The Pennsylvania State University. The Graduate School. Department Of Mechanical & Nuclear Engineering

Transcription

1 The Pennsylvania State University The Graduate School Department Of Mechanical & Nuclear Engineering FEASIBILITY STUDY FOR THE CONCEPTUAL DESIGN OF A DUAL-CORE BOILING SUPERHEAT REACTOR A Thesis in Nuclear Engineering by Jacob Wayne Ross 29 Jacob Wayne Ross Submitted in Partial Fulfillment of the Requirements for the Degree of Master of Science May 29

2 The thesis of Jacob Wayne Ross was reviewed and approved* by the following: ii Lawrence E. Hochreiter Professor of Nuclear & Mechanical Engineering Thesis Advisor Kostadin N. Ivanov Distinguished Professor of Nuclear Engineering Arthur T. Motta Professor of Nuclear Engineering & Material Science Jack S. Brenizer, Jr. J. Lee Everett Professor of Nuclear Engineering Department of Mechanical & Nuclear Engineering Chair of Nuclear Engineering *Signatures are on file in the Graduate School

3 ABSTRACT iii For research concerning economical applications of high temperature reactor technology, a novel approach for creating a Boiling Superheat Reactor (BSR) by augmenting an Advanced Boiling Water Reactor (ABWR) vessel with a steam superheating reactor core is examined. This design utilizes Very High Temperature Reactor (VHTR) assembly inverted design concepts to develop a superheater assembly. The superheater assemblies are placed in the ABWR by modifying the ABWR s reactor vessel head and steam dryers to accommodate the superheating reactor core. A conservative estimation demonstrates that this BSR design provides an overall thermal power of 518 MW at 43.4% efficiency. The resultant electrical power output of 225 MW from the BSR over the 136 MW of the ABWR provides a substantial increase of 89 MWe.

4 TABLE OF CONTENTS iv LIST OF FIGURES...vii LIST OF TABLES...x ACKNOWLEDGEMENTS...xii Chapter 1 Introduction History Technical Concerns Feasibility Investigation...7 Chapter 2 Design Overview Pin Cell Design Assembly Design Previous BSR Design Feasibility Study Dual Core BSR Design...21 Chapter 3 Materials Investigation Thermochemical Interaction Analysis Zirconium Hydride Moderator Uranium Carbide Fuel Steel Cladding & Canning Summary...48 Chapter 4 Thermal-Hydraulic Investigation Thermal Hydraulic Program Thermal-Hydraulic Methodology Results & Analysis Axial Profile Study Fuel Radius Study Core Radius Study Core Height Study Fuel Study Twisted-Tape Insert Study Optimistic Study Summary...72 Chapter 5 Criticality Investigation Methodology...73

5 5.2 Results & Analysis Summary...81 v Chapter 6 Conclusions...82 Bibliography...86 Appendix A Thermochemical Data...89 Appendix B Thermal Hydraulic Program & Property Information...91 B.1 Thermal-Hydraulic Program...92 B.2 Property Reference Files & Data Appendix C Thermal Hydraulic Program Input Files C.1 Default Input C.2 Axial Power Shape Study Input C.3 Fuel Radius Study Input C.4 Core Radius Study Input...12 C.5 Core Height Study Input C.6 Fuel Study Input C.7 Twisted-Tape Study Input C.8 Optimistic Study Input Appendix D Thermal Hydraulic Program Output Files D.1 Default Output D.2 Axial Power Shape Output D.3 Fuel Radius Study Output...13 D.4 Core Radius Study Output D.5 Core Height Study Output D.6 Fuel Study Output D.7 Twisted-Tape Study Output D.8 Optimistic Study Output Appendix E Thermal Hydraulic Figures...14 E.1 Axial Power Shape Plots...14 E.2 Fuel Radius Study Plots E.3 Core Radius Study Plots E.4 Core Height Study Plots E.5 Fuel Study Plots E.6 Twisted-Tape Study Plots...16 E.7 Optimistic Study Plots...164

6 Appendix F MCNP Input Files vi

7 LIST OF FIGURES vii Figure 1-1: BONUS reactor design [2]...3 Figure 1-2: Pathfinder reactor vessel and coolant flow pathways [5]...5 Figure 2-1: Hexagonal pin cell for proposed BSR fuel assembly...11 Figure 2-2: Assembly cross section with pin cell boundaries shown...16 Figure 2-3: Conceptual BSR vessel design...23 Figure 2-4: Cross section of superheater core with circle for inner vessel rim...24 Figure 2-5: Cross section of boiler core with circle for inner vessel rim...25 Figure 2-6: Density and heat capacity as a function of temperature at 25 bar [11]...28 Figure 2-7: Thermal conductivity as a function of temperature at 25 bar [11]...28 Figure 3-1: Fuel temperature coefficient vs. temperature increase from the nominal temperature (UO 2 : 145 K, UC Fuel: 11 K)...41 Figure 3-2: Negative reactivity insertion vs. temperature increase from nominal temperature (UO 2 : 145 K, UC Fuel: 11 K)...42 Figure E-1: Linear Heat Rate distribution for the different axial power shapes...14 Figure E-2: Reynolds number distribution for the different axial power shapes Figure E-3: Density distribution for the different axial power shapes Figure E-4: Convective heat transfer coefficient distribution for the different axial power shapes Figure E-5: Pressure distribution for the different axial power shapes Figure E-6: Fluid temperature distribution for the different axial power shapes Figure E-7: Maximum fuel temperature distribution for the different axial power shapes Figure E-8: Linear Heat Rate distribution for the different outer fuel radii Figure E-9: Reynolds number distribution for the different outer fuel radii Figure E-1: Density distribution for the different outer fuel radii...145

