.4.1.3.10 Shaft Seal Leakage

MPS3 UFSAR5.4-9Rev. 30Leakage along the reactor coolant pump shaft is controlled by three identical mechanical face-type sealing stages in series assembled as a single cartridge assembly, such that reactor coolant leakage to the containment is minimized.

Since leakage flow through a given seal stage is in parallel with the staging coil for that stage, effectively by-passing the coil, cavity pressures in the seal can change with variations in seal leakage. The pressure differential across the leaking seal stage will decrease while the two non-leaking seals equally share an increase in pressure differential (of equal magnitude to the loss of pressure differential across the leaking seal stage). 5.4.1.3.11 Seal Discharge PipingThe seal reduces the leakoff pressure to that of the volume control tank. Seal return water from each pump seal is piped to a common manifold, through the seal water return filter , and through the seal water heat exchanger wh ere the temperature is reduced to that of the volume control tank.The seal leakoff line connected to the upper stage directs seal l eakage to the containment sump.

5.4.1.4 Tests and Inspections The reactor coolant pumps can be inspected in accordance with the ASME Code,Section XI, for inservice inspection of nuclear reactor coolant systems.The pump casing is cast in one piece, thus eliminating the inservice inspection of welds in the casing. Support feet are cast integral with the casing to eliminate a weld region.

The design enables disassembly and removal of the pump internals for visual access to the internal surface of the pump casing.The reactor coolant pump quality assurance program is given in Table 5.4-2.

5.4.2 STEAM

GENERATORS The nuclear steam supply system (NSSS) uses four Model F steam generators as shown on Figure 5.4-3. Analysis of conditions that might compromise the reactor coolant boundary are addressed in this section.

5.4.2.1 Steam Generator Materials 5.4.2.1.1 Selection and Fabrication of Materials All pressure boundary materials used in the stea m generator are selected and fabricated in accordance with the requirements of Section III of the ASME Code. A general discussion of materials specifications is given in Section 5.2.3, with types of materials listed in Tables 5.2-2 and 5.2-3. Fabrication of reactor coolant pressure boundary materials is also discussed in Section 5.2.3, particularly in Sections 5.2.3.3 and 5.2.3.4.

MPS3 UFSAR5.4-10Rev. 30Testing has justified the selection of corrosion resistant In conel 600, a nickel-c hromium-iron alloy (ASME SB-163), for the steam generator tubes. Th e channel head divider plate is Inconel (ASME SB-168). The interior surfaces of the reactor cool ant channel head, nozzles, and manways are clad with austenitic stainless steel.

The primary side of the tube sheet is weld clad with Inconel (ASME SFA-5.14). The tubes are then seal welded to the tube sheet cladding. These fusion welds, performed in compliance with Sections III and IX of the ASME Code, are dye penetrant inspected and leak proof tested before each tube is hydraulically expanded th e full depth of the tube sheet bore.

Code cases used in material selection are disc ussed in Section 5.2.1. The extent of conformance with Regulatory Guides 1.84, "Design and Fabricat ion Code Case Acceptability ASME Section III Division 1," and 1.85 "Materials Code Case Ac ceptability ASME Section III Division 1," is discussed in Section 1.8.

During manufacturing, cleaning is performed on the primary and secondary sides for the steam generator in accordance with written procedures which follow the guidance of Regulatory Guide 1.37, Quality Assurance Require ments for Cleaning of Flui d Systems and Associated Components of Water-Cooled Nuclear Power Plants, and ANSI Standard N45.2.1-1973, Cleaning of Fluid Systems and Associated Components for Nuclear Power Plants. On site cleaning and cleanliness control standards are described in the Quality Assurance Program Description Topical Report. Cleaning process specificat ions are discussed in Section 5.2.3.4.

The fracture toughness of the materials is discussed in Section 5.2.3.3. Adequate fracture toughness of ferritic materials in the reactor coolant pressure boundary is provided by compliance with 10 CFR Part 50, Appendix G, Fracture Toughness Requirement s, and Paragraph NB-2300 of Section III of the ASME Code.

5.4.2.1.2 Steam Generator Design Effects on Material Several features have been introduced into the Model F steam generator to minimize the deposition of contaminants from the secondary side flow. Such deposits could otherwise produce a local environment in which adve rse conditions could develop and re sult in material attack. The support plates are made of corrosion resistant stainless steel 405 alloy and incorporate a four-lobe hole design (quatrefoil) that provides greater flow area adjacent to the tube outer surface and eliminates the need for interstitial flow holes.

The resulting increase in flow provides higher sweeping velocities at the tube/tube support plate intersections. Figure 5.4-4 illustrates the quatrefoil broached holes. This modification in the support plate design is a major factor contributing to the increased circulation ratio. The increased circulation results in increased flow in the interior of the bundle, as well as increased horizontal veloci ty across the tube sheet reducing the tendency for sludge deposition. The effect of the increased circulation on the vibrational stability of the tube bundl e has been analyzed with considerat ion given to flow induced excitation frequencies. The unsupported span length of tubing in the U-bend region and the corresponding optimum number of anti-vibration bars has be en determined. The anti-vibration bars are fabricated from square Inconel barstock, which is then ch rome plated to improve frictional characteristics. Also, due to the increased circulation ratio, the moisture separating equipment has MPS3 UFSAR5.4-11Rev. 30 been modified to maintain an adequate margin with respect to the moisture carryover. To provide added strength as well as resistance to vibrati on, the quatrefoil tube support plate thickness has been increased. In addition, 12 peripheral supports also provide stability to th e plates so that tube fretting or wear due to flow induced plate vibr ations at the tube support contact regions is minimized.Assurance against significant flow induced tube vibration has been obtained by a combination of analysis and testing.Combining both vortex shedding and turbulence effects in a conservative manner, the maximum predicted local tube wear dept h of a 60 year operating design objective is less than 0.008 inch.

This value is considerably below the pluggi ng limit for a Model F steam generator tube.

5.4.2.1.3 Compatibility of Steam Generator Tubi ng with Primary and Secondary Coolants As mentioned in Section 5.4.2.1.1, corrosion tests which subjected the steam generator tubing material, Inconel 600 (ASME SB-163), to simula ted steam generator wa ter chemistry have indicated that the loss due to general corrosi on over the 60 year opera ting design objective is insignificant compared to the tube wall thickness. Testing to investig ate the susceptibility of this material to stress corrosion in caustic and chlo ride aqueous solutions ha s indicated the Inconel 600 has excellent resistance to general and pitting type corrosion in severe operating water conditions. Many reactor years of successful operation have shown the same low general corrosion rates as indicated by the laboratory tests.

Operating experience has revealed areas on sec ondary surfaces and in crevice regions where localized corrosion rates were significantly greater than the low general corrosion rates. Intergranular attack intergranular stress corr osion cracking and tube wall thinning were experienced in localized areas, although not at the same location or under the same environmental conditions (water chemistry, temp erature and sludge composition).

The secondary side water chemistry program as described in Section 10.3.5 minimizes the possibility for developing localized corrosion a nd essentially eliminates the secondary side tube wall thinning phenomenon. Successful all volatile treatment "AVT" operation requires

maintenance of low concentrations of impurities in the steam generator bulk water, thus reducing the potential for formation of highly concentrated solutions in localized areas, which is the precursor of corrosion. By restric tion of the total alkalinity in the steam generator and prohibition of extended operation with free alkalinity, the AV T program should minimize the possibility for recurrence of intergranular corr osion in localized areas due to excessive levels of free caustic.

Laboratory testing has shown that the Inconel 600 tubing is compatible with the AVT environment. Isothermal corrosion testing in high purity water has shown that commercially

produced Inconel 600 exhibiting norma l microstructures test ed at normal engine ering stress levels does not suffer intergranular stress corrosion cracking in extended expos ure to high temperature water. These tests also showed that no general t ype of corrosion occurred. A series of autoclave tests in reference secondary wate r with planned excursions have produced no corrosion attached after 1,938 days of testing on any as produced, Inconel 600 tube samples.

MPS3 UFSAR5.4-12Rev. 30 Successful secondary side water chemistry c ontrols combined with a comprehensive steam generator inservice inspection program as desc ribed in Section 5.4.2.2, assure that the steam generators will provide reliable service. The inspection program will also facilitate detection of any unanticipated steam ge nerator tube degradation.Increased margin against primar y and secondary side stress corr osion cracking has been obtained by the use of thermally treated Inconel 600 tubing. Thermal treatment of Inconel tubes has been shown to be particularly effective in resisting caustic cracking. Tubing used in the Model F is thermally treated in accordance with a laboratory derived treatment process. In addition, the low rows of tubes were thermally stre ss relieved prior to installation. This further reduces the potential for stress corrosion cracking in the small radius U-bends.

The tube support plates used in the Model F are ferritic stainless steel, which has been shown in laboratory tests to be resistant to corrosion in the AVT environment. If corrosion of ferritic stainless steel was to occur, due to concentrat ion of contaminants, the volume of the corrosion products is essentially e quivalent to the volume of the parent material consumed. This would be expected to preclude denting. The support plates are also designed with quatrefoil tube holes rather than cylindrical holes. The quatrefoil tu be hole design promotes high velocity flow along the tube and should sweep impurities away from the support plate location.

Additional measures are incorporated in the Mo del F design to prevent areas of dryout in the steam generator and accumulations of sludge in lo w velocity areas. Modifications to the wrapper have increased water velocities ac ross the tube sheet. A flow distribution baffle is provided which forces the low flow area to the center of the bundle. In creased capacity blow down pipes have been added to enable continuous blowdown of the stea m generators at a high volume. The intakes of these blowdown pipes are located below the center cut out section of the flow distribution baffle in the low velocity region where sludge may be expected to accumulate. Continuous blowdown

should provide protection against inleakage of impurities from the condenser.

5.4.2.1.4 Cleanup of Secondary Side Materials Several methods are employed to clean opera ting steam generators of corrosion causing secondary side deposits. Sludge lancing, a procedur e in which a hydraulic jet inserted through an access opening (handhole) loosen s deposits which are removed by means of a suction pump, can be performed when the need is indicated by the results of steam generator tube inspection. Six 6-inch access ports are provided for sludge lancing and inspection. Three of these are located above the tube sheet and three above the flow distribution baffle. Continuous blowdown is performed to regulate water chemistry. The location of the blowdown piping suction, adjacent to the tube sheet and in a region of relatively low flow velocity, facilitates the rem oval of particulate impurities to minimize the accumulation on the tube sheet.

5.4.2.2 Steam Generator Inservice Inspection The steam generator is designed to permit in spection of ASME Code Class 1 and 2 parts, including individual tubes. The design includes a number of openings to provide access to both the primary and secondary sides of the steam generator. The specifi ed inspection program MPS3 UFSAR5.4-13Rev. 30 complies with the edition of the ASME Code , Division 1,Section XI required by 10 CFR 50.55a, effective January 5, 1977. The openings include four manways, two for access to both chambers of the reactor coolant channel h ead inlet and outlet sides and two in the stea m drum for inspection and maintenance of the moisture separators; six, 6 inch handholes, three located just above the tube sheet secondary surface and three located just above the flow distribution baffle; and two, 2.5 inch inspection ports located on the tube lane diameter between the upper tube support plate and

the Row 1 U-bend. Additional access to the tube U-bend is provided through each of the three deck plates. For proper functioning of the steam generator, some of the deck plate openings are covered with welded, but remova ble, hatch plates. Inspection/ac cess to the primary sides is provided by two, 16 inch manways located in the channel head.Regulatory Guide 1.83, "Inservice inspection of PWR Steam Generator tubes," and Generic Letter 2004-01, "Requirements for Steam Generator tube inspecti ons," provide recommendations concerning the inspection of tubes, which cover in spection equipment, baseline inspections, tube selection, sampling and frequency of inspection, methods of record ing, and required actions based on findings. Regulatory Guide 1.121, Basis for Plugging Degraded PWR Steam Generator Tubes, provides recommendations concerning the tube plugging. Agreement with Regulatory Guides

1.83 and 1.121 is discussed in Sect ion 1.8. The minimum requirements for inservice inspection of steam generators, including tube plugging criteria, are established as part of the Technical Specifications. The inse rvice inspection program for the react or coolant boundary is discussed in Section 5.2.4.

5.4.2.3 Design BasisSteam generator design data are gi ven in Table 5.4-3. Code classifi cations for the steam generator components are given in Section

3.2. Although

the ASME classificati on for the secondary side is specified to be Class 2, the current philosophy is to design all pressu re retaining parts of the steam generator, and thus both the primary and secondary pressure boundaries, to satisfy the criteria specified in Section III of the ASME Code fo r Class 1 components. The design stress limits, transient conditions, and combined loading cond itions applicable to the steam generator are discussed in Section 3.9N.1. Estimate s of radioactivity levels antici pated in the secondary side of the steam generators during normal operation a nd the bases for the estimates are given in Chapter 11. The accident analysis of a steam ge nerator tube rupture is discussed in Chapter 15.

A design objective of the internal moisture separa tor equipment is that moisture carryover should not exceed 0.25 percent by weight under the following conditions:1.Steady state operating up to 100 percent of fu ll load steam flow with water at the normal operation level2.Loading or unloading at a ra te of 5 percent of full pow er steam flow per minute in the range of 15 to 100 percent of full load steam flow3.A step load change of 10 percent of full power in the range of 15 to 100 percent full load steam flow MPS3 UFSAR5.4-14Rev. 30 The water chemistry on the reactor side, select ed to provide the necessary boron content for reactivity control, should minimi ze corrosion of RCS surfaces. The effectiveness of the water chemistry of the steam side in affecting co rrosion control is discussed in Chapter 10.

Compatibility of steam generator tubing with both primary and secondary coolants is discussed further in Section 5.4.2.1.3.The steam generator is designed to minimize unacceptable damage from mechanical or flow induced vibration. Tube support adequacy is discussed in the Design Evaluation Section. The tubes and tube sheet are analyzed and confirme d to withstand the maxi mum accident loading conditions as they are defined in Section 3.9N.1. Further consider ation is given in the Design Evaluation Section to the effect of tube wall thinning on accident condition stresses.

5.4.2.4 Design DescriptionThe steam generator is a Model F, vertical shell and U-tube eva porator , with integral moisture separating equipment. Figure 5.4-3 shows the model, indicating several of its improved design features described in th e following paragraphs.

On the primary side, the reactor coolant flows through the inverted U-tubes, entering and leaving through nozzles located in the hemispherical bottom head of the steam generator. The head is divided into inlet and outlet chambers by a vertic al divider plate extending from the apex of the head to the tube sheet.

Steam is generated on the shell side, flows upwar d, and exits through the outlet nozzle at the top of the vessel. Feedwater enters the steam generato r at an elevation above the top of the U-tubes, through a feedwater nozzle. The wa ter is distributed circumferent ially around the steam generator by means of a feedwater ring and then flows thr ough an annulus between the tube wrapper and shell. The feedwater enters the ring via a weld ed thermal sleeve connect ion and leaves through inverted "J" tubes located at the flow holes at the top of the ring. The "J" tubes are arranged to distribute the bulk of the colder feedwater to the hot leg side of the tube bundle. The feed ring is designed to minimize conditions which can result in water hammer occurr ences in the feedwater piping. At the bottom of the wrapper, the water is directed toward the center of the tube bundle by a flow distribution baffle. This baffle arrangement serves to minimize the tendency in the

relatively low velocity fluid for sludge deposit ion. Flow blocking devices discourage the water from flowing up the bypass lane as it enters th e tube bundle, where it is converted to a steam-water mixture. Subsequently, the steam-water mi xture from the tube bundle rises into the steam drum section, where 16 individual ce ntrifugal moisture separators remove most of the entrained water from the steam. The steam continues to the secondary separators for further moisture removal, increasing its quality to a designed minimum of 99.75 percent. The moisture separators reintroduce the separated water, which is combin ed with entering feedwater to flow back down the annulus between the wrapper and shell for recirculation through the steam generator. The dry steam exits from the steam generator through the outlet nozzle which is provided with a steam flow restrictor (Section 5.4.4).

MPS3 UFSAR5.4-15Rev. 30 5.4.2.5 Design Evaluation Forced ConvectionThe effective heat transfer coeffi cient is determined by the physical characteristics of the Model F steam generator and the fluid conditions in the primary and secondary sy stems for the "nominal" 100 percent design case. It includes a conservati ve allowance for fouling and uncertainty. A designed heat transfer area is provided to permit the achievability of the full design heat removal rate.Natural Circulation Flow The driving head created by the cha nge in coolant density as it is heated in the core and rises to the outlet nozzle initiates convection circulation. This circulation is enhanced by the fact that the steam generators, which provide a heat sink, are at a higher elevation than the reactor core which is the heat source. Thus, natural circulation is provided for the removal of decay heat during hot shutdown in the unlikely event of loss of forced circulation.Mechanical and Flow-Induced Vibrati on Under Normal Operation Conditions In the design of the steam generators, the possi bility of degradation of tubes due to either mechanical or flow-induced exci tation is thoroughly evaluated. Th is evaluation includes detailed analysis of the tube support system s, as well as an extensive rese arch program with tube vibration model tests.

In evaluating degradation due to vi bration, consideration is given to sources of excitation, such as those generated by primary fluid fl owing within the tubes, mechan ically induced vibration, and secondary fluid flow on the outside of the tubes. During normal operation, the effects of primary fluid flow within the tubes and mechanically induced vibration are consider ed to be negligible and should cause little concern. T hus, the primary source of tube vibrations is the hydrodynamic excitation by the secondary fluid on the outside of the tubes. In general, three vibration mechanisms have been identified:1.Vortex shedding2.Fluidelastic excitation3.TurbulenceVortex shedding does not provide detectable tube bundle vibration. There are several reasons why this happens:1.Flow turbulence in the downcomer a nd tube bundle inlet region inhibit the formation of Von Karman's vortex train.

MPS3 UFSAR5.4-16Rev. 302.The spatial variations of cross-flow velocities along the tube precludes vortex shedding at a single frequency

.3.Both axial and cross-flow velocity components exist on the tubes. The axial flow component disrupts the V on Karman vortices.

Fluidelastic excitation was obser ved during the testing. The amplit udes of the vibrations were smaller than those of the turbulent flow induced vibrations. Therefore, fl uidelastic excitation is excluded from consideration as a factor in steam genera tor tube bundle vibrations.

Flow-induced vibrations due to flow turbulence cause stresses in the tubes that ar e more than two orders of magnitude below the endurance limit (30,000 psi) of the tube material. Therefore, the contribution to fatigue is negligible, and fatigue degradation from flow-i nduced vibration is not anticipated.

