ML17212A045

Millstone Power Station Unit 2 Final Safety Analysis Report, Rev. 35, Chapter 4, Reactor Coolant System

ML17212A045
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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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Millstone Power Station Unit 2 Safety Analysis Report Chapter 4 MPS2 UFSAR 4-i Rev. 35CHAPTER 4-REACTOR COOLANT SYSTEM Table of ContentsSection Title Page4.1GENERAL SYSTEM DESCRIPTION...............................................................4.1-14.2DESIGN BASIS..................................................................................................4.2-1 4.2.1Design Parameters......................................................................................4.2-14.2.2Codes Adhered To......................................................................................4.2-4 4.2.3Quality Control Cl assification....................................................................4.2-64.2.4Part-Loop Oper ation...................................................................................4.2-64.3SYSTEM COMPONEN T DESIGN....................................................................4.3-14.3.1Reactor Vessel............................................................................................4.3-1 4.3.2Steam Generator.........................................................................................4.3-24.3.2.1Flow Induced Vi bration..............................................................................4.3-44.3.2.2Tube Thinning.............................................................................................4.3-54.3.2.3Potential Effects of Tube Ruptures.............................................................4.3-54.3.2.4Composition of Sec ondary Fluid................................................................4.3-64.3.3Reactor Coolant Pumps..............................................................................4.3-64.3.4Reactor Coolant Piping.............................................................................4.3-134.3.5Pressurizer.................................................................................................4.3-13 4.3.6Quench Tank.............................................................................................4.3-16 4.3.7Valves.......................................................................................................4.3-174.3.8Instrumentation A pplication.....................................................................4.3-234.3.8.1Temperature..............................................................................................4.3-234.3.8.1.1Hot Leg Temper ature................................................................................4.3-234.3.8.1.2Cold Leg Temper ature..............................................................................4.3-234.3.8.1.3Surge Line Te mperature...........................................................................4.3-244.3.8.1.4Pressurizer Vapor Phas e Temperature......................................................4.3-244.3.8.1.5Pressurizer Water Phas e Temperature......................................................4.3-244.3.8.1.6Spray Line Te mperature...........................................................................4.3-244.3.8.1.7Relief and Safety Valve Di scharge Temper ature.....................................4.3-244.3.8.1.8Quench Tank Temperatures......................................................................4.3-244.3.8.1.9Reactor Vessel Flange Seal Leakage Temperature...................................4.3-254.3.8.1.10RCS High Point Vents L eakage Temperature..........................................4.3-254.3.8.2Pressure.....................................................................................................4.3-254.3.8.2.1Pressurizer Pr essure..................................................................................4.3-254.3.8.2.2Pressurizer Pr essure..................................................................................4.3-264.3.8.2.3Pressurizer Pr essure..................................................................................4.3-264.3.8.2.4Quench Tank Pr essure..............................................................................4.3-264.3.8.3Level.........................................................................................................

4.3-26 MPS2 UFSAR Table of Contents (Continued)

Section Title Page 4-ii Rev. 354.3.8.3.1Pressurizer Level.......................................................................................4.3-264.3.8.3.2Pressurizer Level.......................................................................................4.3-274.3.8.3.3Quench Tank Le vel...................................................................................4.3-274.3.8.4Reactor Coolant Loop Flow......................................................................4.3-274.3.8.5Reactor Coolant Pump Instrumentation....................................................4.3-274.3.8.5.1Pump Seal Temp eratures..........................................................................4.3-274.3.8.5.2Motor Stator Te mperatures.......................................................................4.3-284.3.8.5.3Motor Thrust Beari ng Temperatures........................................................4.3-284.3.8.5.4Pump Controlled Bleed-O ff Temperature................................................4.3-284.3.8.5.5Antireverse Device Be aring Temperature................................................4.3-284.3.8.5.6Upper and Lower Guide Be aring Temperature........................................4.3-284.3.8.5.7Lube Oil Cooler Inlet a nd Outlet Temperature.........................................4.3-284.3.8.5.8Lower Bearing Oil Temperature...............................................................4.3-294.3.8.5.9Pump Seal Pre ssures.................................................................................4.3-294.3.8.5.10Motor Oil Lift Pressure.............................................................................4.3-294.3.8.5.11Lube Oil Filter Pres sure Differen tial........................................................4.3-294.3.8.5.12Pump Controlled Bleed-Off Flow.............................................................4.3-294.3.8.5.13Lube Oil and Antireverse Devi ce Lube Oil Fl ow Switch.........................4.3-294.3.8.5.14Motor Oil Reservoir Level........................................................................4.3-294.3.8.5.15Vibration Instru mentation.........................................................................4.3-304.3.8.5.16Reverse Rotati on Switch...........................................................................4.3-304.3.8.5.17(Deleted)...................................................................................................4.3-30 4.3.8.5.18RCP Underspeed Reactor Trip.................................................................4.3-304.3.9Reactor Coolant Ve nting System..............................................................4.3-304.3.10Permanent Reactor Cavity Seal................................................................4.3-324.4MATERIALS COMPAT IBILITY......................................................................4.4-14.4.1Materials Exposed to Coolant.....................................................................4.4-14.4.2Insulation

....................................................................................................

4.4-14.4.3Coolant Chemistry......................................................................................4.4-24.5SYSTEM DESIGN EVALUATION...................................................................4.5-14.5.1Prevention of Britt le Fracture.....................................................................4.5-14.5.1.1Initial Nil-Ductilit y Transition Reference Temperature.............................4.5-14.5.1.2Nil-Ductility Transition Refe rence Temperature Shift...............................4.5-24.5.1.3Operational Li mits......................................................................................4.5-34.5.1.4Pressurized Thermal Shock........................................................................4.5-54.5.2Seismic Design...........................................................................................4.5-54.5.2.1Piping..........................................................................................................

4.5-6 MPS2 UFSAR Table of Contents (Continued)

Section Title Page 4-iii Rev. 354.5.2.2Vessels........................................................................................................4.5-64.5.2.3Pumps and Valves.......................................................................................4.5-74.5.3Overpressure Pr otection..............................................................................4.5-84.5.3.1Overpressure Protection Du ring Normal Operation...................................4.5-84.5.3.2Low Temperature Overpressu rization Protection.......................................4.5-84.5.4Reactor Vessel Th ermal Shock...................................................................4.5-84.5.5Leak Detection............................................................................................4.5-94.5.6Prevention of Stainless St eel Sensitization...............................................4.5-104.5.7References.................................................................................................

4.5-154.6TESTS AND INSPECTIONS.............................................................................4.6-14.6.1General........................................................................................................4.6-14.6.2NIL Ductility Transition Re ference Temperature......................................4.6-14.6.3Surveillance Program..................................................................................4.6-44.6.4Nondestructive Tests...................................................................................4.6-64.6.5Additional Te sts..........................................................................................4.6-84.6.6In-Service Insp ection................................................................................4.6-114.ASEISMIC ANALYSIS OF RE ACTOR COOLANT SYSTEM.........................4.A-14.A.1INTRODUCTION.....................................................................................4.A-1 4.A.2METHOD OF ANALYSIS.......................................................................4.A-14.A.2.1General.......................................................................................................4.A-1 4.A.2.2Mathematical Models................................................................................4.A-24.A.2.2.1Reactor Coolant System - Coupled Components......................................4.A-24.A.2.2.2Pressurizer..................................................................................................4.A-3 4.A.2.2.3Surge Line..................................................................................................4.A-34.A.2.3Calculations...............................................................................................4.A-44.A.2.3.1General.......................................................................................................4.A-44.A.2.3.2Frequency Analysis....................................................................................4.A-54.A.2.3.3Mass Point Response Analysis..................................................................4.A-54.A.2.3.4Seismic Reaction Analysis.........................................................................4.A-64.A.3RESULTS..................................................................................................4.A-74.A.4EFFECTS OF THERMAL SHIELD REMOVAL....................................4.A-74.A.5EFFECTS OF REPLACEMENT STEAM GENERATORS.....................4.A-74.A.6CONCLUSION..........................................................................................4.A-84.A.7References..................................................................................................

4.A-8 MPS2 UFSAR 4-iv Rev. 35CHAPTER 4-REACTOR COOLANT SYSTEM List of Tables Number Title4.1-1Reactor Coolant System Volumes4.2-1Principal Design Parameters of Reactor Coolant System4.2-2ATable of Loading Combinat ions and Primary Stress Limits4.2-2BTable of Loading Combin ations and Primary Stress Limits for the Replacement Reactor Vessel Head a nd Replacement pres surizer4.2-3Reactor Coolant Syst em Code Requirements 4.2-4Comparison with Safety Guide 26 4.3-1Reactor Vessel Parameters 4.3-2Steam Generator Parameters 4.3-3Main Steam Safety Valve Parameters 4.3-4Tech Pub Review PKG for T4.3-4R eactor Coolant Pump Parameters4.3-5Reactor Coolant Piping Parameters 4.3-6Pressurizer Parameters 4.3-7Quench Tank Parameter 4.3-8Pressurizer Spray (RC-100E , RC-100F) Valve Parameters4.3-9Power-Operated Relief Valve Isolation Valve Parame ters (RC-403, RC-405)4.3-10Pressurizer Power-Operated Relief Valve Parameters (RC-402, RC-404) 4.3-11Pressurizer Safety Va lve Parameters (RC-200, RC-201) 4.3-12Active and Inactive Valves in the Reactor Coolant System Boundary4.4-1Materials Exposed to Coolant 4.4-2Reactor Coolant Chemistry 4.5-1Reactor Coolant System Component Nozzl es, Nozzle Sizes and Nozzle Materials4.5-2Reactor Coolant System Heatup and Cooldown Limits4.6-1RTNDT Determination for Reactor Vessel Ba se Metal Millstone Unit Number 24.6-2Charpy V-notch and Drop Wei ght Test Values - Pressuri zer Millstone Unit Number 24.6-3Charpy V-notch and Drop Weight Test Values - Steam Generator MPS2 UFSAR List of Tables (Continued)

Number Title 4-v Rev. 354.6-4Charpy V-notch Values - Piping4.6-5Plate and Weld Meta l Chemical Analysis4.6-6Beltline Mechanical Te st Properties - Reactor Ve ssel Surveillance Materials4.6-7Tensile Test Properties - Reac tor Vessel Surveillance Materials4.6-8Summary of Specimens Provided for Each Exposure Location4.6-9Capsule Removal Schedule 4.6-10Inspection of Reactor Coolant Syst em Components Durin g Fabrication and Construction4.6-11Reactor Coolant System Inspection C-E Requirements 4.6-12RT PTS Values at 54 EFPY4.6-13Adjusted Reference Temperatures (ART) Projections 4.A-1Natural Frequencies and Dominant Degrees of Freedom4.A-2Seismic Loads on Reactor Coolant Sy stem Components for Operational Basis Earthquake MPS2 UFSARNOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

4-vi Rev. 35CHAPTER 4 - REACTOR COOLANT SYSTEM List of Figures Figure Title4.1-1P&ID for Reactor Coolant System & Pump (Sheet 1)4.1-2Reactor Coolant System Arrangement-Elevation4.1-3Reactor Coolant Syst em Arrangement-Plan4.3-1Reactor Vessel4.3-2Steam Generator 4.3-3Reactor Coolant Pump 4.3-4P&ID Reactor Coolant Pump 4.3-5Reactor Coolant Pump Seal Area 4.3-6Reactor Coolant Pump Predicted Performance 4.3-7Pressurizer 4.3-8Temperature Control Program 4.3-9Pressurizer Level Setpoint Program 4.3-10Pressurizer Level Control Program 4.3-11Quench Tank 4.3-12Permanent Reactor Cavity Seal Plate 4.5-1Reactor Coolant System Pr essure Temperature Limitations for 7 Full Power Years4.5-2Reactor Coolant System Pressure - Temp Limita tions During Plant Heatup/

Cooldown After 7 Years In tegrated Neutron Flux4.5-3Reactor Coolant System Pressure Temperat ure Limitations For 0 to 2 Years of Full Power4.5-4Reactor Coolant System He atup Limitations for 54 EFPY4.5-5Reactor Coolant System Cooldown Limitations for 54 EFPY4.6-1Location of Surveillance Capsule Assemblies 4.6-2Typical Surveillance Capsule Assembly4.6-3Typical Charpy Impact Compartment Assembly4.6-4Typical Tensile-Monitor Compartment Assembly4.6-5Base Metal - WR (Transverse) Plate C-506-1 Impact Energy vs Temperature MPS2 UFSAR List of Figures (Continued)NOTE: REFER TO THE CONTROLLED PLANT DRAWING FOR THE LATEST REVISION.

Figure Title 4-vii Rev. 354.6-6Base Metal - WR (Transverse) Pl ate C-506-1 Lateral Expansion versus Temperature4.6-7Base Metal - RW (Longitudi nal) Plate C-506-1 Impact Energy versus Temperature4.6-8Base Metal - RW (Longitudinal) Plate C-506-1 Lateral Expansion vs Temperature4.6-9WELD Metal PLATE C-506-2/C-506-3 IMPACT ENERGY vs Temperature4.6-10Weld Metal, Plate C-506-2/C-506-3 Lateral Expansion vs Temperature 4.6-11HAZ Metal, Plate C-506-1 Impa ct Energy versus Temperature 4.6-12HAZ Metal, Plate C-506-1 Lateral Expansion versus Temperature 4.6-13SRM (HSST Plate 01MY - Longitudinal)

Impact Energy versus Temperature4.6-14SRM (HSST Plate 01MY - Longitudinal)

Lateral Expansion versus Temperature4.A-1Reactor Coolant System Seismic Analysis Model MS24.A-1AReactor Coolant System - Seis mic Analysis Model MS2 and RV144.A-2RV14 Reactor and Intern als Seismic Analysis Model4.A-3Pressurizer Seismic Analysis Model 4.A-4Surge Line Seismic Analysis Model MPS2 UFSAR4.1-1Rev. 35CHAPTER 4 - REACTOR COOLANT SYSTEM

4.1 GENERAL

SYSTEM DESCRIPTION The function of the reacto r coolant system is to remove heat from the reactor core and internals and transfer it to the secondary (steam generating) system. The reactor coolant system, which is entirely located within the contai nment building, consists of two heat transfer loops connected in parallel across the reactor pressure vessel. Each loop contains one steam generator, two reactor coolant pumps, connecting piping, and flow and temperature instrumentation. Coolant system pressure is maintained by a pressurizer connected to one of the loop hot legs.

A piping and instrumentation diagram of the reactor coolant system is shown in Figure 4.1-1. The legends for the piping and instrumentation di agram are given in Figures 9.1-1, 9.1-2 and 9.1-3. Elevation and plan views of the reactor coolant system are shown in Figure 4.1-2 and Figure 4.1-3, respectively. During operation, the four pumps circulate water through the reactor vessel where it serves as both coolant and moderator for the core. The heated water enters the two steam generators, transferring heat to the secondary (steam) system, and then returns to the pumps to repeat the cycle. System pressure is maintained by regulating the water temperature in the pressurizer where steam and water are held in thermal equilibrium. Steam is either formed by the pressurizer heaters or condensed by the pressurizer spray to limit the pr essure variations cause d by the contraction or expansion of the reactor coolant. The pressurizer is located with its base at a higher elevation than the reactor coolant loop piping.

This eliminates the need for a separate pressurizer drain, and ensures that the pressurizer is dr ained before maintenance operations.

The Reactor Coolant System (RCS) is protected against overpressure by two ASME Section III Code approved spring-loaded safety valves. In addition, two solenoid-operated power relief valves (PORVs) are provided as described in Section 4.3.7. Both the safety valves and the PORVs are connected to the top of the pressurizer. Steam discharged from the valves is cooled and condensed by water in a quench tank. In the unlikely event that the discharge exceeds the capacity of the quench tank, the tank is relieved via a r upture disc to the cont ainment atmosphere. The rupture disc is provided as the tank code over pressure protection device. The quench tank is located at a level lower than the pressurizer. This ensures that any power-operated relief valve or pressurizer valve leakage from the pressurizer, or any discharge for these valves, drains to the quench tank.

Overpressure protection for the secondary side of the stea m generators is provided by ASME Code safety valves located in the main steam line pipes upstream of th e steam line isolation valves. Power-operated steam dum p and bypass valves are provide d to prevent opening of the secondary safety valves followi ng a loss-of-load incident. The s econdary pressure protection is described in Sections 10.3 and 4.3.2.

To maintain reactor coolant chemistry within the limits discussed in Section 4.4.3 and to control pressurizer level, a continuous but variable bleed flow from one loop upstream of the reactor coolant pump is maintained. This bleed fl ow is controlled by pressurizer level.

MPS2 UFSAR4.1-2Rev. 35 Constant coolant makeup is added by charging pumps in the chem ical and volume control system. Two charging nozzles and one letdown nozzle are provided on the reactor c oolant piping for these operations. An inlet nozzle on each of the four reactor vessel inlet pipes allows injecti on of borated water into the reactor vessel from the safety injection system in the event emergency core cooling is needed.

During a normal plant shutdown, these nozzles are also used to supply shutdown cooling flow from the low pressure safety injection pumps. An outlet nozzle on one reactor vessel outlet pipe is used to remove shutdown cooling flow. Vent and drain connections in the reactor cool ant piping are provided fo r draining the reactor coolant system to the radioactive waste pr ocessing system for maintenance operations. A connection is also provided on the quench tank for draining it to the radi oactive waste processing system following a relief valve or safety valve discharge. Other reactor coolant loop penetrations include sampling connections (Section 9.6) and instrument connections. The nozzle identifications are tabul ated in Figure 4.1-3. Norm al draining of the RCS is through the chemical and volume control system.Where required to reduce heat losses and protect personnel from high te mperatures, components and piping in the reactor coolant system are insulated with a material compatible with the temperatures involved. All insula tion material used on stainless steel has a soluble chloride content of less than 600 ppm to minimize the pos sibility of chloride-induced stress corrosion.

Electroslag welding was not used in the construction of any reactor coolant boundary component.

The major reactor coolant system components are designed for a 40 year service life. To assure that this objective can be attained, strict quality assurance stan dards as outlined in Sections 4.6.4and 4.6.5 were followed.

Protection provided the reactor cool ant system against environmental factors such as fires, floods and missiles is described in other sections (see Chapters 1, 5, 9 and 11).

A tabulation of the RCS volumes is contained in Table 4.1-1.

MPS2 UFSAR4.1-3Rev. 35TABLE 4.1-1 REACTOR COOLANT SYSTEM VOLUMES ComponentVolume (ft 3)Reactor Vessel4652Steam Generators3386Reactor Coolant Pumps449 Pressurizer1500Piping: Hot Leg280Piping: Cold Leg752 Piping: Surge Line32Quench Tank217 MPS2 UFSAR4.1-4Rev. 35FIGURE 4.1-1 P&ID FOR REACTOR CO OLANT SYSTEM & PUMP (SHEET 1)

The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS2 UFSAR4.1-5Rev. 35FIGURE 4.1-1 P&ID FOR REACTOR CO OLANT SYSTEM & PUMP (SHEET 2)

The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS2 UFSAR4.1-6Rev. 35FIGURE 4.1-1 P&ID FOR REACTOR COOL ANT SYSTEM & PUMP (SHEETS 3)

The figure indicated above represents an engineering controlled drawing that is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS-2 FSAR JUN 10 1982 Rev. 24.2 FIGURE 4.1-2 REACTOR COOLANT SYSTEM ARRANGEMENT-ELEVATION MPS-2 FSAR JUN 10 1982 Rev. 24.2 FIGURE 4.1-3 REACTOR COOL ANT SYSTEM ARRANGEMENT-PLAN MPS2 UFSAR4.2-1Rev. 35

4.2 DESIGN

BASIS

4.2.1 DESIGN

PARAMETERS The reactor coolant system operated initially at a core power level of 2560 MWt but has since been uprated to a core power level 2700 MWt. The major systems and components which bear significantly on the acceptability of the site have been evaluate d for operation at a core power level of 2700 MWt (NSSS power of 2715 MWt).

The reactor design described in Chapter 3 predicates hot leg temperature, cold leg temperature, minimum reactor coolant flow and reactor vessel pressure drop. These thermodynamic and hydrodynamic data are used in th e design of the steam generator, reactor coolant pump, and reactor coolant piping as described in S ection 4.3 for each of these components.

The principal design parameters for the reactor coolant system are listed in Table 4.2-1. The design parameters for each of the major components are given in Section 4.3. The reactor coolant system is designated a Class 1 sy stem for seismic design and is de signed to the criteria for load combinations and stresses which are presented in Table 4.2-2A and 4.2-2B. Seismic Analysis is discussed in Appendix 4.A.

The system design temperature and pressure are conservativel y established and exceed the combined normal operating value and the change due to anticipate d operating transients. They include the effects of instrument error and the response characteris tics of the cont rol system. The change due to the anticipated transients also considers the effect of reactor core thermal lag, coolant transport time, system pressure drop and the characteristics of the safety and relief valves.

The following design cyclic tran sients, which include conservati ve estimates of the operational requirements for the components discussed in Section 4.3, were used in the fatigue analyses required by the applicable codes listed in Tabl e 4.2-3; the applicable operating condition category as designated by ASME Section II I is indicated in each case. a.Five-hundred heatup and cool down cycles during the system's 40 year design life at a heating and cooling rate of 100

°F/hr between 70

°F and 532°F.The replacement reactor vessel head is designed for 200 steady state and transient operating cycles for plant heatup and cooldown at a rate of 100

°F/hr between 70

°F and 532°F.

The replacement pressurizer is analyzed for 500 heatup cycles at a rate of 100°FÚhour and 500 cooldown cycl es at a rate of 200

°F/hour.Category: Normal Condition.b.Pressurizer spray piping is limited to 160 plant heatup and cooldown cycles.

Primary manway studs of the replaced st eam generators are limited to 200 heatup MPS2 UFSAR4.2-2Rev. 35 and cooldown cycles at a heat ing and cooling rate of 100

°F/hr between 70

°F and 532°F. c.Fifteen-thousand power cha nge cycles over the range of 15 percent to 100 percent of full load with a ramp load change of five percent of full load per minute increasing and decreasing.

The replacement reactor vessel closure head is anal yzed to 15,000 cycles for plant loading and unloading at 5% of full load/minute.

Category: Normal Condition.d.Primary manway studs for th e replaced steam generators are limited to 1000 cycles with a ramp load change of 5% per minute decreasing and 30% per hour increasing (plant loading/unloading).e.Two-thousand cycles of ten percent of fu ll load step power changes, increasing from an initial power level of 15 and 90 percent of full power and decreasing from an initial power level between 15 and 100 percent of full power

.Category: Normal Condition. f.Ten cycles of hydrostatic testing the reactor coolant system at 31 10 psig and a temperature at least 60

°F above the Nil Ductility Transition Temperature (NDTT) of the component having the highest NDTT. The replacement pressurizer is analyzed to 10 cycles of hydrostatic test at 3125 psia and 70

°F - 400°F, above the minimum RTNDT + 60°F.Category: Test Condition. g.Two-hundred cycles of leak testing at 2485 psig and at a temp erature at least 60

°F greater than the NDTT of the co mponent having the highest NDTT. Category: Test Condition; evaluated as Upset Condition. h.Primary manway studs for the replaced st eam generators are limited to 80 cycles of leak testing at 2485 psig.i.10 6 cycles of normal variations of

+/- 100 psi and

+/- 6 F at operating temperature and pressure. The replacement pressurizer is analyzed for 10 6 cycles of normal variations of

+/- 100 psi and

+/- 7°F at operating temperature and pressure.

MPS2 UFSAR4.2-3Rev. 35 Category: Normal Condition. j.Four-hundred reactor trip s from 100 percent power. Category: Upset Condition. k.Primary manway studs for the replaced steam generators are limited to 80 bolt preloading cycles from unbolted state.

The manway studs for the replaced pre ssurizer are analyzed for 100 bolt/unbolt cycles. The vent port studs for the repl aced pressurizer are analyzed for 200 bolt/

unbolt cycles.

Category: Normal Conditionl.Primary manway studs for th e replaced steam generators are limited to 1500 cycles of 10% of full load step power changes, increasing from an in itial power level of 15% to 90% of full power and decreasing from initial power level between 15%

and 100% of full power

.Category: Normal Conditionm.Primary manway studs for th e replaced steam generators are limited to 200 reactor trips from 100% power

.Category: Upset Condition In addition to the above list of normal design tr ansients, the following abnormal transients were also considered when arriving at a satisfactory us age factor as defi ned in Section III of the ASME Boiler and Pressure Vessel Code; however, emergency condition transi ents were not used to form the basis for the code design of the components based on Paragraph N-417.10(f) of ASME III, 1968 Edition, Summer 1968 addenda.1.Forty cycles of loss of turbine load fr om 100 percent power wi th a delayed, reactor trip. Category: Upset Condition. 2.Forty cycles of total loss of reactor coolant flow when at 100 percent power.

Category: Upset Condition. 3.Five cycles of complete loss of secondary system pressure. Category: Emergency Condition.

MPS2 UFSAR4.2-4Rev. 35The reactor coolant system and its associated c ontrols are designed to accommodate plant step load changes of

+/- ten percent of full power and ramp changes of

+/- five percent of full power per minute without reactor trip. The system will accept, without damage, a complete loss of load.

4.2.2 CODES

ADHERED TO The Codes adhered to and compone nt classifications are listed in Table 4.2-3 and conform to 10 CFR Part 50, Section 50.55a. The construction permit date was December 11, 1970; thus in all instances, code dates identified below meet or exceed those required. The impact properties of all materials which form a part of the pressure boundary meet the re quirements of the ASME Boiler and Pressure Vessel Code Section III, Paragraph N330, at a temperature of 40

°F. The impact properties of the replacement reactor vessel closure head and replacement pressurizer meet the requirements of ASME Section III, NB 2300.

In general, code editions and addenda in effect on the date of the original purchase order to a manufacturer apply in the design, manufacture a nd testing of those co mponents. The code editions and addenda which apply for the components in Tables 1.2-1 and 4.2-3 are specified in applicable FSAR subsections.

Full compliance with Safety Gu ide 26 was not possible since most of the components covered by Safety Guide 26 were purchased and fabrications begun prior to the March 23, 1972, issue date for this guide. Table 4.2-4 provided a comparis on of those components which are not in compliance with Safety Guide 26.The instrument air system for Millstone Unit 2 is not required for safe shutdown. Therefore, Safety Guide 26 does not apply to the design codes for this system.

The codes and standards used for the component s of the diesel oil supply are as follows: Tanks (Above Ground) - API 650Tanks (Below Ground) - NFPA Number 30

Piping - ANSI B31.1.0 (MOD C)

Valves - ANSI B31.1.0 (MOD C)

In addition, seismic ca tegory 1 requirements a nd Quality Assurance Pr ogram as outlined in Appendix 1B (located in th e original FSAR dated August 1972) were employed in the manufacture.

The ANS N18.2 system of quality group classifications has been utilized by the NSSS vendor for Millstone Unit 2. The AEC has voted affirmat ively for the adoption of N18.2 as an ANSI standard. It is more definitive than Safety Guide 26, is supporte d by the major NSSS suppliers and by the utility industry, and compli ance with its provisions provides a satisfactory alternative to the Safety Guide.

MPS2 UFSAR4.2-5Rev. 35The following Code Cases were utilized for materials and construction of components.ASME CODE CASES Steam Generators 1332-4 Requirements for Steel Forgings 1359-1 Ultrasonic Examination of Forgings 1335-2 Requirements for Bolting Materials 1336 Requirements for Nickel-Chromiu m Iron Alloy (all product forms) N-71-13 Component Supports N-10 UT of Pressure Vessel Welds N-20 Steam Generator Tubes N-294-4 Nonwelded Components N-474-1 Inconel 690 Material Reactor Vessel 1335-2 Requirements for Bolting Materials 1336 Requirements for Nickel-Chromiu m Iron Alloy (all product forms) 1359-1 Ultrasonic Examination of Forgings N-4-12 Material Requirements for CEDM motor housing N-525 Material Requirements for Instrumentation Nozzles and Head Vent Nozzle Pressurizer N-405-1 Socket Welds N-2142-2 Classification UNS N06052 Filler Material ANSI B31 CODE CASES Piping 1477-1 Use of 1970 Addenda of ANSI B31.7 70 Design Criteria for Nuclear Piping Under Abnormal Conditions 74 Weld Reinforcement for B31.0 Piping 83 Weld Reinforcement for B31.7 Piping For the most severe loading combination, which includes the Design Basis Earthquake loads, the primary stresses in the ASME Code Section III, Class 2 and 3 components and component supports are limited to levels comparable to the emergency stress limits defined in ASME Code Section III for Class 1 components.

MPS2 UFSAR4.2-6Rev. 35

4.2.3 QUALITY

CONTROL CLASSIFICATION The major components of the reacto r coolant system have been placed in the safety classes as defined by ANS N 18.2 "Safety Criter ia for the Design of Stationary Pressurized Water Reactor Plants." 4.2.4 PART-LOOP OPERATION The maximum temperature of the hot leg and cold leg will be less than the maximum temperatures for design power at design flow. R eactor power operation with less than 4 reactor coolant pumps operating or natura l circulation is not allowed. However, decay heat will be transferred to the steam generator for both cases. Current Technical Speci fications restrictions prohibit other than four reactor coolant pump power operations.

The adequacy of natural circul ation for decay heat removal af ter reactor shutdown has been verified analytically and by tests on the Palisides reactor. The core T in the analysis has been shown to be lower than the normal full power T; thus, the thermal and mechanical loads on the core structure are less severe than normal design conditions.

To assess the margin available in a post-co astdown situation, a st udy was made assuming termination of pump coas tdown 100 seconds after reactor trip, with immediate flow decay to the stable natural circulation condition. It should be recognized that pump rotation will not have stopped for substantially longer than 100 seconds. With the maximum decay heat load 100 seconds after trip, the syst em will sustain stable na tural circulation flow ad equate to give a power to flow ratio of less than 0.9.

Heat removed from the co re during natural circulation may be rejected either by dumping to the main condenser or to the atmosphe re; the rate of heat removal may be controlled to maintain core T within allowable limits. The analytical techniques are verified by tests completed on the Palisades reactor (AEC Docket Number 50-225).

MPS2 UFSAR4.2-7Rev. 35TABLE 4.2-1 PRINCIPAL DESIGN PARAMETERS OF RE ACTOR COOLANT SYSTEMDesign Thermal Power, (NSSS) MWt 2715, Btu/hr 9.26 x 10 9 Design Pressure, psig 2485 Design Temperature (Except Pressurizer, 700

°F), °F 650 Design Coolant Flow Rate, gpm 325,000 (1) Cold Leg Temperature, Normal Service, °F 550 (1) Hot Leg Temperature, Normal Service, °F 604 (1) Normal Operating Pressure, psia 2250 (1) System Volume, ft 3 (Without Pressurizer) 9,551 Pressurizer Water Volume, ft 3 942 at 65% level Pressurizer Steam Volume, ft 3 590 at 65% level (1)RCS piping and vessel design parameters. Vo lumetric flow rate is based on 122 x 10 6 lbm/hr mass flow rate and cold leg density. See FSAR Sec tion 14 for principal RCS parameters used in Safety Analyses.

MPS2 UFSARMPS2 UFSAR4.2-8Rev. 35 Notes: (a) This load combination is not applied to the piping run wi thin which a pipe break is considered to have occurred.(b) For loading combinations 2 and 3, stress limits for vessels, with the symbol P M changed to P L , should also be used in evaluating the effects of local loads impos ed on vessels and/or piping.(c) These stress limits are used for cylindrical structures (e.g., CRDM housings) in the vessel design.(d) See Table 4.2-2B for replacement reactor vessel closure head loading combinations and primary stress limits.

P M = Calculated Primary Membrane Stress P B = Calculated Primary Bending Stress TABLE 4.2-2A TABLE OF LOADING COMB INATIONS AND PRIMARY STRESS LIMITS Loading CombinationsVesselsPrimary Stress Limits PipingSupports Working Stress1.Design Loading + Design Earthquake (OBE)P M < S M , P B + P L 1.5S M P M < S M , P B + P L 1.5S M 2.Normal Operating Loadings +

Maximum Hypothetical Earthquake (DBE). P M S D P M < S D Within YieldSee Note b.P B + P L 2.25S m See Note c.

P B 1.5[1-(P M/S D)2]S D P B 4S D/ cos[P M/ 2S D] 3.Normal Operating Loadings + Pipe Rupture + Maximum Hypothetical Earthquake (DBE).

P M S L P M S L Deflection of s upports limited to maintain supported equipment within

limits shown in columns 1 and 2See Notes a and b.