8 Figure E-11: Convective heat transfer coefficient distribution for the different outer fuel radii Figure E-12: Pressure distribution for the different outer fuel radii Figure E-13: Fluid Temperature distribution for the different outer fuel radii Figure E-14: Maximum fuel temperature distribution for the different outer fuel radii Figure E-15: Linear heat rate distribution for the different numbers of assemblies Figure E-16: Reynolds number distribution for the different numbers of assemblies Figure E-17: Density distribution for the different numbers of assemblies Figure E-18: Convective heat transfer coefficient distribution for the different numbers of assemblies Figure E-19: Pressure distribution for the different numbers of assemblies...15 Figure E-2: Fluid temperature distribution for the different numbers of assemblies...15 Figure E-21: Maximum fuel temperature distribution for the different numbers of assemblies Figure E-22: Linear heat rate distribution for the different core heights Figure E-23: Reynolds number distribution for the different core heights Figure E-24: Density distribution for the different core heights Figure E-25: Convective heat transfer distribution for the different core heights Figure E-26: Pressure distribution for the different core heights Figure E-27: Fluid temperature distribution for the different core heights Figure E-28: Maximum fuel temperature distribution for the different core heights..155 Figure E-29: Linear heat rate distribution for the different fuel composition Figure E-3: Reynolds number distribution for the different fuel composition Figure E-31: Density distribution for the different fuel composition viii

9 Figure E-32: Convective heat transfer coefficient distribution for the different fuel composition Figure E-33: Pressure distribution for the different fuel composition Figure E-34: Fluid temperature distribution for the different fuel composition Figure E-35: Maximum fuel temperature distribution for the different fuel composition Figure E-36: Linear heat rate distribution for the twisted-tape insert comparison...16 Figure E-37: Reynolds number distribution for the twisted-tape insert comparison...16 Figure E-38: Density distribution for the twisted-tape insert comparison Figure E-39: Convective heat transfer coefficient distribution for the twisted-tape insert comparison Figure E-4: Pressure distribution for the twisted-tape insert comparison Figure E-41: Fluid temperature distribution for the twisted-tape insert comparison Figure E-42: Maximum fuel temperature distribution for the twisted-tape insert comparison Figure E-43: Linear heat rate distribution for the optimistic case comparison Figure E-44: Reynolds number distribution for the optimistic case comparison Figure E-45: Density distribution for the optimistic case comparison Figure E-46: Convective heat transfer coefficient distribution for the optimistic case comparison e Figure E-47: Pressure distribution for the optimistic case comparison Figure E-48: Fluid temperature distribution for the optimistic case comparison Figure E-49: Maximum fuel temperature distribution for the optimistic case comparison ix

10 LIST OF TABLES x Table 1-1: American BSRs to produce electrical power [1, 2, 3, 4]...2 Table 2-1: Geometric comparison of BSR pin cell to GE-13 pin cell...12 Table 2-2: Geometric comparison of BSR assembly to GE-13 assembly...17 Table 2-3: Power output comparison between ABWR and older BSR-type designs..2 Table 2-4: Geometric and fuel quantity comparison of superheater and boiler cores...26 Table 2-5: Power output comparison between the ABWR, the SCWR, and the dual core BSR design...26 Table 3-1: Possible assembly chemical reactions for a temperature of 298 K (25 C)...32 Table 3-2: Possible assembly chemical reactions for a temperature of 123 K (75 C)...32 Table 3-3: Mass and molar volume changes between ZrH 2 and ZrO Table 3-4: Parameters affecting fuel temperature coefficient...45 Table 3-5: Type Caption Here...46 Table 4-1: Thermal-hydraulic program input parameter list...51 Table 4-2: Default case input values...64 Table 4-3: Core output power and temperature for the axial power shapes...66 Table 4-4: Core output power and temperature for the different fuel outer radii...67 Table 4-5: Core output power and temperature for the different core radii...68 Table 4-6: Core output power and temperature for the different core heights...69 Table 4-7: Core output power and temperature for the different fuel compositions...7 Table 4-8: Core output power and temperature for the twisted-tape insert option...7 Table 4-9: Core output power and temperature for the optimistic estimate comparison...71

