ML22193A073
Text
Millstone Power Station Unit 2 Safety Analysis Report Chapter 4: Reactor Coolant System
Table of Contents tion Title Page GENERAL SYSTEM DESCRIPTION ............................................................... 4.1-1 DESIGN BASIS .................................................................................................. 4.2-1 1 Design Parameters ...................................................................................... 4.2-1 2 Codes Adhered To ...................................................................................... 4.2-4 3 Quality Control Classification .................................................................... 4.2-6 4 Part-Loop Operation ................................................................................... 4.2-6 SYSTEM COMPONENT DESIGN .................................................................... 4.3-1 1 Reactor Vessel ............................................................................................ 4.3-1 2 Steam Generator ......................................................................................... 4.3-2 2.1 Flow Induced Vibration .............................................................................. 4.3-4 2.2 Tube Thinning............................................................................................. 4.3-5 2.3 Potential Effects of Tube Ruptures ............................................................. 4.3-5 2.4 Composition of Secondary Fluid ................................................................ 4.3-6 3 Reactor Coolant Pumps .............................................................................. 4.3-6 4 Reactor Coolant Piping ............................................................................. 4.3-13 5 Pressurizer................................................................................................. 4.3-13 6 Quench Tank............................................................................................. 4.3-16 7 Valves ....................................................................................................... 4.3-17 8 Instrumentation Application ..................................................................... 4.3-23 8.1 Temperature .............................................................................................. 4.3-23 8.1.1 Hot Leg Temperature................................................................................ 4.3-23 8.1.2 Cold Leg Temperature .............................................................................. 4.3-23 8.1.3 Surge Line Temperature ........................................................................... 4.3-24 8.1.4 Pressurizer Vapor Phase Temperature ...................................................... 4.3-24 8.1.5 Pressurizer Water Phase Temperature ...................................................... 4.3-24 8.1.6 Spray Line Temperature ........................................................................... 4.3-24 8.1.7 Relief and Safety Valve Discharge Temperature ..................................... 4.3-24 8.1.8 Quench Tank Temperatures...................................................................... 4.3-24 8.1.9 Reactor Vessel Flange Seal Leakage Temperature................................... 4.3-25
tion Title Page 8.1.10 RCS High Point Vents Leakage Temperature .......................................... 4.3-25 8.2 Pressure ..................................................................................................... 4.3-25 8.2.1 Pressurizer Pressure .................................................................................. 4.3-25 8.2.2 Pressurizer Pressure .................................................................................. 4.3-26 8.2.3 Pressurizer Pressure .................................................................................. 4.3-26 8.2.4 Quench Tank Pressure .............................................................................. 4.3-26 8.3 Level ......................................................................................................... 4.3-26 8.3.1 Pressurizer Level....................................................................................... 4.3-26 8.3.2 Pressurizer Level....................................................................................... 4.3-27 8.3.3 Quench Tank Level................................................................................... 4.3-27 8.4 Reactor Coolant Loop Flow...................................................................... 4.3-27 8.5 Reactor Coolant Pump Instrumentation.................................................... 4.3-27 8.5.1 Pump Seal Temperatures .......................................................................... 4.3-27 8.5.2 Motor Stator Temperatures ....................................................................... 4.3-28 8.5.3 Motor Thrust Bearing Temperatures ........................................................ 4.3-28 8.5.4 Pump Controlled Bleed-Off Temperature ................................................ 4.3-28 8.5.5 Antireverse Device Bearing Temperature ................................................ 4.3-28 8.5.6 Upper and Lower Guide Bearing Temperature ........................................ 4.3-28 8.5.7 Lube Oil Cooler Inlet and Outlet Temperature......................................... 4.3-28 8.5.8 Lower Bearing Oil Temperature............................................................... 4.3-29 8.5.9 Pump Seal Pressures ................................................................................. 4.3-29 8.5.10 Motor Oil Lift Pressure............................................................................. 4.3-29 8.5.11 Lube Oil Filter Pressure Differential ........................................................ 4.3-29 8.5.12 Pump Controlled Bleed-Off Flow............................................................. 4.3-29 8.5.13 Lube Oil and Antireverse Device Lube Oil Flow Switch......................... 4.3-29 8.5.14 Motor Oil Reservoir Level........................................................................ 4.3-29 8.5.15 Vibration Instrumentation......................................................................... 4.3-30 8.5.16 Reverse Rotation Switch........................................................................... 4.3-30 8.5.17 (Deleted) ................................................................................................... 4.3-30 8.5.18 RCP Underspeed Reactor Trip ................................................................. 4.3-30 9 Reactor Coolant Venting System.............................................................. 4.3-30
tion Title Page 10 Permanent Reactor Cavity Seal ................................................................ 4.3-32 MATERIALS COMPATIBILITY ...................................................................... 4.4-1 1 Materials Exposed to Coolant..................................................................... 4.4-1 2 Insulation .................................................................................................... 4.4-1 3 Coolant Chemistry ...................................................................................... 4.4-2 SYSTEM DESIGN EVALUATION ................................................................... 4.5-1 1 Prevention of Brittle Fracture ..................................................................... 4.5-1 1.1 Initial Nil-Ductility Transition Reference Temperature. ............................ 4.5-1 1.2 Nil-Ductility Transition Reference Temperature Shift ............................... 4.5-2 1.3 Operational Limits ...................................................................................... 4.5-3 1.4 Pressurized Thermal Shock ........................................................................ 4.5-5 2 Seismic Design ........................................................................................... 4.5-5 2.1 Piping .......................................................................................................... 4.5-6 2.2 Vessels ........................................................................................................ 4.5-6 2.3 Pumps and Valves....................................................................................... 4.5-7 3 Overpressure Protection.............................................................................. 4.5-8 3.1 Overpressure Protection During Normal Operation ................................... 4.5-8 3.2 Low Temperature Overpressurization Protection....................................... 4.5-8 4 Reactor Vessel Thermal Shock................................................................... 4.5-8 5 Leak Detection ............................................................................................ 4.5-9 6 Prevention of Stainless Steel Sensitization ............................................... 4.5-10 7 References................................................................................................. 4.5-14 TESTS AND INSPECTIONS ............................................................................. 4.6-1 1 General........................................................................................................ 4.6-1 2 NIL Ductility Transition Reference Temperature ...................................... 4.6-1 3 Surveillance Program.................................................................................. 4.6-4 4 Nondestructive Tests................................................................................... 4.6-6 5 Additional Tests .......................................................................................... 4.6-8 6 In-Service Inspection ................................................................................ 4.6-11
tion Title Page SEISMIC ANALYSIS OF REACTOR COOLANT SYSTEM ......................... 4.A-1
.1 Introduction................................................................................................ 4.A-1
.2 Method of Analysis.................................................................................... 4.A-1
.2.1 General....................................................................................................... 4.A-1
.2.2 Mathematical Models ................................................................................ 4.A-2
.2.2.1 Reactor Coolant System - Coupled Components ...................................... 4.A-2
.2.2.2 Pressurizer.................................................................................................. 4.A-3
.2.2.3 Surge Line.................................................................................................. 4.A-3
.2.3 Calculations ............................................................................................... 4.A-4
.2.3.1 General....................................................................................................... 4.A-4
.2.3.2 Frequency Analysis.................................................................................... 4.A-5
.2.3.3 Mass Point Response Analysis .................................................................. 4.A-5
.2.3.4 Seismic Reaction Analysis......................................................................... 4.A-6
.3 Results........................................................................................................ 4.A-7
.4 Effects of Thermal Shield Removal........................................................... 4.A-7
.5 Effects of Replacement Steam Generators ................................................ 4.A-7
.6 Conclusion ................................................................................................. 4.A-8
.7 References.................................................................................................. 4.A-8
List of Tables mber Title 1 Reactor Coolant System Volumes 1 Principal Design Parameters of Reactor Coolant System 2A Table of Loading Combinations and Primary Stress Limits 2B Table of Loading Combinations and Primary Stress Limits for the Replacement Reactor Vessel Head and Replacement pressurizer 3 Reactor Coolant System Code Requirements 4 Comparison with Safety Guide 26 1 Reactor Vessel Parameters 2 Steam Generator Parameters 3 Main Steam Safety Valve Parameters 4 Tech Pub Review PKG for T4.3-4Reactor Coolant Pump Parameters 5 Reactor Coolant Piping Parameters 6 Pressurizer Parameters 7 Quench Tank Parameter 8 Pressurizer Spray (RC-100E, RC-100F) Valve Parameters 9 Power-Operated Relief Valve Isolation Valve Parameters (RC-403, RC-405) 10 Pressurizer Power-Operated Relief Valve Parameters (RC-402, RC-404) 11 Pressurizer Safety Valve Parameters (RC-200, RC-201) 12 Active and Inactive Valves in the Reactor Coolant System Boundary 1 Materials Exposed to Coolant 2 Reactor Coolant Chemistry 1 Reactor Coolant System Component Nozzles, Nozzle Sizes and Nozzle Materials 2 Reactor Coolant System Heatup and Cooldown Limits 1 RTNDT Determination for Reactor Vessel Base Metal Millstone Unit Number 2 2 Charpy V-Notch and Drop Weight Test Values - Pressurizer Millstone Unit Number 2 3 Charpy V-notch and Drop Weight Test Values - Steam Generator
mber Title 4 Charpy V-notch Values - Piping 5 Plate and Weld Metal Chemical Analysis 6 Beltline Mechanical Test Properties - Reactor Vessel Surveillance Materials 7 Tensile Test Properties - Reactor Vessel Surveillance Materials 8 Summary of Specimens Provided for Each Exposure Location 9 Capsule Removal Schedule 10 Inspection of Reactor Coolant System Components During Fabrication and Construction 11 Reactor Coolant System Inspection C-E Requirements 12 Reactor Coolant System Inspection C-E Requirements (Continued) 13 RTPTS Values at 54 EFPY 14 Adjusted Reference Temperatures (ART) Projections
-1 Natural Frequencies and Dominant Degrees of Freedom
-2 Seismic Loads on Reactor Coolant System Components for Operational Basis Earthquake
List of Figures mber Title 1 P&ID for Reactor Coolant System & Pump (Sheet 1) 2 Reactor Coolant System Arrangement-Elevation 3 Reactor Coolant System Arrangement-Plan 1 Reactor Vessel 2 Steam Generator 3 Reactor Coolant Pump 4 P&ID Reactor Coolant Pump 5 Reactor Coolant Pump Seal Area 6 Reactor Coolant Pump Predicted Performance 7 Pressurizer 8 Temperature Control Program 9 Pressurizer Level Setpoint Program 10 Pressurizer Level Control Program 11 Quench Tank 12 Permanent Reactor Cavity Seal Plate 1 Reactor Coolant System Pressure Temperature Limitations for 7 Full Power Years 2 Reactor Coolant System Pressure - Temp Limitations During Plant Heatup/
Cooldown After 7 Years Integrated Neutron Flux
-3 Reactor Coolant System Pressure Temperature Limitations For 0 to 2 Years of Full Power 4 Reactor Coolant System Heatup Limitations for 54 EFPY 5 Reactor Coolant System Cooldown Limitations for 54 EFPY 1 Location of Surveillance Capsule Assemblies 2 Typical Surveillance Capsule Assembly 3 Typical Charpy Impact Compartment Assembly 4 Typical Tensile-Monitor Compartment Assembly 5 Base Metal - WR (Transverse) Plate C-506-1 Impact Energy vs Temperature
List of Figures (Continued) mber Title 6 Base Metal - WR (Transverse) Plate C-506-1 Lateral Expansion versus Temperature 7 Base Metal - RW (Longitudinal) Plate C-506-1 Impact Energy versus Temperature 8 Base Metal - RW (Longitudinal) Plate C-506-1 Lateral Expansion vs Temperature 9 Weld Metal Plate C-506-2/C-506-3 Impact Energy vs Temperature 10 Weld Metal, Plate C-506-2/C-506-3 Lateral Expansion vs Temperature 11 HAZ Metal, Plate C-506-1 Impact Energy versus Temperature 12 HAZ Metal, Plate C-506-1 Lateral Expansion versus Temperature 13 SRM (HSST Plate 01MY - Longitudinal) Impact Energy versus Temperature 14 SRM (HSST Plate 01MY - Longitudinal) Lateral Expansion versus Temperature
-1 Reactor Coolant System Seismic Analysis Model MS2
-1A Reactor Coolant System - Seismic Analysis Model MS2 and RV14
-2 RV14 Reactor and Internals Seismic Analysis Model
-3 Pressurizer Seismic Analysis Model
-4 Surge Line Seismic Analysis Model
GENERAL SYSTEM DESCRIPTION function of the reactor coolant system is to remove heat from the reactor core and internals transfer it to the secondary (steam generating) system. The reactor coolant system, which is rely located within the containment building, consists of two heat transfer loops connected in llel across the reactor pressure vessel. Each loop contains one steam generator, two reactor lant pumps, connecting piping, and flow and temperature instrumentation. Coolant system sure is maintained by a pressurizer connected to one of the loop hot legs.
iping and instrumentation diagram of the reactor coolant system is shown in Figure 4.1-1. The nds for the piping and instrumentation diagram are given in Figures 9.1-1, 9.1-2 and 9.1-3.
vation and plan views of the reactor coolant system are shown in Figure 4.1-2 and Figure 4.1-espectively. During operation, the four pumps circulate water through the reactor vessel where erves as both coolant and moderator for the core. The heated water enters the two steam erators, transferring heat to the secondary (steam) system, and then returns to the pumps to at the cycle.
tem pressure is maintained by regulating the water temperature in the pressurizer where steam water are held in thermal equilibrium. Steam is either formed by the pressurizer heaters or densed by the pressurizer spray to limit the pressure variations caused by the contraction or ansion of the reactor coolant. The pressurizer is located with its base at a higher elevation than reactor coolant loop piping. This eliminates the need for a separate pressurizer drain, and ures that the pressurizer is drained before maintenance operations.
Reactor Coolant System (RCS) is protected against overpressure by two ASME Section III e approved spring-loaded safety valves. In addition, two solenoid-operated power relief es (PORVs) are provided as described in Section 4.3.7. Both the safety valves and the PORVs connected to the top of the pressurizer. Steam discharged from the valves is cooled and densed by water in a quench tank. In the unlikely event that the discharge exceeds the capacity he quench tank, the tank is relieved via a rupture disc to the containment atmosphere. The ure disc is provided as the tank code over pressure protection device. The quench tank is ted at a level lower than the pressurizer. This ensures that any power-operated relief valve or surizer valve leakage from the pressurizer, or any discharge for these valves, drains to the nch tank.
rpressure protection for the secondary side of the steam generators is provided by ASME e safety valves located in the main steam line pipes upstream of the steam line isolation es. Power-operated steam dump and bypass valves are provided to prevent opening of the ondary safety valves following a loss-of-load incident. The secondary pressure protection is cribed in Sections 10.3 and 4.3.2.
maintain reactor coolant chemistry within the limits discussed in Section 4.4.3 and to control surizer level, a continuous but variable bleed flow from one loop upstream of the reactor lant pump is maintained. This bleed flow is controlled by pressurizer level.
rations.
inlet nozzle on each of the four reactor vessel inlet pipes allows injection of borated water into reactor vessel from the safety injection system in the event emergency core cooling is needed.
ing a normal plant shutdown, these nozzles are also used to supply shutdown cooling flow m the low pressure safety injection pumps. An outlet nozzle on one reactor vessel outlet pipe is d to remove shutdown cooling flow.
t and drain connections in the reactor coolant piping are provided for draining the reactor lant system to the radioactive waste processing system for maintenance operations. A nection is also provided on the quench tank for draining it to the radioactive waste processing em following a relief valve or safety valve discharge. Other reactor coolant loop penetrations ude sampling connections (Section 9.6) and instrument connections. The nozzle tifications are tabulated in Figure 4.1-3. Normal draining of the RCS is through the chemical volume control system.
ere required to reduce heat losses and protect personnel from high temperatures, components piping in the reactor coolant system are insulated with a material compatible with the peratures involved. All insulation material used on stainless steel has a soluble chloride tent of less than 600 ppm to minimize the possibility of chloride-induced stress corrosion.
ctroslag welding was not used in the construction of any reactor coolant boundary component.
major reactor coolant system components are designed for a 40 year service life. To assure this objective can be attained, strict quality assurance standards as outlined in tions 4.6.4and 4.6.5 were followed.
tection provided the reactor coolant system against environmental factors such as fires, floods missiles is described in other sections (see Chapters 1, 5, 9 and 11).
bulation of the RCS volumes is contained in Table 4.1-1.
Component Volume (ft3)
Reactor Vessel 4652 Steam Generators 3386 Reactor Coolant Pumps 449 Pressurizer 1500 Piping: Hot Leg 280 Piping: Cold Leg 752 Piping: Surge Line 32 Quench Tank 217
figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
FIGURE 4.1-1 P&ID FOR REACTOR COOLANT SYSTEM & PUMP (SHEET 2) figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
FIGURE 4.1-1 P&ID FOR REACTOR COOLANT SYSTEM & PUMP (SHEETS 3) figure indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
FIGURE 4.1-2 REACTOR COOLANT SYSTEM ARRANGEMENT-ELEVATION Revision 4006/30/22 MPS-2 FSAR 4.1-7
FIGURE 4.1-3 REACTOR COOLANT SYSTEM ARRANGEMENT-PLAN Revision 4006/30/22 MPS-2 FSAR 4.1-8
1 DESIGN PARAMETERS reactor coolant system operated initially at a core power level of 2560 MWt but has since n uprated to a core power level 2700 MWt. The major systems and components which bear ificantly on the acceptability of the site have been evaluated for operation at a core power l of 2700 MWt (NSSS power of 2715 MWt).
reactor design described in Chapter 3 predicates hot leg temperature, cold leg temperature, imum reactor coolant flow and reactor vessel pressure drop. These thermodynamic and rodynamic data are used in the design of the steam generator, reactor coolant pump, and tor coolant piping as described in Section 4.3 for each of these components.
principal design parameters for the reactor coolant system are listed in Table 4.2-1. The gn parameters for each of the major components are given in Section 4.3. The reactor coolant em is designated a Class 1 system for seismic design and is designed to the criteria for load binations and stresses which are presented in Table 4.2-2A and 4.2-2B. Seismic Analysis is ussed in Appendix 4.A.
system design temperature and pressure are conservatively established and exceed the bined normal operating value and the change due to anticipated operating transients. They ude the effects of instrument error and the response characteristics of the control system. The nge due to the anticipated transients also considers the effect of reactor core thermal lag, lant transport time, system pressure drop and the characteristics of the safety and relief valves.
following design cyclic transients, which include conservative estimates of the operational uirements for the components discussed in Section 4.3, were used in the fatigue analyses uired by the applicable codes listed in Table 4.2-3; the applicable operating condition category esignated by ASME Section III is indicated in each case.
- a. Five-hundred heatup and cooldown cycles during the systems 40 year design life at a heating and cooling rate of 100F/hr between 70F and 532F.
The replacement reactor vessel head is designed for 200 steady state and transient operating cycles for plant heatup and cooldown at a rate of 100F/hr between 70F and 532F.
The replacement pressurizer is analyzed for 500 heatup cycles at a rate of 100Fhour and 500 cooldown cycles at a rate of 200F/hour.
Category: Normal Condition.
- b. Pressurizer spray piping is limited to 160 plant heatup and cooldown cycles.
Primary manway studs of the replaced steam generators are limited to 200 heatup
- c. Fifteen-thousand power change 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 analyzed to 15,000 cycles for plant loading and unloading at 5% of full load/minute.
Category: Normal Condition.
- d. Primary manway studs for the 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 full 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 3110 psig and a temperature at least 60F 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 70F - 400F, above the minimum RTNDT + 60F.
Category: Test Condition.
- g. Two-hundred cycles of leak testing at 2485 psig and at a temperature at least 60F greater than the NDTT of the component having the highest NDTT.
Category: Test Condition; evaluated as Upset Condition.
- h. Primary manway studs for the replaced steam generators are limited to 80 cycles of leak testing at 2485 psig.
- i. 106 cycles of normal variations of 100 psi and 6 F at operating temperature and pressure.
The replacement pressurizer is analyzed for 106 cycles of normal variations of 100 psi and 7F at operating temperature and pressure.
- j. Four-hundred reactor trips 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 pressurizer are analyzed for 100 bolt/unbolt cycles. The vent port studs for the replaced pressurizer are analyzed for 200 bolt/
unbolt cycles.
Category: Normal Condition
- l. Primary manway studs for the replaced steam generators are limited to 1500 cycles of 10% of full load step power changes, increasing from an initial power level of 15% to 90% of full power and decreasing from initial power level between 15%
and 100% of full power.
Category: Normal Condition
- m. Primary manway studs for the replaced steam generators are limited to 200 reactor trips from 100% power.
Category: Upset Condition ddition to the above list of normal design transients, the following abnormal transients were considered when arriving at a satisfactory usage factor as defined in Section III of the ASME ler and Pressure Vessel Code; however, emergency condition transients were not used to form basis for the code design of the components based on Paragraph N-417.10(f) of ASME III, 8 Edition, Summer 1968 addenda.
- 1. Forty cycles of loss of turbine load from 100 percent power with 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.
ute without reactor trip. The system will accept, without damage, a complete loss of load.
2 CODES ADHERED TO Codes adhered to and component classifications are listed in Table 4.2-3 and conform to CFR Part 50, Section 50.55a. The construction permit date was December 11, 1970; thus in all ances, code dates identified below meet or exceed those required. The impact properties of all erials which form a part of the pressure boundary meet the requirements of the ASME Boiler Pressure Vessel Code Section III, Paragraph N330, at a temperature of 40F. The impact perties of the replacement reactor vessel closure head and replacement pressurizer meet the uirements of ASME Section III, NB 2300.
eneral, code editions and addenda in effect on the date of the original purchase order to a ufacturer apply in the design, manufacture and testing of those components. The code ions and addenda which apply for the components in Tables 1.2-1 and 4.2-3 are specified in licable FSAR subsections.
compliance with Safety Guide 26 was not possible since most of the components covered by ety Guide 26 were purchased and fabrications begun prior to the March 23, 1972, issue date this guide. Table 4.2-4 provided a comparison of those components which are not in pliance with Safety Guide 26.
instrument air system for Millstone Unit 2 is not required for safe shutdown. Therefore, ety Guide 26 does not apply to the design codes for this system.
codes and standards used for the components of the diesel oil supply are as follows:
Tanks (Above Ground) - API 650 Tanks (Below Ground) - NFPA Number 30 Piping - ANSI B31.1.0 (MOD C)
Valves - ANSI B31.1.0 (MOD C) addition, seismic category 1 requirements and Quality Assurance Program as outlined in endix 1B (located in the original FSAR dated August 1972) were employed in the ufacture.
ANS N18.2 system of quality group classifications has been utilized by the NSSS vendor for lstone Unit 2. The AEC has voted affirmatively for the adoption of N18.2 as an ANSI dard. It is more definitive than Safety Guide 26, is supported by the major NSSS suppliers and he utility industry, and compliance with its provisions provides a satisfactory alternative to the ety Guide.
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-Chromium 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-Chromium 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 the most severe loading combination, which includes the Design Basis Earthquake loads, the ary stresses in the ASME Code Section III, Class 2 and 3 components and component ports are limited to levels comparable to the emergency stress limits defined in ASME Code tion III for Class 1 components.
major components of the reactor coolant system have been placed in the safety classes as ned by ANS N 18.2 Safety Criteria for the Design of Stationary Pressurized Water Reactor nts.
4 PART-LOOP OPERATION maximum temperature of the hot leg and cold leg will be less than the maximum peratures for design power at design flow. Reactor power operation with less than 4 reactor lant pumps operating or natural circulation is not allowed. However, decay heat will be sferred to the steam generator for both cases. Current Technical Specifications restrictions hibit other than four reactor coolant pump power operations.
adequacy of natural circulation for decay heat removal after reactor shutdown has been fied analytically and by tests on the Palisides reactor. The core T in the analysis has been wn to be lower than the normal full power T; thus, the thermal and mechanical loads on the structure are less severe than normal design conditions.
assess the margin available in a post-coastdown situation, a study was made assuming ination of pump coastdown 100 seconds after reactor trip, with immediate flow decay to the le natural circulation condition. It should be recognized that pump rotation will not have ped for substantially longer than 100 seconds. With the maximum decay heat load 100 onds after trip, the system will sustain stable natural circulation flow adequate to give a power ow ratio of less than 0.9.
t removed from the core during natural circulation may be rejected either by dumping to the n condenser or to the atmosphere; the rate of heat removal may be controlled to maintain core within allowable limits. The analytical techniques are verified by tests completed on the sades reactor (AEC Docket Number 50-225).
TABLE 4.2-1 PRINCIPAL DESIGN PARAMETERS OF REACTOR COOLANT SYSTEM ign Thermal Power, (NSSS) MWt 2715, Btu/hr 9.26 x 109 ign Pressure, psig 2485 ign Temperature (Except Pressurizer, 700F), F 650 ign Coolant Flow Rate, gpm 325,000 (1) d Leg Temperature, Normal Service, F 550 (1)
Leg Temperature, Normal Service, F 604 (1) mal Operating Pressure, psia 2250 (1) tem Volume, ft3 (Without Pressurizer) 9,551 ssurizer Water Volume, ft3 942 at 65% level ssurizer Steam Volume, ft 590 at 65% level CS piping and vessel design parameters. Volumetric flow rate is based on 122 x 106 lbm/hr mass flow rate and cold leg density. See FSAR Section 14 for principal RCS parameters used in Safety Analyses.
Primary Stress Limits Loading Combinations Vessels Piping Supports Working Stress
See Note b. PB + PL 2.25Sm See Note c. PB 1.5[1-(PM/SD)2]SD PB 4SD/ cos[PM/
2SD]
- 3. Normal Operating Loadings + Pipe PM SL PM SL Deflection of supports limited to Rupture + Maximum Hypothetical maintain supported equipment within Earthquake (DBE). limits shown in columns 1 and 2 See Notes a and b. PB 1.5[1-(PM/SL)2]SL PL + PB 3SM See Note c. PB 4SL/ cos[PM/
2SL]
Notes:
(a) This load combination is not applied to the piping run within which a pipe break is considered to have occurred.
(b) For loading combinations 2 and 3, stress limits for vessels, with the symbol PM changed to PL, should also be used in evaluating effects of local loads imposed 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.
PM = Calculated Primary Membrane Stress PB = Calculated Primary Bending Stress PL = Calculated Primary Local Membrane Stress SM = Tabulated Allowable Stress Limit at Temperature from ASME Boiler and Pressure Vessel Code,Section III or ANSI B31.7.
SY = Tabulated Yield at Temperature, ASME Boiler and Pressure Vessel Code,Section III.
SD = Design Stress SD = SY (for ferritic steels)
SD = = 1.2SM (for austenitic steels)
SL = SY + 1/3 (Su - SY)
Su = Tensile Strength of Material at Temperature The following typical values are selected to illustrate the conservatism of this approach for establishing stress limits. Units are 103 lbs/
square inch Material SY (1) SU SM (1) SP SL A-106B 25.4 60.0 (2) 17.0 25.4 36.9 SA-533B 41.4 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 (1) From ASME Boiler and Pressure Vessel Code,Section III, 1968 ED., at 650F.
(2) Minimum value at room temperature which is approximately the same at 650F for ferritic materials.