P B 1.5[1-(P M/S L)2]S L P L + P B 3S M See Note c.P B 4S L/ cos[P M/ 2S L]

MPS2 UFSARMPS2 UFSAR4.2-9Rev. 35 P L = Calculated Primary Local Membrane Stress S M = Tabulated Allowable Stress Limit at Te mperature from ASME Boiler and Pressure Vessel Code,Section III or ANSI B31.7.

S Y = Tabulated Yield at Temperat ure, ASME Boiler and Pressure Vessel Code,Section III.

S D = Design Stress S D = S Y (for ferritic steels)

S D = = 1.2S M (for austenitic steels)

S L = S Y + 1/3 (S u - S Y) S u = Tensile Strength of Material at Temperature(1) From ASME Boiler and Pressure Vessel Code,Section III, 1968 ED., at 650

°F. (2) Minimum value at room temperature which is approximately the same at 650

°F for ferritic materials.

(3) EstimatedThe following typical values are selected to illustrate the conservatism of this approach for establ ishing stress limits. Units are 10 3 lbs/square inch Material S Y (1)S U S M (1)S P S L A-106B 25.4 60.0 (2) 17.025.436.9SA-533B41.4 80.0 (2)26.7 41.4 54.3 SA-508, CL2 41.4 80.0 (2)26.7 41.4 54.3 304 SS 17.0 54.0 (3)15.3 18.4 29.3 316 SS 18.5 58.2 (3)16.7 20.0 31.7 MPS2 UFSAR4.2-10Rev. 35Where: P m = General Primary Membrane Stress Intensity P 1 = Local Membrane Stress Intensity P b = Bending Stress IntensityQ = Secondary Stress Intensity P e = Expansion Stress Intensity U i = Actual Service condition cycles divided by allowable cycles, ba sed on calculat ed alternating stress and fatigue design curve.

S TH = Thermal Stress Range S m = Design Stress Intensity S u = Tensile Strength S y = Yield Stressy' = Maximum allowabl e range of thermal stress on an elastic basis divided by S y.TABLE 4.2-2B TABLE OF LOADING COMBINATIONS AND PRIMARY STRESS LIMITS FOR THE REPLACEMENT REACTOR VESSEL HEAD AND REPLACEMENT PRESSURIZER Loading Combinations ASME Code Subsection Design P m S m NB-3221.1Service Level A (Normal)P 1 1.5S m NB-3221.2Service Level B (Upset)P 1 + P b 1.5S m NB-3221.3 P 1 + P b + Q 3.0S m NB-3222.2 P e 3.0S m NB-3222.3U i 1.0 NB-3222.4 S TH y' S y NB-3222.5Service Level D (Faulted)P m Lesser of 2.4S m and 0.7S u NB-3225 Appendix F P 1 1.5P m NB-3225 Appendix F P 1 + P b 1.5P m NB-3225 Appendix F MPS2 UFSAR4.2-11Rev. 35The following typical values in ks i are selected for the replacement reactor vessel head materials at 650°F.Material Sy Su SmSA508 Grade 3 Class 141.580.026.7ASME Section II, Part DSB 167 (Alloy 690)208020.0ASME C ode Case N525/Section II, Part DSB 166 (Alloy 690)27.580.023.3ASME Section II, Part DSA 182 F316LN17.862.816.0ASME Section II, Part D SA 312 TP316L15.361.713.8ASME Section II, Part D The following typical values in ksi are selected for the pressurizer materials at 650

°F. Material Sy Su SmSA508 Grade 3 Class 253.990.030ASME Section II, Part D SA 182 Grade F 31618.567.016.7ASME Section II, Part D MPS2 UFSAR4.2-12Rev. 35Note:The spare and original safety valves are used interchangeably. Refurbishment and retesting of the safety valves are performed periodically and the safety valves (spare and/or original) are then installed into the system rotationally.TABLE 4.2-3 REACTOR COOLANT SYSTEM CODE REQUIREMENTS Components CodesReactor Vessel (excluding replacement Reactor Vessel Closure Head and Nozzles), Original Upper Shell of Steam Generator 1. ASME,Section III, Class A, 1968 Edition, Addenda through Summer 1969 Replacement Reactor Vessel Closure Head and Nozzles, Replacement Pressurizer

1. ASME Section III, 1998 Edition through 2000 Addenda. Replacement Lower Steam Generator1. AS ME,Section III, Class I, 1983 Edition, Addendum to Summer of 1984 Reactor Coolant Pumps

1. Draft ASME Code for Pumps and Valves for Nuclear Power, Class 1, November 1968, including March 1970 Addenda.

2. ASME Section III, paragraph N153 in Summer 1969 Addenda.

3. ASME Section III, Appendix IX. Quench Tank ASME Section III, Class C, 1968Pressurizer Safety Valves

1. ASME Section III, Class A, 1968 Edition, Addenda through Summer of 1970, Code Case

1344-1 Piping 1. ANSI B31.7, Class 1, 1969 Edition.

2. ASME Section III, paragraph N153 in Summer

1969 Addenda.

3. Code Case 70 to B31.7.Secondary Safety ValvesASME Standard Code for Pumps and Valves for Nuclear Power, Clas s 2, March, 1970 Draft MPS2 UFSARMPS2 UFSAR4.2-13Rev. 35TABLE 4.2-4 COMPARISON WITH SAFETY GUIDE 26 SystemComponents Classification Per Safety Guide 26Applicable Code Per Safety Guide 26 Code Used In ManufactureRequirements Imposed In Addition to Code Requirements Used in ManufactureReactor Building Closed Cooling Water SystemPressure VesselsQuality Group CASM E Section III, Class 3 ASME Section VIII

Division I Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a), Spot radiographed Piping (excluding Containment Penetrations)Quality Group CASM E Section III, Class 3 ANSI B31.1.0 MOD

B Seismic Category I, 10% random

radiography of butt welds for piping 4 inch and larger, material identification, manufactured under Quality Assurance Program of Appendix 1B (a)PumpsQuality Group CASME Section III, Class 3Standards of the Hydraulic Institute All pressure containing parts

where hydrostatically tested at a minimum of 1.5 times the design pressure seismic I manufactured under Quality Assurance Program of Appendix 1B (a)Valves (excluding Containment)Quality Group CASM E Section III, Class 3 ANSI B31.1.0 MOD BSeismic Category I, MT-PT

Examination, material traceability on pressure retaining parts, manufactured under Quality Assurance Program of Appendix 1B (a)

MPS2 UFSARMPS2 UFSAR4.2-14Rev. 35Safety Injection System Shutdown Heat ExchangersQuality Group CASM E Section III, Class 2 ASME Section III, Class 3, TEMA R Seismic Category I Refueling Water TankQuality Group BASM E Section III, Class 2 ASME Section III, Class 3 Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a)Piping HCD(C)Quality Group BASME Section IIIANSI B31.1.0 MOD(C)Seismic Category I, Material

Identification per ASTM SpecificationPumpsQuality Group BASME Section III, Class 2 ASME Code for Pumps and Valves for Nuclear Power, Class II Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a) (See Question 4.4)Valves HCD(C) OnlyQuality Group BASM E Section III, Class 2 ANSI B31.1.0 MOD(C)Seismic Category I, Material

Identification per ASTM Specification Piping 4 inch

HCD-3 and 6 inch HCD-3Quality Group BASM E Section III, Class 2ANSI B31.1.0Seismic Category ITABLE 4.2-4 COMPARISON WITH SAFETY GUIDE 26 (CONTINUED)SystemComponents Classification Per Safety Guide 26Applicable Code Per Safety Guide 26 Code Used In ManufactureRequirements Imposed In Addition to Code Requirements Used in Manufacture MPS2 UFSAR 4.2-15 Rev. 35MPS2 UFSAR Auxiliary Feedwater System Condensate Storage TankQuality Group BASM E Section III, Class 2AWWA D100 NFPA Volume 6 Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a) later modified to API 620 or equivalent based on Design Change to a pressurized tank Piping (excluding Containment

Penetrations)Quality Group BASM E Section III, Class 2 ANSI B31.1.0 MOD B Seismic Category I, 10% random

radiography of butt welds for piping 4 inch and larger, material

identification manufactured under Quality Assurance Program of Appendix 1B PumpsQuality Group BASME Section III, Class 2 ASME Code for Pumps and Valves for Nuclear Power, Class II Seismic Category I manufactured under Quality Assurance Program of Appendix 1B (a)Valves (excluding Containment

Isolation)Quality Group BASM E Section III, Class 3 ASNE B31.1.0 MOD BSeismic Category I, MT-PT

Examination, material traceability on pressure retaining parts, manufactured under Quality Assurance Program of Appendix 1B (a)TABLE 4.2-4 COMPARISON WITH SAFETY GUIDE 26 (CONTINUED)SystemComponents Classification Per Safety Guide 26Applicable Code Per Safety Guide 26 Code Used In ManufactureRequirements Imposed In Addition to Code Requirements Used in Manufacture MPS2 UFSAR 4.2-16 Rev. 35MPS2 UFSARService Water SystemPipingQuality Group CA SME Section III, Class 3ANSI B31.1.0Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a)PumpsQuality Group C, Class 3ASME Section IIIASME Section VIIIPerform Test according to ASME PTC 8.2 1965, Seismic Category I manufactured under Quality

Assurance Program of Appendix 1B (a)ValvesQuality Group CA SME Section III, Class 3ANSI B31.1.0Seismic Category I, manufactured under Quality Assurance Program

of Appendix 1B (a)Vital Chilled Water SystemPipingQuality Group CA SME Section III, Class 3ANSI B31.1.0Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a)ValvesQuality Group CA SME Section III, Class 3ANSI B31.1.0Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a)Condenser/

EvaporatorsQuality Group CASM E Section III, Class 3ASME Section VIII, Division I Seismic Category I, manufactured under Quality Assurance Program of Appendix 1B (a)(a)Appendix 1B was located in the original FSAR dated August 15, 1972.TABLE 4.2-4 COMPARISON WITH SAFETY GUIDE 26 (CONTINUED)SystemComponents Classification Per Safety Guide 26Applicable Code Per Safety Guide 26 Code Used In ManufactureRequirements Imposed In Addition to Code Requirements Used in Manufacture MPS2 UFSAR4.3-1Rev. 35

4.3 SYSTEM

COMPONENT DESIGN

4.3.1 REACTOR

VESSELThe reactor vessel (Figure 4.3-1) is supported by three pads welded to the underside of the reactor vessel nozzles. The arrangement of the vessel supports, allows radial growth of the reactor vessel due to thermal expansion while ma intaining it centered and restra ined from movement caused by seismic disturbances. Departure fr om levelness of not more than 0.002 inch pe r foot of flange diameter is maintained during cons truction to facilitate proper asse mbly of reactor internals. The design parameters for the reactor vessel are given in Table 4.3-1.The vessel closure flange is a forged ring with a machined ledge on the inside surface to support the reactor internals and core. No other ring forgings are used for reactor vessel shell sections. The flange is drilled and tapped to receive the closure studs and is machined to provide a mating surface for the reactor vessel cl osure seal. The vessel closure contains 54 studs, 7 inches in diameter, with eight threads per inch. The stud material is ASTM A-540, Grade B24, with a minimum yield strength of 130,000 psi. The tens ile stress in each stud when elongated for operational conditions is approxima tely 40 ksi. Calculations show that 32 uniformly distributed studs can fail before the closure will separate at design pressure. However, 16 uniformly distributed broken studs or four adjacent broken studs will cause O-ring leakage.

Six radial nozzles on a common pl ane are located just below th e vessel closure flange. Extra thickness in this vessel-nozzle course provides most of the reinforcement required for the nozzles.

Additional reinforcement is provided for the individual nozzle attachments. A boss located around each outlet nozzle on the insi de diameter of the vessel wa ll provides a mating surface for the internal structure which guides the outlet coolant flow. This boss a nd the outlet sleeve on the core support barrel are machined to a common contour to reduce core bypass leakage. A fixed hemispherical head is attached to the lower end of the shell. Ther e are no penetrations in the lower head.The removal top closure head is hemispherical.

The closure head is single piece low alloy steel forging replaced during refueling out age 16. All surfaces in contact with reactor coolant are clad with a quarter inch nominal thickness weld deposi t similar to type 304 stai nless steel. The nozzles in the reactor vessel head suppor t the Control Element Drive M echanisms (CEDMs). The CEDM, In-Core Instrumentation (ICI) and vent nozzles are constructed from Inconel alloy 690 material to minimize the susceptibility to Primary Water Stress Corrosion Cr acking. Threaded housing flanges made of stainless steel are joined to the guide tubes by Gas Tungsten Arc Welding. The head flange is drilled to matc h the vessel flange stud bolt locati ons. The 54 stud bolts are fitted with spherical washers located between the cl osure nuts and head flange to maintain stud alignment during head flexing due to boltup. To ensure uniform lo ading of the closure seal, the studs are tensioned with hydraulic stud tensioners. Stud elongation is then measured to ensure proper preload on all the studs.Flange sealing is accomplished by a double-seal arrangement utilizing two silver-plated Ni-Cr-Fe alloy, self-energized O-rings. The space between the two rings is monitored to allow detection of any inner ring leakage. The control element drive mechanism (CEDM) nozzles (Ni-Cr-Fe alloy MPS2 UFSAR4.3-2Rev. 35 through the head, stainless steel fl ange) terminate with threaded and seal-welded flanges at the upper end. There are eight instrume ntation nozzles with Grayloc c onnectors to maintain pressure boundary. In addition to these nozzles, there is a three-quarter inch vent connection.The core is supported from an internally machined core support ledge.

4.3.2 STEAM

GENERATOR The nuclear steam supply system (NSSS) utilizes two steam genera tors (Figure 4.3-2) to transfer the heat generated in the reac tor coolant system (RCS) to the secondary system and produce steam at the warranted st eam pressure and quality. The design parameters for the steam generators are given in Table 4.3-2.

The steam generator is a vertical U-tube heat exchanger. The st eam generator operates with the reactor coolant in the tube side and the secondary fluid in the shell side.

Reactor coolant enters the steam generator thr ough the inlet nozzle, flows through three-quarter inch OD U-tubes, and leaves through two outlet nozzles. Vertical partition plates in the lower head separate the inlet and outlet plenums. The pl enums are stainless steel clad, while the primary side of the tube sheet is Ni-Cr-Fe clad. The vertical U-tubes are Ni-Cr-Fe alloy. The tube-to-tube sheet joint is welded on the primary side.

Feedwater enters the steam gene rator through the feedwater nozzle where it is distributed via a feedwater distribution ring having top discharge "J" nozzles whic h direct the flow into the downcomer. The downcomer is an annular passage formed by the inner surface of the steam generator shell and the cylindrical shell wr apper which encloses the vertical U-tubes.At the bottom of the downcomer, the secondary wate r is directed upward past the vertical U-tubes where it is boiled to produce steam in the evaporator. The heat tran sfer area is determined by the required heat transfer, the thermal driving force and the heat transfer coefficient. The heat transfer coefficient in the evaporator is ca lculated from e xperimental data.

Upon exiting from the vertical U-tube heat transfer surface, the steam water mixture enters centrifugal-type separators. Thes e impart a centrifugal motion to the mixture and separate the steam from the water. The water leaves the primary se parator through the bottom of the separator housing and is directed in to the downcomer where it is mixed with the feedwater. Final drying of the steam from the centrifugal pr imary separators is accomplished by directing the steam through secondary cyclones. The moisture content of th e outlet steam is no greater than 0.2 percent at design flow.

The steam generator primary side pressure loss is determined by summing the losses due to friction in the tubes, in the t ube bends, entrances and exits, a nd the steam gene rator inlet and outlet plenums and nozzles.

The steam generator shell is constructed of carbon steel. Manways and handholes are provided for easy access to the steam generator internals.

MPS2 UFSAR4.3-3Rev. 35The power-operated steam dump valves and steam bypass valves obviate opening of the main steam safety valves following turbine and reactor trip from full power. The steam dump and bypass system is described in Section 10.

Overpressure protection for the sh ell side of the steam generators and the main steam line piping up to the inlet of the turbine stop valve and provided by 16 spring-loaded ASME Code safety valves which discharge to atmosphere. Eight of these safety valves are mounted on each of the main steam lines outsid e the containment upstream of the steam lined isolation valves. The opening pressure of the valves is set in accordance with ASME Code allowances. Parameters for the main steam safety valves are given in Table 4.3-3.Instrumentation has been added to each safety relief valve (SRV) to provide a main control board annunciator alarm indication of valve closed/not closed, (Regulatory Guide 1.97, Rev. 2, D-18 variable).

The main control board annunciator window (C05 D17A/B) is a split wi ndow and will alarm when any of the SRV's on Steam Generator 1 or 2, respectively, are "not closed."

Upon receiving this alarm, the ope rator would use the control room alarm typer or the control room cathode ray tube (CRT) display showing the SRV "graphics," to determine which valve on Steam Generator 1 or 2 is not closed.

The steam generators are vertically mounted on be aring plates which allow lateral motion due to thermal expansion of the reactor coolant piping. Stops are provided to limit this motion in case of a coolant pipe rupture.

The top of each unit is restrain ed from sudden lateral movement by suitable stops and hydraulic snubbers mounted rigidly to the concrete structure.

In addition to the transients listed in Section 4.2.1, each steam gene rator is also designed for the following conditions such that no component is stressed beyond the allowa ble limit as described in ASME Code,Section III (Table 4.3-2):a.Four-thousand cycles of transient pres sure differentials of 85 psi across the primary head divider plate due to starting and stopping the primary coolant pumps (RCP).Category: Normal Condition.b.Ten cycles of hydrostatic testing of the secondary side at 1235 psig, the primary side is at atmospheric pressure.Category: Test Condition.

MPS2 UFSAR4.3-4Rev. 35c.Two-hundred cycles of leak testing of th e secondary side at 985 psig, the primary side will be pressurized to 165 psig.Category: Test Condition.d.Fifteen-thousand cycles of adding 600 gpm of 70

°F feedwater with the plant in hot standby condition.

Category: Normal Condition.

In addition to the normal design transients li sted above, and those li sted in Section 4.2.2, the following additional abnormal tran sient was also considered in arriving at a satisfactory usage factor as defined in Sect ion III of the ASME Code:

Eight cycles of adding a maximum of 650 gpm of 70

°F feedwater with the steam generator secondary side dry and at 620

°F.Category: Emergency Condition.

The unit is capable of withsta nding these conditions for the pres cribed numbers of cycles in addition to the prescribed opera ting conditions without exceeding the allowable cumulative usage factor as prescribed in ASME Code,Section III.The steam generators are locate d at a higher elevation than the reactor vessel. The elevation difference creates natural circulation sufficient to remove core decay heat following coastdown of all RCPs.The steam generators are equipped with a nitrogen addition system which has the capability of admitting N 2 to the bottom blowdown headers to mix the chemicals in the steam generators during wet layup, and also thr ough nozzles in the transition c ones for blanketing the steam generators and steam lines above the water level.

A flowmeter provides nitr ogen flow control and visual verification that flow is occurring. A drain line on the steam generator upper vent line allows checking that the steam generator has not been flooded. A test gauge may be installed on this line to measure steam generator overpressure.

In addition to the transients listed in this se ction, and those in Section 4.2.1 the following factors were considered in the design of the steam generators.

4.3.2.1 Flow Induced VibrationThe steam generator has also been design ed to ensure that crit ical vibration fre quencies will be well out of the range expected during norma l operation and during a bnormal conditions. The steam generator tubing, tube sl eeves, and tubing supports are de signed and fabricated with considerations given to both sec ondary side flow induced vibrati ons and RCP induced vibrations.

In addition, the heat transfer tubi ng, tube sleeves, and tube supports are designed so that they will MPS2 UFSAR4.3-5Rev. 35 not be structurally da maged under the loss of secondary pre ssure conditions that may produce a fluid velocity in the tube bundle four times design velocity.

Because the RCPs have a rotati onal speed of 900 rpm (less normal slip) th e tube bundle design has considered the imposition of exciting freque ncies of 14 to 15 cps and 70 and 75 cps. The lower frequency range is defined as a mechanical vibration result ing from the transmission of a mechanical impulse at the freque ncy of pump rotation. The upper fre quency range is defined to be a sinusoidal pressure variation of

+/-6 psi in the primary piping that contains the pump. The pressure variation results from the impeller vanes interacting wi th the cutwater vane at the volute outlet during each revolution of the impeller.

It has been found that all tubes and tube sections that will experience fo rcing functions from cross flow and parallel flow have natural frequencies sufficiently diff erent from the fr equency of the forcing function that they will not experience damagi ng vibrations. The mechanical excitation frequency is sufficiently different from the lo west natural frequency for out-of-plane or lateral vibration in any tube span that critical vibration will not occur.

4.3.2.2 Tube Thinning The original Combustion Engineer ing (CE) Steam Generators' marg in of tube-wall thinning that could be tolerated without ex ceeding the allowable faulted stress limits under postulated condition of a design basis largest pipe break in the reactor coolant pr essure boundary (RCPB) during reactor operation was 0.008 inches.

0.0012 inch excess material had been intentionally been provided in the tube wall thickness to accommodate the estimated degr adation of tubes during the service lifetime. CE expected negligible tube wall thi nning when operating under the sp ecified secondary chemistry requirements.

Because of the use of volatile secondary water chemistry there has been no tube wall thinning experienced on the steam generators. The new steam generator tubes are designed to be at least structurally equivalent to the original and in compliance to Regulat ory Guide 1.121. Therefore, tube wall thinning or othe r forms of tube degrada tion can be structurally accommodated with the same degree of margin as the original.

4.3.2.3 Potential Effects of Tube Ruptures The steam generator tube rupture accident in a penetrat ion of the barrier between the RCS and the main steam system. The integrity of this barrier is significant from the standpoint of radiological safety in that a leaking steam generator tube allows the transf er of reactor cool ant into the main steam system. Radioactivity contai ned in the reactor coolant woul d mix with water in the shell side of the affected steam generator. This ra dioactivity would be tran sported by steam to the turbine and then to the condenser, or directly to the condenser via the main steam dump and bypass system. Noncondensible radioactive gases in the condenser are removed by the condenser air ejector system and discharged to the plant ve nt. Analysis of a steam generator tube rupture accident, assuming complete severance of a tube, is presented in Section 14.14.

MPS2 UFSAR4.3-6Rev. 35 4.3.2.4 Composition of Secondary Fluid Radioactivity concentrations in the secondary si de of the steam genera tor is dependent upon the radioactive concentration of th e RCS, the primary to secondary leak rate, and the operating history of the steam generator blowdown system.

4.3.3 REACTOR

COOLANT PUMPS The reactor coolant is circulated by four single speed, vertical, single suction, centrifugal type pumps (Figure 4.3-3). The discharge nozzle is horiz ontal and the suction no zzle is in the bottom vertical position. The pressure co ntaining components ar e designed and fabricat ed in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Class A.

The design flow for the RCP is determined from the reactor mass flow. This mass flow is converted to volumetric flow at the full power cold inlet temperature to determine the pump design flow. The maximum pressure loss at the design flow rate for the reactor vessel, steam generator, and piping is determin ed by adding an allowance for uncertainty to the best estimate for each pressure drop. These maximum values ar e used to establish the RCP design head. The RCP is designed to produce the mi nimum reactor design flow at the maximum expected system pressure loss.The minimum RCS pressure at a ny given temperature is limited by required net positive suction head (NPSH) for the RCP during por tions of plant heatup and cooldown. To ensure that the pump NPSH requirements are met unde r all operating conditions, an ope rating curve is used which gives permissible RCS pressure as a function of reac tor coolant temperature. The RCP NPSH restriction on this curve is determined by using the NPSH requirement for each pump combination and correcting it fo r pressure and temper ature instrument errors and pressure measurement location. The NPSH required versus pump flow is supplied by the pump vendor.

Plant operation below this curve is prohibited. At low RCS temperatures and pressures, other considerations require the maxi mum pressure versus temperatur e curve to be above the NPSH curve.The pump impeller is pinned and bolted to the shaft.

A close clearance thermal barrier assembly is mounted above the water lubricated bearing to reta rd heat flow from the pump to the seal cavity which is located above the thermal barrier. The th ermal barrier assembly al so tends to isolate the hot fluid in the pump from the cool er fluid above and, in the event of a seal failure, serves as an additional barrier to reduce leakage from the pump. Each pump is equi pped with replaceable casing wear rings. A water lubricated bearing is located in the fluid between the impeller and thermal barrier to provide shaft support. Additiona l shaft support is provi ded by bearings in the electric motor which is directly connect ed to the pump shaft by a rigid coupling.

The shaft seal assembly located above the thermal barrier consists of four, face type mechanical seals, three full-pressure seals mounted in tande m and a fourth low pres sure backup vapor seal designed to withstand operating system pressure with the pump stopped. The performance of the

shaft seal system is monitored by pressure and temperature sens ing devices in the seal system (Figure 4.3-4). A controlled bleed of f flow through the pump seals is maintained to cool the seals MPS2 UFSAR4.3-7Rev. 35and to equalize the pressure drop across each seal. The controlled bleed-off flow is collected and processed by the chemical and volum e control system. Any minor leak age past the vapor seal (the last mechanical seal), is drained to the containment sump via the containment trench and collected in the radioactive waste processi ng system. Normal vapor seal leakage is minor, (approximately 0 to.08 GPM per RCP), and is consider ed to be negligible leakage to the containment atmosphere.

The seals are cooled by circulating the controlled bleed off through the heat exchanger mounted integrally within the pump cover assembly. No damage would result in the event of pump operation without cooling water for up to five minutes. To reduce plant downtime and personnel exposure to radiation during seal maintenance, the seal system is contained in a cartridge which can be removed and replaced as a unit. The seal cartridge can be repl aced without draining the pump casing. The seal detail is shown in Figure 4.3-5.A motor-mounted flywheel reduces the rate of flow decay upon loss of pump power. The combined inertia of the pump mo tor and flywheel is 100,000 lbm-ft

2. Flow coastdown characteristics are disc ussed in Section 14.6.

The RCPs are typical centrifuga l volumetric flow machines. Th e pump response following a Loss-of-Coolant Accident (LOCA) is predicted using generally acce pted methods as described in Appendix 1 of CENPD-26. CENPD-26 is a propri etary report entitled "C ombustion Engineering Analytical Techniques for Evalua ting Loss of Coolant Accidents".

A spectrum of breaks in the RCP discharge line have been analyzed and the results follow a predic table pattern. Assuming loss of electrical power to the pump at the start of the LOCA, it is seen that the pumps initially lose speed because the volumetric flow through the pump is not sufficient to sustain the nominal speed of rotation. The volumetric flow increases during the transient, accel erating the pump to its maximum speed. The extent of the in itial loss of speed varies with the break size. The larger the break size, the less the initial deceleration and the higher the maximum speed attained. The pump attains its maximum speed following a double-ended discharge break. The calculated torque imposed on the impeller follows the same trend as the speed, wi th the maximum value occurring following the double-ended discharge break.

The need for a disengaging device to prevent motor overspeed following a LOCA has been evaluated. In view of the fact that the maximum anticipated pump speed is well within the safe operating limits for all rotating parts, a means to disengage th e motor from the pump is not necessary.

A break in the suction piping caus es the reactor coolant to flow through the pump opposite to the normal direction of flow, decelera ting the rotation of the pump until it is brought to rest against the anti-reverse rotation device.

The pump/motor assembly includes motor bearing oil coolers, seal chamber, controls and instruments. Cooling water is provided from the reactor building closed cooling water (RBCCW) system. The design parameters for the RCPs are given in Table 4.3-4. The RCP instrumentation is described in Section 4.3.8.5.

The RCP and motor are supported by four support lugs weld ed to the volute. The pump is supported by four spring assemb lies employed between the support lugs and the floor below.

MPS2 UFSAR4.3-8Rev. 35 Movement in the horizontal plane to compensate for pipe thermal grow th and contraction is permitted. Vertical movement is not restrained.

The major pump components wetted by the primary fluid are constructed of austenitic stainless steel to minimize corrosion. These materials are listed in Table 4.3-4. The mech anical seals consist of a rotating tungsten (1) carbide ring riding over a hard carbon face. The desi gn life of this seal arrangement is at least 50,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or 5.7 years. Each seal is desi gned to accept a pressure drop equal to full operating system pressure, but normally operates at one-third this pressure drop.

The predicted pump performance curve is shown in Figure 4.3-6. The air-cooled, self-ventilated pump motor is sized for continuou s operation at the flows resulting from four-pump operation or partial pump operation with 0.76 specific gravity water. The motor service factor is sufficient to allow 500 heatup cycles during which the nominal horsepower lo ad will decrease from 6000 to 4500 over a period of seven hours. The motors are designed to start and accelerate to rated speed under full load when 70 percent or more of rated normal volta ge is applied. The motors are contained within standard drip-proof enclosures and are equi pped with electrical insulation suitable for a zero to 100 percent humidity and radiation environment of 30 R/hr.

The design requirements of the RCPs include a minimum inertia for the rotating assembly of 100,000 lbm

- ft 2; to achieve this total, a flyw heel with an in ertia of 70,000 lbm

- ft 2 has been incorporated.

Each original RCP motor flywheel assembly consists of two solid discs bolted together, shrink fitted onto and keyed to the shaft above the rotor. The dimensions of each disc are:Outside diameter, inches 75 Thickness, inches 6 Weight, each, lb 7,250 The selection of material, machining and manuf acturing operations, quality control, and the rigorous acceptance criteria establ ished to assu re the integrity of the flywheel and to minimize operating stresses include the following:A.The principal stress is 30 percent of the yiel d point of the flywheel material (based on the tensile tests per ASTM-A-20) at the desi gn overspeed of 125% of the normal operating speed not considering keyway stress concentr ation factors. The mi nimum keyway fillet radius is one eight inch.B.The bore in the flywheel was flame cut, with a minimum of one half inch of stock left on the radius for machining to final dimensions.C.All flywheel discs have passed the following nondestructive testing:

(1) or silicon MPS2 UFSAR4.3-9Rev. 351.Testing by steel vendor prior to machininga.One-hundred percent Ultrasonic In spection per ASME Code,Section III, paragraph N-321.1b.One-hundred percent Magnetic Particle Examination per ASME Code,Section III, paragraph N-322.2.2.Testing after finish machininga.One-hundred percent Ultrasonic Insp ection of the flats and edges, performed in compliance with ASME Code,Section III, paragraph N-321.1b.Liquid penetrant inspection performe d in compliance with ASME Code,Section III, paragraph N-322.3 on the bore and each side of each disc for eight inches radially from the bore.D.The finish of the flywheel bore and the finish on each side of each disc for eight inches radially from the bore are held free of nicks, center punch marks, stencil marks, holes or other stress concentrations.E.Welding was not performed on flywheel discs.F.The keyway fillet radius on the bottom of the keyway is 0.125 inches minimum.

The flywheel material, which is pressure vessel quality , vacuum-imp roved steel plate, exceeds the requirements of ASTM-A-516, Grade 70 ev en though it was originally specified for ASTM-A-516 Grade 65. To improve the fracture t oughness properties of th e material, the flame cut discs, with the one-half inch allowance for machining, were heat treated as follows:1.heated to 1650

°F +/-25°F and held for minimum of 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />;2.water quenched to below 400

°F;3.tempered at 1140

°F for one-half hour per inch of thickness, air cooled.

The composition of the material as certif ied by the steel vendor is as follows:ASTM-E-30 Melt NumberASTM-E-30 Slab Number C (Weight %)

Mn (Weight %)

P (Weight %)

S (Weight %)Si (Weight %)B37253 & 4 0.210.970.0080.0250.23A817620.221.160.0060.0200.24 MPS2 UFSAR4.3-10Rev. 35 The grain size of this silicone deoxidized steel is ASTM-E-112 size 7-8. Th e tensile properties of the flywheel material as certified by the steel vendor are as follows: Tensile tests per ASTM-A-20 The Charpy values measured for the flywheel ma terial at 40

°F are substantiall y higher than the data compiled on SA-516 grade 70 material by the Research & Pr oduct Development Department of CE. The report titled "Longitudinal and transverse Charpy V-notch imp act and dropweight test data for normalized and tempered SA-516 gr ade 70 material" issu ed on August 26, 1971, and further identified by Laboratory Number X-24053 and R&PD Projec t Number 420001, was prepared for the Industrial Coopera tive Program of the Material Division of the Pressure Vessel Research Committee of the Welding Research Council.

This indicates that the toughness pr operties of these wheels are bett er than typical SA-516 Grade 70. Therefore, the nil-ductility transition (NDT) temperat ure is lower than the highest value of

-10°F reported in that report.

