Thermal–hydraulic and thermo-structural analysis of first wall for Indian DEMO blanket module (original) (raw)
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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.
Fusion Engineering and Design, 2008
Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) blanket with ferritic steel (FS) as a structural material in the International Thermonuclear Experimental Reactor (ITER) program. The design and the performance of the KO HCML test blanket module (TBM) and the preliminary results of the safety analyses such as activation, decay heat, and accident analysis by a loss of coolant are introduced briefly in this paper. KO HCML TBM uses He as a coolant and Li is used as a tritium breeder by considering its potential advantages. Two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. Performance analyses were performed with the MCCARD code for the neutronics, the CFX-10 code for the thermal-hydraulics, and with the ANSYS-10 code for the thermal-mechanical analysis. For the safety analyses, the activation and decay heat were obtained from the MCCARD and Origen codes. From the obtained decay heat, an accident analysis was performed.
Recent Progress in the WCLL Breeding Blanket Design for the DEMO Fusion Reactor
IEEE Transactions on Plasma Science, 2018
The water-cooled lithium-lead (PbLi) breeding blanket is one of the candidate systems considered for the implementation in the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This concept employs PbLi liquid metal as tritium breeder and neutron multiplier, water pressurized at 15.5 MPa as the coolant, and EUROFER as the structural material. The current design is based on the single module segment approach and follows the requirements of the DEMO-2015 baseline design. The module is constituted by a basic toroidal-radial cell that is recursively repeated along the poloidal direction where the liquid metal flows along a radial-poloidal path. The heat generated by the fusion reactions is extracted by means of separate cooling systems for the breeding zone and the first wall. A back supporting structure is dedicated to withstand loads of the module during normal and off-normal operations. Water and PbLi manifolds are integrated with primary heat transport and tritium extraction systems. The status of the conceptual design is presented, critically discussing its rationale and main features as supported by neutronic, thermal-hydraulic, magneto-hydrodynamic, and thermo-mechanic analyses. Recent results are outlined, pointing out open issues and development needs.
Basic concepts of DEMO and a design of a helium-cooled molten lithium blanket for a testing in ITER
Fusion Engineering and Design, 2007
Basic concepts and the performance of DEMO for an early realization have been investigated with a limited extension of its plasma physics and technology from the second phase of the International Thermonuclear Experimental Reactor (ITER) operation (EPP phase). With the same plasma size as that of ITER, net electric power up to 600 MW is possible with β N > 4.0, H > 1.0 and a divertor heat load of H div < 15 MW/m 2 . Through a consideration of the requirements for a DEMO-relevant blanket concept, Korea has proposed a He cooled molten lithium (HCML) blanket as an ITER TBM. It uses He as a coolant and Li is used as a tritium breeder by considering its potential advantages. Low activation Ferritic Steel (FS) is used as a structural material and two layers of graphite are inserted as a reflector in the breeder zone to increase the tritium breeding ratio (TBR) and the shielding performances. The design and the performance of the KO HCML test blanket module (TBM) are being modified in terms of its He coolant efficiency and its optimized path with a performance analysis; with a 3D Monte Carlo analysis (MCCARD code) for the neutronics; with the CFD code (CFX-10) for the thermal-hydraulics; with ANSYS-10 for the thermo-mechanical analysis.