Summarizing the results of analysis and tests of st eam generator for vibration, it can be stated that a check of all modes of tube vibration mechanisms has been comp leted. The conclusions that can be drawn are that the primary sour ce of tube vibration is fluid tu rbulence and the magnitude of the vibration is so small that when combined with its total random nature, its contribution to tube fatigue is negligible. Therefor e, fatigue degradation due to flow induced vibration is not anticipated.Allowable Tube Wall Thinni ng Under Accident Conditions An evaluation has been performed to determine th e extent of tube wall thinning that can be tolerated under accident conditions. The worst case loading conditions are assumed to be imposed upon uniformly thinned tubes, at th e most critical location in the steam generator. Under such a postulated design basis accident, vibration is of short enough duration that there is no endurance problem to be considered. The steam generator tubes, existing originally at their minimum wall thickness and reduced by a conservative general corrosion and erosion loss, can be shown to provide an adequate safety margin (i.e., sufficient wall thickness, in addition to the minimum required for a maximum stress less than the allowable stress limit, as it is defined by the ASME Code).The results of a study made on "D series" (0.75 inch nominal diameter, 0.043 inch nominal wall thickness) tubes under accident loadings are discussed in WCAP-7832 (1973). These results demonstrate that a minimum wall thickness of 0.026 inch would have a maximum faulted condition stress (i.e., due to combined LOCA a nd safe shutdown earthquake loads) that is less than the allowable limit. This th ickness is 0.010 inch less than the minimum "D seri es" tube wall thickness of 0.039 inch, which is reduced to 0.036 inch by the assumed ge neral corrosion and erosion rate. Thus, an adequate safety margin is exhibited. The corrosion rate is based on a conservative weight loss rate for Inconel tubing in flowing 650

°F primary side reactor coolant fluid. The weight loss, when equated to a thinning rate and projected over a 60 year design

objective with appropriate reducti on after initial hours, is equiva lent to 0.083 mil thinning. The assumed corrosion rate of 3 mils leaves a conservative 2.917 mils for general corrosion thinning on the secondary side.

MPS3 UFSAR5.4-17Rev. 30The Model F steam generator was analyzed using similar assumptions of general corrosion and erosion rates. The overall similarity between prev ious tubes studied and the Model F tubes makes it reasonable to expect the same ge neral results, that is, to conclude that the ability of the Model F steam generator tubes to withstand accident loadings is not impa ired by a lifetime of general corrosion losses. This is confirmed by specific analysis.

5.4.2.6 Quality Assurance The steam generator nondestructive examin ation program is given in Table 5.4-4.Radiographic inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code.

Liquid penetrant inspection is performed on weld deposited tube sheet cladding, channel head cladding, divider plate to tube sheet and to channel head weld ments, tube to tube sheet weldments, and weld deposit cladding. Liquid penetrant inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code.Magnetic particle inspection is performed on the tube sheet forging, channel head casting, nozzle forgings, and the following weldments:1.Nozzle to shell2.Support brackets 3.Instrument connection (secondary)4.Temporary attachments for removal5.All accessible pressure retain ing welds after hydrostatic test Magnetic particle inspection and acceptance standards are in accord ance with the requirements of Section III of the ASME Code.

Ultrasonic tests are performed on the tube sheet forgings, tube sheet cladding, secondary shell and head plates, and nozzle forgings. Inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code.

The heat transfer tubing is subjected to eddy current test ing and ultrasonic examination. Inspection and acceptance standards are in accordance with the requi rements of Sect ion III of the ASME Code. Hydrostatic tests are performed in accordance with Section III of the ASME Code.

In addition, the heat transfer t ubes are subjected to a hydrostatic test pressure not less than 1.25 times the primary side design pressure prior to installation into the vessel.

MPS3 UFSAR5.4-18Rev. 30

5.4.3 REACTOR

COOLANT PIPING 5.4.3.1 Design BasesThe reactor coolant system (RCS) piping is designed and fabricated to accommodate the system pressures and temperature attain ed under all expected modes of plant operation or anticipated system interactions. Stresses are maintained within the limits of Section III of the ASME Nuclear Power Plant Components Code. Section 5.2 pr ovides code and mate rial requirements.

Materials of construction are specified to minimize corrosion/er osion and ensure compatibility with the operating environment.

The piping in the RCS is Safety Class 1 and is designed and fabr icated in accordance with ASME Section III, Class 1 requirements.

Stainless steel pipe conforms to ANSI B36.19 for sizes 0.5 inch through 12 inches and wall thickness Schedules 40S through 80S.

Stainless steel pipe outsid e of the scope of ANSI B36.19 conforms to ANSI B36.10.The minimum wall thickness of the loop pipe and fittings are not less than that calculated using the ASME III Class 1 formula of Paragraph NB

-3641.1(3) with an allowable stress value of 17,550 psi. The pipe wall thickne ss for both bypass and pressurizer surge lines is Schedule 160.

The minimum pipe bend radius is 5 nominal pipe diameters; ovality does not exceed 8 percent.

All butt welds, branch connection nozzle welds, and boss welds are of a full penetration design.

Section 5.2.3 discusses processing a nd minimization of sensitization.

Flanges conform to ANSI B16.5.

Socket weld fittings and socket joints conform to ASNI B16.11.

Section 5.2.4 discusses inservice inspection.

5.4.3.2 Design DescriptionTable 5.4-5 gives principal design da ta for the reactor cooling piping.

Pipe and fittings are cast, for ged, or seamless without longitudinal or electroslag welds, and comply with the requirements of the ASME Code, Sec tion II, Parts A and C,Section III, and Section IX.

The RCS piping is specified in the smallest size s consistent with syst em requirements. This design philosophy results in the reactor inlet and outlet piping diameters given in Table 5.4-5. The MPS3 UFSAR5.4-19Rev. 30 line between the steam generator and the pump suction is larger to reduce pressure drop and improve flow conditions to the pump suction.

The reactor coolant piping and fittings which make up the loops are austenitic stainless steel.

There is no electroslag welding on these component

s. All smaller piping which comprise part of the RCS such as the pressurizer surge line, spray and relief line, loop drai ns and connecting lines to other systems are also austen itic stainless steel.

The nitrogen supply line for the pressurizer relief tank is carbon steel. All joints and connections are welded, except for the pressurizer relief and the pressurizer code safety valves, where flanged joints are used. Thermal sleeves are

installed in the crossover leg at the 2 inch ch emical volume and control system (CHS) charging line connection with each reactor coolant loop. Th e other thermal sleeves are on the pressurizer where the surge line connects and where the spray line connects to the pressurizer. Thermal sleeves are used where thermal stresses could deve lop due to rapid changes in fluid temperature during normal operational transients.

All piping connections from auxi liary systems are made above th e horizontal centerline of the reactor coolant piping, with the exception of:1.Residual heat removal pump suction lines, which are 45 degrees down from the horizontal centerline. This enables the water level in the RCS to be lowered in the reactor coolant pipe wh ile continuing to operate the residual heat removal system, should this be required for maintenance.2.Loop drain lines and the connection for temporary level measurement of water in the RCS during refueling and maintenance operation. 3.The differential pressure taps for flow measurement, which are downstream of the steam generators on the first 90 degree elbow. 4.The hot leg sample connections and the cold leg high pressure safety injection, chemical and volume control charging, pressurizer spray, reactor plant gaseous drains and instrumentation connections , which are located on the horizontal centerline.

Penetrations into the coolant flow path are limited to the following:1.The spray line inlet connections extend into the cold leg piping in the form of a scoop so that the velocity head of the re actor coolant loop flow adds to the spray driving force.2.The reactor coolant sample system taps protrude into the main stream to obtain a representative sample of the reactor coolant.3.The hot and cold narrow range, fast re sponse resistance te mperature detectors (R TDs) are located in thermowells that extend into the reactor coolant pipe.

MPS3 UFSAR5.4-20Rev. 304.The wide range hot and cold RTDs are lo cated in thermowells that extend into hot and cold legs of the reactor coolant piping.

The RCS piping includes those sections of pi ping in terconnecting the reactor vessel, steam generator, and reactor coolant pum

p. It also includes the following:1.Charging line and alternate charging line from the system isolation valve up to the branch co nnections on the reactor coolant loop. 2.Letdown line and excess letd own line from the branch connections on the reactor coolant loop to the system isolation valve. 3.Pressurizer spray lines from the reactor c oolant cold legs to the spray nozzle on the pressurizer vessel. 4.Residual heat removal lines to or from the reactor coolant loops up to the designated check valve or isolation valve. 5.Safety injection lines from the designate d check valve to the reactor coolant loops. 6.Accumulator lines from the designated ch eck valve to the reactor coolant loops. 7.Loop fill, loop drain, sample, and instru ment lines to or from the designated isolation valve to or from the reactor coolant loops.

NOTE: Lines with a 3/8 inch flow restricting orifice qualify as Safety Class 2; in the event of a break in one of these Safe ty Class 2 lines, the normal makeup system is capable of providing makeup flow while maintaining pressurizer water level. In the case of the pressurizer steam space a 0.25 inch orifice was installed to provide a class break.8.Pressurizer surge line from one reactor coolant loop hot leg to the pressurizer vessel inlet nozzle. 9.Pressurizer spray, sample connection with scoop, reactor coolant temperature RTD thermowell installation boss, and the thermowell itself (see Note under Item 7).10.All branch connection nozzles attached to reactor coolant loops. 11.Pressure relief lines from nozzles on top of the pressu rizer vessel up to and through the power operated pressurizer relief va lves and pressurizer safety valves. 12.Auxiliary spray line from the isolation va lve to the pressurizer spray line header. 13.Sample lines from pressurizer to the isolation valve (see Note under Item 7).

MPS3 UFSAR5.4-21Rev. 3014.Loop stop valve bypass lines. 15.Reactor vessel head vent piping from the reactor vessel head to the pressurizer relief tank (Section 5.4.15

). Section 5.2 discusses details of the materials of construction and codes used in the fabrication of reactor coolant piping and fittings.5.4.3.3 Design Evaluation Section 3.9 discusses piping load and stress evaluation for normal operating loads, blowdown loads, and combined normal, blowdown and seismic loads.

5.4.3.3.1 Material Corrosion/Erosion EvaluationThe water chemistry is selected to minimize corrosion.

Periodic analysis of the coolant chemical composition is performed to verify that the reactor coolant quality meets the specifications.

The design and construction are in compliance with ASME Section XI. Pursuant to this, all pressure containing welds out to the second valve that delineates the RCS boundary are available for examination with removal insulation.Components constructed with stainless steel operate satisfactorily under normal plant chemistry conditions in pressurized water reactor systems, because chlorides, fluorides, and particularly oxygen, are controlled to very low levels. (Section 5.2.3)Periodic analysis of the coolant chemical composition is performed to m onitor the adherence of the system to desired reactor coolant water quality listed in Ta ble 5.2-4. Maintenance of the water quality to minimize corrosion is accomplished using the chemical and volume control system and sampling system which are described in Chapter 9.

5.4.3.3.2 Sensitized Stainless Steel Section 5.2.3 discusses sens itized stainless steel.

5.4.3.3.3 Contaminant Control Contamination of stainless steel and Inconel by copper , low me lting temperature alloys, mercury and lead is prohibited. Prior to application of thermal insulation, the austenitic stainless steel surfaces are cleaned and analyzed in accordance with Regulatory Guide 1.37 as described in Section 1.8.

MPS3 UFSAR5.4-22Rev. 30 5.4.3.4 Tests and InspectionsTable 5.4-6 gives the RCS piping NDE program.

V olumetric examination is performed throughout 100 percent of the wall volum e of each pipe and fitting in accordance with the applicable requirements of Section III of the ASME Code for all pipe 27.5 inches and larger. All unacceptable de fects are eliminated in accordance with the requirements of the same section of the code.

A liquid penetrant examinat ion is performed on both the entire out side and inside surfaces of each finished fitting in accordance with the criteria of ASME Section III. Acceptan ce standards are in accordance with the applicable re quirements of ASME Section III.The pressurizer surge line conforms to SA-376 Grade 304, with supplementary requirements S2 (transverse tension tests), and S6 (ultrasonic test

). The S2 requirement applies to each length of pipe. The S6 requirement applies to 100 percen t of the piping wall volume. The material is examined in accordance with ASME Code,Section III and ASME Section II, SA 655, 1977 edition. The end of pipe sections, branch ends a nd fittings are machined back to provide a smooth weld transition adjacent to the weld path.

5.4.4 MAIN STEAM LINE FLOW RESTRICTOR 5.4.4.1 Design Basis The outlet nozzle of the st eam generator contains a flow restrictor designed to limit steam flow in the unlikely event of a break in the main steam li ne. With a restrictor, a large increase in steam flow creates a backpressure whic h limits further increase in flow. Several protective advantages are thereby provided: rapid rise in containment pr essure is prevented, the rate of heat removal from the reactor coolant is maintained within acceptable limits, thrust forces on the main steam line piping are reduced, and stresses on internal st eam generator components, particularly the tube sheet and tubes, are maintained within acceptable limits. Another de sign objective is to minimize waterhammer type loads and unrecovered pressure loss across the restrictor during normal operation.

5.4.4.2 Design DescriptionThe flow restrictor consists of seven Inconel (ASME SB-163) vent uri inserts which are inserted into the holes in an integral steam outlet low al loy steel forging. The inserts are arranged with one venturi at the centerline of the outlet nozzle and the other six equall y spaced around it. After insertion into the low allow steel forging holes, the Inconel venturi nozzles are welded to the Inconel cladding on the inner surface of the forging.

5.4.4.3 Design Evaluation The flow restrictor design has been sufficiently analyzed to assure its structural adequacy. The equivalent throat diameter of the steam generato r outlet is 16 inches, and the resultant pressure MPS3 UFSAR5.4-23Rev. 30 drop through the restrictor at 100 percent steam flow is approximately 3.1 psi. This is based on a design flow rate of 4.07 x 10 6 lb/hr. Materials of constructi on and manufacturing of the flow restrictor are in accordance with Section III of the ASME Code.

5.4.4.4 Tests and Inspections Since the restrictor is not a part of the steam system pressure boundary , no tests and inspection beyond those conducted during fabrication are performed.

5.4.5 MAIN STEAM ISOLATION SYSTEMThe main steam isolation system is described in Sections 6.2.4 and 10.3.

5.4.6 REACTOR

CORE ISOLATION COOLING SYSTEMThis section is not applicable to the Millstone 3 reactor core, as it applies to a boiling water reactor core design and Millstone 3 has a pressurized water reactor.

5.4.7 RESIDUAL

HEAT REMOVAL SYSTEMThe residual heat removal system (RHS) transfers heat from the reactor coolant system (RCS) to the component cooling system (CCP) to reduce the temperature of the reactor coolant to the cold shutdown temperature at a controlled rate during the second part of normal plant cooldown or a safety grade cold shutdown (SGCS) and maintain s this temperature until the plant is started up again. The RHS may be aligned to the RCS for cooldown operation once RHS entry conditions are achieved (RCS temperature and pressure reduced to at or below 350

°F and 375 psig, respectively).

Parts of the RHS also serve as parts of the emergency core cooling system (ECCS) during the injection phase of a loss-of-c oolant accident (Section 6.3).

The RHS may be used to transfer refueling water between the refueling cavity and the refueling water storage tank at the beginning a nd end of the refueling operations.Relief valves in the RHS pump suction lines fr om the RCS provide low temperature overpressure protection for the reactor vessel when the RHS is unisolated from the RCS (Section 5.2.2.11).

Nuclear plants employing the same RHS design as the Millstone 3 Steam Electric Station are given in Section 1.3.

5.4.7.1 Design BasesRHS design parameters are listed in Table 5.4-7.

The RHS is designed to operate in conjunction with other plant sy stems to reduce the temperature of the RCS during the sec ond phase of plant cooldown.

MPS3 UFSAR5.4-24Rev. 30 The RHS is capable of being pl aced in operation approximately four hours after reactor shutdown when the temperature and pressure of the RCS are approximately 350

°F and 375 psig, respectively. Assuming that two h eat exchangers and two pumps are in service and that each heat exchanger is supplied with component cooling wate r at design flow and temperature, the RHS is designed to reduce the temperature of the reactor coolant from 350

°F to 200°F within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. However, during normal cooldown only one RHS heat exchanger and pump are used for cooling until the reactor coolant te mperature is reduced to 260

°F. This limitation is imposed based on the Technical Specifications requirement of 1 RHS train being operable in Mode 4 to mitigate a LOCA event and the issue raised by the West inghouse Owners Group that flashing in the RHS suction line would occur due to th e elevated temperature of the wate r trapped in the suction line in conjunction with the rapid depressurization upon RHS pump start. Consequently, one RHS train remains aligned to the RWST for use as an injection path to the RCS until the RCS temperature

corresponds to the saturation temperature of the RWST elevat ion head above the RHR pump suction which would prevent flashing. Assuming this configuration, the RHS is capable of cooling the reactor coolant from 350

°F to 200°F within 41.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> with one reactor coolant pump operating and within 72.25 hou rs with 2 reactor coolant pum ps operating. The heat load by the RHS during the cooldown transient includes residual and deca y heat from the core and reactor coolant pump heat.

The RHS is also designed to opera te in conjunction with the othe r systems of the cold shutdown design to achieve and maintain cold shutdown us ing only safety grade systems, as required by Branch Technical Position RSB 5-1. See Safety Grade Cold Shutdown (one and/or two RHS trains(s) in service), Section 5.4.7.2.3.5.

The RHS is designed to be isolated from the RCS whenever the RCS pressure exceeds the RHS design pressure. The RHS is isolated from the RC S on the suction side by three normally closed, motor-operated valves in series on each suction line. Two of the motor-operated valves are interlocked to prevent its opening if RCS pressu re is greater than 412.5 psia and alarm in the control room if RCS pressure exceeds 440 psig and the valve is open. If the plant is in Mode 1, 2, or 3, the operator is required to close all three suction valves. If the plant is in mode 4, 5, or 6 and the RCS pressure increases to 750 psig, the operator is required to close the motor-operated valve closest to the pump. (These inte rlocks are discussed in detail in Sections 5.4.7.2.4 and 7.6.2.) The third motor-operated valve is closed and deener gized at the motor control center (MCC). The motor-operated valve closest to the pump suction is closed and deenergized at the MCC.

The RHS is isolated from the RCS on the discharge side by three chec k valves in each return line. Also provided on the discharge side is a normally open motor-opera ted valve downstr eam of each RHS heat exchanger.Each inlet line to the RHS is equi pped with a pressure relief valve sized to relieve the flow of one charging pump at the relief valve set pressure. These relief valves are provided to protect the RHS system (and the reactor pressure vessel when the RHS is unisolated from the RCS) from inadvertent overpressurization dur ing plant cooldown or startup.Each discharge line from the RH S to the RCS is equipped with a pressure relief va lve designed to relieve the maximum possible back leakage thr ough the valves isolating the RHS from the RCS.

MPS3 UFSAR5.4-25Rev. 30 The RHS is designed for a single nuclear power uni t and is not shared am ong nuclear power units.

The RHS is designed to be fully operable from the control room for normal operation except for opening the outermost and inner most pump suction valve in each train. These valves are closed and deenergized at the MCC. The MCCs for the i nnermost valves are located in the ESF building on the 36 foot elevation. The MCCs for the outermost valves are located in the auxiliary building in the vicinity of the rod drive control center. These MCCs are accessible should RHS operability be required after an accident (FSAR Table 12.3-3). Manual operations required of the operator are: closing the suction valves to the RWST, opening the suction isolation valves, positioning the flow control valves downstream of the RHS heat exchangers, and starting the RHS pumps.