Since the normal operating te mperature of the flywheel is approximately 100

°F, a substantial margin exists between the computer K I for large hypothetical cracks, and the toughness K IC. The critical crack size therefore, is greater th an five inches from the bore of the wheel.

Crack growth calculations indicate that the number of starting cycles to cause a reasonably small crack to grow to critical size is orders of magnitude greater than the number of cycles expected during the life of Millstone Unit 2.The loads that are considered for the calculation of the stresses in the flywheel are the combined primary stresses in the flywheel at normal operating speed. They include the stress due to interference fit on the shaft as well as the stress due to centrifugal force.

Normal operating speed of the fl ywheel is 900 rpm (les s normal slip). The flywheel has a design overspeed of rated rpm plus 25 percent, wh ich equals 1,125 rpm. Th e maximum tangential stresses in the flywheel at normal operating speed are 25 percent of the material yield strength.

The maximum tangential stresses in the flywhe el at design overspeed are 30 percent of the material yield strength.

Drop weight test (DWT) - test at +40

°F per ASTM-E-208.Melt NumberSlab NumberTensile Strength PSI0.2% Offset Yield Strength PSIElongation in 2 inch (%)Charpy V-notch at 40°F (ft-lbf)DWTB3725376700490003010410991OKB372547560050700311039597OK A81762767005150029847383OK MPS2 UFSAR4.3-11Rev. 35 The RCP flywheels are accessibl e for 100 percent in place volum etric ultrasonic examination during in-service inspection. Two (2) access panels, 180

° apart, are provide d on the outside of each RCP motor.

The replacement RCP motor flywheel is a one piece forging, 75 inches OD by 12 inches thick, shrink fitted onto and keyed to the motor shaft above the rotor.The replacement RCP motors were procured as Quality Assura nce items in accordance with 10 CFR 21 and 10 CFR 50 Appendix B. They are fully interchangeable with the original RCP motors.The selection of material, machining and manuf acturing operations, quality control, and the rigorous acceptance criteria established to assure the integrity of the flywheel and to minimize operating stresses include the following:A.The principal stress does not ex ceed 30 percent of the yield poi nt of the flywheel material (based on the tensile tests pe r ASTM-A-370) at the design ove rspeed of 125 percent of the normal operating speed not considering keyw ay stress concentration factors. The minimum keyway fillet ra dius is one-eighth inch.B.If the bore in the flywheel was flame cut, a minimum of one-half inch of stock was left on the radius for machining to final dimensions.C.All flywheel discs passed the following nondestructive testing:

Prior to machining:a.The flywheel material was subjected to a 100 percent volumetric ultrasonic examination using procedures and acceptan ce criteria as specified in Paragraphs NB2532.1 and NB2532.2 of the ASME B&PV Code,Section III.The composition of the materi al as certified by the steel vendors is as follows: The heat treatment was as follows:1.Heated to 1,560

°F to 1,580

°F and held for a minimum of 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Chemical Composition (weight percent)Heat NumberCSiMnPSCrMoNiV3206470.160.040.230.0040.0031.600.433.600.015151000.200.080.330.0070.0051.610.443.540.03 MPS2 UFSAR4.3-12Rev. 352.Water quenched.3.Tempered at 1,110

°F to 1,185

°F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, air cooled.Tensile tests were performed on the flywheel material per ASTM A370.

The NDT temperature of the flyw heel material, as obtained from DWT performed in accordance with the specification ASTM E 208 was no higher than -30

°F.The Charpy V-notch (C v) upper-shelf energy level in the "weak" direction of the flywheel material was at least 50 ft-lbs. A minimum of three C v specimens were tested from each forging in accordance with ASTM A 370.

The mechanical propertie s of the flywheel material as ce rtified by the steel vendors are as follows: Fracture Toughness The minimum static fracture toughness of the mate rial at the normal opera ting temperature of the flywheel was equivalent to a critical stress intensity factor, K IC , of at least 150-ksi in. Compliance was demonstrated by either of the following:a.Testing of the actual material to establish the K IC value at the normal operating temperature.b.Determining that the normal opera ting temperature is at least 100

°F above the RTNDT.In-service inspection includes 100 perc ent ultrasonic examination of th e flywheel of one (1) of the four (4) RCPs during each inspection interval. The acceptance criteria is in accordance with ASME - Boiler and Pressure Vessel C ode,Section III, for Class I vessels.

Mechanical PropertiesHeat NumberTensile Strength (psi)0.2% Offset Yield Strength (psi)Elongation in 2 inches

(%)Charpy V-notch at 20

°F Energy (ft-lb f)Drop Weight Test at -30

°F320647110,00093,70021.713012692.9No break515100112,00096,00022.4140130122No break MPS2 UFSAR4.3-13Rev. 35

4.3.4 REACTOR

COOLANT PIPING The RCS piping consists of two l oops which connect the steam gene rators to the reactor vessel.

Each loop can be considered to consist of 42 inch ID "hot le g" piping connecting the reactor vessel outlets to the steam generator inlets a nd 30 inch ID piping connecting the steam generator outlets to the RCPs and the coolant pumps to the reactor vessel inlets. The two 30 inch piping segments are referred to as the "pump suction leg" and the "cold leg" respectively. A 12 inch schedule 160 surge line connects one loop hot leg to the pressurizer. Desi gn parameters for the reactor coolant piping are given in Table 4.3-5.

The reactor coolant piping is desi gned and fabricated in accordanc e with the rules and procedures of ANSI B31.7, Class I. The antici pated transients listed in Section 4.2.1 form the basis for the required fatigue analysis to ensure an adequate usage factor.

The reactor coolant piping is fabricated from SA 516 Gr 70 carbon steel mill clad internally with roll bonded type 304L stainless stee

l. A minimum clad thickness of one-eighth inch is maintained. The 12 inch surge line is fabricated from ASTM A351 Gr CF8M alloy steel.Thermal sleeves are installed in the surge line nozzle, charging nozzles and shutdown cooling inlet nozzle to reduce thermal shock effects from auxiliary system. Clad sections of piping are fitted, where necessary, with safe ends for field welding to stainless steel components.

In response to industry experience, half nozzle replacements have been performed on selected instruments and sampling nozzles as a mitigati on technique against pre ssurized water stress corrosion cracking (PWSCC).

The piping is shop fabricated and shop welded into subassemblies to the greatest extent practicable to minimize the amount of field welding. Fabrication of piping and subassemblies is done by shop personnel experienced in making large heavy wall welds. Welding procedures and operations meet the requirements of Section IX of the ASME Boiler and Pressure Vessel Code.

All welds are 100 percent radiogra phed and liquid-penetrant or magne tic-particle tested and all reactor coolant piping penetrations are attached in accordance with the requirements of ANSI B31.7. Cleanliness standards consis tent with nuclear service are maintained during fabrication and erection. There are no di ssimilar metal field welds.

4.3.5 PRESSURIZERThe

pressurizer maintains RCS operating pressure and compensa tes for changes in coolant volume during load changes. Table 4.3-6 gives design parameters for the pressurizer. The pressurizer is shown in Figure 4.3-7.

Pressure is maintained by controlling the temp erature of the saturate d liquid volume in the pressurizer. At full load nominal conditions, slightly more than one-half the pressurizer volume is occupied by saturated water, a nd the remainder by saturated stea

m. A number of the pressurizer heaters are operated continuously to offset the heat losses and the continuous minimum spray, MPS2 UFSAR4.3-14Rev. 35thereby maintaining the steam and water in thermal equilibrium at the saturation temperature corresponding to the desired system pressure.During load changes, the pressurizer limits pressure va riations caused by expa nsion or contraction of the reactor coolant. The average reactor coolant temperature is programmed to vary as a function of load as shown in Figure 4.3-8. A reducti on in load is followed by a decrease in the average reactor coolant temperature to the pr ogrammed value for the lower power level. The resulting contraction of the coolant lowers the pressurizer water level causing the reactor system pressure to decrease. The pressure reduction is partially compensated by flashing of pressurizer water into steam. All pressurizer heaters are automatically ener gized on low system pressure, generating steam and further limiti ng pressure decrease. Should the water level in the pressurizer drop sufficiently below its setpoi nt, the letdown control valves cl ose to a minimum value and the available charging pumps in the chemical and volume control system (CVCS) are automatically started to add coolant to the system and restore pressurizer level.When steam demand is increased, the average reactor coolant temperature is automatically raised in accordance with the coolant temperature program (Figure 4.3-8). The expanding coolant from the reactor coolant piping hot leg enters the bottom of the pressurizer (in-surge), compressing the steam and raising system pressure. The increase in pressure is moderated by the condensation of steam during compression a nd by the decrease in bulk temperatur e in the liquid phase. Should the pressure increase be large enough, the pressuri zer spray valves open, sp raying coolant from the RCP discharge (cold leg) into the pressurizer steam space. The relatively cold spray water condenses some of the steam in the steam space, limiting the system pressure increase. The programmed pressurizer water le vel is a power dependent func tion. A high level error signal produced by an in-surge causes the letdown control valves to open, releasing coolant to the CVCS and restoring the pressurizer to the prescribed level.

Small pressure and coolant volume variations are accommodated by th e steam volume which absorbs flow into the pressurizer and by the water volume which allo ws flow out of the pressurizer. The total volume of the pressurizer is determined by consider ation of the following factors:a.Sufficient water volume is necessary to prevent draining the pressurizer as the result of a reactor trip or a loss-of-load accident. In order to preclude the initiation of safety injection and of automatic inje ction of concentrated boric acid by the charging pumps, the pressurizer is desi gned so that the minimum pressure observed during such transients is above the setpoint of th e safety injection actuation signal (SIAS);b.The heaters should not be uncovered by the out-sur ge following load decreases; ten percent step decrease, and five percent per minute ramp decrease; c.The steam volume should be sufficient to yield acceptable pressure response to normal system volume changes dur ing load change transients; MPS2 UFSAR4.3-15Rev. 35d.The water volume should be minimized to reduce the ener gy release and resultant containment pressure during a LOCA;e.The steam volume should be sufficient to accept the reactor coolant in-surge resulting from loss-of-load without the water level reaching the safety and power-operated relief valve (PORV) nozzles;f.During load following transients, the to tal coolant volume change and associated char ging and letdown flows should be kept as small as practical and be compatible with the capacities of the volume control tank, charging pumps, and letdown control valves in the CVCS.To account for these factors and to provide adequate mar g in at all power levels , the water level in the pressurizer is programmed as a function of average cool ant temperature as shown in Figure 4.3-9. High or low water le vel error signals result in the control actions shown in Figure 4.3-10 and described above.

The pressurizer heaters are single uni t, direct immersion heaters whic h protrude vertically into the pressurizer through sleeves welded in the lower hea

d. Each heater is intern ally restrained from high amplitude vibrations and can be individually removed for maintenance during plant shutdown. Approximately 20 percent of the heater s are connected to proportional controllers which adjust the heat input as required to account for steady stat e losses and to maintain the desired steam pressure in the pressurizer.

These heaters are separated into two banks (approximately 160 kW each) and are provided with diverse vital power.

The remaining heaters are connected to on-off controllers. These heaters, called backup heaters, are normally deenergized but are turned on by a low pressurizer pressure signal or high level error signal. This latter featur e is provided since load increases result in an in surge of relatively cold coolant into the pressurizer, decreasing the temp erature of the water volume. The action of the CVCS restoring the level results in a pressure undershoot below the desired operating pressure. To minimize the pressure undershoot, the backup heaters are energized earlier in the transient, contributing more heat to the water before the low pressure setti ng is reached. An interlock will prevent operation of the backup heaters if the high level error signal occurs concurrent with a high pressurizer pressure signal. A low-low pressurizer level signal deenergizes all heaters to prevent heater burnout.

The pressurizer spray is supplied from each of the RCP discharges on one loop to the pressurizer spray nozzle. Automatic spray control valves control the amount of spray as a function of pressurizer pressure; both of the spray control valves functi on in response to th e signal from the controller. These components are sized to use the differentia l pressure between the pump discharge and the pressurizer to pass the amount of spray required to prevent the pressurizer steam pressure from opening the PORVs during normal load following transients. A small continuous flow is maintained through the sp ray lines at all times to keep the spray lines and the surge line warm, reducing thermal shock during plant transients. This continuous flow also aids in keeping the chemistry and boric acid concen tration of the pressurizer water e qual to that of the coolant in MPS2 UFSAR4.3-16Rev. 35 the heat transfer loops. An aux iliary spray line is provided from the charging pumps to permit pressurizer spray during plant heatup, or to allow cooling if the RCPs are shut down.

In the event of an abnormal transient which causes a sustained increase in pressurizer pressure, at a rate exceeding the control capacity of the spray, a high pressure trip level will be reached. This signal trips the reactor and opens the two PORVs. The steam discharged by the relief valves is piped to the quench tank where it is condensed. In accordance with Section III of the ASME Boiler and Pressure Vessel Code, the RCS is pr otected from overpressure by two spring-loaded safety valves. The discharge from the safety valves is also piped to th e quench tank. See Section 4.3.7 for the safety valve design parameters.

The pressurizer is supported by a cy lindrical skirt welded to the lo wer head. Since the pressurizer surge line has sufficient flexibility, no provision s are made for horizontal movement and the skirt is bolted rigidly to the floor.

The pressurizer assembly was replaced in 2006 with a new pressurizer asse mbly fabricated from materials that are less suscepti ble to primary water stress co rrosion cracking. The replacement pressurizer is fabricated and installed to the same design criteria as the original pressurizer with some improvements. The cylindrical shell sections, upper and lower heads including the large bore nozzles of the replacement pressurizer are forged components, thereby minimizing the welds and weld inspections. Safe ends made of stainless steel are provided as required on the large bore nozzles to facilitate field welds to the connecting piping. The interi or surface of the pressurizer is clad with weld deposited stainless steel. The heater sleeves, inst rument nozzles and the vent/pass nozzle are fabricated from stainless steel instead of the originally used Inconel Alloy 600 material. The nozzles are attach ed to the pressurizer by j-groove welds to the clad buildup. The total number of heaters is reduced from 120 to 60. The heat output of each heater is roughly twice that of the replaced heater, thus maintain ing the same amount of total heat output.

A six and one-half inch inside di ameter vent port was added on th e upper head of the replacement pressurizer. This vent port is a substitute for the removal of the pressurizer manway during routine

refueling outages with RCS noz zle dams installed. The vent port is sized to avoid a RCS pressurization that would excee d the design pressure of th e RCS nozzle dams following a postulated loss of shutdown cooling with the reactor vessel he ad on and reactor coolant water levels one foot below the reactor vessel flange or lower. Fo r non-routine outages where nozzle dams are installed and r eactor coolant water levels are higher than one foot below the reactor vessel flange and the reactor vess el head is on, the pres surizer manway will need to be removed in order to avoid a RCS pressurization that would exceed the nozzle dam design pressure following a postulated loss of shutdown cooling. In addition, removal of either the vent port or the pressurizer manway provides an adequate RCS vent path for low temperature overpressure protection in Mode 5 and Mode 6 when the head is on the reactor vessel.

4.3.6 QUENCH

TANKThe quench tank is designed to prevent the discharg e of the pressurizer re lief or safety valves from being discharged to the containment. The quench tank is shown in Figure 4.3-11. The steam discharged into the quench tank from the pressurizer is discharged under water by a sparger to MPS2 UFSAR4.3-17Rev. 35 enhance condensation. The normal quench tank water volume of 135 cubic feet is sufficient to condense the steam released from the pressurizer safety and relief valves. The quench tank is sized to accommodate the steam released as a result of a loss-of-load accident followed immediately by an uncontrolled rod withdrawal accident with no coolant letdown or pressurizer spray.The water temperature rise in the quench tank is limited to 281

°F, assuming a maximum initial water temperature of 120

°F. The gas volume in the tank is sufficient to limit the maximum tank pressure after the above steam release to 35 psia. The contents of the quench tank are cooled by recirculation through the primary drain tank and quench tank cooler. The temperatur e of the water in the quench tank is indicated on the main control console. A high temperature alarm is also provided. A high quench tank temperature alerts the operator to cool the tank contents.

A measurement channel provides a quench tank pressure indicati on on the main control console and actuates a high pressure alar

m. High quench tank pressure indicates that the tank has received a discharge from the safety or reli ef valves, or from the HPSI test line relief valve. The operator would then take action to restore th e tank to normal operating conditions.

Quench tank level indication and high and low level alarms are al so provided on the main control console.The quench tank can condense the steam discharged during a loss-of-lo ad accident as described in Section 14.2 without exceeding the rupture disc set point, which is ra ted for 96 psig at 72

°F and 89 psig at 350

°F, assuming normal closing of the safety valves at the end of the accide nt. It is not designed to accept a continuous unc ontrolled safety valve discharge. The rupture disc on the quench tank provides code overpressure protection of the tank. The rupture disc vents to the containment. The quench tank parameters are given in Table 4.3-7.

The tank normally contai ns demineralized water under a nitrogen overpressure. The sparger, spray header, nozzles and rupture disc fittings are stainless steel. The tank is designed and fabricated in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Class C.

The quench tank is located at a level lower than the pressurizer. This ensures that any PORV or pressurizer safety valve leakage from the pressurizer, or any discharge fro m these valves, drains to the quench tank.

4.3.7 VALVES

The design parameters for the pressurizer sp ray valves (RC-100E, RC-100F) are given in Table 4.3-8. The PORV isolation valve (RC-403, RC-405) parameters are given in Table 4.3-9.

The position of each valve on loss of actuating signal (failure position) is selected to ensure safe operation. System redundancy is considered when specifying the failure position of any given valve. Valve position indication is pr ovided at the main control panel.

MPS2 UFSAR4.3-18Rev. 35 Manually operated valves in the RCS have backseats to limit stem leakage when in the open position. Globe valves are installe d with flow entering the valve under the seat. This arrangement will reduce stem leakage during normal operation or when closed.The two PORVs, designated RC-402 and RC-404 on Figure 4.1-1, relieve suffic ient pressure to avoid opening of the RCS safety valves. The relief valves are actu ated by the high RCS pressure trip signal. Parameters for these valves are given in Table 4.3-10.

The valves are solenoid-operated pow er relief valves. The two half capacity valves are located in parallel pipes which are connected to the pressurizer relief valve nozzle in the inlet side and to the relief line piping to the quench tank on the outlet side. A motor-actuated isolation valve is provided upstream of each of the relief valves to permit isolating the valve for maintenance or in case of valve leakage.The capacity of the PORVs is sufficient to pass the maximum steam surge associated with a continuous control rod withdrawal accident starting from low power. Assuming that a reactor trip is effected on a high pressure signal, the capacity of the PORVs is sufficient so that the safety valves do not open. The relief valve capacity is also large enough so that the safety valves do not open during a loss-of-load accident from full power. This assumes normal operation of the pressurizer spray system, and r eactor trip on high pressure. The PORVs also function to provide low temperature overpressurization protection (LTOP) to the RCS. This is accomplished by manually selecting a redu ced valve setpoint as described in Section 7.4.8.Two safety valves, designated RC-200 and RC-201 on Figure 4.1-1 are located on the pressurizer to provide overpressure protection for the RCS. They are totally en closed, backpressure compensated, spring-loaded safety valves meeting ASME Code requirements. Parameters for these valves are given in Table 4.3-11.

Both safety valves (RC-200, 201) are equipped with acoustic valve position monitors. These monitors will provide operators with an indication of the status of the safety valves. The PORVs have been upgraded to monitor their position fr om the control room by the position indication lights instead of the acoustic m onitoring method previously use

d. The valve monitoring system conforms with NUREG-0578 C lessons learned from TMI Task Force Report.The safety valves pass sufficient pressurizer steam to limit the RCS pressure to 110 percent of design (2,735 psig), following a co mplete loss of turbine genera tor load without simultaneous reactor trip. Reactor trip occurs on a high RCS pressure signal. To determine the maximum steam flow, the only other pressure relieving system assumed operational is th e steam system safety valves. Conservative values for all system parameters, delay times, and core moderator coefficient are assumed. This analysis is given in Section 14.2, Lo ss-of-Load External Electrical Load and/or Turbine Trip.

The mounting of pressure relievi ng devices (safety valves and re lief valves) with in the RCPB and on the main steam lines out side of the containment is in accordance with the applicable provisions of ASME Boiler and Pressure Vessel Code Section III.

MPS2 UFSAR4.3-19Rev. 35The pressurizer safety and relief valves are connected to nozzles on the top of the pressurizer vessel. The loads which the nozzles experience during normal plant operation and when the valves are relieving are included in the specification for the pressurizer.

All overpressure relief valves a nd their connected piping (i.e., headers, header connections and discharge piping) are designed to withstand the following conditions without exceeding the applicable codes primary stress allowable: maximum loads due to valve discharge thrust, internal pressure, dead weight and earthquake applied simultaneously. When more than one relief valve is attached to a piping system, the loads due to all relief valves discharg ing simultaneously are applied to the system al ong with the above menti oned primary loads. In addition, the loads from the most critical combination of valves discharging are applied.

The local stresses in the main steam line outside the containmen t at the connection of the relief valves were computed as specified in "Welding Research Council Bulletin 107" and contai ned below the allowable primary stress level.The pressurizer safety and relief valve discharg e piping system was modi fied by deleting the loop seals upstream of the Pressurizer Safety Valves (PSVs) and Power-Operated Relief Valves (PORVs). The piping for both the PSVs and PORV s was again modified during the replacement of the pressurizer. The PORVs were also replaced with upgraded PORV s and credited as an acceptable pressurizer steam sp ace vent path during post accident conditions. The new piping configuration was evaluated to address NUREG-0737, Acti on Item II.D.1 (relief valve and safety valve testing).

The thermal-hydraulic analysis was performed to ca lculate the transient flui d time-history forcing functions acting on the pipe segmen ts due to each safety or relief valve discharge. These forcing functions are combined in a common entry mode to maximize the forces acting on the discharge piping. The potential case of water discharge through relief valves during low temperature modes of operation was also considered.

The structural reanalysis of the PSV and PORV discharge piping system included the normal plant loading, plant transients described in Section 4.2.1, seismic loading, and the fluid transient forcing functions. Seismic analys is of the piping system was performed by the modal analysis response spectra method. D ynamic response of the piping system to the PSV and PORV discharge loads was performed by the time-history modal superposition method.

Pipe primary and secondary stress intensities and fatigue usage fa ctors were found to be within the Code allowable values. The pipe supports of the PSV and PORV discharge piping system were modified to accommodate the load and displacements resulting from the reanalysis.The main steam relief valve system is designed so that the blowdown force is transmitted directly to the structure by mechanical/structural devices, and not through the piping. These relief valves are provided with discharging stacks to dir ect steam blowdown to atmosphere. Stacks are designed so that backpressure does not result in a va lve reaction force.

Pumps and valves within the RCPB are classified as either active or inactive components. Active components are those whose operabili ty is relied upon to perform a safety function, as well as MPS2 UFSAR4.3-20Rev. 35 reactor shutdown function, during the transients or events considered in the respective operating condition categories. Inactive components are t hose whose operability (e.g., valve opening or closure, pump operation or trip) are not relied upon to perform the system function during the transients or events considered in the respec tive operating condition cat egories. Thus, certain pumps and valves (classified as active components) within th e RCPB are required not only to serve as pressure-retaining compone nts (as in the case of passive components such as vessel and piping) but also to operate reliably to perform a safety function such as safe shutdown of the reactor and mitigation of the consequences of a pipe break accident under the loading combinations considered in RCPB as either act ive or inactive. The only pumps within the RCPB are the RCPs which are inac tive under faulted conditions.

Active components outside of the RCPB have been designed to function as required when subjected to the loadings and the maximum pressure and temper ature occurring under normal and accident conditions in th e areas in which the components are located. Anticipated temperature and pressure transients which would have an effect on operability of components were specified as design requirements.A complete list of the active pumps and valves located within the RCPB was provided previously as part of Amendment 14, and is found as Table 4.3-12.The following is a list of activ e pumps and valves located outside the RCPB whose operability is relied upon to perform a safety function such as sa fe shutdown of the reac tor or mitigation of the consequences of a postulated pipe break in the RCPB.

Pumps Component QuantityMaterials Code Test Code Normal Position Post LOCA Position High Pressure Safety Injection3Draft ASME Code for Pumps and Valves for Nuclear Power, Class II, Nov. 1968ASI Standard 610, ASME Power Test Code PTC- 8.2 and Standards of the Hydraulic InstituteOffOn Low Pressure Safety Injection2Draft ASME Code for Pumps and Valves for Nuclear Power, Class II, Nov. 1968ASI Standard 610 Standard of the Hydraulic Institute, and ASME Power Test Code PTC-8.2OffOn MPS2 UFSAR4.3-21Rev. 35 Containment Spray Pumps2Draft ASME Code for Pumps and Valves for Nuclear Power, Class II, 1968Standard of the

Hydraulic InstituteOffOnValves Component QuantityMaterials Code Test Code Normal Position Post LOCA Position High Pressure Safety Injection Pump Suction (Check)2Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968ClosedOpen High Pressure Safety Injection Pump Discharge (Check)4Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968ClosedOpen Low Pressure Safety Injection Pump Discharge (Check)2Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968ClosedOpen Containment Isolation Valves-Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968-As Required As Required Containment Spray (Motor Operated)2Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968ClosedOpen Containment Sump Discharge (Check)2Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968ClosedOpen Pumps Component QuantityMaterials Code Test Code Normal Position Post LOCA Position MPS2 UFSAR4.3-22Rev. 35 For inactive valves and pumps other than the RCPs, the rules of the Pump and Valve Code (March 1970 Draft) for design conditions are applied in evaluating the loadings produced by the Emergency and Faulted Conditions.For active components, additiona l requirements are imposed on th e design to assure operability during the faulted operating conditions. As appropriate, these add itional requirements consist of simulated tests and/or supplemen tary calculations which demonstr ate that the active component will perform its required function during the specified conditions. Where calculations are employed, the primary stre sses produced by the faulted conditions are limited to values less than the Emergency Condition limits of Subparagra ph HB-3224.1 in all regions of the active component where deformations may impair the required function.

For the RCPs, as appropriate, the stress criteria for Vessels, as discussed above, are applied in evaluating the Emergency and Faulte d Conditions (elastic analysis). These limi ts are consistent with the "Design by Analysis" me thod of Article 4, ASME Section III Code which is applied in the design calculation. For design c onditions other than those explic itly addressed by Section III of the ASME Code, and where desi gn calculations are used to eval uate stresses and deformations in pumps, the methods and criter ia applied are in accordance wi th Article 4, ASME Section III Code. The calculations will include the effects of gross and local st ructural discontinuities and the loadings produced by geometrical eccentricities. Current state-of-the-art analytical methods including finite element technique s, are employed in the calculat ions. Experimental techniques were not employed.Refueling Water Storage Tank Discharge (Check)2Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968ClosedOpen during Injection Mode.

Closed during

Recirculation Mode.Shutdown Cooling Heat Exchanger Bypass (2-SI-

306) (Air Operated)1Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968OpenOpen Shutdown Cooling Heat Exchanger Flow Control (2-SI-657)

(Air Operated)1Draft ASME Code for Pumps and Valves for Nuclear Power, Nov. 1968ClosedClosedValves Component QuantityMaterials Code Test Code Normal Position Post LOCA Position MPS2 UFSAR4.3-23Rev. 35

4.3.8 INSTRUMENTATION

APPLICATIONThe measurement channels necessary for opera tional control and protection of the RCS are described below. A brief explanat ion of the purpose of each meas urement channel is made and a summary of resulting action is given. A detailed desc ription of critical inst rument channel actions may be found in Chapter 7.Four independent measurement channels are provided for each parameter which initiates protective system action. Two i ndependent signals are required to initiate protective action, thereby, preventing spurious acti ons resulting from the failure of one measurement channel. This arrangement results in a high degree of protective measurement channel reliability in terms of initiating action when required and avoiding unnecessary action from spurious signals. Two independent measurement channels are provided for parameters which are critical to operational control. These control channels are separate fr om the protective measurement channels. To avoid control conflicts, control action is derived from only one channe l at any time, while the second channel serves as a backup. This allows continued operation of the facility if one channel fails and permits maintenance on the failed channel during operation. This arra ngement results in increased availability.

4.3.8.1 Temperature 4.3.8.1.1 Hot Leg Temperature Each of the two hot legs contai ns five narrow range channels to measure coolant temperature leaving the reactor vessel. Four of these channels are used to furnish a hot leg temperature signal to the reactor protective system. The fifth hot leg temperature measur ement channel provides a signal to the average temperature computer which is a part of the reactor regulating system (RRS).

The average temperature of the loop is recorded on a two-pen re corder in the ma in control room. The second pen on each loop's reco rder records an average temper ature reference signal received from the RRS.

A high temperature alarm is provided on this channel to alert th e operator to a high temperature condition. The temperature from this measurement ch annel is indicated in the main control room in addition to being recorded. The other hot leg temperature channels are also displayed in the main control room.

4.3.8.1.2 Cold Leg Temperature Each of the four cold legs contains three temperature meas urement channels. The cold leg Resistance Temperature Detectors (RTD) are located downstream of the RCPs. Two channels from each cold leg (four per heat transfer loop) are used to furnish a cold leg coolant temperature signal to the reactor protecti ve system (RPS). All eight of these cold leg temperature measurements are indicated in the main control room. Two of the remaining four cold leg temperature measurement channels, one each fr om opposite loops, are narr ow range temperature measurements that input to th e RRS for calculation of average temperature and the Feedwater regulating system for dynamic comp ensation in single element contro

l. These loops also provide MPS2 UFSAR4.3-24Rev. 35 control room indication of narrow range temper ature. The remaining two cold leg measurement channels, one each from opposite loops, are wide range temper ature measurements that provide input to the subcooled margin monitor (Inadequate Core Cooling ICC) and Low Temperature Over Pressure (LTOP) channels. These loops al so provide control indi cation and recording of wide range temperature.

4.3.8.1.3 Surge Line TemperatureThis measurement channel provides an indication of sur ge line temperature in the main control room. A low surge line temperatur e condition activates an alarm in the main control room. The low-temperature alarm during norma l operation is an indication that the continuous spray rate has decreased.

4.3.8.1.4 Pressurizer Vapor Phase TemperatureThe pressurizer vapor phase RTD is located on the upper dome of the pressurizer. This channel provides a wide range te mperature indication of the temper ature of the steam phase in the pressurizer.

4.3.8.1.5 Pressurizer Water Phase TemperatureThe pressurizer water phase RTD is located at an elevation below the top of the pressurizer heaters. This channel provides a wide range temperature indication in the main control room and is used during plant heatup and cooldown.

4.3.8.1.6 Spray Line TemperatureAn RTD in each spray line provides a temperatur e indication and a low te mperature alarm in the control room for each spray line. A low temp erature alarm during normal operation is an indication the continuous spray rate has decreased.

4.3.8.1.7 Relief and Safety Valve Discharge TemperatureTemperatures in the pressurizer safety valv e and PORV discharge lines are measured and indicated in the main control room. A high temperat ure in one of these lines is an indication that the associated valve may be leaking. High temperature alarms are pr ovided to alert the operator to this condition.

4.3.8.1.8 Quench Tank Temperatures The temperature of the water in the quench tank is indicated in the main control room. A high temperature alarm is also provi ded. A high quench tank temperatur e alerts the operator to the requirement for cooling of the tank contents.

MPS2 UFSAR4.3-25Rev. 35 4.3.8.1.9 Reactor Vessel Flange S eal Leakage TemperatureThis RTD is located in the reactor vessel flange leak-off line. The channel is displayed in the main control room and actuates a high temperature alarm. A high temperat ure is indicative of excessive leakage past the first reactor vessel flange seal.

4.3.8.1.10 RCS High Point Vents Leakage Temperature Thermocouple installed on the downs tream side of solenoid valve tr ain of the reactor vessel head vent system is utilized to monitor leakage past the system solenoid valves. Under normal operating conditions, the thermocouple will meas ure the ambient temper ature in the piping downstream of the solenoid valve train. The output of each thermocouple is continuously recorded by the plant computer. Note that monitoring of the leakage past the PORVs in the pressurizer steam space vent path is described in Section 4.3.8.1.7, above.

Any leakage through the system va lves will cause an increased temperature in the downstream piping which will be detected by the thermocouple.

At a predetermined set point, an alarm will be actuated, identifying a high temp erature reading on the appropria te thermocouple. Once a high temperature alarm is received, furt her actions will be governed by the Technical Spec ifications for reactor coolant system leakage.

4.3.8.2 Pressure 4.3.8.2.1 Pressurizer Pressure Four independent narrow range pr essure channels are provided for initiation of protective systems action. The pressure transmitters are connected to the upper portion of the pressurizer via the upper level measurement no zzles and measure pressurizer vapor pressure. All four channels are indicated in the main control room and actuate separate high, low , or low low pressure alarms in the control room.