(3) Estimated
MITS FOR THE REPLACEMENT REACTOR VESSEL HEAD AND REPLACEMENT PRESSURIZER Loading Combinations ASME Code Subsection esign Pm Sm NB-3221.1 rvice Level A (Normal) P1 1.5Sm NB-3221.2 rvice Level B (Upset) P1 + Pb 1.5Sm NB-3221.3 P1 + Pb + Q 3.0Sm NB-3222.2 Pe 3.0Sm NB-3222.3 Ui 1.0 NB-3222.4 STH y' Sy NB-3222.5 rvice Level D (Faulted) Pm Lesser of 2.4Sm and 0.7Su NB-3225 Appendix F P1 1.5Pm NB-3225 Appendix F P1 + Pb 1.5Pm NB-3225 Appendix F ere:
= General Primary Membrane Stress Intensity Local Membrane Stress Intensity Bending Stress Intensity Secondary Stress Intensity Expansion Stress Intensity Actual Service condition cycles divided by allowable cycles, based on calculated alternating ss and fatigue design curve.
= Thermal Stress Range
= Design Stress Intensity Tensile Strength Yield Stress Maximum allowable range of thermal stress on an elastic basis divided by Sy.
PRESSURIZER (CONTINUED) following typical values in ksi are selected for the replacement reactor vessel head materials 50F.
Material Sy Su Sm 508 Grade 3 Class 1 41.5 80.0 26.7 ASME Section II, Part D 167 (Alloy 690) 20 80 20.0 ASME Code Case N525/Section II, Part D 166 (Alloy 690) 27.5 80.0 23.3 ASME Section II, Part D 182 F316LN 17.8 62.8 16.0 ASME Section II, Part D 312 TP316L 15.3 61.7 13.8 ASME Section II, Part D following typical values in ksi are selected for the pressurizer materials at 650F.
Material Sy Su Sm 508 Grade 3 Class 2 53.9 90.0 30 ASME Section II, Part D 182 Grade F 316 18.5 67.0 16.7 ASME Section II, Part D
Components Codes actor Vessel (excluding replacement 1. ASME,Section III, Class A, 1968 Edition, actor Vessel Closure Head and Addenda through Summer 1969 zzles), Original Upper Shell of Steam nerator placement Reactor Vessel Closure 1. ASME Section III, 1998 Edition through 2000 ad and Nozzles, Replacement Addenda.
ssurizer placement Lower Steam Generator 1. ASME,Section III, Class I, 1983 Edition, Addendum to Summer of 1984 actor Coolant Pumps 1. Draft ASME Code for Pumps and Valves for Nuclear Power, Class 1, November 1968, including March 1970 Addenda.
- 2. 2. ASME Section III, paragraph N153 in Summer 1969 Addenda.
- 3. ASME Section III, Appendix IX.
ench Tank ASME Section III, Class C, 1968 ssurizer Safety Valves 1. ASME Section III, Class A, 1968 Edition, Addenda through Summer of 1970, Code Case 1344-1 ing 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.
condary Safety Valves ASME Standard Code for Pumps and Valves for Nuclear Power, Class 2, March, 1970 Draft e: 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.
Requirements Imposed In Classification Per Applicable Code Code Used In Addition to Code Requiremen System Components Safety Guide 26 Per Safety Guide 26 Manufacture Used in Manufacture Reactor Building Pressure Quality Group C ASME Section III, ASME Seismic Category I, manufacture Closed Cooling Vessels Class 3 Section VIII under Quality Assurance Progra Water System Division I of Appendix 1B (a), Spot radiographed Piping Quality Group C ASME Section III, ANSI B31.1.0 Seismic Category I, 10% random (excluding Class 3 MOD B radiography of butt welds for Containment piping 4 inch and larger, material Penetrations) identification, manufactured und Quality Assurance Program of Appendix 1B (a)
Pumps Quality Group C ASME Section III, Standards of the All pressure containing parts whe Class 3 Hydraulic hydrostatically tested at a minimu Institute of 1.5 times the design pressure seismic I manufactured under Quality Assurance Program of Appendix 1B (a)
Valves Quality Group C ASME Section III, ANSI B31.1.0 Seismic Category I, MT-PT (excluding Class 3 MOD B Examination, material traceabilit Containment) on pressure retaining parts, manufactured under Quality Assurance Program of Appendix 1B (a)
Requirements Imposed In Classification Per Applicable Code Code Used In Addition to Code Requiremen System Components Safety Guide 26 Per Safety Guide 26 Manufacture Used in Manufacture Safety Injection Shutdown Quality Group C ASME Section III, ASME Seismic Category I System Heat Class 2 Section III, Exchangers Class 3, TEMA R
Refueling Quality Group B ASME Section III, ASME Seismic Category I, manufacture Water Tank Class 2 Section III, under Quality Assurance Progra Class 3 of Appendix 1B (a)
Piping Quality Group B ASME Section III ANSI B31.1.0 Seismic Category I, Material HCD(C) MOD(C) Identification per ASTM Specification Pumps Quality Group B ASME Section III, ASME Code for Seismic Category I, manufacture Class 2 Pumps and under Quality Assurance Progra Valves for of Appendix 1B (a) (See Question Nuclear Power, 4.4)
Class II Valves Quality Group B ASME Section III, ANSI B31.1.0 Seismic Category I, Material HCD(C) Only Class 2 MOD(C) Identification per ASTM Specification Piping 4 inch Quality Group B ASME Section III, ANSI B31.1.0 Seismic Category I HCD-3 and 6 Class 2 inch HCD-3
Requirements Imposed In Classification Per Applicable Code Code Used In Addition to Code Requiremen System Components Safety Guide 26 Per Safety Guide 26 Manufacture Used in Manufacture Auxiliary Feedwater Condensate Quality Group B ASME Section III, AWWA D100 Seismic Category I, manufacture System Storage Tank Class 2 NFPA Volume 6 under Quality Assurance Progra of Appendix 1B (a) later modified to API 620 or equivalent based o Design Change to a pressurized tank Piping Quality Group B ASME Section III, ANSI B31.1.0 Seismic Category I, 10% random (excluding Class 2 MOD B radiography of butt welds for Containment piping 4 inch and larger, material Penetrations) identification manufactured unde Quality Assurance Program of Appendix 1B Pumps Quality Group B ASME Section III, ASME Code for Seismic Category I manufactured Class 2 Pumps and under Quality Assurance Progra Valves for of Appendix 1B (a)
Nuclear Power, Class II Valves Quality Group B ASME Section III, ASNE B31.1.0 Seismic Category I, MT-PT (excluding Class 3 MOD B Examination, material traceabilit Containment on pressure retaining parts, Isolation) manufactured under Quality Assurance Program of Appendix 1B (a)
Requirements Imposed In Classification Per Applicable Code Code Used In Addition to Code Requiremen System Components Safety Guide 26 Per Safety Guide 26 Manufacture Used in Manufacture Service Water Piping Quality Group C ASME Section III, ANSI B31.1.0 Seismic Category I, manufacture System Class 3 under Quality Assurance Progra of Appendix 1B (a)
Pumps Quality Group C, ASME Section III ASME Perform Test according to ASME Class 3 Section VIII PTC 8.2 1965, Seismic Category manufactured under Quality Assurance Program of Appendix 1B (a)
Valves Quality Group C ASME Section III, ANSI B31.1.0 Seismic Category I, manufacture Class 3 under Quality Assurance Progra of Appendix 1B (a)
Vital Chilled Water Piping Quality Group C ASME Section III, ANSI B31.1.0 Seismic Category I, manufacture System Class 3 under Quality Assurance Progra of Appendix 1B (a)
Valves Quality Group C ASME Section III, ANSI B31.1.0 Seismic Category I, manufacture Class 3 under Quality Assurance Progra of Appendix 1B (a)
Condenser/ Quality Group C ASME Section III, ASME Seismic Category I, manufacture Evaporators Class 3 Section VIII, under Quality Assurance Progra Division I of Appendix 1B (a)
(a)Appendix 1B was located in the original FSAR dated August 15, 1972.
1 REACTOR VESSEL reactor vessel (Figure 4.3-1) is supported by three pads welded to the underside of the reactor sel nozzles. The arrangement of the vessel supports, allows radial growth of the reactor vessel to thermal expansion while maintaining it centered and restrained from movement caused by mic disturbances. Departure from levelness of not more than 0.002 inch per foot of flange meter is maintained during construction to facilitate proper assembly of reactor internals. The gn parameters for the reactor vessel are given in Table 4.3-1.
vessel closure flange is a forged ring with a machined ledge on the inside surface to support reactor internals and core. No other ring forgings are used for reactor vessel shell sections. The ge is drilled and tapped to receive the closure studs and is machined to provide a mating ace for the reactor vessel closure seal. The vessel closure contains 54 studs, 7 inches in meter, with eight threads per inch. The stud material is ASTM A-540, Grade B24, with a imum yield strength of 130,000 psi. The tensile stress in each stud when elongated for rational conditions is approximately 40 ksi. Calculations show that 32 uniformly distributed s can fail before the closure will separate at design pressure. However, 16 uniformly ributed broken studs or four adjacent broken studs will cause O-ring leakage.
radial nozzles on a common plane are located just below the vessel closure flange. Extra kness in this vessel-nozzle course provides most of the reinforcement required for the nozzles.
itional reinforcement is provided for the individual nozzle attachments. A boss located und each outlet nozzle on the inside diameter of the vessel wall provides a mating surface for internal structure which guides the outlet coolant flow. This boss and the outlet sleeve on the support barrel are machined to a common contour to reduce core bypass leakage. A fixed ispherical head is attached to the lower end of the shell. There are no penetrations in the lower d.
removal top closure head is hemispherical. The closure head is single piece low alloy steel ing replaced during refueling outage 16. All surfaces in contact with reactor coolant are clad h a quarter inch nominal thickness weld deposit similar to type 304 stainless steel. The nozzles he reactor vessel head support the Control Element Drive Mechanisms (CEDMs). The CEDM, ore Instrumentation (ICI) and vent nozzles are constructed from Inconel alloy 690 material to imize the susceptibility to Primary Water Stress Corrosion Cracking. Threaded housing ges made of stainless steel are joined to the guide tubes by Gas Tungsten Arc Welding. The d flange is drilled to match the vessel flange stud bolt locations. The 54 stud bolts are fitted h spherical washers located between the closure nuts and head flange to maintain stud nment during head flexing due to boltup. To ensure uniform loading of the closure seal, the s are tensioned with hydraulic stud tensioners. Stud elongation is then measured to ensure per preload on all the studs.
nge sealing is accomplished by a double-seal arrangement utilizing two silver-plated Ni-Cr-Fe y, self-energized O-rings. The space between the two rings is monitored to allow detection of inner ring leakage. The control element drive mechanism (CEDM) nozzles (Ni-Cr-Fe alloy
ndary. In addition to these nozzles, there is a three-quarter inch vent connection.
core is supported from an internally machined core support ledge.
2 STEAM GENERATOR nuclear steam supply system (NSSS) utilizes two steam generators (Figure 4.3-2) to transfer heat generated in the reactor coolant system (RCS) to the secondary system and produce m at the warranted steam pressure and quality. The design parameters for the steam generators given in Table 4.3-2.
steam generator is a vertical U-tube heat exchanger. The steam generator operates with the tor coolant in the tube side and the secondary fluid in the shell side.
ctor coolant enters the steam generator through the inlet nozzle, flows through three-quarter OD U-tubes, and leaves through two outlet nozzles. Vertical partition plates in the lower d separate the inlet and outlet plenums. The plenums are stainless steel clad, while the primary of the tube sheet is Ni-Cr-Fe clad. The vertical U-tubes are Ni-Cr-Fe alloy. The tube-to-tube et joint is welded on the primary side.
dwater enters the steam generator through the feedwater nozzle where it is distributed via a water distribution ring having top discharge J nozzles which direct the flow into the ncomer. The downcomer is an annular passage formed by the inner surface of the steam erator shell and the cylindrical shell wrapper which encloses the vertical U-tubes.
he bottom of the downcomer, the secondary water is directed upward past the vertical U-tubes re it is boiled to produce steam in the evaporator. The heat transfer area is determined by the uired heat transfer, the thermal driving force and the heat transfer coefficient. The heat transfer fficient in the evaporator is calculated from experimental data.
n exiting from the vertical U-tube heat transfer surface, the steam water mixture enters trifugal-type separators. These impart a centrifugal motion to the mixture and separate the m from the water. The water leaves the primary separator through the bottom of the separator sing and is directed into the downcomer where it is mixed with the feedwater. Final drying of steam from the centrifugal primary separators is accomplished by directing the steam through ondary cyclones. The moisture content of the outlet steam is no greater than 2.0 percent at gn flow.
steam generator primary side pressure loss is determined by summing the losses due to tion in the tubes, in the tube bends, entrances and exits, and the steam generator inlet and et plenums and nozzles.
steam generator shell is constructed of carbon steel. Manways and handholes are provided for y access to the steam generator internals.
ass system is described in Section 10.
rpressure protection for the shell side of the steam generators and the main steam line piping to the inlet of the turbine stop valve and provided by 16 spring-loaded ASME Code safety es which discharge to atmosphere. Eight of these safety valves are unted on each of the main steam lines outside the containment upstream of the steam lined ation valves. The opening pressure of the valves is set in accordance with ASME Code wances. Parameters for the main steam safety valves are given in Table 4.3-3.
rumentation has been added to each safety relief valve (SRV) to provide a main control board unciator alarm indication of valve closed/not closed, (Regulatory Guide 1.97, Rev. 2, D-18 able).
main control board annunciator window (C05 D17A/B) is a split window and will alarm n any of the SRV's on Steam Generator 1 or 2, respectively, are not closed.
n receiving this alarm, the operator would use the Plant Process Computer display of the m Generator Safety Valves or local indication, to determine which valve on Steam Generator 2 is not closed.
steam generators are vertically mounted on bearing plates which allow lateral motion due to mal expansion of the reactor coolant piping. Stops are provided to limit this motion in case of olant pipe rupture.
top of each unit is restrained from sudden lateral movement by suitable stops and hydraulic bbers mounted rigidly to the concrete structure.
ddition to the transients listed in Section 4.2.1, each steam generator is also designed for the owing conditions such that no component is stressed beyond the allowable limit as described SME Code,Section III (Table 4.3-2):
- a. Four-thousand cycles of transient pressure 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.
Category: Test Condition.
- d. Fifteen-thousand cycles of adding 600 gpm of 70F feedwater with the plant in hot standby condition.
Category: Normal Condition.
ddition to the normal design transients listed above, and those listed in Section 4.2.2, the owing additional abnormal transient was also considered in arriving at a satisfactory usage or as defined in Section III of the ASME Code:
Eight cycles of adding a maximum of 650 gpm of 70F feedwater with the steam generator secondary side dry and at 620F.
Category: Emergency Condition.
unit is capable of withstanding these conditions for the prescribed numbers of cycles in ition to the prescribed operating conditions without exceeding the allowable cumulative usage or as prescribed in ASME Code,Section III.
steam generators are located at a higher elevation than the reactor vessel. The elevation erence creates natural circulation sufficient to remove core decay heat following coastdown of RCPs.
steam generators are equipped with a nitrogen addition system which has the capability of itting N2 to the bottom blowdown headers to mix the chemicals in the steam generators ng wet layup, and also through nozzles in the transition cones for blanketing the steam erators and steam lines above the water level. A flowmeter provides nitrogen flow control and al verification that flow is occurring. A drain line on the steam generator upper vent line ws checking that the steam generator has not been flooded. A test gauge may be installed on line to measure steam generator overpressure.
ddition to the transients listed in this section, and those in Section 4.2.1 the following factors e considered in the design of the steam generators.
2.1 Flow Induced Vibration steam generator has also been designed to ensure that critical vibration frequencies will be l out of the range expected during normal operation and during abnormal conditions. The m generator tubing, tube sleeves, and tubing supports are designed and fabricated with siderations given to both secondary side flow induced vibrations and RCP induced vibrations.
ddition, the heat transfer tubing, tube sleeves, and tube supports are designed so that they will
ause the RCPs have a rotational speed of 900 rpm (less normal slip) the tube bundle design considered the imposition of exciting frequencies of 14 to 15 cps and 70 and 75 cps. The er frequency range is defined as a mechanical vibration resulting from the transmission of a hanical impulse at the frequency of pump rotation. The upper frequency range is defined to be nusoidal pressure variation of 6 psi in the primary piping that contains the pump. The sure variation results from the impeller vanes interacting with the cutwater vane at the volute et during each revolution of the impeller.
as been found that all tubes and tube sections that will experience forcing functions from cross and parallel flow have natural frequencies sufficiently different from the frequency of the ing function that they will not experience damaging vibrations. The mechanical excitation uency is sufficiently different from the lowest natural frequency for out-of-plane or lateral ation in any tube span that critical vibration will not occur.
2.2 Tube Thinning original Combustion Engineering (CE) Steam Generators margin of tube-wall thinning that ld be tolerated without exceeding the allowable faulted stress limits under postulated dition of a design basis largest pipe break in the reactor coolant pressure boundary (RCPB) ng reactor operation was 0.008 inches.
12 inch excess material had been intentionally been provided in the tube wall thickness to ommodate the estimated degradation of tubes during the service lifetime. CE expected ligible tube wall thinning when operating under the specified secondary chemistry uirements.
ause of the use of volatile secondary water chemistry there has been no tube wall thinning erienced on the steam generators. The new steam generator tubes are designed to be at least cturally equivalent to the original and in compliance to Regulatory Guide 1.121. Therefore, wall thinning or other forms of tube degradation can be structurally accommodated with the e degree of margin as the original.
2.3 Potential Effects of Tube Ruptures steam generator tube rupture accident in a penetration of the barrier between the RCS and the n steam system. The integrity of this barrier is significant from the standpoint of radiological ty in that a leaking steam generator tube allows the transfer of reactor coolant into the main m system. Radioactivity contained in the reactor coolant would mix with water in the shell of the affected steam generator. This radioactivity would be transported by steam to the ine and then to the condenser, or directly to the condenser via the main steam dump and ass system. Noncondensible radioactive gases in the condenser are removed by the condenser ejector system and discharged to the plant vent. Analysis of a steam generator tube rupture dent, assuming complete severance of a tube, is presented in Section 14.14.
ioactivity concentrations in the secondary side of the steam generator is dependent upon the oactive concentration of the RCS, the primary to secondary leak rate, and the operating ory of the steam generator blowdown system.
3 REACTOR COOLANT PUMPS reactor coolant is circulated by four single speed, vertical, single suction, centrifugal type ps (Figure 4.3-3). The discharge nozzle is horizontal and the suction nozzle is in the bottom ical position. The pressure containing components are designed and fabricated in accordance h the ASME Boiler and Pressure Vessel Code,Section III, Class A.
design flow for the RCP is determined from the reactor mass flow. This mass flow is verted to volumetric flow at the full power cold inlet temperature to determine the pump gn flow. The maximum pressure loss at the design flow rate for the reactor vessel, steam erator, and piping is determined by adding an allowance for uncertainty to the best estimate each pressure drop. These maximum values are used to establish the RCP design head. The P is designed to produce the minimum reactor design flow at the maximum expected system sure loss.
minimum RCS pressure at any given temperature is limited by required net positive suction d (NPSH) for the RCP during portions of plant heatup and cooldown. To ensure that the pump SH requirements are met under all operating conditions, an operating curve is used which s permissible RCS pressure as a function of reactor coolant temperature. The RCP NPSH riction on this curve is determined by using the NPSH requirement for each pump bination and correcting it for pressure and temperature instrument errors and pressure surement location. The NPSH required versus pump flow is supplied by the pump vendor.
nt operation below this curve is prohibited. At low RCS temperatures and pressures, other siderations require the maximum pressure versus temperature curve to be above the NPSH ve.
pump impeller is pinned and bolted to the shaft. A close clearance thermal barrier assembly is unted above the water lubricated bearing to retard heat flow from the pump to the seal cavity ch is located above the thermal barrier. The thermal barrier assembly also tends to isolate the fluid in the pump from the cooler fluid above and, in the event of a seal failure, serves as an itional barrier to reduce leakage from the pump. Each pump is equipped with replaceable ng wear rings. A water lubricated bearing is located in the fluid between the impeller and mal barrier to provide shaft support. Additional shaft support is provided by bearings in the tric motor which is directly connected to the pump shaft by a rigid coupling.
shaft seal assembly located above the thermal barrier consists of four, face type mechanical s, three full-pressure seals mounted in tandem and a fourth low pressure backup vapor seal gned to withstand operating system pressure with the pump stopped. The performance of the t seal system is monitored by pressure and temperature sensing devices in the seal system ure 4.3-4). A controlled bleed off flow through the pump seals is maintained to cool the seals
mechanical seal), is drained to the containment sump via the containment trench and collected he radioactive waste processing system. Normal vapor seal leakage is minor, (approximately 0 8 GPM per RCP), and is considered to be negligible leakage to the containment atmosphere.
seals are cooled by circulating the controlled bleed off through the heat exchanger mounted grally within the pump cover assembly. No damage would result in the event of pump ration without cooling water for up to five minutes. To reduce plant downtime and personnel osure to radiation during seal maintenance, the seal system is contained in a cartridge which be removed and replaced as a unit. The seal cartridge can be replaced without draining the p casing. The seal detail is shown in Figure 4.3-5.
motor-mounted flywheel reduces the rate of flow decay upon loss of pump power. The bined inertia of the pump motor and flywheel is 100,000 lbm-ft2. Flow coastdown racteristics are discussed in Section 14.6.
RCPs are typical centrifugal volumetric flow machines. The pump response following a s-of-Coolant Accident (LOCA) is predicted using generally accepted methods as described in endix 1 of CENPD-26. CENPD-26 is a proprietary report entitled Combustion Engineering lytical Techniques for Evaluating Loss of Coolant Accidents. A spectrum of breaks in the P discharge line have been analyzed and the results follow a predictable pattern. Assuming of electrical power to the pump at the start of the LOCA, it is seen that the pumps initially speed because the volumetric flow through the pump is not sufficient to sustain the nominal ed of rotation. The volumetric flow increases during the transient, accelerating the pump to its imum speed. The extent of the initial loss of speed varies with the break size. The larger the ak size, the less the initial deceleration and the higher the maximum speed attained. The pump ins its maximum speed following a double-ended discharge break. The calculated torque osed on the impeller follows the same trend as the speed, with the maximum value occurring owing the double-ended discharge break.
need for a disengaging device to prevent motor overspeed following a LOCA has been luated. In view of the fact that the maximum anticipated pump speed is well within the safe rating limits for all rotating parts, a means to disengage the motor from the pump is not essary.
reak in the suction piping causes the reactor coolant to flow through the pump opposite to the mal direction of flow, decelerating the rotation of the pump until it is brought to rest against anti-reverse rotation device.
pump/motor assembly includes motor bearing oil coolers, seal chamber, controls and ruments. Cooling water is provided from the reactor building closed cooling water (RBCCW) em. The design parameters for the RCPs are given in Table 4.3-4. The RCP instrumentation is cribed in Section 4.3.8.5.
RCP and motor are supported by four support lugs welded to the volute. The pump is ported by four spring assemblies employed between the support lugs and the floor below.
major pump components wetted by the primary fluid are constructed of austenitic stainless l to minimize corrosion. These materials are listed in Table 4.3-4. The mechanical seals sist of a rotating tungsten (1) carbide ring riding over a hard carbon face. The design life of this 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 designed to accept a pressure p equal to full operating system pressure, but normally operates at one-third this pressure drop.
predicted pump performance curve is shown in Figure 4.3-6. The air-cooled, self-ventilated p motor is sized for continuous operation at the flows resulting from four-pump operation or ial pump operation with 0.76 specific gravity water. The motor service factor is sufficient to w 500 heatup cycles during which the nominal horsepower load will decrease from 6000 to 0 over a period of seven hours. The motors are designed to start and accelerate to rated speed er full load when 70 percent or more of rated normal voltage is applied. The motors are tained within standard drip-proof enclosures and are equipped with electrical insulation able for a zero to 100 percent humidity and radiation environment of 30 R/hr.
design requirements of the RCPs include a minimum inertia for the rotating assembly of
,000 lbm - ft2; to achieve this total, a flywheel with an inertia of 70,000 lbm - ft2 has been rporated.
h original RCP motor flywheel assembly consists of two solid discs bolted together, shrink d 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 selection of material, machining and manufacturing operations, quality control, and the rous acceptance criteria established to assure the integrity of the flywheel and to minimize rating stresses include the following:
The principal stress is 30 percent of the yield point of the flywheel material (based on the tensile tests per ASTM-A-20) at the design overspeed of 125% of the normal operating speed not considering keyway stress concentration factors. The minimum keyway fillet radius is one eight inch.
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.
All flywheel discs have passed the following nondestructive testing:
or silicon
- a. One-hundred percent Ultrasonic Inspection per ASME Code,Section III, paragraph N-321.1
- b. One-hundred percent Magnetic Particle Examination per ASME Code,Section III, paragraph N-322.2.
- 2. Testing after finish machining
- a. One-hundred percent Ultrasonic Inspection of the flats and edges, performed in compliance with ASME Code,Section III, paragraph N-321.1
- b. Liquid penetrant inspection performed 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.
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.
Welding was not performed on flywheel discs.
The keyway fillet radius on the bottom of the keyway is 0.125 inches minimum.
flywheel material, which is pressure vessel quality, vacuum-improved steel plate, exceeds the uirements of ASTM-A-516, Grade 70 even though it was originally specified for TM-A-516 Grade 65. To improve the fracture toughness properties of the material, the flame discs, with the one-half inch allowance for machining, were heat treated as follows:
- 1. heated to 1650F 25F 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 400F;
- 3. tempered at 1140F for one-half hour per inch of thickness, air cooled.
composition of the material as certified by the steel vendor is as follows:
STM-E-30 ASTM-E-30 C Mn P S Si elt Number Slab Number (Weight %) (Weight %) (Weight %) (Weight %) (Weight %)
725 3&4 0.21 0.97 0.008 0.025 0.23 176 2 0.22 1.16 0.006 0.020 0.24
sile tests per ASTM-A-20 p weight test (DWT) - test at +40F per ASTM-E-208.
0.2% Offset Melt Slab Tensile Elongation in Charpy V-notch at Yield Strength DWT umber Number Strength PSI 2 inch (%) 40°F (ft-lbf)
PSI 725 3 76700 49000 30 104 109 91 OK 725 4 75600 50700 31 103 95 97 OK 176 2 76700 51500 29 84 73 83 OK Charpy values measured for the flywheel material at 40F are substantially higher than the compiled on SA-516 grade 70 material by the Research & Product Development Department E. The report titled Longitudinal and transverse Charpy V-notch impact and dropweight test for normalized and tempered SA-516 grade 70 material issued on August 26, 1971, and her identified by Laboratory Number X-24053 and R&PD Project Number 420001, was pared for the Industrial Cooperative Program of the Material Division of the Pressure Vessel earch Committee of the Welding Research Council.
s indicates that the toughness properties of these wheels are better than typical SA-516 Grade Therefore, the nil-ductility transition (NDT) temperature is lower than the highest value of F reported in that report.
ce the normal operating temperature of the flywheel is approximately 100F, a substantial gin exists between the computer KI for large hypothetical cracks, and the toughness KIC. The cal crack size therefore, is greater than five inches from the bore of the wheel.
ck growth calculations indicate that the number of starting cycles to cause a reasonably small k to grow to critical size is orders of magnitude greater than the number of cycles expected ng the life of Millstone Unit 2.
loads that are considered for the calculation of the stresses in the flywheel are the combined ary stresses in the flywheel at normal operating speed. They include the stress due to rference fit on the shaft as well as the stress due to centrifugal force.
mal operating speed of the flywheel is 900 rpm (less normal slip). The flywheel has a design rspeed of rated rpm plus 25 percent, which equals 1,125 rpm. The maximum tangential sses in the flywheel at normal operating speed are 25 percent of the material yield strength.
maximum tangential stresses in the flywheel at design overspeed are 30 percent of the erial yield strength.
h RCP motor.
replacement RCP motor flywheel is a one piece forging, 75 inches OD by 12 inches thick, nk fitted onto and keyed to the motor shaft above the rotor.
replacement RCP motors were procured as Quality Assurance items in accordance with CFR 21 and 10 CFR 50 Appendix B. They are fully interchangeable with the original RCP ors.
selection of material, machining and manufacturing operations, quality control, and the rous acceptance criteria established to assure the integrity of the flywheel and to minimize rating stresses include the following:
The principal stress does not exceed 30 percent of the yield point of the flywheel material (based on the tensile tests per ASTM-A-370) at the design overspeed of 125 percent of the normal operating speed not considering keyway stress concentration factors. The minimum keyway fillet radius is one-eighth inch.