Manual actions, including those required for safety grade cold shutdown, are also discussed in Sections 5.4.7.2.3.5, 5.4.7.2.6 and 5.4.7.2.7. By nature of its redundant two tr ain design, the RHS is designed to accept all major co mponent single failures with the only effect being an extension in the required cooldown time. There are no motor-operated valves in the RHS that are subject to common mode flooding. Provisions to protect th e equipment from flooding are discussed in Section 3.4. For two low probability electrical system single failures , i.e., failure in the suction isolation valve interlock circuitry, or emergency generator failure in conjunction with loss of off site power, limited operator action outside the control room is required to open the suction isolation valves. The spurious operation of a single RHS motor-operated valve can be accepted without loss of cooling function as a result of the redundant two train design. Missile protection, protection against dynamic effects associated with the postu lated rupture of piping, and seismic design are discussed in Sections 3.5, 3.6, and 3.7, respectively.

5.4.7.2 System Design 5.4.7.2.1 Schematic Piping and Instrumentation Diagrams The RHS, as shown on Figures 5.4-5 and 5.4-6, cons ists of two residual heat exchangers, two residual heat removal pumps, and the associated piping, valves, and instrumentation necessary for operational control. The suction lines to the RHS are connected to the hot legs of two reactor coolant loops, while the return lines are connected to the cold legs of each of the reactor coolant loops.These return lines are also the ECCS low head injection lines (Figure 6.3-1).The RHS suction lines are isolated from the RCS by three normally- closed motor-operated valves in series. The two normally closed isolation valves inside containment in each RHS suction line receive power from the same Class 1E sour ce as the RHS pump in that line while the valve outside containment is powered by the opposite train. This arrangeme nt ensures that single failure requirements for RHS accessibility and isolation are met. Each discharge line is isolated from the RCS by three check valves located inside the containment and by a normally open motor-operated valve located outside the containment. (The check valves and the motor-operated valve on each discharge line are not part of the RHS; these valv es are shown as part of the ECCS; Figure 6.3-1.)

MPS3 UFSAR5.4-26Rev. 30 During RHS operation, reactor coolant flows from the RCS to the residual heat removal pumps, through the tube side of the residua l heat exchangers, and back to the RCS. The heat is transferred to the component cooling water circulating through th e shell side of the residual heat exchangers.

Coincident with operation of the RHS, a portion of the reactor coolant flow may be diverted from downstream of the residual heat exchangers to the CHS low pressure letdown line for cleanup and/or pressure control. By regulating the diverted flow rate and the charging flow, the RCS pressure may be controlled.

Pressure regulation is necessary to maintain the pressure range dictated by the fracture prevention criteria requirements of the reac tor vessel and by the shaft seal differential pressure and net positive suction head requirement s of the reactor coolant pumps.

The RCS cooldown rate is manually controlled by regulating the reactor coolant flow through the tube side of the residual heat exchangers. A line containing a fl ow control valve bypasses each residual heat exchanger and is used to maintain a constant return flow to the RCS. Instrumentation is provided to monitor system pressure, temperature, and total flow.

The RHS may be used for filling the refueling cavity before refueling. After refueling operations, water is pumped back to the refueling water stor age tank until the water level is brought down to the flange of the reactor vessel. The remainder of the water is removed via a drain connection at the bottom of the refueling canal.

When the RHS is in operation, the water chemistry is the same as that of the reactor coolant.

Provision is made for the pro cess sampling system (Section 9.3.2) to extract samples from the flow of reactor coolant downstream of the residual heat exchangers. A local sampling point is also provided on each residual heat removal train, between the pump and heat exchanger.

The RHS functions in conjunction wi th the high head portion of the ECCS to provide injection of borated water from the refueling water storage ta nk into the RCS cold legs during the injection phase following a loss-of-coolant accident.

Long term recirculation is performed by the containment recirc ulation system discussed in Sections 6.2.2 and 6.3.

The use of the RHS as part of the ECCS is more completely described in Section 6.3.

Description of Com ponent Interlocks:

The RHS pumps, in order to perform their ECCS function, are interlocked to start automatically on receipt of a safety injection signal (Section 6.3

).Two of the RHS suction isolation valves in each inlet line from the RCS are separately interlocked to prevent their being opened when RCS pressure is greater than 412.5 psia. In addition, an alarm will annunciate in the c ontrol room if RCS pressure exceeds 440 psig and the valve is open. If the plant is in Mode 1, 2, or 3, the operator is required to close all MPS3 UFSAR5.4-27Rev. 30 three suction valves. If the plant is in mode 4, 5, or 6 and the RCS pressure increases to 750 psig, the operator is required to close th e motor-operated valve closest to the pump.

These interlocks are described in more detail in Sections 5.4.7.2.4 and 7.6.2. It should be noted that these valves can also be controlled from the Auxiliary Shutdown Panel (ASP).

Valve 8701A is not interlocked with RCS pressure lo w to open to provide one train of RHR cooling when the control room is inaccessible. The innermost and outermost RHS suction isolation valves in each inlet line are closed and deenergized at the MCCs.

The RHS suction isolation valves from the RCS are als o interl ocked to prevent their being opened unless the isolation valves in the following lines are closed:

1.Recirculation line from the residual heat exchanger outlet to th e suction of the high head safety injection pumps2.Residual heat removal pump suction line from the refueling water storage tankThe motor-operated valves in the RHS mini-flow bypass lines are interlocked to open when the residual heat removal pump dischar g e flow is less than approximately 772 gpm and close when the flow ex ceeds approximately 1,542 gpm.The motor-operated isolation valves in the recirculation lines fr om the residual heat exchanger outlet to the suctions of the high head safety injection pumps are interlocked such that they cannot be opened unless either of the series RHS suct ion isolation valves from the RCS in the corresponding subsystem is closed. A high CCP temperature interlock will signal the RHS heat exchanger bypass valve to open. This interlock is in ef fect when the Control Room main board (MB2) "Normal-Cooldown" switch is in the Cooldown position. See also, FSAR Section 9.2.2.1.5

.5.4.7.2.2 Equipment and Component DescriptionsThe materials used to fabricate RHS components are in accordance with the applicable code requirements. All parts of compone nts in contact with borated water are fabricated or clad with austenitic stainless steel or e quivalent corrosion resistant mate rial. Component parameters are given in Table 5.4-8.

Residual Heat Removal Pumps Two pumps are installed in the RH S. The pumps are sized to deli ver reactor coolant flow through the residual heat exchangers to meet the plant cooldown requirements. Th e use of two separate RHR trains assures that cooling capacity is only partially lost should one pump become inoperative.

The RHS pumps are protected fr om overheating and loss of su ction flow by mini-flow bypass lines that assure flow to the pump suction s hould the pump suction be isolated or the RCS pressure be above the shutof f head of the pum

p. A valve located in e ach mini-flow line is regulated by a signal from the flow transmitters located in each pump discharge header. The MPS3 UFSAR5.4-28Rev. 30 control valves open when the RHS pump discharge flow is less th an 772 gpm and close when the flow exceeds 1,542 gpm.A pressure sensor in each pump discharge header provides a signal for an indicator in the control room. A high pressure alarm is also actuated by the pressure sensor.The two pumps are vertical, centrifugal units with mechanical seals on the shafts. All pump surfaces in contact with reactor coolant are austenitic stainless steel or equivalent corrosion resistant material.

The RHS pumps also function as the low h ead safety injection pumps in the ECCS.

(See Section 6.3 for further information and for the RHS pump performance curves.)

Residual Heat Exchangers Two residual heat exchangers are installed in th e system. The heat exchanger design is based on heat load and temperature differences between reactor coolant and component cooling water existing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown when the temperature difference between the two systems is small.

The installation of two heat exchangers in separa te and independent residual heat removal trains assures that the heat removal capacity of the syst em is only partially lost if one train becomes inoperative.The residual heat exchangers are of the shell and U-tube type. Reactor c oolant circulates through the tubes, while component cooli ng water circulates through the shell. The tubes are welded to the tube sheet to prevent leakage of reactor coolant.The residual heat exchangers also function as part of the ECCS (Section 6.3).Residual Heat Removal System Valves Valves that perform a modulating function are equipped with two sets of packings and an intermediate leakoff connection that discharges to the drain header.Manual and motor-operated valves have backseats to facilitate repacking and to limit stem leakage when the valves are open. Motor-operate d valves are stopped in the open direction by limit switches and therefore must be back seated manually. Leakage connections are provided where required by valve size and fluid conditions.

The RHS heat exchanger outlet butterfly valves have been provide d with actuator throttle limiters that have been set to prevent full opening of the valves in the event of a loss of the (non safety) Instrument Air. The RHS heat exchanger bypass butterfly valves have b een modified to fail open in the event of a loss of Instrument Air. Upon loss of air, the outlet valves will fail open to the pre-set open position and the bypass va lves will fail full open to allow continued cooldown without adversely affecting CCP piping with an RCS temperature as high as 350

°F.

MPS3 UFSAR5.4-29Rev. 30 5.4.7.2.3 System Operation 5.4.7.2.3.1 Reactor StartupGenerally, while at cold shutdown condition, decay heat from the reactor core is being removed by the RHS. The number of pumps and heat exchan gers in service depends upon the heat load at the time.At initiation of the plant startup, the RCS is co mpletely filled, and the pressurizer heaters are energized. The RHS is operating and is connected to the CHS via the low pressure letdown line to control reactor coolant pressure. During this time, the RHS acts as an alternate letdown path. The manual valves downstream of the residual heat exchangers leading to the letdown line of the CHS are opened. The control valve in the line from th e RHS to the letdown line of the CHS is then manually adjusted in the control room to permit letdown flow.Steam bubble formation in the pressurizer is accomplished by incr easing the letdown flow above the charging flow with the pressurizer heaters energized. The reactor coolant pumps are normally started to heat up the system after the pressuri zer bubble has been forme

d. When the pressurizer water level reaches the no-load pr ogrammed setpoint, pressurizer le vel control is shifted to the normal operational means. The RHS is then isolat ed from the RCS and the system pressure is controlled by normal letdown, pressurizer spray, and pressurizer heaters.

5.4.7.2.3.2 Power Generation and Hot Standby Operation During power generation and hot st andby operation, the RHS is not in service but is aligned for operation as part of the ECCS.

5.4.7.2.3.3 Plant Shutdown Plant shutdown is defined as th e operation which brings the plan t from no-load temperature and pressure to a cold shutdown condition (i.e., to a subcritical condition with the reactor coolant temperature no greater than 200

°F).5.4.7.2.3.4 Normal Cold ShutdownThe initial phase of a normal plant shutdown is accomplished by transferring heat from the RCS to the steam and power conversi on system. Circulation of the r eactor coolant is provided by the reactor coolant pumps and heat removal is accomplished by using the steam generators and dumping steam to the condenser.

In conjunction with this portion of the cool down, the reactor coolant is borated to the concentration required for cold shutdown and depressurized to a pressure permitting RHS operation. Boration and makeup for the contraction of the RCS due to cooling are performed using the charging, letdown, and ma keup control portions of the CHS.

MPS3 UFSAR5.4-30Rev. 30 The depressurization func tion is performed by initiating pressurizer spray from the discharge of the operating reactor coolant pump.When the reactor coolant temperature and pressure are reduced to at or below 350

°F and 375 psig, no less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> afte r reactor shutdown, the second phase of cooldown starts with the RHS being placed in operation.

Startup of the RHS includes a wa rmup period during which time re actor coolant flow through the heat exchangers is limited to minimize thermal shock. The rate of heat removal from the reactor coolant is manually controlled by regulating the coolant flow through the residual heat exchangers. By adjusting the control valves dow nstream of the residual heat exchangers, the mixed mean temperature of the return flow is c ontrolled. Coincident with the manual adjustment, each heat exchanger bypass valve is automatically regulated to give the required total flow.The reactor cooldown rate is limited by RCS equipment cooling rates based on allowable stress limits, as well as the operating temperature lim its of the component coo ling water system. As the reactor coolant temperature decreases, the reactor coolant flow through the residual heat exchanger is increased by adjusting the control valve in each heat exchanger's tube side outlet line.Modifications to the RHS system have been made to preclude overheating of the RHS heat exchanger (shell side) c ooling water piping (CCP system) in the event of a loss of Instrument Air during a Normal or safety grade cold shutdown (SGCS) cooldown. The RHS heat exchanger outlet butterfly valves have been provided with actuator throttle limiters that have been set to prevent full opening of the valves in the event of a loss of the (non safety) Instrument Air. The RHS heat exchanger bypass butterfly valves have been modified to fail open in the event of a loss of Instrument Air. Upon loss of air, the outlet valves will fail open to the pre-se t open position and the bypass valves will fail full open to allow continued cooldown without adversely affecting CCP piping with an RCS temperature as high as 350

°F. The changes have no effect on the RHS injection flowpath when RHS is used during the SI phase following a LOCA. See also FSAR Section 6.3.2.2.5.

During plant shutdown with the RHS in operation, operation with a steam bubble in the pressurizer is maximized to provide RCS pressu re control. The RCS is augmented by regulating the charging flow rate and the rate of letdown from the RHS to the CHS.

After the reactor coolant is reduced below a temperature of 160

°F and the reactor coolant pump is stopped, cooling of the pressuri zer is continued by providing auxiliary spray from the CHS.After the reactor coolant pressure is reduced and the temperature is 140

°F or lower, the RCS may be opened for refueling or maintenance.

5.4.7.2.3.5 Safety Grade Cold ShutdownWhile the plant shutdown basis is hot standby for those events involving a primary or secondary system piping passive failure, it is cold shutdown for those events which are initiated from normal MPS3 UFSAR5.4-31Rev. 30 operating conditions. In accordance with the functional requirements of Branch Technical Position RSB 5-1, safety grade cold shutdown is defined as the ability to take the plant from normal operating conditions to cold shutdown, with or without off site power, with the most limiting single failure, using only safety relate d equipment and limited action outside of the control room, and within a reasonabl e period of time following shutdown.

Should portions of the normal shut down systems be unavailable, th e operator should maintain the plant in a hot standby condition wh ile making the normal systems functional. However, for cases in which the Demineralized Water Storage Tank (DWST) is the exclusive s ource of demineralized water, the operator should use any of the normal systems available in c onjunction with safety grade backups for those systems which cannot be made available in order to ensure cold shutdown can be achieved without depleting the DWST. The safety grade provisions are to be used only upon the inability to make available the equipment normally used for the given function.In the extreme case where all of the normal shutdown systems are unavailable, a safety grade cold shutdown would be accomplished. For Millstone 3, this is postulated to occur as a resu lt of a safe shutdown earthquake (SSE), coincident with a loss of off site power, and the loss of one RHS train due to the loss of one vital bus as the most limiting single failure. Under these circumstances:1.Circulation of reactor coolant is acco mplished by natural ci rculation until RHS cooling is initiated. (The reactor c oolant pumps are assumed to be stopped.)2.Heat removal is accomplished with th e steam generators and water from the DWST via the auxiliary feedwater syst em, and steaming through the main steam safety or steam generator atmo s pheric relief bypass valves.3.Makeup/boration is accomplis hed with the char ging pumps.4.Letdown is accomplished via the reactor vessel head vent system to the pressurizer relief tank.5.RCS depressurization is accomplished using the pressurizer power-operated relief valves.6.Cooldown continues until RHS entry conditions are achieved, at which time one RHS train is placed in service. Cool down by steaming through the atmospheric relief bypass valves would continue in parallel with RHS cooling (concurrent steaming) only until such time that RHS could independently remove the required decay and sensible heat from the RCS.A safety grade cold shutdown would be implemented in three phases:1.Boration: Borated water from the boric acid storage system is added to the RCS in order to maintain a constant shutdow n margin at lower reactor coolant temperatures. Auxiliary feedwater drawn from the DWST is used to remove decay MPS3 UFSAR5.4-32Rev. 30 heat from the RCS and is released as st eam through the main st eam safety valves.

The plant is maintained at hot standby for a maximum of six hours in order to complete this boration phase. For de tails on safety grade boration, see Section 9.3.4.2.6.2.Steam Generator Cooling:

Boration is terminated. Auxi liary feedwater drawn from the DWST is used to reduce the RCS temp erature to RHS entr y conditions and is released as steam through at least two steam generator atmospheric relief bypass valves. The plant is cooled to RHS entry conditions within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from termination of boration phase.3.RHS System Cooling: Auxiliary feedwa ter cooling is terminated once the RHS can cool the plant independently. The RCS temperature is reduced to cold shutdown conditions via the RHS System.Depressurization of the RCS throughout the safety grade cold shutdown process is accomplished

by the safety grade solenoid-operate d pressurizer PORVs. In order to ensure that the RCS is not repressurized by the safety injection accumulators, the motor-operated accumulator isolation valves are closed prior to RCS pressure dropping below the accumulator discharge pressure. Two of the accumulator isolation valves are powered from the orange safety train while the other two valves are powered from the purple safety tr ain. Additional protecti on against inadvertent repressurization of the RCS by the accumulators is provided by redundant Class 1E solenoid-operated accumulator vent valves which permit venting of the accumulators in the remote event that the discharge line should fail to isolate.

The RHS is designed to operate in conjunction with the other safe ty grade systems of the cold shutdown design in order to addr ess the functional re quirements of SRP Se ction 5.4.7 and Branch Technical Position RSB 5-1. The SRP requires that plant safety systems have the capacity to bring the reactor to conditions permitting the operation of the RHS w ithin a reasonable period of time, defined as 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, assuming a singl e failure of an active component with only either on site or off site power available. The BTP requires that the plant have the capacity to bring the reactor to a cold shutdown condition, within a reasonable pe riod of time following shutdown, assuming the most limiting failure. Therefore, the SRP, in conjunction with the BTP, re quire that the plant be capable of achieving RHS entry conditions within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of reactor trip and achieving cold shutdown within an unspecified a dditional reasonable period of time.

The Millstone 3 safety grade cold shutdown desi gn enables the nuclear steam supply to be taken from hot standby to cold shutdown conditions using only safety grade systems, with or without off site power, and with the most limiting single failure. The safety grade cold shutdown design also enables the RCS to be taken from hot standby to conditions that will permit initiation of RHS operation within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, and then to cold s hutdown within an additional 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Therefore, the Millstone 3 licensing basis is to achieve cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of reactor trip.

Per Section 14.2.7.9, the Millstone 3 instrument air sy stem is non-safety related; therefore, the safety grade cold shutdown design must be capable of achieving co ld shutdown without the use of instrument air.

MPS3 UFSAR5.4-33Rev. 30With instrument air available and a single failure, cold shutdown conditions can be achieved within 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of reactor trip.With instrument air unavailable and a single failure, li mited operator action out side of the control room is required. A loss of instrument air causes the RHS heat exchanger bypass valves to fail open, thus reducing the flow to the RHS heat exchanger. This arrangement reduces the heat removal rate in order to pr otect the reactor plant component cooling water system from overheating. In order to increase the heat removal rate as the RCS temperature decreases, operator action is required to throttle the operating heat exchanger bypass valve. This limited operator action is justified since it is required only afte r a single failure and, if initiated 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after reactor trip, will result in cold shutdown conditions within 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> of reactor trip.With instrument air unavailable and no single failure, the opera tor would sequence on the second RHS train once the RCS temperature was reduced to 260

°F, and still achieve cold shutdown conditions within 68.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Therefore, in all cases, a safety grade cold shutdown can be achieved without challenging the usable DWST inventory or overheating the reac tor plant component cooling water system, and within the licensing basis of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

5.4.7.2.3.6 Refueling Both residual heat removal pumps may be utiliz ed during refueling to pump borated water from the refueling water storage tank to the refuelin g cavity. During this operation, the residual heat removal pumps are stopped, the isol ation valves in the inlet lines of the RHS are closed, the isolation valves to the refueling water storag e tank are opened, and the residual heat removal pumps are restarted.