The protection actions these pressure signals initiate are:1.Reactor trip on high pr imary system pressure. The reacto r trip signals are also used to open the PORVs;2.Safety injection system actuation on low low primary system pressure;3.Reactor trip on a low primary system pressure. The set point is a function of the coolant temperatures in the h ot and cold legs. The variable se t point has high and low limits alarmed in the control room and is not allowed to decrease below 1865 psia.

MPS2 UFSAR4.3-26Rev. 35 4.3.8.2.2 Pressurizer PressureTwo independent pressure channels provide narr ow range pressure signa ls for controlling the pressurizer heaters and spray valv es. The output of one of these ch annels is manually selected to perform the control function. During norma l operation, a small group of heaters are proportionally controlled to of fset heat losses. If the pressure falls below a low pressure set point, all of the heaters are energized. If the pressure increased above the high pressure set point, the spray valves are proportionally ope ned to increase the spray flow rate as pressure rises. An interlock will prevent operation of the backup heaters in the ev ent of a high level error signal concurrent with a high pressure condition. These two channe ls are also used to provide pressurizer pressure signals to the RRS. The two channels are cont inuously recorded in the main control room and are provided with high and low pressure alarms.

4.3.8.2.3 Pressurizer PressureTwo low-range pressure measurement channels provide a control r oom indication of RCS pressure during plant startup and shutdown in the main cont rol room. They also provide independent pressure signals to the shutdown cooling suction isolation valves (refer to Section 9.3.4.1) which prevent thes e valves from opening above a selected set point. If the shutdown cooling suction valves are open when th e pressure exceeds a selected set point (280 psia), an annunciator receives a signal to alarm from these pressure channels. The channels also provide signals to actuate the PORVs for LTOP.

These two instrument channels are independent and redundant.

4.3.8.2.4 Quench Tank Pressure This measurement channel provides a quench tank pr essure indication in the main control room and actuates a high pressure alar

m. High quench tank pressure indicates that the tank has received a discharge from the safety or reli ef valves, or from the HPSI test line relief valve. The operator will then take action to restore th e tank to normal operating conditions.

4.3.8.3 Level 4.3.8.3.1 Pressurizer LevelTwo pressurizer level channels are used to provide two independent level signals for control of the pressurizer liquid level. These si gnals are used to deener gize th e pressurizer heaters on low low pressurizer level to prevent heat er burn out, provide input to one pen in the two-pen recorder in the control room, and actuate high and low pressurizer level alarms in the main control room. The second pen on the level recorder records the programmed pressurizer level setpoint computed by the RRS as a function of the average reactor c oolant temperature. The level transmitters are compensated for the steam and water densitie s existing in the pressurizer during normal operation.

The liquid level in the pressurize r is programmed to vary as a f unction of average reactor coolant temperature. This level set point is computed by the RRS and furnished to controls associated MPS2 UFSAR4.3-27Rev. 35with these measurement channels. One of the two measurement channels is manually selected to furnish an actual liquid level signal to the contro ls. If these two signals differ, the level control adjusts the CVCS charging or letdown flow rates to make the difference zero. Each of the level channels is indicated in the main control room and is equipped wi th a high and low level alarm in the control room.

4.3.8.3.2 Pressurizer LevelOne wide range pressurizer channe l is provided for main control room indication of pressurizer level during plant startup and shutdown.

4.3.8.3.3 Quench Tank Level A quench tank level channe l indicates quench tank level in the main control room. The transmitter also activates high and low level alarms in the main control room.

4.3.8.4 Reactor Coolant Loop FlowFor independent differential pre ssure measurement channels are provided in each heat transfer loop to measure the pressure drop across the steam generators. Four pressure taps are located in each hot leg piping sectio n just before the elbow entering the steam generator and four pressure nozzles are located in the steam generator outlet plenum. Four differential pressure transmitters are connected between the four hot leg nozzles and the steam generator nozzles, resulting in four steam generator differential pressures.

The outputs of the transmitters ar e sent to four analog summing devices in the low total flow trip logic. Each summing device receives two differential pre ssure signals with the summation of these signals representing the total core flow at all times, even during coast down transients.

The summing devices provide four independent total flow signals.

The four signals are indicated separately in the main control r oom and activate separate low-flow alarms. In the RPS, they are compared with the low-flow reactor trip set point, selected by the operator to correspond with the number of operating RCPs. If two channels indicate flow which is less than the flow set point, the reactor is tripped.

4.3.8.5 Reactor Coolant Pump InstrumentationThe RCPs and motors are equipped with the in strumentation necessary for proper operation and to warn of incipient failures. (See Figure 4.3-4.) A description of the major channels follows:

4.3.8.5.1 Pump Seal Temperatures The reactor coolant temperature in the lower s eal cavity may be indicated in the main control room through its selection on a multiposition switch. Th e switch also permits display of all of the remaining temperature measurem ent channels for each pump. Th e pump seal temperature is alarmed to alert the operators to a high-temper ature condition. A high temp erature condition is an MPS2 UFSAR4.3-28Rev. 35 indication that the integral h eat exchanger is not performing satisfactorily or is a backup indication that the com ponent cooling water fl ow has decreased or supply temperature has increased.

4.3.8.5.2 Motor Stator TemperaturesEach RCP motor is provided with six RTD's em bedded in the stator wi ndings. During initial pump testing the highest reading RTD was selected for this temperature measurement control.

The signal output of this RTD may be selected for indicat ion in the main control room by a multiposition switch. Should stator temperature exceed a predetermi ned limit, a high temperature alarm will be sounded in the control room. High temper ature is detrimental to motor winding insulation life, and may be caused by high ambient temperature, reduction in the cooling air flow to the stator or inadequate time delay between successive starts of the motor.

4.3.8.5.3 Motor Thrust Bearing TemperaturesTemperatures of the motor upward thrust beari ng, and downward thrust bearing may be indicated in the main control room through selection on a multiposition swit ch. A high temperature alarm is provided for each pump in the main control room which is annunciated if any one of the bearing temperatures in the pump exceeds a safe value.

4.3.8.5.4 Pump Controlled Bleed-Off Temperature The temperature of the controlled bleed-off flow may be displa yed in the main control room through its selection on a multiposit ion switch. An alarm signal is provided should the controlled bleed off temperature exceed a high limit. A high temper ature condition is an indication that the integral heat exchanger is not operating properly.

4.3.8.5.5 Antireverse Device Bearing TemperatureThis measurement channel provide s a status on the operating te mperature of the antireverse device bearing. Display in the ma in control room is available through selection on a multiposition switch. A high temperature alarm is provided to alert plant operators to an abnormal condition.

4.3.8.5.6 Upper and Lower Guide Bearing Temperature The upper and lower guide be aring temperatures may be displayed in the main control room at the option of the operators. An alarm signal is provided if the high temperature limit is exceeded.

4.3.8.5.7 Lube Oil Cooler Inlet and Outlet TemperatureThe inlet and outlet temperatures of the lube oi l cooler are available for disp lay in the main control room and are alarmed to alert plant operators to high temperature conditions.

MPS2 UFSAR4.3-29Rev. 35 4.3.8.5.8 Lower Bearing Oil Temperature The lower bearing oil temperature may be displayed in the main control room at the option of the plant operators.4.3.8.5.9 Pump Seal PressuresThe middle, upper, and controlled bleed-off pump seal cavities in each pump are provided with pressure sensors which generate a signal proportional to the pre ssure within the cavity. The pressure in any seal may be selected for displa y in the main control room through its selection on a three-position switch.

A high and low pressure alarm is annunciated for the middle and upper seal measurement channels. Abnormally, high pressure in th e upper and middle seal cavities indicates a failed or failing lo wer or middle seal. A low pressure condition in the middle seal cavity indicates a failed or failing upper seal.

A recorder in the cont rol room records seal pressures in order to rec ognize pump seal degradation.

4.3.8.5.10 Motor Oil Lift Pressure This pressure measurement channe l provides a signal to the RCP c ontrol circuit to prevent motor start with insufficient motor thrust bearing oil pressure.4.3.8.5.11 Lube Oil Filter Pressure DifferentialThere are three lube oil filters and one strainer , each with differential pr essure indication capability. These measurement channels provide alarms for high differential pressure across any of the filters or the strainer, indicating clogging.

4.3.8.5.12 Pump Controlled Bleed-Off Flow Flow instruments are used to measure the controlled bleed of f flow from th e pump upper seal cavity to the chemical and vo lume control system. These instrume nts provide a remote indication of the flow rate in the East and West electrical pe netration rooms, and annunciate high and low flow alarms in the control room. Th ere is also an input to the plan t computer and a recorder in the control room which records flow.

4.3.8.5.13 Lube Oil and Antireverse De vice Lube Oil Flow Switch These measurement channels provide alarm signals when a low flow condition exists in the thrust bearing lube oil loop or in the antireverse device lube oil flow

.4.3.8.5.14 Motor Oil Reservoir LevelFloat-type sensors or differential pressure transmitters are used to produce signals proportional to the oil levels in the upper and lower motor oil reservoirs. These signals are used to provide an indication on the main control panel of the oil level in each oil reservoir. Either oil reservoir level MPS2 UFSAR4.3-30Rev. 35 may be selected for display in the control ro om through use of a two-position switch. Each signal also annunciates a high or low le vel alarm in the control room.

4.3.8.5.15 Vibration Instrumentation Motor vibration is sensed by two velomitor probes attached to the upper motor frame. Excessive motor vibration will cause an al arm on the plant process computer.

Pump shaft orbit is monitored by two proximity probes mounted 90 degrees to each other in a horizontal plane. A separate key phasor probe provides a single pulse per revolut ion as a reference for determining angular changes in shaft orbit. Outputs from the probes are also recorded.

4.3.8.5.16 Reverse Rotation Switch Reverse rotation of an RCP is sensed by a reverse rotation switch. This switch actuates an alarm in the control room. Reverse rotation indicates fa ilure of the mechanical anti-reverse device.

4.3.8.5.17 (Deleted)4.3.8.5.18 RCP Underspeed Reactor Trip An RCP underspeed trip has been added which uses a speed sensor on the RCP motor. The sensor generates a frequency signa l proportional to speed which is in turn converted to a voltage. This voltage signal is maintained by the RPS, w ith one pump signal bei ng monitored by only one channel in the RPS. When any two pump speeds drop below 830 RPM a reactor trip will occur.

This trip is designed to protect against simultan eous loss of all pumps and the resulting loss of coolant flow, with a faster reaction or trip time, than is provided by the steam generator differential pressure signal.

4.3.9 REACTOR

COOLANT VENTING SYSTEMThe RCS venting system provides the capability for removing noncondensible gases collected in the system in order to allow for satisfactory long term core cooling.

The two important safety functi ons enhanced by this venting cap ability are core cooling and containment integrity. For events within the pres ent design basis for nuc lear power plants, the capability to vent noncondensible gases will provide additional assu rance that the requirements of 10 CFR 50.44 will be met. For events beyond the de sign basis, this ve nting capability will substantially increase the ability to deal with large qua ntities of noncondensible gas.Reactor Vessel Head Vent SystemThe reactor coolant head vent system is equipped with two (2) solenoid-operated globe valves in series in each piping train. Each valve has remote-manual control capability from the control room with open and closed position indicati on. Provision of two (2) solenoid operated globe valves in series for each vent train minimizes the probability of a vent path failing to close, once MPS2 UFSAR4.3-31Rev. 35 opened. The power source for each valve train is an independent redundant DC emergency bus, energized from separate redundant battery systems. In addition, the valves also receive power from redundant independent AC emergency buses. All valves fa il closed upon lo ss of power supply to the actuator. The solenoid ve nt valves are qualified to IEEE-344-1975.

A manual isolation valve is inst alled downstream of the solenoid ope rated vent valves. This valve is normally open and is inte nded to isolate leakage that may develop through the solenoid-operated valves. Also, a manual isolation valve is located just upstr eam of the two solenoid valves on each train of the RCS head vent piping, capable of isolating the affected train. Piping upstream of the valve is designed to reactor coolant system pressure rating, providing an at-power isolation capability.The discharge sparger for the reactor coolant head vent system is located in the vicinity of the containment air coolers where the vented gases wi ll be cooled, mixed with additional containment air, and discharged into the lower elevati on of the containment.

Uniform mixing of the containment post accident atmosphere is provided by the post-accid ent recirculation system. The RCS vent system piping has been analyzed in accordance with ASME,Section III, of the Boiler and Pressure Vessel Code, for the Class 1 portion of the system. The balance of the system has been analyzed pursuant to ANSI B31.1 Power Piping Code.Pressurizer Steam Space Vent System Venting of the noncondensible gases can also be performed from the pressurizer. The pressurizer venting system consists of two flow paths with redundant isolati on valves in each flow path. Each flow path has a power operated re lief valve and a block valve. The block valve in each train is normally open. During the pressurizer replacement project the two (2) power operated relief valves were upgraded so they can be credited to perform the venting function post accident. The PORV's are EEQ qualified and rece ive power from battery backed independent vital DC buses.

The noncondensibles are vented via the PORV discharge piping into the quench tank and then to the waste gas header as necessary. If the pressure reaches in excess of the rupture disc relief pressure, the noncondensibles woul d be released into the containment. The PORV inlet and discharge piping has been analyzed in accordance with ASME,Section III, of the Boiler and Pressure Vessel Code, Class 1 and Class 2 and 3 requirements respectively.The reactor vessel and pres surizer vents were designed to utili ze existing penetrations within each vessel. The system can pass in ex cess of the gas volume equivalent to one-half the RCS volume in one (1) hour. Although the RCS vents are larger than the size co rresponding to the definition of LOCA (10 CFR 50, Appendix A), consequences of ruptures of the vents are bounded by the results of current small break loss of coolant accident (SBLOCA) analyses.

Since the vent lines are sized larger than the size corresponding to the definition of LOCA, the system is equipped with two (2) solenoid-operated globe valves in series in each piping train.

Each valve has remote-manual control capability from the control room with open and closed position indication. A manual isolation valve is installed downstream of the solenoid operated vent valves on each of the reactor and pressurizer vent systems. This valve is normally open and is intended to isolate leakage that may develop through the solenoid operated valves. Also, a manual MPS2 UFSAR4.3-32Rev. 35isolation valve is located just upstream of the two solenoid valves on each train of the RCS head vent piping, capable of isolating the affected train.Piping upstream of the valve is designed to reactor coolant system pres sure rating, providing an at-power isolation capability.

The design of the venting system minimizes the proba bility of a vent path failing to close, once opened. This has been accomplishe d by providing two (2) solenoid opera ted globe valves in series for each vent train. The power source for eac h valve train is an independent redundant DC emergency bus, energized from separate redundant battery systems. In ad dition, the valves also receive power from redundant independent AC emergency buses. All valves fail closed upon loss of power supply to the actuator.The discharge sparger for the RCS vent system is located in the vicinity of the containment air coolers where the vented gases will be coole d, mixed with additional containment air, and discharged into the lower elevation of the cont ainment. Uniform mixing of the containment post-accident atmosphere is provided by the post-acciden t recirculation system.The RCS vent system piping has been analyzed in accordance with ASME,Section III, of the Boiler and Pressure Vessel Code, for the Class 1 portion of the system. The balance of the system has been analyzed pursuant to ANSI B31.1 Powe r Piping Code. The sole noid vent valves are qualified to IEEE-344-1975.

4.3.10 PERMANENT REACTOR CAVITY SEAL The permanent reactor cavity seal is designed to c ontain refueling water in the refueling cavity for fuel shuf fle during outages. The reactor cavity seal consists of seal membrane sub-assembly and support structure sub-assembly. The seal membrane sub-assembly consists of a stainless steel membrane, inner and outer legs attached to the reactor vessel seal ledge and th e embedment ring by circumferential fillet welds all around the reactor vessel.

The support structure assembly functions as the load bearing assembly consisting of radial members around the annulus, resting on the reactor vessel flange and the embedment ring. The reactor cavity seal is shown in Figure 4.3-12.The reactor cavity seal has multiple openings, which provide access to the ex-core nuclear instrumentation as well as ventilation for the reactor cavity cooling air. These openings have hinged cover plates, which are closed for flooding the transfer canal during refueling. O-rings are installed to provide watertight seal to prevent leakage into the reactor cavity. Prior to reactor operation, the hinged cover plates will be laid back to allow for vent ilation of cooling air from the reactor cavity.

The cavity seal is designed to with stand the pressure load s from the refueling water as well as the motions of the reactor ve ssel and containment buildi ng due to seismic displacements. It can also accommodate axial and radial gr owth from the normal and transi ent thermal conditions of the reactor vessel. The cavity seal is also designe d and tested to withstand loads imposed by a dropped fuel assembly. The cavity seal is designed for 80 heat up/c ool down cycles and 50 cycles of maximum allowed water static head.

MPS2 UFSAR4.3-33Rev. 35 Leak before break analysis is applied as the basi s for design of the reactor cavity seal and neutron shielding. Compartmental pressuri zation due to reactor coolant lo op rupture is excluded for the design of reactor cavity seal and the neutron shielding in accordance with Reference 3.A-29.

The support structure is fabricated from 300 series stainless steel conforming to ASME Section II Part A. The seal membra ne is allowed to deflec t such that it can rest on the support structure during flooding. The cavity seal membrane, although not required, is designed and analyzed to the guidelines of ASME B&PV Code Section III, Appendix XIII and ASME Section II, Parts A and C, 1995 Edition, through 1996 Addenda.

MPS2 UFSAR4.3-34Rev. 35TABLE 4.3-1 REACTOR VESSEL PARAMETERS Design Pressure, psig 2485 Design Temperature, °F 650 Nozzles Inlet (4 each), inches 30 Outlet (2 each), ID inches 42 CEDM (69), ID, inches 2.718 Instrumentation (8), nominal, inches 4 5/8 Vent (1), nominal, inches 0.75 Dimension Inside Diameter, nominal, inches 172 Overall Height, Including CEDM Nozzles, inches 503.75 Height, Vessel Without Head, inches 408-9/16 Wall Thickness, minimum, inches 8-5/8 Upper Head Thickness, minimum inches 7-3/8 Lower Head Thickness, minimum inches 4-3/8 Cladding Thickness, Repl acement Closure Head, inches nominal 0.25 Cladding Thickness, Repl acement Closure Head, inches minimum after machining 1/8 Cladding Thickness, Bottom Head, minimum 3/16 Cladding Thickness, Remainder of vessel, minimum, inches 1/8 Material Shell/Bottom Head SA-533-65 Grade B, Class 1 Steel Closure Head SA 508 Grade 3 Class 1 Carbon Steel Forgings A-508-64, Class 2 Cladding Stainless Steel (1) and NiCrFe Alloy (1)Weld deposited austenitic stainless steel with a composition approximately equivalent to SA-240, type 304 in contact with coolant.

MPS2 UFSAR4.3-35Rev. 35CEDM Nozzles Ni-Cr-Fe Alloy welded to SA-182, F316LN Instrumentation Nozzles Ni-Cr-Fe Alloy welded to SA-182, F316LN Dry Weights Closure Head lb. 147,900 Vessel, without flow skirt, lb. 682,000 Studs, Nuts and Washers, lb. 38,900 Total lb. 878,900 Volumes Bottom of Core, ft 3 1113.11 Center of Core, ft 3 1680.93 Top of Core, ft 3 2248.74 Full Vessel, ft 3 4651.27 MPS2 UFSAR4.3-36Rev. 35TABLE 4.3-2 STEAM GENERATOR PARAMETERS Number 2 Type Vertical U-Tube Number of Tubes (Design/Actual ) 8523 Tube Outside Diameter, inches 0.750 Heat Transfer Rate, each, Btu/hr 4.63 x 10 9 Nozzles and Manways Primary Inlet Nozzle (1 each), ID, inches 42 Primary Outlet Nozzle (2 each), ID, inches 30 Steam Nozzle (1 each), ID, inches 34 Feedwater Nozzle (1 each), nominal, inches 18 Instrument Taps (12 each), nominal, inches 1

Primary Manways (2 each), ID, inches 18 Secondary Manways (2 each), ID, inches 16 Secondary Handhole (4 each), ID, inches 8

Bottom Blowdown (2 each), nominal, inches 4 Nitrogen Addition (1 each), nominal, inches 1 Wet Layup (1 each), nominal, inches 2 Primary Side Design Design Pressure, psig 2485 Design Temperature, °F 650 Design Thermal Power (NSSS), MWt 2715Coolant Flow (Each), lb/hr 74 x 10 6Normal Operating Pressure, psia 2250 Coolant Volume, each, ft 3 1693 Secondary Side Design Design Pressure, psig 1000 Design Temperature, °F 550 Normal Operating Steam Pre ssure, psia 880, Full Load Normal Operating Steam Temperature, °F 530, Full Load Blowdown Flow, Design, Maximum, Each, lb/hr 112,000 MPS2 UFSAR4.3-37Rev. 35Steam Flow, Each, lb/hr 5.9 x 10 6 Steam Moisture Content, Maximum, percent 0.20 Feedwater Temperature, °F 435 Number of Steam Primary Separators, each Steam Generator 170 Number of Secondary Separators, each Steam Generator 170 Dimensions Overall Height, Including Support Skirt, inches 749 Upper Shell Outside Diameter, inches 239.75 Lower Shell Outside Diameter, inches 166 Weights Dry, lb. 1,070,400 Flooded, lb. 1,666,600 Operating, lb. 1,283,000 MPS2 UFSAR4.3-38Rev. 35TABLE 4.3-3 MAIN STEAM SAFETY VALVE PARAMETERS Main Steam Piping Design Pressure, psig 1,100 (NOTE)Main Steam Piping Design Temperature, °F 600 Fluid Saturated SteamTotal Capacity (16 valves) lb/hr 12,704,960 Material Body ASTM A105 Gr IIDisc ASTM 565 Gr 616 or ASME SB-637 UNS N07750 Type 3 Trim ASTM A451 Gr CPF8 NOTE:ASME Section III Code requires pressure rise to be limited to no more than 10% above piping design pressure. Operability of the sa fety valves ensures that the main steam system pressure will be limited to within 110% of its design pressure. Design pressure for main steam safety valves - 1035 psig.Valve NumberSet Pressure psiaCapacity lb/hr (each valve) 1, 2 1000 794,060 3, 4 1005 794,060 5, 6 1015 794,060 7, 8 1025 794,060 9, 10 1035 794,06011, 12 1045 794,060 13, 14, 15, 16 1050 794,060 MPS2 UFSAR4.3-39Rev. 35TABLE 4.3-4 TECH PUB REVIEW PKG FOR T4.3-4REACTOR COOLANT PUMP PARAMETERS Synchronous Speed, rpm 900Type Vertical, Limited Leakage, Centrifugal Shaft Seals Type N-9000 AssemblyStationary Face Ring Carbon-Morganite CNFJ Rotating Face Ring Tungsten or Silicon Carbide Design Pressure, psig 2,485Design Temperature, °F 650Normal Operating Pressure, psig 2,235Maximum Operating Temperature, °F 549 Design Flow, gpm 81,200 Total Dynamic Head, ft, Minimum 243 Maximum Flow (one-pump Operation), gpm 120,000 Dry Weight, lb.

168,050 Flooded Weight, lb.

175,050Reactor Coolant Volume, ft 3 per pump 112 Shaft Material ASTM A-182 Type F-304 Casing Material ASME SA-351 Gr CF8M Casing Wear Ring Material ASTM A-351 Gr CF8 Hydrostatic Bearing Bearing Material ASTM A-351 Gr CF8 Journal Material ASTM A-351 Gr CF8Motor Voltage, volts (source) 6,900 Voltage, volts (rated) 6,600 Frequency, Hz 60 Phase 3 Horsepower (rated)/rpm 6,500/887 Synchronous Speed, rpm900 MPS2 UFSAR4.3-40Rev. 35Total Seal Assembly Leakage (Normal and Stand-by Operation)

Three Pressure Seals Operating, gpm ** Two Pressure Seals Operating, gpm **

One Pressure Seal Operating, gpm **

• NMAC, Main Coolant Pump Seal Maintenance Guidelin es, Final Report-December 1993 can be used to calculate these volumes more accurately at different conditions. ** Six (6) resistance temperature detectors (RTD s) were imbedded into the stator (2 per phase) during fabrica tion. During first full load plant operation, all 6 RTDs were monitored. The RTD reading the highest temper ature was chosen for use. The other five RTDs remain as spares.*** Vibration Monitoring System consists of tw o (2) velomitors on the motor casing, and two (2) proximity probes and one (1) key phasor on the pump shaft.InstrumentationQuantity (per pump)Seal Temperature1Pump Casing Differential Temperature1Seal Pressure3Controlled Bleed off Flow1Controlled Bleed off Temperature1 Motor Oil Level2Motor Bearing Temperature5Motor Stator Temperature *6 Reverse Rotation Flow1Vibration Monitoring System ***5Oil Lift Pressure1 Lubrication Oil Temperature3 MPS2 UFSAR4.3-41Rev. 35TABLE 4.3-5 REACTOR COOLANT PIPING PARAMETERS Number of loops 2 Flow per loop, lb/hr 61 x 10 6 Pipe Size Reactor outlet, ID, inches 42Reactor inlet, ID, inches 30Surge line, nominal, inches 12 Design Pressure, psig 2485Design Temperature, °F 650Velocity, Hot leg, ft/sec 40.4Velocity, Cold leg, ft/sec 36.3 MPS2 UFSAR4.3-42Rev. 35TABLE 4.3-6 PRESSURIZER PARAMETERS Design Pressure, psig 2,485 Design Pressure, °F 700 Normal Operating Pressure, psia 2250 Normal Operating Temperature, °F 653Internal Free Volume, ft 3 1500 Normal Operating Water Volume, Full Power, ft 3 800Normal Steam Volume, Full Power, ft 3 700 Installed Heater Capacity, kW 1600(Note: Total Heater Capacity may be less due to Heater unavailability)

Spray Flow, Maximum, gpm 375 Spray Flow, Continuous, gpm

1.5 Nozzles

Surge Line (1) nominal, inches 12 Safety Valves (2) nominal, inches 4Relief Valve (1) nominal, inches 4 Spray (1) nominal, inches 4 Heater Sleeve (60) ID, inches 0.905 Manual Vent (1) nominal inches Manway nominal, inches 16 Alternate Vent Port nominal, inches

6.5 Instrument

Nozzles Level (4) nominal, inches 1 Temperature (2) nominal, inches 1 Pressure (2) nominal, inches 1 Materials Vessel SA-508 Grade 3, Class 2 Cladding - Cylinder Shell, Upper and Lower Head Type 308 Stainless Steel (1)

MPS2 UFSAR4.3-43Rev. 35 Dimensions Overall Length, inches (bottom of support skirt to tip of relief nozzle) 434.06 Outside Diameter, inches 106.56 Inside Diameter, inches (with cladding) 96.16 Cladding Thickness, inches (minimum) 1/8 Dry Weight, Including Heaters, lb.

202,731 Flooded Weight, Incl uding Heaters, lb.

297,433 (1)Weld deposited austenitic stainl ess steel in contact with coolant.

MPS2 UFSAR4.3-44Rev. 35TABLE 4.3-7 QUENCH TANK PARAMETER Design Pressure, psig 100 INT/15 EXT Design Temperature, °F 350 Normal Operating Pressure, psig 3 Normal Operating Temperature, °F 120 Internal Volume, ft 3 217 Normal Water Volume, ft 3 135 Normal Gas Volume, ft 3 82 Blanket Gas Nitrogen

NozzlesPressurizer discharge (1), inch, nominal 10 Demineralized water (1), inch, nominal 2 Rupture Disc (1), inch, nominal 18 Drain (1), inch, nominal 3 Temp. Instrument (1), inch, nominal 1 Level Instrument (1), inch, nominal 1 Pressure and Level Instru ment (1) inch, nominal 1 Vent (1), inch., nominal

1.5 Vessel

Material ASTM-A-240 TP 304 Dimensions Overall Length, inches 145.5 Outside Diameter, inches 60 Dry Weight, lb.

4600 Flooded Weight, lb.

18,120 MPS2 UFSAR4.3-45Rev. 35TABLE 4.3-8 PRESSURIZER SPRAY (RC-100E, RC-100F) VALVE PARAMETERS Service - Pressurizer Spray Control Design Temperature, °F 650 Design Pressure, psig 2485 Flow, gpm 440 Pressure Drop, psi 8.5 - 40 Failure Position Failed Closed Manufacturer Fisher Controls Co.

Design Code Pump and Valve Code, Nov. 1968 Draft, Class I Seismic Class I Materials Body 316 SST ASTM A351-CF8M TABLE 4.3-9 POWER-OPERATED RELIEF VALVE ISOLATION VALVE PARAMETERS (RC-403, RC-405)Service - Pressurizer Power-Operated Relief Valve Isolation Design Temperature, °F 675 Design Pressure, psig 2,485 Actuator Electric Motor Failure Position As Is ANSI Class 2,500 lb Manufacturer Velan Valve Company Design Code Pump and Valve Code Nov. 1968 Draft, Class I Seismic Class I Materials Body ASTM A182 Grade F316 MPS2 UFSAR4.3-46Rev. 35TABLE 4.3-10 PRESSURIZER POWER-OPERATED RELIEF VALVE PARAMETERS (RC-402, RC-404)

Design Pressure, psig 2485 Design Temperature, °F 675 Fluid Saturated Steam 0.1% (wt) Boric Acid Number 2 Capacity, lb/hr (minimum) each 153,000 Type Solenoid Operated Set Pressure, psig 2,385 and 400 psig (low temperature overpressurization)

Failure Position Closed Design Code ASME Section III, 1977 Edition through Winter 1979 Addenda Materials Body 316L SS, SA182 MPS2 UFSAR4.3-47Rev. 35TABLE 4.3-11 PRESSURIZER SAFETY VALVE PARAMETERS (RC-200, RC-201)

Design Pressure 2,485 Design Temperature, °F 675 Fluid Saturated Steam, 0.1% (wt) Boric Acid Set Pressure RC-200, psig 2,485 RC-201, psig 2,485 Capacity, lb/hr, at set pressure RC-200 294,000 RC-201 294,000 Type Spring loaded-balances bellows, enclosed bonnet Accumulation, %

3 Back pressure Compensation Yes Blowdown, %

12 Design Code ASME Section III, Class A, 1968 Edition, Addenda through Summer of 1970, Code Case 1344-1 Materials Body 316 SST, ASTM A 182 MPS2 UFSAR4.3-48Rev. 35TABLE 4.3-12 ACTIVE AND INACTIVE VALVES IN THE REACTOR COOLANT SYSTEM BOUNDARYLineValve Type/

NumberClassification Active - A Inactive - INormal Position Post LOCA PositionShutdown CoolingMotor / 2AClosedClosedCharging aAir / 2AOpen / ClosedOpen / ClosedCheck / 2AOpen / ClosedOpen / Closed LetdownAir / 2A Open ClosedManual / 1I Open OpenAuxiliary SprayAir / 1 I Closed ClosedCheck / 1I Closed ClosedPressurizer SprayaAir / 2 IOpen / ClosedClosedManual / 6I Open OpenPressurizer ReliefMotor / 2I Open OpenSolenoid / 2 IClosedClosedPressurizer Safety2 IClosedClosedSafety Injection TankMotor / 4I Open OpenCheck / 4AClosed OpenLeakage Control (SIS)Air / 4 IClosedClosedSafety InjectionCheck / 8AClosed Open LPSI HeaderCheck / 4AClosed OpenMotor / 4AClosed Open HPSI HeaderCheck / 8AClosed OpenMotor / 8AThrottled OpenOpen Drain: Reactor Coolant LoopManual / 13IClosedClosed Air / 1 I Open OpenChargingManual / 4I Closed ClosedPressurizer SprayManual / 4I Closed Closed LetdownManual / 4I Closed Closed MPS2 UFSAR4.3-49Rev. 35Safety InjectionManual / 14IClosedClosedAuxiliary SprayManual / 4IClosedClosed Shutdown CoolingManual / 2IClosedClosedVent / Test: Reactor VesselManual / 2IClosedClosedVent / Test: PressurizerManual / 2IClosedClosedVent / Test: Pressurizer SprayManual / 3IClosedClosedVent / Test: LetdownManual / 3IClosedClosedVent / Test: ChargingManual / 4IClosedClosed Vent / Test: Safety InjectionManual / 10IClosedClosedVent / Test: Auxiliary SprayManual / 2IClosedClosedSamplingManual / 3IOpenOpen Shutdown Cooling Relief1IClosedCloseda.Valves may be open or shut during normal operation or postincident.TABLE 4.3-12 ACTIVE AND INACTIVE VALVES IN THE REACTOR COOLANT SYSTEM BOUNDARYLineValve Type/

NumberClassification Active - A Inactive - INormal Position Post LOCA Position MPS-2 FSAR JUN 10 1982 FIGURE 4.3-1 REACTOR VESSEL MPS-2 FSARMAY 1994FIGURE 4.3-2 STEAM GENERATOR MPS-2 FSARJULY 1998FIGURE 4.3-3 REAC TOR COOLANT PUMP MPS2 UFSAR4.3-53Rev. 35FIGURE 4.3-4 P&ID REACT OR COOLANT PUMP The figures indicated above represents an engineer ing controlled drawing th at is Incorporated by Reference in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing number and the controlled plant drawing for the latest revision.