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.
All flywheel discs passed the following nondestructive testing:
r to machining:
- a. The flywheel material was subjected to a 100 percent volumetric ultrasonic examination using procedures and acceptance criteria as specified in Paragraphs NB2532.1 and NB2532.2 of the ASME B&PV Code,Section III.
composition of the material as certified by the steel vendors is as follows:
mical Composition (weight percent)
Heat C Si Mn P S Cr Mo Ni V umber 0647 0.16 0.04 0.23 0.004 0.003 1.60 0.43 3.60 0.01 5100 0.20 0.08 0.33 0.007 0.005 1.61 0.44 3.54 0.03 heat treatment was as follows:
- 1. Heated to 1,560F to 1,580F 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 />.
- 3. Tempered at 1,110F to 1,185F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, air cooled.
sile tests were performed on the flywheel material per ASTM A370.
NDT temperature of the flywheel material, as obtained from DWT performed in accordance h the specification ASTM E 208 was no higher than -30F.
Charpy V-notch (Cv) upper-shelf energy level in the weak direction of the flywheel erial was at least 50 ft-lbs. A minimum of three Cv specimens were tested from each forging in ordance with ASTM A 370.
mechanical properties of the flywheel material as certified by the steel vendors are as ows:
chanical Properties Tensile 0.2% Offset Elongation Heat Charpy V-notch at 20F Drop Weight Strength Yield Strength in 2 inches umber Energy (ft-lbf) Test at -30F (psi) (psi) (%)
0647 110,000 93,700 21.7 130 126 92.9 No break 5100 112,000 96,000 22.4 140 130 122 No break cture Toughness minimum static fracture toughness of the material at the normal operating temperature of the heel was equivalent to a critical stress intensity factor, KIC, of at least 150-ksi in.
mpliance was demonstrated by either of the following:
- a. Testing of the actual material to establish the KIC value at the normal operating temperature.
- b. Determining that the normal operating temperature is at least 100F above the RTNDT.
ervice inspection includes 100 percent ultrasonic examination of the flywheel of one (1) of the (4) RCPs during each inspection interval. The acceptance criteria is in accordance with ME - Boiler and Pressure Vessel Code,Section III, for Class I vessels.
RCS piping consists of two loops which connect the steam generators to the reactor vessel.
h loop can be considered to consist of 42 inch ID hot leg piping connecting the reactor sel outlets to the steam generator inlets and 30 inch ID piping connecting the steam generator ets to the RCPs and the coolant pumps to the reactor vessel inlets. The two 30 inch piping ments are referred to as the pump suction leg and the cold leg respectively. A 12 inch edule 160 surge line connects one loop hot leg to the pressurizer. Design parameters for the tor coolant piping are given in Table 4.3-5.
reactor coolant piping is designed and fabricated in accordance with the rules and procedures NSI B31.7, Class I. The anticipated transients listed in Section 4.2.1 form the basis for the uired fatigue analysis to ensure an adequate usage factor.
reactor coolant piping is fabricated from SA 516 Gr 70 carbon steel mill clad internally with bonded type 304L stainless steel. A minimum clad thickness of one-eighth inch is maintained.
12 inch surge line is fabricated from ASTM A351 Gr CF8M alloy steel.
rmal sleeves are installed in the surge line nozzle, charging nozzles and shutdown cooling t nozzle to reduce thermal shock effects from auxiliary system. Clad sections of piping are d, where necessary, with safe ends for field welding to stainless steel components.
esponse to industry experience, half nozzle replacements have been performed on selected ruments and sampling nozzles as a mitigation technique against pressurized water stress osion cracking (PWSCC).
piping is shop fabricated and shop welded into subassemblies to the greatest extent ticable to minimize the amount of field welding. Fabrication of piping and subassemblies is e by shop personnel experienced in making large heavy wall welds. Welding procedures and rations meet the requirements of Section IX of the ASME Boiler and Pressure Vessel Code.
welds are 100 percent radiographed and liquid-penetrant or magnetic-particle tested and all tor coolant piping penetrations are attached in accordance with the requirements of ANSI
.7. Cleanliness standards consistent with nuclear service are maintained during fabrication erection. There are no dissimilar metal field welds.
5 PRESSURIZER pressurizer maintains RCS operating pressure and compensates for changes in coolant ume during load changes. Table 4.3-6 gives design parameters for the pressurizer. The surizer is shown in Figure 4.3-7.
ssure is maintained by controlling the temperature of the saturated liquid volume in the surizer. At full load nominal conditions, slightly more than one-half the pressurizer volume is upied by saturated water, and the remainder by saturated steam. A number of the pressurizer ters are operated continuously to offset the heat losses and the continuous minimum spray,
ing load changes, the pressurizer limits pressure variations caused by expansion or contraction he reactor coolant. The average reactor coolant temperature is programmed to vary as a ction of load as shown in Figure 4.3-8. A reduction in load is followed by a decrease in the rage reactor coolant temperature to the programmed value for the lower power level. The lting contraction of the coolant lowers the pressurizer water level causing the reactor system sure to decrease. The pressure reduction is partially compensated by flashing of pressurizer er into steam. All pressurizer heaters are automatically energized on low system pressure, erating steam and further limiting pressure decrease. Should the water level in the pressurizer p sufficiently below its setpoint, the letdown control valves close to a minimum value and the ilable charging pumps in the chemical and volume control system (CVCS) are automatically ted to add coolant to the system and restore pressurizer level.
en steam demand is increased, the average reactor coolant temperature is raised in accordance h the coolant temperature program (Figure 4.3-8). The expanding coolant from the reactor lant piping hot leg enters the bottom of the pressurizer (in-surge), compressing the steam and ing system pressure. The increase in pressure is moderated by the condensation of steam ng compression and by the decrease in bulk temperature in the liquid phase. Should the sure increase be large enough, the pressurizer spray valves open, spraying coolant from the P discharge (cold leg) into the pressurizer steam space. The relatively cold spray water denses some of the steam in the steam space, limiting the system pressure increase. The grammed pressurizer water level is a power dependent function. A high level error signal duced by an in-surge causes the letdown control valves to open, releasing coolant to the CVCS restoring the pressurizer to the prescribed level.
all pressure and coolant volume variations are accommodated by the steam volume which orbs flow into the pressurizer and by the water volume which allows flow out of the surizer. The total volume of the pressurizer is determined by consideration of the following ors:
- 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 injection of concentrated boric acid by the charging pumps, the pressurizer is designed so that the minimum pressure observed during such transients is above the setpoint of the safety injection actuation signal (SIAS);
- b. The heaters should not be uncovered by the out-surge 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 during load change transients;
- 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 total coolant volume change and associated charging 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.
account for these factors and to provide adequate margin at all power levels, the water level in pressurizer is programmed as a function of average coolant temperature as shown in ure 4.3-9. High or low water level error signals result in the control actions shown in ure 4.3-10 and described above.
pressurizer heaters are single unit, direct immersion heaters which protrude vertically into the surizer through sleeves welded in the lower head. Each heater is internally restrained from h amplitude vibrations and can be individually removed for maintenance during plant tdown. Approximately 20 percent of the heaters are connected to proportional controllers ch adjust the heat input as required to account for steady state losses and to maintain the red steam pressure in the pressurizer. These heaters are separated into two banks proximately 160 kW each) and are provided with diverse vital power.
remaining heaters are connected to on-off controllers. These heaters, called backup heaters, normally deenergized but are turned on by a low pressurizer pressure signal or high level error al. This latter feature is provided since load increases result in an in surge of relatively cold lant into the pressurizer, decreasing the temperature of the water volume. The action of the CS restoring the level results in a pressure undershoot below the desired operating pressure.
minimize the pressure undershoot, the backup heaters are energized earlier in the transient, tributing more heat to the water before the low pressure setting is reached. An interlock will vent operation of the backup heaters if the high level error signal occurs concurrent with a high surizer pressure signal. A low-low pressurizer level signal deenergizes all heaters to prevent ter burnout.
pressurizer spray is supplied from each of the RCP discharges on one loop to the pressurizer y nozzle. Automatic spray control valves control the amount of spray as a function of surizer pressure; both of the spray control valves function in response to the signal from the troller. These components are sized to use the differential pressure between the pump harge and the pressurizer to pass the amount of spray required to prevent the pressurizer steam sure from opening the PORVs during normal load following transients. A small continuous is maintained through the spray lines at all times to keep the spray lines and the surge line m, reducing thermal shock during plant transients. This continuous flow also aids in keeping chemistry and boric acid concentration of the pressurizer water equal to that of the coolant in
he event of an abnormal transient which causes a sustained increase in pressurizer pressure, at te exceeding the control capacity of the spray, a high pressure trip level will be reached. This al trips the reactor and opens the two PORVs. The steam discharged by the relief valves is d to the quench tank where it is condensed. In accordance with Section III of the ASME ler and Pressure Vessel Code, the RCS is protected from overpressure by two spring-loaded ty valves. The discharge from the safety valves is also piped to the quench tank. See tion 4.3.7 for the safety valve design parameters.
pressurizer is supported by a cylindrical skirt welded to the lower head. Since the pressurizer e line has sufficient flexibility, no provisions are made for horizontal movement and the skirt olted rigidly to the floor.
pressurizer assembly was replaced in 2006 with a new pressurizer assembly fabricated from erials that are less susceptible to primary water stress corrosion cracking. The replacement surizer is fabricated and installed to the same design criteria as the original pressurizer with e improvements. The cylindrical shell sections, upper and lower heads including the large e nozzles of the replacement pressurizer are forged components, thereby minimizing the welds weld inspections. Safe ends made of stainless steel are provided as required on the large bore zles to facilitate field welds to the connecting piping. The interior surface of the pressurizer is with weld deposited stainless steel. The heater sleeves, instrument nozzles and the vent/pass zle are fabricated from stainless steel instead of the originally used Inconel Alloy 600 erial. The nozzles are attached to the pressurizer by j-groove welds to the clad buildup. The l number of heaters is reduced from 120 to 60. The heat output of each heater is roughly twice of the replaced heater, thus maintaining the same amount of total heat output.
x and one-half inch inside diameter vent port was added on the upper head of the replacement surizer. This vent port is a substitute for the removal of the pressurizer manway during routine eling outages with RCS nozzle dams installed. The vent port is sized to avoid a RCS surization that would exceed the design pressure of the RCS nozzle dams following a tulated loss of shutdown cooling with the reactor vessel head on and reactor coolant water ls one foot below the reactor vessel flange or lower. For non-routine outages where nozzle s are installed and reactor coolant water levels are higher than one foot below the reactor sel flange and the reactor vessel head is on, the pressurizer manway will need to be removed in er to avoid a RCS pressurization that would exceed the nozzle dam design pressure following ostulated loss of shutdown cooling. In addition, removal of either the vent port or the surizer manway provides an adequate RCS vent path for low temperature overpressure ection in Mode 5 and Mode 6 when the head is on the reactor vessel.
6 QUENCH TANK quench tank is designed to prevent the discharge of the pressurizer relief or safety valves m being discharged to the containment. The quench tank is shown in Figure 4.3-11. The steam harged into the quench tank from the pressurizer is discharged under water by a sparger to
d to accommodate the steam released as a result of a loss-of-load accident followed ediately by an uncontrolled rod withdrawal accident with no coolant letdown or pressurizer y.
water temperature rise in the quench tank is limited to 281F, assuming a maximum initial er temperature of 120F. The gas volume in the tank is sufficient to limit the maximum tank sure after the above steam release to 35 psia. The contents of the quench tank are cooled by rculation through the primary drain tank and quench tank cooler. The temperature of the water he quench tank is indicated on the main control console. A high temperature alarm is also vided. A high quench tank temperature alerts the operator to cool the tank contents.
easurement channel provides a quench tank pressure indication on the main control console actuates a high pressure alarm. High quench tank pressure indicates that the tank has received scharge from the safety or relief valves, or from the HPSI test line relief valve. The operator ld then take action to restore the tank to normal operating conditions.
nch tank level indication and high and low level alarms are also provided on the main control sole.
quench tank can condense the steam discharged during a loss-of-load accident as described in tion 14.2 without exceeding the rupture disc set point, which is rated for 96 psig at 72F and psig at 350F, assuming normal closing of the safety valves at the end of the accident. It is not gned to accept a continuous uncontrolled safety valve discharge. The rupture disc on the nch tank provides code overpressure protection of the tank. The rupture disc vents to the tainment. The quench tank parameters are given in Table 4.3-7.
tank normally contains demineralized water under a nitrogen overpressure. The sparger, y header, nozzles and rupture disc fittings are stainless steel. The tank is designed and icated in accordance with the ASME Boiler and Pressure Vessel Code,Section III, Class C.
quench tank is located at a level lower than the pressurizer. This ensures that any PORV or surizer safety valve leakage from the pressurizer, or any discharge from these valves, drains he quench tank.
7 VALVES design parameters for the pressurizer spray valves (RC-100E, RC-100F) are given in le 4.3-8. The PORV isolation valve (RC-403, RC-405) parameters are given in Table 4.3-9.
position of each valve on loss of actuating signal (failure position) is selected to ensure safe ration. System redundancy is considered when specifying the failure position of any given
- e. Valve position indication is provided at the main control panel.
reduce stem leakage during normal operation or when closed.
two PORVs, designated RC-402 and RC-404 on Figure 4.1-1, relieve sufficient pressure to id opening of the RCS safety valves. The relief valves are actuated by the high RCS pressure signal. Parameters for these valves are given in Table 4.3-10.
valves are solenoid-operated power relief valves. The two half capacity valves are located in llel pipes which are connected to the pressurizer relief valve nozzle in the inlet side and to the ef line piping to the quench tank on the outlet side. A motor-actuated isolation valve is vided upstream of each of the relief valves to permit isolating the valve for maintenance or in of valve leakage.
capacity of the PORVs is sufficient to pass the maximum steam surge associated with a tinuous control rod withdrawal accident starting from low power. Assuming that a reactor trip ffected on a high pressure signal, the capacity of the PORVs is sufficient so that the safety es do not open. The relief valve capacity is also large enough so that the safety valves do not n during a loss-of-load accident from full power. This assumes normal operation of the surizer spray system, and reactor trip on high pressure. The PORVs also function to provide temperature overpressurization protection (LTOP) to the RCS. This is accomplished by ually selecting a reduced valve setpoint as described in Section 7.4.8.
o safety valves, designated RC-200 and RC-201 on Figure 4.1-1 are located on the pressurizer provide overpressure protection for the RCS. They are totally enclosed, backpressure pensated, spring-loaded safety valves meeting ASME Code requirements. Parameters for e valves are given in Table 4.3-11.
h safety valves (RC-200, 201) are equipped with acoustic valve position monitors. These nitors will provide operators with an indication of the status of the safety valves. The PORVs e been upgraded to monitor their position from the control room by the position indication ts instead of the acoustic monitoring method previously used. The valve monitoring system forms with NUREG-0578 C lessons learned from TMI Task Force Report.
safety valves pass sufficient pressurizer steam to limit the RCS pressure to 110 percent of gn (2,735 psig), following a complete loss of turbine generator load without simultaneous tor trip. Reactor trip occurs on a high RCS pressure signal. To determine the maximum steam
, the only other pressure relieving system assumed operational is the steam system safety es. Conservative values for all system parameters, delay times, and core moderator fficient are assumed. This analysis is given in Section 14.2, Loss-of-Load External Electrical d and/or Turbine Trip.
mounting of pressure relieving devices (safety valves and relief valves) within the RCPB and he main steam lines outside of the containment is in accordance with the applicable provisions SME Boiler and Pressure Vessel Code Section III.
es are relieving are included in the specification for the pressurizer.
overpressure relief valves and their connected piping (i.e., headers, header connections and harge piping) are designed to withstand the following conditions without exceeding the licable codes primary stress allowable: maximum loads due to valve discharge thrust, internal sure, dead weight and earthquake applied simultaneously. When more than one relief valve is ched to a piping system, the loads due to all relief valves discharging simultaneously are lied to the system along with the above mentioned primary loads. In addition, the loads from most critical combination of valves discharging are applied. The local stresses in the main m line outside the containment at the connection of the relief valves were computed as cified in Welding Research Council Bulletin 107 and contained below the allowable primary ss level.
pressurizer safety and relief valve discharge piping system was modified by deleting the loop s upstream of the Pressurizer Safety Valves (PSVs) and Power-Operated Relief Valves RVs). The piping for both the PSVs and PORVs was again modified during the replacement he pressurizer. The PORVs were also replaced with upgraded PORVs and credited as an eptable pressurizer steam space vent path during post accident conditions. The new piping figuration was evaluated to address NUREG-0737, Action Item II.D.1 (relief valve and safety e testing).
thermal-hydraulic analysis was performed to calculate the transient fluid time-history forcing ctions acting on the pipe segments due to each safety or relief valve discharge. These forcing ctions are combined in a common entry mode to maximize the forces acting on the discharge ng. The potential case of water discharge through relief valves during low temperature modes peration was also considered.
structural reanalysis of the PSV and PORV discharge piping system included the normal t loading, plant transients described in Section 4.2.1, seismic loading, and the fluid transient ing functions. Seismic analysis of the piping system was performed by the modal analysis onse spectra method. Dynamic response of the piping system to the PSV and PORV discharge s was performed by the time-history modal superposition method.
e primary and secondary stress intensities and fatigue usage factors were found to be within Code allowable values. The pipe supports of the PSV and PORV discharge piping system e modified to accommodate the load and displacements resulting from the reanalysis.
main steam relief valve system is designed so that the blowdown force is transmitted directly he structure by mechanical/structural devices, and not through the piping. These relief valves provided with discharging stacks to direct steam blowdown to atmosphere. Stacks are gned so that backpressure does not result in a valve reaction force.
ps and valves within the RCPB are classified as either active or inactive components. Active ponents are those whose operability is relied upon to perform a safety function, as well as
ure, pump operation or trip) are not relied upon to perform the system function during the sients or events considered in the respective operating condition categories. Thus, certain ps and valves (classified as active components) within the RCPB are required not only to e as pressure-retaining components (as in the case of passive components such as vessel and ng) but also to operate reliably to perform a safety function such as safe shutdown of the tor and mitigation of the consequences of a pipe break accident under the loading binations considered in RCPB as either active or inactive. The only pumps within the RCPB the RCPs which are inactive under faulted conditions.
ive components outside of the RCPB have been designed to function as required when jected to the loadings and the maximum pressure and temperature occurring under normal and dent conditions in the areas in which the components are located. Anticipated temperature and sure transients which would have an effect on operability of components were specified as gn requirements.
omplete list of the active pumps and valves located within the RCPB was provided previously art of Amendment 14, and is found as Table 4.3-12.
following is a list of active pumps and valves located outside the RCPB whose operability is ed upon to perform a safety function such as safe shutdown of the reactor or mitigation of the sequences of a postulated pipe break in the RCPB.
ps Post Normal LOCA Component Quantity Materials Code Test Code Position Position gh Pressure 3 Draft ASME Code ASI Standard 610, Off On fety Injection for Pumps and ASME Power Test Valves for Nuclear Code PTC- 8.2 and Power, Class II, Standards of the Nov. 1968 Hydraulic Institute w Pressure 2 Draft ASME Code ASI Standard 610 Off On fety Injection for Pumps and Standard of the Valves for Nuclear Hydraulic Institute, Power, Class II, and ASME Power Nov. 1968 Test Code PTC-8.2
Post Normal LOCA Component Quantity Materials Code Test Code Position Position ntainment 2 Draft ASME Code Standard of the Off On ray Pumps for Pumps and Hydraulic Institute Valves for Nuclear Power, Class II, 1968 ves Normal Post LOCA Component Quantity Materials Code Test Code Position Position gh Pressure 2 Draft ASME Code Closed Open fety Injection for Pumps and mp Suction Valves for Nuclear heck) Power, Nov. 1968 gh Pressure 4 Draft ASME Code Closed Open fety Injection for Pumps and mp Discharge Valves for Nuclear heck) Power, Nov. 1968 w Pressure 2 Draft ASME Code Closed Open fety Injection for Pumps and mp Discharge Valves for Nuclear heck) Power, Nov. 1968 ntainment - Draft ASME Code - As As Required lation Valves for Pumps and Required Valves for Nuclear Power, Nov. 1968 ntainment 2 Draft ASME Code Closed Open ray (Motor for Pumps and erated) Valves for Nuclear Power, Nov. 1968 ntainment 2 Draft ASME Code Closed Open mp Discharge for Pumps and heck) Valves for Nuclear Power, Nov. 1968
Normal Post LOCA Component Quantity Materials Code Test Code Position Position fueling Water 2 Draft ASME Code Closed Open during rage Tank for Pumps and Injection Mode.
scharge Valves for Nuclear Closed during heck) Power, Nov. 1968 Recirculation Mode.
utdown 1 Draft ASME Code Open Open oling Heat for Pumps and changer Valves for Nuclear pass (2-SI- Power, Nov. 1968
- 6) (Air erated) utdown 1 Draft ASME Code Closed Closed oling Heat for Pumps and changer Flow Valves for Nuclear ntrol Power, Nov. 1968 SI-657) ir Operated) inactive valves and pumps other than the RCPs, the rules of the Pump and Valve Code (March 0 Draft) for design conditions are applied in evaluating the loadings produced by the ergency and Faulted Conditions.
active components, additional requirements are imposed on the design to assure operability ng the faulted operating conditions. As appropriate, these additional requirements consist of ulated tests and/or supplementary calculations which demonstrate that the active component perform its required function during the specified conditions. Where calculations are loyed, the primary stresses produced by the faulted conditions are limited to values less than Emergency Condition limits of Subparagraph HB-3224.1 in all regions of the active ponent where deformations may impair the required function.
the RCPs, as appropriate, the stress criteria for Vessels, as discussed above, are applied in luating the Emergency and Faulted Conditions (elastic analysis). These limits are consistent h the Design by Analysis method of Article 4, ASME Section III Code which is applied in design calculation. For design conditions other than those explicitly addressed by Section III he ASME Code, and where design calculations are used to evaluate stresses and deformations umps, the methods and criteria applied are in accordance with Article 4, ASME Section III
- e. The calculations will include the effects of gross and local structural discontinuities and the ings produced by geometrical eccentricities. Current state-of-the-art analytical methods uding finite element techniques, are employed in the calculations. Experimental techniques e not employed.
measurement channels necessary for operational control and protection of the RCS are cribed below. A brief explanation of the purpose of each measurement channel is made and a mary of resulting action is given. A detailed description of critical instrument channel actions be found in Chapter 7.
r independent measurement channels are provided for each parameter which initiates ective system action. Two independent signals are required to initiate protective action, eby, preventing spurious actions resulting from the failure of one measurement channel. This ngement results in a high degree of protective measurement channel reliability in terms of ating action when required and avoiding unnecessary action from spurious signals. Two pendent measurement channels are provided for parameters which are critical to operational trol. These control channels are separate from the protective measurement channels. To avoid trol conflicts, control action is derived from only one channel at any time, while the second nnel serves as a backup. This allows continued operation of the facility if one channel fails and mits maintenance on the failed channel during operation. This arrangement results in increased ilability.
8.1 Temperature 8.1.1 Hot Leg Temperature h of the two hot legs contains five narrow range channels to measure coolant temperature ing the reactor vessel. Four of these channels are used to furnish a hot leg temperature signal he reactor protective system. The fifth hot leg temperature measurement channel provides a al to the average temperature computer which is a part of the reactor regulating system (RRS).
average temperature of the loop is recorded on a two-channel recorder in the main control
- m. The second channel on each loops recorder records an average temperature reference al received from the RRS.
igh temperature alarm is provided on this channel to alert the operator to a high temperature dition. The temperature from this measurement channel is indicated in the main control room ddition to being recorded. The other hot leg temperature channels are also displayed in the n control room.
8.1.2 Cold Leg Temperature h of the four cold legs contains three temperature measurement channels. The cold leg istance Temperature Detectors (RTD) are located downstream of the RCPs. Two channels m each cold leg (four per heat transfer loop) are used to furnish a cold leg coolant temperature al to the reactor protective system (RPS). All eight of these cold leg temperature surements are indicated in the main control room. Two of the remaining four cold leg perature measurement channels, one each from opposite loops, are narrow range temperature surements that input to the RRS for calculation of average temperature and the Feedwater ulating system for dynamic compensation in single element control. These loops also provide
ut to the subcooled margin monitor (Inadequate Core Cooling ICC) and Low Temperature r Pressure (LTOP) channels. These loops also provide control indication and recording of e range temperature.
8.1.3 Surge Line Temperature s measurement channel provides an indication of surge line temperature in the main control
- m. A low surge line temperature condition activates an alarm in the main control room. The
-temperature alarm during normal operation is an indication that the continuous spray rate has reased.
8.1.4 Pressurizer Vapor Phase Temperature pressurizer vapor phase RTD is located on the upper dome of the pressurizer. This channel vides a wide range temperature indication of the temperature of the steam phase in the surizer.
8.1.5 Pressurizer Water Phase Temperature pressurizer water phase RTD is located at an elevation below the top of the pressurizer ters. This channel provides a wide range temperature indication in the main control room and sed during plant heatup and cooldown.
8.1.6 Spray Line Temperature RTD in each spray line provides a temperature indication and a low temperature alarm in the trol room for each spray line. A low temperature alarm during normal operation is an cation the continuous spray rate has decreased.
8.1.7 Relief and Safety Valve Discharge Temperature peratures in the pressurizer safety valve and PORV discharge lines are measured and cated in the main control room. A high temperature in one of these lines is an indication that associated valve may be leaking. High temperature alarms are provided to alert the operator to condition.
8.1.8 Quench Tank Temperatures temperature of the water in the quench tank is indicated in the main control room. A high perature alarm is also provided. A high quench tank temperature alerts the operator to the uirement for cooling of the tank contents.
s RTD is located in the reactor vessel flange leak-off line. The channel is displayed in the main trol room and actuates a high temperature alarm. A high temperature is indicative of excessive age past the first reactor vessel flange seal.
8.1.10 RCS High Point Vents Leakage Temperature rmocouple installed on the downstream side of solenoid valve train of the reactor vessel head t system is utilized to monitor leakage past the system solenoid valves. Under normal rating conditions, the thermocouple will measure the ambient temperature in the piping nstream of the solenoid valve train. The output of each thermocouple is continuously rded by the plant computer. Note that monitoring of the leakage past the PORVs in the surizer steam space vent path is described in Section 4.3.8.1.7, above.
leakage through the system valves will cause an increased temperature in the downstream ng which will be detected by the thermocouple. At a predetermined setpoint, an alarm will be ated, identifying a high temperature reading on the appropriate thermocouple. Once a high perature alarm is received, further actions will be governed by the Technical Specifications for tor coolant system leakage.
