ML16256A126
| ML16256A126 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 08/25/2016 |
| From: | Entergy Operations |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16256A115 | List:
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| References | |
| W3F1-2016-0053 | |
| Download: ML16256A126 (19) | |
Text
WSES-FSAR-UNIT-3 1.3-1 Revision 11-A (02/02) 1.3 COMPARISONS 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS Table 1.3.1 presents a summary of the characteristics of Waterford 3 for Cycle 1. The table presents comparative data for San Onofre Units 2 and 3; Arkansas Nuclear One, Unit 2; and St. Lucie Unit 1.
The San Onofre Units 2 and 3, and Arkansas Nuclear One, Unit 2 designs were selected for comparison because of the basic similarity of the reactor cores and the Reactor Coolant Systems. Also they are well advanced in terms of licensing relative to Waterford 3. St. Lucie Unit 1 was selected because of the basic similarity in the containment design.
1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION 1.3.2.1 General This section contains a discussion of all significant changes that have been made in the Waterford 3 design since submittal of the PSAR until the docketing of the FSAR. Changes considered as significant include changes in design bases or criteria for safety-related structures, systems or components, plant arrangement, mode of system operation, type of equipment, or gross changes in component or system capacity. In general changes have been made to further increase the safety margin and operating flexibility of Waterford 3.
1.3.2.2 Site Characteristics Additional site studies and field tests have been made since the submittal of the PSAR. No significant site characteristic changes have been brought to light that would require a design change.
1.3.2.3 Design Criteria 1.3.2.3.1 NRC Regulatory Guides Many design changes are the result of the evolution of NRC's interpretation of the safety requirements to comply with the General Design Criteria of 10CFR50, Appendix A. This evolution has resulted in the promulgation of many Regulatory Guides which were not addressed in the PSAR. The intent of these new guides have been evaluated against the plant design and changes have been initiated where practical to bring Waterford 3 into general compliance.
1.3.2.3.2 Design Codes a)
Containment Vessel Code Case
¨(DRN 01-758)
The steel containment vessel has been designed to ASME Section III, Class "MC" rather than Class "B" requirements as stated in the PSAR. Class "B" was an earlier version of the ASME Code for containment vessels which was superseded by Class "MC" in 1971. The containment vessel design is described in Section 3.8.
(DRN 01-758) b)
Changes in Concrete Quality Control Program Concrete slump test frequency and maximum temperature have been changed in accordance with ASTM C-143 and ACI-305-72, respectively. The code requirements for concrete are discussed in Section 3.8
WSES-FSAR-UNIT-3 1.3-2 1.3.2.3.3 Pipe Break Criteria High and moderate energy piping systems have been analyzed in accordance with NRC Branch Technical Positions APCSB 3-1 and MEB 3-1 as discussed in LP&L letter LPL3690 dated July 11, 1975. (See Section 3.6.)
1.3.2.3.4 Tornado Design Criteria As discussed in Section 3.3, protection against multiple tornado generated missiles has been extended to systems and components required for safe shutdown.
1.3.2.4 Reactor Changes to fuel rod design have been made in order to reduce fuel densification problems. The number and design of fuel spacer grids has been changed to enhance fuel performances under seismic loading.
Fuel design is described in Section 4.2.
1.3.2.5 Reactor Structures 1.3.2.5.1 Reactor Vessel Grillage Foundation The reactor vessel grillage has been modified to incorporate a positive system of restraints to prevent upward motion of the reactor vessel following a LOCA. This Support System is discussed in Section 5.4.
1.3.2.5.2 Reactor Vessel Cavity The design of the reactor vessel cavity has been modified to provide a reduction in neutron streaming from the cavity, and activation of Ar-40 in the containment atmosphere. The reactor vessel cavity is discussed in Sections 3.8 and 6.2.
1.3.2.6 Engineered Safety Features 1.3.2.6.1 Hot Leg Injection Capability Provision has been made in the Safety Injection System to permit introduction of safety injection fluid during recirculation through both the hot and cold legs, thereby ensuring adequate long term post-LOCA cooling. The Safety Injection System is described in Section 6.3.
WSES-FSAR-UNIT-3 1.3-3 1.3.2.6.2 Containment Ductwork Declassification The containment ductwork ring header has been changed from Safety Class 2, seismic Category I to NNS, non-seismic Category I. The ring header is not required for the Containment Cooling System to perform its design function.