The reactor vessel head is lifted slightly. The refueling water is then pumped into the reactor vessel through the normal RHS retu rn lines and into the refueling cavity through the open reactor vessel. The reactor vessel head is gradually raised as the water level in the refueling cavity increases. After the water level reaches the nor mal refueling level, th e residual heat removal pumps are stopped, the inlet isolation valves ar e opened, the refueling water storage tank supply valves are closed, the residual heat removal pumps are restarted, and the re sidual heat removal is resumed.During refueling, the RHS is maintained in se rvice with the number of pumps and heat exchangers in operation as required by the heat load.

Following refueling, either the residual heat removal pumps or the spent fuel purification system pumps can be used to drain the refueling cavity.

If the residual heat removal pumps are used to drain the refueling cavity, the refueling water level is lowered to the top of the reactor vessel flange by pumping water from the RCS to the refu eling water storage tank.

The vessel head is then replaced and the normal RHS fl ow path reestablished. The rema inder of the water is removed from the refueling canal via a drain connection in the bottom of the canal as described in Section 9.1.3.2.

MPS3 UFSAR5.4-34Rev. 30 5.4.7.2.4 ControlEach inlet line to the RHS is equi pped with a pressure relief valve sized to relieve the flow of one charging pump at the relief valve set pressure. Th ese relief valves also protect the RHS system (and the reactor pressure vessel when the RHS is unisolated from the RCS) from inadvertent overpressurization duri ng plant cooldown or start up. Each valve has a relief flow capacity of 560 gpm at a set pressure of 440 psig. An analysis has been conduc ted to confirm the capability of the RHS relief valve to prevent overpressurization in the RHS. All cred ible events were examined for their potential to overpressurize the RHS. These events included nor mal operating conditions, infrequent transients, and abnorma l occurrences. The analysis conf irmed that one relief valve has the capability to keep the RHS maximum pressure within code limits.Each discharge line from the RHS to the RCS is equipped with a pressure relief valve to relieve the maximum possible back leak age through the valve separating the RHS from the RCS. Each valve has a relief flow capacity of 20 gpm at a set pressure of 600 psig.

These relief valves are located in the low pressure safety inj ection portion of the ECCS (Figure 5.4-5).The fluid discharged by the suction side relief va lves is collected in the pressurizer relief tank. The fluid discharged by the discharge side relief va lves is collected in the primary drains transfer tank (Section 9.3.3).The design of the RHS includes three motor-operated gate isolation valves in series on each inlet line between the high pressure RCS and the lower pressure RHS. They are closed during normal operation and are opened only for residual heat removal during a plant cooldown after the RCS pressure is reduced to 375 psig or lower and RCS temperature is reduced to approximately 350

°F. During a plant startup, the inlet is olation valves are shut after drawing a bubbl e in the pressurizer and prior to increasing RCS pressure above 425 psig. Two of the th ree isolation valves in each inlet line are provided with "prevent-open" inte rlocks. It should be noted that when controlling valve 8701A from the ASP, the RCS low pressure interlock is not available. This design feature allows one train of RHR cooling when the control room is inaccessible. Although spurious

opening of these two isolation valves in series is considered unlikely, the third isolation valve in each inlet train is closed and deenergized at the MCC to preven t overpressure of RHS piping. The isolation valves clos est to the pump suctions are deenergized at the MCC to prevent a fire induced spurious hot short from damaging the valve in the credited train rendering that train non-functional. The two interlocked valves in each RHS subsystem are separa tely and independently interlocked with pressure signals to prevent their being opened whenever the RCS pressure is greater than approximately 412.5 psia.The two interlocked valves in each RHS subsystem are also separately a nd independently alarmed if RCS pressure signal is 440 psig and the valve is open. If the plant is in Mode 1, 2, or 3, the operator is required to close all th ree suction valves. If the plant is in mode 4, 5, or 6 and the RCS pressure increases to 750 psig, the operator closes the motor-operated valve closest to the pump.The use of two independently powered motor-operate d valves in each of the two inlet lines, along with two independent pressure interlock signals for each function, assure s a design which meets applicable single failure criteria.

Not only more than one single failure, but also different failure MPS3 UFSAR5.4-35Rev. 30mechanisms must be postulated to defeat the f unction of preventing possi ble exposure of the RHS to normal RCS operating pressure. These productive interlock desi gns, in combination with plant operating procedures, provide diverse means of accomplishing the protective function. For further information on the instrumentation and control features refer to Section 7.6.2.The RHS inlet isolation valves are provided with red-green position indicator lights on the main control board and the auxiliary shutdown panel.

The indicator lights for the innermost RHS suction MOV are extinguished when deenergized at the MCC in MODES 1, 2 and 3.

Isolation of the low pressure RHS from the high pressure RCS is provided on the discharge side by a normally open motor-operated valve and three check valves in series. These check valves are located in the ECCS and their te sting is described in Section 6.3.4.2.

5.4.7.2.5 Applicable Codes and Classifications The entire RHS is designed as Nuclear Safety Cl ass 2, except the suction isolation valves inside containment which are Class 1. Component codes and classifications for the RHS and the other systems relied upon for safety grade cold shutdown are given in Section 3.2.

5.4.7.2.6 System Reliability ConsiderationsGeneral Design Criterion 34 requires that a system to remove residual heat be provided. The safety function of this system is to transfer fission product decay heat and other residual heat from the core at a rate sufficient to prevent fu el or pressure boundary design limits from being exceeded. Safety grade systems ar e provided in the plant design to perform this safety function. The safety grade systems which perform this function for all plant conditions, except LOCA, are:1.The RCS and steam generators, which ope rate in conjunction with the auxiliary feedwater system;2.The steam generator safety valves;3.The steam generator atmospheric relief bypass valves; 4.The residual heat removal system (RHS) which operates in conjunction with the reactor plant component coo ling water system;5.The service water system.

For LOCA conditions, the safety grade system which performs th e function of removing residual heat from the reactor core is the ECCS, which operates in conjunction with the char ging pump cooling water system, safety injection pump cool ing water system and the service water system.

The auxiliary feedwater system, al ong with the steam gene rator safety valves and steam generator atmospheric relief bypass valves, provides a completely separate, independent, and diverse means of performing the safety functio n of removing residual heat, whic h is normally performed by the MPS3 UFSAR5.4-36Rev. 30RHS system when RCS temperature is less than 350

°F. The auxiliary feedwater system is capable of performing this function for an extende d period of time following plant shutdown.In order to achieve conditions that permit initiation of RHS operation, two other functions (boration and depressurization) must be performed. The borati on function is normally provided by the CHS. Certain initiating HELB events, postulated to occur in the operating CHS pump discharge piping, when combined with a single act ive failure of the standby CHS pump to start, may lead to a loss of all charging. In addition, all charging may be lost as a result of certain

postulated fire conditions (s ee FSAR Section 9.5.1 and the FPER for SIH system performance requirements). For these conditions, the SIH pumps will provide the required RCS inventory and boration flow to achieve safe shutdown. When the reactor coolant pumps are not available, due to loss of off site power or following a manual pump trip, the depressurization function may be provided by the CHS. The normal function and inhere nt reliability of the CHS is discussed in detail in Section 9.3.4.1.

The RHS is provided with two residual heat rem oval pumps, and two residual heat removal heat exchangers arranged in two separate, independent flowpaths. To assure reliability, each residual heat removal pump is connected to a different emergency bus. Each residual heat re moval train is isolated from the RCS on the suction side by three motor-operated valves in series. Each motor-operated valve receives power via a separate motor control center, and one of the three valves in series in the same train receives power from a different emergency bus than do the other two valves and the pump. Two of the suction isolat ion valves in each RHS subsystem are also interlocked and alarmed to prevent exposure of the RHS to the normal operating pressure of the RCS (Section 5.4.7.2.4).

RHS operation for normal conditions and for major failures is accomplished from the control room with limited operator action outside the control room. The redundancy in the RHS system design provides the system with the capability to maintain its cooling function even with major single failures, such as failure of an RHS pump, valve, or heat exchanger, since the redundant train can be used for continued heat removal.

Should it be necessary to take the plant to co ld shutdown conditions using only safety grade systems, portions of the RCS (Section 5.4.15) and the ECCS (Section 6.3) are al so relied upon for boration, letdown, makeup and depressurization. These safety grade provisions would be used only upon failure of the equipment nor mally used for the given function.Boration is accomplished by using the centrifugal charging pumps to supply borated water from the boric acid tanks to the RCS via the charging bypass line or the high head safety injection lines in the ECCS. See Section 9.3.4.2 for further details.Letdown to accommodate boration a nd any other addition to the RC S inventory is provided by the reactor vessel head vent system letdown path to the pressuri zer relief tank. See Section 5.4.15.2 for further details.

Depressurization is accomplished by discharging RCS inventory via the safety grade pressurizer power-operated relief valves. Two parallel lines are provided with solenoid-actuated valves which MPS3 UFSAR5.4-37Rev. 30 can be remotely operated to relieve to the pressu rizer relief tank. The ECCS accumulators are also provided with safety grade is olation and venting capability in order to ensure that depressurization can be completed.

The pressurizer relief tank, the ve ssel head letdown valves, and th e pressurizer relief valves are described in Sections 5.4.11, 5.4.12, and 5.4.13, respectively, and are shown on Figure 5.1-1.

The systems used for boration/inventory control and for depressurization are remotely operable with either on site or off site power availabl e and assuming the most limiting single failure. A failure modes and effects analysis (FMEA) of th e portions of the RCS, ECCS, and CHS that are used for safety grade cold shutdown is included in the RHS - Cold Shutdown Operations - FMEA (Table 5.4-9). The reliability of these systems ensures that conditions permit RHS operation can

be allowed.RHS operation for normal conditions, even with a major failure is accomplished from the control room with limited operator action outside the control room. The redundancy in the RHS design provides the system with the capability to mainta in its cooling function even with a major single failure, such as failure of a residual heat re moval pump, valve, or heat exchanger or of an emergency power source, without impact on the redundant train's continued heat removal. The only effect would be an extensio n of the time required for cool down. The capability of the RHRS or safety grade cooldown is demonstrated in the RHS - Cold Shutdown Operation - FMEA (Table 5.4-9).

5.4.7.2.7 Manual Actions The RHS is designed to be fully operable from the control room for normal operation except for opening the outermost and innermost pump suction valve in each trai

n. The outermost and innermost valves are closed and deenergized at the MCC. The outermost MOVs' MCCs are located in the auxiliary building in the vicinity of the rod drive control center. The innermost MOV's MCCs are located in the ESF building on the 36 foot level. The MCCs are accessible should RHS operability be required after an accident (FSAR Table 12.3-3). Manual operations required of the operator include: opening the suction and discharge isolation valves, positioning the flow control valves downstream of the residual heat exchangers , and starting the residual heat removal pumps. If the plant is in mode 1, 2, or 3, all three of the RHR isolation valves in each flow path require manual closure upon alarm of valve open and RCS pressure greater than 440 psig. If the plant is in mode 4, 5, or 6 and the RCS pressure increases to 750 psig, the operator is required to close the motor-operated valve closest to the pump.Assuming the most limiting single failure, the RHS can still be operated with limited operator action required outside of the control room, with the only effect being an extension in the cooldown time. Manual operation consists of opening one of the suction/isolation valves, and in the event instrument air is not available, thro ttling the operating heat ex changer bypass valve to increase RHS heat exchanger flow; see Section 5.4.7.2.3.5.

MPS3 UFSAR5.4-38Rev. 30 5.4.7.3 Performance Evaluation The performance of the RHS syst em in reducing reactor coolant temperature is evaluated through the use of heat balance calculations on the RCS and CCP at stepwise intervals following the initiation of RHS operation. Heat removal th rough the RHS and CCP heat exchangers is calculated at each interval by use of standard water-to-water heat exchanger performance correlations; the resultant fluid temperatures for the RHS and CCP systems are calculated and used as input to the next interval's heat balance calculation.

Assumptions utilized in the series of heat ba lance calculations descri bing plant RHS cooldown are as follows:1.RHS operation is initiated no earlier than four hours after reactor shutdown.2.RHS operation begins at a reactor coolant hot leg temperature of 350

°F or below.3.Thermal equilibrium is maintained throughout th e RCS during the cooldown.4.Component cooling water outlet temper ature from the RHS heat exchanger is limite d to 145°F for normal and 145

°F for a safety grade cold shutdown.5.One reactor coolant pump is assumed runni ng until the coolant temperature is at 160°F for normal two-train cooldown. At this temperature, the reactor co olant pump is stopped. For safety grade cooldown with one or two trains, the reactor coolant pumps are assumed to be stopped.

5.4.7.4 Preoperational TestingPreoperational testing of the RHS is addressed in Chapter 14.

5.4.8 REACTOR

WATER CLEANUP SYSTEM This is a BWR requirement and, as such, does not apply to Millstone 3, which is a PWR plant.

5.4.9 MAIN STEAMLINES AN D FEEDWATER PIPING Main steamlines and feedwater pi ping are discussed in Sections 10.3 (main steam supply system), 10.4.7 (condensate and feedwater systems), and 10.4.9 (auxiliary feedwater system).

5.4.10 PRESSURIZER 5.4.10.1 Design BasesThe general configuration of the pressurizer is shown on Figure 5.4-8. The design data of the pressurizer are given in Table 5.4-10. Codes and material requirem ents are provided in Section 5.2.

MPS3 UFSAR5.4-39Rev. 30 The pressurizer provides a point in the RCS wh ere liquid and vapor can be maintained in equilibrium under saturated conditions for pres sure and control purposes, for steady state operations and during transients.

5.4.10.1.1 Pressurizer Surge LineThe surge line is sized to minimize the pressure drop between the RCS and the safety valves in order to obtain maximum allowable dischar ge flow from the safety valves, as necessary.The surge line and the thermal sleeves at each e nd are designed to withstand the thermal stresses resulting from volume surges of relatively hot ter or colder water which may occur during operation. The pressurizer surge line nozzle diameter is given in Table 5.4-10 and the pressurizer surge line dimensions are shown on Figure 5.1-1.

5.4.10.1.2 Pressurizer The volume of the pressurizer is equal to, or gr eater than, the minimum volume of steam, water, or total of the two which satisfies all of the following requirements:1.The combined saturated water volume and steam expansion volume is sufficient to provide the desired response to system volume changes2.The water volume is sufficient to prev ent the heaters from being uncovered during a step load increase of ten percent at full power3.The steam volume is large enough to ac commodate the sur ge resulting from 50 percent reduction of full load with auto matic reactor control and 40 percent steam dump without the water level reaching the high level reactor trip point4.The steam volume is large enough to prevent water relief through the safety valves following a loss of load with the high wate r level initiating a reactor trip, without reactor control or steam dump5.The pressurizer will not empty foll owing reactor trip and turbine trip6.The emergency core cooling signal is not activated during reactor trip and turbine trip 5.4.10.2 Design Description 5.4.10.2.1 Pressurizer Surge LineThe pressurizer surge line connects the pressurize r to one reactor hot leg providing for continuous coolant volume pressure adjustments between the RCS and the pressurizer

.

MPS3 UFSAR5.4-40Rev. 30 5.4.10.2.2 Pressurizer The pressurizer as shown on Figure 5.4-8 is a ver tical, cylindrical vessel with hemispherical top and bottom heads constructed of carbon steel, wi th austenitic stainless steel cladding on all internal surfaces exposed to the reactor coolant. A st ainless steel liner or tube may be used in lieu of cladding in some nozzles.

The sur ge line nozzle and removable electric heaters are installed in the lower pressurizer head. The heaters are removable for maintenance or replacement. A thermal sleeve is provided to minimize stresses in the surge line nozzle. A re taining screen is located above the nozzle to prevent any foreign matter from entering the RCS. Baffles in the lower section of the pressurizer prevent an insurge of cold water from flowing directly to the steam/water interface and assist mixing.

Spray line nozzles, relief and safety valve connect ions are located in the upper head of the vessel. Spray flow is modulated by automatically controlled air-operated valves. The spray valves also can be operated manually by a switch in the control room.

A small continuous spray flow is provided through a manual bypass valve around the power-operated spray valves to assure that the pressurizer liquid is homogeneous with the coolant and to prevent excessive cooling of the spray piping.During an outsurge from the pressurizer, flashing of water to steam and generating of steam by automatic actuation of th e heaters retain the pres sure above the minimum allowable limit. During an insurge from the RCS, the sp ray system, which is fed from two cold legs, condenses steam in the vessel to prevent the pressurizer pressure from reaching the setpoint of the power-operated relief valves for normal design transients. Heaters are energized on high water level during insurge to heat the subcooled surge water that en ters the pressurizer from the reactor coolant loop.

Material specifications are provided in Table 5.2-7 for the pressurizer, pressurizer relief tank, and the surge line. Design transients for the components of the RCS are discussed in Section 3.9(N).1.

Additional details on the pressurizer design cycle analysis are given in Section 5.4.10.3.5.Spray Line Temperatures Temperatures in the spray lines from the cold legs of two loops are measured and indicated.

Alarms to warn the operator from these signa ls are actuated by low sp ray water temperature. Alarm conditions indicate insuffic ient flow in the spray lines.Safety and Relief Valve Discharge Temperatures Temperatures in the pressurizer safety and relief valve dischar ge lines are measured and indicated. An increase in a discharge line temperature is an indication of leakage or relief through the associated valve.

MPS3 UFSAR5.4-41Rev. 30 5.4.10.3 Design Evaluation 5.4.10.3.1 System Pressure Whenever a steam bubble is present within the pr essurizer , RCS pressure is maintained by the pressurizer. Analyses indicate that proper cont rol of pressure is maintained for the operating conditions.A safety limit has been set to ensure that the RCS pressure does not exceed the maximum transient value allowed under the ASME Code, Secti on III, and thereby assure continued integrity of the RCS components.Evaluation of plant conditions of operation which follow indicate that this safety limit is not reached.During startup and shutdown, the rate of temperat ure change in the RCS is controlled by the operator. Heatup rate is controlled by pump energy and by the pressurizer electrical heating capacity. This heatup rate takes into account the continuous spray flow provided to the pressurizer. When the reactor core is shutdown, the heaters are deenergized.

When the pressurizer is filled with water, i.e., during initial syst em heatup, and near the end of the second phase of plant cooldown, RCS pressure is maintained by the letdown flow rate via the Residual Heat Removal System.

5.4.10.3.2 Pressurizer PerformanceThe normal operating water volume at full load conditions is a percentage of the free internal vessel volume. Under part load conditions, the water volume in the vessel is reduced for proportional reductions in plant load. The various plant operating transients are analyzed and the

design pressure is not exceeded with the pressu rizer design parameters as given in Table 5.4-10.

5.4.10.3.3 Pressure Setpoints The RCS design and operating pressure together wi th the safety , power relief and pressurizer spray valves setpoints, and the protection system setpoint pressures are listed in Table 5.4-11. The design pressure allows for operating transient pressure changes. The selected design margin considers core thermal lag, coolant transport times and pressure drops, instrumentation and control response characteristics, and system relief valve characteristics.