11 Table 5-1: Material information for reflected assembly section base case...74 xi Table 5-2: Case 1: No ZrH 2 quantized resonance scattering information...75 Table 5-3: Case 2: No carbon quantized resonance scattering information...76 Table 5-4: Case 3: Increase in ZrH 1.6 moderator density...76 Table 5-5: Case 4: Increase in ZrH 1.6 moderator temperature...76 Table 5-6 Case 5: Increase in UC fuel temperature...77 Table 5-7: Case 6: Change from UC to UO 2 fuel...77 Table 5-8: Case 7: Change to lower temperature parameters corresponding to the core inlet...78 Table 5-9: Case 8: Change to higher temperature parameters corresponding to the core outlet...78 Table 5-1: Case 9: Change to lower temperature parameters and liquid water in the channel...79 Table 5-11: Critical eigenvalue and neutron energy range data...8 Table 6-1: Power output comparison between the ABWR, the SCWR, and the dual core BSR design...83 Table A-1: A selected thermochemical list of carbon-containing compounds...89 Table A-2: A selected thermochemical list of hydrogen-containing compounds, excluding the carbon containing compounds...9 Table A-3: A selected thermochemical list of oxygen-containing compounds, excluding the carbon and hydrogen containing compounds...9 Table A-4: A selected thermochemical list of metallic elements...9

12 ACKNOWLEDGEMENTS xii Dr. Lawrence E. Hochreiter was the professor who originally conceived of the idea to attempt to revive Boiling Superheat Reactor (BSR) technology. Although this reactor design is significantly different than the one he conceived, he provided the impetus for further research into this current design and is the inspiration for it. He believed that one of the most economically feasible ways to improve reactor technology was to build off of what currently exists. He endeavored to find ways to increase the performance of existing reactors in a safe and reliable manner. The BSR concept was an avenue toward that goal. It is my hope that this research will honor his goal to improve nuclear reactor technology. He will be greatly missed.

13 Chapter 1 Introduction One area of interest in the Department of Energy s Generation IV new reactor designs is to increase the performance of nuclear reactors such that they are more cost effective through higher thermodynamic efficiencies and/or power densities. Several new designs are being considered, each of which requires significant technological investment. Based on the research presented here, it may be possible to use today s technology to modify an existing Boiling Water Reactor (BWR) to generate superheated steam such that the plant efficiency and power density is significantly improved. At the very least a new Boiling Superheat Reactor (BSR) could be constructed largely based on existing BWR technology. A conceptual design effort is initiated to investigate this possibility. BSR technology is not a new idea, but has been largely forgotten in the wake of standardization for Light Water Reactor (LWR) technology during last century. Because of the research provided in the past, it is instructive to review previous BSR technology for the purpose of identifying possible problems and limitations. 1.1 History Boiling superheat reactors were originally designed and built in the late 195 s and early 196 s as part of the Atomic Energy Nuclear Development effort. A similar effort also took place in Russia during the same time period. All these reactors were

14 small, low power devices because they were intended to provide a demonstration of the 2 concept and to generate basic design and engineering information that would allow the design concept to be scaled to larger, more commercial sizes. All the designs built were integral superheat designs with a separate boiler and superheat region in a single reactor core. The list of the US BSRs is given in Table 1-1. Table 1-1: American BSRs to produce electrical power [1, 2, 3, 4] Plant Operation Power (MWt) Exit Temp. ( C) BORAX-V BONUS Pathfinder The Boiling Reactor Experiment V (BORAX-V) Reactor was designed to contain superheat assemblies with boiling assemblies, or boiling assemblies alone. The flexibility in the design allowed the reactor to operate in three modes: (1) a pure boiler mode, (2) a boiling water annular region with a centrally located superheater region, and (3) a central boiling water region with an annular superheater. By designing a reactor to operate in a range of modes and conditions, significant BWR and BSR research could be achieved. Unfortunately, the research program was terminated before a number of reactor performance and safety research projects could be conducted [1]. The BOiling NUclear Superheater (BONUS) Reactor was designed with a central boiler region and a peripheral superheat region. Built in Puerto Rico, this reactor was a prototype explicitly designed for commercial operation. An example of this endeavor includes using a superheat fuel assembly design that consisted of separate fuel, steam, and water moderator regions that, although mechanically complicated, allowed the

15 superheat fuel to operate with a lower fuel enrichment. The core layout is shown in 3 Figure 1-1. Figure 1-1: BONUS reactor design [2]

16 4 As can be seen, the reactor inlet is directly below the boiler region of the core and is heated in a single pass through it. From there, steam and water were separated, with the water level located several feet above the core. Heated water was removed from the core to be recycled. The saturated steam rose and entered steam dryers located on the sides of the vessel. From there, the saturated steam was sent down into the superheating annular region through isolation piping. This steam makes two passes through the superheated region before rising through more steam isolation piping, and then out of the vessel. The heated water surrounds the superheated assemblies in order to provide neutron moderation. The heated water is what allows the superheated assemblies operated with enrichments consistent with LWR assemblies. More safety features were incorporated to accommodate commercial operation as well as a high temperature for the steam [2, 3]. The Pathfinder Reactor had an upward flowing peripheral boiler region and a downward flowing central superheat region with the superheated steam exiting through the bottom of the reactor vessel. The flow path as well as the overall design for the Pathfinder vessel and internals can be seen in Figure 1-2. The superheat fuel elements were thin concentric annuli of highly enriched UO 2 fuel enclosed in stainless steel cladding. Approximately 22% of the total reactor power for the Pathfinder Reactor was produced in the superheat region and the exit conditions for the steam was 44 C at a pressure of 3.72 MPa for a steam superheat value of approximately 177 C. The Pathfinder superheating resulted in a 1% improvement in cycle efficiency as compared to other Boiling Water Reactors (BWRs) during that time period [2]. However, sustained full power operation was never achieved for the Pathfinder Reactor and the plant was shutdown in 1967 [4].