This duct system is discussed in Section 6.2.
1.3.2.6.3 Containment Vessel Installation, Testing and Inspection Several changes have been made to the installation and testing procedures for the containment vessel.
These changes include upgrading of containment vessel inner surface coating and cleaning requirements, deletion of radiography of randomly selected welds (15 percent) following postweld heat treatment, revision to welder qualification test record retention procedures, and acceptance of gas metal arc welding processes. The containment vessel is discussed in Section 6.2 and 3.8.
1.3.2.7 Instrumentation and Control 1.3.2.7.1 Reactor Protection System (RPS)
The Reactor Protection System (RPS) described in the PSAR has been expanded and some portions modified in order to provide automatic protection against axial xenon oscillations and to implement design improvements.
a)
The following changes were made to meet the requirement for automatic protection against axial xenon oscillations:
1)
The high local power density trip is added; 2)
The thermal margin/low pressure trip is replaced by the low DNBR trip; 3)
The core protection calculators (CPCS) are added to provide the high local power density and low DNBR trips and the thermal margin/low pressure calculator is eliminated.
b)
As a consequence of the above addition of the CPCS, the following design changes are implemented:
1)
The low reactor coolant flow trip function is incorporated into the low DNBR trip; 2)
Reactor coolant flowrate is calculated by use of reactor coolant pump speed instead of being inferred by differential pressure measurements; 3)
CEA position signals are incorporated into the RPS.
c)
A high logarithimic power level trip has replaced the high rate of change of power trip in order to provide improved protection against inadvertent boron dilution. The RPS also addresses the unplanned withdrawal of CEAs as the previous trip did.
WSES-FSAR-UNIT-3 1.3-4 Revision 9 (12/97)
The Engineered Safety Feature Actuation System (ESFAS) has been changed in the following areas:
a)
The Emergency Feedwater Actuation Signal (EFAS) is added to the ESFAS.
b)
Diverse signals for Containment Isolation have been added.
c)
Variable setpoints for SIAS on low pressurizer pressure and Main Steam Isolation Signal on low steam generator pressure are added.
d)
The group testing capability is added.
Further discussion of the RPS is found in Section 7.2.
1.3.2.7.2 Core Operation Limit Supervisory System
A non-safety-related Core Operating Limit Supervisory System (COLSS) has been added. The COLSS consists of sensors, algorithms implemented in the plant monitoring computer, and other equipment to monitor selected Nuclear Steam Supply System parameters and process the parameter information so that a comprehensive, on-line calculation of the margin to specified limiting conditions of operation is available at all times. The COLSS also provides the operator with an alarm so that he can maintain the reactor core within the limiting conditions of operation during steady-state operation by initiating a power reduction whenever any one of the monitored core conditions reaches its specified limiting condition of operation. This system is described further in Section 7.7.
1.3.2.7.3 Movable In-Core Instrument System This system has been deleted.
1.3.2.7.4 Deletion of Containment Purge Isolation Signal (CPIS) and High Containments Radiation Input to the Plant Protection System Since a fuel handling accident can occur only during shutdown, CPIS on high radiation should be independent of the Plant Protection System (PPS). Therefore, the PPS has been changed accordingly.
This change also allows the deenergizing of the PPS for maintenance and inspection during shutdown.
The PPS is discussed in Section 7.3.
1.3.2.8 Electric Power Extensive redesign of all electrical and I&C cable trays, control boards and cabinets to meet the separation criteria of Regulatory Guide 1.75 (1/75) has been made. Compliance with RG 1.75 is discussed in Section 8.3.
1.3.2.9 Auxiliary Systems 1.3.2.9.1 High Density Spent Fuel Storage Additional storage capacity of spent fuel has been provided for by means of high density poison racks.
The design of these racks is described in Section 9.1.
WSES-FSAR-UNIT-3 1.3-5 1.3.2.9.2 Fire Protection System The Fire Protection System has been modified as a result of the new NRC requirements (10CFR50, Appendix R). The Fire Protection System is discussed in Section 9.5.
1.3.2.10 Steam and Power Conversion System There are no significant changes in the final design of the Steam and Power Conversion System from that described in the PSAR other than all volatile water treatment has been provided.
