• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.323-327
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• 2020

In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% ⁶Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687℃. As regards the tolerable temperature of beryllium (650℃), the design of blanket module is done in the way that beryllium temperature never exceed 600℃. The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4287-4294
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• 2023

In order to meet the energy demand, energy production must be done continuously. Hydrogen seems to be the best alternative for this energy production, because it is both an environmentally friendly and renewable energy source. In this study, the hydrogen fuel production of the peaceful nuclear explosives (PACER) fusion blanket as the energy source integrated with Fe-Cl thermochemical water splitting cycle have been investigated. Firstly, neutronic analyzes of the PACER fusion blanket were performed. Necessary neutronic studies were performed in the Monte Carlo calculation method. Molten salt fuel has been considered mole-fractions of heavy metal salt (ThF₄, UF₄ and ThF₄+UF₄) by 2, 6 and 12 mol. % with Flibe as the main constituent. Secondly, potential of the hydrogen fuel production as a result of the neutronic evaluations of the PACER fusion blanket integrated with Fe-Cl thermochemical cycle have been performed. In these calculations, tritium breeding (TBR), energy multiplication factor (M), thermal power ratio (1 - 𝜓), total thermal power (P_hpf) and mass flow rate of hydrogen (ṁ_H2) have been computed. As a results, the amount of the hydrogen production (ṁ_H2) have been obtained in the range of 232.24x10⁶ kg/year and 345.79 x10⁶ kg/year for the all mole-fractions of heavy metal salts using in the blanket.

• Nuclear Engineering and Technology
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• v.35 no.6
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• pp.515-522
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• 2003

Tritium behavior in the solid breeder blanket is one of the key factors in determining tritium self-sufficiency, as well as safety, of fusion reactors. Recently, a model has been developed to describe the tritium behavior in solid breeder material, which can predict the tritium release and inventory in the blanket. However, the model has limitation to account for tritium solubility effects, mainly existing as LiOT, especially inside the $Li_{2}O$ solid breeder. In order to improve the capability of predicting the LiOT precipitation in $Li_{2}O$ solid breeder, a new logic is developed and integrated in the existing tritium release and inventory calculation code. With the logic developed in this work, the code can have capabilities to analyze tritium release and inventories in $Li_{2}O$ under steady and transient conditions. It can be found that tritium inventory as LiOT is an important mechanism under pulsed operation, and the amount of inventory becomes higher as the tritium generation rate increases and the temperature decreases. Also, the temperature limits for the generation of LiOT precipitation are determined. Therefore the developed logic helps understand the tritium transport mechanism in $Li_{2}O$ solid breeder.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.423-432
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• 2005

Japanese activities on fusion structural materials R&D have been well organized under the coordination of university programs and JAERI/NIMS programs more than two decades. Where, two categories of structural materials have been studied, those are; reduced activation martensitic/ferritic steels (RAFs) as reference material and vanadium alloys and SiC/SiC composite materials as advanced materials. The R&D histories of these candidate materials and the present status in Japan are reviewed with the emphasis on materials behavior under radiation damage. The importance of IFMIF and technology development for blanket R&D including ITER-TBRG activity is emphasized and the current status of those activities in Japan is also presented.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.582-582
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• 2012

Characteristics of a fusion-driven transmutation reactor was investigated. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4274-4281
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• 2023

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1060-1067
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• 2018

A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial $k_{eff}$ level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially $^{241}Am$ is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1736-1746
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• 2021

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion-fission reactor.

• Proceedings of the KWS Conference
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• 2008.05a
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• pp.13-13
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• 2008