17 Figure 1-2: Pathfinder reactor vessel and coolant flow pathways [5] 5

18 1.2 Technical Concerns 6 The primary technical concern for BSRs was chloride stress corrosion of the stainless steel cladding material and other stainless steel components in the reactor vessel [2]. Today, it can be concluded that the stress corrosion cracking issue for BSRs and LWRs in general has been solved through improved water chemistry controls. There have also been many improvements in nuclear grade steels since the late 196s indicating that issues with previous BSR designs would not be a problem with modern technology. This is evident from the research into Supercritical Water Reactors (SCWRs). Stainless steel and Inconel alloys have been used in reactor technology for decades and continue to be researched for the SCWR [6]. The SCWR was designed to be operated at pressures of 25 MPa with an exit temperature of 5 C and a thermodynamic efficiency of 44.8% [7]. Much of the technology derived from research of the SCWR can be applied to BSR technology. A problem that some of the BSR older designs did not have, that would be critical to any new design, would be the superheat region fuel enrichment. To compensate for a lack of moderation from the steam for the superheat region, researchers of the 196 s chose to either (1) enrich the superheat fuel assemblies to above 9%, or (2) provide explicit water channels around the superheated assemblies. The first option, from the BORAX-V and Pathfinder reactors, is no longer acceptable in the current regulatory environment. The second option, from the BONUS Reactor, proved to be complicated from a mechanical design standpoint; decreased the superheat region power density; and required more steel to be inserted into the reactor core, impacting neutron economy.

19 1.3 Feasibility Investigation 7 In order to design a physically realistic and economically viable reactor, a number of constraints have to be met. For the fuel composition, the Nuclear Regulatory Commission (NRC) sets the fissile enrichment limit at 5%. In order to achieve and maintain reactor criticality for this enrichment, a significant amount of moderation must be present. Because the density of the steam is roughly 3% that of water at the pressures of a typical BWR, steam cannot provide the necessary moderation to sustain criticality. Thus a moderator substance must be placed in the superheater assemblies. For the fuel cladding and containment, zirconium alloys currently used in LWRs are not well suited to the required high temperature steam environment of a superheated assembly. Although there is a great amount of research being conducted in this area to allow zirconium alloys to withstand a highly corrosive steam environment, the problem of strength at high temperature remains a problem [8, 9]. Thus other cladding materials are proposed to ensure cladding integrity throughout a normal BWR fuel cycle. For the layout of the vessel, there are strict economic tradeoffs for the previous BSR designs. Superheater assemblies possess inferior heat transfer properties compared to boiler assemblies due to single phase heat transfer of a gas being drastically inferior to saturated boiling heat transfer. However, the higher the temperature the steam can be heated to, the higher the efficiency of the thermodynamic cycle. By replacing boiling assemblies with superheater assemblies in a single reactor core, the designers traded higher power density and lower efficiency for lower power density and higher efficiency.

20 To be economically attractive, one must be able to gain the advantages of both types of 8 assemblies while minimizing the disadvantages. To investigate the feasibility of the BSR concept, this thesis outlines the overall design of the reactor. Chapter 2 starts with outlining the basic fuel pin cell structure, moving onto the assembly and core design, and ending with the overall vessel layout. Chapter 3 analyzes possible materials issues and places them in perspective by comparing them to issues observed in other known reactors. With the materials selection verified, Chapter 4 provides a thermal-hydraulic investigation of this design to justify various design decisions as well as gives information concerning reactor performance and limitations. Finally, Chapter 5 consists of a criticality analysis that gives insight into various materials effects as well as information on accident scenarios.

21 Chapter 2 Design Overview In order to compete with current reactor designs, this reactor design must have capabilities that other reactors do not possess. The most directly competitive reactor design being investigated by the world community is that of the Supercritical Water Reactor (SCWR). There are many advantages of the SCWR design compared to existing LWR technology. It produces higher temperature coolant, resulting in significantly higher thermodynamic cycles efficiencies. Another advantage is the simplicity of design. Because the SCWR produces a supercritical fluid, there is no need to add special structures between the outlet of the SCWR and the entrance to the power turbines to separate a two-phase fluid as required by BWRs. For Pressurized Water Reactors (PWRs), steam generators are required to produce lower pressure steam from the higher pressure water of the reactor core. For BWRs, the reactor vessel height must be significantly increased in order to accommodate steam separators and dryers. Since the BSR is based on BWR technology with steam separators and dryers required, the BSR design must be able to compete with the SCWR in ways other than simplicity of design. This BSR design must directly compete where it can, such as outlet coolant temperature, and must counter with other design advantages when it cannot, such as design simplicity. Given the competition, this BSR design generates superheated steam at a temperature comparable to that of a SCWR by modifying the internal structure of a