1.3.2.11 Radioactive Waste Management Additional radwaste handling capability has been provided for by installing means to allow the use of portable solidification and demineralization systems.
1.3.2.12 Radiation Protection Since the PSAR, the normal sampling system has been extensively rerouted and a new sampling panel with a modified equipment configuration has been introduced in the RAB design. A new sampling system has been installed for Post-Accident Sampling (P.A.S.S.) and additional shielding and sample tubing have been installed for this purpose. These systems are discussed in Subsection 12.5.3.
1.3.2.13 Conduct of Operations Since the PSAR, extensive modifications to the Plant Security System have been made including the addition of much more sophisticated equipment, such as TV monitors and electronic card readers. The Plant Security System is summarized in Section 13.6.
A separate Administration Building has been added due to increases in the operating staff.
1.3.2.14 Initial Test and Operations There are no significant changes in initial tests and operations affecting plant design from that described in the PSAR.
1.3.2.15 Accident Analyses The methods used to analyze some of the accidents have been revised to take into account expansion of the RPS (see Subsection 1.3.2.7.1). In addition, there have been extensive refinements to computer codes, analytical investigations and tests to demonstrate compliance with 10CFR50, Appendix K,.
Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors.
The offsite accident doses presented in the PSAR were calculated using the atmospheric dispersion factors (X/Q) based on preliminary meteorological data. The dose analyses presented in Chapter 15 are based on X/Q values obtained from the onsite meteorological monitoring program described in Section 2.3. The onsite data demonstrates the conservatism of the X/Q values used in the PSAR accident analyses.
1.3.2.16 Technical Specifications Details of safety limiting settings and limiting conditions of operation have changed as a result of changes enumerated in other parts of this subsection.
WSES-FSAR-UNIT-3 1.3-6 Subject coverage is in accordance with NRC Standard Technical Specifications as revised for Waterford 3.
1.3.2.17 Quality Assurance The major change in the QA program since the submittal of the PSAR has been the commitment of WASH 1309, "Guidance on Quality Assurance Requirements During the Construction Phase of Nuclear Power Plants." Quality Assurance during operation is described in detail in the QA Program Manual.
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 1 of 13) Revision 11-A (02/02)
¨(DRN 01-758)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 (DRN 01-758)
Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Hydraulic and Thermal Design Parameters Rated core heat output, Mwt 3,390 4.4 3,390 2,815 2,560 Rated core heat output, Btu/hr 11,570 x 106 4.4 11,570 x 106 9,608 X 106 8,737 X 106 Heat generated in fuel, %
97.5 4.4 97.5 96.5 97.5 System pressure, nominal, psia 2,250 4.4 2,250 2,250 2,250 System pressure, minimum steady state, psia 2,200 4.4 2,200 2,200 2,200 Hot channel factors, Heat flux, Fq 2.35 2.35 2.35 2.85 Enthalpy rise, FH 1.55 4.4 1.55 1.55 2.02 DNB ratio at nominal conditions 2.07 (CE-1) 4.4 2.07 (CE-1) 2.26 (W-3) 2.30 (W-3)
Coolant flow Total flowrate, lb/hr 148 x 106 4.4 148 x 106 120.4 x 106 122 x 106 Effective flowrate for heat transfer, lb/hr 144.2 x 106 4.4 144.2 x 106 116.2 x 106 117.5 x 106 Effective flow area for heat transfer, ft2 54.7 4.4 54.7 44.7 53.5 Average velocity along fuel rods, ft/sec 16.4 4.4 16.4 16.4 13.6 Average mass velocity, lb/hr-ft2 2.64 x 106 4.4 2.64 x 106 2.60 x 106 2.20 x 106 Coolant temperatures, F Nominal inlet 553 4.4 553 553.5 538.9 Design inlet 553 4.4 556 556 5 544 Average rise in vessel 58 4.4 58 58.5 55 Average rise in core 60 4.4 60 60.5 56 Average in core 583 4.4 583 583.75 572
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 2 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Hydraulic and Thermal Design Parameters (Cont.)
