APEX ADVANCED FERRITIC STEEL, FLIBE SELF-COOLED FIRST WALL AND BLANKET DESIGN (original) (raw)
Related papers
Molten salt self-cooled solid first wall and blanket design based on advanced ferritic steel
Fusion Engineering and Design, 2004
As an element in the U.S. Advanced Power Extraction (APEX) program, the solid first wall and blanket design team assessed innovative design configurations with the use of advanced nano-composite ferritic steel (AFS) as the structural material and FLiBe as the tritium breeder and coolant. The goal for the assessment is to search for designs that can have high volumetric power density and surface heat-flux handling capability, with assurance of fuel self-sufficiency, high thermal efficiency and passive safety for a tokamak power reactor. We selected the re-circulating flow configuration as our reference design. Based on the recommended material properties of AFS we found that the reference design can handle a maximum surface heat flux of 1 MW/m 2 , and a maximum neutron wall loading of 5.4 MW/m 2 , with a gross thermal efficiency of 47%, while meeting all the tritium breeding, structural design and passive safety requirements. This paper will cover the results of the following areas of assessment: material design properties, FW/blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal-hydraulics analysis including MHD effects, structural analysis; molten salt and helium closed cycle power conversion system; and safety and waste disposal of the re-circulating coolant first wall and blanket design.
High strength and heat resistant chromium steels for sodium-cooled fast reactors
2004
This report provides the results of a preliminary phase of a project supporting the Advanced Nuclear Fuel Cycle Technology Initiative at ANL. The project targets the Generation IV nuclear energy systems, particularly the area of reducing the cost of sodium-cooled fast-reactors by utilizing innovative materials. The main goal of the project is to provide the nuclear heat exchanger designers a simplified means to quantify the cost advantages of the recently developed high strength and heat resistant ferritic steels with 9 to 13% chromium content. The emphasis in the preliminary phase is on two steels that show distinctive advantages and have been proposed as candidate materials for heat exchangers and also for reactor vessels and near-core components of Gen IV reactors. These steels are the 12Cr-2W (HCM12A) and 9Cr-1MoVNb (modified 9Cr-1Mo). When these steels are in tube form, they are referred to in ASTM Standards as T122 and T91, respectively. A simple thermal-hydraulics analytical model of a counter-flow, shell-and-tube, oncethrough type superheated steam generator is developed to determine the required tube length and tube wall temperature profile. The single-tube model calculations are then extended to cover the following design criteria: (i) ratio of the tube stress due to water/steam pressure to the ASME B&PV Code allowable stress, (ii) ratio of the strain due to through-tube-wall temperature differences to the material fatigue limit, (iii) overall differential thermal expansion between the tube and shell, and (iv) total amount of tube material required for the specified heat exchanger thermal power. Calculations were done for a 292 MW steam generator design with 2200 tubes and a steam exit condition of 457°C and 16 MPa. The calculations were performed with the tubes made of the two advanced ferritic steels, 12Cr-2W and 9Cr-1MoVNb, and of the most commonly used steel, 2¼Cr-1Mo. Compared to the 2¼Cr-1Mo results, the 12Cr-2W tubes required 29% less material and the 9Cr-1MoVNb tubes required 25% less material. Also, with the advanced steels, the thermal strains in the tubes and differential thermal expansion between tubes and shell were significantly better. For steam generators with higher steam exit temperatures, the benefits of the advanced steels become much larger. A thorough search for the thermal and mechanical properties of the two advanced steels was conducted. A summary of the search results is provided. It shows what is presently known about these two advanced steels and what still needs to be determined so that they can be used in nuclear heat exchanger designs. Possible follow up steps are outlined.
Safety assessment of two advanced ferritic steel molten salt blanket design concepts
Fusion Engineering and Design, 2004
In this article, we explore some of the safety issues associated with two advanced ferritic steel (AFS) molten salt blanket designs from the Advanced Power Extraction (APEX) design study [M.A. Abdou, The APEX Team, On the exploration of innovative concepts for fusion chamber technology, Fus. Eng. Des. 54 ]. In particular, we examine radiological inventories, decay heat, waste disposal ratings, and toxic chemical inventories of these design concepts. In addition, we predict the thermal response of these blanket designs during accident conditions, and the mobilization of the radiological inventories and site boundary dose from the release of this mobilized material during a worst-case confinement-boundary-bypass accident. The molten salts being proposed for these blanket concepts are Flibe and Flinabe, and the structural material is a nano-composite strengthened ferritic steel alloy called 12YWT. The estimated dose at the site boundary is less than the no-evacuation limit of 10 mSv for a ground level release during conservative weather conditions if plant isolation occurs within 5 days.