22 General Electric Advanced Boiling Water Reactor (ABWR) vessel to incorporate fuel 1 assemblies for a second superheater core that will superheat the separated steam from the boiling reactor core. This second reactor core would be cooled explicitly by steam from the ABWR reactor core. The effect of the added superheater core would be to increase the power output and efficiency of the existing design. The assemblies are designed according to what would be feasible with respect to enrichment, vessel geometry, material, and thermal hydraulic limitations. The assemblies consist of an inverted design, similar to a VHTR assembly, with a sealed steel can containing zirconium hydride moderator and fuel annuli. Steel tubes are aligned with the center of the fuel annuli to provide paths for steam flow. The assembly cross section is shown in Section 2.2. In designing a reactor core, the simplest and most important element is the fuel pin cell. Once a suitable pin cell has been determined, the next step is to design an assembly that can contain and support the pin cells while accommodating other structures such as reactivity control structures, core support plates, and end flow pathways. Last, the core arrangement as it relates to the other reactor vessel internal structures must be determined. 2.1 Pin Cell Design In researching possible physical dimensions for the BSR pin cells and assemblies, current BWR reactor dimensions were used as a template. This decision made the most sense because the BSR s reactivity would be controlled with control blades in a manner

23 identical to that of BWRs. The design of the fuel pin cell is shown in Figure 2-1 with 11 Table 2-1 displaying data comparing geometric information of both the BSR pin cell with that of a standard General Electric (GE)-13 pin cell. Figure 2-1: Hexagonal pin cell for proposed BSR fuel assembly

24 Table 2-1: Geometric comparison of BSR pin cell to GE-13 pin cell BSR Pin Cell GE-13 Pin Cell Radius to Side (mm) Pin Cell Area (mm 2 ) Fuel Outer Radius (mm) Fuel Area (mm 2 ) Cladding Outer Radius (mm) 12 Cladding Thickness (mm) To match existing BWR pin cell designs as much as possible, the area of the pin cell closely correlates to that of a GE-13 fuel pin cell. The deviation between the areas is due to the fact that the pin cells on the ends of the assembly must have sufficient moderator between the fuel and the assembly canning. The fuel area for a single pin cell is 54% of that compared to the GE-13 assembly pin cell. However, this deficiency is partially compensated for by the design of more pin cells in the BSR superheater assembly than the GE-13 assembly. A calculation to compare the overall fuel content is shown in Section 2.4. The hexagonal pin cell was chosen because it has a better packing fraction than a square pin cell of the same area. For LWRs, the square pin cell is favorable because the extra room from the lower packing fraction allows more coolant to flow between the fuel rods. In the superheater pin cell, the extra room is not necessary because the moderator does not provide heat removal from the assembly. For a square pin cell, the moderator is not spread out very evenly. This provides uneven neutron moderation in the azimuthal direction: low moderation in regions between the pin cell faces, high moderation along the diagonals of the pin cell. With a hexagonal pin cell, the moderator is much more

25 13 evenly spaced, resulting in more uniform burnup of the uranium carbide fuel annulus and a more even temperature distribution of the moderator. The inverted design is chosen in order to prevent thermal neutrons from being absorbed in the cladding. If a neutron born in the fuel moves into the ZrH 1.6 moderator, it is effectively thermalized and moves back into the fuel or an adjacent fuel element. The only exception to this is if the neutron is absorbed in the canning or moderator. Nonetheless, there are virtually no neutrons being absorbed in the cladding that were thermalized in the ZrH 1.6 moderator. The main benefit of this pin cell design is to optimize neutron economy. As noted in Table 2-1, the cladding is significantly thinner than that of the GE-13 fuel. The reason for this is that steel is a significant neutron absorber and minimizing the amount of this material near the fuel raises the neutron economy of the pin cell. Also, thinner cladding results in less thermal resistance to heat transfer from the steam to the fuel. The proposition that this thin cladding can be feasible is due to a number of reasons. A significant benefit to this design is that the channel tube is in a state of compression instead of tension. This can help resist corrosion as the material would be forced to contract instead of expand. Compression of the tubing would originate from two sources. First, the UC fuel annuli expands in two radial directions from thermal expansion and fission product build up, eventually contacting and compressing the cladding inward. Because the UC fuel is an annulus, the expansion in two radial directions is less detrimental and more subdued compared to a fuel pellet that expands only in the outward direction. Also, because the UC fuel operates at a significantly lower

26 14 temperature, with a maximum of 75 C, it expands far less thermally than a UO 2 pellet would. The second source of compressive stress occurs in the void swelling of the steel cladding. Better swelling resistant steels have been researched over the years, but some compression can be beneficial to the integrity of the cladding [6, 7]. If these expansion effects can be properly managed without detrimentally deforming the channel, many of the stress corrosion effects associated with cladding in tension could be mitigated or possibly eliminated. Another benefit is the differential gas pressure between the coolant side and the fuel-moderator side. For a normal LWR, the cladding starts out in compression until the cladding creeps down onto the fuel. During fuel burnup, fission gases build up in the fuel rod such that at the end of the cycle, the internal rod pressure can be significantly higher than the external coolant pressure. In the BSR assembly, the fission gases can permeate the entire volume of the assembly including the moderator as well as the fuel, giving much more volume for the fission products, thus reducing pressure buildup. Also, there is an equilibrium pressure and temperature associated with the decomposition of the ZrH 1.6 moderator into zirconium metal and hydrogen gas. As the pressure increases from fission product gas buildup, more hydrogen gas will reintegrate back into the moderator to establish a new decomposition equilibrium pressure. These effects allow the pressure gradient for the cladding to be more evenly managed for the BSR superheater assembly.