MPS-2 FSARJULY 1998FIGURE 4.3-5 RE ACTOR COOLANT PUMP SEAL AREA MPS-2 FSAR JUN 10 1982FIGURE 4.3-6 REACTOR COOLANT PUMP PREDICTED PERFORMANCE MPS-2 FSARFIGURE 4.3-7 PRESSURIZER MPS-2 FSAR Rev. 24.2FIGURE 4.3-8 TEMPERATURE CONTROL PROGRAM MPS-2 FSAR Rev. 24.2FIGURE 4.3-9 PRESSURIZER LEVEL SETPOINT PROGRAM MPS-2 FSARFIGURE 4.3-10 PRESSURIZER LEVEL CONTROL PROGRAM MPS-2 FSARMAY 1998FIGURE 4.3-11 QUENCH TANK MPS-2 FSARFIGURE 4.3-12 PERMANENT RE ACTOR CAVITY SEAL PLATE MPS2 UFSAR4.4-1Rev. 35

4.4 MATERIALS

COMPATIBILITY

4.4.1 MATERIALS

EXPOSED TO COOLANT The materials exposed to the reactor coolant ha ve shown satisfactory performance in operating reactor plants. A listin g of materials is given in Table 4.4-1.

4.4.2 INSULATION

Piping and equipment are insulated with gra nular-type, reflective-type and NUKON fiberglas-type material compatible with the temperature and functions involved. All insulation material used on stainless steel has a low (< 600 ppm) soluble chloride conten t to minimize the possibility of chloride induced stress corr osion. Removable meta l reflective-type thermal insulation is provided on weld areas of the reactor coolant system subject to inservice inspection. The replacement pressurizer is enti rely covered with reflective me tallic insulation, taking into consideration the containment sump clogging concerns as deta iled in NRC Regulatory Guide 1.82. The chemical make up of the insulation conforms to NRC Regulatory Guide 1.36. Nonremovable metal reflective-type thermal insula tion is provided on the reactor cavity wall. The primary channel heads of the tw o steam generators ar e covered with the NUKON fiberglas-type material.Removable blanket insulation is provided on the CEDM, instrumentation nozzle areas of the replacement reactor vessel head. Interconnected rigid panel insulation covers the lower portion of the head and the head flange for easy removal.

The thickness of insulation is such that the exterior surface temperature is not higher than approximately 50

°F above the maximum containment ambient (120

°F). All insulation support attachments are atta ched prior to final stress relief. The possibility of leakage of reactor coolant onto the reactor ve ssel head or other part of the reactor coolant pressure boundary causing corrosion of the pressure boundary has been investigated by Combustion Engineering (CE).

Detailed laboratory examinations have shown that: a.Reactor coolant (containing boric acid) alone, at temperatures greater than about 250°F, does not result in signifi cant corrosion of low all oy steels. Therefore, under normal operating conditions, corrosion of the pressure boundary is not a concern.

Below this temperature, boric acid solu tions can result in significant corrosion.

This corrosion is controlled with an aggressive preventative maintenance program and procedures to evaluate all unidentified reactor coolant leakage. b.Boric acid solutions dripped through the ca lcium silicate insula tion to be used on this plant do not initiate attack.

MPS2 UFSAR4.4-2Rev. 35The results of these tests have therefore shown that reactor coolant system leakage onto surfaces of the reactor coolant pressure boundary will not affect the in tegrity of the pressure boundary when calcium silicate insulation, NUKON fibergla s-type insulation or metal reflective-type insulation is used.

4.4.3 COOLANT

CHEMISTRY Control and variation of the reac tor coolant chemistry is a func tion of the chemical and volume control system. Sampling lines are provided from the reactor coolant pipi ng to provide a means for taking periodic sample of the coolant for chemical analysis. Table 4.4-2 contains the Reactor Coolant Chemistry parameters and limits listing. The water chemistr y is maintained as follows:

At temperatures below 250

°F, no upper limit on dissolved O 2 is specified.1.Hydrazine should only be added duri ng subcritical heatup at 1.5 times the measured oxygen concentration. 2.Consistent with concen tration of additives.

All wetted surfaces in the reactor coolant system are compatible with the above water chemistry.

MPS2 UFSAR4.4-3Rev. 35TABLE 4.4-1 MATERIALS EXPOSED TO COOLANT ReactorReplacement Closure Head Cladding Weld Deposit Type 309 SS (layer 1)Type 308 SS (subsequent layers)Vessel Cladding Weld Deposited Type 308 SS

• Vessel Internals 304 SS and Ni-Cr-Fe Alloy Fuel Cladding Zircaloy-4 Control Element Drive Mechanisms Ni-Cr-Fe Piping Base Metal SA 516 Gr 70 Carbon Steel ** Piping Cladding Austenitic Stainless Steel Type 304 L Steam GeneratorBottom Head Cladding Weld Deposited Type 308 SS

• Tube Sheet Cladding Weld Deposited Ni-Cr-Fe Alloy Tubes Ni-Cr-Fe Alloy Pumps Casing Austenitic Stainless Steel, Type 316 Internals Austenitic Stainless Steel, Type 316 and Type 304Pressurizer Cladding - Cylinder Shell, Upper and Lower Head Weld Deposited Type 308 and 309 SS

• Heater Sheath SA 213 TP 316 Heater Sleeve SA 182 Grade F316* Weld Deposited Austenitic Stainless Steel in contact with coolant.** Piping Base Material is exposed on instrument nozzles that have been subjected to a half nozzle replacement.

MPS2 UFSAR4.4-4Rev. 35(a)During power operation li thium is coordinated with boron to maintain a pH(t) of 7.0, but 7.4, consistent with the Primary Chemistry Control Program. Lithium is added to the RCS during plant startup, but prior to reactor criticality, and is in specification per the Primary Chemistry Control Program within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after criticality. Lithium may be removed from the reactor coolant immediately before, or dur ing, shutdown periods to aid in the cleanup of corrosion products. By eval uation, a maximum lithium concentration of 4.5 ppm is permissible with a target lithium concentration of 4.3 ppm for 100% power operations. (b)The temperature at which th e Oxygen limit applies is > 250

°F. (c)The at power operation residual Oxyg en concentration control value is ppm. (d)During plant startup, Hydrazine may be used to control disso lved Oxygen concentration at .1 ppm. (e)RCS boron concentration is maintained as necessary to ensure core reactivity or shutdown margin requirements are met. Although the RC S and related auxiliary systems containing reactor coolant are designed for a maximum concentration of 2620 ppm boron, it should be noted the design basis for the TSP baskets in the containment sump assumes the RCS, SITs, and RWST are at a maximu m boron concentration of 2400 ppm.TABLE 4.4-2 REACTOR COOLANT CHEMISTRYPARAMETERREACTOR COOLANT LIMITS Suspended Solids, ppm maximum 0.35 prior to reactor startup pH at 25°FDetermined by the concentration of boric acid and lithium present. Consistent with the Primary Chemistry Control Program. (a) Chloride, ppm Cl

-, maximum 0.15 Fluoride, ppm F

-, maximum 0.10 Hydrogen as H 2 , cc (STP)/Kg H 2O25-50 Dissolved O 2 , ppm maximum 0.1 (b) (c) (d)

Lithium as Li 7 , ppm Consistent with the Primary Chemistry Control

Program (a) Boron, ppm 0-2620 (e) Conductivity, µS/cm at 25

°CRelative to Lithium and Boron concentration MPS2 UFSAR4.5-1Rev. 35

4.5 SYSTEM

DESIGN EVALUATION

4.5.1 PREVENTION

OF BRITTLE FRACTURETo protect against non-ductile failure, the requirements of 10 CFR 50 Appendix G have been implemented and the requirements of 10 CFR 50.61 have been satisfied.10 CFR 50 Appendix G provides fracture toughness requirements which ensure sufficient margins of safety against non-ductile failure during normal operation, including anticipated operational occurrences and inserv ice hydrostatic tests. The requi rements of this Appendix are implemented by Figure 4.5-4 which provides ma ximum heatup and cool down rates for the reactor coolant system and maximu m reactor coolant system pressure (as indicated by pressurizer pressure) as a function of reactor vessel inlet temperature. Additional discussions addressing the development of these limits ar e provided in Sections 4.5.1.2 and 4.5.1.3.10 CFR 50.61 provides additional fracture toughness re quirements to protec t against non-ductile failure of the reactor vessel during pressurized therma l shock (PTS) events. Compliance with these requirements is demonstrat ed by ensuring that the end-of-life reference temperature (RT PTS) for the reactor vessel beltline materials stays below the established limits using the prescribed methods. Additional discu ssion addressing 10 CFR 50.61 is provided in Section 4.5.1.4.

4.5.1.1 Initial Nil-Ductility Trans ition Reference Temperature.

The original design requirements for carbon and lo w alloy materials which form the pressure boundary of the reactor coolant system (RCS) have impact properties which meet the requirements of paragr aph N-330 of the Summer 1969 ASME Bo iler and Pressure Vessel Code,Section III, at 40

°F or less.To address changes in regulations and demonstrate compliance with 10 CFR 50 Appendix G and 10 CFR 50.61 requirements, the original design requirements of N-330 we re supplemented and the materials initial Nil-Ductility Transition Reference Temperature (RTNDT) were subsequently established.

The impact properties for the replacement reactor vessel closure head meet the requirements of Article NB 2331 of the ASME B&PV Code,Section III, 1998 Edition through 2000 Addenda. Reference temperature RTNDT of - 40°F was established in accordance with SA 508 Supplementary Requirement (S10) and NB 2300.

The replacement reactor vessel closure head and the replacement pressurizer were evaluated for protection against non-duc tile failure in accordance with the methodology presented in ASME Section III, Appendix G for Class 1 components. The maximum stre ss intensity factors for the transients meet the fracture toughness require ments of ASME Section III Appendix G for a postulated defect of 1/

10th thickness of the reactor vessel head and 1/4th thickness of the pressurizer.

MPS2 UFSAR4.5-2Rev. 35 In the case of the replacement steam generators, the materials were required to satisfy the requirements of ASME,Section III, 1983 Edit ion through the Summer 1984 Addenda, Article NB-2331, and establish a maximum RTNDT at a temperature of 0

°F.4.5.1.2 Nil-Ductility Transition Reference Temperature Shift The flux of fast neutrons at th e reactor vessel wall is governed by the reactor core load design, arrangement of the reactor internals and average power leve ls, among other factors. The integration of this flux over time , called fluence, is monitored by dosimetry materials included in the surveillance capsules located near the vessel inner wall. In the time since original plant startup the thermal shield has been removed and the co re load design has changed. In addition, small crack-arresting holes have been drilled in the core barrel, affecting the flux for one vessel plate.

Based on results from survei llance capsule dosimetry retrieved through cycle 14, the known configuration history, neutron flux modeling calculations described below, and further assuming future reactor operation with the low leakage design at 2700 Mwt and 90 percent plant capacity factor, the maximum fluence at the end of the period of extended opera tion of 60 years will be approximately 3.83 x 10 19 n/cm 2 (E > 1Mev). These fluence estimates are fully described in Reference 4.5-2. The following summ ary describes the key methods of the referenced analysis.

Discussion of Fluence Calculations The prediction of neutron fluenc e at various locations was base d on an analysis of neutron transport for given configurations and source developed to model the reactor. To further refine the model accuracy, the results of the fluence analysis were correlated with measurements of actual fluence based on dosimetry retrieve d from surveillance capsules. Estimations of future fluence are then based on a combination of the currently measured fluence and the extrapolation of calculated cycle 14 flux though the end of licensed plant life. T hus, it is assumed there will be no significant change to reactor configur ation or core load design.

The three dimensional discrete ordinates transport computer code DORT was used in the surveillance capsule W-83 analysis to model neutron transport within the reactor. The code results describe the space and energy dependent neutron flux present in the reactor.

The reactor was modeled as a 1/8th segment of the core, the reactor internals, core barrel, thermal shield (through cycle 5), expl icit representations of th e surveillance capsules at 6

° and 14°, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, the primary biological cla dding and shield wall.

Note that a variation of this model, in which ma terial composition of the surveillance capsules was redefined as water, was utilized to determin e the maximum neutron expo sure at the pressure vessel wall in octants of the core that do not contain surveillance capsules.The actual activation or fission measured for the retrieved capsule dosimetry was compared to the calculated activation. The ratio of measured to calc ulated activation, M/C, was determined to fall well within the criteria spec ified in Regulatory Guide 1.190.

MPS2 UFSAR4.5-3Rev. 35Shift Prediction The shift prediction methodology fo r the determination of the ad justed reference temperature (ART) is done in accordance with Revision 2 to Regulatory Guide 1.99 (dated May 1988). The ART, for each material in the beltlin e is given by the following expression: ART = Initial RTNDT + RTNDT + MarginThe shift in RTNDT is based on fluence predictions, de scribed above, for the time period corresponding to 54 EFPY. Table 4.6-13 pr ovides the results of the calculation.

4.5.1.3 Operational Limits All components in the RCS are de signed to withstand the ef fects of cyclic loads due to RCS temperature and pressure changes. These cy clic loads are introduc ed by normal unit load transients, reactor trips and startup and shutdown operation.

During unit startup and shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown is based upon a rate of 100

°F/hr and for cyclic operation.

The maximum allowable RCS pressu re at any temperature is base d upon the stress limitations for brittle fracture considerations. These limitatio ns are derived by using the rules contained in Section III of the ASME Code including Appendix G, Protecti on Against Nonductil e Failure and the rules contained in 10 CFR 50, Appendix G, Fracture T oughness Requirements.

FSAR Figures 4.5-4 and 4.5-5 provide the RCS pressure-tempera ture limitations during plant heatup and cooldown. Figures 4.5-4 and 4.5-5 are valid for the period up to and including fifty four years of full power inte grated neutron flux, as determ ined using the rationale of Section 4.5.1.2, and was develope d using the rules of Appendix G, "Protection Against Nonductile Failure" of the ASME Boiler and Pressure Vessel Code,Section XI, 2002 Addenda. Using these rules the belt line material of the reactor vessel is established as the controlling component section throughout plant life. This established an uppe r boundary on RCS pressure as a function of RCS temperature a nd allowable heatup and cooldown ra tes to ensure prevention of nonductile failure.

The limitations for normal heatup and cooldown rates and the applicable temperature ranges are summarized in Table 4.5-2.

The RCS pressure-temperature li mits provided by Figures 4.5-4 a nd 4.5-5 have been corrected to indicated pressurizer pressure versus indicated cold leg temperature.

Indicated cold leg temperature is the best available indication of the reactor vessel downcomer temperature and will normally be monitored as RCS cold leg te mperature when reactor coolant pumps are operating or natural circulation is occurring. In th e instances where the shutdown cooling (SDC) system is operating without RCP's, the SDC system return temper ature will be MPS2 UFSAR4.5-4Rev. 35 used. Appropriate corrections which account for dynamic fl ow losses, static elevation differences and instrumentation uncertainties have been applied.

Also shown is an allowable region for shut down cooling system ope ration. This region is established based upon the design pressure-temperature ratings of components of the shutdown cooling system and the normal ope ration of this system as described in Chapter 6 (Section 6.3) and Chapter 9 (Section 9.3).

The reactor vessel beltline material consists of six plates. The NDT temperatures (TNDT) of each plate was established by drop weight test (DWT). Charpy tests were then performed to determine at what temperature the plates exhibited 50 ft-lbs absorbed ener gy and 35 mils lateral expansion. (Data points were based on av erage of three specimens.)

Similar testing was not performed on all remaining material in the RCS. However, sufficient impact testing was performed to meet appropriate design code requirements and a conservative RTNDT of 50°F has been established for longitudinal direction.

As a result of fast neutron irradi ation in the region of the core, th ere will be an increase in the RTNDT with operation. The techniques used to predict the integrated fast neutron (E 1 Mev) fluxes of the reactor vess el are described in Section 4.5.1.2 of the FSAR.

Since the neutron spectra and flux measured at the samples and reac tor vessel inside radius should be nearly identical, the measured reference transition temperat ure shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured samp le exposure by applicat ion of the calculate d azimuthal neutron flux variation.The actual shift in RTNDT will be established periodically during plant operation by testing of reactor vessel material sample s which are irradiated cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.6.3 and shown in Figure 4.6-1 of the FSAR. To compensate for any increase in the RTNDT caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown.

In addition to the requirements provided by ASME Section XI, Appendix G and 10 CFR 50, Appendix G, the following items were also considered in the development of the RCS pressure-temperature limits provided by Figures 4.5-4 and 4.5-5.Lowest Service Temperature - As indicated previously, an RTNDT for all material with the exception of the reactor vessel beltline was established at 50

°F. ASME III, Art. NB-2332(b) requires a lowest service temperature of RTNDT + 100°F for piping, pumps and valves. Below this temperature a pressure of 20 perc ent of the system hydrostatic te st pressure cannot be exceeded.

MPS2 UFSAR4.5-5Rev. 35a.Maximum Pressure for Shutdown Cooli ng - This pressure is established by considering the design pressure of the s hutdown cooling system, shutof f head of the Low Pressure Safety Injection (LPSI) pumps, elevation head from the pressurizer to the LPSI pumps, margin to SDC safety valve setpoints and the design temperature of the shutdown cooling system.

4.5.1.4 Pressurized Thermal ShockIn accordance with 10 CFR 50.61, reactor pressure ve ssel belt line materials have been reviewed to establish a reference temperature for pressurized thermal shock (RT PTS). This review evaluated core loading patterns and the actu al or best estimate of copper and nickel in the vessel material. In addition, the reactor vessel material compositi on and properties were co mpared to those of surveillance capsule materials fr om which actual tests and measurements were taken. A summary of this review is as follows:a.Copper/Nickel Content Best estimate copper/nickel content values for the Reactor Vessel Beltline Plates and welds are given in Table 4.6-12.*Plates; Full chemistry results available for beltline plates.*Welds; Where chemistry results were unavailable, nickel content was conservatively estimated using data available for the type of wire used.b.Core Configuration The neutron fluence values given in Table 4.6-12 have been used and have been calculated as described in Section 4.5.1.2. These end-of-life fluence values represent the most recent surveillan ce capsule (capsule W-83) evaluation.Calculated RT PTS values have been obtained using the above assumptions. Table 4.6-12 provides the results of the calcula tions. This table will be updated wh enever 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 at 54 effective full power years.

4.5.2 SEISMIC

DESIGN The nuclear steam supply system (NSSS) is designed to withst and the loads imposed by the maximum hypothetical accident and the maximum seismic disturbance wit hout loss of functions required for reactor shutdown a nd emer gency core cooling. The method of combining stresses produced by these simultaneous condi tions is described in Section 4.2.1.

MPS2 UFSAR4.5-6Rev. 35 A seismic analysis of the RCS is given in Appendix 4.A.

The RCS components are consider ed Class 1 for seismic design.

Loadings which result from earthquake conditions are categori zed and, in combinations with other specified loadings, are evaluated in accordance with the rules of ASME Boiler and Pressure Vessel Code,Section III.a.Operating Basis Earthquake (OBE) - The OBE condition is categorized as an upset condition. In evaluations using normal a nd upset conditions, loadings resulting from the OBE shall be considered to o ccur during normal operation at full power. 200 cycles of the OBE have been specifi ed in the system design. The procedure used to account for the number of earthquake cycles during one seismic event includes consideration of the number of significant motion peaks expected to occur during the event. The number of si gnificant motion peaks during one seismic event would be expected to be equivalent in severity to no more than 40 full load cycles about a mean value of zero and with an amplitude equal to the maximum response produced during the entire event. Based upon this consideration and the assumption that seismic events equivalent to 5 OBEs will occur during the life of the plant, Category I systems, components and equipment are designed for a total of 200 full load cycles.b.Design Basis Earthquake (D BE) - Two faulted conditions , which include loadings resulting from the DBE, are defined.*Loadings resulting from the combin ed ef fects of the DBE and normal operation at full power.*Loading resulting from the combined effects of the DBE, normal operation at full power and pipe rupture conditions.

4.5.2.1 Piping The primary stress limits applie d in evaluating the emer gency a nd faulted conditions for the RCS piping are specified as follows:The RCS piping is designed in accordance with the requirements fo r Class I piping of ANSI B31.7, Code for Nuclear Power Piping.

The primary stress limits of ANSI B31.7, Case 70, Design Criteria for Nuclear Power Piping Under Abnormal Conditions, are applied in evaluating the emer gency and faulte d conditions, except in the case of Faulted Condition (1) noted above in Section 4.5.2.b. In this case, the primary stress limits for emergency conditions (Case 70) are applied.

4.5.2.2 Vessels The primary stress limits applied in evaluating the emer gency and fa ulted conditions for vessels in the RCS are specified as follows:

MPS2 UFSAR4.5-7Rev. 35The reactor vessel, steam generators and pressu rizer were designed in accordan ce with the rules for Class A vessels of the ASME Boiler and Pressure Ve ssel Code Section III. The replacement steam generator subassemblies ar e designed in accordance with the rules of Class I vessels, ASME Section III. The primary stress limits of Section III, Paragraph N-417-10, Stress Limitations for Emergency C onditions, are applied in evaluating the emergency conditions and in evaluating Faulted Condition (1) noted above in Section 4.5.2.b.With respect to Faulted Condition (2) in Section 4.5.2.b, the primary stress limits of Section III, N-417-11, Stress Limitations fo r Faulted Conditions, modified as follows, are applied.a.In lieu of the value suggested in N-417.11b, the yield strength value to be used in applying the limit analysis procedure will be equal to tabulated yield strength plus one-third of the difference between the tensile strength and the tabulated yield strength, with values taken at temperature.b.In the piping run within which a pipe br eak is considered to have occurred, the stress criteria for this condi tion need not be applied to th e relevant nozzles or to the nozzle-vessel region within the limits of reinforcement given in N-454(a), except

in the case of nozzles integral with com ponent support assemblies. In the case of nozzles integral with compone nt support assemblies, the crit eria is applicable in all regions which sustain support reactions.The replacement reactor vessel closure head is designed in accordance with ASME Section III, Class 1 vessels, 1998 Edition thro ugh 2000, Addenda. The primary stress limits of Section III, NB 3225 Appendix F, F-1331.1 (a, b a nd c) are applied in evalua ting the faulted conditions (Table 4.2-2B). The load combination for the fa ulted condition (Service Le vel D) is defined in Section 4.5.2.b. DBE and LOCA load s are combined using the S quare Root Sum of Squares (SRSS) method for evaluating the faulted condition. The emergency conditions (Servi ce Level C) are bounded by the design conditions and therefore, the replacement reactor vessel head is evaluated using the primary stress limits of NB 3221.The pressurizer was replaced to the design requirements of ASME Boiler and Pressure Vessel Code Section III, Subsection NB, 1998 Edition through 2000 Addenda. The pr imary stress limits for the emergency and faulted c onditions are applied in accordan ce with subsect ion NB 3000 of the design code.

4.5.2.3 Pumps and Valves The primary stress limits applied in evaluati ng the emer gency and faulted conditions for the pumps and valves in the RCS are specified as follows:

The RCS pumps and valves are de signed in accordance with the rules of ASME Code for Pumps and Valves for Nuclear Power -

March 1970, Draft. Supplementary to these rules, the primary stress limits used for Vessel, as discussed under Item b. above, are applied in evaluating the emergency and faulted conditions for the reactor c oolant pumps (RCP). In the case of valves, the MPS2 UFSAR4.5-8Rev. 35rules of the Pump and Valve Code are considered adequate to assure th at, as regards primary stresses in the pressure boundary, the piping, not the valves, will be the limiting element.

Therefore, no supplementary crite ria which limit the pr imary stresses duri ng abnormal conditions are necessary.

4.5.3 OVERPRESSURE

PROTECTION 4.5.3.1 Overpressure Protection Du ring Normal Operation The RCS is structurally designed for operation at 2485 psig and 650

°F (pressurizer 700

°F). Operation of the system 2235 psig nominal and 600

°F will result in material stresses 90 percent of design values. Detailed structur al analyses have been perfor med by the component vendors and reviewed independently by Comb ustion Engineering (CE) for all portions of the system. Welding materials used have physical properties superior to the materials which they join. Inspection procedures and tests specified a nd independently reviewed by CE were carried out to assure that pressure-containing components ha ve the maximum integrity obt ainable with present code-approved inspection techniques.

The RCS is protected against overpressure by tw o ASME Code approved safety valves which limit system pressure to a maximum of 110 pe rcent of design. In addition, two solenoid-operated relief valves are provided as described in Section 4.3.7.

4.5.3.2 Low Temperature Overpr essurization Protection The RCS low temperature ove rpressurization protection (L TOP) system, along with administrative procedures, provides protection against exceeding the ASME Section III, Appendix G (Protection Against Br ittle Fracture) require ments during cold plant conditions; i.e., temperature

< 275°F. The LTOP system consists of two redundant relief trains each with one power operated relief valve (PORV) with a setpoint of 400 psig and associated relief piping as described in Section 7.4.8.

4.5.4 REACTOR

VESSEL THERMAL SHOCKAn analysis of the thermal stresses produced in th e reactor vessel wall due to the operation of the safety injection system has been performed. Th e analysis has been reported in a CE report "Thermal Shock Analysis on Re actor Vessels due to Emergenc y Core Cooling System (ECCS) Operation," A-68-9-1, and was submitted for the record on Docket Number. 50-309, Maine Yankee Atomic Power Station. The re sults show that there will be no failure of the reactor vessel due to brittle fracture.Work has also been performed to refine the surface heat transfer coeffi cient. The temperature quench data obtained during the heat treatment of several heavy section steel plates was reviewed. With this background, CE planned and conducted additional quench tests to develop experimental heat transfer coefficients.

MPS2 UFSAR4.5-9Rev. 35 These tests were performed on a plate approximately 2 foot by 2 foot by one-half foot thick and instrumented with 11 thermocouples. The plate was heated to 550

°F and lowered quickly into an agitated (turbulent) water bath at 80

°F, nearly duplicating the temperature c onditions which would be present in th e reactor during ECCS operation. The te mperature of all thermocouples was recorded throughout the cooldown of the plate.Subsequently, the data was compared to a heat transfer computer model of the plate to obtain an effective heat transfer coefficient. A detailed report covering this work, entitled "Experimental Determination of Limiting Heat Transfer Coefficients during Quenching of Thick Steel Plates in Water" (A-68-10-2, December 13, 1968), was submitted to the AEC (now NRC) and made part of the public record. The report concludes that an effective heat transfer coefficient of 300 Btu/hr-ft 2-F provides a realistic upper limit for thick steel plates que nched in highly agitated room temperature water.The stress near the tip of axia l and circumferential vessel crac ks of various depth has been determined by the finite elemen t method. This work was reported by a CE report, "Finite Element Analysis of Structural Integrity of a Reactor Pressure Vessel during Emergency Core Cooling,"

A-70-19-2, January 1970, and is part of the public record.

These reports substantiate the an alytical conclusion that a vessel failure will not occur due to ECCS operation. An acute crack, even if formed, will not propagate.

4.5.5 LEAK DETECTION Methods are provided to alert the operator of the presence of leak age from the RCS in a timely manner to allow detection and isolation of the leak to ensure the leakage does not exceed acceptable limits. Detection of leaks from the RCS can be accomplished by one or a combination of the below listed means.Leak Detection Within the Containment Leaks within the containment may be indicated by:a.Increased pressure and temperature in the containment;b.Monitoring the normal containment sump level;c.An increase in airborne activity as measured by the contai nment air radiation monitor system. The sensitivity and response time of the par ticulate and gaseous detectors are dependent on many factors. Although the airborne activity detectors may very well give an early warning of an RCS leak, any correlation of these radiation monitor readings and the RCS leak age rate will be weak. These monitors are best used for trending purposes and a trigger to check other indications for

leakage source; MPS2 UFSAR4.5-10Rev. 35d.Monitoring the reactor building closed cooling water (RBCCW) temperature to and from the containment air recirc ulation (CARS) and cooling units.Reactor Coolant System Leakage from the RCS is indicated by the level in the pressurizer or in the primary drain tank and/

or high RCS makeup flow from the ch emical and volume control system.Relief and Safety Valves Located on the Reactor Coolant System Piping from the relief and safety valves located on the pressurizer is provided with temperature sensors with readout in the main control room. Any te mperature increase will indicate relief or safety valve leakage. In additi on, an increase in pressurizer quench tank level, pressure, or temperature will also indicate leakage.Reactor Vessel Head Closure The space between the double O-ring s eal is monitored to detect an increase in pressure, which indicates a leak past the inner O-ring. A leak indicator in cont ainment indicates pressure. Upon high temperature a contro l room alarm is sounded.Leakage Through Steam Generator Tubes or Tube Sheet An increase in radioactivity indi cated by radiation moni tors for the gases fr om the condenser air ejectors or steam generator bl owdown system monitors will indicate leakage through steam generator tubes to the secondary side. The N-16 radiation monito rs will indica te primary to secondary leakage when the reactor is at power.

4.5.6 PREVENTION

OF STAINLESS STEEL SENSITIZATION Sensitization of stainl ess steel occurs when unst abilized T ype 300 Series st ainless material is held in the temperature range of 900-1400

°F for sufficient time to fo rm a continuous network of chromium carbide precipitates.

Sensitization occu rs after approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> at 900

°F, as compared to one hour at 1400

°F. Stabilized Type 300 Series stai nless material avoids continuity of chromium carbide precipitat es in the grain boundaries by careful control of metal chemistry.

There are no furnace sensitized st ainless steels in the reactor coolant pressure boundary (RCPB).

Sensitization is preclude d from the NSSS through mate rials selection and c ontrol to all welding and heat treating procedures.Major portions of the RCPB in CE's nuclear plants are shown in preceding tables in this section to be formed by carbon steels and a high nickel base alloy. None of these mate rials is susceptible to furnace sensitization (a conti nuous network of iron-chromium grain boundary carbides) in the sense of unstabilized Type 300 Seri es stainless steels. All internal carbon steel surfaces are weld-deposit or roll-bond clad with Inconel or stai nless steel, to preclude excessive corrosion.

MPS2 UFSAR4.5-11Rev. 35 Internal surfaces of the reactor vessel, pressurizer and steam gene rator primary side are overlaid with Type 308, 309 or 308L weld deposited metal. Weld metal composition is carefully controlled to overcome interface dilution and promote an austenoferritic duplex structure. Therefore, during the stress relief heat treatment (1150

+/-25°F) required by the ASME Code for the pressure vessel, a continuous network of chromium carbide precipitates is not formed in the Type 308 or 308L weld overlay even though this material has been su bjected to a furnace heat treatment. The delta ferrite acts as a carbon sink and prevents continuity of carbide precipitates.

Extensive testing has confirmed that, properly formulated (a duplex structure), Type 308 weld deposited metal does not form a continuous car bide network within grain boundaries even following a typical vessel post weld heat treatment (viz, 1150

°F for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />). Hence, the material is immune to intergrannular corrosion.

The replacement reactor vessel head is overlaid inside with stainless steel weld material. The first layer is type 309 stainless steel and all subsequent layers including th e final layer that is in contact with the reactor coolant is type 308L stainless steel.

Susceptibility to underclad / re heat cracking was minimized by c ontrolling the welding heat input and chemical composition of the RV Head forging to maintain the Delta G function less than 0.1 in accordance with th e correlations deve loped by the vendor (Mitsubi shi Heavy Industries). In addition, in order to comply wi th the intent of Re g Guide 1.43, examinations were performed on three representative penetration holes in the RV Head "J" groove preparations pr ior to depositing the buttering to specifically check for underclad / reheat cracking on the susceptible area up to 1Ú2 inch below the fusion line between the base material and cladding.The bimetallic weld on the CEDM and instrumentat ion nozzles for the replacement head joins the stainless steel threaded connector or flange to the alloy 690 penetration tube. Additionally, the materials for welding to Alloy 690 have lower sus ceptibility to PWSCC th an the original alloy 82Ú182 welds.

