• Nuclear Engineering and Technology
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• 제46권5호
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• pp.675-680
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• 2014

Pulverization of two different sized micro-$B_4C$ particles (${\sim}10{\mu}m$ and ${\sim}150{\mu}m$) was investigated using a STS based high energy ball milling system. Shapes, generation of the impurities, and reduction of the particle size dependent on milling time and initial particle size were investigated using various analytic tools including SEM-EDX, XRD, and ICP-MS. Most of impurity was produced during the early stage of milling, and impurity content became independent on the milling time after the saturation. The degree of particle size reduction was also dependent on the initial $B_4C$ size. It was found that the STS nanoparticles produced from milling is strongly bounded with the $B_4C$ particles forming the $B_4C$/STS composite particles that can be used as a neutron absorbing nanocomposite. Based on the morphological evolution of the milled particles, a schematic pulverization model for the $B_4C$ particles was constructed.

• 한국세라믹학회지
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• 제54권2호
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• pp.121-127
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• 2017

Gadolinium tetraboride ($GdB_4$) was synthesized by reduction of $Gd_2O_3$ using boron carbide in presence of carbon. Effect of temperature on product quality was investigated. Pure $GdB_4$ powder was obtained in vacuum at $1500^{\circ}C$. Pressureless sintering experiments revealed that sintering takes place only above $1600^{\circ}C$. A maximum density of 77.1% of the theoretical value was obtained at $1800^{\circ}C$ by pressureless sintering. Hot pressing resulted in 95.5% of theoretical density at the lower temperature of $1700^{\circ}C$ under 35 MPa pressure. Hardness and fracture toughness of dense $GdB_4$ were measured and found to be 21.4 GPa and $2.3MPa{\cdot}m^{1/2}$, respectively. After exposure to air at $900^{\circ}C$, the formation of a porous and non-protective oxide layer was observed.

• 대한금속재료학회지
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• 제48권8호
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• pp.780-789
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• 2010

The aim of this study is to evaluate the microstructure and tensile property of in-situ (TiB+TiC) particulate reinforced titanium matrix composites (TMCs) synthesized by the investment casting process. Boron carbide ($1,500{\mu}m$ and $150{\mu}m$) was added to the titanium matrix during vacuum induction melting, which can provide the in-situ reaction of $5Ti+B_4C{\rightarrow}4TiB+TiC$. 0.94, 1.88 and 3.76 wt% of $B_4C$ were added to the melt. The phases identification of the in-situ synthesized TMCs was examined using scanning electron microscopy, an X-ray diffractometer, an electron probe micro-analyzer and transmission electron microscopy. Tensile properties of TMCs were investigated in accordance with the reinforcement size and volume fraction. The improvement of tensile property of titanium matrix composites was caused by load transfer from the titanium matrix to the reinforcement and by grain refinement of titanium matrix and reinforcements.

• 한국주조공학회지
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• 제37권4호
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• pp.115-122
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• 2017

The aim of this research is to evaluate the wear properties of (TiB+TiC) paticulate reinforced titanium matrix composites (TMCs) by in-situ synthesis. Different particle sizes (1500, $150{\mu}m$) and contents (0.94, 1.88 and 3.76 mass% for Ti, 1.98 and 3.96 mass% for the Ti6Al4V alloy) of boron carbide were added to pure titanium and to a Ti6Al4V alloy matrix during vacuum induction melting to provide 5, 10 and 20 vol.% (TiB+TiC) particulate reinforcement amounts. The wear behavior of the (TiB+TiC) particulate reinforced TMCs is described in detail with regard to the coefficient of friction, the hardness, and the degree of reinforcement fragmentation during sliding wear. The worn surfaces of each sliding wear condition are shown for the three types of wear studied here: transfer layer wear, particle cohesion wear and the development of abrasive areas. The fine reinforcements of TMCs were easily fragmented from the Ti matrix as compared to coarse reinforcements, and fragmented debris accelerated the decrease in the wear resistance.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1994년도 가을 학술발표회 논문집
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• pp.399-404
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• 1994

Heavyweight(or High density) concrete, which is generally for shiedling structures, differs from normal weight concrete by having a higher density and special compositions to improve its attenuation properties. There are setting 7 Beam Ports around the reactor of the KMRR Project(Korea Multi-purpose Research Reactor) conducted by the KAERI(Korea Atomic Energy Research Institute). High density(p=5.0t/$\textrm{m}^3$) and Heavyweight(p=3.5t/$\textrm{m}^3$) concrete were placed around the Beam Ports in order to shield radiation. This paper was discussed about construction of High density concrete. High density concrete was placed with method of Preplace Aggregate. Coarse metallic aggregate(steel shot) was used. Boron, boron carbide(B4C), was used to capture effctively the neutrom. The mock-up test was carried out. And the consturction of High density concrete was performed successfully.