27 2.2 Assembly Design 15 In the same manner that the pin cell is hexagonal, the assembly is also made hexagonal. The cross sectional area for the assembly is exactly the same as that of the GE-13 fuel assembly. However, a GE-13 assembly possesses 74 pin cells compared to the 91 pin cells for the BSR superheater assembly. This is due to the better packing fraction of the hexagonal pin cell compared to the square pin cell. Also, the BSR superheater assembly does not have water rods. The diagram for the assembly is shown in Figure 2-2. Comparison information between the BSR assembly and a GE-13 assembly is displayed in Table 2-2. The outer light blue region is the empty steam space for the placement of control tri-blades. Because the reactivity of the reactor will be controlled by control tri-blades that possess the same thicknesses as the BWR control blades, a space matching that of the GE-13 assembly must be provided. The thin grey lining indicates the presence of canning with steel. The design requires canning to ensure containment of the ZrH 1.6 moderator and fuel from the steam environment. The canning for the superheater assembly performs the same containment functions as that of fuel cladding. However, because this assembly boxing material encompasses the entire assembly, it is denoted canning. The thickness of the canning is designed to be a little more than half that of BWR zirconium alloy canning. The resulting thickness is 1 mm. The reason for this change is the same as that for the cladding: to limit parasitic neutron losses associated with thicker canning. In this design, both the canning and the channel tubes provide the structural support for the assembly. A lower and upper support plate is welded onto the ends of the assembly in order to seal the ZrH 1.6

28 16 moderator, UC fuel, and fission products into the assembly. These support plates would be identical to allow the assembly to be symmetrical about its axial midplane. By allowing this symmetry, core reload designers would have the option of actually flipping the assemblies upside-down in order to further optimize their reload pattern. Figure 2-2: Assembly cross section with pin cell boundaries shown The assembly contains extra moderator near the canning to provide some additional reflecting capability to the neutrons attempting to exit the assembly. Note that the moderator does not provide any structural support other than as a filler material. For added lateral support for the steam tubes, it may be necessary to provide some spacer grid support. A zirconium alloy might be preferred since it could act as a sacrificial anode for the steel and provide additional zirconium metal to react with excess hydrogen. Steelhydride interaction has been studied in many respects for both the SCWR and the use of

29 hydride fuels in standard TRIGA Reactors [6, 15]. In TRIGA reactors, reactors can 17 operate nominally at 75 C and up to 12 C in transient situations [1]. However, there may be a need to investigate the possibility of providing some kind of material liner between the steel and moderator to minimize steel hydride interactions. Table 2-2: Geometric comparison of BSR assembly to GE-13 assembly BSR Assembly GE-13 Assembly Assembly Gap (mm) Assembly Lattice Area (mm 2 ) Canning Thickness (mm) Area for Fuel Pin Cells (mm 2 ) Assembly Fuel Area (mm 2 ) Previous BSR Design Feasibility Study In designing the reactor core and vessel, a careful review was undertaken to understand all the possible disadvantages from the previous designs. Since the Pathfinder reactor is among the simplest BSR configurations, a comparison is made with this design compared to an existing BWR design. The main focus is the internal structure of the vessel and the added component complexity created by incorporating superheater assemblies for a two-region single-core design. Also, qualitative predictions are made as to the impact to the neutron economy of the reactor core. For the lower core plate, some of the holes would need to be covered for the superheated region and water-tight sealed U-tubes placed so that steam coming from the top can be redirected into the assemblies. This design feature was used in the BONUS

30 18 reactor and is shown in Figure 1-1. Another option would be to create a lower plenum for the superheat region would also have to be added that is separated from the boiling region. This design feature was used in the Pathfinder reactor and is shown in Figure 1-2. Whatever method is chosen to isolate the steam from water at the lower core plate, it adds structures and increases complexity compared to a standard BWR lower core plate design. The control blades must be changed in some way so they can be inserted through the lower core plate without letting the steam coming from the top mix with the water below in any way. A possibility would be to weld control blade shells onto the lower core plate for all the superheater blades to provide steam/water isolation. In the case of inserting control blades from the top of the core, the problem of providing isolation between the steam and water remains. For the BONUS and Pathfinder reactors, control blades were only used in the boiler core. Although this design simplifies matters significantly, most modern BWRs rely on a multitude of control blades to provide fuel cycle optimization. If no control blades are used in the superheat region, significant optimization capability is removed. For the upper core plate region, the plate may have to be split into two plates. One would be for the boiling region, the other for the superheat region. Another option would be to provide vertical support only with the steam piping. Either way, special accommodations may be needed for the saturated steam tubing as well as the boiling/superheat steam separator lining. For the steam separation equipment, many of the steam separators and dryers must be removed and structures added to accommodate the superheated steam exit tubes as well as the saturated steam inlet tubes.