The J-groove weld design for the guide tubes is modified to control the amo unt of overwelding to a minimum and thus minimize residual weld stresses. The weld surface finish was also improved to facilitate surface examinations.

All other Type 300 Series st ainless steel used either is not subj ected to a furnace sensitization heat treatment or, as is the case of cladding on the primary piping, is of Type 304L (low carbon, 0.03% max.) composition a nd is not susceptible to the format ion of continuous chromium carbide grain boundary networks.

The replacement pressurizer is overl aid inside with stainless steel weld material. Susceptibility to underclad/reheat cracking was precluded by performa nce testing of the cl adding weld procedure specifications in accordance with NRC Regulatory Guide 1.43.

The RCS pump casing is CF8M (Cast 316), whic h again is a duplex material. The casting is solution annealed after welding; hence, this component will not have a sensitized structure.

MPS2 UFSAR4.5-12Rev. 35 Nitrogen-enhanced stainless stee l was not used in the fabrication of any RCPB component.

Because carbon steel piping is used in the RCS, no safe ends are required on the reactor vessel, or steam generator primary nozzles. Where small diameter solid stai nless pipes are employed (or in the instance of welding the coolant pump casing to carbon steel), an Inconel-182 weld deposit is built up on the nozzle prior to vess el post-weld heat treatment. Thereafter, an annealed stainless steel safe end is shop welded to the Inc onel-182 buildup using alloy 82/182 weld metal.

In joining small diameter annealed solid stainless steel piping, as is used in the pressurizer surge line, charging pump lines and safety injection systems, some carbide precipitation will occur as a result of welding. However, the precipitation that occurs in the weld heat affected zone does not sensitize the material in the context of forming continuous gr ain boundary carbide precipitates.

Samples from typical such welds pass the industry accepted standard for intergrannular corrosion susceptibility (i.e., Strauss Test - ASTM-A393). Metallographic examin ation of such welds reveal that only discontinuous grai n boundary precipitates are present.

The following four weldi ng processes are used to weld stainless steel in CE's NSSSs. Welding processes are performed in accordance with wr itten procedures, as pr ovided in the Quality Assurance Program. Nitrogen is not used as a purge gas in the welding process in lieu of argon or helium gas.

• Shielded Metal Arc (SMA)*Gas Tungsten Arc (GTA)*Gas Metal Arc (GMA)*Submerged Arc (SA)a.Shielded metal arc (SMA) is a process wherein coales cence is produced by heating with an electric arc drawn between a fl ux covered metal electrode and the work.b.In the gas tungsten arc (GTA), coalescen ce is produced by heating with an electric arc drawn between a tungsten el ectrode and the work. Filler metal, if required, is added by feeding a bare metal rod or wire into the weld pool. Shielding of the weld is obtained from an inert gas mixture.c.With gas metal arc (GMA), coalescence is ef fected by heating with an arc drawn between a continuous feed wire electrode and the work. Shielding of the weld is obtained from an external ly supplied inert mixture.d.Submerged arc (SA) produces coalescence by heating with an arc or arcs drawn between a bare metal (fille r) electrode or electrodes and the work. The arc and weld are shielded by a blanke t of granular fusible flux.

MPS2 UFSAR4.5-13Rev. 35Table 4.5-1 lists the nozzles on the steam generator, reactor vesse l, pressurizer and piping. The table also indicates the size of the nozzle, base material of th e nozzle and, where applicable, the material of the nozzle safe end.

The procedures used in welding nozzles within CE manufacturing facil ities are generally as follows: (1) For nozzles with stainless steel safe ends, the safe e nds are not attached until after final stress relief, and (2) the stainless steel safe end is welded to Inconel buttering on the alloy steel and the weld made using Inconel weld wire. With this procedur e furnace sensitizing of stainless steel is precluded.

During manufacture of the core st ructures, various parts of the core structure are tested for sensitization using the Strauss Test (ASTM A393). Test specimens consist of: (1) mockups of various welded joints, and (2) monitoring specimens included in any heat treatment of various components. None of the specim ens tested in conjunction with fabrication of reactor vessel internals for previous CE plants have failed the Strauss Test. The replacement reactor vessel head and replacement pressurizer weld materials and welding are controlled in accordance with Regulatory Guide 1.44, May 1973, "Control of the use of Sensitized Steel", to preclude sensitization of austenitic stainless steels.

The typical weld heat input with the above processes as used by CE to joint Type 300 Series stainless steel varies from 6000 joules per inch GTA to 96,000 joules per inch SA. To avoid weld heat affected zone sensitizati on, CE limits the inte rpass temperature on multipass welds in stainless steels to 350

°F maximum. The replacement pressuri zer uses only low carbon stainless steel materials. The CE interpass temperature limi ts do not apply to the replacement pressurizer. AREVA limited the interpass temperature fo r low carbon stainless steel materials to 250

°C (482°F), which is sufficient to prevent sensitization. The combination of normal heat input using the above welding procedures an d control of interpass temper ature assures minimum carbide precipitation in the weld heat affe cted zone. Samples from large welds have been examined in the laboratory and none has failed the Strauss Test.

In field welding operations, Bechte l uses welding procedures that limit heat input to the weld areas, and thus preclude th e possibility of sensitization of austenitic stainless steel

s. Most of the welding employed is of the manual SMA process; a minor amount of GMA welding is also used.

Neither one of these processes w ould be classified as an exces sively high heat input welding procedure.

Further precautions employed to preclude field sensitization of austenitic stainless steels consist of:a.Preheat and interpass temperatures are limited to 350

°F maximum.b.Controlled welding sequence is used to minimize heat input.c.The practice of block welding is prohibited.

MPS2 UFSAR4.5-14Rev. 35d.Postweld heat treatment is prohibite d on equipment and/or parts that are completely or partially fabricated of austenitic stainless steel. During the fabrication of the replacement pres surizer post weld heat treatment was performed on the safety/relief nozzles and spray noz zle safe ends. Nozzle welds were post heat-treated using a Post Weld Heat Treatment (PWHT) procedure specially qualified for the heat of material used for the safe ends in accordance with Regulatory Guide 1.44 to preclude sensitization.e.Application of heat to correct weld distortions resulting in d imensional deviations in equipment and/or parts fabricated of austenitic stainless steel is prohibited.In preparing for, and engineering, the field welding requirements, close liaison is maintained

between Bechtel and CE. Detailed welding parameters prepared by CE are submitted to CE for review and mutual concurrence and approval before they are adopted for use. Bechtel quality assurance procedures fo r field welding are discussed in A ppendix 1B. Appendix 1B was located in the original FSAR dated August 15, 1972.

The delta ferrite content of all austenitic stainless steel weld metals used to fabricate CE's RCS components is controlled to 5-18% in the as-d eposited condition. Delta ferrite content is confirmed from chemical analysis and the Schaeffler or Mc Kay Diagrams. In addition, a calibrated ferrite measuring instru ment (Seven Gauge or similar) is used. The ferrite requirement is met for each heat and/or lot of filler metal used in fabrication. The ferr ite content in the weld materials and welding for the replacement reactor vessel head and replacemen t pressurizer are in accordance with Reg Guide 1.31, "Control of Ferrite in Stainless Steel weld metal."Where field welding of austenitic stainless steels is required in the RCS only the inert GTA and manual SMA processes are used.

The welding procedures were qualified in accordance with Section IX of the ASME Boiler & Pressure Vessel Code. Tensile test sp ecimens were taken to exhibit the tensil e strength of the welded joints and bend tests were taken to indicate the ductility of the weld joints. Filler materi al compositions are in accordance with the ASME SFA/AWS filler material specification and are selected on the basis of the austenitic stainless steels to be welded.

In addition to these requirements the filler materials must be capable of depositing 8 to 25%

ferrite. This is verified for each heat or lot of fi ller material by plotting the heat analysis for bare wire or the analysis of an al l weld metal deposit for covered electrodes on the Schaeffler, DeLong or equivalent diagram to determine its ferrite conten t and the acceptability of the filler materials. The austenitic stainless steel materials includi ng cladding are analyzed for delta ferrite in accordance with NB-2433. The FN shall be 5FN to 15FN. The FN for undiluted ER309L deposit shall be in the range of 5FN - 22FN. By control of the welding processes, the filler materials and the welding parameters (by specifying a maximum interpass temperature of 350

°F), the welds should contain sufficient delta-ferrite in the austenitic matrix to avoid hot cracking in the austenitic stainless steel welds.

MPS2 UFSAR4.5-15Rev. 35 4.

5.7 REFERENCES

4.5-1W. G. Counsil (NU) letter to J. R. Miller (NRC), "Millst one Nuclear Power Station, Unit No. 2, Proposed Revisions to Technical Specifications, Pressure-Temperature Curves" (January 4, 1984), Attachment 2 - CE Report TR-N-MCM-008, "Evaluation of Irradiated Capsule W-97" (April, 1982).4.5-2S. E. Scace (DNC) to U.S. NRC, "Mil lstone Nuclear Power Station, Unit No.2, Submittal of Third Reactor Vessel Surveillance Capsule Report," (February 2003), Enclosure - WCAP-16012 Revision 0, "Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," (February 2003).

MPS2 UFSAR4.5-16Rev. 35TABLE 4.5-1 REACTOR COOLANT SY STEM COMPONENT NOZZLES, NOZZLE SIZES AND NOZZLE MATERIALS Steam GeneratorComponent(Number)SizeMaterial Primary Inlet(1)42 inch ID Car bon steel safe ends clad with stainless steel Primary Outlet(2)30inch ID Carbon steel safe ends clad with stainless steelPressure Taps(4)1inch Schedule 160Inconel B-166Steam Outlet(1)36 inch ID Carbon steel Feedwater(1)18 inch Schedule 80Carbon steelBottom Blowdown(2)4 inch ID Carbon steel Liquid Level(12)1 inch Schedule 80/

Schedule 160 Carbon steel/Inconel B-166Nitrogen Addition(1)1 inch Schedule 80Carbon steelWet Layup(1)2 inch Schedule 160Carbon steelReactor Vessel and HeadComponent(Number)SizeMaterial Primary Outlet(2)42 inch ID Car bon steel clad with stainless steel Primary Inlet(4)30 inch ID Car bon steel clad with stainless steel CEDM/HJTC(69)2.718 inch IDNi-Cr-Fe and stainless steel Instrumentation(8)4.625 inch IDNi-Cr-Fe and stainless steelVent(1)0.75 inch Schedule 80Inconel SB-167 PipeComponent(Number)SizeMaterialSurge Nozzle(1)12 inch Schedule 160A-105 Gr II with A-351 Gr CF8M safe end MPS2 UFSAR4.5-17Rev. 35Pressure(8)0.75 inch Schedule 160Inconel B-166 with A-182 Type 316 safe endRTD(25)1 inch nominalInconel B-166Shutdown Cooling(1)12 inch Schedule 140A-105 Gr II with A-351 Gr CF8M safe endSpray(2)3 inch Schedule 160A-105 Gr II with A-182 Type 316 SS safe endSafety Injection(4)12 inch Schedule 140A-182 F1 with A-351 Gr CF8M safe endCharging Inlet(2)2 inch Schedule 160A-105 Gr II with A-182 Type 316 SS safe endSampling(2)0.75 inch Schedule 160Inconel B-166 with A-182 Type 316 SS safe endDrain/Letdown(5)2 inch Schedule 160A-105 Gr II with A-182 Type 316 SS safe endPump(8)30 inch IDA-516 Gr 70 clad with A-240, Type 304L SS with A-351, CF8M safe end PipeComponent(Number)SizeMaterial MPS2 UFSAR4.5-18Rev. 35TABLE 4.5-2 REACTOR CO OLANT SYSTEM HEATUP AND COOLDOWN LIMITS Cooldown Heatup aa.These limitations apply to hydrostatic and leak test conditions.

Indicated Cold Leg TemperatureLimit Indicated Cold Leg TemperatureLimit220°F 50°F/hour 200°F 60°F/hour> 220°F 100°F/hour200°F < T 275°F 80°F/hour> 275°F 100°F/hour MPS-2 FSAR JUN 29 1984 Rev. 24.2FIGURE 4.5-1 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 7 FULL POWER YEARS MPS-2 FSAR Rev. 24.2JUN 10 1982FIGURE 4.5-2 REACTOR COOLANT SYSTEM PRESSURE - TEMP LIMITATIONS DURING PLANT HEATUP/COOLDOWN AFTER 7 YEARS INTEGRATED NEUTRON FLUX MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.5-3 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 2 YEARS OF FULL POWER MPS-2 FSAR Rev. 24.2FIGURE 4.5-4 REACTOR CO OLANT SYSTEM HEATUP LIMITATIONS FOR 54 EFPY MPS-2 FSAR Rev. 24.2FIGURE 4.5-5 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS FOR 54 EFPY MPS2 UFSAR4.6-1Rev. 35

4.6 TESTS

AND INSPECTIONS

4.6.1 GENERAL

Shop inspection and tests of all major components are performed at the vendors' plants prior to shipment. An inspection at the site is performed to assure that no damage ha s occurred in transit. Testing of the reactor c oolant system (RCS) are pe rformed at the site upon completion of the plant construction. These tests will include hydrostatic tests of all fluid systems.

A complete visual inspection of all welds and joints are performed prior to the instal lation of the insulation. All field welds are radiographically and dye penetrant insp ected in accordance wi th the requirements of Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.In addition to the code-required examinations on the replacement reactor vessel head, base line examinations were performed prior to the com ponent being place in serv ice. These baseline examinations include a full volumetric examination of the CEDM nozzle base material, a minimum of 2 inches from the high point of the J-groove weld down to the threaded portion.

Additionally eddy current examin ation was performed on the wett ed surface of at least 28 peripheral CEDM nozzles in the bimetallic weld area.

A hot flow test of the reactor coolant loop up to zero power ope rating pressure and temperature without the core installed will be made. The system will be checked for vi bration and cleanliness.

Auxiliary systems will be checked for performance (see Chapter 13).

4.6.2 NIL DUCTILITY TRANSITION REFERENCE TEMPERATURE The carbon and low alloy steels whic h form the reactor coolant pr essure boundary are required to satisfy the requirements of 10 CFR 50, Appendix G which utilizes the nil ductility transition reference temperature (RTNDT) as the basis for establishing operational limitations for the reactor coolant system.

The original materials associated with the reactor coolant pressure boundary were ordered and tested to the requirements of ASME Code,Section III, Paragraph N-330. Th e impact properties of these materials were required to meet the acceptance criteria noted in Paragraph N-330 at +40

°F. These tests determined the nil ductility transition temperature (NDTT) for these materials. As the original design requirements were insufficient to directly establish the initial RTNDT of the pressure boundary materials and comply with the requirements of 10 CFR 50, Appendix G, the RTNDT was either estimated using the procedures of MTEB 5-2, performance of supplemental testing of surplus material in accordance with NB-2331 of Section III to the ASME Code or developed from generic impact testing.In the case of the reactor vessel, the beltline mate rials will experience a sh ift (an increase) in nil ductility transition reference temperature due to neutron irradiation. This increase is calculated in accordance with regulatory positions and is described in Section 4.5.1.3.

MPS2 UFSAR4.6-2Rev. 35To assure that the change in the fracture toughness prope rties behave in the expected fashion, a reactor vessel material surveillance program is conducted. A description of the surveillance program is provide d in Section 4.6.3.

The design material toughness test requirements were as follows:Reactor Vessel Carbon and low alloy steel materials which form a part of the pressure boundary shall meet the requirements of the ASME C ode,Section III, Para graph N-330 at a temperature of +40

°F. It shall be an objective that the materials meet this requirement at -10

°F. Charpy tests sh all be performed and the results used to plot a transition curve of impact values vs. temperature extending from fully brittle to fully ductile behavior. The actual NDT temperat ure of inlet and outlet nozzles, vessel and head flanges, and shell and lower head materials shall be de termined by drop weight tests per ASTM E-208.Replacement Reactor Vessel Closure Head The impact properties of carbon and low alloy steel materials including weld filler metals for the replacement head shall meet the requirements of ASME Section III, Subsecti on NB 2331, 1998 Edition through 2000 Addenda. Charpy V-notch transition curves were established in accordance with SA 508 supplementary requirements (S3) for temperatures showing, upper shelf energy, lower shelf energy and transition. The actual TNDT shall be determined by drop weight test in accordance with ASTM E208. Reference temperature, RTNDT shall be determined in accordance with NB 2331.Steam Generator and Pressurizer It shall be an objective that impact properties of all ferritic steel materials which form a part of the pressure boundary shall meet the requirements of the ASME code Section III, at a temperature of +10°F; alternate higher temperature levels up to 40

°F may be used only if the material fails at +10°F. Such higher temperature leve ls, if applicable, shall be de termined and documented. This objective is applicable to the pre ssurizer and the original steam generators, of which the original steam drums are still in use. For the steam generator replaceme nt subassemblies, the maximum allowable RTNDT as defined in paragraph NB-2311(a) of ASME Section III Code is 0

°F. For the replacement pressurizer ferritic steel materials, the RTNDT shall be performed at a temperature of 10 degrees F or less. The actual TNDT is determined by drop weight test in accordance with NB 2331 to ASTM-208-91.

Reactor Coolant Piping Materials used to fabricate the pi pe fittings shall be specified, examined and tested to satisfy as a minimum the requirements of Chapter I-III of the American Nu clear Society (ANS) Code for Pressure Piping B31.7, Class 1.

MPS2 UFSAR4.6-3Rev. 35 Impact properties of carbon steel materials, including welds, shall have a minimum V-notch value of 20 ft-lb (average of three specimens) or 15 ft-lb (a ny individual specimen) at 40

°F. It shall be a design objective that the material s meet this re quirement at 10

°F. Weld procedure qualifications and weld metal certifications records may serv e to demonstrate impact properties of welds.

The initial toughness test data for these components are provided in Tables 4.6-1, 4.6-2, 4.6-3 and 4.6-4. Table 4.6-1 also provides RTNDT values.Toughness test data are summed as follows: a.The maximum NDT temperature for the reactor vessel as obtained from drop weight tests is +10

°F. Drop weight tests were conducted only for material used in the reactor vessel.b.The maximum temperature correspondi ng to the 50 ft-lb value of the C v fracture energy for the reactor vessel is +65

°F.Refer to the tables presented above the Charpy V-notch data at 10

°F for the steam generators, pressurizer, and reactor coolant piping.c.The minimum upper-shelf C v energy value for the strong direction of the material used in the reactor vessel is 103 ft-lb.

The upper shelf-C v energy was not determined for the material used in fabricating the steam generator, pressurizers, or reactor coolant piping. The data was not obtained for the weak direction in the material used to fabricate the reactor vessels, thus branch technical position MTEB5.2 is used to establish transverse (WR) properties.

Due to regulatory changes during th e construction of the f acility , it was necessary to establish the RTNDT of the reactor coolan t pressure boundary materials to co mply with the requirements of 10 CFR 50 Appendix G. In most instances, this has been accomplished by utilizing the guidance of MTEB 5.2 to estimate the RTNDT of the material based upon th e available data. An initial RTNDT of 50°F has been established for the reacto r coolant pressure boundary materials (excluding the reactor vessel beltline) based on MTEB 5-2. However, in the case of the reactor vessel beltline materials, the RTNDT was not determined through a combination of methods including utilizing the guidance of MTEB 5-2, test ing surplus materials to the requirements of NB-2300, and utilizing generic data. The initial RTNDT values for the beltline materials are provided in Table 4.6-12. In additi on, the primary side of the stea m generator has been replaced and the materials RTNDT values have been determined from testing in accordance with NB-2300.

MPS2 UFSAR4.6-4Rev. 35

4.6.3 SURVEILLANCE

PROGRAMThe surveillance program is implemented to monitor the radiation-i nduced changes in the mechanical and impact propertie s of the pressure vessel mate rials in accordance with the requirements of 10 CFR 50, Appendix H. Changes in the impact properties of the material are evaluated by the comparison of pre-irradiation and post-irradiation Charpy impact test specimens.

Changes in mechanical properties are evaluated by the compar ison of pre-irradiation and post-irradiation data from tensile test specimens. Three metallurgically different materials represen tative of the pressure vessel are investigated. These are base metal, weld metal, and weld HAZ material. In addition to the materials from the reactor vessel, materials from a standard heat of A533B, ma de available through the Heavy Section Steel Technology (HSST)

Program, are also used. This reference material is fully processed and heat treated and is used for Charpy impact specimens so th at a comparison may be made between the irradiations in various operating power reactors a nd in experimental reactors. A complete record of the chemical analysis, fabr ication history and mechan ical properties of all surveillance test materials is maintained.

The exposure locations and a summary of the sp ecimens at each locat ion is presented in Table 4.6-8. The pre-irradiation NDT temperature of each plate in the intermediate and lower vessel shell courses is determin ed from the drop weight tests a nd correlated with Charpy impact tests.Base metal test specimens are fabri cated from sections of the shell plate in either the intermediate or the lower shell course whic h exhibits the highest unirradiated NDT temperat ure. All base material test specimens are cut from the same shell plate. This material is heat treated to a condition of the base metal in the completed reactor vessel.

Weld metal and HAZ material are produced by we lding together two plat e sections from the intermediate or lower shell course of the reactor vessel. All HAZ test materials are also fabricated from the plate which exhibits the highest unirradiated NDT temperature.The material used for weld metal and HAZ test specimens was adjacent to the test material used for ASME Code,Section III tests and was at least one plate thickness from any water-quenched edge. The procedures used for ma king the shell girth welds in the reactor vessel was followed in the preparation of the weld meta l and HAZ test materials. The procedures for inspection of the reactor vessel welds was followed for inspection of the welds in the test materials. The welded plate was heat treated to a conditi on which is representative of th e final heat treat ed condition of the completed reactor vessel.

Additional information from the baseline su rveillance program in cludes the chemical composition of the surveillance pl ate and weldment (made from two separate heats of weld wire), which are given in Table 4.6-5. The baseline mech anical properties of the base metal (WR and RW), weld metal, heat affected zone (HAZ), and standard referen ce material (SRM), are shown in Tables 4.6-6 and 4.6-7. The Charpy V-notch impact energy and lateral expansion data as a function of test temperature ar e shown in Figures 4.6-5 through 4.6-14.

MPS2 UFSAR4.6-5Rev. 35 The test specimens are contained in six irradiation capsule assemb lies. The axial position of the capsules bisected by the midplate of the core. The circumferential locations include the peak flux regions. The reactor vessel surv eillance program was designed in accordance with ASTM E185 (no edition specified). The program complies with ASTM E185-73 and 10 CFR 50, Appendix H.

The location of the surveillance capsule assemblies is shown in Figure 4.6-1. A typical surveillance capsule assembly is shown in Figure 4.6-2. A typi cal Charpy impact compartment assembly is shown in Figure 4.6-3. A typical tensil e monitor compartment as sembly is shown in Figure 4.6-4.

Fission threshold detectors (U-238) were inserted into each surveillance capsule to measure the fact neutron flux. Threshold detectors of Ni, Ti, Fe , S, and Cu with known Co content have been selected for this application to monitor the fast neutron exposure. Cobalt is included to monitor the thermal neutron exposure.

The selection of threshold det ectors is based on the recomm endations of ASTM E-261. "Method for Measuring Neutron Flux by Radioactive Tec hniques." Activation of th e specimen material will also be analyzed to determine the amount of exposure.

The maximum temperature of the encapsulated specimens will be monitored by including in the surveillance capsules small pieces of low-me lting-point eutectic al loys or pure metals individually sealed in quartz tubes.

The temperature monitors will provide an indi cation of the highest temperature to which the surveillance specimens were expos ed but not the time-temperatu re history or the variance between the time-temperature history of different specimens. These factors, however, will affect the accuracy of the estimated vessel material NDT temperature to only a small extent.

Test specimens removed from the surveillance capsules are tested in accordance with ASTM Standard Test Methods for Tension and Impact Testing. The data obtained from testing the irradiated specimens will be compared with the unirradia ted data and an assessment of the neutron embrittlement of the pressu re vessel material will then be made. This a ssessment of the NDT temperature shift is based on the temperature shift in the average Charpy curves, the average curves being considered representative of the material.The periodic analysis of the surveillance samples permit the monitoring of the neutron radiation effects upon the vessel materials.

If, with due allowance for uncer tainties in NDT temperature determination, the measured NDT te mperature shift turns out to be greater than predicted, then appropriate limitations would be imposed on permissible operating pressure-temperature combinations and transients to ensure that the existing reactor vessel stresses are low enough to preclude brittle fracture failure.The original six surveillance capsules were insert ed into their designated holders during the final reactor assembly operation. Each capsule remains in the reactor for the tentative schedule listed in Table 4.6-9. Table 4.6-9 shows the target fluence for each of the capsules.

MPS2 UFSAR4.6-6Rev. 35The fluence values in Table 4.6-9 are accurate within +30 percent.

4.6.4 NONDESTRUCTIVE

TESTS Prior to and during fabrication of the reactor vessel, nondestructiv e tests based upon Section III of the ASME Boiler and Pressure Vessel Code were performed on all welds, forgings and plates as follows: All full penetration pressure retaining welds were 100 percent radiographed to the standards of paragraph N-624 of Section III of the ASME Boiler and Pressure Vessel Code. Other pressure retaining welds such as thos e used for the attachment of mechanism housings, vents and instrument housings or the repla cement reactor vessel head and J groove welds were inspected to additional inspection criteria listed in Table 4.6-11.All forgings were inspected by ul trasonic testing, using longitudina l beam techniques. In addition, ring forgings were tested using shear wave techniques. Rejection under longitudinal beam inspection, with calibration so that the first back re flection is at least 75 pe rcent of screen height, was based on interpretation of indications causing complete lo ss-of-back reflection. Rejection under shear wave inspection was ba sed on indications, exceeding th e amplitude of the indication from a calibration notch whose depth is three percent of the forging thickness not exceeding 2Ú8 inch with a length of 1 inch.All forgings were also subjected to magnetic-particle examinat ion or liquid-penetrant testing depending upon the material. Rejection was based on Section III of the ASME Code, paragraph N626.3 for magnetic-particle and paragraph N627.3 for dye penetrant testing.

Plates were ultrasonically tested using longitudinal ultrasoni c testing techniques. Rejection under longitudinal beam testing performed in accordance with ASME Code, with calibration so that the first back reflection is at least 50 percent of scr een height, was based on defects causing complete loss of back reflection. Any defect which showed a total loss of back reflection which could not be contained within a circle whose diameter is the greater of three inches or one-half the plate thickness was unacceptable. Two or more defects smaller than described above which caused a complete loss-of-back reflection were unacceptable unless separated by a minimum distance equal to the greatest diameter of the larger defect unless the defects were c ontained within the area described above.

Nondestructive testing of the vessel was performed during several stages of fabrication with strict quality control in critical areas such as frequent calibration of test instruments, metallurgical inspection of all weld rod a nd wire, and strict adherence to the nondestructive testing requirements of Section III of the ASME Boiler and Pressure Vessel Code.

The detection of flaws in irregu lar geometries was facilitated be cause most nondestructive testing of the materials was completed while the material was in its simplest form. Nondestructive inspection during fabrication was sc heduled so that full penetrati on welds were capable of being radiographed to the extent required by Section III of the ASME Boiler and Pressure Vessel Code.

MPS2 UFSAR4.6-7Rev. 35 The new replacement reactor vessel closure head is a single piece forging. The head was replaced during refueling outage 16. Prior to and during fabrication of the replacement reac tor vessel head, nondestructive tests based upon Section III of the ASME Boiler and Pressure Vessel Code 1998 Edition through 2000 Addenda were performed on the forging and welds.

Ultrasonic and magnetic particle examinations were performed on the head forging in accordance with ASME Section III, NB 2000 and NB 5000 respectively. H ead cladding was ultrasonically examined for both bond and defects using a calibration block typi cal of the cladding and base material. Any indications that produce amplitude e qual to or greater than the amplitude received from the three-eighths inch flat bottom hole, regardless of length, were unacceptable (see Table 4.6-11 for additional owne rs inspection requirements).

Pressure retaining welds were 100% radiographed and liquid penetr ant examined in accordance with NB 5000. Canopy seal welds were liquid pene trant examined. J-groove welds were liquid penetrant examined at half thic kness and again at fina l surface. No indicati ons were allowed for the final PT of the J welds on the CEDM and instrument and ve nt nozzles and attachments of mechanism housings. Hydrostatic tests at the shop were conducted to 3107 psig. Visual examinations, magnetic particle a nd liquid penetrant tests were pe rformed to reveal any surface discontinuities.

In addition to the pre-service examinations required by Section XI of the ASME Boiler and Pressure Vessel Code, augmented ultrasonic and eddy current exa mination were performed on the inconel nozzle bore material to meet the ex amination requirements of NRC Bulletin 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles", August 3, 2001 and NRC Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity" March 18, 2002.Each of the vessel studs received one ultrasonic test and one magnetic-pa rticle inspection during the manufacturing process.

The ultrasonic test was a radial longitudinal beam inspection, and a discontinuity which causes an indication with a height which exceeded 20 percent of the height of the adjusted first back reflection was cause for re jection. Any discont inuity which prevents the production of a first back reflection of 50 percent of the screen height was also cause for rejection.The magnetic-particle inspection was performed on the finished studs. Linear axially aligned defects whose lengths are grea ter than one inch long and linear nonaxi al defects were unacceptable.

The vessel studs are elongated by the stud tensioners during each installation of the vessel head.

The amount of elongation for the de sired preload was specified by the vendor to fall within a predetermined acceptance range. Maintenance procedures for vessel head installation are in full compliance with the vendor specified range.The replacement pressurizer assembly consists of upper and lower heads and two shells attached by circumferential seam welds.

The shells and the heads are forged components. The large bore MPS2 UFSAR4.6-8Rev. 35nozzles (safety, relief, spray and surge) are inte gral with the upper and lo wer head eliminating the nozzle to shell welds. The attachment weld be tween the lower head prolongation and the support skirt is greater than one "t" li mit of the ASME Boiler and Pressure Vessel Code Section XI, Subsection IWB jurisdictional boundary. Ther efore ASME Section XI, Subsection IWF jurisdiction applies to the attachment weld.Non destructive examinations were performed on forgings and welds in accordance with ASME Section III, Boiler and Pressure Vessel Code 1998 Edition thr ough 2000 Addenda. Ultrasonic and magnetic particle examinations were performed on the forged components in accordance with ASME Section III, NB 2000 and 5000. The heater sleeves, vent and instrument nozzles are J-groove weld attachments to the reinforced weld buildup of the replacement pressurizer. The partial penetration welds were examined by li quid penetrant tests in accordance with NB-5245. The electrical heater sheaths were completely liquid penetrant inspected in accordance with NB-5000. In addition to the code require ments, additional inspection criteria listed in Table 4.6-11 for the pressurizer were required for acceptability.

Hydrostatic test of the replacement pressurizer was conducted at 3125 psia.

Prior to and during fabrication of the components of the RCS, nondestructive testing based upon the requirements of Section III of the ASME Boiler and Pressure Ve ssel Code is used to determine the acceptance criteria for various size flaws. The requirements for the Class A vessels are the same as the reactor vessel. Vessels designated as Class C were fabricated to the standards of Subsection C, Article 21 of Section III of the ASME Code.Table 4.6-10 summarizes the components inspecti on program during fabric ation and construction.

In addition to the inspections conducted during fabricat ion and construction of the pressurizer shown in Table 4.6-10, pre-service inspections were performed on several components of the replacement pressurizer. UT insp ections were performed on the circumferential welds of the pressurizer shell and heads. Magnetic particle in spection of the skirt to vessel was performed. UT and PT inspections were performed on the safety, relief, spray and surge nozzles.

Periodic tests and examinations of the RCS ar e conducted after startup on a regular basis.

For preoperational and in-service structure surveillance of the RCS, refer to the Technical Specifications; tests for RCS in tegrity after the system is closed following normal opening, modification or repair are specified in Technical Specifications.

4.6.5 ADDITIONAL

TESTS During design and fabrication of the reactor vessel, additional operations beyond the requirements of the ASME Boiler and Pressure Vessel Code,Section III were performed by the vendor. Table 4.6-11 summarizes the additional tests by components.