Average in vessel 582 4.4 582 582.75 571.5 Nominal outlet of hot channel 642 4.4 642 652 640 Average film coefficient, Btu/hr-ft2-F 6,160 4.4 6,160 6,170 5,300 Average film temperature difference, F 30 4.4 30 31 35 Heat transfer at 100% power Active heat transfer surface area, ft2 62,000 4.4 62,000 51,000 48,400 Average heat flux, Btu/hr-ft2 182,400 4.4 182,400 182,200 176,000 Maximum heat flux, Btu/hr/ft2 428,000 4.4 428,000 425,800 501,300 Average thermal output, KW/ft (Fuel Rod Only) 5.34 4.4 5.34 5.34 5.94 Maximum thermal output, KW/ft (Fuel Rod Only) 12.5 4.4 12.5 12.5 17 Maximum clad surface temperature at nominal 657.0 4.4 657.0 657 657 Pressure, F Fuel center temperature, F maximum at 100% power 3,420 4.4 3,420 3,420 3,890 Core Mechanical Design Parameters Fuel assemblies Design CEA 4.2 CEA CEA CEA Rod pitch, in.
0,506 4.2 0.5063 0.5063 0.58 Cross-section dimensions, in.
7.972 x 7.972 4.2 7.972 x 7.972 7.98 x 7.98 7.98 x 7.98 Fuel weight (as UO2 ), lbm 223.9 x 103 4.2 223.9 x 103 183,640 207,200 Total weight, lbm 310,744 4.2 314,867 256,827 271,280 Number of grids per assembly 11 4.2 11 12 8
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 3 of 13) Revision 11-A (02/02)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Core Mechanical Desiltn Parameters (Cont.)
Fuel rods Number 49,580 4.2 49,500 40,716 36,896
¨(DRN 01-758)
Outside diameter, in.
0.382 4.2 0,382 0,382 0.44 Diametral gap, in.
0.007 4.2 0,007 0,007 0.0085 Clad thickness, in.
0.025 4.2 0,025 0,025 0,026 (DRN 01-758)
Clad material Zircaloy-4 4.2 Zircaloy-4 Zircaloy Zircaloy Fuel pellets Material UO2 sintered 4.2 UO2 sintered UO2 sintered UO2 sintered Diameter, in.
0.325 4.2 0.325 0.325 0.3795 Length, in.
0.390 4.2 0.390 0.390 0.650 Control assemblies Neutron absorber (See Table 4.2-1) 4.2 (See Table 4.2-1)
B4 C/Ag-In-Cd B4 C/SS Cladding material Inconel 625 4.2 Inconel 625 NiCrFe alloy NiCrFe alloy Clad thickness 0.035 4.2 0.035 0.035 0.040 Number of assembly, full/part-length 83/8 4.2 83/8 73/8 73/8 Number of rods per assembly 4,5/5 4.2 4,5/5 5
5 Nuclear Design Data Structural characteristics Core diameter, in. (equivalent) 136 4.2 136 123 136 Core height, in. (active fuel) 150 4.2 150 150 136.7
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 4 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Nuclear Design Data (Cont.)
H2 O/U, Unit cell (cold) 3.35 4.3 3.35 1.63 Number of fuel assemblies 217 4.2 217 177 217 UO2 Rods per assembly, unshimed/shimed Batch A 236 4.3 236 176 176 Batch B 236/220 4.3 236/220 164 164 Batch C 236/224 or 220 4.3 236/224 or 220 176/164 176/164/164 Performance characteristics loading technique 3-batch mixed 4.3 3-batch mixed 3-batch mixed 3-batch mixed central zone central zone central zone central zone Fuel discharge burnup, MWD/MTU Average first cycle 12,731 4.3 12,731 12,500 12,800 Feed enrichment, wt%
Region 1 1.87 4.3 1.87 1.93 1.93 Region 2 2.38 4.3 2.38 2.27 2.33 Region 3 2.88 4.3 2.88 2.94 2.82 Control characteristics effective multiplication (beginning of life)
Cold, no power, clean 1,170 4.3 1,170 1,182 1.170 Hot, no power, clean 1,125 4.3 1,125 1,136 1.134 Hot, full power, Xe equilibrium 1,067 4.3 1,067 1,075 1.078 Control assemblies Total rod worth (hot), %
11.35 4.3 11.35 12.3 11.0
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 5 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Boron concentrations for criticality:
Zero power no rods inserted, clean, ppm 832 832 Cold/Hot 899/832 4.3 899/832 1,004/987 945/935 At power with no rods inserted, 719/452 4.3 719/452 870/612 820/590 clean/equilibrium xenon, PPM Kinetic characteristics, range over life Moderator temperature coefficient, p/F See Table 4.3-4 4.3 See Table 4.3-4
-0.5 x 10-4 0.4 x 104 to to
-3.1 x 10-4
-2.1 x 10-4 Moderator pressure coefficient, p/psi
+0.7 x 10-6 4.3
+0.7 x 10-6
+0.45 x 10-6
+0.49 x 10-6 to to
+2.97 x 10-6
+2.55 x 10-6 Moderator void coefficient, p/% Void
-0.36 x 10-3 4.3
-0.36 x 10-3
-0.28 x 10-3
-0.26 x 10-3 to to
-1.47 x 10-3
-1.35 x 10-3 Doppler coefficient, p/F
-1.13 x 10-5 4.3
-1.13 x 10-5
-1.18 x 10-5
-1.45 x 105 to to to to
-1.67 x 10-5 1.67 x 10-5
-1.78 x 10-5
-1.07 x 10-5 Reactor Coolant System-Code Requirements Component Reactor vessel ASME III Class 1 5.2 ASME III Class 1 ASME III Class A ASME III Class A Steam generator Tube side ASME III Class 1 5.2 ASME III Class 1 ASME III Class A ASME III Class A Shell side ASME III Class 2 5.2 ASME III Class 2 ASME III Class A ASME III Class A
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 6 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Nuclear Design Data (Cont.)
Pressurizer ASME III Class 1 5.2 ASME III Class 1 ASME III Class 1 ASME III Class A Pressurizer relief (or quench) tank ASME VIII Div. 1 5.4 ASME VIII Div. 1 ASME III Class C ASME III Class C Pressurizer safety valves ASME III Class 1 5.2 ASME III Class 1 ASME III Class A ASME III Class Reactor coolant piping ASME III Class 1 5.2 ASME III Class 1 ASME III Class 1 USAS B31.7 (USAS B31.1)
(USAS B31.1)
Principal Design Parameters of the Reactor Coolant System Operating pressure, psig 2,235 5.1 2,235 2,235 2,235 Reactor inlet temperature, F 553 5.1 553 553.5 539.7 Reactor outlet temperature, F 611.2 5.1 611.2 612.5 595.1 Number of loops 2
5.1 2
2 2
Design pressure, psig 2,485 5.1 2,485 2,485 2,485 Design temperature, F 650 5.1 650 650 650 Hydrostatic test pressure (cold), psig 3,110 3,110 3,110 3,110 Total coolant volume, ft3 10,300 (without 10,300 (without 9,376 11.101 pressurizer) pressurizer)
Principal Design Parameters of the Reactor Vessel Material See Table 5.2-3 5.2 See Table 5.2-2 SA-533, Grade B SA-533, Grade B Class 1, low Class, 1, low alloy steel, alloy steel, internally clad internally clad with Type 304 with Type 304 austenitic SS austenitic SS
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 7 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Principal Design Parameters of the Reactor Vessel (contd)
Design pressure, psig 2,485 5.4 2,485 2,485 2,485 Design temperature, F 650 5.4 650 650 650 Operating pressure, psig 2,235 5.4 2,235 2,235 2,235 Inside diameter of shell, in.
172 5.4 172 157 172 Outside diameter across nozzles, in.
253 253 238 253 Overall height of vessel and enclosure head, 43-6-1/2 5.4 43-6-1/2 43-4-1/6 41-11-3/4 ft-in. to top of CEDM nozzle Minimum clad thickness, in.
1/8 5.4 1/8 1/8 5/16 Principal Design Parameters of the Steam Generators Number of Units 2
5.4 2
2 2
Type Vertical U-tube 5.5 Vertical Untube Vertical U-tube Vertical U-tube with integral with integral with integral with integral moisture separator moisture separator moisture seperator moisture separator Tube material Inconel(ASME SB-163) 5.4 Inconel(ASME SB-163)
NiCrFe alloy NiCrFe alloy Shell material SA-533 Gr. B SA-533 Gr B SA-533 Gr B SA-533 Gr. B Class 1 and Class 1 and Class 1 and Class 1 and SA-516, Gr. 70 SA-516, Gr. 70 SA-516, Gr. 70 SA-516, Cr. 70 Tube side design pressure, psig 2,485 5.4 2,485 2,485 2,485 Tube side design temperature, F 650 5.4 650 650 650 Tube side design flow, lb/hr 74 x 106 5.4 74 x 106 60.2 x 106 61 x 106 Shell side design pressure, psia 1,100 5.4 1,100 1,100 1,000 Shell side design temperature, F 560 5.4 560 560 550
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 8 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Principal Design Parameters of the Steam Generators (Cont)
Operating pressure, tube side, nominal, psig 2,235 5.4 2,235 2,235 2,235 Operating pressure, shell side, maxim-, psig 985 985 985 885 Maximum moisture at outlet at full load, %
0.2 5.4 0.2 0.2 0.2 Hydrostatic test pressure, tube side (cold) psig 3,110 3,110 3,110 3,110 Steam pressure, at full power, psia 900 5.4 900 900 815 Steam temperature, at full power, F 532 5.4 532 531.95 520.3 Principal Design Parameters of the Reactor Coolant Pumps Number of units 4
5.4 4
4 4
Type Vertical, single Vertical, single Vertical, single Vertical, single stage radial flow stage radial flow stage centrifugal stage centrifugal with bottom with bottom with bottom with bottom suction and suction and suction and suction and horizontal horizontal horizontal horizontal discharge discharge discharge discharge Design pressure, psig 2,485 5.4 2,485 2,45 2,485 Design temperature, F 650 5.4 650 650 650 Operating pressure, nominal psig 2,235 5.4 2,235 2,235 2,235 Suction temperature, F 553 5.4 553 553.5 540 Design capacity, gal/min 99,000 5.4 99,000 80,000 80,000 Design head, ft 310 5.4 310 275 250 Hydrostatic test Pressure (cold), psig 3,110 3,110 3,110 3,110
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 9 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Principal Design Parameters of the Reactor Coolant Pumps (Contd)
Motor type AC induction, AC induction, AC induction, AC induction, single speed single speed single speed single speed Motor rating, hp 9,700 9,700 6,500 6,500 Principal Design Parameters of the Reactor Coolant Piping Material SA-516, Gr 70 SA-516, Gr 70 SA-516, Gr 70 with nominal with nominal with nominal 7/32 SS Clad 7/32 SS Clad 7/32 SS Clad Hot leg ID, in.
42 5.4 42 42 42 Cold leg ID, in.
30 5.4 30 30 30 Between pump and steam generator ID, in.
30 5.4 30 30 30 Engineered Safety Feature Safety injection system No. of high pressure pumps 3
6.3 3
3 3
No. of low pressure pumps 2
6.3 2
2 2
Containment spray No. of pumps 2
6.2 2
2 2
Containment fan coolers No. of units 4
6.2 4
4 4
Air flow capacity, each at emergency conditions, ft3/min 35,000 6.2 31,000 50,000 55,800 Safety injection tanks, number 4
6.3 4
4 4
Emergency power Diesel-generator unit 2
8.3 4 (for two units) 2 2
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 10 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Principal Design Parameters of the Reactor Coolant Piping (Contd)
Containment System Parameters Type Steel containment Steel-lined Steel-lined Steel containment vessel with prestressed post prestressed post vessel with cylindrical cylindrical shell, tensioned con-tensioned con-shell, hemisperical hemisperical dome crete cylinder, crete cylinder, dome and ellipsoidal and ellipsoidal curve dome roof.
curved dome roof bottom - ASME Code, bottom - ASME Code,Section III, Class B,Section III, Class MC, surrounded by reinforced surrounded by concrete Shield Building reinforced concrete Shield Building Design Parameters - Containment Inside Diameter, ft.
140 3.8 150 116 140 Height, ft.
240.5 3.8 172 207 232 Free volume, ft3 2,677,000 6.2 2,335,000 1,780,000 2,500,000 Reference accident Pressure, psig 44 3.8 60 54 44 Steel Thickness, in.
Vertical Wall 1.90 3.8 Not applicable Not applicable 1.91 Hemispherical Head 0.95 Not applicable Not applicable 0.95 Knuckles 2.25 Not applicable Not applicable 2.25 Concrete Thickness, ft.
Vertical Wall Not Applicable 3.8 4 1/3 3 3/4 Not applicable Dome Not Applicable 3 3/4 3 1/4 Not applicable
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 11 of 13) Revision 9 (12/97)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 Principal Design Parameters of the Reactor Coolant Piping (Contd)
Design Parameters - Shield Building 3.8 Not applicable Not applicable Inside Diameter, ft-148 148 Height, ft. (top of foundation to top of dome) 249.5 230.5 Concrete Thickness, ft.
Vertical Wall 3
3 Dome 2.5 2.5 Containment Leak Prevention and Mitigation Systems Leak-tight pene-6.2 Leak-tight pen-Leak-tight pene-Leak-tight pene-tration Automatic tration, and tration, and tration. Automatic isolation where continuous steel continuous stee isolation where required.
liner. Automatic liner. Automatic required.
isolation where isolation where required. The ex-required.
haust from pene-tration rooms to vent.
Gaseous Effluent Purge Discharge through 6.2 Discharge thru Discharge thru Discharge thru vent.
vent.
vent.
vent.
RADIOACTIVE WASTE MANAGEMENT SYSTEM Liquid Waste Processing Systems Reactor Coolant Waste Holdup Tank 11.2 (BMS)
Number 4
1/2 4
4 Capacity (Gal.), each 45,000 6,000/25,000 51,270 40,000
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 12 of 13) Revision 11-B (06/02)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 RADIOACTIVE WASTE MANAGEMENT SYSTEM (Cont'd)
Degasifier
Number Flash Tank*
1 (Gas Stripper) 1 Flash Tank
Capacity (gpm)
Concentrators Number 2
1 (For 2 units) 1 1
Capacity (gpm) 20 50 gpm 20 2
Gaseous Waste Processing Systems Waste Gas Decay Tank 11.3 Number 3 3
6 (For 2 units) 3 3
Capacity (ft ), each 600 500 300 144 Pressure (psig) 380 150 380 190 Hold-up Time (days) 60 30 30 30 ELECTRIC SYSTEMS Number of Offsite Circuits 7
8.2.1.1 8
3 3
Number of Incoming Lines to Startup Transformers 2
8.2 2
2 2
Number of Startup Transformers 2
8.2 4
1+1(shared) 2 Number of Main Unit Transformers (Three Phase) 2 8.2 1
3 (single phase) 2 Number of 4.16 KV Engineered Safety Features System Buses 3
8.3 3
2 3
Number of 480V Engineered Safety Features System Buses (Power Centers) 3 8.3 3
2 4
Number of 120V AC Vital Buses 8
8.3 4
4 3
Number of Standby Diesel Generators 2
8.3 2
2 2
Diesel Generator Rating (KW) 4400 8.3 4700 2850 3500
(*) The Flash Tank is inactive per ER-W3-00-0225-00-00.
WSES-FSAR-UNIT-3 TABLE 1.3-1 (Sheet 13 of 13) Revision 11-A (02/02)
PLANT PARAMETER COMPARISON FOR WATERFORD 3 CYCLE 1 Waterford 3 Reference San Onofre St. Lucie Item Cycle 1 Section Units 2 and 3 ANO-2 Unit 1 INSTRUMENTATION SYSTEMS*
Reactor Protective System 7.2 7.2 7.2 7.2 Reactor and Reactor Coolant System 7.7.1.1 7.7.1.1 7.7.1.1 7.7.1.1 7.1.1.2 7.7.1.2 7.7.1.2 7.7.1.2 Steam and Feedwater Control System 7.7.1.3 7.7.1.3 7.7.1.3 7.7.1.3 Nuclear Instrumentation 7.2.1.1 7.2.1.1 7.2.1.1 7.2.1.1 Non-Nuclear Process Instrumentation 7.5.1.5 7.5.1.5 7.5.1.5 7.5.1.5 CEA Position Instrumentation 7.5.1.3 7.5.1.3 7.5.1.3 7.5.1.3
¨ (DRN 01-758)
- This section is not suited for tabular description. SAR section numbers have been included for the location of the detailed description of each system.