8.2 Pressure 8.2.1 Pressurizer Pressure r independent narrow range pressure channels are provided for initiation of protective systems on. The pressure transmitters are connected to the upper portion of the pressurizer via the er level measurement nozzles and measure pressurizer vapor pressure. All four channels are cated in the main control room and actuate separate high, low, or low low pressure alarms in control room.
protection actions these pressure signals initiate are:
- 1. Reactor trip on high primary system pressure. The reactor 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 hot and cold legs. The variable set point has high and low limits alarmed in the control room and is not allowed to decrease below 1865 psia.
o independent pressure channels provide narrow range pressure signals for controlling the surizer heaters and spray valves. The output of one of these channels is manually selected to orm the control function. During normal operation, a small group of heaters are portionally controlled to offset heat losses. If the pressure falls below a low pressure set point, of the heaters are energized. If the pressure increased above the high pressure set point, the y valves are proportionally opened to increase the spray flow rate as pressure rises. An rlock will prevent operation of the backup heaters in the event of a high level error signal current with a high pressure condition. These two channels are also used to provide surizer pressure signals to the RRS. The two channels are continuously recorded in the main trol room and are provided with high and low pressure alarms.
8.2.3 Pressurizer Pressure o low-range pressure measurement channels provide a control room indication of RCS sure during plant startup and shutdown in the main control room. They also provide ependent pressure signals to the shutdown cooling suction isolation valves (refer to tion 9.3.4.1) which prevent these valves from opening above a selected set point. If the tdown cooling suction valves are open when the pressure exceeds a selected set point (280
), an annunciator receives a signal to alarm from these pressure channels. The channels also vide signals to actuate the PORVs for LTOP. These two instrument channels are independent redundant.
8.2.4 Quench Tank Pressure s measurement channel provides a quench tank pressure indication in the main control room actuates a high pressure alarm. High quench tank pressure indicates that the tank has received scharge from the safety or relief valves, or from the HPSI test line relief valve. The operator then take action to restore the tank to normal operating conditions.
8.3 Level 8.3.1 Pressurizer Level o pressurizer level channels are used to provide two independent level signals for control of the surizer liquid level. These signals are used to deenergize the pressurizer heaters on low low surizer level to prevent heater burn out, provide input to one channel in the two-channel rder in the control room, and actuate high and low pressurizer level alarms in the main control
- m. The second channel on the level recorder records the programmed pressurizer level setpoint puted by the RRS as a function of the average reactor coolant temperature. The level smitters are compensated for the steam and water densities existing in the pressurizer during mal operation.
liquid level in the pressurizer is programmed to vary as a function of average reactor coolant perature. This level set point is computed by the RRS and furnished to controls associated
sts the CVCS charging or letdown flow rates to make the difference zero. Each of the level nnels is indicated in the main control room and is equipped with a high and low level alarm in control room.
8.3.2 Pressurizer Level wide range pressurizer channel is provided for main control room indication of pressurizer l during plant startup and shutdown.
8.3.3 Quench Tank Level uench tank level channel indicates quench tank level in the main control room. The transmitter activates high and low level alarms in the main control room.
8.4 Reactor Coolant Loop Flow independent differential pressure measurement channels are provided in each heat transfer p to measure the pressure drop across the steam generators. Four pressure taps are located in h hot leg piping section just before the elbow entering the steam generator and four pressure zles are located in the steam generator outlet plenum. Four differential pressure transmitters connected between the four hot leg nozzles and the steam generator nozzles, resulting in four m generator differential pressures.
outputs of the transmitters are sent to four analog summing devices in the low total flow trip
- c. Each summing device receives two differential pressure signals with the summation of e signals representing the total core flow at all times, even during coast down transients.
summing devices provide four independent total flow signals. The four signals are indicated arately in the main control room and activate separate low-flow alarms. In the RPS, they are pared with the low-flow reactor trip set point, selected by the operator to correspond with the ber of operating RCPs. If two channels indicate flow which is less than the flow set point, the tor is tripped.
8.5 Reactor Coolant Pump Instrumentation RCPs and motors are equipped with the instrumentation necessary for proper operation and arn of incipient failures. (See Figure 4.3-4.) A description of the major channels follows:
8.5.1 Pump Seal Temperatures reactor coolant temperature in the lower seal cavity may be indicated in the main control m through its selection on a multiposition switch. The switch also permits display of all of the aining temperature measurement channels for each pump. The pump seal temperature is med to alert the operators to a high-temperature condition. A high temperature condition is an
eased.
8.5.2 Motor Stator Temperatures h RCP motor is provided with six RTDs embedded in the stator windings. During initial p testing the highest reading RTD was selected for this temperature measurement control.
signal output of this RTD may be selected for indication in the main control room by a tiposition switch. Should stator temperature exceed a predetermined limit, a high temperature m will be sounded in the control room. High temperature is detrimental to motor winding lation life, and may be caused by high ambient temperature, reduction in the cooling air flow he stator or inadequate time delay between successive starts of the motor.
8.5.3 Motor Thrust Bearing Temperatures peratures of the motor upward thrust bearing, and downward thrust bearing may be indicated he main control room through selection on a multiposition switch. A high temperature alarm is vided for each pump in the main control room which is annunciated if any one of the bearing peratures in the pump exceeds a safe value.
8.5.4 Pump Controlled Bleed-Off Temperature temperature of the controlled bleed-off flow may be displayed in the main control room ugh its selection on a multiposition switch. An alarm signal is provided should the controlled d off temperature exceed a high limit. A high temperature condition is an indication that the gral heat exchanger is not operating properly.
8.5.5 Antireverse Device Bearing Temperature s measurement channel provides a status on the operating temperature of the antireverse ice bearing. Display in the main control room is available through selection on a multiposition tch. A high temperature alarm is provided to alert plant operators to an abnormal condition.
8.5.6 Upper and Lower Guide Bearing Temperature upper and lower guide bearing temperatures may be displayed in the main control room at the on of the operators. An alarm signal is provided if the high temperature limit is exceeded.
8.5.7 Lube Oil Cooler Inlet and Outlet Temperature inlet and outlet temperatures of the lube oil cooler are available for display in the main trol room and are alarmed to alert plant operators to high temperature conditions.
lower bearing oil temperature may be displayed in the main control room at the option of the t operators.
8.5.9 Pump Seal Pressures middle, upper, and controlled bleed-off pump seal cavities in each pump are provided with sure sensors which generate a signal proportional to the pressure within the cavity. The sure in any seal may be selected for display in the main control room through its selection on ree-position switch. A high and low pressure alarm is annunciated for the middle and upper measurement channels. Abnormally, high pressure in the upper and middle seal cavities cates a failed or failing lower or middle seal. A low pressure condition in the middle seal ity indicates a failed or failing upper seal. A recorder in the control room records seal sures in order to recognize pump seal degradation.
8.5.10 Motor Oil Lift Pressure s pressure measurement channel provides a signal to the RCP control circuit to prevent motor t with insufficient motor thrust bearing oil pressure.
8.5.11 Lube Oil Filter Pressure Differential re are three lube oil filters, each with differential pressure indication capability. These surement channels provide alarms for high differential pressure across any of the filters, cating clogging.
8.5.12 Pump Controlled Bleed-Off Flow w instruments are used to measure the controlled bleed off flow from the pump upper seal ity to the chemical and volume control system. These instruments provide a remote indication he flow rate in the East and West electrical penetration rooms, and annunciate high and low alarms in the control room. There is also an input to the plant computer and a recorder in the trol room which records flow.
8.5.13 Lube Oil and Antireverse Device Lube Oil Flow Switch se measurement channels provide alarm signals when a low flow condition exists in the thrust ring lube oil loop or in the antireverse device lube oil flow.
8.5.14 Motor Oil Reservoir Level at-type sensors or differential pressure transmitters are used to produce signals proportional to oil levels in the upper and lower motor oil reservoirs. These signals are used to provide an cation on the main control panel of the oil level in each oil reservoir. Either oil reservoir level
8.5.15 Vibration Instrumentation tor vibration is sensed by two velomitor probes attached to the upper motor frame. Excessive or vibration will cause an alarm on the plant process computer.
p shaft orbit is monitored by two proximity probes mounted 90 degrees to each other in a zontal plane. A separate key phasor probe provides a single pulse per revolution as a reference determining angular changes in shaft orbit. Outputs from the probes are also recorded.
8.5.16 Reverse Rotation Switch erse rotation of an RCP is sensed by a reverse rotation switch. This switch actuates an alarm he control room. Reverse rotation indicates failure of the mechanical anti-reverse device.
8.5.17 (Deleted) 8.5.18 RCP Underspeed Reactor Trip RCP underspeed trip has been added which uses a speed sensor on the RCP motor. The sensor erates a frequency signal proportional to speed which is in turn converted to a voltage. This age signal is maintained by the RPS, with one pump signal being monitored by only one nnel in the RPS. When any two pump speeds drop below 830 RPM a reactor trip will occur.
s trip is designed to protect against simultaneous loss of all pumps and the resulting loss of lant flow, with a faster reaction or trip time, than is provided by the steam generator erential pressure signal.
9 REACTOR COOLANT VENTING SYSTEM RCS venting system provides the capability for removing noncondensible gases collected in system in order to allow for satisfactory long term core cooling.
two important safety functions enhanced by this venting capability are core cooling and tainment integrity. For events within the present design basis for nuclear power plants, the ability to vent noncondensible gases will provide additional assurance that the requirements of CFR 50.44 will be met. For events beyond the design basis, this venting capability will stantially increase the ability to deal with large quantities of noncondensible gas.
ctor Vessel Head Vent System reactor coolant head vent system is equipped with two (2) solenoid-operated globe valves in es in each piping train. Each valve has remote-manual control capability from the control m with open and closed position indication. Provision of two (2) solenoid operated globe es in series for each vent train minimizes the probability of a vent path failing to close, once
m redundant independent AC emergency buses. All valves fail closed upon loss of power ply to the actuator. The solenoid vent valves are qualified to IEEE-344-1975.
anual isolation valve is installed downstream of the solenoid operated vent valves. This valve ormally open and is intended to isolate leakage that may develop through the solenoid-rated valves. Also, a manual isolation valve is located just upstream of the two solenoid valves ach train of the RCS head vent piping, capable of isolating the affected train. Piping upstream he valve is designed to reactor coolant system pressure rating, providing an at-power isolation ability.
discharge sparger for the reactor coolant head vent system is located in the vicinity of the tainment air coolers where the vented gases will be cooled, mixed with additional containment and discharged into the lower elevation of the containment. Uniform mixing of the tainment post accident atmosphere is provided by the post-accident recirculation system. The S vent system piping has been analyzed in accordance with ASME,Section III, of the Boiler Pressure Vessel Code, for the Class 1 portion of the system. The balance of the system has n analyzed pursuant to ANSI B31.1 Power Piping Code.
ssurizer Steam Space Vent System ting of the noncondensible gases can also be performed from the pressurizer. The pressurizer ting system consists of two flow paths with redundant isolation valves in each flow path. Each path has a power operated relief valve and a block valve. The block valve in each train is mally open. During the pressurizer replacement project the two (2) power operated relief es were upgraded so they can be credited to perform the venting function post accident. The RVs are EEQ qualified and receive power from battery backed independent vital DC buses.
noncondensibles are vented via the PORV discharge piping into the quench tank and then to waste gas header as necessary. If the pressure reaches in excess of the rupture disc relief sure, the noncondensibles would be released into the containment. The PORV inlet and harge piping has been analyzed in accordance with ASME,Section III, of the Boiler and ssure Vessel Code, Class 1 and Class 2 and 3 requirements respectively.
reactor vessel and pressurizer vents were designed to utilize existing penetrations within each sel. The system can pass in excess of the gas volume equivalent to one-half the RCS volume in (1) hour. Although the RCS vents are larger than the size corresponding to the definition of CA (10 CFR 50, Appendix A), consequences of ruptures of the vents are bounded by the lts of current small break loss of coolant accident (SBLOCA) analyses.
ce the vent lines are sized larger than the size corresponding to the definition of LOCA, the em is equipped with two (2) solenoid-operated globe valves in series in each piping train.
h valve has remote-manual control capability from the control room with open and closed ition indication. A manual isolation valve is installed downstream of the solenoid operated t valves on each of the reactor and pressurizer vent systems. This valve is normally open and is nded to isolate leakage that may develop through the solenoid operated valves. Also, a manual
tor coolant system pressure rating, providing an at-power isolation capability.
design of the venting system minimizes the probability of a vent path failing to close, once ned. This has been accomplished by providing two (2) solenoid operated globe valves in series each vent train. The power source for each valve train is an independent redundant DC rgency bus, energized from separate redundant battery systems. In addition, the valves also ive power from redundant independent AC emergency buses. All valves fail closed upon loss ower supply to the actuator.
discharge sparger for the RCS vent system is located in the vicinity of the containment air lers where the vented gases will be cooled, mixed with additional containment air, and harged into the lower elevation of the containment. Uniform mixing of the containment post-dent atmosphere is provided by the post-accident recirculation system.
RCS vent system piping has been analyzed in accordance with ASME,Section III, of the ler and Pressure Vessel Code, for the Class 1 portion of the system. The balance of the system been analyzed pursuant to ANSI B31.1 Power Piping Code. The solenoid vent valves are lified to IEEE-344-1975.
10 PERMANENT REACTOR CAVITY SEAL permanent reactor cavity seal is designed to contain refueling water in the refueling cavity for shuffle during outages. The reactor cavity seal consists of seal membrane sub-assembly and port structure sub-assembly. The seal membrane sub-assembly consists of a stainless steel mbrane, inner and outer legs attached to the reactor vessel seal ledge and the embedment ring circumferential fillet welds all around the reactor vessel. The support structure assembly ctions as the load bearing assembly consisting of radial members around the annulus, resting the reactor vessel flange and the embedment ring. The reactor cavity seal is shown in ure 4.3-12.
reactor cavity seal has multiple openings, which provide access to the ex-core nuclear rumentation as well as ventilation for the reactor cavity cooling air. These openings have ged cover plates, which are closed for flooding the transfer canal during refueling. O-rings are alled to provide watertight seal to prevent leakage into the reactor cavity. Prior to reactor ration, the hinged cover plates will be laid back to allow for ventilation of cooling air from the tor cavity.
cavity seal is designed to withstand the pressure loads from the refueling water as well as the ions of the reactor vessel and containment building due to seismic displacements. It can also ommodate axial and radial growth from the normal and transient thermal conditions of the tor vessel. The cavity seal is also designed and tested to withstand loads imposed by a pped fuel assembly. The cavity seal is designed for 80 heat up/cool down cycles and 50 cycles aximum allowed water static head.
gn of reactor cavity seal and the neutron shielding in accordance with Reference 3.A-29.
support structure is fabricated from 300 series stainless steel conforming to ASME Section II A. The seal membrane is allowed to deflect such that it can rest on the support structure ng flooding. The cavity seal membrane, although not required, is designed and analyzed to the delines of ASME B&PV Code Section III, Appendix XIII and ASME Section II, Parts A and 995 Edition, through 1996 Addenda.
ign Pressure, psig 2485 ign Temperature, F 650 zles 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 ension 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, Replacement Closure Head, inches nominal 0.25 Cladding Thickness, Replacement Closure Head, inches minimum after machining 1/8 Cladding Thickness, Bottom Head, minimum 3/16 Cladding Thickness, Remainder of vessel, minimum, inches 1/8 erial 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 CEDM Nozzles Ni-Cr-Fe Alloy welded to SA-182, F316LN Instrumentation Nozzles Ni-Cr-Fe Alloy welded to SA-182, F316LN Weld deposited austenitic stainless steel with a composition approximately equivalent to SA-
, type 304 in contact with coolant.
Weights Closure Head lb. 147,900 Vessel, without flow skirt, lb. 682,000 Studs, Nuts and Washers, lb. 38,900 Total lb. 878,900 umes Bottom of Core, ft3 1113.11 Center of Core, ft3 1680.93 Top of Core, ft3 2248.74 Vessel, ft3 4651.27
mber 2 e Vertical U-Tube mber of Tubes (Design/Actual ) 8523 e Outside Diameter, inches 0.750 t Transfer Rate, each, Btu/hr 4.63 x 109 zles 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 mary Side Design Design Pressure, psig 2485 Design Temperature, F 650 Design Thermal Power (NSSS), MWt 2715 Coolant Flow (Each), lb/hr 74 x 106 Normal Operating Pressure, psia 2250 Coolant Volume, each, ft3 1693 ondary Side Design Design Pressure, psig 1000 Design Temperature, F 550 Normal Operating Steam Pressure, psia 880, Full Load Normal Operating Steam Temperature, F 530, Full Load Blowdown Flow, Design, Maximum, Each, lb/hr 112,000
Steam Flow, Each, lb/hr 5.9 x 106 Steam Moisture Content, Maximum, percent 2.0 Feedwater Temperature, F 435 Number of Steam Primary Separators, each Steam Generator 170 Number of Secondary Separators, each Steam Generator 170 ensions Overall Height, Including Support Skirt, inches 749 Upper Shell Outside Diameter, inches 239.75 Lower Shell Outside Diameter, inches 166 ghts Dry, lb. 1,070,400 Flooded, lb. 1,666,600 Operating, lb. 1,283,000
n Steam Piping Design Pressure, psig 1,100 (NOTE) n Steam Piping Design Temperature, F 600 d Saturated Steam Valve Number Set Pressure psia Capacity lb/hr (each valve) 2 1000 794,060 4 1005 794,060 6 1015 794,060 8 1025 794,060 10 1035 794,060 12 1045 794,060 14, 15, 16 1050 794,060 al Capacity (16 valves) lb/hr 12,704,960 erial Body ASTM A105 Gr II Disc ASTM 565 Gr 616 or ASME SB-637 UNS N07750 Type 3 Trim ASTM A451 Gr CPF8 TE: ASME Section III Code requires pressure rise to be limited to no more than 10%
above piping design pressure. Operability of the safety 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.
PARAMETERS chronous Speed, rpm 900 e Vertical, Limited Leakage, Centrifugal ft Seals Type N-9000 Assembly Stationary Face Ring Carbon-Morganite CNFJ Rotating Face Ring Tungsten or Silicon Carbide ign Pressure, psig 2,485 ign Temperature, F 650 mal Operating Pressure, psig 2,235 ximum Operating Temperature, F 549 ign Flow, gpm 81,200 al Dynamic Head, ft, Minimum 243 ximum Flow (one-pump Operation), gpm 120,000 Weight, lb. 168,050 oded Weight, lb. 175,050 ctor Coolant Volume, ft3 per pump 112 ft Material ASTM A-182 Type F-304 ing Material ASME SA-351 Gr CF8M ing Wear Ring Material ASTM A-351 Gr CF8 rostatic Bearing Bearing Material ASTM A-351 Gr CF8 Journal Material ASTM A-351 Gr CF8 tor Voltage, volts (source) 6,900 Voltage, volts (rated) 6,600 Frequency, Hz 60 Phase 3 Horsepower (rated)/rpm 6,500/887 Synchronous Speed, rpm 900
Instrumentation Quantity (per pump)
Seal Temperature 1 Pump Casing Differential Temperature 1 Seal Pressure 3 Controlled Bleed off Flow 1 Controlled Bleed off Temperature 1 Motor Oil Level 2 Motor Bearing Temperature 5 Motor Stator Temperature
- 6 Reverse Rotation Flow 1 Vibration Monitoring System *** 5 Oil Lift Pressure 1 Lubrication Oil Temperature 3 al 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 Guidelines, Final Report-December 1993 can be used to calculate these volumes more accurately at different conditions.
Six (6) resistance temperature detectors (RTDs) were imbedded into the stator (2 per phase) during fabrication. During first full load plant operation, all 6 RTDs were monitored. The RTD reading the highest temperature was chosen for use. The other five RTDs remain as spares.
Vibration Monitoring System consists of two (2) velomitors on the motor casing, and two (2) proximity probes and one (1) key phasor on the pump shaft.
mber of loops 2 w per loop, lb/hr 61 x 106 e Size Reactor outlet, ID, inches 42 Reactor inlet, ID, inches 30 Surge line, nominal, inches 12 ign Pressure, psig 2485 ign Temperature, F 650 ocity, Hot leg, ft/sec 40.4 ocity, Cold leg, ft/sec 36.3
ign Pressure, psig 2,485 ign Pressure, F 700 mal Operating Pressure, psia 2250 mal Operating Temperature, F 653 rnal Free Volume, ft3 1500 mal Operating Water Volume, Full Power, ft3 800 mal Steam Volume, Full Power, ft3 700 alled Heater Capacity, kW 1600 (Note: Total Heater Capacity may be less due to Heater unavailability) ay Flow, Maximum, gpm 375 ay Flow, Continuous, gpm 1.5 zles Surge Line (1) nominal, inches 12 Safety Valves (2) nominal, inches 4 Relief Valve (1) nominal, inches 4 Spray (1) nominal, inches 4 Heater Sleeve (60) ID, inches 0.905 nual Vent (1) nominal inches Manway nominal, inches 16 Alternate Vent Port nominal, inches 6.5 rument Nozzles Level (4) nominal, inches 1 Temperature (2) nominal, inches 1 Pressure (2) nominal, inches 1 erials Vessel SA-508 Grade 3, Class 2 Cladding - Cylinder Shell, Upper and Lower Head Type 308 Stainless Steel (1)
Weld deposited austenitic stainless steel in contact with coolant.
ensions 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 Weight, Including Heaters, lb. 202,731 oded Weight, Including Heaters, lb. 297,433
ign Pressure, psig 100 INT/15 EXT ign Temperature, F 350 mal Operating Pressure, psig 3 mal Operating Temperature, F 120 rnal Volume, ft3 217 mal Water Volume, ft3 135 mal Gas Volume, ft3 82 nket Gas Nitrogen zles Pressurizer 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 Instrument (1) inch, nominal 1 Vent (1), inch., nominal 1.5 sel Material ASTM-A-240 TP 304 ensions Overall Length, inches 145.5 Outside Diameter, inches 60 Weight, lb. 4600 oded Weight, lb. 18,120
vice - Pressurizer Spray Control ign Temperature, F 650 ign Pressure, psig 2485 w, gpm 440 ssure Drop, psi 8.5 - 40 ure Position Failed Closed nufacturer Fisher Controls Co.
ign Code Pump and Valve Code, Nov. 1968 Draft, Class I mic Class I erials Body 316 SST ASTM A351-CF8M TABLE 4.3-9 POWER-OPERATED RELIEF VALVE ISOLATION VALVE PARAMETERS (RC-403, RC-405) vice - Pressurizer Power-Operated Relief Valve Isolation ign Temperature, F 675 ign Pressure, psig 2,485 uator Electric Motor ure Position As Is SI Class 2,500 lb nufacturer Velan Valve Company ign Code Pump and Valve Code Nov. 1968 Draft, Class I mic Class I erials Body ASTM A182 Grade F316
(RC-402, RC-404) ign Pressure, psig 2485 ign Temperature, F 675 d Saturated Steam 0.1% (wt) Boric Acid mber 2 acity, lb/hr (minimum) each 153,000 e Solenoid Operated Pressure, psig 2,385 and 400 psig (low temperature overpressurization) ure Position Closed ign Code ASME Section III, 1977 Edition through Winter 1979 Addenda erials Body 316L SS, SA182
ign Pressure 2,485 ign Temperature, F 675 d Saturated Steam, 0.1% (wt) Boric Acid Pressure RC-200, psig 2,485 RC-201, psig 2,485 acity, lb/hr, at set pressure RC-200 294,000 RC-201 294,000 e Spring loaded-balances bellows, enclosed bonnet umulation, % 3 k pressure Compensation Yes wdown, % 12 ign Code ASME Section III, Class A, 1968 Edition, Addenda through Summer of 1970, Code Case 1344-1 erials Body 316 SST, ASTM A 182
SYSTEM BOUNDARY Classification Valve Type/ Active - A Post LOCA Line Number Inactive - I Normal Position Position utdown Cooling Motor / 2 A Closed Closed arging a Air / 2 A Open / Closed Open / Closed Check / 2 A Open / Closed Open / Closed tdown Air / 2 A Open Closed Manual / 1 I Open Open xiliary Spray Air / 1 I Closed Closed Check / 1 I Closed Closed ssurizer Spraya Air / 2 I Open / Closed Closed Manual / 6 I Open Open ssurizer Relief Motor / 2 I Open Open Solenoid / 2 I Closed Closed ssurizer Safety 2 I Closed Closed ety Injection Tank Motor / 4 I Open Open Check / 4 A Closed Open akage Control (SIS) Air / 4 I Closed Closed fety Injection Check / 8 A Closed Open SI Header Check / 4 A Closed Open Motor / 4 A Closed Open SI Header Check / 8 A Closed Open Motor / 8 A Throttled Open Open ain: Reactor Manual / 13 I Closed Closed olant Loop Air / 1 I Open Open arging Manual / 4 I Closed Closed Pressurizer Spray Manual / 4 I Closed Closed tdown Manual / 4 I Closed Closed Safety Injection Manual / 14 I Closed Closed Auxiliary Spray Manual / 4 I Closed Closed
Classification Valve Type/ Active - A Post LOCA Line Number Inactive - I Normal Position Position Shutdown Cooling Manual / 2 I Closed Closed nt / Test: Reactor Manual / 2 I Closed Closed ssel nt / Test: Manual / 2 I Closed Closed ssurizer nt / Test: Manual / 3 I Closed Closed ssurizer Spray nt / Test: Letdown Manual / 3 I Closed Closed nt / Test: Charging Manual / 4 I Closed Closed nt / Test: Safety Manual / 10 I Closed Closed ection nt / Test: Auxiliary Manual / 2 I Closed Closed ray mpling Manual / 3 I Open Open utdown Cooling 1 I Closed Closed lief Valves may be open or shut during normal operation or post-incident.
figures indicated above represents an engineering controlled drawing that is Incorporated by erence in the MPS-2 FSAR. Refer to the List of Effective Figures for the related drawing ber and the controlled plant drawing for the latest revision.
1 MATERIALS EXPOSED TO COOLANT materials exposed to the reactor coolant have shown satisfactory performance in operating tor plants. A listing of materials is given in Table 4.4-1.
2 INSULATION ng and equipment are insulated with granular-type, reflective-type and NUKON fiberglas-material compatible with the temperature and functions involved. All insulation material d on stainless steel has a low ( 600 ppm) soluble chloride content to minimize the possibility chloride induced stress corrosion. Removable metal reflective-type thermal insulation is vided on weld areas of the reactor coolant system subject to inservice inspection. The acement pressurizer is entirely covered with reflective metallic insulation, taking into sideration the containment sump clogging concerns as detailed in NRC Regulatory Guide
. The chemical makeup of the insulation conforms to NRC Regulatory Guide 1.36.
removable metal reflective-type thermal insulation is provided on the reactor cavity wall. The ary channel heads of the two steam generators are covered with the NUKON fiberglas-type erial.
ovable blanket insulation is provided on the CEDM, instrumentation nozzle areas of the acement reactor vessel head. Interconnected rigid panel insulation covers the lower portion of head and the head flange for easy removal.
thickness of insulation is such that the exterior surface temperature is not higher than roximately 50F above the maximum containment ambient (120F). All insulation support chments are attached prior to final stress relief.
possibility of leakage of reactor coolant onto the reactor vessel head or other part of the tor coolant pressure boundary causing corrosion of the pressure boundary has been stigated by Combustion Engineering (CE).
ailed laboratory examinations have shown that:
- a. Reactor coolant (containing boric acid) alone, at temperatures greater than about 250F, does not result in significant corrosion of low alloy steels. Therefore, under normal operating conditions, corrosion of the pressure boundary is not a concern.
