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In this study, the structural integrity of liquid metal fast reactor fuel assembly has been established for different parameters considering the optimum fuel design. Analytical calculation of added mass effect due to lead bismuth eutectic and verification through previously presented theories, has been established. The integrity of the hexagonal wrapper of fuel assembly has been guaranteed over the entire operating temperature range. Effect of temperature on the density of lead bismuth eutectic, the subsequent change in added mass of lead bismuth eutectic, the effect on natural frequencies and effect on stresses on wrapper, has been studied in detail. A simple empirical relationship is presented for estimation of added mass effect for lead bismuth eutectic type fast reactors for any desired temperature. An approach for assessment of fast reactor fuel assembly performance has been outlined and calculated results are presented. Nuclear seismic rules require that systems and components which are important to safety, shall be capable of bearing earthquake effects and their integrity and functionality should be guaranteed. Mode shapes, natural frequencies, stresses on wrapper and seismic aspect has also been considered using ANSYS. Modal analysis has been compared in vacuum and lead bismuth eutectic using the calculated added mass.

The corrosion of Liquid lead or lead-bismuth eutectic (LBE) on the structural steels is a serious problem for their application in fast reactors or accelerator-driven transmutation systems. This paper focuses on the modeling and experimental study on LBE corrosion in the spent fuel transmutation systems, as well as the effects of corrosion products on heat transfer and flow. By studying the corrosion and deposit of the oxidation products on the pipe wall, the empirical correlations of thermal resistance and fouling factor can be theoretically developed and experimentally verified. The comparison of fouling factors obtained with each oxygen level is presented, the relationship between fouling factors and oxygen concentrations is correlated, and the effects of oxidation on heat transfer are demonstrated qualitatively by wetting conditions of the samples.

Radiation damage to structural material of fusion facilities is of high concern for safety. The superconducting tokamak EAST will conduct D-D plasma experiments with the neutron production of 1015 neutrons per second. To evaluate the material radiation damage a programme system has been devised with the Monte Carlo transport code MCNP-4C, the inventory code FISPACT99, a specific interface, and the fusion evaluated nuclear data library FENDL-2. The key nuclear responses, i.e. fast neutron flux, displacement per atom, and the helium and hydrogen production, are calculated for the structural material SS-316L of the first wall, and the vacuum vessel, using this programme. The results demonstrate that the radiation damage to the structural material is so little that it will not lead to any significant change of material properties according to the reference design. This indicates that there is a large potential space for EAST to test advanced operation regime from the viewpoint of structural material safety.

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis.

Preliminary analysis and calculation of liquid metal Li17Pb83 magnetohydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket. To decrease the liquid metal MHD pressure drop, Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts. The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed. Finally, the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.

The concept of the liquid Li17Pb83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR >1.05) and annual output of 100 kg or more fissile 239Pu (FBR >0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimizated calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (BR = TBR + FBR) is listed corresponding to different cases.

This paper proposes a conceptual structure of segmental water-cooled Center Conductor Post (CCP) to be flexible in installment and replacement. Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100 MW fusion power. Numerical simulation by using a commercial code PHOENICS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.

Neutronics optimization calculations have been performed for the tritium breeding blankets with solid ceramic breeder Li2O and liquid eutectic breeder Li17Pb83, respectively, based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport code MCNP/4B. The effects of beryllium, 6Li enrichment and various structural materials on Tritium Breeding Ratio have been systematically analyzed.

The self-cooling molten-salt Flibe blanket of FFHR is numerically analyzed, resulting the optimum first wall to be as thin as 5mm and the heat flux up to 0.25MW/m 2 to be feasible with adopting V-4Cr-4Ti as the structural material. An alternative concept of free surface using a capillary force is shown to be feasible even in helical systems, where a spiral flow is formed and drastically enhances the heat transfer efficiency. The nuclear property of Flibe blanket is modified with increasing Be amount and adopting carbon reflector, resulting the local TBR of 1.3. As an optional technique, 50% enrichment of Li-6 gives the maximum TBR of 1.4.

Liquid metal, mainly lead and lead-bismuth eutectic (LBE) has been widely applied in the advanced nuclear reactors and accelerator-driven transmutation (ADS) systems for a series of excellent physical and chemical properties, but LBE is very corrosive to the structural materials such as iron and stainless steel in medium and high temperature circumstance which limits its development seriously. Research works of anticorrosion for LBE has been carried out all over the world. In this paper, A kinetic model for corrosion and corrosion products precipitation in non-isothermal LBE circuits is developed to predict corrosion and precipitation behaviors, the mass transfer convection-diffusion equation is established and the corrosion/corrosion products precipitation rate along the circuit is predicted considering the effects of temperature profile of the entire circuit which are different from local parameters, the overall estimate of the corrosion/precipitation situation is obtained, at the end, the solid-phase oxygen control technique is discussed, a program developed for dissolution rate of the PbO particles in the mass exchanger is proposed.Copyright © 2013 by ASME

Fusion reactor is considered as one of the solutions for the sustaining development of nuclear energy. International Thermonuclear Experimental Reactor (ITER) is the biggest fusion reactor research plan in the world. High-intensity D-T fusion neutron generator can generate 14 MeV neutrons, and it matches the neutrons generated in ITER and be competently used for imitating the neutron environment in nuclear fusion reactor, which is important for the relevant experimental researches of blanket materials of fusion reactors. It can also be used for validating the correctness and reliability of the simulations and analyses in fusion basic studies, and can guide the subsequent material improvement and innovation of calculation methodology. A rotating tritium target system for D-T fusion neutron generator with a neutron yield of 1012 n·-1, i.e., a high intensity D-T fusion neutron generator, is proposed in this paper and the design, main parameters, technical difficulties and heat transfer enhancement method are introduced. The key and innovative technology of this rotating target system is the integration of the sprayed water cooling, mechanical seal and magnetic fluid seal technologies, which focuses on the heat transfer of the high heat power density in the target system. The most important technical index is that the maximum temperature on the target should not be above 200 ℃ as the tritium ions run away heavily from the tritium target when the target temperature is bigger than 200 ℃. To investigate the heat transfer characteristics of this rotating target system, the effects of water layer thickness, water flow rate and rotating speed on the heat transfer of this rotating target system are analyzed by computational fluid dynamics method. And the heat transfer processes of the target system under different heat power densities are also simulated and studied. The analysis results show that big water layer thickness, big water flow rate and high rotating speed are good for the heat transfer enhancement of the rotating target system, but the effects of the changes of the water layer thickness and water flow rate on the heat transfer process are both very small. Due to the design index, the heat power density on the target should be under a limit value, which is about 12 kW·cm-2 in the calculation results of this paper.

In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R&D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. c