Fusion Engineering and Design, 2010
The Helium-Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two TBM to be installed in an ITER equatorial port since day 1 of operation, with the specific aim to investigate the main concept functionalities and issues such as high efficiency helium cooling, resistance to thermo-mechanical stresses, manufacturing techniques, as well as tritium transport, magneto-hydrodynamics effects and corrosion. In particular, in order to show a DEMO-relevant thermo-mechanical and thermal-hydraulic behavior, the HCLL-TBM has to meet several requirements especially as far as its coolant thermofluid-dynamic conditions and its thermal-mechanical field are concerned. The present paper is focused on the assessment of the HCLL-TBM thermal-mechanical performances under both nominal and accidental load conditions, by adopting a computational approach based on the Finite Element Method. A realistic 3D finite element model of the whole HCLL-TBM, in the horizontal first wall design has been set up, consisting of about 597,000 elements and 767,000 nodes. In particular, since the thermal fields of both the module and the coolant are strictly coupled, the helium flow domain has been modeled too and a thermal contact model has been set up to properly simulate the convective heat transfer between the structure wall and the coolant. Pure conductive heat transfer has been assumed within the Pb-Li eutectic alloy of the breeder units. The volumetric density of the nuclear deposited power, recently calculated at Department of Nuclear Engineering of the University of Palermo by the MCNP 4C code, has been applied as distributed thermal load in order to assess the potential influence on the module thermo-mechanical performances of the markedly non-uniform poloidal and toroidal distributions that have been predicted within the Segment Box. Different loading scenarios have been considered as to the heat flux onto the module First Wall. Steady state and transient thermal-mechanical analyses have been carried out and the detailed spatial distributions of the thermal and mechanical fields obtained as to the considered loading scenarios are reported, together with a critical analysis intended to verify their compliance with the agreed design criteria and the DEMO relevance requirements.
Journal of Fusion Energy, 2018
An upgraded form of China fusion engineering test reactor (CFETR) was investigated for the safety performance. In the current study, modification of the designs were presented with relative tolerance. The steady state were calculated for the new design using Relap5 code. Two accidents were simulated i.e., in-vessel and In-box loss of coolant accident. These accidents were simulated in helium cooled ceramic blanket (HCCB) system for the purpose to investigate the safety measures of the CFETR. It is utmost important to ensure the safety performance of the reactor. In this research, sudden break at blanket system was assumed and calculated different parameters including temperature, pressure and coolant fluxes to observe the differences in pattern during the accident under limited time domain. The research is very important because the design of HCCB is new and there is a need to conduct steady state and transient state of the reactor in order to make sure and authenticate the design and to safer the reactor. Keywords Safety studies Á Modified design of HCCB Á Inbox LOCA Á In-vessel LOCA Á Parametric analysis Á Authentication of design and safety
Advanced high performance solid wall blanket concepts
Fusion Engineering and Design, 2002
First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D-T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability.
Review of blanket designs for advanced fusion reactors
Fusion Engineering and Design, 2008
The dominating fraction of the power generated by fusion in the reactor is captured by neutron moderation in the blanket surrounding the plasma. From this, the efficiency of the fusion plant is predominated by the technologies applied to make electricity or hydrogen from the neutrons. The main blanket concepts addressed in this paper are advanced ceramic breeder concepts, dual coolant blankets as well as self-cooled liquid metal and Flibe blankets. Two important questions that are addressed are: (i) Can we draw a bottom line conclusion on the most promising concept(s)? (ii) What are the common issues to be resolved independently from individual design and layout proposals to define a feasible route towards advanced fusion reactors? For ceramic breeder concepts, a key issue in the long term could be the limitation of beryllium as the considered multiplier in terms of world sources and achievable temperature levels. For liquid metal blankets, attractive long-term visions have been developed but major technological challenges also exist for the in-vessel blanket technology and the corresponding subsystems. The paper proposes a strategic conclusion derived from the review of blanket designs for advanced fusion reactors.
Blankets – key element of a fusion reactor – functions, design and present state of development
Kerntechnik, 2018
Blankets are key elements of a future fusion power reactor, as they breed the fusion fuel tritium, extract the heat from the reactor for power generation and contribute to the nuclear shielding of the plasma confining magnetic field coils. On the way to the engineering implementation of fusion, in particular the blanket design approach has changed substantially. Novel blanket designs require, already from the beginning, incorporating close coupling of plasma physics with engineering physics to develop robust solutions coping with thermal, mechanical and also electrodynamic loads – not only during the stationary operating phase, but also during transients. Simultaneously, nuclear licensing feasibility as well as component failure safety must be part of the design approach. This article describes advanced blanket design approaches undertaken in the past years by the example of the helium cooled pebble bed blanket (HCPB), aiming at an efficient blanket engineering design, starting from...