5.4.10.3.4 Pressurizer SprayTwo separate, automatically controlled spray valv es with remote manual overrides are used to initiate pressurizer spray. In parallel with each spray valve is a manual throttle valve which permits a small continuous flow through both spra y lines to reduce thermal stresses and thermal shock when the spray valves open, and to help maintain uniform water chemistry and temperature in the pressurizer. Temperature sens ors with low alarms are provided in each spray line to alert the MPS3 UFSAR5.4-42Rev. 30operator to insufficient bypass flow. The layout of the common spray line piping to the pressurizer forms a water seal which prevents the steam buildup back to the c ontrol valves. The spray rate is selected to prevent the pressurizer pressure from reaching the operating setpoint of the power relief valves during a step reduction in pow er level of ten percent of full load.The pressurizer spray lines and valves are larg e enough to provide adequa te spray using as the driving force the differential pressure between the surge line c onnection in the hot leg and the spray line connection in the cold leg. The spray line inlet connections ex tend into the cold leg piping in the form of a scoop so that the velocity head of the reactor coolan t loop flow adds to the spray driving force. The spray valves and spray line connections are arranged so that the spray will operate when one reactor cool ant pump is not operating. The line ma y also be used to assist in equalizing the boron concentration between the reactor coolant loops and the pressurizer.

A flow path from the chemical and volume control system to the pressurizer spray line is also provided. This additional facility provides auxiliary spray to the vapor space of the pressurizer during cooldown when the reactor coolant pumps are not operating. The thermal sleeves on the pressurizer spray connection and the spray piping are designed to withstand the thermal stresses resulting from the introduction of cold spray water.

5.4.10.3.5 Pressurizer Design Analysis The occurrences for pressurizer design cy cle analysis are defined as follows:1.The temperature in the pressurizer vess el is always, for design purposes, assumed to equal saturation temperature for th e existing RCS pressure, except in the pressurizer steam space subsequent to a pressure increase. In this case the temperature of the steam space will exceed the saturation temperature since an isentropic compression of the steam is assumed.

The only exceptions of the a bove occur when the pressuri zer is filled water solid during plant startup and cooldown or poten tially during transients, such as an Inadvertent ECCS Actuation, CVCS ma lfunction or a feedwater line break.2.The temperature shock on the spray nozzle is assumed to equal the temperature of the nozzle minus the co ld leg temperature and the temperature shock on the surge nozzle is assumed to equal the pressurizer water space temper ature minus the hot leg temperature.3.Pressurizer spray is assumed to be initia ted instantaneously to its design flow rate as soon as the RCS pressu rizer pressure increases above 2260 psig. Spray is assumed to be terminated as soon as the RCS pressure falls below 2260 unless otherwise noted.4.Consistent with 3 above, unless otherwis e noted, pressurizer spray is assumed to be initiated once per occurrence of each transient condition. Th e pressurizer sur ge MPS3 UFSAR5.4-43Rev. 30 nozzle is also assumed to be subject to one temperature transient per transient condition, unless otherwise noted.5.At the end of each upset condition transient, the RCS is assumed to return to a no-load condition with pressure and temper ature changes contro lled within normal limits.6.Temperature changes occurring as a result of pressurizer spray are assumed to be instantaneous. Temperature changes occurring on the surge nozzle are also assumed to be instantaneous.7.Whenever spray is initiated in the pr essurizer, the pressurizer water level is assumed to be at the no load level.

5.4.10.3.6 Natural Circulation Following Loss of Off Site Power One bank of pressurizer backup heaters (manua lly connected to an emer gency power source within 60 minutes) is sufficient to maintain natural circulation following a loss of off site power.

5.4.10.4 Inspection and Testing Requirements The pressurizer is designed and constructed in accordance with ASME Code Section III.

T o implement the requirements of ASME Code Section XI the following welds are designed and constructed to present a smooth transition surface between the parent metal and the weld metal.

The path is ground smooth for ultrasonic inspection.1.Support skirt to the pressurizer lower head2.Surge nozzle to the lower head 3.Safety, relief, and spray nozzles to the upper head4.Nozzle to safe end attachment welds5.All girth and longitudina l full penetration welds 6.Manway attachment welds The liner within the safe end nozzle region extends beyond the weld region to maintain a uniform geometry for ultrasonic inspection.Peripheral support rings are furnished for the removable insulation modules.

The pressurizer quality assurance program is given in Table 5.4-12.

MPS3 UFSAR5.4-44Rev. 30 5.4.10.5 Instrumentation RequirementsRefer to Chapter 7 for details of the instrumentation associated with pressurizer pressure, level, and temperature.5.4.11 PRESSURIZER RELIEF DISCHARGE SYSTEMThe pressurizer relief discharge system collects , cools and directs for processing the steam and water dischar ged from the various safety and relief valves in the containment. The system consists of the pressurizer relief tank, the safety and relief valve discharge piping, the relief tank spray header and associated piping, and the tank nitrogen supply, the vent to containment and the drain to the reactor plant gaseous drains. Table 5.4-14 shows these valves with reference to their FSAR figures.5.4.11.1 Design Basis Codes and materials of the pressu rizer relief tank Figure 5.4-7 and associated piping are given in Section 5.2. Design data for the tank are given in Table 5.4-13.The system design is based on the requirement to absorb a discharge of steam equivalent to 110 percent of the full power pressurizer steam volume. The steam volume requirement is approximately that which would be experienced if the plant were to suffer a complete loss of load accompanied by a turbine trip but wi thout the resulting reactor trip.

The minimum volume of water in the pressurizer relief tank is determined by the energy content of the steam to be condensed and cooled, by the assumed initial temperature of the water, and by the desired final temperature of the water volume. The initial water temperature is assumed to be 120°F, which corresponds to the design maximum expected containment temperature for normal conditions. Provision is made to permit cooling the tank should the water temperature rise above 120°F during plant operation. The de sign final temperature is 200

°F, which allows the contents of the tank to be drained directly to the reacto r plant gaseous drains system without cooling.The vessel saddle supports and anchor bolt arrangement are designe d to withstand the loadings resulting from a combination of nozzle loadings acting simultaneously with the vessel dead weight loadings.5.4.11.2 System Description The piping and instrumentation diagram for the pr essurizer relief dischar ge system is given on Figure 5.1-1.The steam and water discharged from the various safety and relief valves inside containment is routed to the pressurizer relief tank if the discharg ed fluid is of reactor grade quality. Table 5.4-14 provides an itemized list of valves discharging to the tank together with references of the corresponding piping and instrumentation diagrams.

MPS3 UFSAR5.4-45Rev. 30 The tank normally contains water and a predomin antly nitrogen atmosphere. In order to obtain effective condensing and cooling of the discharged steam, the tank is installed horizontally and the steam is discharged through a sparger pipe located near the bottom, under the water level. The sparger holes are designed to insure a resultant steam velocity close to sonic.The tank is also equipped with an internal spray and a drain which are used to cool the water following a discharge. Cold wate r is drawn from the primary grad e water system, and the content of the tank is drained to the reac tor plant gaseous drains system.

The nitrogen gas blanket is used to control the at mosphere in the tank and to allow room for the expansion of the original water plus the condensed steam discharge. The tank gas volume is calculated using a final pressu re based on an arbitrary design pressure of 100 psig. The design discharge raises the worst case initial conditio ns to 50 psig, a pressure low enough to prevent fatigue of the rupture disks. Prov ision is made to permit the gas in the tank to be periodically analyzed to monitor the concen tration of hydrogen and/or oxygen.The contents of the vessel are drained to the reactor plant gaseous drains system.5.4.11.2.1 Pressurizer Relief Tank The general configuration of th e pressurizer relief tank is shown on Figure 5.4-7. The tank is a horizontal, cylindrical vessel with elliptical dished heads. The ve ssel is constructed of austenitic stainless steel and is overpressure protected in accordanc e with ASME Code Section VIII, Division 1, by means of two safety head s with stainless steel rupture discs.

A flanged nozzle is provided on the tank for the pressurizer dischar ge line connection to the sparger pipe. The tank is also equipped with an in ternal spray connected to a cold water inlet and with a bottom drain, which are used to cool the tank following a discharge.5.4.11.3 Safety EvaluationThe pressurizer relief discharge system does not co nstitute part of the r eactor coolant pressure boundary per 10 CFR 50, Section 50.2, si nce all of its components ar e downstream of the reactor coolant system safety and relief valves. Thus, General Design Criteria 14 and 15 are not applicable. Furthermore, complete failure of the auxiliary systems serving the pressurizer relief tank does not impair the capability for safe plant shutdown.

The design of the system piping layout and piping restraints is consistent with Regulatory Guide 1.46. Compliance to Regulatory Guide 1.46 by restraining the safety and relief valve discharge piping so that integrity and operability of the valves are maintained in the event of a rupture.

Regulatory Guide 1.67 is not appl icable since the system is not an open discharge system.The pressurizer relief discharge system is capable of handling the design discharge of steam without exceeding the design pressure and temperature. The volume of water in the pressurizer relief tank is capable of absorbing the heat from the assumed discharge maintaining the water MPS3 UFSAR5.4-46Rev. 30 temperature below 200

°F. If a discharge exceeding the design basis should occur, the relief device on the tank would pass the discharge th rough the tank to the containment sumps.

The rupture discs on the relief tank have a relief capacity equal to or greater than the combined capacity of the pressurizer safety valves. The tank design pressure is twice the calculated pressure resulting from the design basis safety valve discharge describe d in Section 5.4.1 1.1. The tank and rupture discs holders are also designed for full vacuum to prevent tank collapse if the contents cool following a discharge without nitrogen being added.The discharge piping from the safety and relief valves to the relief tank is sufficiently large to prevent backpressure at the safety valves from exceeding 20 percent of the setpoint pressure at full flow.5.4.11.4 Instrumentation RequirementsThe pressurizer relief tank pressure transmitter provides an indication of pressure relief tank pressure. An alarm is provided to indicate high tank pressure.

The pressurizer relief tank level transmitter suppl ies a signal for an indi cator with high and low level alarms.

The temperature of the water in the pressu rizer relief tank is indicated, and an alarm actuated by high temperature informs the operator that c ooling of the tank contents is required.5.4.11.5 Inspection and Testing Requirements The system components are subject to non destructive and hydrosta tic testing during construction in accordance with Section VIII, Division 1 of the ASME Code (Table 5.4-12).

During plant operation, periodic visual inspections and preventive main tenance are conducted on the system components according to normal industrial practice.

5.4.12 VALVES 5.4.12.1 Design Bases As noted in Section 5.2, all valves out to and including the second valve normally closed or capable of automatic or remote closure, lar ger than three-quarter inch, are ANS Safety Class 1, and ASME III, Code Class 1 valves. All three-quarter inch or smaller valves in lines connected to the reactor coolant system (RCS) are Class 2 si nce the interface with the Class 1 piping is provided with suitable or ificing for such valves. Exceptions to this are the pressurizer steam space isolation valves and RHS PLTB lines which are ASME III Class 1 with the orifice located downstream of the isolation valve. Design data for the RCS are given in Table 5.4-15.

For a check valve to qualify as part of the RCS it must be located inside the containment system.

When the second of two normally open check valves is considered part of the RCS (as defined in MPS3 UFSAR5.4-47Rev. 30 Section 5.1), and verification of proper valve closure is requi red, non-intrusive techniques (e.g., radiography) are employed to perform this verification.To ensure that the valves meet the design objectives, the materials of construction minimize corrosion/erosion and ensure compatibility with the environment, leakage is minimized to the extent practicable by design, and stresses are mainta ined within the limits of the ASME Section III Code.

5.4.12.2 Design DescriptionAll valves in the RCS are constructed primaril y of stainless steel.

All manual and motor-operated valves in the RCS wh ich are 2.5 inches and larger were originally provided with double packed stuf fing boxes and intermediate lantern ring leakoff connections. The PORV block valves were replaced with valves that do not require leakoff connections. All throttling control valves, regardless of size, ar e provided with double stuf fing boxes and with stem leakoff connections. All leakoff connections are piped to a closed collection system. Leakage to the atmosphere is essentially zero for these valves.

Gate valves at the engineered safety features interface are wedge design and are essentially straight through. The wedges are flexwedge or solid. All gate va lves have backseats. Check valves are either swing type or tilting disc for size 2.5 inches and larger. All check valves which contain radioactive fluid are stai nless steel and do not have body pene trations other than the inlet, outlet and bonnet. The check hinge is services through the bonnet. The acc umulator check valve is designed such that at the required flow the resulting pressure drop is within the specified limits.

All operating parts are contained within the body. The disc has limited rotation to provide a change of seating surface and al ignment after each valve opening.The reactor coolant loop stop valves are remotely controlled motor- operated gate valves which permit any loop to be isolated from the reactor vessel. One valve is installed on each hot leg and one on each cold leg. The design of the valve is basically the same as noted above with the additional feature that each set of packing is capable of being tight ened independently of the other sets of packing. Also, the valve is a paralleled disc design. RCS parameters are given in Table 5.4-15.

5.4.12.3 Design Evaluation The design/analysis requirements fo r Class 1 valves, as discussed in Section 5.2, limit stress to levels which ensure the structural integrity of the valves. In addition, the testing programs described in Subsection 3.9N.3.2.2 demonstrate the ability of the valves to operate as required during anticipated and postulated coolant conditions.

Reactor coolant chemistry parameters are specifi ed in the design specifications to assure the compatibility of valve construction materials with the reactor coolant. To ensure that the reactor coolant continues to meet these parameters, the chemical composition of the coolant is analyzed periodically as discussed in the technical specifications.

MPS3 UFSAR5.4-48Rev. 30The above requirements and procedures, coupled with the previous ly described design features for minimizing leakage, ensure that the valves perform their intended func tions as required during plant operation.

5.4.12.4 Tests and InspectionsTests and examinations of RCS va lves are performed in accordance with the requirements of the ASME Code,Section III. There are no full penetration welds within valve body walls. Valve nondestructive examinations are given in Table 5.4-16.

The tests and inspection di scussed in Section 3.9 are performed to ensure the operability of active valves. In place operational testing is performed on valves as required by the ASME Code,Section XI, as indicated in the Technical Specifications.Valves are accessible for disassembly and internal visual inspection to the extent practical.

Inservice inspection is discussed in Section 5.2.4.

5.4.13 SAFETY AND RELIEF VALVES 5.4.13.1 Design Bases The pressurizer safety valves are designed to accommodate the maximum surge resulting from complete loss of load. Sizing of the pressurizer safety valves is discussed in Section 5.2.2. The pressurizer power-operated relief valves are designed to limit pressurizer pressure to a value below the fixed high pressure reacto r trip setpoint. They are designed to fail to the closed position on power loss.

5.4.13.2 Design Description The pressurizer safety valves are of the pop type. The valves are spring loaded, open by direct fluid pressure action, and have back pressure compensation features.

The piping connecting the pressurize r nozzles to their respective safety valves are shaped in the form of a loop seal. However, loop seal drains are mainta ined open to eliminate the formation of a water seal. EPRI testing showed water seals cause substantial safety valve discharge pipe loads.

Condensate resulting from normal heat losses drains back to the pressurizer via a drain tapped to the low point of the loop seal. The pressurizer power operated relief valves are solenoid operated valves which are operated automatically or by remote manual control. Th e pressurizer power operated relief valves are provided with a positive positi on indication in the control room (open/closed indication lights which are activated by limit switches).

Remotely operated stop valves are provided to isolate the power operated relief valves if excessive leakage develops. Positive position i ndication (open/closed) fo r the stop valves is located in the control room.

MPS3 UFSAR5.4-49Rev. 30The power operated relief valves (PORVs), the PORV block valves, and pressurizer level instrumentation are powered from the Class IE power sy stem (Section 8.3.1).Temperatures in the pressurizer safety and relief valve discharge lines are measured and indicated in the control room. An increase in a discharge line temperature is an indication of leakage or relief through the associated valve.Valves identical to Millstone 3's power operated relief valves (PORVs) a nd safety valves were tested, in a program conducted by EPRI under full flow, expected saturated steam operating conditions. Power operated relief valve tests we re completed in August 1981 and safety valve tests were completed in December 1981. Testing of valves was performed and test results were evaluated by Westinghouse and a report was generated. Additionally, testing of valves similar to the PORV block valves, is documented in Duke Engineering Services Inc. Report TR-161, "Dynamic Test Program for BW/IP International, Inc. - Valve Division Inch Parallel Disk Gate Motor-Operated Valve," dated 9/8/97. This test verified valve operability under full flow conditions. The PORVs were analyzed by Westinghouse in 1998 and qualified for operation during subcooled water conditions. The block valves were similarly qualified to ope rate with subcooled water in accordance with GL89-10 requireme nts per the licensee's program.

Design parameters for the pressurizer safety and power operated relief valves are given in Table 5.4-17.5.4.13.3 Design EvaluationThe pressurizer safety valves prevent RCS pressu re from exceeding 1 10 percent of system design pressure, in compliance with the ASME Code.The pressurizer power operated relief valves prevent actuation of the fixed reactor high pressure trip for all design transi ents up to and including the design step load decreases with steam dump. The relief valves also limit th e opening of the spring loaded safety valves. The pressurizer power-operated relief valves also provi de a safety grade means to depressurize the RCS for safety grade cold shutdown. See Section 5.4.7.2.3.5.

5.4.13.4 Inspection and Testing RequirementsTests and examination of pressuri zer safety and relief valves are performed in accordance with the requirements of the ASME Code,Section III. Th ere are no full penetration welds within valve body walls. Valve nondestructive examinations are given in Table 5.4-16.

The tests and inspections discussed in Section 3

.9 are performed to ensure the operability of active valves. In place operati onal testing is performed on valv es as required by the ASME OM Code as indicated in th e technical specifications.

MPS3 UFSAR5.4-50Rev. 30Valves are accessible for disassembly and internal visual inspection to the extent practical.

Inservice inspection is discussed in Section 5.2.4.

5.4.14 COMPONENT SUPPORTS Component supports are part of a safety system that permits m ovement to accommodate thermal expansion of the reactor coolant loops during plant operation while providing restraint to the reactor coolant system (RCS) comp o nents during accident conditions.Included as part of the reactor coolant system are four steam generators, four reactor coolant pumps, one pressurizer, and the reactor vessel.

Supports for these components are designed to maintain their necessary safety functions dur ing normal operating condi tions, postulated safe shutdown earthquake (SSE) conditions , and accidents such as postula ted extremes of pipe rupture acting concurrently with the SSE. Postulated pi pe ruptures are assume d to be double-ended or longitudinal. These failures are assumed to occur in either the reactor c oolant piping, pressurizer surge piping, or main steam line piping (Secti on 3.6). Section 3.7B discus ses seismic design of supports for reactor coolant piping. Detailed desi gn bases and results of qualification analyses are contained in Section 3.9B.3.

5.4.14.1 DescriptionThe supports are comprised of forged, cast, and we lded steel sections. Li near type supports are used in all cases except for the RP V support which is a plate and sh ell support. The attachments to the supported equipment are non-in tegral and are bolted to or bear against the components.