Fusion Engineering and Design, 1993
The TITAN-II reactor is a compact, high-neutron-wall-loading (18 MW/m 2) design. The TITAN-II fusion power core (FPC) is cooled by an aqueous lithium-salt solution that also acts as the breeder material. The use of an aqueous solution imposes special constraints on the selection of structural and breeder material because of corrosion concerns, hydrogen embrittlement, and radiolytic effects. In this paper, the materials engineering and design considerations for the TITAN-II FPC are presented. Material selection criteria, based on electrochemical corrosion mechanisms of aqueous solutions coupled with radiolysis of water by ionizing radiation, resulted in the choice of a low-activation ferritic steel as structural material for TITAN-II. Stress corrosion cracking, hydrogen embrittlement, and changes in the ductile-to-brittle transition temperature of ferritic alloys are discussed. Lithium-nitrate (LiNO 3) salt was chosen over lithium hydroxide (LiOH) because it is less corrosive and reduces the net radiolytic decomposition rate of the water. The dissolved salt in the coolant changes the thermophysical properties of the coolant results in trade-offs between the lithium concentration in the coolant, neutronics performance, thermal and structural design. The TITAN-II design requires a neutron multiplier to achieve an adequate tritium breeding ratio. Beryllium is the primary neutron multiplier, assuming a maximum swelling of about 10% based on continuous self-limiting microcracking/sintering cycles.
Ferritic/martensitic steels for next-generation reactors
Journal of Nuclear Materials, 2007
Concepts for the next generation of nuclear power reactors designed to meet increasing worldwide demand for energy include water-cooled, gas-cooled, and liquid-metal-cooled reactors. Reactor conditions for several designs offer challenges for engineers and designers concerning which structural and cladding materials to use. Depending on operating conditions, some of the designs favor the use of elevated-temperature ferritic/martensitic steels for in-core and out-of core applications. This class of commercial steels has been investigated in previous work on international fast reactor and fusion reactor research programs. More recently, international fusion reactor research programs have developed and tested elevated-temperature reduced-activation steels. Steels from these fission and fusion programs will provide reference materials for future fission applications. In addition, new elevated-temperature steels have been developed in recent years for conventional power systems that also need to be considered for the next generation of nuclear reactors.
Selection Criteria for Fusion Reactor Structures
Journal of Thermal Engineering
Fusion energy is the ultimate energy to cover Mankind's energy needs forever. However, taming the fusion energy is the greatest technological challenge the humanity is facing. Development of structural materials to withstand against the extreme conditions in the course of fusion power plant operation is one of the toughest nuts to be cracked. A great number of structural materials have been investigated for fusion reactor applications, such as steels (austenitic stainless steels and ferritic/martensitic steels), vanadium alloys, refractory metals and alloys (niobium alloys, tantalum alloys, chromium and chromium alloys, molybdenum alloys, tungsten and tungsten alloys), and composites (SiCf/SiC and Carbon Fibre Composite CFC composites). Steels have extensive technological data base and significantly lower cost compared to other refractory metals and alloys. Ferritic steels and modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. However, steels cannot withstand high neutron wall loads to build an economically competitive fusion reactor. Some refractory metals and alloys (niobium alloys, tantalum alloys, molybdenum alloys, tungsten and tungsten alloys) can withstand high neutron wall loads. But, in addition to their very limited technological data base, they have high residual radioactivity and prohibitively high production costs. A protective, flowing liquid zone to protect the first wall of a fusion reactor from direct exposure to the fusion reaction products could extend the lifetime of the first wall to the expected lifetime of the fusion reactor. In that context, a fusion-fission (hybrid) with a multi-layered spherical blanket has been investigated, which is composed of a first wall made of oxide dispersed steel (ODS, 2 cm); neutron multiplier and coolant zone made of LiPb; ODS-separator (2 cm); a molten salt FLIBE coolant and fission zone; ODS-separator (2 cm); graphite reflector. Calculations are conducted for a liquid wall with variable thickness, containing Flibe + heavy metal salt (UF4 or ThF4) is used for first wall protection. The content of heavy metal salt is chosen as 4 and 12 mol%. A flowing wall with a thickness of ~ 60 cm can extend the lifetime of the solid first wall structure to a plant lifetime of 30 years for 9Cr-2WVTa and V-4Cr-4Ti, whereas the SiCf/SiC composite as first wall needs a flowing wall with a thickness of ~ 85 cm to maintain the radiation damage limit.
Neutronics Assessment of Molten Salt Breeding Blanket Design Options
Fusion Science and Technology, 2005
Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be~1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (~3 kg) over the blanket lifetime of~3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied.
Fusion Engineering and Design, 2000
Among the concepts considered in the Advanced Power Extraction (APEX) study is the Hecooled refractory metal FW and blanket concept. Refractory metals exhibit high operating temperature and can offer good capability for withstanding high power density operation that is the focus of the APEX study. In this paper, we assess the impact of using various refractory metal on the nuclear heating profiles across the blanket and power multiplication, PM, and on the tritium breeding profiles and tritium breeding ratio, TBR. The refractory metals considered with liquid lithium breeder are W, TZM, and Nb-1Zr. The impact of Li-6 enrichment on these profiles and on TBR and PM is also assessed. Comparison of these nuclear characteristics is also made to other liquid breeder (Flibe and Li-Sn). Because the moderation power of these breeders to neutron energy varies among them, the damage to the structure is different with various structure/breeder combinations. The damage parameters (DPA, helium and hydrogen production)
Fusion Engineering and Design, 2006
An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiC f /SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 • C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.