Below this temperature, boric acid solutions 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 calcium silicate insulation to be used on this plant do not initiate attack.
n calcium silicate insulation, NUKON fiberglas-type insulation or metal reflective-type lation is used.
3 COOLANT CHEMISTRY trol and variation of the reactor coolant chemistry is a function of the chemical and volume trol system. Sampling lines are provided from the reactor coolant piping to provide a means taking periodic sample of the coolant for chemical analysis. Table 4.4-2 contains the Reactor lant Chemistry parameters and limits listing. The water chemistry is maintained as follows:
At temperatures below 250F, no upper limit on dissolved O2 is specified.
- 1. Hydrazine should only be added during subcritical heatup at 1.5 times the measured oxygen concentration.
- 2. Consistent with concentration of additives.
wetted surfaces in the reactor coolant system are compatible with the above water chemistry.
TABLE 4.4-1 MATERIALS EXPOSED TO COOLANT ctor Replacement 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 **
ng Cladding Austenitic Stainless Steel Type 304 L m Generator Bottom Head Cladding Weld Deposited Type 308 SS
- Tube Sheet Cladding Weld Deposited Ni-Cr-Fe Alloy Tubes Ni-Cr-Fe Alloy ps Casing Austenitic Stainless Steel, Type 316 Internals Austenitic Stainless Steel, Type 316 and Type 304 ssurizer 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.
PARAMETER REACTOR COOLANT LIMITS spended Solids, ppm maximum 0.35 prior to reactor startup at 25F Determined by the concentration of boric acid and lithium present. Consistent with the Primary Chemistry Control Program. (a) loride, ppm Cl-, maximum 0.15 oride, ppm F-, maximum 0.10 drogen as H2, cc (STP)/Kg H2O 25-50 ssolved O2, ppm maximum 0.1 (b) (c) (d) hium as Li7, ppm Consistent with the Primary Chemistry Control Program (a) ron, ppm 0-2620 (e) nductivity, S/cm at 25C Relative to Lithium and Boron concentration
) During power operation lithium 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 during, shutdown periods to aid in the cleanup of corrosion products. By evaluation, a maximum lithium concentration of 4.5 ppm is permissible with a target lithium concentration of 4.3 ppm for 100% power operations.
) The temperature at which the Oxygen limit applies is > 250F.
) The at power operation residual Oxygen concentration control value is ppm
) During plant startup, Hydrazine may be used to control dissolved Oxygen concentration at ppm
) RCS boron concentration is maintained as necessary to ensure core reactivity or shutdown margin requirements are met. Although the RCS 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 maximum boron concentration of 2400 ppm.
1 PREVENTION OF BRITTLE FRACTURE protect against non-ductile failure, the requirements of 10 CFR 50 Appendix G have been lemented and the requirements of 10 CFR 50.61 have been satisfied.
CFR 50 Appendix G provides fracture toughness requirements which ensure sufficient gins of safety against non-ductile failure during normal operation, including anticipated rational occurrences and inservice hydrostatic tests. The requirements of this Appendix are lemented by Figure 4.5-4 which provides maximum heatup and cooldown rates for the tor coolant system and maximum reactor coolant system pressure (as indicated by pressurizer sure) as a function of reactor vessel inlet temperature. Additional discussions addressing the elopment of these limits are provided in Sections 4.5.1.2 and 4.5.1.3.
CFR 50.61 provides additional fracture toughness requirements to protect against non-ductile ure of the reactor vessel during pressurized thermal shock (PTS) events. Compliance with e requirements is demonstrated by ensuring that the end-of-life reference temperature (RTPTS) the reactor vessel beltline materials stays below the established limits using the prescribed hods. Additional discussion addressing 10 CFR 50.61 is provided in Section 4.5.1.4.
1.1 Initial Nil-Ductility Transition Reference Temperature.
original design requirements for carbon and low alloy materials which form the pressure ndary of the reactor coolant system (RCS) have impact properties which meet the uirements of paragraph N-330 of the Summer 1969 ASME Boiler and Pressure Vessel Code, tion III, at 40F or less.
address changes in regulations and demonstrate compliance with 10 CFR 50 Appendix G and CFR 50.61 requirements, the original design requirements of N-330 were supplemented and materials initial Nil-Ductility Transition Reference Temperature (RTNDT) were subsequently blished.
impact properties for the replacement reactor vessel closure head meet the requirements of cle NB 2331 of the ASME B&PV Code,Section III, 1998 Edition through 2000 Addenda.
erence Temperature RTNDT of - 40F was established in accordance with SA 508 plementary Requirement (S10) and NB 2300.
replacement reactor vessel closure head and the replacement pressurizer were evaluated for ection against non-ductile failure in accordance with the methodology presented in ASME tion III, Appendix G for Class 1 components. The maximum stress intensity factors for the sients meet the fracture toughness requirements of ASME Section III Appendix G for a tulated defect of 1/10th thickness of the reactor vessel head and 1/4th thickness of the surizer.
-2331, and establish a maximum RTNDT at a temperature of 0F.
1.2 Nil-Ductility Transition Reference Temperature Shift flux of fast neutrons at the reactor vessel wall is governed by the reactor core load design, ngement of the reactor internals and average power levels, among other factors. The gration of this flux over time, called fluence, is monitored by dosimetry materials included in surveillance capsules located near the vessel inner wall. In the time since original plant startup thermal shield has been removed and the core load design has changed. In addition, small k-arresting holes have been drilled in the core barrel, affecting the flux for one vessel plate.
ed on results from surveillance capsule dosimetry retrieved through cycle 14, the known figuration history, neutron flux modeling calculations described below, and further assuming re reactor operation with the low leakage design at 2700 Mwt and 90 percent plant capacity or, the maximum fluence at the end of the period of extended operation of 60 years will be roximately 3.83 x 1019 n/cm2 (E > 1Mev). These fluence estimates are fully described in erence 4.5-2. The following summary describes the key methods of the referenced analysis.
cussion of Fluence Calculations prediction of neutron fluence at various locations was based on an analysis of neutron sport for given configurations and source developed to model the reactor. To further refine the del accuracy, the results of the fluence analysis were correlated with measurements of actual nce based on dosimetry retrieved from surveillance capsules. Estimations of future fluence are based on a combination of the currently measured fluence and the extrapolation of calculated le 14 flux though the end of licensed plant life. Thus, it is assumed there will be no significant nge to reactor configuration or core load design.
three dimensional discrete ordinates transport computer code DORT was used in the eillance capsule W-83 analysis to model neutron transport within the reactor. The code results cribe the space and energy dependent neutron flux present in the reactor.
reactor was modeled as a 1/8th segment of the core, the reactor internals, core barrel, thermal ld (through cycle 5), explicit representations of the surveillance capsules at 6and 14, the sure vessel cladding and vessel wall, the insulation external to the pressure vessel, the mary biological cladding and shield wall.
e that a variation of this model, in which material composition of the surveillance capsules redefined as water, was utilized to determine the maximum neutron exposure at the pressure sel wall in octants of the core that do not contain surveillance capsules.
actual activation or fission measured for the retrieved capsule dosimetry was compared to the ulated activation. The ratio of measured to calculated activation, M/C, was determined to fall l within the criteria specified in Regulatory Guide 1.190.
shift prediction methodology for the determination of the adjusted reference temperature T) is done in accordance with Revision 2 to Regulatory Guide 1.99 (dated May 1988). The T, for each material in the beltline is given by the following expression:
ART = Initial RTNDT + RTNDT + Margin shift in RTNDT is based on fluence predictions, described above, for the time period esponding to 54 EFPY. Table 4.6-14 provides the results of the calculation.
1.3 Operational Limits components in the RCS are designed to withstand the effects of cyclic loads due to RCS perature and pressure changes. These cyclic loads are introduced by normal unit load sients, reactor trips and startup and shutdown operation.
ing unit startup and shutdown, the rates of temperature and pressure changes are limited. The gn number of cycles for heatup and cooldown is based upon a rate of 100F/hr and for cyclic ration.
maximum allowable RCS pressure at any temperature is based upon the stress limitations for tle fracture considerations. These limitations are derived by using the rules contained in tion III of the ASME Code including Appendix G, Protection Against Nonductile Failure and rules contained in 10 CFR 50, Appendix G, Fracture Toughness Requirements.
R Figures 4.5-4 and 4.5-5 provide the RCS pressure-temperature limitations during plant tup and cooldown. Figures 4.5-4 and 4.5-5 are valid for the period up to and including fifty years of full power integrated neutron flux, as determined using the rationale of tion 4.5.1.2, and was developed using the rules of Appendix G, Protection Against ductile Failure of the ASME Boiler and Pressure Vessel Code,Section XI, 2002 Addenda.
ng these rules the belt line material of the reactor vessel is established as the controlling ponent section throughout plant life. This established an upper boundary on RCS pressure as nction of RCS temperature and allowable heatup and cooldown rates to ensure prevention of ductile failure.
limitations for normal heatup and cooldown rates and the applicable temperature ranges are marized in Table 4.5-2.
RCS pressure-temperature limits provided by Figures 4.5-4 and 4.5-5 have been corrected to cated pressurizer pressure versus indicated cold leg temperature.
cated cold leg temperature is the best available indication of the reactor vessel downcomer perature and will normally be monitored as RCS cold leg temperature when reactor coolant ps are operating or natural circulation is occurring. In the instances where the shutdown ling (SDC) system is operating without RCPs, the SDC system return temperature will be
o shown is an allowable region for shutdown cooling system operation. This region is blished based upon the design pressure-temperature ratings of components of the shutdown ling system and the normal operation of this system as described in Chapter 6 (Section 6.3)
Chapter 9 (Section 9.3).
reactor vessel beltline material consists of six plates. The NDT temperatures (TNDT) of each e was established by drop weight test (DWT). Charpy tests were then performed to determine hat temperature the plates exhibited 50 ft-lbs absorbed energy and 35 mils lateral expansion.
ta points were based on average of three specimens.)
ilar testing was not performed on all remaining material in the RCS. However, sufficient act testing was performed to meet appropriate design code requirements and a conservative NDT of 50F has been established for longitudinal direction.
a result of fast neutron irradiation in the region of the core, there will be an increase in the NDT with operation. The techniques used to predict the integrated fast neutron (E 1 Mev) es of the reactor vessel are described in Section 4.5.1.2 of the FSAR.
ce the neutron spectra and flux measured at the samples and reactor vessel inside radius should nearly identical, the measured reference transition temperature shift for a sample can be lied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the erence in calculated flux magnitude. The maximum exposure of the reactor vessel will be ained from the measured sample exposure by application of the calculated azimuthal neutron variation.
actual shift in RTNDT will be established periodically during plant operation by testing of tor vessel material samples which are irradiated cumulatively by securing them near the de wall of the reactor vessel as described in Section 4.6.3 and shown in Figure 4.6-1 of the R. To compensate for any increase in the RTNDT caused by irradiation, limits on the pressure-perature relationship are periodically changed to stay within the stress limits during heatup cooldown.
addition to the requirements provided by ASME Section XI, Appendix G and 10 CFR 50, endix G, the following items were also considered in the development of the RCS pressure-perature limits provided by Figures 4.5-4 and 4.5-5.
est Service Temperature - As indicated previously, an RTNDT for all material with the eption of the reactor vessel beltline was established at 50F. ASME III, Art. NB-2332(b) uires a lowest service temperature of RTNDT + 100F for piping, pumps and valves. Below this perature a pressure of 20 percent of the system hydrostatic test pressure cannot be exceeded.
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.
1.4 Pressurized Thermal Shock ccordance with 10 CFR 50.61, reactor pressure vessel belt line materials have been reviewed stablish a reference temperature for pressurized thermal shock (RTPTS). This review evaluated loading patterns and the actual or best estimate of copper and nickel in the vessel material. In ition, the reactor vessel material composition and properties were compared to those of eillance capsule materials from which actual tests and measurements were taken. A summary his 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 4.6-13.
- 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 4.6-13 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 surveillance capsule (capsule W-83) evaluation.
culated RTPTS values have been obtained using the above assumptions. Table 4.6-13 provides results of the calculations. This table will be updated whenever changes in core loadings, eillance measurements, or other information indicate a significant change in the RTPTS ected values, as required by 10 CFR 50.61(b)(1). The values that were calculated do not eed the RTPTS screening criteria of 270F for plates, forgings, and axial weld materials, and F for circumferential weld materials at 54 effective full power years.
2 SEISMIC DESIGN nuclear steam supply system (NSSS) is designed to withstand the loads imposed by the imum hypothetical accident and the maximum seismic disturbance without loss of functions uired for reactor shutdown and emergency core cooling. The method of combining stresses duced by these simultaneous conditions is described in Section 4.2.1.
RCS components are considered Class 1 for seismic design. Loadings which result from hquake conditions are categorized and, in combinations with other specified loadings, are luated 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 and upset conditions, loadings resulting from the OBE shall be considered to occur during normal operation at full power. 200 cycles of the OBE have been specified 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 significant 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 (DBE) - Two faulted conditions, which include loadings resulting from the DBE, are defined.
- Loadings resulting from the combined effects 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.
2.1 Piping primary stress limits applied in evaluating the emergency and faulted conditions for the RCS ng are specified as follows:
The RCS piping is designed in accordance with the requirements for 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 emergency and faulted 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.
2.2 Vessels primary stress limits applied in evaluating the emergency and faulted conditions for vessels in RCS are specified as follows:
replacement steam generator subassemblies are 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 Conditions, are applied in evaluating the emergency conditions and in evaluating Faulted Condition (1) noted above in Section 4.5.2.b.
h respect to Faulted Condition (2) in Section 4.5.2.b, the primary stress limits of Section III, 17-11, Stress Limitations for 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 break is considered to have occurred, the stress criteria for this condition need not be applied to the 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 component support assemblies. In the case of nozzles integral with component support assemblies, the criteria is applicable in all regions which sustain support reactions.
replacement reactor vessel closure head is designed in accordance with ASME Section III, ss 1 vessels, 1998 Edition through 2000, Addenda. The primary stress limits of Section III, NB 5 Appendix F, F-1331.1 (a, b and c) are applied in evaluating the faulted conditions ble 4.2-2B). The load combination for the faulted condition (Service Level D) is defined in tion 4.5.2.b. DBE and LOCA loads are combined using the Square Root Sum of Squares SS) method for evaluating the faulted condition. The emergency conditions (Service Level C) bounded by the design conditions and therefore, the replacement reactor vessel head is luated using the primary stress limits of NB 3221.
pressurizer was replaced to the design requirements of ASME Boiler and Pressure Vessel e Section III, Subsection NB, 1998 Edition through 2000 Addenda. The primary stress limits the emergency and faulted conditions are applied in accordance with subsection NB 3000 of design code.
2.3 Pumps and Valves primary stress limits applied in evaluating the emergency and faulted conditions for the ps and valves in the RCS are specified as follows:
RCS pumps and valves are designed in accordance with the rules of ASME Code for Pumps Valves for Nuclear Power - March 1970, Draft. Supplementary to these rules, the primary ss limits used for Vessel, as discussed under Item b. above, are applied in evaluating the rgency and faulted conditions for the reactor coolant pumps (RCP). In the case of valves, the
refore, no supplementary criteria which limit the primary stresses during abnormal conditions necessary.
3 OVERPRESSURE PROTECTION 3.1 Overpressure Protection During Normal Operation RCS is structurally designed for operation at 2485 psig and 650F (pressurizer 700F).
ration of the system 2235 psig nominal and 600F will result in material stresses 90 percent of gn values. Detailed structural analyses have been performed by the component vendors and ewed independently by Combustion Engineering (CE) for all portions of the system. Welding erials used have physical properties superior to the materials which they join. Inspection cedures and tests specified and independently reviewed by CE were carried out to assure that sure-containing components have the maximum integrity obtainable with present code-roved inspection techniques.
RCS is protected against overpressure by two ASME Code approved safety valves which t system pressure to a maximum of 110 percent of design. In addition, two solenoid-operated ef valves are provided as described in Section 4.3.7.
3.2 Low Temperature Overpressurization Protection RCS low temperature overpressurization protection (LTOP) system, along with inistrative procedures, provides protection against exceeding the ASME Section III, endix G (Protection Against Brittle Fracture) requirements during cold plant conditions; i.e.,
perature 275F. The LTOP system consists of two redundant relief trains each with one er operated relief valve (PORV) with a setpoint of 400 psig and associated relief piping as cribed in Section 7.4.8.
4 REACTOR VESSEL THERMAL SHOCK analysis of the thermal stresses produced in the reactor vessel wall due to the operation of the ty injection system has been performed. The analysis has been reported in a CE report ermal Shock Analysis on Reactor Vessels due to Emergency Core Cooling System (ECCS) ration, A-68-9-1, and was submitted for the record on Docket Number. 50-309, Maine kee Atomic Power Station. The results show that there will be no failure of the reactor vessel to brittle fracture.
rk has also been performed to refine the surface heat transfer coefficient. The temperature nch data obtained during the heat treatment of several heavy section steel plates was reviewed.
h this background, CE planned and conducted additional quench tests to develop experimental t transfer coefficients.
ated (turbulent) water bath at 80F, nearly duplicating the temperature conditions which ld be present in the reactor during ECCS operation. The temperature of all thermocouples was rded throughout the cooldown of the plate.
sequently, the data was compared to a heat transfer computer model of the plate to obtain an ctive heat transfer coefficient. A detailed report covering this work, entitled Experimental ermination of Limiting Heat Transfer Coefficients during Quenching of Thick Steel Plates in er (A-68-10-2, December 13, 1968), was submitted to the AEC (now NRC) and made part of public record. The report concludes that an effective heat transfer coefficient of 300 Btu/hr-F provides a realistic upper limit for thick steel plates quenched in highly agitated room perature water.
stress near the tip of axial and circumferential vessel cracks of various depth has been rmined by the finite element method. This work was reported by a CE report, Finite Element lysis of Structural Integrity of a Reactor Pressure Vessel during Emergency Core Cooling, 0-19-2, January 1970, and is part of the public record.
se reports substantiate the analytical conclusion that a vessel failure will not occur due to CS operation. An acute crack, even if formed, will not propagate.
5 LEAK DETECTION hods are provided to alert the operator of the presence of leakage from the RCS in a timely ner to allow detection and isolation of the leak to ensure the leakage does not exceed eptable limits. Detection of leaks from the RCS can be accomplished by one or a combination he below listed means.
k Detection Within the Containment ks 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 containment air radiation monitor system. The sensitivity and response time of the particulate 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 leakage rate will be weak. These monitors are best used for trending purposes and a trigger to check other indications for leakage source;
- d. Monitoring the reactor building closed cooling water (RBCCW) temperature to and from the containment air recirculation (CARS) and cooling units.
kage from the RCS is indicated by the level in the pressurizer or in the primary drain tank and/
igh RCS makeup flow from the chemical and volume control system.
ief and Safety Valves Located on the Reactor Coolant System ng from the relief and safety valves located on the pressurizer is provided with temperature sors with readout in the main control room. Any temperature increase will indicate relief or ty valve leakage. In addition, an increase in pressurizer quench tank level, pressure, or perature will also indicate leakage.
ctor Vessel Head Closure space between the double O-ring seal is monitored to detect an increase in pressure, which cates a leak past the inner O-ring. A leak indicator in containment indicates pressure. Upon h temperature a control room alarm is sounded.
kage Through Steam Generator Tubes or Tube Sheet increase in radioactivity indicated by radiation monitors for the gases from the condenser air tors or steam generator blowdown system monitors will indicate leakage through steam erator tubes to the secondary side. The N-16 radiation monitors will indicate primary to ondary leakage when the reactor is at power.
6 PREVENTION OF STAINLESS STEEL SENSITIZATION sitization of stainless steel occurs when unstabilized Type 300 Series stainless material is held he temperature range of 900-1400F for sufficient time to form a continuous network of omium carbide precipitates. Sensitization occurs after approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> at 900F, as pared to one hour at 1400F. Stabilized Type 300 Series stainless material avoids continuity hromium carbide precipitates in the grain boundaries by careful control of metal chemistry.
re are no furnace sensitized stainless steels in the reactor coolant pressure boundary (RCPB).
sitization is precluded from the NSSS through materials selection and control to all welding heat treating procedures.
or portions of the RCPB in CEs nuclear plants are shown in preceding tables in this section to ormed by carbon steels and a high nickel base alloy. None of these materials is susceptible to ace sensitization (a continuous network of iron-chromium grain boundary carbides) in the se of unstabilized Type 300 Series stainless steels. All internal carbon steel surfaces are weld-osit or roll-bond clad with Inconel or stainless steel, to preclude excessive corrosion.
rnal surfaces of the reactor vessel, pressurizer and steam generator primary side are overlaid h Type 308, 309 or 308L weld deposited metal. Weld metal composition is carefully controlled vercome interface dilution and promote an austenoferritic duplex structure. Therefore, during stress relief heat treatment (1150 25F) required by the ASME Code for the pressure vessel,
ite acts as a carbon sink and prevents continuity of carbide precipitates.
ensive testing has confirmed that, properly formulated (a duplex structure), Type 308 weld osited metal does not form a continuous carbide network within grain boundaries even owing a typical vessel post weld heat treatment (viz, 1150F for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />). Hence, the material mmune to intergrannular corrosion.
replacement reactor vessel head is overlaid inside with stainless steel weld material. The first r is type 309 stainless steel and all subsequent layers including the final layer that is in contact h the reactor coolant is type 308L stainless steel.
ceptibility to underclad / reheat cracking was minimized by controlling the welding heat input chemical composition of the RV Head forging to maintain the Delta G function less than 0.1 ccordance with the correlations developed by the vendor (Mitsubishi Heavy Industries). In ition, in order to comply with the intent of Reg Guide 1.43, examinations were performed on e representative penetration holes in the RV Head J groove preparations prior to depositing buttering to specifically check for underclad / reheat cracking on the susceptible area up to inch below the fusion line between the base material and cladding.
bimetallic weld on the CEDM and instrumentation nozzles for the replacement head joins the nless steel threaded connector or flange to the alloy 690 penetration tube. Additionally, the erials for welding to Alloy 690 have lower susceptibility to PWSCC than the original alloy 82 welds.
J-groove weld design for the guide tubes is modified to control the amount of overwelding to inimum and thus minimize residual weld stresses. The weld surface finish was also improved acilitate surface examinations.
other Type 300 Series stainless steel used either is not subjected to a furnace sensitization heat tment or, as is the case of cladding on the primary piping, is of Type 304L (low carbon,
% max.) composition and is not susceptible to the formation of continuous chromium carbide n boundary networks.
replacement pressurizer is overlaid inside with stainless steel weld material. Susceptibility to erclad/reheat cracking was precluded by performance testing of the cladding weld procedure cifications in accordance with NRC Regulatory Guide 1.43.
RCS pump casing is CF8M (Cast 316), which again is a duplex material. The casting is tion annealed after welding; hence, this component will not have a sensitized structure.
ogen-enhanced stainless steel was not used in the fabrication of any RCPB component.
ause carbon steel piping is used in the RCS, no safe ends are required on the reactor vessel, or m generator primary nozzles. Where small diameter solid stainless pipes are employed (or in instance of welding the coolant pump casing to carbon steel), an Inconel-182 weld deposit is
oining small diameter annealed solid stainless steel piping, as is used in the pressurizer surge
, charging pump lines and safety injection systems, some carbide precipitation will occur as a lt of welding. However, the precipitation that occurs in the weld heat affected zone does not sitize the material in the context of forming continuous grain boundary carbide precipitates.
ples from typical such welds pass the industry accepted standard for intergrannular corrosion eptibility (i.e., Strauss Test - ASTM-A393). Metallographic examination of such welds al that only discontinuous grain boundary precipitates are present.
following four welding processes are used to weld stainless steel in CEs NSSSs. Welding cesses are performed in accordance with written procedures, as provided in the Quality urance Program. Nitrogen is not used as a purge gas in the welding process in lieu of argon or um 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 coalescence is produced by heating with an electric arc drawn between a flux covered metal electrode and the work.
- b. In the gas tungsten arc (GTA), coalescence is produced by heating with an electric arc drawn between a tungsten electrode 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 effected by heating with an arc drawn between a continuous feed wire electrode and the work. Shielding of the weld is obtained from an externally supplied inert mixture.
- d. Submerged arc (SA) produces coalescence by heating with an arc or arcs drawn between a bare metal (filler) electrode or electrodes and the work. The arc and weld are shielded by a blanket of granular fusible flux.
le 4.5-1 lists the nozzles on the steam generator, reactor vessel, pressurizer and piping. The e also indicates the size of the nozzle, base material of the nozzle and, where applicable, the erial of the nozzle safe end.
procedures used in welding nozzles within CE manufacturing facilities are generally as ows: (1) For nozzles with stainless steel safe ends, the safe ends are not attached until after l stress relief, and (2) the stainless steel safe end is welded to Inconel buttering on the alloy
ing manufacture of the core structures, various parts of the core structure are tested for sitization using the Strauss Test (ASTM A393). Test specimens consist of: (1) mockups of ous welded joints, and (2) monitoring specimens included in any heat treatment of various ponents. None of the specimens tested in conjunction with fabrication of reactor vessel rnals for previous CE plants have failed the Strauss Test. The replacement reactor vessel head replacement pressurizer weld materials and welding are controlled in accordance with ulatory Guide 1.44, May 1973, Control of the use of Sensitized Steel, to preclude sitization of austenitic stainless steels.
typical weld heat input with the above processes as used by CE to joint Type 300 Series nless steel varies from 6000 joules per inch GTA to 96,000 joules per inch SA. To avoid weld t affected zone sensitization, CE limits the interpass temperature on multipass welds in nless steels to 350F maximum. The replacement pressurizer uses only low carbon stainless l materials. The CE interpass temperature limits do not apply to the replacement pressurizer.
EVA limited the interpass temperature for low carbon stainless steel materials to 250C 2F), which is sufficient to prevent sensitization. The combination of normal heat input using above welding procedures and control of interpass temperature assures minimum carbide ipitation in the weld heat affected zone. Samples from large welds have been examined in the ratory and none has failed the Strauss Test.
ield welding operations, Bechtel uses welding procedures that limit heat input to the weld s, and thus preclude the possibility of sensitization of austenitic stainless steels. Most of the ding employed is of the manual SMA process; a minor amount of GMA welding is also used.
ther one of these processes would be classified as an excessively high heat input welding cedure.
ther precautions employed to preclude field sensitization of austenitic stainless steels consist
- a. Preheat and interpass temperatures are limited to 350F maximum.
- b. Controlled welding sequence is used to minimize heat input.
- c. The practice of block welding is prohibited.
