• Nuclear Engineering and Technology
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• 제41권4호
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• pp.427-446
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• 2009

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical $CO_2$ Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.43-53
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• 2018

It is important to clarify the characteristics of flow regimes underlying the debris bed formation behavior that might be encountered in core disruptive accidents of sodium-cooled fast reactors. Although in our previous publications, by applying dimensional analysis technique, an empirical model, with its reasonability confirmed over a variety of parametric conditions, has been successfully developed to predict the regime transition and final bed geometry formed, so far this model is restricted to predictions of debris mixtures composed of spherical particles. Focusing on this aspect, in this study a new series of experiments using nonspherical particles have been conducted. Based on the knowledge and data obtained, an extension scheme is suggested with the purpose of extending the base model to cover the particle-shape influence. Through detailed analyses and given our current range of experimental conditions, it is found that, by coupling the base model with this scheme, respectable agreement between experiments and model predictions for the regime transition can be achieved for both spherical and nonspherical particles. Knowledge and evidence from our work might be utilized for the future improvement of design of an in-vessel core catcher as well as the development and verification of sodium-cooled fast reactor severe accident analysis codes in China.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.589-600
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• 2022

Investigations on the molten-pool sloshing behavior are of essential value for improving nuclear safety evaluation of Core Disruptive Accidents (CDA) that would be possibly encountered for Sodium-cooled Fast Reactors (SFR). This paper is aimed at synthesizing the knowledge from our recent studies on molten-pool sloshing behavior with solid particles conducted at the Sun Yat-sen University. To better visualize and clarify the mechanism and characteristics of sloshing induced by local Fuel-Coolant Interaction (FCI), experiments were performed with various parameters by injecting nitrogen gas into a 2-dimensional liquid pool with accumulated solid particles. It was confirmed that under different particle-bed conditions, three representative flow regimes (i.e. the bubble-impulsion dominant, transitional and bed-inertia dominant regimes) are identifiable. Aimed at predicting the regime transitions during sloshing process, a predictive empirical model along with a regime map was proposed on the basis of experiments using single-sized spherical solid particles, and then was extended for covering more complex particle conditions (e.g. non-spherical, mixed-sized and mixed-density spherical particle conditions). To obtain more comprehensive understandings and verify the applicability and reliability of the predictive model under more realistic conditions (e.g. large-scale 3-dimensional condition), further experimental and modeling studies are also being prepared under other more complicated actual conditions.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1303-1314
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• 2016

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical $CO_2$ cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.9-18
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• 2024

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents' acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristics are analyzed through BALI experiments with similar condition. In this study, computational fluid dynamics (CFD) via Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) for BALI experiments were performed. Steady-state CFD analysis was performed on three turbulence models, and SST k-ω model was in good agreement with the experimental measurement temperature within the maximum error range of 1.9%. LES CFD analysis was performed based on the RANS analysis results and it was confirmed that the temperature and wall heat flux for depth was consistent within an error range of 1.0% with BALI experiment. The LES CFD analysis results were compared with those of the Lagrangian-based solver. LES matched the temperature distribution better than SOPHIA, but SOPHIA calculated the position of boundary between stratified layer and convective layer more accurately. On the other hand, Lagrangian-based solver predicted several small eddy behaviors of the convective layer and LES predicted large vortex behavior. The vibration characteristics near the cooling part of the BALI experimental device were confirmed through Fast Fourier Transform (FFT) investigation. It was found that the power spectral density for pressure at least 10 times higher near the side cooling than near the top cooling.

• 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.

• 콘크리트학회논문집
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• 제28권2호
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• pp.197-204
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• 2016

본 논문에서는 타 재료에 비해 수소 원소 함유량이 높아 고속 중성자 차폐에 유리한 합성 고분자 화합물과 중성자 포획단면적이 큰 붕소화합물을 각각 혼입한 모르타르를 대상으로 중성자 차폐성능을 분석하였다. 합성 고분자 화합물의 종류, 형상, 크기, 함량 및 붕소화합물의 함량에 따라 총 16개 모르타르 배합을 설계하였으며, 각 배합의 슬럼프 플로우, 28일 압축 및 인장강도를 측정하고, 고속 중성자 및 열중성자에 대한 차폐실험을 수행하였다. 합성 고분자 화합물의 선량 투과율은 모르타르 대비 최대 38.5%까지 감소하였으며, 공기 중 선량의 26.3%까지 차폐하였다. 붕소화합물을 혼입한 모르타르의 열중성자 차폐율은 최대 90.3%로 대부분의 열중성자를 차폐하였다. 비록 화합물 혼입에 의해 모르타르의 기본 특성은 저하되었으나, 표면 개질, 특수 혼화제 첨가 등 지속적인 연구를 통하여 성능 저하를 최소화할 수 있을 것으로 판단되며, 중성자 투과성능 역시 다양한 타입의 시험체 조합을 통한 레이어 시스템 도입 등으로 다양한 투과성능에 따른 맞춤형 설계를 제공할 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제2권2호
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• pp.85-95
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• 1970

두개의 threshold detector로서 인자로의 폭발사고시에 방출되는 속 중성자의 속도분포를 측정하고 그로부터 속 중성자의 인체흡수선량을 계산하였다. 이때 속 중성자의 속도분포는 하나의 스펙트럼 매개변수에 의하여 결정된다는 가정으로부터 얻어지는데 이 매개변수는 threshold detector의 반응율을 측정하므로서 구해진다. 속 중성자의 인체흡수선량은 속 중성자의 속도분포 변화에 따라 큰 변동이 없었으나 threshold detector의 평균반응단면적은 크게 변하였다. 따라서 속 중성자의 속도분포에 관계없이 threshold detector의 평균반응단면적을 고정된 값으로 취하여 속 중성자선량을 계산한다면 큰 오차를 일으키게 될 것이라는 것을 보여주었다. 한편 핵분열에서 방출되는 속 중성자의 속도분포에 대한 세 해석적 표현인 즉 Watt, Cranberg및 Maxwellian 공식들로부터 속 중성자 선량을 계산하여 서로 비교하였다. Watt 및 Cranberg 공식들로 부터 얻어진 속 중성자선량은 Maxwellian 공식으로부터 얻어진 그것보다 약간 높은 값을 보여 주었으며 Watt 공식에 의한 선량계산치는 Cranberg 공식에 의한 그것과 비슷한 값을 보여주었다.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.700-711
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• 2015

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called "self-wastage phenomena." A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).