Attachment to the interface to building structure is achieved by embedded anchor bolts and shear lugs.5.4.14.1.1 Reactor Vessel Struct ural Support (R VSS)The support for the reactor vessel (t he neutron shield tank) is a cylindrical double-wall structure that surrounds and supports the re actor pressure vessel, and acco mmodates all applicable loading conditions. The RVSS transfers all loading conditio ns from the reactor vessel to the primary shield wall through groutings, and to the concrete anchors at its base. The RVSS also provides support for the out-of-core neutron detector moni tors. The annular portion of the tank is filled with water to provide neutron shielding and a thermal barrier for protection of the surrounding structural concrete. The water is circulated through an external h eat exchanger to maintain proper cooling for the system. The reactor vessel is supported at four nozzles on leveling devices mounted on top of the neutron shield tank. Th e functional requirement of the RPV leveling devices is to provide vertical adjustment at each RPV nozzle restraint pa d during installation of the reactor vessel. During all plant conditions, the leveling device is designed to transfer only downward vertical loads from the RPV to the RV SS. Upward and side loads from the RPV are resisted by gib keys and gib gussets. The RVSS is shown on Figures 5.4-9 and 5.4-10.

5.4.14.1.2 Steam Generator SupportsThe supports for each steam generator consist of vertical, upper and lower lateral supports.

MPS3 UFSAR5.4-51Rev. 30Four individual column assemblies provide the vertical support for each steam generator. Each column assembly consists of a lower clevis, column lug, exte nsion tube and upper column clevis. The upper clevises are bolted to the steam gene rator tube sheet and the lower clevises are anchored to the concrete floor. The four vertical column assemblies transmit vertical forces from the steam generator to the cubicle floor.

The lateral (upper and lower) supports are pr ovided by eight double acti ng hydraulic snubbers. Each lateral support has four hydraulic snubber assemblies whic h permit motion of the steam generator due to thermal expansion of the RCS. Vertical steam generator thermal motions are accommodated by the upper lateral support assembly.

The hydraulic snubbers are designed to lock and resist dynamic forces which result from seismic and/or pipe rupture conditions.

The lower lateral support assembli es are bolted to the steam genera tor tube sheet and the concrete wall. The upper lateral support asse mblies are bolted to the steam generator restraint ring and the concrete wall. The steam generator supports are shown on Figures 5.4-11 and 5.4-12.

5.4.14.1.3 Reactor Coolant Pump SupportsThe reactor coolant pump is supported by three pin ended columns which prov ide vertical support while allowing free movement in the horizontal plane. Three independent hydraulic snubber assemblies, connected to the pump support and the reactor shield wa ll, provide lateral support for the pump during dynamic loading conditions while allowing thermal expansion of the RCS. The pump supports are shown on Figures 5.4-11 and 5.4-13.

5.4.14.1.4 Pressurizer SupportThere is one pressurizer located in the pressu rizer cu bicle of the co ntainment building. The pressurizer is an integral part of the RCS and is connected to the hot leg of Loop 2 by the surge line (Section 5.4.10).

The pressurizer is skirt mounted to a ring girder which is suspended from the operating floor by four hanger columns. Four horizontal support re straints, which attach the ring girder to the building structure, prevent all mo tions except vertical translati on and horizontal rotation. Integral lugs located on the pressurizer near the center of gravity fit into striker plate as semblies embedded in the concrete floor at elevation 51 feet 4 inches These bracket s allow thermal expansion of the pressurizer but resist horizontal and torsional displacements re sulting from seismic and/or blowdown forces. The pressurizer support is shown on Figure 5.4-14.

5.4.14.1.5 Pressurizer Safety Valve Supports The pressurizer safety valves are mounted on a ri ng girder that is loca ted on the upper portion of the pressurizer. Support flanges are bolted to th e valve body which in turn are connected to the ring girder. The ring girder is s upported by four columns welded to the ring and pin-connected to clevis lugs attached to the pressurizer gussets. The supports are designed to withstand the loads MPS3 UFSAR5.4-52Rev. 30imposed by the safety relief piping which consist of dead weight, thermal, seismic, and simultaneous discharge of all safety valves. Th e pressurizer safety va lve supports are shown on Figure 5.4-15.

5.4.14.2 Design Basis The final designs, established for these supports, are based on maximum load combinations.

These load combinations are derived from a sy stem analysis using bl owdown forcing functions developed by Westinghouse. The maximum load s are a combination of dead weight, SSE (horizontal plus vertical), and a pipe rupture condition fo r each support structure. These maximum loads are used as the faul ted conditions for support design.From these final support designs, the inertia, stiffness, and damp ing quantities are evaluated and used as a basis for further refinement by elasti c structural dynamic analysis, as described in Section 3.7B.3. The designs are revi sed according to the computer re sults for more uniform stress distribution. This cycle is repeated, as required, to achieve sufficient opt imization of structural design efficiency for the supports (Section 3.9B.3).

The loading categories, load combinations, and st ress limits for the supports are shown in Table 5.4-18. The steam generator, reactor coolant pump a nd pressurizer supports are classified as linear type supports. For these linear type supports subject to design, normal, upset, and emergency operating loads, the stress limits are based on the elastic analysis of the ASME III Code, Subsection NA, Appendix XVII-2000. The reactor ve ssel support, neutron shield tank, is a combination linear and plate and sh ell-type support. The stress limi ts for linear type supports are per Appendix XVII-2000 and the st ress limits for plate a nd shell type supports are per NF-3220 of the ASME Code,Section III. Faulted operating conditions for all component supports have been analyzed in accordance with Appendix F of the ASME III Code.

5.4.14.3 EvaluationDetailed evaluation ensures the de sign adequacy and structural integrity of the reactor coolant loop and the primary equipment supports system. This detailed evaluation is made by comparing the analytical results with established criteria for acceptability as desc ribed in Section 3.9B.3.

Structural analyses are performe d to demonstrate design adequacy of the plant in case of an OBE or SSE and/or LOCA conditions.

Loads which the system is expected to encounter during plant operation (thermal, weight, pressure) are applied and stresses are compared to allowable values as described in Section 3.9B.3.

The safe shutdown earthquake (SSE) and design basis LOCA resulting in a rapid depressurization of the system, are required design conditions for public health and safety. The methods used for the analysis of the SSE and LOCA conditions are given in Section 3.9B.1.

MPS3 UFSAR5.4-53Rev. 30 5.4.14.4 Tests and InspectionsWeld inspection and standards are specified in accordance with Section V of the ASME Code. Welder qualifications and welding procedures are specified in accor dance with Section IX of the ASME Code.

5.4.15 REACTOR VESSEL HEAD VENT SYSTEMThe reactor vessel head vent system (RVHVS) (Figure 5.1-

1) removes non condensable gases or steam from the reactor vessel h ead. This system is designed to mitigate a possible condition of inadequate core cooling or impaired natural circulation resulting from the accumulation of nonconsumable gases in the RCS. Additionally, the system provides the safety grade letdown path to the pressurizer relief tank for a safety gr ade cold shutdown. The design of the RVHVS is in accordance with the requi rements of NUREG-0737.

5.4.15.1 Design BasisThe RVHVS is designed to remove non consumable gases or steam from the reactor coolant system via remote manual operations from the control room. The system discharges to the pressurizer relief tank. Additionally, a letdown flow path is provided from the reactor vessel head vent to the excess letdown heat exchanger in the chemical and volume control system (CHS). The RVHVS is designed to vent a volume of hydrogen at system de sign pressure and temperature approximately equivalent to one half of the reactor coolant system volume in one hour.

The system provides for venting the reactor vessel head by using only safety grade equipment. Letdown to accommodate boration during a safety grade cold shutdown is also provided by this path. To ensure reliability of this function, the le tdown line is provided with parallel solenoid valves. The valves are designed to fail closed such that both lines can always be isolated, and the two valves in the same line are powered by the sa me power train such that at least one line can always be made available. Downstream of these isolation valves, the safety grade path directs letdown to the pressurizer relief tank via parallel solenoid valves.

All piping and equipment from the ve ssel head vent up to and in cluding the second isolation valve in each flow path are designed and fabricated in accordance with ASME Section III, Class 1 requirements. The piping and equipment in the flow paths from the isolation valve to the modulating valves and from the isolation valv es to the excess letdown heat exchanger are designed and fabricated in accordance with AS ME Section III, Class 2 requirements. The remainder of the piping and equipment is non-nuclear safety.The isolation valves are included in the Wes tinghouse valve operability program which is an acceptable alternative to Regulatory Guide 1.48. These valves are qualified to IEEE 323-1974, 344-1975, and 382-1972 (Section 3.11).

All supports and support structures comply with the requirements of the ASME Code.

MPS3 UFSAR5.4-54Rev. 30The analysis of the rector vessel head vent piping is based on the following plant operating conditions defined in the ASME Code, Section III:1.Normal ConditionPressure, deadweight, and thermal expansion analysis of the vent piping during:a.Normal reactor operation with the vent isolation valves closed andb.Post refueling venting2.Upset Condition (including safety grade cold shutdown)

Loads generated by the operating basis earthquake (OBE) and by valve thrust during venting3.Faulted Condition Loads generated by the safe shutdown earthquake (SSE). Loads generated by valve thrust during venting. In accordance with ASME III, faulted conditions are not included in fatigue evaluations.

The Class 1 piping used for the reactor vessel head vent is 1 inch schedule 160 and, therefore, in accordance with ASME III, is analyzed following the procedures of NC-3600 for Class 2 piping.

For all plant operating conditions listed above , the piping stresses are shown to meet the requirements of equations (8), (9), and (10) or (1 1) of ASME III, NC-3600, with a design temperature of 650

°F and a design pressure of 2,485 psig.

5.4.15.2 System DescriptionThe RVHVS consists of two parallel flow paths wi th redundant isolation valves in each flow path.

The venting operation uses only one of these flow paths at any one time. The equipment design parameters are listed in Table 5.4-19.

The active portion of the system consists of four one inch open/close so lenoid-operated isolation valves connected to the existing 1 inch vent pipe, which is located near the center of the reactor vessel head. The system design with two valves in series in each fl ow path minimizes the possibility of reactor c oolant pressure boundary leakage. The isol ation valves in one flow path are powered by one vital power suppl y and the valves in the second flow path are powered by a second vital power supply. The isolation valv es are fail closed normally closed valves.

The vent system piping is supported to ensure th at the resulting loads a nd stresses on the piping and on the vent connection to vessel head are acceptable.

MPS3 UFSAR5.4-55Rev. 30 5.4.15.3 Safety Evaluation If one single active failure prevents a venting operation through one flow path, the redundant path is available for venting. The two isolation valves in ea ch flow path provide a similar method of isolating the venting system. With two valves in series, the failure of any one valve or power supply will not inadvertently open a vent pat

h. Thus, the combination of safety grade train assignments and valve failure mode s will not prevent vessel head venting nor venting isolation with any single active failure.The RVHVS has two normally deenergized valves in series in each flow path. This arrangement eliminates the possibility of an opened flow path due to the spuri ous movement of one valve. As such, power lockout to any valve is not considered necessary.A break of the RVHVS line would result in a small LOCA of not greater than one inch diameter.

Such a break is similar to those analyzed in WCAP-9600 (1979). Since a break in the head vent line would behave similarly to the hot leg break case presented in WCAP-9600, the results presented therein are applicable to a RVHVS line break. This postulated vent line break, therefore, results in no calculated core uncovery.

5.4.15.4 Inspection and Testing Requirements Inservice inspection is conducted in accordance with Sections 5.2.4 and 6.6.

5.4.15.5 Instrumentation Requirements The system is operated from the control room a nd the auxiliary shutdown panel. The isolation valves have stem position switches. The position indication from each valve is monitored in the control room by status lights.

5.4.16 REFERENCES FOR SECTION 5.45.4-1WCAP-8163, 1973, "Reactor Coolant Pump Integrity in LOCA, Westinghouse."5.4-2WCAP-8768, Revision 2, 1978, "Safety Rela ted Research and Development for Westinghouse Pressurized Water Reactor, Program Summaries," Winter 1977-Summer 1978, Westinghouse.5.4-3WCAP-9600, 1979. "Report on Sm all Break Accidents for Westinghouse NSSS System," (Section 3.2).5.4-4WCAP-7832, Evaluation of Steam Generator Tube, T ube Sheet and Divider Plate Under Combined LOCA plus SSE conditions, December 1973, Westinghouse.5.4-5WOG-87-102, 5/12/87, Mode 4 LOCA Concern Interim Guidance.

MPS3 UFSAR5.4-56Rev. 305.4-6OG-90-30, 6/1/90, Shutdown LO CA Analysis Concerns That Relate to the Interim Guidance.5.4-7WOG-90-48, 3/6/90, RHR System Oper ability During Mode 4 LOCA.

MPS3 UFSAR5.4-57Rev. 30TABLE 5.4-1 REACTOR COOLANT PUMP DESIGN PARAMETERS Unit design pressure (psig) 2,485 Unit design temperature (°F)650 (a)Unit overall height (ft) 26.3 Seal water injection (gpm) 8 Seal water return (gpm)

2.5 Component

cooling water flow (gpm) (b)216 Maximum continuous com ponent cooling water inlet temperature (°F)105 Chilled water flow (gpm) (c)220Maximum chilled water inlet temperature (°F)45Total weight, dry (lb) 187,852Pump Design flow, best estimate (gpm)100,400 Developed head, best estimate (ft)289NPSH required (ft)Figure 5.4-2

Suction temperature, thermal design (°F)556.8Pump discharge nozzle, inside diameter (in)27-1/2Pump suction nozzle, inside diameter (in)31Speed (rpm)1,186Water volume (ft 3)80 (d)Motor Type:Drip proof, squirrel-cage induction, with water/air coolers Power (hp) 7,000Voltage (V) 6,600Phase 3 Frequency (Hz) 60 Insulation classClass B, thermalasatic epoxy insulation MPS3 UFSAR5.4-58Rev. 30NOTES:a.Design temperature of pressu re-retaining parts of the pump assembly exposed to the reactor coolant and injection water on the high pressure side of the controlled leakage seal shall be that temperature determined for the parts for a reactor coolant loop temperature of 650°F.b.Component cooling water is supplied to the thermal barrie r at 40 gpm, the upper bearing cooler at 170 gpm, and the lower bearing cooler at 6 gpm.c.Chilled water is supplied to the air coolers at a flow rate of 1 10 gpm to each air cooler.d.Composed of reactor coolant in the casing a nd of seal injection and cooling water in the thermal barrier

.Current (amp):Starting3,000 @ 6,600 VNormal input, hot reactor coolant506

+/- 10Normal input, cold reactor coolant664

+/- 13 Pump moment of inertia, max (lb/ft 2):Flywheel70,000Motor22,500 Shaft520Impeller1,980TABLE 5.4-1 REACTOR COOLANT PUMP DESIGN PARAMETERS MPS3 UFSAR5.4-59Rev. 30*Either a UT over the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or a surface examination (MT/PT) of exposed surfaces defined by the volume of the disassembled flywheel.TABLE 5.4-2 REACTOR COOLANT PUMP NON-DESTRUCTIVE EXAMINATION PROGRAMRTUTPTMT Castings yesyesForgings Main shaftyesyesMain studsyesyesPlate Flywheel yes *yes *yes *Weldments CircumferentialyesyesInstrument connectionsyes NOTES:RT - Radiographic UT - Ultrasonic PT - Dye penetrant

MT - Magnetic particle MPS3 UFSAR5.4-60Rev. 30TABLE 5.4-3 STEAM GENERATOR DESIGN DATADesign pressure, reactor coolant side (psig) 2,485 Design pressure, steam side (psig) 1,185 Design pressure, primary to secondary (psi) 1,600 Design temperature, r eactor coolant side (°F) 650 Design temperature, steam side (°F) 600 Design temperature, pr imary to secondary (°F) 650 Total heat transfer surface area (ft

2) 55,000 Maximum moisture carryover (weight percent) 0.25 Overall height (ft-in) 67-8 Number of U-tubes 5,626 U-tube nominal diameter (in.)

0.688 Tube wall nominal thickness (in.)

0.040 Number of manways 4 Inside diameter of manways (in.)

16 Number of handholes 6 Number of inspection ports 2 Design fouling factor (ft 2-hr-°F/Btu) 0.00006 Steam flow (lb/hr) 4.075 x 10 6

MPS3 UFSAR5.4-61Rev. 30TABLE 5.4-4 STEAM GENERATOR NONDESTRUCTIVE EXAMINATION PROGRAMRTUTPTMTETTube Sheet Forgingyesyes Cladding yes *yesChannel Head (if fabricated) Fabrication yes**yes ***yesCladdingyes Secondary Shell and Head PlatesyesTubes yesyesNozzles (Forgings) yesyesWeldments Shell, longitudinalyesyes Shell, circumferentialyesyes

Cladding (channel head-tube sheet joint cladding restoration) yesPrimary nozzles to fab headyesyesManways to fab headyesyesSteam and feedwater nozzle to shellyesyesSupport bracketsyes Tube to tube sheetyes Instrument connections (primary and secondary)yesTemporary attachments after removalyes

After hydrostatic test (all major pressure boundary welds and complete cast channel head - where accessible) yesNozzle safe ends (if weld deposit)yesyes MPS3 UFSAR5.4-62Rev. 30NOTES:RT - Radiographic UT - Ultrasonic PT - Dye penetrantMT - Magnetic particle ET - Eddy current Flat surfaces only**Weld deposit***Base material only MPS3 UFSAR5.4-63Rev. 30TABLE 5.4-5 REACTOR COOLANT PI PING DESIGN PARAMETERS Reactor inlet piping inside diameter (in) 27-1/2 Reactor inlet piping, nomin al wall thickness (in) 2.32 Reactor outlet piping inside diameter (in) 29 Rector outlet piping, nominal wall thickness (in) 2.45 Coolant pump suction piping inside diameter (in) 31 Coolant pump suction piping, nominal wall thickness (in) 2.60Pressurizer surge line piping, nominal pipe size (in) 14Pressurizer surge line piping, nominal wall thickness (in) 1.406 Reactor Coolant Loop Piping Design/operating pressure (psig)2485 / 2235 Design temperature (°F)650Pressurizer Surge Line Design pressure (psig)2485 Design temperature (°F)680Pressurizer Safety Valve Inlet Line Design pressure (psig)2485 Design temperature (°F)680Pressurizer (Power-Operated) Relief Valve Inlet Line Design pressure (psig)2485

Design temperature (°F)680Reactor Head Vent Piping Design pressure (psig)2485

Design temperature (°F)650 Nominal pipe size (in) 1Wall thickness (schedule) 160 MPS3 UFSAR5.4-64Rev. 30Pressurizer Relief Tank Inlet Line Design pressure (psig)600 Design temperature (°F)600Loop Stop Valve Bypass Line Design pressure (psig)2485 Design temperature (°F)650 Loop stop valve bypass line nominal pipe size (in) 8 Loop stop valve bypass line nominal wall thickness (in) 0.906TABLE 5.4-5 REACTOR COOLANT PI PING DESIGN PARAMETERS MPS3 UFSAR5.4-65Rev. 30NOTES:*RT - Radiographic UT - Ultrasonic PT - Dye PenetrantTABLE 5.4-6 REACTOR COOLANT PIPI NG QUALITY ASSURANCE PROGRAM RT*UT*PT *Fittings and Pipe (Castings) yesyesFittings and Pipe (Forgings) yesyesWeldments 1. Circumferentialyesyes2. Nozzle to runpipe yesyes (Except no RT for nozzles less than 6 inches)3. Instrument connections yes Castings yes yes (after finishing)Forgings yesyes (after finishing)

MPS3 UFSAR5.4-66Rev. 30TABLE 5.4-7 DESIGN BASES FOR RESI DUAL HEAT REMOVAL SYSTEM OPERATIONSGCSNORMALReactor coolant system initial pressure (psig) 375375Reactor coolant system initial temperature (°F)350350 Maximum component cooling water supply temperature (°F)113110Maximum component cooling water outlet temperature

°F145145 Cooldown time, after re actor shutdown (hr)7236Reactor coolant system temper ature at end of cooldown (°F)200200 MPS3 UFSAR5.4-67Rev. 30TABLE 5.4-8 RESIDUAL HEAT REMOVAL SYSTEM COMPONENT DATA Residual Heat Removal Pump Number2Design pressure (psig)600 Design temperature (°F)400 Design flow (gpm) 4,000 Design head (ft) 350 NPSH required at 3,800 gpm (ft) 18NPSH required at runout flow 5,500 gpm (ft)25 Power (hp) 400 Residual Heat Exchanger Number2 Design heat removal capacity (Btu/hr) 35.27 x 10 6Estimated UA (Btu/hr

°F)3.5 x 10 6Tube Side Shell Side Design pressure (psig)600175

Design temperature (°F)400 200 Design flow (lb/hr) 1.98 x 10 6 3.3 x 10 6Inlet temperature (°F)120 92.2 Outlet temperature (°F)102.2 102.9 MaterialAustenitic Stainless SteelCarbon Steel FluidReactor Coolant Water Component Cooling MPS3 UFSARMPS3 UFSAR5.4-68Rev. 30TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks1.Motor-operated gate valve 8701A (8701B analogous)a.Fails to open on demandProvides isolation of fluid flow from the RCS to the suction of RHR pump 1 (pump 2)a. Failure blocks reactor coolant flow from hot leg of RC loop 1 (loop 4) through train "A" (train "B") of RHRS. Failure reduces redundancy of RHR coolant trains provided. No effect on safety for system operation. Plant cooldown requirements will be met by reactor coolant flow from hot leg of RC loop 4 (loop 1) flowing through train "B" (train "A") of RHRS, however, time required to reduce RCS temperature will be extended.a. Valve open/close position indication at CB; RC loop 1 (loop 4) hot leg pressure indication at CB: RHR train "A" (train "B") discharge flow indication and low flow alarm at CB; and RHR pump 1 (pump 2) discharge pressure indication and low flow alarm at CB; and RHR pump 1 (pump 2) discharge pressure indication at CB.1. Valve is electrically interlocked with RWST to RHR

suction line isolation valve 8812A (8812B),

with RHR to charging pump

suction line isolation valve 8804A (8804B) and with a "prevent-open" pressure interlock PT405/

PT-405A, PT403/

403A of RC loop 1 (loop 4) hot leg.