- d. Postweld heat treatment is prohibited on equipment and/or parts that are completely or partially fabricated of austenitic stainless steel. During the fabrication of the replacement pressurizer post weld heat treatment was performed on the safety/relief nozzles and spray nozzle 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.
reparing for, and engineering, the field welding requirements, close liaison is maintained ween Bechtel and CE. Detailed welding parameters prepared by CE are submitted to CE for ew and mutual concurrence and approval before they are adopted for use. Bechtel quality rance procedures for field welding are discussed in Appendix 1B. Appendix 1B was located he original FSAR dated August 15, 1972.
delta ferrite content of all austenitic stainless steel weld metals used to fabricate CEs RCS ponents is controlled to 5-18% in the as-deposited condition. Delta ferrite content is firmed from chemical analysis and the Schaeffler or McKay Diagrams. In addition, a brated ferrite measuring instrument (Seven Gauge or similar) is used. The ferrite requirement et for each heat and/or lot of filler metal used in fabrication. The ferrite content in the weld erials and welding for the replacement reactor vessel head and replacement pressurizer are in ordance with Reg Guide 1.31, Control of Ferrite in Stainless Steel weld metal.
ere field welding of austenitic stainless steels is required in the RCS only the inert GTA and ual SMA processes are used. The welding procedures were qualified in accordance with tion IX of the ASME Boiler & Pressure Vessel Code. Tensile test specimens were taken to ibit the tensile strength of the welded joints and bend tests were taken to indicate the ductility he weld joints. Filler material compositions are in accordance with the ASME SFA/AWS filler erial specification and are selected on the basis of the austenitic stainless steels to be welded.
ddition to these requirements the filler materials must be capable of depositing 8 to 25%
ite. This is verified for each heat or lot of filler material by plotting the heat analysis for bare e or the analysis of an all weld metal deposit for covered electrodes on the Schaeffler, DeLong quivalent diagram to determine its ferrite content and the acceptability of the filler materials.
austenitic stainless steel materials including cladding are analyzed for delta ferrite in ordance with NB-2433. The FN shall be 5FN to 15FN. The FN for undiluted ER309L deposit l be in the range of 5FN - 22FN. By control of the welding processes, the filler materials and welding parameters (by specifying a maximum interpass temperature of 350F), the welds uld contain sufficient delta-ferrite in the austenitic matrix to avoid hot cracking in the enitic stainless steel welds.
7 REFERENCES 1 W. G. Counsil (NU) letter to J. R. Miller (NRC), Millstone 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).
2 S. E. Scace (DNC) to U.S. NRC, Millstone 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).
SIZES AND NOZZLE MATERIALS m Generator Component (Number) Size Material mary Inlet (1) 42 inch ID Carbon steel safe ends clad with stainless steel mary Outlet (2) 30inch ID Carbon steel safe ends clad with stainless steel ssure Taps (4) 1inch Schedule 160 Inconel B-166 am Outlet (1) 36 inch ID Carbon steel edwater (1) 18 inch Schedule 80 Carbon steel ttom Blowdown (2) 4 inch ID Carbon steel uid Level (12) 1 inch Schedule 80/ Carbon steel/Inconel B-166 Schedule 160 rogen Addition (1) 1 inch Schedule 80 Carbon steel t Layup (1) 2 inch Schedule 160 Carbon steel ctor Vessel and Head Component (Number) Size Material mary Outlet (2) 42 inch ID Carbon steel clad with stainless steel mary Inlet (4) 30 inch ID Carbon steel clad with stainless steel DM/HJTC (69) 2.718 inch ID Ni-Cr-Fe and stainless steel trumentation (8) 4.625 inch ID Ni-Cr-Fe and stainless steel nt (1) 0.75 inch Schedule 80 Inconel SB-167
e Component (Number) Size Material rge Nozzle (1) 12 inch Schedule 160 A-105 Gr II with A-351 Gr CF8M safe end ssure (8) 0.75 inch Schedule Inconel B-166 with A-182 Type 160 316 safe end D (25) 1 inch nominal Inconel B-166 utdown Cooling (1) 12 inch Schedule 140 A-105 Gr II with A-351 Gr CF8M safe end ray (2) 3 inch Schedule 160 A-105 Gr II with A-182 Type 316 SS safe end fety Injection (4) 12 inch Schedule 140 A-182 F1 with A-351 Gr CF8M safe end arging Inlet (2) 2 inch Schedule 160 A-105 Gr II with A-182 Type 316 SS safe end mpling (2) 0.75 inch Schedule Inconel B-166 with A-182 Type 160 316 SS safe end ain/Letdown (5) 2 inch Schedule 160 A-105 Gr II with A-182 Type 316 SS safe end mp (8) 30 inch ID A-516 Gr 70 clad with A-240, Type 304L SS with A-351, CF8M safe end
Cooldown Heatup a dicated Cold Leg Indicated Cold Leg Temperature Limit Temperature Limit 20F 50F/hour 200F 60F/hour 20F 100F/hour 200F < T 275F 80F/hour 275F 100F/hour These limitations apply to hydrostatic and leak test conditions.
FIGURE 4.5-1 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 7 FULL POWER YEARS
GURE 4.5-2 REACTOR COOLANT SYSTEM PRESSURE - TEMP LIMITATIONS RING PLANT HEATUP/COOLDOWN AFTER 7 YEARS INTEGRATED NEUTRON FLUX
FIGURE 4.5-3 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 2 YEARS OF FULL POWER
FIGURE 4.5-4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS FOR 54 EFPY
IGURE 4.5-5 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS FOR 54 EFPY
1 GENERAL p inspection and tests of all major components are performed at the vendors plants prior to ment. An inspection at the site is performed to assure that no damage has occurred in transit.
ting of the reactor coolant system (RCS) are performed at the site upon completion of the plant struction. These tests will include hydrostatic tests of all fluid systems. A complete visual ection of all welds and joints are performed prior to the installation of the insulation. All field ds are radiographically and dye penetrant inspected in accordance with the requirements of tion III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel e.
ddition to the code-required examinations on the replacement reactor vessel head, base line minations were performed prior to the component being place in service. These baseline minations include a full volumetric examination of the CEDM nozzle base material, a imum of 2 inches from the high point of the J-groove weld down to the threaded portion.
itionally eddy current examination was performed on the wetted surface of at least 28 pheral CEDM nozzles in the bimetallic weld area.
ot flow test of the reactor coolant loop up to zero power operating pressure and temperature hout the core installed will be made. The system will be checked for vibration and cleanliness.
iliary systems will be checked for performance (see Chapter 13).
2 NIL DUCTILITY TRANSITION REFERENCE TEMPERATURE carbon and low alloy steels which form the reactor coolant pressure boundary are required to sfy the requirements of 10 CFR 50, Appendix G which utilizes the nil ductility transition rence temperature (RTNDT) as the basis for establishing operational limitations for the reactor lant system.
original materials associated with the reactor coolant pressure boundary were ordered and ed to the requirements of ASME Code,Section III, Paragraph N-330. The impact properties of e materials were required to meet the acceptance criteria noted in Paragraph N-330 at +40F.
se tests determined the nil ductility transition temperature (NDTT) for these materials. As the inal design requirements were insufficient to directly establish the initial RTNDT of the sure boundary materials and comply with the requirements of 10 CFR 50, Appendix G, the NDT was either estimated using the procedures of MTEB 5-2, performance of supplemental ing of surplus material in accordance with NB-2331 of Section III to the ASME Code or eloped from generic impact testing.
he case of the reactor vessel, the beltline materials will experience a shift (an increase) in nil tility transition reference temperature due to neutron irradiation. This increase is calculated in ordance with regulatory positions and is described in Section 4.5.1.3.
gram is provided in Section 4.6.3.
design material toughness test requirements were as follows:
ctor Vessel bon and low alloy steel materials which form a part of the pressure boundary shall meet the uirements of the ASME Code,Section III, Paragraph N-330 at a temperature of +40F. It shall n objective that the materials meet this requirement at -10F. Charpy tests shall be performed the results used to plot a transition curve of impact values vs. temperature extending from y brittle to fully ductile behavior. The actual NDT temperature of inlet and outlet nozzles, sel and head flanges, and shell and lower head materials shall be determined by drop weight s per ASTM E-208.
lacement Reactor Vessel Closure Head impact properties of carbon and low alloy steel materials including weld filler metals for the acement head shall meet the requirements of ASME Section III, Subsection NB 2331, 1998 tion through 2000 Addenda. Charpy V-notch transition curves were established in accordance h SA 508 supplementary requirements (S3) for temperatures showing, upper shelf energy, er shelf energy and transition. The actual TNDT shall be determined by drop weight test in ordance with ASTM E208. Reference Temperature, RTNDT shall be determined in accordance h NB 2331.
m Generator and Pressurizer all be an objective that impact properties of all ferritic steel materials which form a part of the sure boundary shall meet the requirements of the ASME code Section III, at a temperature of F; alternate higher temperature levels up to 40F may be used only if the material fails at F. Such higher temperature levels, if applicable, shall be determined and documented. This ective is applicable to the pressurizer and the original steam generators, of which the original m drums are still in use. For the steam generator replacement subassemblies, the maximum wable RTNDT as defined in paragraph NB-2311(a) of ASME Section III Code is 0F. For the acement pressurizer ferritic steel materials, the RTNDT shall be performed at a temperature of degrees F or less. The actual TNDT is determined by drop weight test in accordance with NB 1 to ASTM-208-91.
ctor Coolant Piping erials used to fabricate the pipe fittings shall be specified, examined and tested to satisfy as a imum the requirements of Chapter I-III of the American Nuclear Society (ANS) Code for ssure Piping B31.7, Class 1.
gn objective that the materials meet this requirement at 10F. Weld procedure qualifications weld metal certifications records may serve to demonstrate impact properties of welds.
initial toughness test data for these components are provided in Tables 4.6-1, 4.6-2, 4.6-3 and
- 4. Table 4.6-1 also provides RTNDT values.
ghness test data are summed as follows:
- a. The maximum NDT temperature for the reactor vessel as obtained from drop weight tests is +10F. Drop weight tests were conducted only for material used in the reactor vessel.
- b. The maximum temperature corresponding to the 50 ft-lb value of the Cv fracture energy for the reactor vessel is +65F.
Refer to the tables presented above the Charpy V-notch data at 10F for the steam generators, pressurizer, and reactor coolant piping.
- c. The minimum upper-shelf Cv energy value for the strong direction of the material used in the reactor vessel is 103 ft-lb.
The upper shelf-Cv 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.
to regulatory changes during the construction of the facility, it was necessary to establish the NDT of the reactor coolant pressure boundary materials to comply with the requirements of CFR 50 Appendix G. In most instances, this has been accomplished by utilizing the guidance MTEB 5.2 to estimate the RTNDT of the material based upon the available data. An initial NDT of 50F has been established for the reactor coolant pressure boundary materials luding the reactor vessel beltline) based on MTEB 5-2. However, in the case of the reactor sel beltline materials, the RTNDT was not determined through a combination of methods uding utilizing the guidance of MTEB 5-2, testing surplus materials to the requirements of
-2300, and utilizing generic data. The initial RTNDT values for the beltline materials are vided in Table 4.6-13. In addition, the primary side of the steam generator has been replaced the materials RTNDT values have been determined from testing in accordance with NB-2300.
surveillance program is implemented to monitor the radiation-induced changes in the hanical and impact properties of the pressure vessel materials in accordance with the uirements of 10 CFR 50, Appendix H. Changes in the impact properties of the material are luated by the comparison of pre-irradiation and post-irradiation Charpy impact test specimens.
nges in mechanical properties are evaluated by the comparison of pre-irradiation and t-irradiation data from tensile test specimens.
ee metallurgically different materials representative of the pressure vessel are investigated.
se are base metal, weld metal, and weld HAZ material. In addition to the materials from the tor vessel, materials from a standard heat of A533B, made available through the Heavy tion Steel Technology (HSST) Program, are also used. This reference material is fully cessed and heat treated and is used for Charpy impact specimens so that a comparison may be e between the irradiations in various operating power reactors and in experimental reactors. A plete record of the chemical analysis, fabrication history and mechanical properties of all eillance test materials is maintained.
exposure locations and a summary of the specimens at each location is presented in le 4.6-8. The pre-irradiation NDT temperature of each plate in the intermediate and lower sel shell courses is determined from the drop weight tests and correlated with Charpy impact s.
e metal test specimens are fabricated from sections of the shell plate in either the intermediate he lower shell course which exhibits the highest unirradiated NDT temperature. All base erial test specimens are cut from the same shell plate. This material is heat treated to a dition of the base metal in the completed reactor vessel.
d metal and HAZ material are produced by welding together two plate sections from the rmediate or lower shell course of the reactor vessel. All HAZ test materials are also fabricated m the plate which exhibits the highest unirradiated NDT temperature.
material used for weld metal and HAZ test specimens was adjacent to the test material used ASME Code,Section III tests and was at least one plate thickness from any water-quenched
- e. The procedures used for making the shell girth welds in the reactor vessel was followed in preparation of the weld metal and HAZ test materials. The procedures for inspection of the tor vessel welds was followed for inspection of the welds in the test materials. The welded e was heat treated to a condition which is representative of the final heat treated condition of completed reactor vessel.
itional information from the baseline surveillance program includes the chemical position of the surveillance plate and weldment (made from two separate heats of weld wire),
ch are given in Table 4.6-5. The baseline mechanical properties of the base metal (WR and
), weld metal, heat affected zone (HAZ), and standard reference material (SRM), are shown in les 4.6-6 and 4.6-7. The Charpy V-notch impact energy and lateral expansion data as a ction of test temperature are shown in Figures 4.6-5 through 4.6-14.
ons. The reactor vessel surveillance program was designed in accordance with ASTM E185 edition specified). The program complies with ASTM E185-73 and 10 CFR 50, Appendix H.
location of the surveillance capsule assemblies is shown in Figure 4.6-1. A typical eillance capsule assembly is shown in Figure 4.6-2. A typical Charpy impact compartment mbly is shown in Figure 4.6-3. A typical tensile monitor compartment assembly is shown in ure 4.6-4.
ion threshold detectors (U-238) were inserted into each surveillance capsule to measure the neutron flux. Threshold detectors of Ni, Ti, Fe, S, and Cu with known Co content have been cted for this application to monitor the fast neutron exposure. Cobalt is included to monitor thermal neutron exposure.
selection of threshold detectors is based on the recommendations of ASTM E-261. Method Measuring Neutron Flux by Radioactive Techniques. Activation of the specimen material also be analyzed to determine the amount of exposure.
maximum temperature of the encapsulated specimens will be monitored by including in the eillance capsules small pieces of low-melting-point eutectic alloys or pure metals vidually sealed in quartz tubes.
temperature monitors will provide an indication of the highest temperature to which the eillance specimens were exposed but not the time-temperature history or the variance ween the time-temperature history of different specimens. These factors, however, will affect accuracy of the estimated vessel material NDT temperature to only a small extent.
t specimens removed from the surveillance capsules are tested in accordance with ASTM ndard Test Methods for Tension and Impact Testing. The data obtained from testing the diated specimens will be compared with the unirradiated data and an assessment of the tron embrittlement of the pressure vessel material will then be made. This assessment of the T temperature shift is based on the temperature shift in the average Charpy curves, the average ves being considered representative of the material.
periodic analysis of the surveillance samples permit the monitoring of the neutron radiation cts upon the vessel materials. If, with due allowance for uncertainties in NDT temperature rmination, the measured NDT temperature shift turns out to be greater than predicted, then ropriate limitations would be imposed on permissible operating pressure-temperature binations and transients to ensure that the existing reactor vessel stresses are low enough to lude brittle fracture failure.
original six surveillance capsules were inserted into their designated holders during the final tor assembly operation. Each capsule remains in the reactor for the tentative schedule listed in le 4.6-9. Table 4.6-9 shows the target fluence for each of the capsules.
4 NONDESTRUCTIVE TESTS r to and during fabrication of the reactor vessel, nondestructive tests based upon Section III of ASME Boiler and Pressure Vessel Code were performed on all welds, forgings and plates as ows:
full penetration pressure retaining welds were 100 percent radiographed to the standards of graph N-624 of Section III of the ASME Boiler and Pressure Vessel Code. Other pressure ining welds such as those used for the attachment of mechanism housings, vents and rument housings or the replacement reactor vessel head and J groove welds were inspected to itional inspection criteria listed in Table 4.6-11.
forgings were inspected by ultrasonic testing, using longitudinal beam techniques. In addition, forgings were tested using shear wave techniques. Rejection under longitudinal beam ection, with calibration so that the first back reflection is at least 75 percent of screen height, based on interpretation of indications causing complete loss-of-back reflection. Rejection er shear wave inspection was based on indications, exceeding the amplitude of the indication m a calibration notch whose depth is three percent of the forging thickness not exceeding inch with a length of 1 inch.
forgings were also subjected to magnetic-particle examination or liquid-penetrant testing ending upon the material. Rejection was based on Section III of the ASME Code, paragraph 6.3 for magnetic-particle and paragraph N627.3 for dye penetrant testing.
es were ultrasonically tested using longitudinal ultrasonic testing techniques. Rejection under gitudinal beam testing performed in accordance with ASME Code, with calibration so that the back reflection is at least 50 percent of screen height, was based on defects causing complete of back reflection. Any defect which showed a total loss of back reflection which could not contained within a circle whose diameter is the greater of three inches or one-half the plate kness was unacceptable. Two or more defects smaller than described above which caused a plete loss-of-back reflection were unacceptable unless separated by a minimum distance al to the greatest diameter of the larger defect unless the defects were contained within the area cribed above.
destructive testing of the vessel was performed during several stages of fabrication with strict lity control in critical areas such as frequent calibration of test instruments, metallurgical ection of all weld rod and wire, and strict adherence to the nondestructive testing uirements of Section III of the ASME Boiler and Pressure Vessel Code.
detection of flaws in irregular geometries was facilitated because most nondestructive testing he materials was completed while the material was in its simplest form. Nondestructive ection during fabrication was scheduled so that full penetration welds were capable of being ographed to the extent required by Section III of the ASME Boiler and Pressure Vessel Code.
destructive tests based upon Section III of the ASME Boiler and Pressure Vessel Code 1998 tion through 2000 Addenda were performed on the forging and welds.
asonic and magnetic particle examinations were performed on the head forging in accordance h ASME Section III, NB 2000 and NB 5000 respectively. Head cladding was ultrasonically mined for both bond and defects using a calibration block typical of the cladding and base erial. Any indications that produce amplitude equal to or greater than the amplitude received m the three-eighths inch flat bottom hole, regardless of length, were unacceptable (see le 4.6-11 for additional owners inspection requirements).
ssure retaining welds were 100% radiographed and liquid penetrant examined in accordance h NB 5000. Canopy seal welds were liquid penetrant examined. J-groove welds were liquid etrant examined at half thickness and again at final surface. No indications were allowed for final PT of the J welds on the CEDM and instrument and vent nozzles and attachments of hanism housings. Hydrostatic tests at the shop were conducted to 3107 psig. Visual minations, magnetic particle and liquid penetrant tests were performed to reveal any surface ontinuities.
addition to the pre-service examinations required by Section XI of the ASME Boiler and ssure Vessel Code, augmented ultrasonic and eddy current examination were performed on the nel nozzle bore material to meet the examination requirements of NRC Bulletin 2001-01, rcumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles, August 3, 1 and NRC Bulletin 2002-01, Reactor Pressure Vessel Head Degradation and Reactor lant Pressure Boundary Integrity March 18, 2002.
h of the vessel studs received one ultrasonic test and one magnetic-particle inspection during manufacturing process.
ultrasonic test was a radial longitudinal beam inspection, and a discontinuity which causes an cation with a height which exceeded 20 percent of the height of the adjusted first back ection was cause for rejection. Any discontinuity which prevents the production of a first back ection of 50 percent of the screen height was also cause for rejection.
magnetic-particle inspection was performed on the finished studs. Linear axially aligned cts whose lengths are greater than one inch long and linear nonaxial defects were cceptable.
vessel studs are elongated by the stud tensioners during each installation of the vessel head.
amount of elongation for the desired preload was specified by the vendor to fall within a determined acceptance range. Maintenance procedures for vessel head installation are in full pliance with the vendor specified range.
replacement pressurizer assembly consists of upper and lower heads and two shells attached circumferential seam welds. The shells and the heads are forged components. The large bore
t is greater than one t limit of the ASME Boiler and Pressure Vessel Code Section XI, section IWB jurisdictional boundary. Therefore ASME Section XI, Subsection IWF sdiction applies to the attachment weld.
destructive examinations were performed on forgings and welds in accordance with ASME tion III, Boiler and Pressure Vessel Code 1998 Edition through 2000 Addenda. Ultrasonic and netic particle examinations were performed on the forged components in accordance with ME Section III, NB 2000 and 5000. The heater sleeves, vent and instrument nozzles are J-ove weld attachments to the reinforced weld buildup of the replacement pressurizer. The ial penetration welds were examined by liquid penetrant tests in accordance with NB-5245.
electrical heater sheaths were completely liquid penetrant inspected in accordance with NB-
- 0. In addition to the code requirements, additional inspection criteria listed in Table 4.6-11 for pressurizer were required for acceptability.
rostatic test of the replacement pressurizer was conducted at 3125 psia.
r to and during fabrication of the components of the RCS, nondestructive testing based upon requirements of Section III of the ASME Boiler and Pressure Vessel Code is used to determine acceptance criteria for various size flaws. The requirements for the Class A vessels are the e as the reactor vessel. Vessels designated as Class C were fabricated to the standards of section C, Article 21 of Section III of the ASME Code.
le 4.6-10 summarizes the components inspection program during fabrication and construction.
ddition to the inspections conducted during fabrication and construction of the pressurizer wn in Table 4.6-10, pre-service inspections were performed on several components of the acement pressurizer. UT inspections were performed on the circumferential welds of the surizer shell and heads. Magnetic particle inspection of the skirt to vessel was performed. UT PT inspections were performed on the safety, relief, spray and surge nozzles.
odic tests and examinations of the RCS are conducted after startup on a regular basis.
preoperational and in-service structure surveillance of the RCS, refer to the Technical cifications; tests for RCS integrity after the system is closed following normal opening, dification or repair are specified in Technical Specifications.
5 ADDITIONAL TESTS ing design and fabrication of the reactor vessel, additional operations beyond the requirements he ASME Boiler and Pressure Vessel Code,Section III were performed by the vendor.
le 4.6-11 summarizes the additional tests by components.
ing the design of the reactor vessel, detailed calculations were performed to assure that the l product would have adequate design margins. A detailed fatigue analysis of the vessel for all
mbustion Engineering (CE) has performed test programs for the determination and verification nalytical solutions to thermal stress problems. Also fracture mechanics and brittle fracture luations have been performed.
thermal and structural analyses of CEDM and ICI head adapters of the replacement reactor sel closure head were performed using finite element models. Structural responses from loads uding branch line pipe break, faulted conditions (combination of line break and seismic),
mal operating mechanical and acoustic excitation were evaluated. ASME Appendix G luations were performed assuming meridional and circumferential flaws at critical locations of reactor vessel closure head. Detailed fatigue analysis using finite element models of bimetallic ds at the interface between alloy 690 and stainless steel for the ICI nozzles were performed.
material used in the reactor vessel was carefully selected and precaution were taken by the sel fabricator to ensure that all material specifications were adhered to. To assure compliance, quality control staff of CE reviewed the mill test reports and the fabricators testing cedures.
welding methods, materials, techniques, and inspections comply with Sections III and IX of ASME Boiler and Pressure Vessel Code. Before fabrication was begun, detailed qualified ding procedures, including methods of joint preparation, together with certified procedure lification test reports, were prepared. Also, prior to fabrication, certified performance lification tests were obtained for each welder and welding operator. Quality control was rcised for all welders and welds by subjection to a complete and thorough testing program in er to ensure maximum quality of welded joints.
ing the manufacture of the reactor vessel, in addition to the areas covered by the ASME Boiler Pressure Vessel Code,Section III, quality control by the vendor included:
- a. preparation of detailed purchase specifications which included cooling rates for test samples;
- b. requiring vacuum degassing for all ferritic plates and forgings;
- c. specification of fabrication instructions for plates and forgings to provide control of material prior to receipt and during fabrication;
- d. use of written instructions and manufacturing procedures which enable continual review based on past and current manufacturing experiences;
- e. performance of chemical analysis of welding electrodes, welding wire, and materials for automatic welding, thereby providing continuous control over welding materials;
- g. and test programs on fabrication of plates up to 15 inches thick to provide information about material properties as thickness increases.
wave ultrasonic testing was performed on 100 percent of all plate material.
dding for the reactor vessel is a continuous integral surface of corrosion-resistant material, inch nominal thickness. The detailed procedure used, i.e., type of weld rod, welding position, ed of welding, nondestructive testing requirements, etc., was in compliance with the ASME ler and Pressure Vessel Code. The cladding is ultrasonically inspected for lack of bond at rvals not to exceed 12 inches WR to the direction of welding. Unbonded areas equal to or in ess of calibration require additional scanning of the surrounding material until the boundary of discontinuity is established. An area of unbounded clad in excess of acceptance standards is ired.
n completion of all postweld heat treatments, the reactor vessel was hydrostatically tested, r which all weld surfaces, including those of welds used to repair material, were magnetic-icle inspected in accordance with Section III, paragraph N-618 of the ASME Boiler and ssure Vessel Code.
veillance of the quality control program was also carried out during the manufacture of the sel by the Windsor Quality Control Section of CE and by Northeast Nuclear Energy Company ECO) with an independent consultant. This work included independent review of ographs, magnetic-particle tests, ultrasonic tests, and dye penetrant tests conducted during the ufacture of the vessel. A review of material certifications and vendor manufacturing and ing procedures was also conducted. Manufacturers records such as heat-treat logs, personnel lification files and deviation files were also included in this review.
nominal cladding thickness of the replacement reactor vessel closure head following hining (i.e., grinding) is quarter inch.
cladding was ultrasonically examined such that each pass of the scanner overlap a minimum percent of the transducer dimension perpendicular to the direction of the scan. Magnetic icle and liquid penetrant tests were conducted following hydrostatic test of the reactor vessel d to find any surface discontinuities. The magnetic particle and liquid penetrant tests were ducted in accordance with ASME Section III subsection NB 5000.
replacement head was fabricated and inspected by Mitsubishi Heavy Industries at their lity. Representatives from Dominion Nuclear, CT (DNC) were present to witness the weld ections and hydrostatic tests at various hold points. DNC also reviewed material ifications, test results conducted during fabrication.
replacement pressurizer has a minimum stainless steel cladding of 1/5 inch. The cladding was asonically examined such that each pass of the scanner overlap a minimum 10 percent of the
e conducted in accordance with ASME Section III subsection NB 5000. In addition to the d inspections, DNC personnel witnessed the hydrostatic test at various hold points at the icators shop for the replacement pressurizer.
6 IN-SERVICE INSPECTION reoperational inspection was performed in compliance with Section XI of the ASME Boiler Pressure Vessel Code, In-Service Inspection of Nuclear Reactor Coolant Systems, 1971, re possible. CE and Southwest Research Institute were retained to assist in the development he in-service inspection program. Access was provided, where possible, to permit inspection the areas listed in the code. In-service inspections are performed in accordance with tion XI.
rge portion of the insulation for the reactor vessel has been placed on the reactor cavity wall ermit inspection of the vessel outer surface. Essentially all vessel internals can be removed so a complete visual internal inspection is possible and a volumetric internal inspection of the sel is also possible. Pads have been welded to the outer surface of the reactor vessel to litate prompt location of welds for inspection purposes.
ess openings are provided on the permanent reactor cavity seal and the neutron shielding und the reactor vessel to facilitate inspection and maintenance of the neutron detector wells ng refueling.
logical shielding around the primary piping in the area of the reactor pressure vessel has been gned to afford access to the circumferential and RW welds, as well as the transition piece-to-zle welds.
primary piping, as well as major components, excluding the reactor pressure vessel, have been vided with easily removable insulation in the areas of all welds and adjacent base metal uiring examination as defined by Section XI. Removable blanket insulation is provided on the DM, vent and instrumentation nozzles areas of the replacement reactor vessel closure head.
nt arrangement and piping has been designed to assure that adequate access exists for either ct personnel access or for remote handling equipment to perform the examinations required Section XI. Service connections, e.g., air, water, electricity, have been located adjacent or in e proximity to each inspection area. Consideration has been given to the type of examination equipment requirements.
ess holes have been provided in the support skirt of each steam generator to provide a means xamining the tube sheet support stay cylinder weld.
