• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.290.1-290
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• 2002

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.345-346
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• 2018

• 한국재료학회:학술대회논문집
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• 한국재료학회 2011년도 춘계학술발표대회
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• pp.15-15
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• 2011

As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.199-214
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• 2009

High burn-up transformation process in low temperature nuclear fuel oxides material was observed in the early sixties in LWR $UO_2$ fuels, but not studied in depth. Increasing progressively the fuel discharge burn-up in PWR power plants, this material transformation was again observed in 1985 and identified as an important process to be accounted for in the fuel simulations due to its expected consequence on fuel heat transfer and therefore on the fission gas release. Fission gas release was one of the major concerns in PWR fuels, mainly during transient or accidents events. The behaviour of such a material in case of rod failure was also an important aspect to analyse. Therefore several national and international programs were launched during the last 25 years to understand the mechanisms leading to the high burn-up structure formation and to evaluate the physical properties of the final material. A large observations database has been acquired, using the more sophisticated techniques available in hot cells. This large database is discussed in this paper, providing basis to build an engineering-model, which is based on phenomenological description data and information accumulated. In addition this paper has the ambition to construct the best logical model to understand restructuring.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.863-868
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• 2020

Co Self-Powered Neutron Detector (SPND) is confronted with the problem of material consumption, which causes the response current can neither reflect the change of neutron flux in time nor be proportional to the neutron flux. In this paper, a deconvolution-based method is established to solve this problem. First of all, a step signal of neutron flux is taken as an example to analyze its performance. When the material consumption of Co SPND is 10%, after compensation, the response current can be in correspondence of neutron flux. Finally, the effects of this model in different Signal-to-Noise Ratio are analyzed, which fully confirms the truth of its excellent performance for compensating Co SPND's signal.

• 한국세라믹학회지
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• 제54권1호
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• pp.9-22
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• 2017

Many new functional materials have been studied for efficient production and storage of energy. Many new materials such as sodium-based and sulfide-based materials have been proposed for energy storage, but research on Li batteries is still dominant. Due to the influence of environmental concerns regarding nuclear energy, interest in and research on fusion power are steadily increasing. For the commercialization of nuclear fusion, a design standard based on a considerable level of physical analysis and modeling is proposed. Nevertheless, limitations of existing materials in nuclear fusion environments limit practical applications. Tritium propagation material for continuous fusion reaction is one of the core materials, and therefore research on this material is being carried out intermittently. The key material for Li-based energy storage and tritium generation is the functional material Li-M-O. In this review, a structural description of functional Li-M-O system materials and technical trends for its applications are introduced.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1537-1544
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• 2020

Molecular dynamics (MD) simulations were conducted to investigate the temperature effects on the primary damage in gallium nitride (GaN) material. Five temperatures ranging from 300 K to 900 K were studied for 10 keV Ga primary knock-on atom (PKA) with inject direction of [0001]. The results of MD simulations showed that threshold displacement energy (E_d) was affected by temperatures and at higher temperature, it was larger. The evolutions of defects under various temperatures were similar. However, the higher temperature was found to increase the peak number, peak time, final time and recombination efficiency while decreasing the final number. With regard to clusters, isolated point defects and little clusters were common clusters and the fraction of point defects increased with temperature for vacancy clusters, whereas it did not appear in the interstitial clusters. Finally, at each temperature, the number of Ga interstitial atoms was larger than that of N and besides that, there were other different results of specific types of split interstitial atoms.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.2026-2033
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• 2019

Fluidelastic instability of a tube array is a key factor of the security of a nuclear power plant. An unsteady model of the fluidelastic instability of a tube array subjected to two-phase flow was developed to analyze the fluidelastic instability of tube bundles in two-phase flow. Based on this model, a computational program was written to calculate the eigenvalue and the critical velocity of the fluidelastic instability. The unsteady model and the program were verified by comparing with the experimental results reported previously. The influences of void fraction and the tube's material properties on the critical velocity were investigated. Numerical results showed that, with increasing the void fraction of the two-phase flow, the tube array becomes more stable. The results indicate that the critical velocities of the tube array made of stainless are much higher than those of the other two tube arrays within void fraction ranging from 20% to 80%.

• 방사성폐기물학회지
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• 제20권2호
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• pp.171-183
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• 2022

The parent and daughter nuclides in a radioactive decay chain arrive at secular equilibrium once they have a large half-life difference. The characteristics of this equilibrium state can be used to estimate the production time of nuclear materials. In this study, a mathematical model and algorithm that can be applied to radio-chronometry using the radioactive equilibrium relationship were investigated, reviewed, and implemented. A Bateman equation that can analyze the decay of radioactive materials over time was used for the mathematical model. To obtain a differential-based solution of the Bateman equation, an algebraic numerical solution approach and two different matrix exponential functions (Moral and Levy) were implemented. The obtained result was compared with those of commonly used algorithms, such as the Chebyshev rational approximation method and WISE Uranium. The experimental analysis confirmed the similarity of the results. However, the Moral method led to an increasing calculation uncertainty once there was a branching decay, so this aspect must be improved. The time period corresponding to the production of nuclear materials or nuclear activity can be estimated using the proposed algorithm when uranium or its daughter nuclides are included in the target materials for nuclear forensics.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.811-820
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• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.