31 19 For the reactor core that is roughly the size of the ABWR core, it is reasonable to assume that the equivalent of at least 2 rings of assemblies would need to be removed in order to accommodate the steam isolation equipment. Some kind of shell or piping structures would have to be constructed. For the Pathfinder reactor shown in Figure 1-2, an entire shroud is placed between superheater and boiler regions. For the BONUS reactor shown in Figure 1-1, a fairly sophisticated steam piping system was necessary. Extra space was required for the piping to isolate the steam, while at the same time allowing water to reside outside the piping to provide moderation. Whatever the method used, a reactor designer would need extra space for the steam isolation equipment. To provide a rough estimate of the performance increase from superheating a single core with two heating regions, some calculations using optimistic numbers can be employed. As a rough estimate, 5 out of 872 assemblies would have to be removed for additional equipment. A portion of the assemblies could be estimated to be superheating. Also, ABWR core information is used to provide the template for the two-region core. Assuming that both regions have the same assembly heights and that there is a 4% penalty in heat transfer of the superheater assemblies compared to the boiler assemblies, the performance estimate can be made. The electric power is given in Eq. 2.1 and data displayed in Table 2-3. P = q L N N η 2.1 e H r ass cycle Where: P e : q : Electric power Linear heat rate

32 2 L H : Heated length N r : Number of fuel rods per assembly N ass : Number of assembly in the core η cycle : Cycle efficiency Table 2-3: Power output comparison between ABWR and older BSR-type designs q L H η N (kw/m) (m) r N cycle P e ass (%) (MWe) ABWR BSR: S: 11.8 S: 91 S: Old Design B: 19.6 B: 62 B: 658 Note S: superheater assembly, B: boiler assembly As can be seen in Table 2-3, the overall power increase between superheating a reactor based on previous designs compared to the existing ABWR design is very small. A power increase of 5 MWe does not provide a justification for designing this kind of BSR or changing the existing ABWR design. It is even less competitive with the SCWR design. Note that this calculation was fairly optimistic because, as will be shown in Chapter 4, more than 4% of the assemblies need to be superheating to reach the temperatures necessary to justify the given cycle efficiency of 44.8%. The optimistic estimate of 25% of the assemblies being superheating is insufficient. If more realistic estimates were used, it would result in power decrease. Also, this assumes power output is not affected by the different neutronics situation in the reactor core. This single core configuration would result in a severe neutronics penalty with all the added steel for the superheat shroud and/or steam piping. It may be possible to make these out of a zirconium alloy, but it would be even less cost

33 21 effective because (1) zirconium alloy is more expensive and (2) the irradiation-induced growth, irradiation-enhanced creep, and lower strength over these components may require them to be replaced much more often than known swelling-resistant steel alloys over the life of the reactor. Overall, the sacrifice of power density for higher cycle efficiency prevents the two-region single-core BSR design from being cost competitive with SCWR technology or even existing BWR technology. 2.4 Dual Core BSR Design The design of a separate superheating core in addition to an existing boiler core provides the ability to bypass the coupled constraints of power density vs. cycle efficiency. The reactor vessel design proposed is shown in Figure 2-3 with reactor cross section information for the ABWR-based boiler core as well as the superheater core displayed in Figures 2-4 and 2-5. Figure 2-3 highlights the fact that the steam dryers can be placed around the perimeter of the superheater core and that reactor control comes from control blades mounted into the vessel head. The purpose of Figures 2-4 and 2-5 is to demonstrate the cross sectional areas of these two reactor cores compared to each other and the inner diameter of the pressure vessel, denoted by the black circles enclosing the reactor cores. The colors are simply to provide visualization of the number of assembly rings of the reactor cores and do not signify any kind of core loading pattern. This design is much different than previous BSR designs due to the fact that an effort is made to incorporate as much existing BWR technology (1) to provide an ability to retrofit an existing design for superheating and (2) to allow an evolutionary path to

34 22 future BSR technology by basing it in existing technology. As such, Figure 2-3 shows a reactor vessel with ¾ of the reactor vessel identical to that of an ABWR, BWR-6, or BWR-5. The ability to build on existing technology, in addition to obtaining higher efficiencies and power densities, can make BSR technology much more cost effective. With the new design outlined, a new comparison can be made concerning the internal structure of the vessel and the added component complexity. For the entire lower core up to the steam separators, there is no major change from existing BWR technology. Changes would start with the steam dryers by removing the central region and orienting the dryer heads into concentric rings around the superheat core. There would be a lower plenum for the steam after the steam dryers with a lower separation plate coupled with the base of the steam dryers and a lower core plate for placement of the superheater assemblies. There would be the core in the center with an upper core plate to separate the saturated steam from the superheated steam. The lower core plate, shroud, and upper could be integrated to allow the entire superheater reactor core be removed from the vessel. This would allow the boiler and superheater cores to be reloaded at the same time. No additional steam flow tubing is necessary beyond what is already present. The greatest change needed would be the extra height required to accommodate the superheater core. If retrofitting an existing reactor, a new vessel head made longer with extra steam outlets to replace the existing vessel outlets would be needed. For new construction, a longer vessel could be made such as the vessel required for the General Electric Economic Simplified Boiling Water Reactor (ESBWR).

35 23 Figure 2-3: Conceptual BSR vessel design Last, the vessel head would need to incorporate control blades in order to provide criticality control of the superheater core. It is advantageous that the blades insert from the top because the reactor is hottest at the top and is the most susceptible to damage in

36 an accident scenario. The neutronics situation for the superheater core is investigated in Chapter Figure 2-4: Cross section of superheater core with circle for inner vessel rim With the design set, the geometric information is displayed in Table 2-4. As can be seen, the superheater core is taller but thinner than the boiler core. It has slightly over 46% more fuel content. Because the power density of the superheater core is lower due to

37 25 inferior heat transfer and lower fuel temperature limits, the amount of uranium per unit thermal power is higher by a factor of 4.6. This means that for the same initial k eff for both cores without burnable absorbers, the superheater core would last 4.6 times longer. It is not guaranteed that the superheater core would have the same k eff as the boiler core, but data on the criticality for the superheater elements is given in Chapter 5. Figure 2-5: Cross section of boiler core with circle for inner vessel rim

38 26 Table 2-4: Geometric and fuel quantity comparison of superheater and boiler cores Core Diameter (m) Heated Length (m) Assemblies Total Fuel Material (kmol) Fuel per Power (mol/mwt) Superheater Core Boiler Core With the new core design data outlined, it is possible to make the preliminary power calculations with data from Chapter 4. Equation 2.1 is used and the comparison data is displayed in Table 2-5. From the data, it is clear that one of the major benefits of the dual core BSR lies in the substantial power increase. The 89 MWe increase from the ABWR amounts to placing a modern American 2-loop Babcock & Wilcox PWR into the ABWR vessel. More detailed thermal-hydraulic information is given in Chapter 4. Table 2-5: Power output comparison between the ABWR, the SCWR, and the dual core BSR design q L H η N (kw/m) (m) r N cycle P e ass (%) (MWe) ABWR SCWR BSR: Dual Core S: 5.59 S: 4.27 S: 91 S: 577 B: 19.6 B: 3.71 B: 62 B: Note S: superheater assembly, B: boiler assembly There are many advantages of this new design compared to the SCWR. The BSR operates at a much lower pressure of 72.4 bar in the middle of the boiler core. The SCWR operates over three times this pressure, at 25 bar. The materials that are required to operate at the temperature and pressure of the SCWR require further testing. However, the materials that are required to operate at the temperatures and pressures of the BSR

39 have been tested in previous designs and fall in the range of current SCWR materials 27 testing. Also, the factor of three increase in pressure requires vessel and piping to be roughly 3 times thicker, increasing the cost of the vessel and piping. The BSR performs competitively in terms of power density as well. The SCWR operates under single phase heat transfer conditions. No major boiling phenomena occur to enhance heat transfer. Because of this, the SCWR must compensate with higher flow velocities to obtain competitive heat transfer rates. Combining this with the high enthalpy rise and phase change to supercritical makes the SCWR very problematic under accident conditions [11]. Figures 2-6 and 2-7 show some of the relevant phase change information that would affect the stability of the SCWR. Of most relevance in Figure 2-6 is the heat capacity. The spike in the heat capacity from the phase transition to a supercritical fluid can have detrimental consequences to the assembly. As the axial power profile changes, this spike can move around, thermally shocking an area that had previously operated with great heat transfer capabilities as the spike moves away, providing much poorer heat transfer afterwards. The phase change in the SCWR leads to a drastic change in the coolant thermophysical properties. For a reactor that undergoes a changing axial power profile, this means that this phase change can move axially in the core. When trying to model accident scenarios, the location of these drastic coolant property changes can make predictions and safety strategies significantly more complex. The advantage of the dualcore BSR is that it splits the enthalpy rise across two cores with different reactor control structures.

40 28 Figure 2-6: Density and heat capacity as a function of temperature at 25 bar [11] Figure 2-7: Thermal conductivity as a function of temperature at 25 bar [11]

41 29 Another advantage lies in the neutron economy. To operate at the temperatures and pressures required for the SCWR, the structural material used is largely steel. When entering or leaving the fuel, neutrons must always pass through the steel cladding. Because the thermal absorption cross section of iron is over 14 times larger than zirconium, using steel as a cladding material negatively impacts the neutron economy of the reactor. For the boiler core in the BSR, the use of steel is largely avoided. For the superheater core, the steel cladding is not placed between the primary moderator and fuel, avoiding much of the neutron losses associated with the SCWR steel cladding. Overall, the dual core BSR design appears to be competitive with the SCWR. Although more structurally complex than the SCWR and more expensive than existing BWR technology, the dual core BSR design is largely based on existing technology and tested materials. Last, this design provides a substantial electrical power increase over all existing and proposed reactor technology. It is worthy of further investigation.

42 Chapter 3 Materials Investigation As defined in Chapter 2, most of the materials used in the fuel assemblies are not standard in commercial LWR technology. Because of the many unknowns concerning the interactions between these materials, an investigation must be performed in order to identify possible problems. First, basic thermochemical information is used to review various chemical reactions between the materials. Next, the materials are investigated specifically for the advantages and disadvantages they offer. This includes interaction dynamics of these materials under accident scenarios. 3.1 Thermochemical Interaction Analysis For chemical compounds there are energies of formation of these compounds from their constituent elements. This energy of formation is highly temperature dependent. With the basic thermochemical data, it is possible to estimate the energy changes due to the chemical reactions that might occur. These estimations have two purposes. First, the calculations indicate whether the reaction is physically possible. For the reactions that are possible, the calculations give estimates for how much free energy is released. The relevant information for these calculations is given in Appendix A [12]. Using this information, it is possible to estimate the nature of possible reactions according to Eq. 3.1.