A ceramic breeder in a poloidal tube blanket for a tokamak reactor (original) (raw)
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Fusion Engineering and Design, 2021
Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16 N and 17 N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase.
Evaluation of the parfait blanket concept for fast breeder reactors
1974
An evaluation of the neutronic, thermal-hydraulic, mechanical and economic characteristics of fast breeder reactor configurations containing an internal blanket has been performed. This design, called the parfait blanket concept, employs a layer of axial blanket fuel pellets at the core midplane in the fuel pins of the inner enrichment zone; otherwise, the design is the same as that of the conventional LMFBR's to which the parfait configuration was compared. Two significant advantages were identified for the parfait blanket concept relative to the conventional design. First, the parfait configuration has a 25% smaller peak fast flux which reduces wrapper tube dilation by 37% and fuel element elongation by 29%; and second, axial and radial flux flattening contribute to a 7. 6% reduction in the peak fuel burnup. Both characteristics significantly diminish the problems of fuel and metal swelling. Other advantages identified for a typical parfait design include: a 25% reduction in the burnup reactivity swing, which reduces control rod requirements; a 7% greater overpower operating margin; an increased breeding ratio, which offsets the disadvantage of a higher critical mass; and more favorable sodium voiding characteristics which counteract the disadvantage of an 8% smaller power Doppler coefficient. All other characteristics investigated were found to differ insignificantly or slightly favor the parfait design.
An Assessment Of Critical Thermal-Hydraulic Problems in A Deuterium-Tritium Solid Breeder Blanket
Nuclear Technology - Fusion, 1983
Steady-state thermal-hydraulic analyses were carried out for the DEMO/STARFIRE fusion reactor based on solid breeder blankets and pressurized water as the coolant. The results of the parametric studies show that a coolant in-tube design, i.e., coolant tubes embedded in solid breeder blanket, with a contact resistance between the coolant tube and the solid breeder tailored to maintain the operating temperature window (i.e., the maximum and the minimum temperature imposed on the solid breeder) is viable. However, design of such a solid breeder blanket will present serious challenges because of uncertainty in the thermophysical properties of breeder materials, the narrow operating temperature window, the close manufacturing tolerances necessary to control the gap conductance, the sensitivity of tritium inventory and tritium extraction to breeder temperature distribution, and the deleterious effect of neutron irradiation on breeder material properties. The study shows that even modest uncertainties in the thermal conductivity of solid breeders, interfacial gap conductances, and operating power levels can have significant impact on blanket design. Therefore, the designer should include the expected variations in these parameters. Experimental programs are needed to quantify the above factors and to develop methods (e.g., insulated coatings) for gap conductance control and in situ recovery of tritium via helium purge gas channels.
Thermal hydraulic analysis of the basic breeder module of the European ceramic BIT DEMO blanket
Fusion Engineering and Design, 1991
The European ceramic breeder-inside-tube blanket design proposed for the DEMO plant is featured by helium-cooled poloidal breeder modules consisting of a shrouded bundle of breeder tubes contained in a pressure tube and surrounded by beryllium blocks. This kind of breeder modules raises some problems as far as the control of pressure losses and hot spots is concerned, problems which might, a priori, jeopardize the feasibility and performance of the blanket. Results of first thermal-hydraulic analyses, performed by means of heat transport and subchannel codes, prove however to be quite encouraging in this respect. The elimination of remaining uncertainties-not crucial for the viability of the proposed module design-will nevertheless require specific experiments.
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.
Overview of design and thermal–hydraulic analysis of Indian solid breeder blanket concept
India has developed two concepts of breeding blanket for the DEMO reactor: one is Lead Lithium Ceramic Breeder (LLCB), and the other one is Helium-cooled Ceramic Breeder (HCCB) concept. Indian HCCB concept is having edge on configuration of helium-cooled solid breeder with RAFMS structure. Li 2 TiO 3 /Li 4 SiO 4 and beryllium are used as the tritium breeder and neutron multiplier, respectively. 2D thermal-hydraulic simulation studies using ANSYS have been performed based on the heat load obtained from neutronics calculations to confirm heat removal under ITER pulsed operation. Transient thermal analysis has been simulated in ANSYS for the ITER relevant operational conditions. Thermal analysis provides important information about the temperature distribution in different materials used and their temperature-time histories. Result of thermal-hydraulic simulations shows that in each cycle, the maximum temperature of all materials remains same. The peak temperatures of all materials are well within their limiting value. Concept designs of HCCB blanket and its thermal hydraulic analysis will be presented in this paper.
A Parametric Study of Water-Cooled Solid Breeder Blanket Designs
Fusion Technology, 1986
An assessment of the role of the design and operating parameters on the performance of solid breeder blankets based on lithium ceramics (Li 2 0 and 'Y-LiA/02) was carried out. The results indicate that not only poor thermophysical properties but also uncertainties associated with the property data base are the design-limiting factors. In addition, the operating conditions such as the upper and the lower temperature limits, the choice of breeder materials either in the form of sintered pellets or in sphere-pac form, the interfacial contact resistance between the coolant channels and the solid breeder, and the diffusion characteristics of tritium and chemical interactions between tritium and the solid breeder play a prominent role in selection of blanket concepts. Designs to account for the expected degradation of the thermophysica/ properties due to thermal sintering and nuclear irradiation lead to high coolant and structural material fractions, and thus may result in a lower tritium breeding ratio. The results of the parametric studies show that watercooled solid breeder blanket designs require a firmer data base for the operating temperature limits, the thermophysical properties, the gap conductances, and the tritium retention and release characteristics of solid breeders.