MILLSTONE UNIT NUMBER 2 Piece Number UPPER SHELF (Reference DROP ENERGY FO Drawing E- CODE WEIGHT RTNDT LONGITUDIN 233-426-1) Number HEAT Number VESSEL LOCATION NDTT T50 (F) T35 (F) TCV (F) (F) DIRECTION 203-02 C-500 4P2989 Vessel Flange +10F 5 50 70 10 149 5P3286-4310V1 204-02 C-511 C-5823-3A Bottom Head Dome -70 61 68 88 28 115 204-03A C-510-1 C-5823-3B Bottom Head Peel -60 -20 -2 18 -42 150 204-03C C-510-2 C-5892-3A -10 -22 -20 0 -60 135 204-03B C-510-3 C-5892-3B -40 20 20 40 -20 140 204-03D C-510-1 C-5823-3B -60 20 -2 18 -42 150 204-03E C-510-2 C-5892-3A -10 -22 -20 0 -60 135 204-03F C-510-3 C-5892-3B -40 20 20 40 -20 140 205-02C C-503-1 9-7395-1-1 Inlet Nozzles -20 45 10 65 5 111 205-02B C-503-2 9-7401-1-2 +10 30 12 50 10 120 205-02A C-503-3 9-7454-1-3 -20 12 12 32 -28 108 205-02D C-503-4 9-7458-1-4 -60 30 20 50 -10 123 205-03A C-508-3 AV2999-9G-1283 Inlet Nozzle Extension -20F 30 2 50 -10 146 205-03B C-508-1 AV2999-9G-1281 - 30 2 50 -10 205-03C C-508-2 AV2999-9G-1282 30 2 50 -10 205-03D C-508-4 AV2999-9G-1262 30 2 50 -10 205-06A C-502-1 9-7356-001 Outlet Nozzles -110 -20 -30 0 -60 132 205-06B C-502-2 9-7375-002 -130 16 -10 36 -24 103 205-07A C-509-1 AV3816-9G-1239 Outlet Nozzle Exten. -20 25 15 45 -15 141 205-07B C-509-2 AV3816-9G-1391 25 15 45 -15 215-01C C-504-1 C-5804-2 Upper Shell +10 62 53 82 22 118 215-01A C-504-2 C-5809-2 +10 62 48 82 22 131
MILLSTONE UNIT NUMBER 2 (CONTINUED)
Piece Number UPPER SHELF (Reference DROP ENERGY FO Drawing E- CODE WEIGHT RTNDT LONGITUDIN 233-426-1) Number HEAT Number VESSEL LOCATION NDTT T50 (F) T35 (F) TCV (F) (F) DIRECTION 215-01B C-504-3 C-5809-1 -10 55 52 75 15 125 215-02A C-505-1 C-5843-1 Intermediate Shell -20 - - - 8.1 118 215-02B C-505-2 C-5843-2 -10 - - - 17.5 123 215-02C C-505-3 C-5843-3 -10 - - - 5 115 215-03A C-506-1 C-5667-1 1/4T Lower Shell +10 - - - 7 112 215-03C C-506-2 C-5667-2 1/4T Lower Shell -40F - - - -33.7 135 215-03B C-506-3 A-5518-1 1/4T -30 - - - -19.2 136 A1 03W62-1-1 Replacement Closure Head - 40 - 44 - 70 - - 40 148 B1 03W62-1-1 Replacement Closure Head - 40 - 18 - 22 - - 40 139 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.
TCV (F) reflects the greater of T50 and T35 and incorporates guidance from MTEB 5-2 where necessary based upon available data.
MILLSTONE UNIT NUMBER 2 Charpy Heat Drop Wt/ V-notch Test Charpy V-notch Lateral Expansi Description of Part Number RTNDT F(1) Temp F Test Orientation Impact Energy (ft-lbs) (mils)
Upper Head T4987 -8 (+19) +52 0 147-147-147 90-87-87 180 160-172-157 87-94-90 Upper Shell T5086 +1 (+28) +61 0 142-162-155 90-98-94 180 144-128-125 90-87-90 Lower Shell T5085 -8 (+19) +52 0 137-140-125 94-90-87 180 127-111-113 87-79-79 Lower Head T4986 -17 (+10) +43 0 177-169-153 94-90-87 180 169-169-138 87-90-87 Support Skirt 4-1575 - +10 Base Ring 179-165-177 87-83-87 Skirt A 134-131-134 79-78-78 Skirt B 131-171-131 76-89-76 Manway / Vent 11746 - +59 Vent 116-119-115 79-83-83 Covers Manway 119-159-127 85-94-89 Lifting Lug 11764 - +59 Lug 1 130-167-133 91-94-93 Lug 2 136-144-174 93-94-94 Manway Stud N9952 - +10 - 37.6-37.6-36.9 29.3-28.9-28.1 Manway Nut 81025 - +10 - 50.1-48-7-49.4 41.0-39.5-42.2 Ventport Stud N9879 - +10 - 36.9-36.1-36.1 27.0-26.6-26.2 Ventport Nut 81025 - +10 - 50.1-48.7-49.4 41.0-39-5-42.2 Notes:
(1) Per the replacement pressurizer vendor, the crack started welds for drop weight testing specimens were deposited in accordance with the 1981 edition of ASTM St Test Method E 208 instead of the specified 1991 edition. A conservative shift in RTNDT of 27°F above the values reported by the fabricator CMTRs has been dete to bound the resulting uncertainty in RTNDT values. The adjusted values are shown in parentheses and were used in relevant ASME Code and analyses.
Steam LOCATION/ Generator RT NDT (F) CHARPY V-NOTCH IMPACT TEST CVN TEST PART Number HEAT MATERIAL Unit Number (1) RESULTS (2) TEMP F NO 5056021 W70029-1 Tubesheet / SA508 2 -70 126.9, 118.8, 104.4 146.5, 125.0, 94.0 -10F Class 3 5056021 W70022-1 1 -60 130.9, 146.5, 152.1 144.6, 134.9, 130.9 0 5056020-2 88D108-1-1 Primary Head / 2 124.6, 147.5, 142.3 136.4, 142.3, 132.7 40 SA508, Class 3 (La) (La) 88C115 195.4, 190.3, 147.5 40 (La)
-30 114.3, 95.8, 112 104, 120, 119 (Me) 30 (ME) 0 173, 181, 151 (ME) 60 5056020-1 88D101-1-1 1 129, 111, 136 (La) 122, 125, 123 (La) 40 88C110 157, 131, 123 (La) 40
-30 90, 98, 98 (ME) 30
-40 98.8, 90, 107 (ME) 20 0 142, 134, 164 (ME) 60 5056024-1 727957 Stay Cylinder / 1 98.1, 95.8, 97.3 112, 98.1, 102 +40 SA508, Class 3 120.2, 113, 116 126, 90.7, 101.7 +40
-35 99.5, 112, 110 95.8, 72.2, 116 +25
-35 129, 135.7, 127.6 123.9, 128.3, 131 +25
Steam LOCATION/ Generator RT NDT (F) CHARPY V-NOTCH IMPACT TEST CVN TEST PART Number HEAT MATERIAL Unit Number (1) RESULTS (2) TEMP F NO 5056024-2 727957 2 -27 106.9, 118, 112.8 104.7, 109.1, 113.6 +33
-35 104.7, 104.7, 109.1 +25
-27 115, 118, 120.9 +33 106, 110, 120 99.5, 104.7, 118 +40 122.4, 107.6, 112.8 126, 102, 109 +40 5056025-1 724015 Safe End-Inlet / 1 -8F 130.5, 127.6, 118 +40 SA508, Class 1 (TANG)
-8F 99.6, 80.4, 110.6 +52 5056025-2 724015 2 -8F 103.2, 118, 109.1 +40 (TANG)
-8F 103.3, 89.9, 97.36 +52 5056026-1, 2, 724015 Safe End-Outlet / 1&2 -8F 115.8, 105.5, 109.1 99.5, 95.88, 88.5 +40 3, 4 SA508, Class 1 (TANG)
+52 5069670-2 87633-2 Manway 1 -22F 109, 121, 124 (Top) +37.4 5069670- SA533, Grade B, -49 108, 108, 91 +10 Class 1 (Bottom) 5069670-5 87659-2 Manway SA 533, 1+2 -49 107.7, 97.5, 115 +10 Grade B, Class 1 (Top)
Steam LOCATION/ Generator RT NDT (F) CHARPY V-NOTCH IMPACT TEST CVN TEST PART Number HEAT MATERIAL Unit Number (1) RESULTS (2) TEMP F NO 5069679-1 87659-2 -40 137, 125.5, 119 +20 (Bottom) 5069670-3 87660-2 2 -40 143, 103, 107 (Top) +10
-31 106, 165, 140 +30 (Bottom) 5062982-1, 2 37314 Vessel Support A533, 1 + 2 -40 181, 178, 140 (Top) +20 Grade B, Class 1 (66961) -31 184, 200, 187 +30 (Bottom) 5062980-1, 2, 37410 1+2 -22 177, 190, 151 (Top) +40 3, 4, 5, 6 (66922) -22 178, 160, 182 +40 (Bottom) 5056024-1 727957 Inlet Nozzle / SA508, 1 124, 113, 115 123, 129, 125 +40 Class 3
-45 71.5, 67.8, 81.8 73, 84.8, 70 -15 5056024-2 727957 2 134, 152, 136 178, 205, 179 +40
-62 58, 89, 87 88, 84, 60 -3 5056023-1 727957 Outlet Nozzle / 1 110, 111, 118 +40 SA508, Class 3
-36 59, 74.5, 73 +42
Steam LOCATION/ Generator RT NDT (F) CHARPY V-NOTCH IMPACT TEST CVN TEST PART Number HEAT MATERIAL Unit Number (1) RESULTS (2) TEMP F NO 5056023-2 727957 2 116, 125, 116 +40
-18 93, 112, 82.6 +42 5056023-3 727957 1 118, 119, 135 +40
-45 75, 67, 86 +15 5056023-4 727957 2 112, 122, 118 +40
-35 81.1, 85.56, 76.7 +25 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 and remove testing in accordance with ASME Section III requirements. Specimen location and orientation 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.
Assembly PIECE CHARPY V-NOTCH Tempe Number Number CODE Number DESCRIPTION MATERIAL VALUES - FT-LB F 503-01 502-20-1 C-4601-1 Pipe Segment SA-516 Grade 70 50 53 49 +10 Hot Leg 502-20-2 C-4601-2 Pipe Segment Grade 70 37 34 38 +10 502-02-1 C-4605-1 Ell Segment Grade 70 55 41 61 +10 502-02-2 C-4605-1 Ell Segment Grade 70 55 41 61 +10 506-02 C-4613-1 Nozzle Forging A 105 Grade 2 43 +10 512-03 C-4611 Nozzle Forging Grade 2 43 36 32 +10 503-02 502-02-3 C-4605 Ell Segment SA 516 Grade 70 55 41 61 +10 Hot Leg 502-02-4 C-4605 Ell Segment Grade 70 55 41 61 +10 502-20-3 C-4601-3 Pipe Segment Grade 70 41 47 43 +10 502-20-4 C-4601-4 Pipe Segment Grade 70 45 47 51 +10 506-08 C-4612-1 Nozzle Forging A 105 Grade 2 24 30 28 +10 504-01 502-12-1 C-4604-1 Pipe Segment SA 516 Grade 70 58 61 63 +10 Cold Leg 502-12-2 C-4604-2 Pipe Segment Grade 70 36 36 36 +10 Pump 502-08-1 C-4608 Ell Segment Grade 70 34 30 39 +10 Discharge 502-08-2 C-4608 Ell Segment Grade 70 34 30 39 +10 507-02-1 C-4615-1 Nozzle Forging SA 105 Grade 2 43 36 32 +10 508-02-1 C-4610-1 Nozzle Forging SA 182 Grade F1 98 95 104 +10 504-03 502-10-9 C-4609-2 Ell Segment SA 516 Grade 70 64 41 66 +10
Assembly PIECE CHARPY V-NOTCH Tempe Number Number CODE Number DESCRIPTION MATERIAL VALUES - FT-LB F Cold Leg 502-10-10 C-4609-2 Ell Segment Grade 70 64 41 66 +10 Pump 502-18-1 C-4602-1 Pipe Segment Grade 70 58 67 62 +10 Discharge 502-18-2 C-4602-2 Pipe Segment Grade 70 58 67 62 +10 508-02-2 C-4610-2 Nozzle Forging SA 182 Grade F1 95 94 110 +10 504-04 502-12-3 C-4604-3 Pipe Segment SA 516 Grade 70 59 66 60 +10 Cold Leg 502-12-4 C-4604-4 Pipe Segment Grade 70 59 66 60 +10 Pump 502-08-3 C-4608 Ell Segment Grade 70 34 30 39 +10 Discharge 502-08-4 C-4608 Ell Segment Grade 70 34 30 39 +10 507-02-2 C-4615-2 Nozzle Forging SA 105 Grade 2 43 36 32 +10 508-02-3 C-4610-3 Nozzle Forging SA 182 Grade F1 106 120 114 +10 507-07-2 C-4616-2R Nozzle Forging SA 105 Grade 2 43 36 32 +10 504-05 502-10-11 C-4609-2 Ell Segment SA 516 Grade 70 64 41 66 +10 Cold Leg 502-10-12 C-4609-2 Ell Segment Grade 70 64 41 66 +10 Pump 502-18-3 C-4602-3 Pipe Segment Grade 70 58 61 57 +10 Discharge 502-18-4 C-4602-4 Pipe Segment Grade 70 58 61 57 +10 508-02-4 C-4610-4 Nozzle Forging SA 182 Grade F1 109 76 92 +10 507-07-1 C-4616-1 Nozzle Forging SA 105 Grade 2 43 36 32 +10 503-05-1 502-04-1 C-4606-1 Ell Segment SA 516 Grade 70 58 51 50 +10 Cold Leg 502-04-2 C-4606-1 Ell Segment Grade 70 58 51 50 +10 Pump Suction
Assembly PIECE CHARPY V-NOTCH Tempe Number Number CODE Number DESCRIPTION MATERIAL VALUES - FT-LB F 503-05-2 502-04-3 C-4606-1 Ell Segment SA 516 Grade 70 58 51 50 +10 Cold Leg 502-04-4 C-4606-1 Ell Segment Grade 70 58 51 50 +10 Pump Suction 503-05-3 502-04-5 C-4606-2 Ell Segment SA 516 Grade 70 44 49 53 +10 Cold Leg 502-04-6 C-4606-2 Ell Segment Grade 70 44 49 53 +10 Pump Suction 503-03-1 502-06-1 C-4607 Ell Segment SA 516 Grade 70 47 55 52 +10 503-05-4 502-04-7 C-4606-2 Ell Segment SA 516 Grade 70 44 49 53 +10 Cold Leg 502-04-8 C-4606-2 Ell Segment Grade 70 44 49 53 +10 Pump Suction Cold Leg 502-06-2 C-4607 Ell Segment Grade 70 47 55 52 +10 Pump Suction 502-16-1 C-4603-1 Pipe Segment SA 516 63 70 64 +10 502-16-2 C-4603-2 Pipe Segment Grade 70 63 70 64 +10 503-03-2 502-06-3 C-4607 Ell Segment SA 516 Grade 70 47 55 52 +10 Cold Leg 502-06-4 C-4607 Ell Segment Grade 70 47 55 52 +10 Pump Suction 502-16-1 C-4603-1 Pipe Segment Grade 70 63 70 64 +10 502-16-2 C-4603-2 Pipe Segment Grade 70 63 70 64 +10 503-03-3 502-06-5 C-4607 Ell Segment SA 516 Grade 70 47 55 52 +10 Cold Leg 502-06-6 C-4607 Ell Segment Grade 70 47 55 52 +10 Pump Suction 502-16-1 C-4603-1 Pipe Segment Grade 70 63 70 64 +10 502-16-2 C-4603-2 Pipe Segment Grade 70 63 70 64 +10
Assembly PIECE CHARPY V-NOTCH Tempe Number Number CODE Number DESCRIPTION MATERIAL VALUES - FT-LB F 503-03-4 502-06-7 C-4607 Ell Segment SA 516 Grade 70 47 55 52 +10 Cold Leg 502-06-8 C-4607 Ell Segment Grade 70 47 55 52 +10 Pump Suction 502-16-1 C-4603-1 Pipe Segment Grade 70 63 70 64 +10 502-16-2 C-4603-2 Pipe Segment Grade 70 63 70 64 503-07-1 502-14-1 C-4604-5 Pipe Segment SA 516 Grade 70 23 21 20 +10 Cold Leg 502-14-2 C-4604-6 Pipe Segment Grade 70 52 57 54 +10 Pump Suction 502-10-1 C-4609-1 Pipe Segment Grade 70 50 43 35 +10 502-10-2 C-4609-1 Pipe Segment Grade 70 50 43 35 +10 507-10-4 C-4614-4R Nozzle Forging SA 105 Grade 2 38 42 43 +10 503-07-2 502-14-1 C-4604-5 Pipe Segment SA 516 Grade 70 23 21 20 +10 Cold Leg 502-14-2 C-4604-6 Pipe Segment Grade 70 52 57 54 +10 Pump Suction 502-10-3 C-4609-1 Ell Segment Grade 70 50 43 35 +10 502-10-4 C-4609-1 Ell Segment Grade 70 50 43 35 +10 507-10-1 C-4614-3R Nozzle Forging SA 105 Grade 2 38 42 43 +10 505-10-1 502-14-3 C-4604-7 Pipe Segment SA 516 Grade 70 52 57 54 +10 Cold Leg 502-14-4 C-4604-8 Pipe Segment Grade 70 58 61 63 +10 Pump Suction 502-10-5 C-4609-1 Ell Segment Grade 70 50 43 35 +10 502-10-6 C-4609-1 Ell Segment Grade 70 50 43 35 +10 507-10-2 C-4614-2 Nozzle Forging SA 105 Grade 2 43 36 32 +10 505-10-2 502-14-3 C-4604-7 Pipe Segment SA 516 Grade 70 52 57 54 +10
Assembly PIECE CHARPY V-NOTCH Tempe Number Number CODE Number DESCRIPTION MATERIAL VALUES - FT-LB F Cold Leg 502-14-4 C-4604-8 Pipe Segment Grade 70 58 61 63 +10 Pump Suction 502-10-7 C-4609-2 Ell Segment Grade 70 64 41 66 +10 502-10-8 C-4609-2 Ell Segment Grade 70 64 41 66 +10 507-10-3 C-4614-1 Nozzle Forging SA 105 Grade 2 43 36 32 +10
Weight Percent Plate 1/4 T-ID Weld 1/4 T-OD Weld Element C-506-1 C-506-2/C-506-3 C-506-2/C-506-3 Si 0.12 0.17 0.15 S 0.014 0.013 0.013 P 0.006 0.015 0.016 Mn 1.26 1.13 1.13 C 0.21 0.12 0.12 Cr 0.10 0.04 0.05 Ni 0.61 0.06 0.06 Mo 0.62 0.54 0.53 V 0.004 0.006 0.007 Cb < 0.01 < 0.01 < 0.01 B 0.0006 0.0003 0.0003 Co 0.011 0.009 0.009 Cu 0.14 0.30 0.21 Al 0.020 < 0.001 < 0.01 W < 0.01 0.01 < 0.01 Ti < 0.01 < 0.01 < 0.01 As 0.011 0.011 0.012 Sn 0.009 0.004 0.003 Zr 0.002 0.002 0.002 N2 0.009 0.008 0.009
NDTT 30 ft-lb 50 ft-lb 35 Mils Lat. RTNDT Upper Shelf RT Yi Material Code Number Orientation a (F) Fix (F) Fix (F) Exp. Fix (F) (F) (ft-lb) (ksi Base Metal C-506-1 Transverse -10 22 66 54 6 106.5 67 Plate (WR)
Base Metal C-506-1 Longitudinal -10 36 84 65 24 b 124.5 67 Plate (RW)
Weld Metal C-506-2 / C-506- Transverse -60 -26 5 2 -55 129.5 76 3 (WR)
HAZ Metal C-506-1 Transverse -30 -44 35 34 -25 123.0 68 (WR)
- a. With respect to the plates major rolling direction for base metal; with respect to the welding direction for weld and HAZ.
- b. Not valid per 10 CFR 50, Appendix G.
Plate Code Test Temp Yield Strength Tensile Strength Elongation R.A.
a Material Number Orientation (F) (ksi) (ksi) TE(%)/UE(%) (%)
Base Metal C-506-1 Transverse (WR) 67 88 27/11 68 Plate 250 64 81 25/10 67 550 58 84 23/09 62 Base Metal C-506-1 Longitudinal 71 67 86 29/12 71 (RW) 250 61 79 26/10 70 550 56 83 26/10 69 Weld Metal C-506-2/C-506-3 Longitudinal 71 76 86 28/11 74 (RW) 250 74 81 26/09 71 550 67 85 25/10 65 HAZ Metal C-506-1 Transverse (WR) 71 68 88 24/09 70 250 60 80 25/08 72 550 62 83 21/07 68
- a. With respect to the plates major rolling direction for base metal; with respect to the welding direction for weld and HAZ.
Base Base Capsule Metal Metal Base Weld Weld HAZ Location on Impact Impact Metal Metal Metal Impac HAZ Reference Total Specimens Vessel Wall L (a) T (b) Tensile Impact Tensile t Tensile Impact (c) Impact Tensile 83 12 12 3 12 3 12 3 - 48 9 97 12 12 3 12 3 12 3 - 48 9 104 12 - 3 12 3 12 3 12 48 9 263 12 - 3 12 3 12 3 12 48 9 277 12 12 3 12 3 12 3 - 48 9 284 12 12 3 12 3 12 3 - 48 9 72 48 18 72 18 72 18 24 288 54 (a) L = Longitudinal (b) T = Transverse (c) Reference material correlation monitors
REMOVAL LEAD TIME FLUENCE APSULE LOCATION FACTOR a (EFPY) b (n/cm2, >1.0MeV) (a)
W-97 97 1.40 3.0 3.24 x 1018 c W-104 104 0.95 10.0 9.49 x 1018 (c)
W-97 d 97 10.0 --
W-83 83 1.31 15.3 1.74 x 1019 (c)
W-277 277 1.31 EOL e See Note (d)
W-263 263 1.31 Standby --
W-284 284 0.97 Standby --
Updated in Capsule W-83 dosimetry analysis Effective Full Power Years (EFPY) from plant startup Plant specific evaluation Flux Monitor EOL is defined as the end-of-license period 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 1019 n/cm2 (E > 1.0 MeV). Capsule W-277 will receive 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 1019 n/cm2 (E > 1.0 MeV).
ABLE 4.6-10 INSPECTION OF REACTOR COOLANT SYSTEM COMPONENTS DURING FABRICATION AND CONSTRUCTION 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 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 Tubes UT, ET Nozzles (Forgings) UT, MT Studs MT Welds Shell, Longitudinal RT, MT Shell, Circumferential RT, MT Liquid penetrant tests of J-welds only.
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 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 Temporary Attachments After Removal MT Heads to Shell and Shell to Shell Welds PT Heater Assembly RT, PT Pumps Castings RT, PT Forgings UT, PT Welds Circumferential RT, PT Instrument Connections PT All Welds After Hydrostatic Test PT
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 end:
RT -Radiographic PT - Dye Penetrant ET - Eddy Current UT - Ultrasonic MT - Magnetic Particle
BLE 4.6-11 REACTOR COOLANT SYSTEM INSPECTION C-E REQUIREMENTS ctor Vessel
<test> C-E Requirements Code Requirements rasonic Testing (UT) 1. UT of Weld Clad for bond. 1. None lacement Reactor Vessel Head (a)
<test> C-E Requirements Code Requirements rasonic Testing (UT)) 1. Indications that produce an ASME Section V, Article 23, SA amplitude greater than the 578 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.
Depth from Clad Surface Length of Indication Up to 0.02 inches 0.375 inches 0.02 inches - 0.06 inches 1.0 inches 0.06 inches - 0.1 inches 3.0 inches Over 0.10 inches 6.0 inches
lacement Reactor Vessel Head (continued)
<test> C-E Requirements Code Requirements uid Penetrant Test 1. The final PT of J-weld and the half-inch area ASME Section III, T) adjacent to the weld for CEDM adapters, NB 5245 instrument tube connections shall allow no indications of defects.
m Generator
<test> C-E Requirements Code Requirements rasonic Test (UT) 1. UT for Defects in Tube Sheet Clad UT of 1. 1. None Weld Clad for bond.
- 2. UT of Weld Clad for bond 2. None lacement Pressurizer (a)
<test> C-E Requirements Code Requirements rasonic Testing 1. Indications that produce an amplitude greater ASME Section V, T) than the amplitude received from the one-eighth Article 23, SA 578 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.
Depth from Clad Surface Length of Indication Up to 0.03 inches 0.375 inches 0.03 inch - 0.06 inch 1.0 inches 0.06 inch - 0.125 inch 3.0 inches Over 0.125 inch 6.0 inches This is a Dominion requirement and not a CE requirement. The reactor vessel head and acement pressurizer were replaced in accordance with Dominion Purchase Specifications.
TABLE 4.6-13 RTPTS VALUES AT 54 EFPY Neutron Initial Fluence Chemical Chemical Chemist Margin Component Content: Content: Ni ry RTNDT Term (10n/ RTPTS Vessel Location Identification Cu (%) (%) Factor (F) (F) cm2) (F)
Intermediate C-505-1 0.13 0.61 91.3 8.1 34 3.83 165.1 Shell C-505-2 0.13 0.62 91.5 17.5 34 3.83 174.7 Course Plates C-505-3 0.13 0.62 91.5 5 34 3.83 162.2 Lower C-506-1 0.15 0.60 110.0 7 34 3.78 188.8 Shell C-506-2 0.15 0.61 110.0 -33.7 34 3.78 163.5 Course Plates C-506-3 0.14 0.66 101.5 -19.2 34 3.78 183.6 Intermediate Shell Axial 2-203 A (Heat A8746) 0.15 0.13 77.7 -56 66 2.83 114.6 Welds 2-203 B/C (Heat 0.15 0.13 77.7 -56 66 3.53 114.6 A8746)
Lower Shell Axial Welds 3-203 A (Heat A8746) 0.15 0.13 77.7 -56 66 3.83 114.6 3-203 B/C (Heat 0.15 0.13 77.7 -56 66 2.50 114.6 A8746)
Intermediate-to- Lower 9-203 (Heat 10137) 0.22 0.04 100.0 -56.3 56 3.78 134.1 Shell Girth Weld 9-203 (Heat 90136) 0.27 0.07 124.3 -56.3 56 3.78 166.7
Neutron Fluence (a), 1019 n/cm2 (E>1 ART Projections (b) 54 MeV) 54 EFPY EFPY Component Vessel Location Identification 1/4T 3/4T 1/4T 3/4T Intermediate C-505-1 2.283 0.811 153.8 128.0 Shell C-505-2 2.283 0.811 163.4 137.6 Course Plates C-505-3 2.283 0.811 150.9 125.1 Lower C-506-1 2.253 0.809 175.2 144.1 Shell C-506-2 2.253 0.809 134.5 103.4 Course Plates C-506-3 2.253 0.809 138.6 110.0 Intermediate Shell Axial Welds 2-203 A (Heat A8746) 2.283 3.811 104.6 82.6 2-203 B (Heat A8746) 1.508 0.536 96.0 73.7 2-203 C (Heat A8746) 1.508 0.536 96.0 73.7 Lower Shell Axial Welds 3-203 A (Heat A8746) 2.283 0.811 104.6 82.6 3-203 B (Heat A8746) 1.490 0.529 95.8 73.4 3-203 C (Heat A8746) 1.490 0.529 95.8 73.4 Intermediate to Lower Shell Girth 9-203 (Heat 10137) 2.253 0.800 121.7 93.5 Weld 9-203 (Heat 90136) 2.253 0.800 151.3 116.2 (a) Maximum neutron fluence for vessel location.