The valve cannot be opened remotely from the CB if one of the indication isolation valves is open or if RC loop pressure exceeds 412.5 psia. The valve can be manually opened.2. Motor-operated gate valve 8702A (8702B analogous)

Same as item

1Same as item 1Same as item 1.Same as item 1.Same as item 1, except for pressure interlock PT405/

405A (PT-403/403A) control.

MPS3 UFSARMPS3 UFSAR5.4-69Rev. 303. RHR pump 1 (RHR pump 2 analogous)

Fails to deliver working fluidProvides fluid flow of reactor coolant through RHR heat exchange 1 (heat exchanger

2) to reduce RCS temperature during cooldown operationFailure results in loss of reactor coolant flow from hot leg of RC loop 1 (loop 4) through train "A" (train "B") of RHRS.

Failure reduces redundancy of

RHR coolant trains provided.

No effect on safety for system operation. Plant cooldown requirements will be met by reactor coolant flow from hot leg of RC loop 4 (loop 1) flowing through train "B" (train "A") of RHRS, however, time required to reduce RCS temperature will be extended.Open pump switchgear circuit breaker indication at CB; circuit breaker close position monitor light for group monitoring of components at CB; common breaker trip

alarm at CB; RC loop 1 (loop 4) hot leg pressure indication at CB; RHR train "A" (train "B") discharge flow indication and low flow alarm CB; and pump discharge pressure indication

at CB.The RHRS shares

components with the ECCS. Pumps are tested as part of the ECCS testing

program (see Section 6.3.4).4. Motor-operated globe valve FCV-610 (FCV-611 analogous)a. Fails closedProvides regulation of fluid flow through miniflow bypass line to suction of RHR pump 1 (pump 2) to protect against overheating of the pump and loss of discharge flow from the pump.a. Failure blocks miniflow line to suction of RHR pump 1 (pump 2) during cooldown operation. No effect on safety for system operation. Plant cooldown requirements will be

met by reactor coolant flow from hot leg of RC loop 4 (loop 1) flowing through train "B" (train "A") of RHRS, however, time required to reduce RCS temperature will be extended.a. Valve open/close position indication at CB; and RHRS train "A" (train "B")

discharge flow indication at

CB.1. Valve is automatically controlled to open when pump discharge is less than 772 gpm and close when the discharge exceeds 1542 gpm. These

flow are nominal valves.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-70Rev. 30b. Fails openb.Failure allows for a portion of RHR heat exchanger 1 (heat exchanger 2) discharge flow to be bypassed to suction of RHR pump 1 (pump 2). RHRS train "A" (train "B") is degraded for the regulation of coolant temperature by RHR heat exchanger 1 (heat exchanger 2). No effect on safety for system operation. Cooldown of RCS within established

specification cooldown rate may be accomplished through

operator action of adjusting throttle valves HCV-606 (HCV-607) and FCV-618 (FCV-619) to compensate for the open miniflow line and controlling cooldown with reduanant RHRS train "B" (train "A").Same as item 4.a.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-71Rev. 305.Air piston operated butterfly valve FCV-618 (FCV-619 analogous)a. Fails to open on demand for flow increase

("Auto" mode CB switch selection)Controls rate of fluid flow bypassed around RHR heat

exchanger 1 (heat exchanger 2) during cooldown operationa.Failure prevents coolant discharged from RHR pump 1 (pump 2) from bypassing RHR heat exchanger 1 (heat exchanger 2) resulting in mixed mean temperature of coolant flow to RCS being low.

RHRS train "A" (t rain "B") is degraded for the regulation of controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS within established specification rate may be

accomplished through operator action of throttling flow control valve HCV-606 (HCV-607) and controlling cooldown with redundant RHRS train "B (train" "A").a. RHR pump 1 (pump 2) discharge flow temperature and RHRS train "A" (train "B") discharge to RCS cold leg flow temperature recording at CB; and RHRS train "A" (train "B")

discharge to RCS cold leg flow indication at CB (TR 612).1. Valve is designed to fail "open" and is

electrically wired so that electrical solenoid of the air diaphragm operator is energized to open the valve. Valve is normally "open" to align RHRS for ECCS operation during plant power operation and load follow.2. Valve is designed for normal plant cooldown operation. It is required for safety grade cold shutdown operations, if only one train of RHS is

available and instrument air is lost.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-72Rev. 30b. Fails to close on demand for flow reduction

("Auto" mode CB switch selection)b. Failure allows coolant discharged from RHR pump 1 (pump 2) to bypass RHR heat

exchanger 1 (heat exchanger 2) resulting in mixed mean temperature of coolant flow to RCS being high. RHRS train "A" (train "B") is degraded for the regulation of controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS within established specification rate may be

accomplished through operation action of throttling flow control valve HCV-606 (HCV-607) and controlling cooldown with redundant

RHRS train "B" (train "A"),

however, cooldown time will be extended.b. Same as item 5.a.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-73Rev. 306. Air piston operated butterfly valve HVC-606 (HCV-607 analogous)a. Fails to close on demand for flow reductionControls rate of fluid flow through RHR heat

exchanger 1 (heat exchanger 2) during cooldown operationa. Failure prevents control of coolant discharge flow from RHR heat exchanger 1 (heat exchanger 2) resulting in loss of mixed mean temperature coolant flow adjustment to RCS. No effect on safety for system operation. Cooldown of RCS within established specification rate may be

accomplished by operator action of controlling cooldown with redundant RHR train "B" (train "A").a. Same methods of detection as those stated for item 5.a.

In addition, monitor light and alarm (valve closed) for group monitoring of

components at CB.1. Valve is designed to fail "open." Valve is normally "open" to align RHRS for ECCS operation during plant power operation and load follow.b. Fails to open on demand for flow increaseb. Same as item 6.a.b. Same as item 6.a.7. Motor-operated gate valve 8812A (8812B) analogous)Fails to close on demandProvides isolation of fluid from the RWST to suction of RHR pump 1 (pump 2) during cooldown operationNo effect on safety for system operation. Plant cooldown requirements will be met by reactor coolant flow from hot leg loop 4 (loop 1) flowing through train "B" (train "A")

of RHRS, however, time required to reduce RCS temperature will be extended.Valve open/closed position indication at CB and valve (closed) monitor light and

alarm at CB.Valve is normally "open" to align RHRS for ECCS operating during plant power operation and load follow. Valve must be closed during plant cooldown to

satisfy electrical interlock to permit valves 8701A and 8702A (8701B and 8702B) to be opened.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-74Rev. 308. Motor-operated gate valve 8716A (8716B analogous)Fails to close on demandProvides separation between the two RHR trains during cooldown

operationFailure reduces the redundancy for isolating RHR trains during cooldown. Negligible effect on system operation. Isolation valve 8716B (8716A) provides backup isolation between the two RHR trains.Same as item 7.9. Centrifugal charging pump 1 (pump 3 analogous)

Fails to deliver working fluidProvides fluid flow of borated water from the BAT or RWST to the RCSFailure reduces redundancy of providing water to the RCS at high RCS pressures. Fluid flow from charging pump 1 (pump

3) will be lost. Minimum flow requirements for boration and makeup will be met by charging pump 3 (pump 1).Charging pump discharge header pressure and flow indication at CB. Open/close pump switchgear circuit breaker indication on CB.

Circuit breaker close position monitor light. For group monitoring of

component at CB. Common breaker trip alarm at CB.1.The charging pumps provide boration and makeup flow to the RCS during safety grade cold shutdown operations. Note (4)2. Analysis of charging pump 2 being on line is analogous to that presented for charging pumps 1

and 3.10.Motor-operated gate valve LCV-112B (LCV-112C analogous)Fails to close on demandProvides isolation of fluid discharge from the VCT to the suction of charging pumpsFailure reduces redundancy of providing VCT discharging isolation. Negligible effect on safety for system operation.

Alternate isolation valve LCV-112C (LCV-112B) provides backup tank discharge isolation.Same as item 7.The charging pumps' suction is isolated from the VCT and aligned to the BAT (for boration) or RWST (for makeup) during safety grade cold shutdown operations.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-75Rev. 3011.Motor-operated gate valve LCV-112D (LCV-112E analogous)Fails to open on demandProvides isolation of fluid discharge from the RWST to the suction of charging pumpsFailure reduces redundancy of providing fluid from RWST to suction of charging pumps.

Negligible effect on safety for

system operation. Alternate isolation valve LCV-112E (LCV-112D) opens to provide backup flow path to suction of charging pumps.Valve open/close position indication at CB (open) monitor light and alarm at

CB.The charging pumps' suction is aligned to the RWST for makeup to the RCS during

ECCS and safety grade cold shutdown operations.12.Motor-operated gate valve 8105 (8106 analogous)Fails to close on demandProvides isolation of fluid flow from the charging pump discharge header to the CVCS normal charging line to the RCSFailure reduces redundancy of providing isolation of charging pump discharge to normal charging line of CVCS.

Negligible effect on safety for

system operation. Alternate isolation valve 8105 (8106) provides backup normal CVCS charging line isolation.Same as item 7 except no valve (closed) monitor alarm for group monitoring.Normal charging line is isolated during safety grade cold shutdown operations.

Boration and makeup flow provided to RCS through redundant ECCS headers to the RCS cold legs.13.Motor-operated gate valve 8468A (8468B analogous)Fails to close on demandProvides isolation barrier to isolate charging pump suction flow paths in the

event of a MELB in charging pump suction headerMELB isolation may be provided by closing isolation valve 8468B (8468A).Same as item 7.Note (4)14. Motor-operated gate valve 8438A (8438B analogous)Fails to close on demandProvides isolation barrier to isolate charging pump discharge flow paths in the event of a HELB in charging pump discharge headerHELB isolation may be provided by closing isolation valve 8438B (8438A).Same as item 7.Note (5)TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-76Rev. 3015.Solenoid operated globe valve HCV-190A (HCV-190B)Fails to open on demandProvides control of fluid flow from charging pump 1 (pump 3) to RCS during plant boration and makeupFailure reduces redundancy of controlling boration and makeup flow to the RCS.

Negligible effect on safety for

system operation. Alternate control valve HCV-190B (HCV-190A) controls flow from charging pump 3 (pump 1).Valve position indication at CB; and charging pump 1 (pump 3) discharge header flow indication at CB.Same as item 12.15a.Motor-operated Globe Valve MV 8116Fails to open on demandProvides isolation of fluid flow from pump/(pumps) to

RCSSame as Item 15Same as Item 15Same as Item 1216.Solenoid operated globe valve 8095A (8095B analogous)a. Fails to open on demandProvides isolation of fluid flow from the RV head to the CHS or PRTa. Failure reduces redundancy of providing flow from the RV head to the CHS or PRT.

Negligible effect on safety for system operation. RV head letdown flow provided by parallel head letdown path through alternate isolation valve 8095B (8095A).a. Valve open/close position indication at CB; and RV head letdown high

temperature indication and

alarm at CB.1. The RC head letdown path to the CHS or PRT provides fluid flow out of the RCS to

accommodate boration flow into the RCS.b. Fails to close on demandb. Failure reduces redundancy of isolating flow from the RV head to the CHS or PRT.

Negligible effect on safety for system operation. RV head letdown flow isolation provided by alternate series isolation valve 8096A (8096B).b. Same as item 16.a.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-77Rev. 3017. Solenoid operated globe valve 8096A (8096B analogous)a. Fails to open on demandSame as item 16a. Same as item 16.a except for alternative isolation valve 8096B (8096A).a. Same as item 16.a.1. Same as item 16.1 except that the RV Letdown Path is to the PRT.b. Fails to close on demandb. Same as item 16.b except for alternative series isolation valve 8095A (8095B).b. Same as item 16.a.18. Solenoid operated globe valve HCV-442A (HCV-442B analogous)Fails to open on demandSame as item 16 except that flow is from the RV head to the PRT Same as item 16.a except that flow is from the RV head to the PRT and the alternative parallel isolation valve HCV-442B (HCV-442A).Valve position indication at CB; RV letdown temperature indication at CB.Same as item 16.1 except that the RV Letdown Path is to the PRT.19.Solenoid operated power

operated relief valve PCV-455A (PCV-456 analogous)a. Fails to open on demandProvides isolation of fluid flow from pressurizer to PRTa. Failure reduces redundance of providing flow from pressurizer to PRT. Negligible effect on safety for system operation. Pressurizer vent flow provided by a parallel pressurizer vent path through alternate isolation valves PCV-456A or PCV-455A.a. Valve open/close position indication at CB; pressurizer power operated relief valve outlet temperature indication

at CB.1. Pressurizer vent path to the PRT provides fluid flow out of the RCS to permit RCS depressurization to RHRS initiation conditions.b. Fails to close on demandb.Failure reduces redundancy of isolating flow from the pressurizer to the PRT.

Negligible effect on safety for system operation. Pressurizer vent flow isolation provided by alternate series isolation valve 8000A (or 8000B).b. Same as item 19.a.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-78Rev. 3020. Motor-operated gate valve 8000A (8000B analogous)a. Fails to close on demandSame as item 19a. Same as item 19.a except pressurizer vent flow isolation provided by alternate series isolation valve PCV-455A (PCV-456).a. Same as item 19.a.1. Same as item 19.1.21. Motor-operated gate valve 8808A (8808B, 8808C, and 8808D analogous)Fails to close on demandProvides isolation of fluid flow from accumulator 1 (accumulator 2, accumulator 3, and accumulator 4) to the RCSFailure prevents isolation of accumulator 1 (accumulator 2, accumulator 3 and accumulator 4) from the RCS. Negligible effect on safety for system

operation. Accumulator 1 (accumulator 2, accumulator 3 and accumulator 4) is depressurized by opening vent isolation valves 8875A (8875B or 8875C or 8875D) and HCV-943A, or vent isolation valves 8875E (8875F, 8875G or 8875H) and HCV-943B.Valve open/closed position indication at CB, valve (closed) monitor light and alarm at CB and accumulator pressure indication and low alarm at

CB.Accumulators are isolated or vented during plant cooldown to not effect RCS depressurization to RHRS initiation conditions.22. Solenoid operated globe valve 8875A (8875B, 8875C and 8875D analogous)Fails to open on demandProvides venting of nitrogen gas from accumulator 1 (accumulator

2, accumulator 3 and accumulator 4) to

containment Failure reduces redundancy for venting accumulator 1 (accumulator 2, accumulator 3 and accumulator 4) to

containment. No effect on safety for system operation.

Accumulator 1 (accumulator 2, accumulator 3, accumulator 4) can be vented by opening vent valves 8875E (8875F, 8875G and 8875H) and HCV-943B or isolated valve 8808A (8808B, 8808C, 8808D).Valve open/closed position indication at CB and accumulator pressure indication and low alarm at

CB.Same as item 21.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-79Rev. 3023. Solenoid operated globe valve 8875E (8875F, 8875G and 8875H analogous) Fails to open on demandSame as item 22Failure reduces redundancy for venting accumulator 1 (accumulator 2, accumulator 3 and accumulator 4) to

containment. No effect on safety for system operation.

Accumulator 1 (accumulator 2, accumulator 3 and accumulator

4) can be vented by opening vent valves 8875A (8875B, 8875C, and 8875D) and HCV-943A or isolated from the RCS by closing isolation valve 8808A (8808B, 8808C, and 8808D).Same as item 22.Same as item 22.24. Solenoid operated globe valve HCV-943A (943B analogous)Fails to open on demandProvides venting of nitrogen gas from accumulators to

containmentVenting can be accomplished via HCV-943B, (943A).Valve position indication at

CB and accumulator pressure indication and low

alarm at CB.Same as item 22.25. Boric acid transfer pump pump 1 (pump 2 analogous)Fails to deliver working fluidProvides fluid of

concentrated boric acid from BAT to charging pump suctionFailure reduces redundancy of providing concentrated boric acid to charging pump suction.

Fluid flow from boric acid transfer pump 1 (pump 2) will be lost. Minimum flow requirements for boration will be met by boric acid transfer pump 2 (pump 1).Pump motor start relay position indication (open) at CB and local pump discharge pressure indication PI-113 (PI-114).The boric acid

transfer pumps provide boration flow to the charging pumps' suction.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-80Rev. 30NOTES:(1)Components 5, 7, 8, 15 and 21 through 24 are components of the ECCS that perform a safety grade cold shutdown function.

Components 9 through 14, 25, 26, and 27 are co mponents of the CVCS that perform a safety grade cold shutdown function.

Components 16 through 20 are components of the RCS th at perform a safety grade cold shutdown function.