During the design of the reactor vessel, detailed cal culations were performed to assure that the final product would have adequate design margins. A detailed fatigue analysis of the vessel for all MPS2 UFSAR4.6-9Rev. 35 design conditions has been performe

d. In those areas which are not amenable to calculation, stress concentrations have been obtained through the use of photo-elastic models. In addition, Combustion Engineering (CE) has performed test programs for the determination and verification of analytical solutions to ther mal stress problems. Also fracture mechanics and brittle fracture evaluations have been performed.The thermal and structural analyses of CEDM and ICI head adapters of the replacement reactor vessel closure head were perfor med using finite element models. Structural responses from loads including branch line pipe brea k, faulted conditions (combination of line break and seismic), normal operating mechanical a nd acoustic excitation were evaluated. ASME Appendix G evaluations were performed assuming meridional and circumferential flaws at critical locations of the reactor vessel closure head. Detailed fatigue analysis using finite element models of bimetallic welds at the interface between alloy 690 and stai nless steel for the ICI nozzles were performed.All material used in the reactor vessel was car efully selected and preca ution were taken by the vessel fabricator to ensure that all material specificat ions were adhered to. To assure compliance, the quality control staff of CE reviewed the mill test reports and the fabricator's testing procedures.

All welding methods, materials, t echniques, and inspections comply with Sections III and IX of the ASME Boiler and Pressure Vessel Code. Before fabricati on was begun, detailed qualified welding procedures, including met hods of joint preparation, togeth er with certified procedure qualification test reports, were prepared. Also, prior to fabr ication, certified performance qualification tests were obtained for each welder and welding operator. Quality control was exercised for all welder s and welds by subjection to a comp lete and thorough testing program in order to ensure maximum quality of welded joints.During the manufacture of the reac tor vessel, in addition to the ar eas covered by the ASME Boiler and Pressure Vessel Code,Section III, quality control by the vendor included: a.preparation of detailed purchase specif ications which included cooling rates for test samp les;b.requiring vacuum degassing for al l ferritic plates and for gings;c.specification of fabrication instructions for plates and for gings to provide control of material prior to rece ipt and during fabrication;d.use of written instructions and manufact uring procedures which enable continual review based on past and curr ent manufacturing experiences;e.performance of chemical analysis of welding electrodes, welding wire, and materials for automatic welding, ther eby providing continuous control over welding materials; MPS2 UFSAR4.6-10Rev. 35f.the determination of NDT temperature th rough use of drop wei ght testing methods as well as Charpy impact tests;g.and test programs on fabri cation of plates up to 15 inches thick to provide information about material pr operties as thickness increases.RW wave ultrasonic testing was performe d on 100 percent of all plate material.

Cladding for the reactor vessel is a continuous in tegral surface of corros ion-resistant material, 5Ú16 inch nominal thickness. The detailed procedure used, i.e., type of we ld rod, welding position, speed of welding, nondestructive testing requirements, etc., was in compliance with the ASME Boiler and Pressure Vessel C ode. The cladding is ultrasonicall y inspected for lack of bond at intervals not to exceed 12 inches WR to the dire ction of weldi ng. Unbonded areas equal to or in excess of calibration requ ire additional scanning of the surr ounding material until the boundary of the discontinuity is established. An area of unbounded clad in excess of acceptance standards is repaired.Upon completion of all postweld he at treatments, the reactor vess el was hydrostatically tested, after which all weld surfaces, incl uding those of welds used to repair material, were magnetic-particle inspected in accordance with Section III, paragraph N-618 of the ASME Boiler and Pressure Vessel Code.

Surveillance of the quality c ontrol program was also carried out during the manufacture of the vessel by the Windsor Quality C ontrol Section of CE and by Northeast Nuclear Energy Company (NNECO) with an independent consultant. This work incl uded independent review of radiographs, magnetic-partic le tests, ultrasonic te sts, and dye penetrant tests conducted during the manufacture of the vessel. A review of material cert ifications and vendor manufacturing and testing procedures was also c onducted. Manufacturers' re cords such as heat-t reat logs, personnel qualification files and deviation files were also included in this review.

The nominal cladding thickness of the replacement reactor ve ssel closure head following machining (i.e., grinding) is quarter inch.

The cladding was ultrasonically examined such that each pass of the scanner overlap a minimum 10 percent of the transducer dimension perpendi cular to the direction of the scan. Magnetic particle and liquid penetrant test s were conducted following hydrosta tic test of the reactor vessel head to find any surface discontinuities. The ma gnetic particle and liqui d penetrant tests were conducted in accordance with AS ME Section III subsection NB 5000.

The replacement head was fabr icated and inspected by Mitsubi shi Heavy Industries at their facility. Representatives from Dominion Nuclear, CT (DNC) were present to witness the weld inspections and hydrostatic test s at various hold points. DNC also re viewed material certifications, test results conducted during fabrication.The replacement pressurizer has a minimum stainless steel cladding of 1/5 inch. The cladding was ultrasonically examined such th at each pass of the scanner overl ap a minimum 10 percent of the MPS2 UFSAR4.6-11Rev. 35 transducer dimension perpendicular to the direction of the sc an. Magnetic particle and liquid penetrant tests were conducted on 100% of the accessible surface to detect surface discontinuities were conducted in accordance with ASME Section III subsec tion NB 5000. In addition to the weld inspections, DNC personnel witnessed the hydrosta tic test at various hold points at the fabricator's shop for the replacement pressurizer.

4.6.6 IN-SERVICE INSPECTION A preoperational inspection was pe rformed in compliance with Sect ion XI of the ASME Boiler and Pressure Vessel Code, In-Service Inspect ion of Nuclear Reactor Coolant Systems, 1971, where possible. CE and Southwest Research Institute were reta ined to assist in the development of the in-service inspection pr ogram. Access was provide d, where possible, to permit inspection of the areas listed in the code. In-service inspections are performed in accordance with Section XI. A large portion of the insulation for the reactor vessel has been placed on the reactor cavity wall to permit inspection of the vessel outer surface. Essentia lly all vessel internals can be removed so that a complete visual internal inspection is po ssible and a volumetric internal inspection of the vessel is also possible. Pads have been welded to the outer surface of the reactor vessel to facilitate prompt location of welds for inspection purposes.

Access openings are provided on the permanent reactor cavity seal and the neutron shielding around the reactor vessel to facil itate inspection and maintenance of the neutron detector wells during refueling.

Biological shielding around the primar y piping in the area of the reactor pressure vessel has been designed to afford access to the circumferential and RW welds, as well as the transition piece-to-nozzle welds.

All primary piping, as well as major components, excluding the reactor pressure vessel, have been provided with easily removable insulation in the areas of all welds and adjacent base metal requiring examination as defined by Section XI.

Removable blanket insula tion is provided on the CEDM, vent and instrumentation nozzles areas of the replacement reactor vessel closure head.Plant arrangement and piping has been designed to assure that adequate access exists for either direct personnel access or for remote handling equipment to perf orm the examinations required by Section XI. Service connections, e.g., air, water, electricity, have been located adjacent or in close proximity to each inspecti on area. Consideration has been gi ven to the type of examination and equipment requirements.

Access holes have been pr ovided in the support skirt of each steam generator to provide a means of examining the tube sheet support stay cylinder weld.

MPS2 UFSARMPS2 UFSAR4.6-12Rev. 35TABLE 4.6-1 RTNDT DETERMINATION FOR REACTOR VESSEL BA SE METAL MILLSTONE UNIT NUMBER 2 Piece Number (Reference Drawing E-233-426-1)CODE Number HEAT Number VESSEL LOCATION DROP WEIGHT NDTT T50 (°F)T35 (°F)T CV (°F)RTNDT (°F)UPPER SHELF C V ENERGY FOR LONGITUDINAL DIRECTION203-02C-5004P2989 5P3286-4310V1Vessel Flange+10F5507010149204-02C-511C-5823-3ABottom Head Dome-7061688828115204-03AC-510-1C-5823-3BBottom Head Peel-60-20-218-42150 204-03CC-510-2C-5892-3A-10-22-200-60135 204-03BC-510-3C-5892-3B-40202040-20140 204-03DC-510-1C-5823-3B-6020-218-42150 204-03EC-510-2C-5892-3A-10-22-200-60135 204-03FC-510-3C-5892-3B-40202040-20140 205-02CC-503-19-7395-1-1Inlet Nozzles-204510655111 205-02BC-503-29-7401-1-2+1030125010120 205-02AC-503-39-7454-1-3-20121232-28108 205-02DC-503-49-7458-1-4-60302050-10123 205-03AC-508-3AV2999-9G-1283Inlet Nozzle Extension-20F30250-10146205-03BC-508-1AV2999-9G-1281-30250-10 205-03CC-508-2AV2999-9G-128230250-10 205-03DC-508-4AV2999-9G-126230250-10 205-06AC-502-19-7356-001Outlet Nozzles-110-20-300-60132 205-06BC-502-29-7375-002-13016-1036-24103 MPS2 UFSARMPS2 UFSAR4.6-13Rev. 35 T50 (°F) corresponds to the temperature at which 50 ft-lb energy is absorbed.

T35 (°F) corresponds to the temperature at which 35 mils lateral expansion is exhibited.

T CV (°F) reflects the greater of T50 and T35 and incorporates guidance from MTEB 5-2 where necessary based upon available data.205-07AC-509-1AV3816-9G-1239Outlet Nozzle Exten.-20251545-15141205-07BC-509-2AV3816-9G-1391251545-15 215-01CC-504-1C-5804-2Upper Shell+1062538222118 215-01AC-504-2C-5809-2+1062488222131 215-01BC-504-3C-5809-1-1055527515125 215-02AC-505-1C-5843-1Intermediate Shell-20---8.1118 215-02BC-505-2C-5843-2-10---17.5123 215-02CC-505-3C-5843-3-10---5115 215-03AC-506-1C-5667-1 1/4TLower Shell+10---7112 215-03CC-506-2C-5667-2 1/4TLower Shell-40F----33.7135215-03BC-506-3A-5518-1 1/4T-30----19.2136 A103W62-1-1Replacement Closure Head 44- 70-- 40148 B103W62-1-1Replacement Closure Head 18- 22-- 40139TABLE 4.6-1 RTNDT DETERMINATION FOR REACTOR VESSEL BA SE METAL MILLSTONE UNIT NUMBER 2 Piece Number (Reference Drawing E-233-426-1)CODE Number HEAT Number VESSEL LOCATION DROP WEIGHT NDTT T50 (°F)T35 (°F)T CV (°F)RTNDT (°F)UPPER SHELF C V ENERGY FOR LONGITUDINAL DIRECTION MPS2 UFSARMPS2 UFSAR4.6-14Rev. 35TABLE 4.6-2 CHARPY V-NOTCH AND DROP WEIGHT TEST VALUES - PRESSURIZER MILLSTONE UNIT NUMBER 2 Description of Part Heat NumberDrop Wt/RTNDT °FCharpy V-notch Test Temp °FCharpy V-notch Test OrientationImpact Energy (ft-lbs)

Lateral Expansion (mils) Upper HeadT4987-8+520

° 147-147-147 90-87-87 180° 160-172-157 87-94-90 Upper ShellT5086+1+61 0° 142-162-155 90-98-94 180° 144-128-125 90-87-90 Lower ShellT5085-8+52 0° 137-140-125 94-90-87 180° 127-111-113 87-79-79 Lower HeadT4986-17+43 0° 177-169-153 94-90-87 180° 169-169-138 87-90-87Support Skirt4-1575-+10Base Ring179-165-177 87-83-87 Skirt A134-131-134 79-78-78 Skirt B131-171-131 76-89-76Manway / Vent Covers11746-+59Vent116-119-115 79-83-83Manway119-159-127 85-94-89 Lifting Lug11764-+59 Lug 1130-167-133 91-94-93 Lug 2136-144-174 93-94-94Manway StudN9952-+10-37.6-37.6-36.929.3-28.9-28.1 Manway Nut 81025-+10-50.1-48-7-49.441.0-39.5-42.2Ventport StudN9879-+10-36.9-36.1-36.127.0-26.6-26.2Ventport Nut 81025-+10-50.1-48.7-49.441.0-39-5-42.2 MPS2 UFSARMPS2 UFSAR4.6-15Rev. 35TABLE 4.6-3 CHARPY V-NOTCH AND DROP WEIGHT TEST VALUES - STEAM GENERATOR PART NumberHEATLOCATION/MATERIALSteam Generator Unit NumberRT NDT (°F) (1)CHARPY V-NOTCH IMPACT TEST RESULTS (2)CVN TEST TEMP °FNOTES5056021W70029-1Tubesheet / SA508 Class 32-70126.9, 118.8, 104.4146.5, 125.0, 94.0-10

°F5056021W70022-11-60130.9, 146.5, 152.1144.6, 134.9, 130.90

°5056020-288D108-1-1Primary Head /

SA508, Class 32124.6, 147.5, 142.3 (La)136.4, 142.3, 132.7 (La)40388C115195.4, 190.3, 147.5 (La)40-30114.3, 95.8, 112 (ME)104, 120, 119 (Me)300173, 181, 151 (ME)605056020-188D101-1-11129, 111, 136 (La)122, 125, 123 (La)40388C110157, 131, 123 (La)40-3090, 98, 98 (ME)30-4098.8, 90, 107 (ME)200142, 134, 164 (ME)605056024-1727957Stay Cylinder /

SA508, Class 3198.1, 95.8, 97.3112, 98.1, 102+40120.2, 113, 116126, 90.7, 101.7+40-3599.5, 112, 11095.8, 72.2, 116+25

-35129, 135.7, 127.6123.9, 128.3, 131+25 MPS2 UFSARMPS2 UFSAR4.6-16Rev. 355056024-27279572-27106.9, 118, 112.8104.7, 109.1, 113.6+33-35104.7, 104.7, 109.1+25-27115, 118, 120.9+33106, 110, 12099.5, 104.7, 118+40122.4, 107.6, 112.8126, 102, 109+405056025-1724015Safe End-Inlet /

SA508, Class 11-8°F130.5, 127.6, 118 (TANG)+404-8°F99.6, 80.4, 110.6

+525056025-27240152-8°F103.2, 118, 109.1 (TANG)+40-8°F103.3, 89.9, 97.36

+52 5056026-1, 2, 3, 4724015Safe End-Outlet /

SA508, Class 11 & 2-8°F115.8, 105.5, 109.1 (TANG)99.5, 95.88, 88.5+404

+525069670-287633-2Manway1-22°F109, 121, 124 (Top)

+37.4 5069670-SA533, Grade B, Class 1-49 108, 108, 91 (Bottom)+105069670-587659-2Manway SA 533, Grade B, Class 11 + 2 -49107.7, 97.5, 115 (Top)+10TABLE 4.6-3 CHARPY V-NOTCH AND DROP WEIGHT TEST VALUES - STEAM GENERATOR (CONTINUED)PART NumberHEATLOCATION/MATERIALSteam Generator Unit NumberRT NDT (°F) (1)CHARPY V-NOTCH IMPACT TEST RESULTS (2)CVN TEST TEMP °FNOTES MPS2 UFSARMPS2 UFSAR4.6-17Rev. 355069679-187659-2-40137, 125.5, 119 (Bottom)+205069670-387660-22-40143, 103, 107 (Top)+10-31106, 165, 140 (Bottom)+305062982-1, 237314Vessel Support A533, Grade B, Class 11 + 2-40181, 178, 140 (Top)+20(66961)-31184, 200, 187 (Bottom)+30 5062980-1, 2, 3, 4, 5, 6374101 + 2-22177, 190, 151 (Top)+40(66922)-22178, 160, 182 (Bottom)+405056024-1727957Inlet Nozzle /

SA508, Class 31124, 113, 115123, 129, 125+40-4571.5, 67.8, 81.873, 84.8, 70-155056024-27279572134, 152, 136178, 205, 179+40-6258, 89, 8788, 84, 60-35056023-1727957Outlet Nozzle /

SA508, Class 31110, 111, 118+40TABLE 4.6-3 CHARPY V-NOTCH AND DROP WEIGHT TEST VALUES - STEAM GENERATOR (CONTINUED)PART NumberHEATLOCATION/MATERIALSteam Generator Unit NumberRT NDT (°F) (1)CHARPY V-NOTCH IMPACT TEST RESULTS (2)CVN TEST TEMP °FNOTES MPS2 UFSARMPS2 UFSAR4.6-18Rev. 35 Notes:(1) RTNDT determined from drop WEIGHT and Charpy V-notch test results.

(2) Charpy-v-notch impact test results are listed in sets of three as tested. Specimen location and orientations were varied an d removed for testing in accordance with ASME Section III requirements. Sp ecimen location and orientat ion have been listed where possible.(3) Specimen location identified as La and Me were orientated in longitudinal and meridional direction, respectively.(4) Specimen location identified as TANG were oriented in tangential direction.-3659, 74.5, 73+425056023-27279572116, 125, 116+40-1893, 112, 82.6+425056023-37279571118, 119, 135+40-4575, 67, 86+155056023-47279572112, 122, 118+40-3581.1, 85.56, 76.7+25TABLE 4.6-3 CHARPY V-NOTCH AND DROP WEIGHT TEST VALUES - STEAM GENERATOR (CONTINUED)PART NumberHEATLOCATION/MATERIALSteam Generator Unit NumberRT NDT (°F) (1)CHARPY V-NOTCH IMPACT TEST RESULTS (2)CVN TEST TEMP °FNOTES MPS2 UFSARMPS2 UFSAR4.6-19Rev. 35TABLE 4.6-4 CHARPY V-NOTCH VALUES - PIPING Assembly Number PIECE NumberCODE NumberDESCRIPTIONMATERIALCHARPY V-NOTCH VALUES - FT-LBTemperature

°F503-01502-20-1C-4601-1Pipe SegmentSA-516 Grade 70505349+10Hot Leg502-20-2C-4601-2Pipe SegmentGrade 70373438+10502-02-1C-4605-1Ell SegmentGrade 70554161+10502-02-2C-4605-1Ell SegmentGrade 70554161+10506-02C-4613-1Nozzle Forging A 105 Grade 243+10512-03C-4611Nozzle ForgingGrade 2433632+10503-02502-02-3C-4605Ell SegmentSA 516 Grade 70554161+10Hot Leg502-02-4C-4605Ell SegmentGrade 70554161+10502-20-3C-4601-3Pipe SegmentGrade 70414743+10502-20-4C-4601-4Pipe SegmentGrade 70454751+10506-08C-4612-1Nozzle ForgingA 105 Grade 2243028+10504-01502-12-1C-4604-1Pipe SegmentSA 516 Grade 70586163+10 MPS2 UFSARMPS2 UFSAR4.6-20Rev. 35 Cold Leg Pump Discharge502-12-2C-4604-2Pipe SegmentGrade 70363636+10502-08-1C-4608Ell SegmentGrade 70343039+10 502-08-2C-4608Ell SegmentGrade 70343039+10507-02-1C-4615-1Nozzle Forging SA 105 Grade 2433632+10508-02-1C-4610-1Nozzle Forging SA 182 Grade F19895104+10504-03502-10-9C-4609-2Ell SegmentSA 516 Grade 70644166+10 Cold Leg Pump Discharge 502-10-10C-4609-2Ell SegmentGrade 70644166+10502-18-1C-4602-1Pipe SegmentGrade 70586762+10502-18-2C-4602-2Pipe SegmentGrade 70586762+10508-02-2C-4610-2Nozzle Forging SA 182 Grade F19594110+10504-04502-12-3C-4604-3Pipe SegmentSA 516 Grade 70596660+10TABLE 4.6-4 CHARPY V-NOTCH VALUES - PIPING Assembly Number PIECE NumberCODE NumberDESCRIPTIONMATERIALCHARPY V-NOTCH VALUES - FT-LBTemperature

°F MPS2 UFSARMPS2 UFSAR4.6-21Rev. 35 Cold Leg Pump Discharge502-12-4C-4604-4Pipe SegmentGrade 70596660+10502-08-3C-4608Ell SegmentGrade 70343039+10 502-08-4C-4608Ell SegmentGrade 70343039+10507-02-2C-4615-2Nozzle Forging SA 105 Grade 2433632+10508-02-3C-4610-3Nozzle Forging SA 182 Grade F1106120114+10507-07-2C-4616-2RNozzle Forging SA 105 Grade 2433632+10504-05502 11C-4609-2Ell SegmentSA 516 Grade 70644166+10 Cold Leg Pump Discharge 502-10-12C-4609-2Ell SegmentGrade 70644166+10502-18-3C-4602-3Pipe SegmentGrade 70586157+10502-18-4C-4602-4Pipe SegmentGrade 70586157+10508-02-4C-4610-4Nozzle Forging SA 182 Grade F11097692+10507-07-1C-4616-1Nozzle Forging SA 105 Grade 2433632+10503-05-1502-04-1C-4606-1Ell SegmentSA 516 Grade 70585150+10 Cold Leg Pump Suction 502-04-2C-4606-1Ell SegmentGrade 70585150+10TABLE 4.6-4 CHARPY V-NOTCH VALUES - PIPING Assembly Number PIECE NumberCODE NumberDESCRIPTIONMATERIALCHARPY V-NOTCH VALUES - FT-LBTemperature

°F MPS2 UFSARMPS2 UFSAR4.6-22Rev. 35503-05-2502-04-3C-4606-1Ell SegmentSA 516 Grade 70585150+10 Cold Leg Pump Suction502-04-4C-4606-1Ell SegmentGrade 70585150+10503-05-3502-04-5C-4606-2Ell SegmentSA 516 Grade 70444953+10 Cold Leg Pump Suction502-04-6C-4606-2Ell SegmentGrade 70444953+10503-03-1502-06-1C-4607Ell SegmentSA 516 Grade 70475552+10503-05-4502-04-7C-4606-2Ell SegmentSA 516 Grade 70444953+10 Cold Leg Pump Suction502-04-8C-4606-2Ell SegmentGrade 70444953+10 Cold Leg Pump Suction502-06-2C-4607Ell SegmentGrade 70475552+10502-16-1C-4603-1Pipe SegmentSA 516637064+10502-16-2C-4603-2Pipe SegmentGrade 70637064+10503-03-2502-06-3C-4607Ell SegmentSA 516 Grade 70475552+10 Cold Leg Pump Suction502-06-4C-4607Ell SegmentGrade 70475552+10502-16-1C-4603-1Pipe SegmentGrade 70637064+10502-16-2C-4603-2Pipe SegmentGrade 70637064+10TABLE 4.6-4 CHARPY V-NOTCH VALUES - PIPING Assembly Number PIECE NumberCODE NumberDESCRIPTIONMATERIALCHARPY V-NOTCH VALUES - FT-LBTemperature

°F MPS2 UFSARMPS2 UFSAR4.6-23Rev. 35503-03-3502-06-5C-4607Ell SegmentSA 516 Grade 70475552+10 Cold Leg Pump Suction502-06-6C-4607Ell SegmentGrade 70475552+10502-16-1C-4603-1Pipe SegmentGrade 70637064+10502-16-2C-4603-2Pipe SegmentGrade 70637064+10503-03-4502-06-7C-4607Ell SegmentSA 516 Grade 70475552+10 Cold Leg Pump Suction502-06-8C-4607Ell SegmentGrade 70475552+10502-16-1C-4603-1Pipe SegmentGrade 70637064+10502-16-2C-4603-2Pipe SegmentGrade 70637064503-07-1502-14-1C-4604-5Pipe SegmentSA 516 Grade 70232120+10 Cold Leg Pump Suction502-14-2C-4604-6Pipe SegmentGrade 70525754+10502-10-1C-4609-1Pipe SegmentGrade 70504335+10 502-10-2C-4609-1Pipe SegmentGrade 70504335+10507-10-4C-4614-4RNozzle Forging SA 105 Grade 2384243+10TABLE 4.6-4 CHARPY V-NOTCH VALUES - PIPING Assembly Number PIECE NumberCODE NumberDESCRIPTIONMATERIALCHARPY V-NOTCH VALUES - FT-LBTemperature

°F MPS2 UFSARMPS2 UFSAR4.6-24Rev. 35503-07-2502-14-1C-4604-5Pipe SegmentSA 516 Grade 70232120+10 Cold Leg Pump Suction502-14-2C-4604-6Pipe SegmentGrade 70525754+10502-10-3C-4609-1Ell SegmentGrade 70504335+10502-10-4C-4609-1Ell SegmentGrade 70504335+10 507-10-1C-4614-3RNozzle Forging SA 105 Grade 2384243+10505-10-1502-14-3C-4604-7Pipe SegmentSA 516 Grade 70525754+10 Cold Leg Pump Suction502-14-4C-4604-8Pipe SegmentGrade 70586163+10502-10-5C-4609-1Ell SegmentGrade 70504335+10 502-10-6C-4609-1Ell SegmentGrade 70504335+10507-10-2C-4614-2Nozzle Forging SA 105 Grade 2433632+10505-10-2502-14-3C-4604-7Pipe SegmentSA 516 Grade 70525754+10 Cold Leg Pump Suction502-14-4C-4604-8Pipe SegmentGrade 70586163+10502-10-7C-4609-2Ell SegmentGrade 70644166+10502-10-8C-4609-2Ell SegmentGrade 70644166+10507-10-3C-4614-1Nozzle Forging SA 105 Grade 2433632+10TABLE 4.6-4 CHARPY V-NOTCH VALUES - PIPING Assembly Number PIECE NumberCODE NumberDESCRIPTIONMATERIALCHARPY V-NOTCH VALUES - FT-LBTemperature

°F MPS2 UFSAR4.6-25Rev. 35TABLE 4.6-5 PLATE AND WELD METAL CHEMICAL ANALYSIS Element Weight Percent Plate C-506-11/4 T-ID Weld C-506-2/C-506-31/4 T-OD Weld C-506-2/C-506-3Si0.120.170.15S0.0140.0130.013 P0.0060.0150.016Mn1.261.131.13C0.210.120.12 Cr0.100.040.05Ni0.610.060.06Mo0.620.540.53 V0.0040.0060.007Cb< 0.01< 0.01< 0.01B0.00060.00030.0003 Co0.0110.0090.009Cu0.140.300.21Al0.020< 0.001< 0.01 W< 0.010.01< 0.01Ti< 0.01< 0.01< 0.01As0.0110.0110.012 Sn0.0090.0040.003Zr0.0020.0020.002 N 2 0.009 0.008 0.009 MPS2 UFSARMPS2 UFSAR4.6-26Rev. 35TABLE 4.6-6 BELTLINE MECHANICAL TEST PROPERTIES - REACTOR VESSEL SU RVEILLANCE MATERIALSMaterial Code NumberOrientation aa.With respect to the plates' major rolling direction for base metal; with respect to the weldi ng direction for weld and HAZ.

NDTT (°F) 30 ft-lb Fix (°F)50 ft-lb Fix (°F)35 Mils Lat.

Exp. Fix (°F)RTNDT (°F)Upper Shelf (ft-lb)RT Yield (ksi)Base Metal PlateC-506-1Transverse (WR)-102266546106.567Base Metal PlateC-506-1Longitudinal (RW)-10368465 24 bb.Not valid per 10 CFR 50, Appendix G.124.567Weld MetalC-506-2 / C-506-3Transverse (WR)-60-2652-55129.576HAZ MetalC-506-1Transverse (WR)-30-443534-25123.068 MPS2 UFSARMPS2 UFSAR4.6-27Rev. 35TABLE 4.6-7 TENSILE TEST PROPERTIES - REACTOR VESSEL SURV EILLANCE MATERIALS Material Plate Code NumberOrientation aa.With respect to the plates' major rolling direction for base metal; with respect to the weld ing direction for weld and HAZ. Test Temp (F) Yield Strength (ksi)Tensile Strength (ksi) Elongation TE(%)/UE(%)

R.A. (%)Base Metal PlateC-506-1Transverse (WR) 67 8827/116825064 8125/106755058 8423/0962Base MetalC-506-1Longitudinal (RW)7167 8629/127125061 7926/107055056 8326/1069Weld MetalC-506-2/C-506-3Longitudinal (RW)7176 8628/117425074 8126/097155067 8525/1065HAZ MetalC-506-1Transverse (WR)7168 8824/097025060 8025/087255062 8321/0768 MPS2 UFSARMPS2 UFSAR4.6-28Rev. 35 (a) L = Longitudinal(b) T = Transverse (c) Reference material correlation monitorsTABLE 4.6-8

SUMMARY

OF SPECIMENS PR OVIDED FOR EACH EXPOSURE LOCATION Capsule Location on Vessel Wall Base Metal Impact L (a) Base Metal Impact T (b) Base Metal TensileWeld Metal ImpactWeld Metal Tensile HAZ Impac tHAZ TensileReference Impact (c)Total Impact Specimens Tensile 83°12123123123-489 97°12123123123-489 104°12-312312312489 263°12-312312312489 277°12123123123-489 284°12123123123-489724818721872182428854 MPS2 UFSAR4.6-29Rev. 35TABLE 4.6-9 CAPSULE REMOVAL SCHEDULECAPSULELOCATION LEAD FACTOR a a.Updated in Capsule W-83 dosimetry analysisREMOVAL TIME (EFPY) b b.Effective Full Power Years (EFPY) from plant startupFLUENCE (n/cm 2 , >1.0MeV) (a) W-9797°1.403.0 3.24 x 10 18 c c.Plant specific evaluationW-104104°0.9510.0 9.49 x 10 18 (c) W-97 d d.Flux Monitor 97°10.0--W-8383°1.3115.3 1.74 x 10 19 (c) W-277277°1.31 EOL e e.EOL is defined as the end-of-license peri od corresponding to the original 40 year license. Capsule W-277 is projected to receive 1.31 times the reactor vessel peak EOL surface fluence of 2.4 x 10 19 n/cm 2 (E > 1.0 MeV). Capsule W-277 will r eceive the vessel peak EOL surface fluence at 23.2 EFPY. It will be removed before it receives twice the peak vessel surface fluence of 4.80 x 10 19 n/cm 2 (E > 1.0 MeV).

See Note (d)W-263263°1.31Standby--W-284284°0.97Standby--

MPS2 UFSAR4.6-30Rev. 35TABLE 4.6-10 INSPECTION OF REACTOR COOLANT SYSTEM COMPONENTS DURING FABRICATION AND CONSTRUCTION1.Reactor Vessel Forgings Flange UT, MT Studs UT, MT Cladding UT, PT Nozzles UT, MT Plates UT, MT Cladding UT, PT Welds Main Seams RT, MT CRD Head Nozzle Connection UT, PT, ET Instrumentation and Vent Nozzles UT, PT Main Nozzles to Shell RT, MT Cladding UT, PT Nozzle Safe Ends RT, MT Vessel Support Buildup UT, MT All Welds - After Hydrostatic Test MT, PT 1 2.Steam Generator Tube Sheet Forging UT, MT Cladding UT, PT Primary Head Forging UT, MT Cladding UT, PT Secondary Shell and Head Plates UT, MT 1.Liquid penetrant tests of J-welds only.

MPS2 UFSAR4.6-31Rev. 35Tubes UT, ET Nozzles (Forgings) UT, MT Studs MT Welds Shell, Longitudinal RT, MT Shell, Circumferential RT, MT Cladding UT, PT Nozzles to Shell RT, MT Tube-to-Tube Sheet PT Instrument Connections MT Temporary Attachments After Removal MT All Welds - After Hydrostatic Test MT Nozzle Safe Ends RT, (MT or PT)

Level Nozzles MT 3.Replacement Pressurizer Heads, Forging RT, MT Cladding UT, PT, MT Shell, Forging RT, MT Cladding UT, PT, MT Heaters, Sheath UT, PT Nozzles (Integral to Heads) UT, MT Studs (Manway, Ventport) UT, PT Welds Shell to Shell and Shell to Heads Circumferential RT, MT, PT, UT Cladding UT, PT, MT Nozzle Safe Ends RT, PT Instrument Connections PT Support Skirt RT, PT, UT MPS2 UFSAR4.6-32Rev. 35Temporary Attachments After Removal MT Heads to Shell and Shell to Shell Welds PT Heater Assembly RT, PT 4.Pumps Castings RT, PT Forgings UT, PT Welds Circumferential RT, PT Instrument Connections PT All Welds After Hydrostatic Test PT 5.Piping Fittings RT, PT Pipe RT, PT Nozzles RT, PT Welds UT, PT Circumferential RT, PT, MT Nozzles to Run Pipe RT, PT Instrument Connections PT Cladding UT, PT Legend: RT -Radiographic PT - Dye Penetrant ET - Eddy Current

UT - Ultrasonic MT - Magnetic Particle MPS2 UFSAR4.6-33Rev. 35TABLE 4.6-11 REACTOR COOLANT SYSTEM INSPECTION C-E REQUIREMENTSReactor Vessel

<test>C-E RequirementsCode RequirementsUltrasonic Testing (UT)1. UT of Weld Clad for bond.1. NoneReplacement Reactor Vessel Head (a)<test>C-E RequirementsCode RequirementsUltrasonic Testing (UT))1. I ndications that produce an amplitude greater than the amplitude received from the

one-eighth inch flat bottom hole and less than the amplitude from the three-eighth inch flat

bottom hole are to be characterized as to size, length, and depth below the surface.