(b) Adjusted reference temperature projections based on best-estimate copper and nickel content, chemistry factor, initial RTNDT and margin given in Table 4.6-13.
FIGURE 4.6-1 LOCATION OF SURVEILLANCE CAPSULE ASSEMBLIES FIGURE 4.6-2 TYPICAL SURVEILLANCE CAPSULE ASSEMBLY FIGURE 4.6-3 TYPICAL CHARPY IMPACT COMPARTMENT ASSEMBLY FIGURE 4.6-4 TYPICAL TENSILE-MONITOR COMPARTMENT ASSEMBLY FIGURE 4.6-5 BASE METAL - WR (TRANSVERSE) PLATE C-506-1 IMPACT ENERGY VS TEMPERATURE
FIGURE 4.6-6 BASE METAL - WR (TRANSVERSE) PLATE C-506-1 LATERAL EXPANSION VERSUS TEMPERATURE
FIGURE 4.6-7 BASE METAL - RW (LONGITUDINAL) PLATE C-506-1 IMPACT ENERGY VERSUS TEMPERATURE
IGURE 4.6-8 BASE METAL - RW (LONGITUDINAL) PLATE C-506-1 LATERAL EXPANSION VS TEMPERATURE
FIGURE 4.6-9 WELD METAL PLATE C-506-2/C-506-3 IMPACT ENERGY VS TEMPERATURE
GURE 4.6-10 WELD METAL, PLATE C-506-2/C-506-3 LATERAL EXPANSION VS TEMPERATURE
FIGURE 4.6-11 HAZ METAL, PLATE C-506-1 IMPACT ENERGY VERSUS TEMPERATURE
FIGURE 4.6-12 HAZ METAL, PLATE C-506-1 LATERAL EXPANSION VERSUS TEMPERATURE
IGURE 4.6-13 SRM (HSST PLATE 01MY - LONGITUDINAL) IMPACT ENERGY VERSUS TEMPERATURE
FIGURE 4.6-14 SRM (HSST PLATE 01MY - LONGITUDINAL) LATERAL EXPANSION VERSUS TEMPERATURE
.1 INTRODUCTION purpose of this appendix is to describe the methods employed and present the results obtained dynamic seismic analyses of the reactor coolant system components. These analyses were ormed to confirm the adequacy of the seismic loadings specified for the design of the ponents and the supports of the reactor coolant system which includes the reactor vessel, the m generators, the reactor coolant pumps, the pressurizer, and the interconnecting reactor lant piping.
amic seismic analysis of the reactor vessel internals was performed separately and is ussed in Appendix 3.A.
.2 METHOD OF ANALYSIS
.2.1 General seismic analysis of the reactor coolant system (RCS) components was performed using either mal mode or direct integration theory in conjunction with time history and response spectrum niques, as appropriate.
e history techniques were employed in the analysis of the reactor vessel, the two steam erators, the four reactor coolant pumps and the interconnecting reactor coolant piping. In the lysis of these components, a single composite mathematical model, which included integral esentations of each of the components and connecting piping, was employed to account for interacting effects of dynamic coupling. The analysis of these dynamically coupled multi-ported components utilized different time dependent input excitations applied simultaneously ach support.
analyses of the pressurizer and the surge line piping employed separate, uncoupled, hematical models and utilized response spectrum techniques.
input data, time histories and response spectra, applied in the analyses were provided by the lysis of the containment structure internal support structure described in Section 5.8.
RCS components were analyzed using either modal or proportional (Rayleigh) methods of ping. Except for analysis of the surge line piping, all modal analyses used a constant damping or of 1% of critical damping for all active modes. In the analysis of the surge line piping, a ping factor of 0.5% of critical damping was used for each mode. When proportional damping used, Alpha and Beta were conservatively selected to provide less than 1% of critical ping at the significant frequencies of response of the major components.
he descriptions of the mathematical models which follow, the spatial orientations are defined he set of orthogonal axes where Y is in the vertical direction, and X and Z are in the horizontal e, in the directions indicated on the appropriate figure. The mathematical representation of section properties of the structural elements employs a 12 by 12 stiffness matrix for the three-ensional space frame models, and employs a 6 x 6 stiffness matrix for the two dimensional e frame model. Elbows in piping runs include the in-plane/out-of-plane bending flexibility ors as specified in the ASME Code,Section III.
.2.2.1 Reactor Coolant System - Coupled Components chematic diagram of the composite mathematical model, designated MS-2, used in the original t design analyses of the dynamically coupled components of the reactor coolant system is ented in Figure 4.A-1 and 4.A-1A. This model, Figure 4.A-1, includes 19 mass points with a l 47 dynamic degrees of freedom. The mass points and corresponding dynamic degrees of dom are distributed to provide appropriate representations of the dynamic characteristic of the ponents, as follows: the reactor vessel, with internals, is represented by 5 mass points with a l of 13 dynamic degrees of freedom; each of the two steam generators are represented by 3 s points with a total of 7 dynamic degrees of freedom; and each of the four reactor coolant ps are represented by 2 mass points with a total of 5 dynamic degrees of freedom. The tively small mass of the interconnecting reactor coolant piping is lumped proportionately with masses of the adjoining components.
Number of Mass Number of Dynamic Degrees of
<component> Points Freedom actor Vessel and Internals 5 13 am Generators (2) 6 14 C. Pumps (4) 8 20 Total 19 47 mathematical model, Figure 4.A-1, as defined, provides a complete three dimensional esentation of the dynamic response of the coupled components to seismic excitations in both horizontal and vertical directions. The mass is distributed at the selected mass points and esponding translational degrees of freedom are retained to include rotary inertial effects of the ponents. The total mass of the entire coupled system is dynamically active in each of the three rdinate directions.
addition to Model MS-2 described above, a second model of the coupled components, gnated Model RV14, was formulated for the original plant design to incorporate a more iled representation of the reactor vessel assembly. With the exception of the representation of reactor vessel assembly, Model RV14 is identical to Model MS-2, Figure 4.A-1. A schematic ram of the representation of the reactor vessel assembly incorporated into Model RV14 is ented in Figure 4.A-2. This more detailed representation of the reactor vessel assembly
esentation of the reactor vessel internals was formulated in conjunction with the analysis of reactor vessel internals discussed in Appendix 3.A, and was designed to simulate the dynamic racteristics of the models used in that analysis. The coupled Model RV14 was used to generate e histories of absolute accelerations at the reactor vessel flange used as forcing functions in the lysis of the reactor vessel internals.
pled model RV14, modified to delete the reactor internals representation of the thermal shield to reflect current design steam generator properties, was also used in analyses to determine effects of the replacement steam generators on RCS responses. This modified model esentation consists of 27 mass points with 63 dynamic degrees of freedom.
Number of Mass Number of Dynamic Degrees of
<component> Points Freedom actor Vessel 5 11 actor Internals 8 18 am Generators(2) 6 14 actor Coolant Pumps (4) 8 20 Total 27 63
.2.2.2 Pressurizer mathematical model employed in the analysis of the pressurizer is shown schematically in ure 4.A-3. This lumped parameter, planer model provides a multi-mass representation of the lly symmetric pressurizer and includes 5 mass points with a total of 6 dynamic degrees of dom.
replacement pressurizer was analyzed using BWSPAN for both operating basis and safe tdown earthquake using the response spectrum method. The seismic excitation is applied at support skirt elevation. The significant change to the analysis from the original pressurizer lysis is that the analysis is performed at 65% water level. The seismic directional responses are bined by the absolute sum of each horizontal and vertical response of the spectrum analysis.
.2.2.3 Surge Line lumped parameter, multi-mass mathematical model employed in the analysis of the surge line hown schematically in Figure 4.A-4. The surge line is modeled as a three dimensional piping with end points anchored at the attachments to the pressurizer and the reactor vessel outlet ng. In the definition of the mathematical model, 10 mass points with a total of 27 dynamic rees of freedom were selected to provide a complete three-dimensional representation of the amic response of the surge line. All supports and restraints defined for the surge line assembly
.2.3 Calculations
.2.3.1 General applied in the analysis, the simultaneous equations of motion for linear structural systems with ous damping can be written, Reference 4.A-1:
MX ** + CX* + KX = - MY ** - K X ms s re:
M = diagonal matrix of lumped masses C = square symmetric damping matrix K = square symmetric stiffness matrix which defines the mass point force-displacement relationship.
Y** = column matrix with elements equal to the absolute acceleration of the datum support in the coordinate direction of the related dynamic degree of freedom of the structural system.
Kms = rectangular matrix of stiffness coefficients which defines the mass point force, non-datum support displacement relationship.
Xs = column matrix of displacements relative to the datum at non-datum supports.
X = column matrix of mass point displacements relative to the datum.
X* = column matrix of mass point velocities relative to the datum.
X** = column matrix of mass point accelerations relative to the datum.
his form, the equations define the dynamic response of a multi-mass structural system jected to time dependent support motion. In the analysis of systems with multiple supports, h as the coupled components of the reactor coolant system, the equations provide for different e-dependent input motions at each of the supports. In this case, one of the supports of the em is designated the reference, or datum, from which the motions of all other points of the ctural system are measured. The reactor vessel support was designated as the datum in the lyses of the coupled components of the reactor coolant system.
mal mode theory, as described in References 4.A-1 and 4.A-2, was employed to reduce the ations of motion to a system of independent equations in terms of the normal modes for the dal superposition time-history and spectrum analyses of the reactor coolant system ponents. For direct integration time-history analyses, the equations of motion were solved
h of the three global coordinate directions: X (east-west), Y (vertical) and Z (north-south). The amic responses to vertical seismic excitation were found for both the case of initial support lacement upward and the case of initial support displacement downward. These responses e combined to determine the most severe combinations produced by the effects of seismic itations in each of the horizontal directions applied simultaneously with either seismic itation in the vertical direction.
.2.3.2 Frequency Analysis eigenvalue analysis was performed utilizing the ICES STRUDL II computer code, erence 4.A-3, to calculate the mode shapes and natural frequencies of the original plant design posite mathematical models. Modifications to the standard ICES STRUDL II program have n implemented by Combustion Engineering to include a Jacobi diagonalization procedure in eigenvalue analysis and to provide appropriate influence coefficients and stiffness matrices for in the response and reaction calculations.
natural frequencies and dominant degrees of freedom calculated are shown in Table 4.A-1 for modes used in the analysis of the reactor coolant system Model MS-2, the surge line and the surizer.
eigenvalue analysis was also performed utilizing the ANSYS computer code, Reference 4.A-o calculate the natural frequencies and mode shapes of modified model RV14 which was used he analysis to evaluate the effects of the replacement steam generators. A comparison of onse frequencies for corresponding mode shapes identified no significant differences from e in Table 4.A-1 (see also Section 4.A.5, Effects of Replacement Steam Generators).
.2.3.3 Mass Point Response Analysis original plant design time history mass point responses to seismic excitation were computed g TMCALC, a C-E code. This code performs a numerical integration of the equations of ion for singly or multiply supported dynamic systems utilizing normal mode theory, erence 4.A-2, and Newmarks Beta-Method with Beta equal to one-sixth, Reference 4.A-4.
the multiply supported system, the separate time histories of each support were imposed on system simultaneously. The results are time history responses of the mass points. The analysis he reactor coolant system utilized modal data for all frequencies through 40 cps.
mass point responses resulting from the spectrum analysis were found utilizing SHAKE, a computer code. This code performs a normal mode response spectrum analysis resulting in modal inertials loads found using the response spectrum for the pressurizer support. The mass nt responses of the surge line were found using an envelope of the support spectra of the rconnected major components.
acement steam generators and with the reactor thermal shield removed.
.2.3.4 Seismic Reaction Analysis original plant design dynamically induced loads at all system design points due to the time ory support excitations and mass point responses were calculated utilizing FORCE, a C-E puter code. This code performs a complete loads analysis of the deformed structure at each emental time step by computing internal and external system reactions (forces and moments) uperposition of the reactions due to the mass point displacements and the non-datum support lacements as follows:
R(t) = CmXm(t)+CsXs(t) re:
R(t) = the matrix of all components of the reactions at the system design points.
Cm = the matrix of mass points displacement influence coefficients.
Xm(t) = the column matrix of time history mass point displacements relative to the datum at each time step.
Cs = the matrix of support displacement influence coefficients.
Xs(t) = the column matrix of time history support displacements relative to the datum at non-datum supports at each time step.
support and mass point displacements due to horizontal and vertical seismic excitations are ed algebraically at each time step. The maximum component of each reaction for the entire e domain, and its associated time of occurrence, are selected.
maximum reactions for the pressurizer and surge line resulting from the response spectrum lysis were found by applying the modal inertial loads for each mode, to the structural model g the STRUDL computer code. The design point reactions due to each modal loading were servatively combined by summing the absolute values of the modal reactions. For the acement pressurizer, the design point reactions due to each modal loading were combined g square root sum of the squares. The surge line analysis included consideration of the tive end displacements. The reactions found by statically imposing the maximum relative lacements of the two ends of the surge line were conservatively included by absolute mation with the inertial response from the spectrum analysis.
system design loads for the coupled components of the reactor coolant system with existing acement steam generators and with the reactor internals thermal shield removed were ulated using the ANSYS computer code.
reactions (forces and moments) at all design points in the system, obtained from the dynamic mic analysis, were compared with the seismic loads in each component design specification.
results of this comparison are summarized in Table 4.A-2 for the points of maximum ulated load.
maximum seismic loads calculated by the time history techniques are the result of a search comparison over the entire time domain of each individual component of load due to the ultaneous application of the horizontal and either vertical excitation. The maximum calculated ponents of load shown in Table 4.A-2 for each design location do not in general occur at the e time, nor for the same combination of horizontal and vertical excitation, and therefore result conservative case.
ept for the replacement pressurizer, the maximum seismic loads calculated by the response ctrum techniques are the result of combining the modal reactions due to the horizontal and the ical excitation on an absolute sum basis.
results shown are for the Operational Basis Earthquake. For conservative determination of lts due to the Design Basis Earthquake, both the calculated results and specification values e been multiplied by a factor of 2.0, rather than 0.17/0.09 = 1.89, the ratio of DBE to OBE imum ground accelerations.
.4 EFFECTS OF THERMAL SHIELD REMOVAL engineering evaluation was made to assess the effects of the thermal shield removal on the amic response characteristics of the reactor vessel and the reactor coolant system. The main ct of the thermal shield removal was a reduction in the weight of the reactor vessel. This uction was approximately two percent (2%). Since the stiffness of the connection between the tor vessel and internals did not change, it was concluded that the dynamic response racteristics of the reactor vessel and, therefore, the reactor coolant system would not change ificantly with the removal of the thermal shield. Reactor vessel flange motions, originally d for a more detailed evaluation of the internal structures in Appendix 3.A, also remain hanged.
.5 EFFECTS OF REPLACEMENT STEAM GENERATORS most significant effects of the replacement steam generators were increases in weights and ter of gravity elevations of approximately 2.7% and 13.2 inches, respectively, for these ponents. Modified model RV14, which includes these changes in addition to those due to oval of the reactor internals thermal shield, was used in a reanalysis of the dynamically pled components of the RCS. The changes in steam generator properties resulted in a uction of approximately 4.5% in the fundamental frequencies in which the steam generator onse is predominant. Corresponding changes in loads which are attributed to the replacement m generators are generally insignificant with no design governing load increases of more than
actor vessel flange motions were compared with those originally used in the detailed luation of the internal structures in Appendix 3.A. This comparison concluded that use of the inal flange motions is conservative for reactor internals design.
.6 CONCLUSION concluded that the seismic loadings specified for the design of the reactor coolant system ponents and supports are adequate. All seismic loads calculated by the dynamic seismic lysis are less than the corresponding loads in the component design specification.
.7 REFERENCES
-1 Przemieniecki, J. S., Theory of Matrix Structural Analysis, Chapter 13, McGraw-Hill Book Company, New York, New York, 1968.
-2 Hurty, W. C., and Rubinstein, M. F., Dynamics of Structures, Chapter 8, Prentice Hall, Inc., Englewood Cliffs, New Jersey, 1964.
-3 ICES STRUDL II Engineering Users Manual, R68-91, Department of Civil Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts.
-4 Newmark, N. M., A Method of Computation for Structural Dynamics, Volume 3, Journal of Engineering Mechanics Division, A.S.C.E., July, 1959.
-5 DeSalvo, G. P., and Swanson, J. A., ANSYS - Engineering Analysis System, Swanson Analysis Systems, Inc., Elizabeth, PA., 1972.
Mode Frequency Dominant Degrees of Freedom Number (cps) Names Directions Locations 1 3.07 M66 Z Pump 1B 2 3.08 M43 Z Pump 2B 3 3.21 M61 Z Pump 1A 4 3.23 M52 Z Pump 2A 5 3.32 RI1 Z Reactor Internals 6 3.32 RI1 X Reactor Internals 7 5.11 M61 X,Y Pump 1A 8 5.14 M52 X,Y Pump 2A 9 5.21 M66 X,Y Pump 1B 10 5.23 M43 X,Y Pump 2B 11 10.51 SG5A, SG5B X Steam Generators 1 & 2 12 10.52 SG5A, SG5B X Steam Generators 1 & 2 13 10.60 M61 X Pump 1A 14 10.74 M52 X Pump 2A 15 10.98 M66 X Pump 1B 16 11.11 M43 X Pump 2B 17 12.14 RI2 Z Reactor Internals 18 12.16 RI2 X Reactor Internals 19 20.40 SG9A Z Steam Generator 1 20 20.42 SG9B Z Steam Generator 2 21 23.65 RI1 Y Reactor Internals 22 27.26 SG9, 10, A & B X Steam Generators 1 & 2 23 27.79 SG5A, SG5B Y Steam Generators 1 & 2 24 27.82 SG5A, SG5B Y Steam Generators 1 & 2 25 28.99 M65, M42 Z Pumps 1B & 2B 26 29.91 M65, M42 Z Pumps 1B & 2B 27 30.37 M42 Z Pump 2B 28 31.43 M60 Z Pump 1A 29 31.92 M51 Z Pump 2A 30 37.51 V1, V3, V4 X, Z Reactor Vessel 31 39.34 SG5B Z Steam Generator 2
Mode Frequency Dominant Degrees of Freedom Number (cps) Names Directions Locations 32 39.41 SG5A Z Steam Generator 1 1 13.78 - X, Z Replacement Pressurizer 2 44.65 - X, Z Replacement Pressurizer 3 48.99 - X Replacement Pressurizer 4 62.35 - X, Z Replacement Pressurizer 5 75.86 - X, Z Replacement Pressurizer 6 75.97 - Y Replacement Pressurizer 7 99.45 - X, Z Replacement Pressurizer 1 5.75 7, 8, H3 Y Surge Line 2 11.29 5, 7, H1 X Surge Line 3 15.98 5, H1 Y Surge Line 4 21.79 8, 9, H3 Z Surge Line 5 25.81 9, H3 Y Surge Line 6 32.12 8, 9 X Surge Line 7 40.35 10 X Surge Line 8 73.64 5 Z Surge Line 9 108.52 3 X Surge Line 10 129.12 10 Y Surge Line 11 151.51 7, H3 X Surge Line 12 155.03 7 Y Surge Line 13 174.76 7, H1 Y Surge Line 14 186.39 H1 X Surge Line 15 229.90 10, 11 Z Surge Line 16 260.38 11 X Surge Line 17 304.86 3 Y Surge Line 18 320.52 7 X Surge Line 19 503.60 H1 X Surge Line 20 525.21 8 Y Surge Line 21 535.87 8 X Surge Line 22 542.64 11 Z Surge Line 23 668.06 4 X Surge Line 24 752.11 4 Z Surge Line
Mode Frequency Dominant Degrees of Freedom Number (cps) Names Directions Locations 25 938.66 11 Y Surge Line 26 1327.41 8 Z Surge Line 27 1807.56 4 Y Surge Line
EARTHQUAKE SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined North-South and Vertical Reactor Vessel Outlet Nozzle Fx 107.0 69.0 Fy 3.0 172.0 Fz 65.0 63.0 Mx 1237.0 2168.0 My 7329.0 7521.0 Mz 382.0 38474.0 MR 7443.0 39263.0 Reactor Vessel Inlet Nozzle Fx 97.0 75.0 Fy 98.0 105.0 Fz 107.0 108.0 Mx 7719.0 22392.0 My 5402.0 10085.0 Mz 16910.0 14087.0 MR 19358.0 28312.0 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined East-West and Vertical Reactor Vessel Outlet Nozzle Fx 196.0 241.0 Fy 39.0 253.0 Fz 4.0 10.0 Mx 67.0 1042.0 My 360.0 972.0 Mz 3277.0 43773.0 MR 3298.0 43796.0 Reactor Vessel Inlet Nozzle Fx 81.0 220.0 Fy 97.0 146.0 Fz 110.0 63.0 Mx 5963.0 9638.0 My 3803.0 5981.0 Mz 11763.0 12770.0 MR 13726.0 17080.0 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined North-South and Vertical Steam Generator Inlet Nozzle Fs 88.5 144.0 Fa 89.0 230.0 Mb 3901.7 4000.0 Mt 2584.0 2640.0 Steam Generator Outlet Nozzles Fs 134.8 202.0 Fa 54.0 104.0 Mb 8990.0 13200.0 Mt 10517.0 14800.0 Combined East-West and Vertical Steam Generator Inlet Nozzle Fs 97.1 144.0 Fa 175.0 230.0 Mb 4239.0 4000.0 Mt 113.0 2640.0 Steam Generator Outlet Nozzle Fs 126.1 202.0 Fa 49.0 104.0 Mb 7681.0 13200.0 Mt 11224.0 14800.0 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined North-South and Vertical Pressurizer Surge Line Nozzle Fx 1.74 2.26 Fy 0.91 4.85 Fz 2.71 8.75 MR 192.7 690.9 Combined East-West and Vertical Pressurizer Surge Line Nozzle Fx 1.07 5.70 Fy 0.59 4.77 Fz 1.22 2.71 MR 121.5 617.1 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined North-South and Vertical Reactor Coolant Pump Nozzle Fx 108.0 83.6 Fy 61.0 93.2 Fz 53.0 114.1 Mx 6289.0 12030.0 My 5632.0 10188.0 Mz 11867.0 10095.0 MR 14564.0 18719.0 Reactor Coolant Pump Outlet Nozzle Fx 137.0 93.2 Fy 98.0 71.3 Fz 43.0 97.8 Mx 10457.0 22770.3 My 2244.0 7514.4 Mz 2244.0 8957.0 MR 7797.0 25596.0 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined East-West and Vertical Reactor Coolant Pump Inlet Nozzle Fx 119.0 143.6 Fy 50.0 104.5 Fz 35.0 33.9 Mx 5291.0 8709.0 My 4479.0 5238.0 Mz 12263.0 16662.5 MR 14087.0 19517.0 Reactor Coolant Pump Outlet Nozzle Fx 134.0 263.8 Fy 97.0 145.8 Fz 37.0 109.9 Mx 5099.0 4072.0 My 2964.0 13317.0 Mz 6269.0 16648.0 MR 8608.0 21796.0 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined North-South and Vertical Reactor Vessel Outlet Piping MR 7443.0 11514.0 Steam Generator Inlet Piping MR 4680.0 4361.0 *
- Note: Piping design is controlled by combined East-West and vertical seismic excitation.
Steam Generator Outlet Piping MR 13836.0 13836.0 Pump Inlet Piping MR 13148.0 13836.0 Pump Outlet Piping MR 13236.0 18528.0 Reactor Vessel Inlet Piping MR 18522.0 18528.0 Combined East-West and Vertical Reactor Vessel Outlet Piping MR 3298.0 11581.0 Steam Generator Inlet Piping MR 4241.0 9658.0 Steam Generator Outlet Piping MR 13601.0 13836.0 Pump Inlet Piping MR 13632.0 13836.0 Pump Outlet Piping MR 8608.0 18528.0 Reactor Vessel Inlet Piping MR 13726.0 18528.0 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined North-South and Vertical Surge Line RCS Nozzle Fx 4.41 6.0 Fy 1.45 4.0 Fz 3.25 4.0 MR 277.6 468.0 Combined East-West and Vertical Surge Line RCS Nozzle Fx 2.55 6.0 Fy 0.84 4.0 Fz 1.31 4.0 MR 155.5 468.0 Combined North-South and Vertical Surge Line Hanger H4 Fx 0.150 0.18 Fy 0.439 2.0 Surge Line Hanger H2 Fy 0.209 1.0 Fz 0.078 0.11 Combined East-West and Vertical Surge Line Hanger H4 Fx 0.083 0.18 Fy 0.335 2.0 Surge Line Hanger H2 Fy 0.145 1.0 Fz 0.016 0.11 Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined North-South and Vertical Reactor Vessel Outlet Support Fy 165.0 392.0 Fz 325.0 663.0 Reactor Vessel Inlet Support Fy 314.0 692.0 FH 191.0 304.0 Steam Generator Lower Support Fy 219.0 431.0 Fz 195.0 397.0 Mx 20345.0 24422.0 My 2335.0 10332.0 Steam Generator Upper Support Fx 20.0 24.0 Fz 219.0 240.0 Pump Support Fy 2.6 9.2 Replacement Pressurizer Support Fy 14.1 80.0 a Fz 80.1 84.0
- Mx 15107.6 22768.9 17036.0
- Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
SEISMIC LOAD COMPONENT AND DESIGN CALCULATED SEISMIC EXCITATION LOCATION COMPONENT MAXIMUM DESIGN BA Combined East-West and Vertical Reactor Outlet Support Fy 309.0 473.0 Fz 47.0 42.0 Reactor Vessel Inlet Support Fy 227.0 469.0 FH 375.0 1020.0 Steam Generator Lower Support Fy 278.0 624.0 Fz 32.0 29.0 My 356.0 455.0 Mz 8792.0 24383.0 Steam Generator Upper Support Fx 281.0 296.0 Fz 12.0 10.0 Reactor Coolant Pump Support Fy 1.6 4.3 Replacement Pressurizer Support Fx 55.6 81.0
- Fy 14.1 80.0
- Mz 15746.9 16949.0
- Forces = Kips; Moments = Inch - Kips MR = [Mx2 + My2 + Mz2]1/2 FH = [Fx2 + Fz2]1/2
- a. Note that the design basis values are based on original pressurizer design and are retained here for historical purposes only.
IGURE 4.A-1 REACTOR COOLANT SYSTEM SEISMIC ANALYSIS MODEL MS2 FIGURE 4.A-1A REACTOR COOLANT SYSTEM - SEISMIC ANALYSIS MODEL MS2 AND RV14 IGURE 4.A-2 RV14 REACTOR AND INTERNALS SEISMIC ANALYSIS MODEL FIGURE 4.A-3 PRESSURIZER SEISMIC ANALYSIS MODEL FIGURE 4.A-4 SURGE LINE SEISMIC ANALYSIS MODEL