• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1339-1348
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• 2018

The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2334-2337
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• 2022

Deuterium-tritium reaction is the most promising one in term of the highest nuclear fusion cross-section for the reactor. So it is one of urgent issues to develop materials and components that are simultaneously resistant to high heat flux and high energy neutron flux in realization of the fusion energy. 2.45 MeV neutron production was reported in D-D reaction in KSTAR and regarded as beam-target is the dominant process. The feasibility study of KSTAR to wide area neutron source facility is done in term of D-D and D-T reactions from the empirical scaling law from the mixed fast and thermal stored energy and its projection to cases of heating power upgrade and DT reaction is done.

• 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.50 no.1
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• 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.

• Journal of Welding and Joining
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• v.16 no.3
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• pp.121-128
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• 1998

The degradation of mechanical properties in the high nitrogen steel HN3 developed for nuclear fusion reactor has been evaluated quantitatively using the small punch(SP) test, X-ray diffraction (XRD) analysis has also been conducted to identify carbides or nitrides precipitated on grain boundaries of the heat treated samples. Mechanical properties of the steel HN3 significantly decreased with increasing heat treatment time and temperature or with decreasing testing temperature. Combination of XRD and metallurgical observation, revealed that the material degradation in the thermally aged steel was caused by precipitation of carbides on the grain boundaries. While the weld metal showed the lowest mechanical properties among various microstructures in GTA weldments. By combining SP test and XRD analysis, cryogenic fracture behaviors and aging degradation for high nitrogen steel could be successfully evaluated in nondestructive manner.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2678-2686
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• 2023

This work demonstrates the results of full-scale numerical experiments of a hybrid thorium-containing fuel plant operating in a state close to critical due to a controlled source of D-T neutrons. The proposed facility represented a level of generated power (~10-100 MW_t) in a small pilot. In this work, the simulation of the D-T neutron plasma source operation in conjunction with the facility blanket was performed. The fission of fuel nuclei and the formation of spatial-energy release were studied in this simulation, in pulsed and stationary modes of the facility operation. The optimization results of neutronic and fluid dynamics studies to level the emerging offsets of the radial energy formed in the volume of the facility multiplying part due to the pulsed operation of the D-T neutron plasma source were presented. The results will be useful in improving the power control-based subcriticality monitoring method in coupled systems of the "pulsed neutron source-subcritical fuel assembly" type.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1002-1007
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• 2020

The purpose of the current study was to evaluate a spectrum formulation set employed to modify the neutron spectrum of D-D fusion neutrons in a IS plasma focus device using GEANT4, MCNPX2.6, and FLUKA codes. The set consists of a moderator, reflector, collimator and filters of fast neutron and gamma radiation, which placed on the path of 2.45 MeV neutron energy. The treated neutrons eliminate cancerous tissue with minimal damage to other healthy tissue in a method called neutron therapy. The system optimized for a total neutron yield of 10⁹ (n/s). The numerical results indicate that the GEANT4 code for the cubic geometry in the Beam Shaping Assembly 3 (BSA3) is the best choice for the energy of epithermal neutrons.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.87-92
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• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.