• Composites Research
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• 제36권6호
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• pp.435-440
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• 2023

본 연구에서는 기존 사용후핵연료(Spent nuclear fuel) 운반/저장 용기에 사용되는 중성자 흡수용 B₄C/Al 복합소재의 열전도도를 개선하기 위해 탄화붕소(B₄C)와 입방정 질화붕소(cBN)를 동시에 강화재로 사용한 알루미늄(Al) 기지 복합소재를 제조하고 평가를 진행하였다. 이를 위해서 교반주조 공정을 통해 복합재 잉곳을 제조하고 이를 압연하여 중성자 흡수용 소재를 성공적으로 제조하였다. 제조된 소재의 평가를 위해 cBN 첨가에 따른 열전도도와 중성자 흡수능 변화를 관찰하였다. 열전도도 측정 결과, B₄C 단일 입자만을 사용한 복합소재 대비 B₄C, cBN을 함께 사용한 복합소재가 동일 체적률 조건 하에서 약 3%의 열전도도 증가가 발생하는 것을 확인하였으며 중성자 흡수 단면적 계산을 통해 중성자 흡수능이 무시할 수 있는 수준으로 저하가 발생하는 것을 확인하였다. 본 연구의 결과를 바탕으로, 중성자 흡수 소재의 새로운 설계 방안을 제시하고 고성능 운반/저장 용기의 개발에 기여할 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2387-2393
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• 2020

In this study, a gradient material for shielding neutrons and gamma rays was developed, which consists of epoxy resin, boron carbide (B₄C), lead (Pb) and a little graphene oxide. It aims light weight and compact, which will be applied on the transportable nuclear reactor. The material is made up of sixteen layers, and the thickness and components of each layer were designed by genetic algorithm (GA) combined with Monte Carlo N Particle Transport (MCNP). In the experiment, the viscosities of the epoxy at different temperatures were tested, and the settlement regularity of Pb particles and B₄C particles in the epoxy was simulated by matlab software. The material was manufactured at 25 ℃, the Pb C and O elements of which were also tested, and the result was compared with the outcome of the simulation. Finally, the material's shielding performance was simulated by MCNP and compared with the uniformity material's. The result shows that the shielding performance of gradient material is more effective than that of the uniformity material, and the difference is most noticeable when the materials are 30 cm thick.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2153-2162
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• 2024

The purpose of this study is to develop a high-energy neutron shielding material applied in proton therapy environment. Composite shielding material consisting of 10.00 wt% boron carbide particles (B₄C), 13.64 wt% surface-modified cross-linked polyethylene (PE), and 76.36 wt% tungsten particles were fabricated by hot-pressure sintering method, where the optimal ratio of the composite is determined by the shielding effect under the neutron field generated in typical proton therapy environment. The results of Differential Scanning Calorimetry measurements (DSC) and tensile experiment show that the composite has good thermal and mechanical properties. In addition, the high energy-neutron shielding performance of the developed material was evaluated using cyclotron proton accelerator with 100 MeV proton. The simulation shows a 99.99% decrease in fast neutron injection after 44 cm shielding, and the experiment result show a 99.70% decrease. Finally, the shielding effect of replacing part of the shielding material of the proton therapy hall with the developed material was simulated, and the results showed that the total neutron injection decreased to 0.99‰ and the neutron dose reduced to 1.10‰ before the enhanced shielding. In summary, the developed material is expected to serve as a shielding enhancement material in the proton therapy environment.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• 공업화학
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• 제7권3호
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• pp.597-604
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• 1996

사용후핵연료 수송용기 등에 사용되는 수지계 중성자 차폐재, KNS-115A, 115B 및 115C를 제조하였다. 기본물질은 에폭시수지이며, 첨가제로는 폴리프로필렌, 수산화알루미늄 및 탄화붕소이다. 이들 중성자 차폐재들은 유동성이 좋아 수송용기와 같은 복잡한 구조에 사용할 수 있다. 개발된 중성자 차폐재들의 차폐능, 연소특성, 난연성, 열적 및 역학적 성질 등을 평가하기 위해 여러 특성시험을 행하였다. 개발된 중성자 차폐재(수소원자 밀도, $6.1{\sim}6.2{\times}10^{22}atoms/cm^3$)들은 외국산 중성자 차폐재(NS-4-FR, $6.0{\times}10^{22}atoms/cm^3$)보다 수소원자 밀도가 높아 차폐능이 우수할 것으로 예측되며, 조사된 제반 특성들은 열분해온도; $267{\sim}270^{\circ}C$, 열전도도; $0.62{\sim}0.72W/m{\cdot}K$, 연소특성; $800^{\circ}C$ 이하, 평균연소시간; 5초 이하, 평균연소길이; 5mm 이하, 인장강도; $2.3{\sim}3.0kg/mm^2$, 압축강도; $5.3{\sim}13.3kg/mm^2$, 굴곡강도; $4.4{\sim}5.4kg/mm^2$ 등을 나타냈다.