(2)List of Acronyms and Abbreviations Auto - Automatic

CB - Control board CVCS - Chemical and volume control systemECCS - Emergency core cooling system

RC - Reactor coolant RCS - Reactor coolant system RHR - Residual heat removal RHRS - Residual heat removal systemRWST - Refueling water storage tank26. Motor-operated globe valve 8104Fails to open on demandProvides isolation of fluid flow from either boric acid transfer pump to charging pump suctionFailure reduces redundancy of providing concentrated boric acid to charging pump suction.

Negligible effect on safety for system operation.

Concentrated boric acid provided to charging pump

suction through alternate isolation valve 8507A/B.Valve open/close position

indication at CB; and boration flow indication (FI-183A) at CB.The charging pumps' suction is aligned to the gravity drain lines (8507A/B) for boration of the RCS during safety grade cold shutdown operations.27. Air operated diaphragm valve 8439Fails to open on demandSame as item 26Same as item 26 Same as item 26 except for flow indication (FT-110).Same as item 26.TABLE 5.4-9 RESIDUAL HEAT REMOVAL SYSTEM - COLD SHUTDOWN OPERATIONS-FAILURE MODES AND EFFECTS ANALYSISComponent (1)Failure ModeFunction (2)Effect on System Operation (3)Failure Detection Methods (3)Remarks MPS3 UFSARMPS3 UFSAR5.4-81Rev. 30BAT - Boric acid tankVCT - Volume control tankMELB - Moderate energy line break HELB - High energy line breakRV - Reactor vesselPRT - Pressurizer relief tank(3)As part of the plant operation; periodic tests, surveillance inspections and instru ment calibrations are made to monitor equ ipment and performance. Failures may be detected during such monitoring of equipment in addition to detection methods noted.(4)Certain initiating MELB events, postulated to occur in the operating CHS pump suct ion piping, when combin ed with a single ac tive failure of the standby CHS pump to start, may lead to a loss of all char ging. For this condition, the SIH pumps will provide th e required RCS inventory and boration flow to achieve safe shutdown.(5)Certain initiating HELB events, postulated to occur in the operating CHS pump dischar ge piping, when combined with a single active failure of the standby CHS pump to start, may lead to a loss of all charging.

For this condition, th e SIH pumps will pro vide the required RCS inventory and borati on flow to achieve safe shutdown.

MPS3 UFSAR5.4-82Rev. 30TABLE 5.4-10 PRESSURIZER DESIGN DATA Design pressure (psig) 2485 Design temperature (°F)680Surge line nozzle diameter (inch) 14 Heatup rate of pressurizer using heaters only (°F/hr)55 Internal volume (ft 3)1800 MPS3 UFSAR5.4-83Rev. 30NOTE:1.At 2335 psig, a pressure signal initiates actuation (opening) of these valves. Remote manual control is also provided.2.Actual setpoints will vary depending on M/A station settings and controller response to plant conditions.TABLE 5.4-11 REACTOR COOLANT SYSTEM DESIGN PRESSURE SETTINGS psigHydrostatic test pressure3107Design pressure2485Safety valves (begin to open)2485High pressure reactor trip2370

Power operated relief valves 2335 (1)High pressure deviation alarm 2310 (2)Pressure spray valves (full open) 2310 (2)Pressure spray valve (begin to open) 2260 (2)Proportional heaters (begin to operate) 2250 (2)Operating pressure 2235 Proportional heater (full operation) 2220 (2)Backup heaters on 2210 (2)Low pressure deviation alarm 2210 (2)Low pressure reactor trip (typical, but variable) 1885 MPS3 UFSAR5.4-84Rev. 30NOTES:1.RT - Radiographic UT - Ultrasonic PT - Dye penetrantMT - Magnetic particle2.UT and ET3.MT or PTTABLE 5.4-12 PRESSURIZER QUALITY ASSURANCE PROGRAMRT 1 UT 1PT 1 MT 1Heads:PlatesyesCladdingyesShell:Platesyes CladdingyesHeaters:Tubing 2yesyesCentering of elementyesNozzle (forgings)yes yes 3 yes 3Weldments:Shell, longitudinalyesyes Shell, circumferentialyesyesCladdingyesNozzle safe end (if forging)yesyes Instrument connectionyesSupport skirt, longitudinal seamyesyesSupport skirt to lower headyesyes Temporary attachments (after removal)yes All external pressure boundary welds after shop hydrostatic test yes MPS3 UFSAR5.4-85Rev. 30TABLE 5.4-13 PRESSURE RELIEF TANK DESIGN DATA Design pressure (psig) 100Rupture disc release pressure (psig)Nominal:91range:86-100 Design temperature (°F)340Total rupture disc relief capacity at 100 psig (lb/hr) 1.6 x 10 6 MPS3 UFSAR5.4-86Rev. 30TABLE 5.4-14 RELIEF VALVE DISCHARGE TO THE PRESSURIZER RELIEF TANKReactor Coolant System 3 Pressurizer safety valvesFigure 5.1-1 2 Pressurizer power operated relief valves Figure 5.1-1 Residual Heat Removal System 2 Residual heat removal pump suction li ne from the Reactor Coolant System hot legs Figure 5.4-6Chemical and Volume Control System 1 Seal water return lineFigure 9.3-81 Letdown lineFigure 9.3-8 MPS3 UFSAR5.4-87Rev. 30TABLE 5.4-15 REACTOR COOLANT SYSTEM DESIGN PARAMETERS Design/normal operating pressure (psig) 2485 / 2235 Preoperational plant hydrotest (psig) 3107Design temperature (°F) 650 MPS3 UFSAR5.4-88Rev. 30NOTES:1.RT - Radiographic UT - Ultrasonic PT - Dye penetrant2.Weld ends only3.Forged stems UT onlyTABLE 5.4-16 NON-DESTRUCTIVE EXAMINATION PROG RAM REACTOR COOLANT SYSTEM VALVESRT (1)UT (1)PT (1)Castings(larger than 4 inches)yesyes(2 inches to 4 inches) yes (2)yesForgings yes (3)yes MPS3 UFSAR5.4-89Rev. 30TABLE 5.4-17 PRESSURIZER VA LVES DESIGN PARAMETERSPressurizer Safety Valves Number3Maximum relieving capacity, ASME rated flow per valve (lb/hr)420,000Set pressure (psig)2485

Design temperature (°F)650 FluidSaturated steam Backpressure Normal (psig)3 to 5Expected during discharge (psig)500Pressurizer Power Relief Valves Number2Design pressure (psig)2485

Design temperature (°F)650 Relieving capacity at 2350 psia, minimum per valve (lb/hr)210,000 FluidSaturated steam Relieving capacity at 2,438 psia, minimum per valve (lb/hr)353,880 FluidSubcooled water MPS3 UFSARMPS3 UFSAR5.4-90Rev. 30TABLE 5.4-18 EQUIPMENT SUPPORTS, LOADING COMBINATIONS, AND DESIGN ALLOWABLE STRESSES EquipmentLoading Category Loading CombinationsStress Limits CodesSteam Generator and Reactor Coolant Pump SupportsDesign, Normal & UpsetDead WeightApp. XVII-2000 and Paragraph NF-3230 ASME Boiler and Pressure Vessel Code,Section III, Subsection NF 1974 Edition through 1974 Winter Addenda+Thermal+1/2 SSE FaultedDead WeightApp. F-1370+SSE

+Pipe rupturePressurizer SupportsDesign, Normal & UpsetDead WeightApp. XVII-2000 and Paragraph NF-3230 ASME Boiler and Pressure Vessel Code,Section III, Subsection NF 1974 Edition+Thermal+1/2 SSE FaultedDead WeightApp. F-1370+SSE

+Pipe rupture MPS3 UFSARMPS3 UFSAR5.4-91Rev. 30RPVSS (Neutron Shield Tank)Design, Normal & UpsetDead WeightApp. XVII-2000 and Paragraph NF-3230 for linear type supports ASME Boiler and Pressure Vessel Code,Section III, Subsection NF 1974 Edition including 1974 Summer

Addenda+Initial Pressurization Paragraph NF-3220

for plate and shell type supports+Thermal+1/2 SSEFaultedDead WeightApp. F-1323.1 for plate and shell type supports+SSEApp. F-1370 for linear type supports

+Pipe rupture+Initial and Asymmetric

PressurizationTABLE 5.4-18 EQUIPMENT SUPPORTS, LOADING COMBINATIONS, AND DESIGN ALLOWABLE STRESSES EquipmentLoading Category Loading CombinationsStress Limits Codes MPS3 UFSAR5.4-92Rev. 30TABLE 5.4-19 REACTOR VESSEL HEAD VE NT SYSTEM EQUIPMENT DESIGN PARAMETERSValves Number of remote valves (6 solenoid, 1 motor operated) 7 Design pressure (psig) 2485 Design temperature (°F) 650 Maximum operating temperature (°F) 620 PipingVent line, nominal diameter (in) 1 Design pressure (psig) 2485 Design temperature (°F) 650 Maximum operating temperature (°F) 620 MPS3 UFSAR5.4-93Rev. 30*Seal leakoff for flowse rve seals will now be located 45 degrees CCW from the second stage outlet pressure connections.

FIGURE 5.4-1 REACTOR COOLANT PUMP UPPER MOTOR RADIAL BEARING MAIN LEAD

CONDUIT BOX No. 1 SEAL INJECTION

WATER PUMP SHAFT CASING SUCTION NOZZLE MODEL 93A-1 IMPELLER DISCHARGE NOZZLE PUMP RADIAL BEARING THERMAL BARRIER COOLING WATER INLET SPOOL PIECE LOWER MOTOR RADIAL

BEARING MOTOR SHAFT FLYWHEEL THRUST

BEARING OIL COOLER CVCS SEAL RETURN (CBO)

LOWER STAGE OUTLET PRESSURE (P2)

MIDDLE STAGE OUTLET PRESSURE (P3)

SEAL LEAKOFF

• MPS-3 FSAR Rev. 20.2FIGURE 5.4-2 REACTOR COOLANT PUMP ESTIMATED PERFORMANCE CHARACTERISTICS TOTAL HEAD - FEET FLOW - THOUSANDS OF GPM TOTAL HEAD REQUIRED NET POSITIVE SUCTION HEAD 110 100 90 80 7 0 60 5 0 4 0 30 20 10 0 0 100 200 300 400 500 600 MPS-3 FSAR Rev. 20.2FIGURE 5.4-3 MODEL F STEAM GENERATOR MPS-3 FSAR Rev. 20.2FIGURE 5.4-4 QUATRFOIL TUBE SUPPORT PLATES MPS3 UFSAR5.4-97Rev. 30 FIGURE 5.4-5 5 (SHEETS 1-3) P&IDS LOW PRESSURE SAFETY INJECTION / CONTAINMENT RECIRCULATION The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-3 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS-3 FSAR May 1998 Rev. 20.2 FIGURE 5.4-6 RESIDUAL HEAT REMOVAL SYSTEM PROCESS FLOW DIAGRAM (MODE A)

MPS-3 FSARPage 1 of 5Rev. 21.3NOTES TO FIGURE 5.4-6Mode A Initiation of Residual Heat Removal System Operation This mode presents the process flow conditions for the initiati on of RHS operation. This begins the second phase of plant cooldown, when the reactor coolant temper ature and pressure have been reduced to 350

°F and 375 psig by use of the steam generators , transferring heat to the secondary side. During Mode A operation, one RHS loop is aligned for RCS cooldown and the second loop remains aligned for safety injection (R WST to RHS connectio n is not shown in Figure 5.4-6).One residual heat removal subsystem takes sucti on from its respective RCS hot leg, discharging through the heat exchanger with the return flow routed to the RCS cold legs. During the initial phases of RHS operation, reactor c oolant flow through the heat exch angers is manua lly limited to control the rate of heat removal. The total flow is automatically regulated by flow control valves in the heat exchanger bypass lines to maintain a constant total return flow. The heat removal rate is limited to both control the RCS cooldown rate to 100

°F/hr, based on equipment stress considerations, and to limit component coo ling water temperature to a maximum of 145

°F.During this initial phase of RHS operation, one or two reactor coolant pumps are maintained in operation. This results in a slight RHS return flow imbalance between the four RCS cold legs due to their different operating pressures. In the data presented, reactor coolant pump Number 2 is assumed operating.Mode B Initiation of Second Residual Heat Removal Loop This mode presents the proces s flow conditions once the reactor coolant temperature has been reduced to < 260

°F. The second RHS loop is aligned to ta ke suction from its respective RCS hot leg, discharging through the heat exchanger with return flow routed to the RCS cold legs.

Mode C End Conditions of Normal Cooldown 140

°F This mode presents the proces s flow conditions for the comple tion of RHS operation, refer to Section 5.4.7.2.3.4 for normal cooldown time details.

The flow distribution of this mode, maintains RCS core cooling by controlling reactor coolant flow through the heat exchangers with bypass flow adjustments.

Reactor coolant pump operation has also been terminated at this time, with all RCS cold legs in equ ilibrium.

MODE A INITIATION OF SINGLE TRAIN RESIDUAL HEAT REMOVAL SYSTEM OPERATIONLocation FluidPressure (psig)Temperature

(°F)Flow(gpm)(lb/hr)1 RC 375 350 4186 1.86E + 6 2 RC 372 350 4186 1.86E + 6 3 RC 481 350 4186 1.86E + 6 4 RC 488 3501176 5.24E + 5 5 RC 488 109 1057 5.24E + 5 6 RC 434 283 4005 1.86E + 6 7 RC 372 350 Note 1 8 RC 482 350 2992 1.33E + 6 9 RC 426 283 4005 1.83E + 6 10 RC 428 283 0 11 RC--0 12 RC N/A N/A 0 0 13 RC N/A N/A 0 0 14 RC N/A N/A 0 0 15 RC N/A N/A 0 0 16 RC N/A N/A 0 0 17 RC N/A N/A 0 0 18 RC N/A N/A 0 0 19 RC N/A N/A 0 0 20 RC N/A N/A 0 0 21 RC N/A N/A 0 0MPS-3 FSARNOTES TO FIGURE 5.4-6 Page 2 of 5Rev. 21.3NOTES:1.Miniflow continues until flow at location 3 is greater than 1542 gpm. The miniflow is then closed.

MPS-3 FSARNOTES TO FIGURE 5.4-6 Page 3 of 5Rev. 21.32.The RCS cold leg distribution is a result of operating reactor coolant pump number 2 during this phase of RHR operation.MODE B INITIATION OF SECOND RESIDUAL HEAT REMOVAL LOOPLocation FluidPressure (psig)Temperature

(°F)Flow(gpm)(lb/hr)1 RC 375 260 4120 1.93E + 6 2 RC 372 260 4120 1.93E + 6 3 RC 492 260 4120 1.93E + 6 4 RC 499 260 2016 9.47E + 5 5 RC 498 124 1915 9.47E + 5 6 RC 433 194 4002 1.93E + 6 7 RC 372 260 0 0 8 RC 492 260 2104 9.88E + 5 9 RC 426 194 4002 1.93E + 6 10 RC 428 194 0 0 11 RC--0 0 12 RC 375 260 4120 1.93E + 6 13 RC 374 260 4120 1.93E + 6 14 RC 496 260 4120 1.93E + 6 15 RC 503 260 1836 8.62E + 5 16 RC 502 127 2284 8.62E + 6 17 RC 428 201 4012 1.93E + 6 18 RC 374 260 2284 1.07E + 6 19 RC 497 260 0 0 20 RC 422 201 4012 1.93E + 6 21 RC 423 201 0 0 MODE C END CONDITIONS OF NORMAL COOLDOWN 140

°FLocation FluidPressure (psig)Temperature

(°F)Flow(gpm)(lb/hr)1 RC 0/Note 1 140 2950 1.45E + 06 2 RC 0 140 2950 1.45E + 06 3 RC 147 140 2950 1.45E + 06 4 RC 153 140 2950 1.45E + 06 5 RC 152 101 2921 1.45E + 06 6 RC 34 101 2921 1.45E + 06 7 RC 0 140 0/Note 2 0/Note 2 8 RC 148 140 0 0 9 RC 34 101 2921 1.45E + 06 10 RC 33 101 0 0 11 RC--0 0 12 RC 0/Note 1 140 2950 1.45E + 06 13 RC 0 140 2950 1.45E + 06 14 RC 147 140 2950 1.45E + 06 15 RC 154 140 2950 1.46E + 06 16 RC 148 99 2920 1.46E + 06 17 RC 34 99 2920 1.45E + 6 18 RC 0 140 0 0 19 RC 149 140 0/Note 2 0/Note 2 20 RC 33 99 2920 1.45E + 6 21 RC 32 99 2920 0MPS-3 FSARNOTES TO FIGURE 5.4-6 Page 4 of 5Rev. 21.3NOTES:1.RCS is assumed depressurized with the water level draine d to th e centerline of reactor coolant piping.2.Conservative design assumpti ons presume that the bypass line is isolated; during normal cooldown operations, the bypass line is typically open.

RHRS VALVE ALIGNMENT CHARTValve NumberOperational Mode A B C 1 O O O 2 O O O 3 C C C 4 O *O *O *5 O *O *O *6 C C C 7 C C C 8 C O O 9 C O O 10 O*O*O*11 O C C 12 O*O*O*13 C C C 14 O O O 15 C O OMPS-3 FSARNOTES TO FIGURE 5.4-6 Page 5 of 5Rev. 21.3NOTES: O = Open C = Closed P = Partially open*Valve disc partially closed by means of a permanent "Travel Limiter" on valve actuator.

MPS-3 FSAR Rev. 20.2FIGURE 5.4-7 PRESSURIZER RELIEF TANK MPS-3 FSAR Rev. 20.2FIGURE 5.4-8 PRESSURIZER MPS-3 FSAR Rev. 20.2 FIGURE 5.4-9 RPV SUPPORT SYSTEM GIB KEY VERTICLE RESTRAINT PAD LEVELING UNIT LUB RITE PLATE REACTOR VESSEL SUPPORT PAD__------J FIGURE 5.4-10 LEVELING DEVICE (TYPICAL)RPV 3UPPORT SYSTEM MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT MPS-3 FSAR Rev. 20.2FIGURE 5.4-11 VERTICAL SUPPORTS (TYP ICAL) REACTOR COOLANT PUMPS AND STEAM GENERATOR MPS-3 FSAR Rev. 20.2FIGURE 5.4-12 LATERAL SUPPORTS (TYPICAL) STEAM GENERATOR MPS-3 FSAR Rev. 20.2FIGURE 5.4-13 LATERAL SUPPORT (TYPICAL) REACTOR COOLANT PUMP II k" EL 5j'-4"REF!-l 1;--\I*{;---II------I I EL 51 1-4"REF.FLOOROPENING I II 91-OT DIA.REF----EL 25'-7"PRESSURIZER SKIRT*A RING GIRDER INTERFACE LATERAL SUPPORT(SEESHEET2OF2 FOR DETAI LS), FIGURE5.4-14 (SHEET1OF 2)FRONTVIEWPRESSURIZER SUPPORT VERTICAL SUPPORTS MILLSTONE NUCLEAR POWER PLANT UNIT 3 FINAL SAF ETY ANALYS IS REPORT 41'.a" (f.CONTAINMENT TOP VIEW FIGURE 5.4-15 (SHEET1OF 2)PRESSURIZER SAFETY VALVE SUPPORT SYSTEM MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT