ML21102A352
| ML21102A352 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/02/2021 |
| From: | Point Beach |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21102A337 | List:
|
| References | |
| NRC-2021-0009 | |
| Download: ML21102A352 (21) | |
Text
Appendix B Table of Contents FSAR APPENDIX B TABLE OF CONTENTS B.1 DOES NOT EXIST B.2 DESIGN PARAMETERS AND PLANT COMPARISONS (Historical) - - - - B.2-1 B.3 INITIAL PLANT DESIGN - - - - - - - - - - - - - - - - - - - - - - - - - B.3-1 UFSAR 2017 Page TOC - B-i of i
Plant Comparisons FSAR Appendix B.2 (Historical)
B.2 DESIGN PARAMETERS AND PLANT COMPARISONS The design parameters of the Point Beach Nuclear Plant were initially provided in a comparison with H. B. Robinson, Indian Point 2, and Ginna Station. This comparison provided an informational reference of similar aspects of similar plants constructed during the same time period to demonstrate that the technology employed at Point Beach was proven in multiple applications.
The design parameters listed were considered valid at the time of license issuance, and have been retained for historical context in the following Table B.2-1 and paragraphs. Note: The information provided is not currently reflective of the PBNP design specifications, and therefore should not be used as the basis of any Safety Evaluation without prior verification against current information provided elsewhere in the FSAR.
In 2003 a measurement uncertainty recapture power uprate was performed increasing the rated thermal power level to 1540 MWt. The tables of this section have not been updated since this Appendix is historical.
Design Highlights The design of each Point Beach unit is based upon proven concepts which have been developed and successfully applied in the construction of pressurized water reactor systems. In subsequent paragraphs, a few of the design features are listed which represent slight variation or extrapolations from units presently operating such as San Onofre and Connecticut-Yankee.
POWER LEVEL - The license application power level of 1518.5 MWt is smaller than the capability of the Prairie Island plant and larger than the capability of the Ginna plant. This level is a reasonable increase over power levels of pressurized water reactors now operating.
REACTOR COOLANT LOOPS - The Reactor Coolant System for each Point Beach unit consists of two loops, the same as the Prairie Island and Ginna Units.
PEAK SPECIFIC POWER - Based on the design hot channel factors, operation at a primary heat output of 1518.5 MWt corresponds to a peak specific power of 16.0 kw/ft. This design rating is slightly lower than that licensed in Ginna (16.5 kw/ft) as well as that of Prairie Island (17.4 kw/ft). The maximum overpower condition is 17.9 kw/ft (112%) compared to 19.6 kw/ft (118%) for Prairie Island and 18.5 kw/ft for Ginna.
FUEL ASSEMBLY DESIGN - The fuel assembly design incorporates the rod cluster control concept in a canless assembly utilizing a spring clip grid to provide support for the 14 x 14 array of fuel rods. This concept incorporates the advantages of the Yankee canless fuel assembly and the Saxton spring clip grid with the rod cluster control scheme. Extensive out-of-pile tests have been performed on this concept and operating experience is available from the San Onofre and Connecticut-Yankee plants.
ENGINEERED SAFETY FEATURES - The engineered safety features provided are similar to those provided for the Connecticut-Yankee plant, augmented by borated water injection accumulators. There is a safety injection system of the Connecticut-Yankee type which can be operated from emergency on-site diesel power. The system design is such that it can be tested UFSAR 2017 Page B.2-1 of 15
Plant Comparisons FSAR Appendix B.2 (Historical) while the plant is at power. There is air recirculation cooling for post-loss-of-coolant conditions which utilizes the normal ventilation fans. A containment spray system provides cool, chemically-treated, borated water spray into the containment atmosphere for additional cooling capacity, and provides a means of rapidly reducing the concentration of airborne halogen fission products in the containment atmosphere.
EMERGENCY POWER - In addition to the multiple ties to outside sources for emergency power, four diesel generator units are provided as backup power supplies for the case of loss of all outside power. Each generator is capable of operating sufficient safety injection and containment cooling equipment to ensure an acceptable post-loss-of-coolant pressure transient in the affected unit, and safe shutdown of the other unit.
NET LOAD REJECTION - Each of the Point Beach units is designed to accept loss of 50% of external load without a reactor or turbine trip. This is accomplished by an automatic control system which dumps steam to the condenser and atmosphere as a short term supplemental load to provide time for the reactor control system to reduce the reactor output without exceeding acceptable core and coolant conditions. No unique or unproven features are required in the reactor control system to accomplish this.
UFSAR 2017 Page B.2-2 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Total Primary Heat Output, MWt 1518.5 2200 2758 1300 1 Total Core Heat Output, Btu/hr 5181x106 7479x106 9413x106 4437x106 2 Heat Generated in Fuel,% 97.4 97.4 97.4 97.4 3 Maximum Thermal Overpower 12% 12% 12% 12% 4 System Pressure, Nominal, psia 2250 2250 2250 2250 5 System Pressure, Minimum Steady State, psia 2220 2220 2220 2220 6 Hot Channel Factors Heat Flux, Fq 2.32 3.23 3.23 3.38 7 Enthalpy Rise, FH 1.60 1.77 1.77 1.77 8 DNB Ratio at Nominal Conditions 2.11 1.81 2.00 2.15 9 Minimum DNBR for Design Transients 1.30 1.30 1.30 1.30 10 Coolant Flow Total Flow Rate, lb/hr 66.7x106 101.5x106 136.3x106 67.3x106 11 Effective Flow Rate for Heat Transfer, lb/hr 63.6x106 97.0x106 130x106 64.3x106 12 Effective Flow Area for Heat Transfer, ft2 27.0 41.8 51.4 27.0 13 Average Velocity Along Fuel Rods, ft/sec 15.0 14.3 15.4 14.7 14 UFSAR 2017 Page B.2-3 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Average Mass Velocity, lb/hr-ft2 2.37x106 2.32x106 2.53x106 2.38x106 15 Coolant Temperatures, °F Nominal Inlet, °F 552.5 546.2 543 551.9 16 Maximum Inlet Due to Instrumentation Error and Deadband, °F 556.5 550.2 547 555.9 17 Average Rise in Vessel, °F 57.6 55.9 53.0 49.5 18 Average Rise in Core, °F 60.0 58.3 55.5 52 19 Average in Core, °F 582.5 575.4 571.0 578.0 20 Average in Vessel, °F 581.3 574.2 569.5 577.0 21 Nominal Outlet of Hot Channel, °F 642.9 642 633.5 634.0 22 Average Film Coefficient, Btu/hr-ft2-F 5600 5400 5790 5590 23 Average Film Temperature Difference, °F 31.0 31.8 30.3 26.9 24 Heat Transfer at 100% Power Active Heat Transfer Surface Area, ft2 28,715 42,460 52,200 28,715 25 Average Heat Flux, Btu/hr-ft2 175,800 171,600 175,600 150,500 26 Maximum Heat Flux, Btu/hr-ft2 491,000 554,200 567,300 508,700 27 Average Thermal Output, kw/ft 5.7 5.5 5.7 4.88 28 Maximum Thermal Output, kw/ft 16.0 17.9 18.4 16.52 29 UFSAR 2017 Page B.2-4 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Maximum Clad Surface Temperature at Nominal Pressure, °F 657 657 657 657 30 Fuel Central Temperature, °F Maximum at 100% Power 3750 4030 4090 3880 31 Maximum at Overpower 4000 4300 4380 4100 32 Thermal Output, kw/ft at Maximum Overpower 17.9 20.0 20.6 18.5 33 CORE MECHANICAL DESIGN PARAMETERS Fuel Assemblies Design RCC Canless RCC Canless RCC Canless RCC Can- 34 14x14 15x15 15x15 less 14x14 Rod Pitch, in. 0.556 0.563 0.563 0.556 35 Overall Dimensions, in. 7.763x7.763 8.426x8.426 8.426x8.426 7.763x7.763 36 Fuel Assemblies Fuel Weight (as UO2), pounds 118,729 176,200 216,000 120,782 37 Total Weight, pounds 154,519 226,200 276,000 152,895 38 Number of Grids per Assembly 7 7 9 9 39 UFSAR 2017 Page B.2-5 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Fuel Rods Number 21,659 32,028 39,372 21,659 40 Outside Diameter, in. 0.422 0.422 0.422 0.422 41 Diametral Gap, in. 0.0065 0.0065 0.0065 0.0065 42 Clad Thickness, in. 0.0243 0.0243 0.0243 0.0243 43 Clad Material Zircaloy Zircaloy Zircaloy Zircaloy 44 Fuel Pellets Material UO2 Sintered UO2 Sintered UO2 Sintered UO2 Sintered 45 Density (% of Theoretical) Unit 1 94-92-91 94-92-91 94-92-91-93 46 94-92-91 Unit 2 94-93-92 Diameter, in. 0.3669 0.3669 0.3669 0.3669 47 Length, in. 0.6000 0.6000 0.6000 0.6000 48 Rod Cluster Control Assemblies Neutron Absorber 5% Cd-15% 5% Cd-15% 5% Cd-15% 5% Cd-15% 49 In-80% Ag. In-80% Ag. In-80% Ag. In-80% Ag.
UFSAR 2017 Page B.2-6 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Cladding Material Type 304 Type 304 Type 304 Type 304 50 SS-Cold SS-Cold Wrkd. SS-Cold SS-Cold Wrkd. Wrkd. Wrkd.
Rod Cluster Control Assemblies Clad Thickness, in. 0.019 0.019 0.019 0.019 51 Number of Clusters 33 53 53 29 52 Number of Control Rods per Cluster 16 20 20 16 53 Core Structure Core Barrel I.D./O.D., in. 109.0/112.5 133.875/ 148.0/152.5 109.0/112.5 54 137.875 Thermal Shield I.D./O.D., in. 115.3/122.5 158.5/164.0 115.3/122.5 55 Structural Characteristics Fuel Weight (as UO2), lbs. 118,729 176,200 216,000 120,130 56 Clad Weight, lbs. 24,260 36,300 44,600 22,440 57 Core Diameter, in. (Equivalent) 96.5 119.5 132.5 96.5 58 Core Height, in. (Active Fuel) 144 144 144 144 59 UFSAR 2017 Page B.2-7 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Reflector Thickness and Composition Top - Water plus Steel, in. 10 10 10 10 60 Bottom - Water plus Steel, in. 10 10 10 10 61 Side - Water plus Steel, in. 15 15 15 15 62 H20/U, (Cold Volume Ratio) 4.20 4.18 4.18 4.08 63 Number of Fuel Assemblies 121 157 193 121 64 UO2 Rods per Assembly 179 204 204 179 65 Performance Characteristics Loading Technique 3 region, 3 region, 3 region, 3 region, 66 non-uniform non-uniform non-uniform non-uniform Fuel Discharge Burnup, MWD/MTU Average First Cycle 15,100 14,500 14,200 ~14,900 67 Equilibrium Region Average 33,000 33,000 24,700 ~24,400 68 Feed Enrichments, w/o Region 1 2.27 1.85 2.2 2.44 69 Region 2 3.03 2.55 2.7 2.78 70 Region 3 3.40 3.10 3.2 3.48 71 Equilibrium 3.40 3.10 UFSAR 2017 Page B.2-8 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Control Characteristics Effective Multiplication (Beginning of Life)
Cold, No Power, Clean 1.211 1.180 1.257 1.188 72 Hot, No Power, Clean 1.167 1.38 1.999 1.137 73 Hot, Fuel Power, Xe and Sm Equilibrium 1.113 1.077 1.152 1.080 74 Rod Cluster Control Assemblies Material 5% Cd-15% 5% Cd-15% 5% Cd-15% 5% Cd-15% 75 In-80% Ag. In-80% Ag. In-80% Ag. In-80% Ag.
Number of RCC Assemblies 37 53 53 33 76 Number of Absorbers per RCC Assembly 16 20 20 16 77 Total Rod Worth See Table See Table See Table 6.8% 78 3.2.1-3 3.2.1-3 3.2.1-3 Boron Concentrations To shut reactor down with no rods inserted, clean (keff = .99) Cold/hot 1598 ppm/ 1250 ppm/ 1480 ppm/ 1160 ppm/ 79 1676 ppm 1210 ppm 1370 ppm 820 ppm To control at power with no rods inserted, clean/equilibrium xenon and samarium 1465 ppm/ 1000 ppm/920 1200 ppm/ 1310 ppm/ 80 1007 ppm ppm 780 ppm 890 ppm UFSAR 2017 Page B.2-9 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Boron Worth, Hot 1% k/k/130 7.3 k/k 1% k/k/89 1% k/k/120 81 ppm ppm ppm Boron Worth, Cold 1% k/k/98 5.6 k/k 1% k/k/72 1% k/k/90 82 ppm ppm ppm Kinetic Characteristics Moderator Temperature Coefficient (k/k/oF) +0.3x10-4 to +0.3x10-4 to -0.3x10-4 to +0.5x10-4 to 83
-3.5x10-4 -3.5x10-4 -3.0x10-4 -3.5x10-4 Moderator Pressure Coefficient (k/k/psi) -0.3x10-6 to -0.3x10-6 to +0.3x10-6 to -0.5x10-6 to 84 3.5x10-6 3.5x10-6 +3.0x10-6 3.5x10-6 Moderator Void Coefficient -0.10 to -0.30 +0.5x10-3 to +0.03 to -0.30 -0.10 to -0.30 85
-2.5x10-3 k/k/g/cm3 k/k/% void k/k/g/cm-3 k/k/g/cm3 Doppler Coefficient (k/k/oF) -1x10-5 to -1x10-5 to -1.1x10-5 to -1.1x10-5 to 86
-1.6x10-5 -1.6x10-5 +1.8x10-5 1.8x10-5 UFSAR 2017 Page B.2-10 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
REACTOR COOLANT SYSTEM - CODE REQUIREMENTS Reactor Vessel ASME III ASME III ASME III ASME III 87 Class A Class A Class A Class A Steam Generator Tube Side ASME III ASME III ASME III ASME III 88 Class A Class A Class A Class A Shell Side ASME III ASME III ASME III ASME III 89 Class C* Class C* Class C* Class C*
Pressurizer ASME III ASME III ASME III ASME III 90 Class A Class A Class A Class A Pressurizer Relief Tank ASME III ASME III ASME III ASME III 91 Class C Class C Class C Class C Pressurizer Safety Valves ASME III ASME III ASME III ASME III 92 Reactor Coolant Piping USAS B31.1 USAS B31.1 USAS B31.1 USAS B31.1 93
- The shell side of the steam generator conforms to the requirements for Class A vessels and is so stamped as permitted under the rules of Section III.
PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT SYSTEM Reactor Primary Heat Output, MWt 1518.5 2200 2758 1300 94 UFSAR 2017 Page B.2-11 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Reactor Primary Heat Output, Btu/hr 5181x106 7508x106 9413x106 4437x106 95 Operating Pressure, psig 2235 2235 2235 2235 96 Reactor Inlet Temperature 552.5 546.2 543 551.9 97 Reactor Outlet Temperature 610.1 602.1 596.0 601.4 98 Number of Loops 2 3 4 2 99 Design Pressure, psig 2485 2485 2485 2485 100 Design Temperature, oF 650 650 650 650 101 Hydrostatic Test Pressure (Cold), psig 3110 3110 3110 3110 102 Coolant Volume, including pressurizer, cu. ft. 6450 9088 12,600 6245 103 Total Reactor Flow, gpm 178,000 268,500 358,800 180,000 104 Material SA-302 SA-302 Grade SA-302 SA-302 105 Grade B, low B, low Grade B, low Grade B, low alloy steel, alloy steel, alloy steel, alloy steel, internally internally clad internally internally clad with with clad with clad with austenitic SS austenitic SS austenitic SS austenitic SS Design Pressure, psig 2485 2485 2485 2485 106 Design Temperature, oF 650 650 650 650 107 Operating Pressure, psig 2235 2235 2235 2235 108 UFSAR 2017 Page B.2-12 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Inside Diameter of Shell, in. 132 155.5 173 132 109 Outside Diameter Across Nozzles, in. 224-1/16 236 262-7/16 219-5/16 110 Overall Height of Vessel & Enclosure Head, ft-in. 39-0 41-6 43' 9-11/16" 39' 1-5/16" 111 Minimum Clad Thickness, in. 5/32 5/32 5/32 5/32 112 PRINCIPAL DESIGN PARAMETERS OF THE STEAM GENERATORS Number of Units 2 3 4 2 113 Type Vertical Vertical U-tube Vertical Vertical 114 U-tube with with U-tube with U-tube with interal- integral- integral- integral-moisture moisture moisture moisture separator separator separator separator Tube Material Inconel Inconel Inconel Inconel 115 Shell Material Carbon Steel Carbon Steel Carbon Steel Carbon Steel 116 Tube Side Design Pressure, psig 2485 2485 2485 2485 117 Tube Side Design Temperature, oF 650 650 650 650 118 Tube Side Design Flow, lb/hr 33.35x106 33.93x106 34.07x106 33.63x106 119 Shell Side Design Pressure, psig 1085 1085 1085 1085 120 Shell Side Design Temperature, oF 556 556 556 556 121 UFSAR 2017 Page B.2-13 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Operating Pressure, Tube Side, Nominal, psig 2235 2235 2235 2235 122 Operating Pressure, Shell Side, Maximum, psi 1020 1020 1015.3 1020 123 Maximum Moisture at Outlet at Full Load, % 1/4 1/4 1/4 1/4 124 Hydrostatic Test Pressure, Tube Side (Cold), psig 3110 3110 3110 3110 125 PRINCIPAL DESIGN PARAMETERS OF THE REACTOR COOLANT PUMPS Number of Units 2 3 4 2 126 Type Vertical, Vertical, single Vertical, Vertical, 127 single stage stage single stage single stage radial flow radial flow radial flow radial flow with bottom with bottom with bottom with bottom suction & suction & suction & suction &
horiz. disch. horiz. disch. horiz. disch. horiz. disch.
Design Pressure, psig 2485 2485 2485 2485 128 Design Temperature, oF 650 650 650 650 129 Operating Pressure, Nominal, psig 2235 2235 2235 2235 130 Suction Temperature, oF 551.5 546.5 556 551.9 131 Design Capacity, gpm 89,000 88,500 90,000 90,000 132 Design Head, ft. 259 261 252 252 133 Hydrostatic Test Pressure (Cold), psig 3110 3110 3110 3110 134 UFSAR 2017 Page B.2-14 of 15
Plant Comparisons FSAR Appendix B.2 (Historical)
Table B.2-1 COMPARISON OF DESIGN PARAMETERS (See General Note)
PBNP U1/U2 Robinson 2 Indian Point 2 R.E. Ginna Reference Thermal and Hydraulic Parameters Final Report Final Report Final Report Final Report Line No.
Motor Type AC induc. AC induc. AC induc. AC induc. 135 single speed single speed single speed single speed air cooled air cooled air cooled air cooled Motor Rating (Nameplate) 6000 HP 6000 HP 6000 HP 5500 HP 136 Material Austenitic SS Austenitic SS Austenitic SS Austenitic SS 137 Hot Leg - I.D., in. 29 29 29 29 138 Cold Leg - I.D., in. 27-1/2 27-1/2 27-1/2 27-1/2 139 Between Pump and Steam Generator - I.D., in. 31 31 31 31 140 Design Pressure, psig 2485 2485 2485 2485 141 UFSAR 2017 Page B.2-15 of 15
Initial Plant Design FSAR Appendix B.3 B.3 INITIAL PLANT DESIGN Research and development (as defined in Section 50.2 of the Commission's regulations) was conducted regarding core design details and parameters, analytical methods for kinetics calculations, thermal shock and its effects on reactor vessel integrity, the safety injection (emergency core cooling) system, xenon stability and related control systems, containment spray additive effectiveness, and capability of reactor internals to resist blowdown forces.
Core Design The nuclear design, including fuel configuration and enrichments, control rod pattern and worths, reactivity coefficients, and boron requirements are presented in Section 3.2 and the thermal-hydraulics design parameters are also in Section 3.2. Section 3.2 presents the fuel, fuel rod, fuel assembly, and control rod mechanical design. The core design incorporates fixed burnable poison rods (Reference 1) in the initial loading and, when necessary, in subsequent core reloads to ensure a negative moderator reactivity temperature coefficient at operating temperature. This improves reactor stability and lessens the consequences of a rod ejection or loss-of-coolant accident. The mechanical design is presented in Section 3.2.
Development Of Analytical Methods For Reactivity Transients From Rod Ejection Accidents A control rod ejection accident is not considered credible since it would require the fracture of a control rod mechanism housing. Nevertheless, the reactivity and associated pressure and temperature transients for this accident have been analyzed. Rod ejection analyses for this plant were performed using the CHIC-KIN code(Reference 2), which uses a point reactor kinetics model and a single channel fuel and coolant description. The rod ejection analysis results are given in Section 14.2.6 of this report, together with a brief description of the CHIC-KIN code.
These analyses show that the temperature and pressure transients associated with a rod ejection accident do not cause any consequential damage to the reactor coolant system. The consequences of a rod ejection accident are now lessened because the moderator coefficient of reactivity is always negative at operating conditions. In addition, the effects of rod ejection are inherently limited in this reactor, in which boric acid chemical shim is employed, since full-length control rods need only to be inserted sufficiently to handle load changes.
The initial cores contain fixed burnable poison rods. These, by allowing a reduction in the chemical shim concentration, ensure that the moderator coefficient of reactivity is always negative at operating temperature. The burnable poison rods, contain borosilicate glass. Critical experiments were conducted at the Westinghouse Reactor Evaluation Center using rods containing 12.8 w/o boron and Zircaloy clad UO2 fuel rods, 2.27% enriched. These values are also typical of this plant's initial core. The experiments showed that standard analytical methods can be used to calculate the reactivity worth of the burnable poison rods. The design basis and critical experiments are described in Reference 1. In-core testing completed in the Saxton reactor showed satisfactory performance of these rods.
UFSAR 1998 Page B.3-1 of 5
Initial Plant Design FSAR Appendix B.3 Safety Injection System (SI) Design The design of the safety injection system includes nitrogen-pressurized accumulators to inject borated water into the reactor coolant system to rapidly and reliably reflood the core following a loss-of-coolant accident. Additional analyses have been performed to demonstrate that the accumulators, in conjunction with other components of the emergency core cooling system, can adequately cool the core for any pipe rupture. These analyses are presented in Section 14.3. The computer code, FLASH-R, used for the blowdown phase of the loss-of-coolant accident was modified to take into account the accumulator injection.
Research and development work has also been performed on the integrity of Zircaloy-clad fuel under conditions simulating those during a loss-of-coolant accident. Under the conservatively elevated temperatures predicted for the fuel rods during loss-of-coolant accident, the clad may burst due to a combination of fuel rod internal gas pressure and the reduction of clad strength with temperature. Burst cladding could block flow channels in the core, so that core cooling by the safety injection system would be insufficient to prevent fuel rod melting. Rod burst experiments have, therefore, been conducted on Zircaloy rods. The results of single-rod tests have been presented to the AEC in WCAP-7379-L Volume I (Westinghouse Proprietary) and Volume II.
The results of multi-rod tests have been reported to the AEC in WCAP-7495-L.
Systems For Reactor Control During Xenon Instabilities Extensive analytical work has been performed on reactor core stability(Reference 3, Reference 4, and Reference 5). These indicated that a core of this size may be unstable against axial power redistribution, but is nominally stable against transverse (denoted X-Y) power oscillations. The plant was, therefore, provided with instrumentation and control equipment which would allow the operator to detect and suppress the axial power oscillations.
The original plant design provided for part-length control rods to control axial power oscillations which could result from the potential of power spatial redistribution caused by instabilities in local xenon concentration. Initial plant operations established that part-length control rods were not necessary for control of axial power oscillation. The part-length control rods at Point Beach Nuclear Plant Units 1 and 2 were subsequently removed.
Control information for axial power oscillation suppression is obtained from four long ion chambers, each divided into an upper and lower section mounted vertically outside the core. Both calculation and experimental measurements at SENA, San Onofre, and Haddam Neck have shown that this out-of-core instrumentation represents in-core power distribution adequately for power distribution control(Reference 5).
The control strategy is based on the difference in output between the top and bottom sections of the long ion chambers. If the operator allows axial power imbalance to exceed operating limits, various levels of protection are invoked automatically. These include generation of alarms, turbine power cutback, blocking of control rod withdrawal, and reactor trip. This capability is described in Section 7.0.
UFSAR 1998 Page B.3-2 of 5
Initial Plant Design FSAR Appendix B.3 Containment Spray Additive For Iodine Removal Initially, sodium thiosulphate, Na2S2O3, was proposed as the iodine removal additive to the boric acid containment spray, but an evaluation program led to the selection of sodium hydroxide, NaOH. The results of the evaluation program are detailed in Reference 6 and are summarized briefly below:
- 1. Chemical Characteristics The Na2S2O3 solution was found to be oxidized by air at the post-accident temperatures in containment. NaOH was not unstable in this way.
- 2. Iodine Removal Characteristics The removal efficiency of the NaOH solution (at pH not less than 9.5) was comparable to that of the Na2S2O3 solution.
- 3. Materials Compatibility Corrosion rates of copper and copper-alloy heat exchanger tubing were reduced by more than an order of magnitude compared with high pH Na2S2O3 solution and were acceptably low (<0.01 mils/month at 1000F) for the application. These tests showed that pitting or local corrosion did not occur.
- 4. Radiolysis The NaOH solution was radiolytically stable, and liberates significantly less net hydrogen than the unstable Na2S2O3 solution.
Therefore, further testing has centered on the use of NaOH as the spray additive leading to the development of a technical basis for its inclusion in the plant engineered safety features as a means of fixing absorbed iodine, enhancing the natural rate of deposition of I2, and thus lowering the calculated off-site thyroid dose resulting from a postulated release of fission products to the containment atmosphere.
Section 6 gives a further discussion of iodine removal by the containment spray system.
Blowdown Capability Of Reactor Internals The forces exerted on reactor internals and the core following a loss-of-coolant accident are computed by employing the BLODWN-2 digital computer program developed for the space-time-dependent analysis of multi-loop PWR plants. This program and the models used are discussed in Section 14.3.3.
Reactor Vessel Thermal Shock Research was performed prior to and following the issuance of the Point Beach Operating Licenses to determine the effect of the addition of cold water from the accumulators to the reactor UFSAR 1998 Page B.3-3 of 5
Initial Plant Design FSAR Appendix B.3 pressure vessel. This research considered three failure modes: the ductile failure mode, the fatigue yielding mode and the brittle failure mode. Analysis of the ductile and fatigue modes determined that reactor vessel integrity is maintained following addition of the accumulator water.
Extensive analysis of the brittle failure mode demonstrated adequate reactor vessel fracture toughness to prevent brittle failure for a period of several years of plant operation.
Subsequently, but before the end of the analyzed period, the NRC issued 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.
10 CFR 50.61 contains screening criteria for material fracture toughness, such that, if the materials of construction of the reactor vessel for a nuclear power plant maintain fracture toughness in compliance with the screening criteria, the functional integrity of the reactor vessel is ensured. It has been demonstrated that Point Beach Units 1 and 2 have adequate fracture toughness to be in compliance with the screening criteria of 10 CFR 50.61 through the end of their Operating Licenses. Therefore, brittle failure of the Point Beach reactor vessels is not a credible failure mode.
Identification Of Contractors The Licensee engaged or approved the engagement of the contractors identified below in connection with the design and construction of the Point Beach Nuclear Plant. However, irrespective of the contractual arrangements discussed below, Wisconsin Electric Power Company is the sole holder of the operating licenses and, as the Licensee, is responsible for the design, construction, and operation of the Point Beach Nuclear Plant.
Point Beach Nuclear Plant was designed and built by Westinghouse Electric Corporation as prime contractor for the Licensee. Westinghouse contracted to provide a complete, safe, and operable nuclear power unit ready for commercial service. The project was directed by Westinghouse from the offices of its Atomic Power Divisions in Pittsburgh, Pennsylvania, and by Westinghouse representatives at the plant site during construction and plant startup. Westinghouse engaged the engineering firm of Bechtel Corporation, San Francisco, California, to provide the design of the structures and non-nuclear portions of the plant and to prepare specifications for the purchase and construction thereof. The Licensee reviewed the designs and specifications prepared by Westinghouse and Bechtel to assure that the general plant arrangements, equipment, and operating provisions were satisfactory.
The plant was constructed under the general direction of Westinghouse through Bechtel as the general contractor who was responsible for the management of all site construction activities and who either performed or subcontracted the work of construction and equipment erection.
NUS Corporation, Washington, D.C., was engaged as consultants on general site studies and meteorology. The firm of Murray and Trettel, Inc. assisted on meteorology. The firm of Dames and Moore, Chicago, Illinois, was engaged as consultants on earth science and geology. The engineering firm of Sargent and Lundy, Chicago, Illinois, was engaged to design cooling water facilities.
In addition, specialists in environmental sciences have participated in developing information concerning the Point Beach site. Harza Engineering Company of Chicago, Illinois, provided assistance in hydrology and the firm of John A. Blume and Associates of San Francisco, UFSAR 1998 Page B.3-4 of 5
Initial Plant Design FSAR Appendix B.3 California, provided assistance is assessing the seismic history of the sites and establishing the ground accelerations associated with the design earthquake.
Stone and Webster Engineering Corporation of Boston, Massachusetts, provided assistance in system planning and site studies.
The Licensee had qualified representatives at the site throughout construction and, with their own personnel and consultants, inspected major components and construction installations. The Licensee's initial operating force performed acceptance testing of all structures and equipment.
REFERENCES
- 1. Wood, P. M., Baller, E. A., et al, Use of Burnable Poison Rods in Westinghouse Pressurized Water Reactors, WCAP-7113 (October 1967).
- 2. Redfield, V. A., CHIC-KIN... A Fortran Program for Intermediate and Fast Transients in a Water Moderated Reactor, WAPD-TM-479 (January 1, 1965).
- 3. Poncelet, C. G. and Christie, A. M., Xenon Induced Spatial Instabilities in Large Pressurized Water Reactors, WCAP-3680-20 (March 1968).
- 4. McGaugh, J. D., The Effect of Xenon Spatial Variations and the Moderator Coefficient on Core Stability, WCAP-2983 (August 1968).
- 5. Westinghouse Proprietary Report, Power Distribution Control in Westinghouse PWRs, WCAP-7208 (October 1968).
- 6. Westinghouse Confidential Report, Investigation of Chemical Additives for Reactor Containment Sprays, WCAP-7153 (March 1968).
- 7. Westinghouse Customer Report, Fracture Mechanics Evaluation of the Wisconsin Electric Power Company and Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit 2 Reactor Vessel, WCAP-8737 (February 1977).
- 8. Westinghouse Customer Report, Fracture Mechanics Evaluation of the Wisconsin Electric Power Company and Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit 1 Reactor Vessel, WCAP-8742 (February 1977).
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