Indications exceeding the following are unacceptable.ASME Section V, Article 23, SA

578Depth from Clad SurfaceLength of Indication Up to 0.02 inches 0.375 inches0.02 inches - 0.06 inches1.0 inches0.06 inches - 0.1 inches3.0 inches

Over 0.10 inches

6.0 inches

MPS2 UFSAR4.6-34Rev. 35 (a) This is a Dominion requirement and not a CE requirement. The reactor vessel head and replacement pressurizer were re placed in accordance with Do minion Purchase Specifications.Replacement Reactor Vessel Head (continued)

<test>C-E RequirementsCode RequirementsLiquid Penetrant test (PT)1.

The final PT of J-weld and the half-inch area adjacent to the weld for CEDM adapters, instrument tube connections

shall allow no indications of defects.ASME Section III, NB 5245Steam Generator

<test>C-E RequirementsCode RequirementsUltrasonic Test (UT)1. UT for Defects in Tube Sheet Clad UT of Weld Clad for bond.

1. None2. UT of Weld Clad for bond

2. NoneReplacement Pressurizer (a)<test>C-E RequirementsCode RequirementsUltrasonic Testing (UT)1. Indications that produce an amplitude greater than the ampl itude received from the one-eighth inch flat bottom hole and less than the amplitude from the three-eighth inch flat bottom hole are to be characterized as to their length, and depth below the surface. Indications exceeding

the following are unacceptable.ASME Section V, Article

23, SA 578Depth from Clad SurfaceLength of Indication Up to 0.03 inches 0.375 inches0.03 inch - 0.06 inch1.0 inches0.06 inch - 0.125 inch3.0 inches Over 0.125 inch

6.0 inches

MPS2 UFSARMPS2 UFSAR4.6-35Rev. 35TABLE 4.6-12 RTPTS VALUES AT 54 EFPYVessel Location Component Identification Chemical Content: Cu (%)Chemical Content: Ni

(%)Chemist ry FactorInitial RTNDT (°F)Margin Term (°F)Neutron Fluence (10 19 n/cm 2)RT PTS (°F)Intermediate C-505-10.130.6191.38.1343.83165.1 ShellC-505-20.130.6291.517.5343.83174.7 Course Plates C-505-30.130.6291.55343.83162.2 LowerC-506-10.150.60110.07343.78188.8 ShellC-506-20.150.61110.0-33.7343.78163.5 Course Plates C-506-30.140.66101.5-19.2343.78183.6Intermediate Shell Axial Welds2-203 A (Heat A8746)0.150.1377.7-56662.83114.6 2-203 B/C (Heat A8746)0.150.1377.7-56663.53114.6Lower Shell Axial Welds3-203 A (Heat A8746)0.150.1377.7-56663.83114.6 3-203 B/C (Heat

A8746)0.150.1377.7-56662.50114.6Intermediate-to- Lower Shell Girth Weld9-203 (Heat 10137)0.220.04100.0-56.3563.78134.19-203 (Heat 90136)0.270.07124.3-56.3563.78166.7 MPS2 UFSARMPS2 UFSAR4.6-36Rev. 35 (a) Maximum neutron fluence for vessel location.(b) Adjusted reference temperature projec tions based on best-estimate copper and nickel content, chemistry factor, initial RTNDT and margin given in Table 4.6-12.TABLE 4.6-13 ADJUSTED REFERENCE TEMPERATURES (ART) PROJECTIONSVessel LocationComponent IdentificationNeutron Fluence (a), 10 19 n/cm 2 (E>1 MeV) 54 EFPYART Projections (b) 54 EFPY1/4T3/4T1/4T3/4TIntermediateC-505-12.2830.811153.8128.0ShellC-505-22.2830.811163.4137.6Course PlatesC-505-32.2830.811150.9125.1LowerC-506-12.2530.809175.2144.1ShellC-506-22.2530.809134.5103.4 Course PlatesC-506-32.2530.809138.6110.0Intermediate Shell Axial Welds2-203 A (Heat A8746)2.2833.811104.682.62-203 B (Heat A8746)1.5080.53696.073.72-203 C (Heat A8746)1.5080.53696.073.7Lower Shell Axial Welds3-203 A (Heat A8746)2.2830.811104.682.63-203 B (Heat A8746)1.4900.52995.873.4 3-203 C (Heat A8746)1.4900.52995.873.4Intermediate to Lower Shell Girth Weld9-203 (Heat 10137)2.2530.800121.793.59-203 (Heat 90136)2.2530.800151.3116.2 MPS-2 FSAR JUN 29 1984 Rev. 24.2FIGURE 4.6-1 LOCATION OF SURVEILLANCE CAPSULE ASSEMBLIES MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-2 TYPICAL SURV EILLANCE CAPSULE ASSEMBLY MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-3 TYPICAL CHARPY IMPACT COMPARTMENT ASSEMBLY MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-4 TYPICAL TENSILE-MONITOR COMPARTMENT ASSEMBLY MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-5 BASE METAL - WR (TRANSVERSE) PLATE C-506-1 IMPACT ENERGY VS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-6 BASE METAL - WR (TRANSVERSE) PLATE C-506-1 LATERAL EXPANSION VERSUS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-7 BASE METAL - RW (LONGITUDINAL) PLATE C-506-1 IMPACT ENERGY VERSUS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-8 BASE METAL - RW (LOGITUDINAL) PLATE C-506-1 LATERAL EXPANSION VS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-9 WELD METAL PLATE C-506-2/C-506-3 IMPACT ENERGY VS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-10 WELD METAL, PLATE C-506-2/C-506-3 LATERAL EXPANSION VS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-11 HAZ METAL, PLATE C-506-1 IMPACT ENERGY VERSUS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-12 HAZ METAL, PL ATE C-506-1 LATERAL EXPANSION VERSUS TEMPERATURE MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4.6-13 SRM (HSST PLATE 01MY - LONGITUDINAL) IMPACT ENERGY VERSUS TEMPERATURE MPS-2 FSAR Rev. 24.2JUN 10 1982 FIGURE 4.6-14 SRM (HSST PLATE 01MY - LO NGITUDINAL) LATERA L EXPANSION VS TEMPERATURE MPS2 UFSAR4.A-1Rev. 35 4.A SEISMIC ANALYSIS OF REACTOR COOLANT SYSTEM 4.A.1 INTRODUCTION The purpose of this appendix is to describe the methods employed and present the results obtained by dynamic seismic analyses of th e reactor coolant sy stem components. Th ese analyses were performed to confirm the adequ acy of the seismic loadings specified for the design of the components and the supports of the reactor coolant system which in cludes the reactor vessel, the steam generators, the reactor coolant pumps, the pressurizer, and the interconnecting reactor coolant piping.

Dynamic seismic analysis of the reactor vessel internals was performed separately and is discussed in Appendix 3.A.

4.A.2 METHOD OF ANALYSIS 4.A.2.1 General The seismic analysis of the reac tor coolant system (RCS) compone nts was performed using either normal mode or direct in tegration theory in conjunction with time history and response spectrum techniques, as appropriate.

Time history techniques were em ployed in the analysis of the reactor vessel, the two steam generators, the four reactor coolant pumps and the interconnecti ng reactor coolant piping. In the analysis of these components, a single composite math ematical model, whic h included integral representations of each of the components and connecting piping, was employed to account for the interacting effects of dynamic coupling. The analysis of these dynamically coupled multi-supported components utilized diff erent time dependent input exc itations applied simultaneously to each support.

The analyses of the pressurizer and the surge line pipi ng employed separate, uncoupled, mathematical models and utilized response spectrum techniques.

The input data, time histories a nd response spectra, app lied in the analyses were provided by the analysis of the containment structure internal support structure described in Section 5.8.

The RCS components were analyzed using either modal or proportional (R ayleigh) methods of damping. Except for analysis of the surge line piping, all modal anal yses used a constant damping factor of 1% of critical damping for all active modes. In the analysis of the surge line piping, a damping factor of 0.5 %

of critical damping was used for each mode. When proportional damping was used, Alpha and Beta were conservatively se lected to provide less than 1% of critical damping at the significant frequencie s of response of the major components.

MPS2 UFSAR4.A-2Rev. 35 4.A.2.2 Mathematical Models In the descriptions of the mathem atical models which follow , the spatial orientations are defined by the set of orthogonal axes where Y is in the vert ical direction, and X and Z are in the horizontal plane, in the directions indicated on the appropriate figure. The mathematical representation of the section properties of the stru ctural elements employs a 12 by 12 stiffness matrix for the three-dimensional space frame models, and employs a 6 x 6 stiffness matrix for the two dimensional plane frame model. Elbow s in piping runs include the in-pla ne/out-of-plane bending flexibility factors as specified in th e ASME Code,Section III.

4.A.2.2.1 Reactor Coolant System - Coupled ComponentsA schematic diagram of th e co mposite mathematical model, designated MS-2, used in the original plant design analyses of the dynamically coupled components of the reactor coolant system is presented in Figure 4.A-1 and 4.A-1A. This model, Figure 4.A-1, includes 19 mass points with a total 47 dynamic degree s of freedom. The mass points a nd corresponding dynamic degrees of freedom are distributed to provide appropriate representations of the dynamic characteristic of the components, as follows: the reactor vessel, with in ternals, is represented by 5 mass points with a total of 13 dynamic degrees of freedom; each of the two steam generators are represented by 3 mass points with a total of 7 dynamic degrees of freedom; and each of the four reactor coolant pumps are represented by 2 mass points with a total of 5 dyna mic degrees of freedom. The relatively small mass of the inte rconnecting reactor coolant piping is lumped proportionately with the masses of the adjoining components.

The mathematical model, Figure 4.A-1, as defined, provides a complete three dimensional representation of the dynamic response of the coupl ed components to seismi c excitations in both the horizontal and vertical direct ions. The mass is distributed at the selected mass points and corresponding translational degrees of freedom are retained to include rotary inertial effects of the components. The total mass of the entire coupled system is dynamical ly active in each of the three coordinate directions.

In addition to Model MS-2 de scribed above, a second model of the coupled components, designated Model RV14, was formul ated for the original plant design to incorporate a more detailed representation of the reactor vessel assembly. With the ex ception of the representation of the reactor vessel assembly, Model RV14 is identical to Model MS-2, Figure 4.A-1. A schematic diagram of the representation of the reactor vessel assembly incorporated into Model RV14 is presented in Figure 4.A-2. This more detailed representation of the reactor vessel assembly

<component> Number of Mass PointsNumber of Dynamic Degrees of FreedomReactor Vessel and Internals513Steam Generators (2)614 R.C. Pumps (4)820Total1947 MPS2 UFSAR4.A-3Rev. 35 consists of 15 mass points with a total of 33 d ynamic degrees of freedom and includes a 10 mass point, 22 dynamic degrees of free dom representation of the re actor vessel internals. The representation of the reactor vessel internals was formulated in c onjunction with the analysis of the reactor vessel internals di scussed in Appendix 3.A, and was designed to simulate the dynamic characteristics of the models used in that analysis. The coupled Model RV14 was used to generate time histories of absolute accelerat ions at the reactor vessel flange used as forcing functions in the analysis of the reactor vessel internals.Coupled model RV14, modified to delete the reactor internals representation of the thermal shield and to reflect current design steam generator properties, was also used in analyses to determine the effects of the replacemen t steam generators on RCS responses. This modified model representation consists of 27 mass point s with 63 dynamic degrees of freedom.

4.A.2.2.2 Pressurizer The mathematical model employed in the analysis of the pressuri zer is shown schematically in Figure 4.A-3. This lumped parameter, planer m odel provides a multi-mass representation of the axially symmetric pressurizer a nd includes 5 mass points with a total of 6 dynamic degrees of freedom.The replacement pressurizer was analyzed using BWSPAN for both operating basis and safe shutdown earthquake using the re sponse spectrum method. The seismi c excitation is applied at the support skirt elevation. The signi ficant change to the analysis from the original pressurizer analysis is that the analysis is performed at 65% water level. Th e seismic directi onal responses are combined by the absolute sum of each horizontal a nd vertical response of the spectrum analysis.

4.A.2.2.3 Surge LineThe lumped parameter, multi-mass mathematical model employed in the analysis of the surge line is shown schematically in Figure 4.A-4. The surge line is modeled as a three dimensional piping run with end points anchored at the attachments to the pressurizer and the reactor vessel outlet piping. In the definition of the mathematical model, 10 mass points with a total of 27 dynamic degrees of freedom were selected to provide a complete three-di mensional representation of the dynamic response of the surge line.

All supports and restraints defined for the surge line assembly

<component> Number of Mass PointsNumber of Dynamic Degrees of FreedomReactor Vessel511Reactor Internals818 Steam Generators(2)614 Reactor Coolant Pumps (4)820Total2763 MPS2 UFSAR4.A-4Rev. 35 are included in the mathematical model. The total mass of the surge line is dynamically active in each of the three c oordinate directions.

4.A.2.3 Calculations 4.A.2.3.1 General As applied in the analysis, the s imultaneous equations of motion for linear structural systems with viscous damping can be wr itten, Reference 4.A-1:

where: M = diagonal matrix of lumped masses C = square symmetric damping matrix K = square symmetric stiffness matrix whic h defines the mass point force-displacement relationship.

= column matrix with elemen ts equal to the absolute acceleration of the datum support in the coordinate direction of the related dynamic degree of freedom of th e structural system.K ms = rectangular matrix of stiffness coeffici ents which defines th e mass point force, non-datum support displacement relationship.

X s = column matrix of displacements re lative to the datum at non-datum supports. X = column matrix of mass point displacements relative to the datum.

= column matrix of mass point velocities relative to the datum.

= column matrix of mass point accel erations relative to the datum.

In this form, the equations define the dynamic response of a multi-mass structural system subjected to time dependent s upport motion. In the analysis of systems with multiple supports, such as the coupled components of the reactor coolant system, the equations provide for different time-dependent input motions at each of the sup ports. In this case, one of the supports of the system is designated the reference, or datum, fr om which the motions of all other points of the structural system are measured. The reactor vessel support was de signated as the datum in the analyses of the coupled component s of the reactor coolant system.Normal mode theory, as descri bed in References 4.A-1 and 4.

A-2, was employed to reduce the equations of motion to a system of independent equations in terms of the normal modes for the modal superposition time-history and spectrum analyses of the reactor coolant system components. For direct integration time-history analyses, the equations of motion were solved MX**CX*KXMY**-K ms X s-=++Y**X*X**

MPS2 UFSAR4.A-5Rev. 35 using numerical integration me thods with an integration ti me step of 0.001 second. In the analyses, the dynamic response of the components was determined fo r seismic input excitations in each of the three global coordinate directions: X (east-west), Y (v ertical) and Z (north-south). The dynamic responses to vertical se ismic excitation were found for bot h the case of initial support displacement upward and the cas e of initial support displacem ent downward. These responses were combined to determine the most severe combinations produced by the effects of seismic excitations in each of the horizontal directions applied simultaneously with either seismic excitation in the vertical direction.

4.A.2.3.2 Frequency Analysis An eigenvalue analysis was performed utiliz ing the ICES STRUDL II computer code, Reference 4.A-3, to calculate the mode shapes and natural frequencies of th e original plant design composite mathematical models.

Modifications to the standard ICES STRUDL II program have been implemented by Combustion Engineering to include a Jacobi diagonalization procedure in the eigenvalue analysis and to provide appropriate influence coefficients and stiffness matrices for use in the response and reaction calculations.

The natural frequencies and dominant degrees of freedom calculated are shown in Table 4.A-1 for all modes used in the analysis of the reactor coolant system Model MS-2, the surge line and the pressurizer.

An eigenvalue analysis was also performed utilizing the ANSYS computer code, Reference 4.A-5, to calculate the natural frequencies and mode shapes of modified model RV14 which was used in the analysis to evaluate the effects of the replacement st eam generators. A comparison of response frequencies for corresponding mode shapes identified no significant differences from those in Table 4.A-1 (see also Section 4.A.5, Effects of Replacement Steam Generators).

4.A.2.3.3 Mass Point Response Analysis The original plant design time hi story mass point responses to se ismic excitation were computed using TMCALC, a C-E code. This code performs a numerical integration of the equations of motion for singly or multiply supported dynami c systems utilizing normal mode theory , Reference 4.A-2, and Newmark's Beta-Method with Beta equal to one-sixth, Reference 4.A-4.

For the multiply supported system, the separate time histories of each support were imposed on the system simultaneously. The results are time history responses of the ma ss points. The analysis of the reactor coolant system utilized m odal data for all frequencies through 40 cps.

The mass point responses result ing from the spectrum analysis were found utilizing SHAKE, a C-E computer code. This code pe rforms a normal mode response spectrum analysis resulting in the modal inertials loads found using the response sp ectrum for the pressurizer support. The mass point responses of the surge line were found using an enve lope of the support spectra of the interconnected major components.

MPS2 UFSAR4.A-6Rev. 35 The transient analysis capability of the ANSYS computer code, Reference 4.A-5, was used to compute the responses of the c oupled components of the reactor coolant system with existing replacement steam generators and with the reactor thermal shield removed.

4.A.2.3.4 Seismic Reaction Analysis The original plant design dynamically induced loads at all system design po ints due to the time history support excitations and mass point respo nses were calcu lated utilizing FORCE, a C-E computer code. This code performs a complete loads analysis of the defo rmed structure at each incremental time step by computing internal and external system reactions (forces and moments) by superposition of the reactions due to the mass point displ acements and the non-datum support displacements as follows:R(t) = C m X m (t)+C s X s (t) where:R(t) = the matrix of all components of the reactions at the system design points.

C m = the matrix of mass points displacement influence coefficients.

X m(t) = the column matrix of time history mass point displacements re lative to the datum at each time step.

C s = the matrix of support displacement influence coefficients.

X s (t) = the column matrix of time history su pport displacements relative to the datum at non-datum supports at each time step.

The support and mass point displace ments due to horizontal and vert ical seismic ex citations are added algebraically at each time step. The maximum component of each reaction for the entire time domain, and its associated t ime of occurrence, are selected.The maximum reactions for the pressurizer and surge line resulting from the response spectrum analysis were found by applying the modal inertial loads for each mode, to the structural model using the STRUDL computer code. The design point reactions due to each modal loading were conservatively combined by summing the absolute values of the modal reactions. For the replacement pressurizer, the design point reactions due to each modal loading were combined using square root sum of the squares. The surge line analysis included consideration of the relative end displacements. The reactions found by statically im posing the maximum relative displacements of the two ends of the surge line were cons ervatively included by absolute summation with the inertial res ponse from the spectrum analysis.The system design loads for the coupled component s of the reactor coolant system with existing replacement steam generators and with the reactor internals thermal shield removed were calculated using the ANSYS computer code.

MPS2 UFSAR4.A-7Rev. 35 4.A.3 RESULTS The reactions (forces and moment s) at all design points in the system, obtained from the dynamic seismic analysis, were compared with the seismic loads in ea ch component design specification.

The res ults of this comparison are summarized in Table 4.A-2 for the points of maximum calculated load. The maximum seismic loads calculated by the time history techniques are the result of a search and comparison over the entire time domain of each individual component of load due to the simultaneous application of the horizontal and either vertical ex citation. The maximum calculated components of load shown in Table 4.A-2 for each design location do not in general occur at the same time, nor for the same combination of horizontal and vertical excitation, and therefore result in a conservative case.

Except for the replacement pressurizer, the maximum seismic lo ads calculated by the response spectrum techniques are the result of combining the modal reactions due to the horizontal and the vertical excitation on an absolute sum basis.

The results shown are fo r the Operational Basis Earthquake. For conservati ve determination of results due to the Design Basis Earthquake, both the calculated re sults and specification values have been multiplied by a factor of 2.0, rath er than 0.17/0.09 = 1.89, the ratio of DBE to OBE maximum ground accelerations.

4.A.4 EFFECTS OF THERMAL SHIELD REMOVAL An engineering evaluation was made to assess the effects of the thermal shield removal on the dynamic response characteristics of the reactor vessel and the r eactor coolant system. The main effect of the thermal shield removal was a reduction in the weight of the reactor vessel. This reduction was approximately two percent (2%). Since the stiffness of the connection between the reactor vessel and internals did not change, it was concluded that the dynamic response characteristics of the reactor ve ssel and, therefore, the reactor c oolant system would not change significantly with the removal of the thermal shield. Reactor vessel fla nge motions, originally used for a more detailed evaluation of the internal structures in Appendix 3.A, also remain unchanged.

4.A.5 EFFECTS OF REPLACEMENT STEAM GENERATORSThe most significant effects of the replacement steam generators were increases in weights and center of gravity elevations of approximately 2.7% and 13.2 inches, respectively, for these components. Modified model RV 14, which includes these changes in addition to those due to removal of the reactor internal s thermal shield, was used in a reanalysis of the dynamically coupled components of the RCS. The changes in steam generator prope rties resulted in a reduction of approximately 4.5%

in the fundamental frequencies in which the steam generator response is predominant. Corresponding changes in loads which are attribut ed to the replacement steam generators are generally in significant with no design governing load increases of more than MPS2 UFSAR4.A-8Rev. 35 7%. The loads from this reanalysis, incorporating the effects of both thermal shield removal and steam generator replacement, are included in Table 4.A-2.

Reactor vessel flange motions were compared with those originally used in the detailed evaluation of the internal structures in Appendix 3.A. This comparison conc luded that use of the original flange motions is conser vative for reactor internals design.

4.A.6 CONCLUSIONIt is concluded that the seismic loadings specif ied for the design of th e reactor coolant system components and supports are ad equate. All seismic loads cal culated by the dynamic seismic analysis are less than the corresponding loads in the component design specification.

4.A.7 REFERENCES4.A-1Przemieniecki, J. S., "Theory of Matrix Structural Analysis," Chapter 13, McGraw-Hill Book Company, New York, New York, 1968. 4.A-2Hurty, W. C., and Rubinstein, M.

F., "Dynamics of Structures," Chapter 8, Prentice Hall, Inc., Englewood Cliffs, New Jersey, 1964. 4.A-3ICES STRUDL II Engineering User s Manual, R68-91, Department of Civil Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts. 4.A-4Newmark, N. M., "A Method of Computat ion for Structural Dynamics", Volume 3, Journal of Engineering Mechanics Division, A.S.C.E., July, 1959.4.A-5DeSalvo, G. P., and Swanson, J. A., "ANS YS - Engineering Analys is System," Swanson Analysis Systems, Inc., Elizabeth, PA., 1972.

MPS2 UFSAR4.A-9Rev. 35TABLE 4.A-1 NATURAL FREQUENCIES AND DOMINANT DEGREES OF FREEDOMMode NumberFrequency (cps)Dominant Degrees of FreedomNamesDirectionsLocations13.07M66ZPump 1B23.08M43ZPump 2B 33.21M61ZPump 1A 43.23M52ZPump 2A 53.32RI1ZReactor Internals 63.32RI1XReactor Internals 75.11M61X,YPump 1A 85.14M52X,YPump 2A 95.21M66X,YPump 1B105.23M43X,YPump 2B1110.51SG5A, SG5BXSteam Generators 1 & 21210.52SG5A, SG5BXSteam Generators 1 & 2 1310.60M61XPump 1A 1410.74M52XPump 2A 1510.98M66XPump 1B 1611.11M43XPump 2B 1712.14RI2ZReactor Internals 1812.16RI2XReactor Internals 1920.40SG9AZSteam Generator 1 2020.42SG9BZSteam Generator 2 2123.65RI1YReactor Internals 2227.26SG9, 10, A & BXSteam Generators 1 & 2 2327.79SG5A, SG5BYSteam Generators 1 & 2 2427.82SG5A, SG5BYSteam Generators 1 & 2 2528.99M65, M42ZPumps 1B & 2B 2629.91M65, M42ZPumps 1B & 2B2730.37M42ZPump 2B2831.43M60ZPump 1A 2931.92M51ZPump 2A MPS2 UFSAR4.A-10Rev. 353037.51V1, V3, V4X, ZReactor Vessel3139.34SG5BZSteam Generator 2 3239.41SG5AZSteam Generator 1113.78-X, ZReplacement Pressurizer 244.65-X, ZReplacement Pressurizer 348.99-XReplacement Pressurizer 462.35-X, ZReplacement Pressurizer 575.86-X, ZReplacement Pressurizer 675.97-YReplacement Pressurizer 799.45-X, ZReplacement Pressurizer 15.757, 8, H3YSurge Line 211.295, 7, H1XSurge Line 315.985, H1YSurge Line 421.798, 9, H3ZSurge Line 525.819, H3YSurge Line 632.128, 9XSurge Line 740.3510XSurge Line 873.645ZSurge Line 9108.523XSurge Line10129.1210YSurge Line11151.517, H3XSurge Line12155.037YSurge Line 13174.767, H1YSurge Line 14186.39H1XSurge Line 15229.9010, 11ZSurge Line 16260.3811XSurge Line 17304.863YSurge Line 18320.527XSurge Line 19503.60H1XSurge Line 20525.218YSurge Line 21535.878XSurge Line 22542.6411ZSurge LineTABLE 4.A-1 NATURAL FREQUENCIES AND DOMINANT DEGREES OF FREEDOMMode NumberFrequency (cps)Dominant Degrees of FreedomNamesDirectionsLocations MPS2 UFSAR4.A-11Rev. 3523668.064XSurge Line24752.114ZSurge Line 25938.6611YSurge Line 261327.418ZSurge Line 271807.564YSurge LineTABLE 4.A-1 NATURAL FREQUENCIES AND DOMINANT DEGREES OF FREEDOMMode NumberFrequency (cps)Dominant Degrees of FreedomNamesDirectionsLocations MPS2 UFSARMPS2 UFSAR4.A-12Rev. 35TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASISCombined North-South and Vertical Reactor Vessel Outlet NozzleFx107.069.0Fy3.0172.0Fz65.063.0 Mx1237.02168.0 My7329.07521.0 Mz382.038474.0

M R7443.039263.0Reactor Vessel Inlet Nozzle Fx 97.0 75.0 Fy 98.0 105.0 Fz 107.0 108.0 Mx7719.022392.0 My5402.010085.0 Mz16910.014087.0 M R19358.028312.0 Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-13Rev. 35Combined East-West and Vertical Reactor Vessel Outlet NozzleFx196.0241.0Fy39.0253.0Fz4.010.0 Mx67.01042.0 My360.0972.0 Mz3277.043773.0

M R3298.043796.0Reactor Vessel Inlet NozzleFx81.0220.0Fy97.0146.0 Fz110.063.0 Mx5963.09638.0 My3803.05981.0 Mz11763.012770.0

M R13726.017080.0TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-14Rev. 35Combined North-South and Vertical Steam Generator Inlet NozzleFs88.5144.0Fa89.0230.0Mb3901.74000.0 Mt2584.02640.0Steam Generator Outlet NozzlesFs134.8202.0Fa54.0104.0 Mb8990.013200.0 Mt10517.014800.0Combined East-West and Vertical Steam Generator Inlet NozzleFs97.1144.0Fa175.0230.0 Mb4239.04000.0 Mt113.02640.0Steam Generator Outlet NozzleFs126.1202.0Fa49.0104.0 Mb7681.013200.0 Mt11224.014800.0TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-15Rev. 35Combined North-South and Vertical Pressurizer Surge Line NozzleFx1.742.26Fy0.914.85Fz2.718.75

M R 192.7 690.9Combined East-West and Vertical Pressurizer Surge Line NozzleFx1.075.70Fy0.594.77 Fz1.222.71

M R 121.5 617.1TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-16Rev. 35 Combined North-South and Vertical Reactor Coolant Pump NozzleFx108.083.6Fy61.093.2Fz53.0114.1 Mx6289.012030.0 My5632.010188.0 Mz11867.010095.0

M R14564.018719.0Reactor Coolant Pump Outlet NozzleFx137.093.2Fy98.071.3 Fz43.097.8 Mx10457.022770.3 My2244.07514.4 Mz2244.08957.0

M R7797.025596.0TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-17Rev. 35Combined East-West and Vertical Reactor Coolant Pump Inlet NozzleFx119.0143.6Fy50.0104.5Fz35.033.9 Mx5291.08709.0 My4479.05238.0 Mz12263.016662.5

M R14087.019517.0Reactor Coolant Pump Outlet NozzleFx134.0263.8Fy97.0145.8 Fz37.0109.9 Mx5099.04072.0 My2964.013317.0 Mz6269.016648.0

M R8608.021796.0TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-18Rev. 35Combined North-South and Vertical Reactor Vessel Outlet PipingM R7443.011514.0Steam Generator Inlet PipingM R4680.04361.0 *

• Note: Piping design is controlled by combined East-West and vertical seismic excitation.Steam Generator Outlet PipingM R13836.013836.0 Pump Inlet Piping M R13148.013836.0Pump Outlet PipingM R13236.018528.0Reactor Vessel Inlet PipingM R18522.018528.0Combined East-West and Vertical Reactor Vessel Outlet Piping M R3298.011581.0Steam Generator Inlet PipingM R4241.09658.0Steam Generator Outlet PipingM R13601.013836.0 Pump Inlet Piping M R13632.013836.0Pump Outlet PipingM R8608.018528.0Reactor Vessel Inlet PipingM R13726.018528.0TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-19Rev. 35Combined North-South and Vertical Surge Line RCS NozzleFx4.416.0Fy1.454.0Fz3.254.0 M R 277.6 468.0Combined East-West and Vertical Surge Line RCS NozzleFx2.556.0Fy0.844.0 Fz1.314.0 M R 155.5 468.0Combined North-South and Vertical Surge Line Hanger H4Fx0.1500.18Fy0.4392.0Surge Line Hanger H2Fy0.2091.0Fz0.0780.11Combined East-West and Vertical Surge Line Hanger H4Fx0.0830.18Fy0.3352.0Surge Line Hanger H2Fy0.1451.0Fz0.0160.11TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-20Rev. 35Combined North-South and Vertical Reactor Vessel Outlet SupportFy165.0392.0Fz325.0663.0Reactor Vessel Inlet SupportFy314.0692.0 F H 191.0 304.0Steam Generator Lower SupportFy 219.0 431.0 Fz 195.0 397.0 Mx20345.024422.0 My2335.010332.0Steam Generator Upper SupportFx 20.0 24.0 Fz 219.0 240.0 Pump SupportFy 2.6 9.2Replacement Pressurizer SupportFy 14.1 80.0 a Fz80.1 84.0

• Mx15107.6 22768.9 17036.0

• TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS2 UFSARMPS2 UFSAR4.A-21Rev. 35Combined East-West and Vertical Reactor Outlet SupportFy309.0473.0Fz47.042.0Reactor Vessel Inlet SupportFy227.0469.0 F H 375.0 1020.0Steam Generator Lower SupportFy 278.0 624.0 Fz 32.0 29.0 My 356.0 455.0 Mz8792.024383.0Steam Generator Upper SupportFx 281.0 296.0 Fz 12.0 10.0 Reactor Coolant Pump SupportFy 1.6 4.3Replacement Pressurizer SupportFx 55.6 81.0

• Fy14.1 80.0

• Mz15746.9 16949.0

• a.Note that the design basis values are ba sed on original pressurizer design and are re tained here for historical purposes only. TABLE 4.A-2 SEISMIC LOADS ON REACTOR COOLANT SYSTEM COMPONENTS FOR OPERATIONAL BASIS EARTHQUAKE (CONTINUED)SEISMIC EXCITATIONCOMPONENT AND DESIGN LOCATIONSEISMIC LOADCOMPONEN TCALCULATE D MAXIMUM DESIGN BASIS Forces = Kips; Moments = Inch - Kips M R = [Mx 2 + My 2 + Mz 2]1/2 F H = [Fx 2 + Fz 2]1/2 MPS-2 FSAR Rev. 24.2JULY 1988FIGURE 4A-1 REACTOR COOLANT SYSTEM SEISMIC ANALYSIS MODEL MS2 MPS-2 FSARMAY 1994 Rev. 24.2FIGURE 4A-1A REACTOR COOLANT SYSTEM - SEISMIC ANALYSIS MODEL MS2 AND RV14 MPS-2 FSARJULY 1998 Rev. 24.2FIGURE4A-2 RV14 REACTOR AND IN TERNALS SEISMIC ANALYSIS MODEL MPS-2 FSARJULY 10 1982 Rev. 24.2FIGURE 4A-3 PRESSURIZER SEISMIC ANALYSIS MODEL MPS-2 FSAR JUN 10 1982 Rev. 24.2FIGURE 4A-4 SURGE LINE SEISMIC ANALYSIS MODEL