• Nuclear Engineering and Technology
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• 제33권4호
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• pp.365-374
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• 2001

In order to examine the in-reactor performance of very-high-density dispersion fuels for high flux performance research reactors, U-l0wt.%Mo microplates containing centrifugally atomized powder were irradiated at low temperature. The U-l0wt.%Mo dispersion fuels show stable in- reactor irradiation behaviors even at high burn-up, similar to U$_3$Si$_2$ dispersion fuels. The atomized U-l0wt.%Mo fuel particles have a fine and a relatively uniform fission gas bubble size distribution. Moreover, only one of third of the area of the atomized fuel cross-sections at 70a1.% burn-up shows fission gas bubble-free zones, This appears to be the result of segregation into high Mo and low Mo.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• 방사성폐기물학회지
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• 제20권3호
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• 대한기계학회논문집 C: 기술과 교육
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• 제3권4호
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• pp.299-305
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• 2015

미래에너지의 해법으로 원자력에너지를 이용한 물분해 수소생산시스템의 핵심기술을 개발하였다. 안전성을 보장할 수 있는 제4세대 원자로인 초고온가스로의 고열을 이용하여 황요오드 열화학적인 방법으로 물을 분해하여 수소를 생산하는 기술이다. 원자력수소생산 핵심기술은 초고온에서의 열을 공급하는 것을 모사하는 초고온 실험기술, 초고온가스로의 안전성을 모사하는 연구, 초고온가스로의 노심과 안전성을 해석할 수 있는 도구의 개발, 초고온가스로에 사용하는 연료제조기술, 물을 분해하여 열화학적인 방법으로 수소를 생산하는 기술로 구성된다. 원자력수소생산에 필요한 핵심기술을 개발하고 실험실 규모로 입증하였으며, 대규모 실용화를 위해서 선결되어할 미완성 기술을 제시하였다. 본 기술은 제4세대 원자로개발 국제공동연구로 수행한 기술로서 향후 미래의 원자로 기술이다.

• 한국환경보건학회지
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• 제22권2호
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• pp.90-95
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• 1996

This study was performed to calculate the degration rate coefficient, operating parameters to meet the effluent standards, and the temperature adjustment coefficients to each parameter of pollution by seasonal variation of concentration and temperature of influent in livestock wastewater treatment by sequencing batch reactor process in field scale. The followings are the conclusions that were derived from this study. 1. In the field, temperature of livestock wastewater in reactor was 20.3$\circ$C in summer and 6.0$\circ$C in winter. The ratio of BOD:TKN: T-P in influent was 100:80:7. BOD loadings in winter and spring were 0.26 and 0.43 kg $BOD/m^3$ day, respectively. Those in summer and fall were 0.25 and 0.13 kg $BOD/m^3$ day, respectively. 2. The degradation rate coefficient for TKN was larger in summer and fall in which temperature was high than that in which temperature was high than that in winter and spring in which concentration was high. On the contrary, the phosphorus uptake rate was larger in winter and spring than that in summer and fall. 3. The hydraulic retention time in winter and spring was longer than that in summer and fall. Especially, in order to meet the standard for TKN of 120 mg/l in winter in which temperature of wastewater was 6.0$\circ$C, as the MLSS concentration was increased from 4, 000 to 7, 000 mg/l, the hydraulic retention time was increased from 212 to 121 hours. But, in order to shorten that less than 121 hours for the economical wastewater treatment, countermeasure to increase temperature of wastewater in the reactor should be considered. 4. the temperature adjustment coefficients for BOD, $COD_{Mn}$, TKN and T-P were 1.0241, 1.0225, 1.0541 and 1.0495, respectively. Namely, the treatment of TKN was most sensitively affected by temperature. For the purpose of the effective removal of nitrogen and phosphorus which are sensitive to temperature, it is necessary to keep the temperature of livestock wastewater more than 20$\circ$C which is the temperature of it in summer.

• Nuclear Engineering and Technology
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• 제38권3호
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• pp.293-300
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• 2006

A liquid metal reactor (LMR) operated at high temperatures is subjected to both cyclic mechanical loading and thermal loading; thus, creep-fatigue is a major concern to be addressed with regard to maintaining structural integrity. The Korea Advanced Liquid Metal Reactor (KALIMER), which has a normal operating temperature of $545^{\circ}C$ and a total service life time of 60 years, is composed of various cylindrical structures, such as the reactor vessel and the reactor baffle. This study focuses on the creepfatigue crack initiation for a cylindrical Y-junction structure made of 316 stainless steel (SS), which is subjected to cyclic axial tensile loading and thermal loading at a high-temperature hold time of $545^{\circ}C$. The evaluation of the considered creep-fatigue crack initiation was carried out utilizing the ${\sigma}_d$ approach of the RCC-MR A16 guide, which is the high-temperature defect assessment procedure. This procedure is based on the total accumulated strain during the service time. To confirm the evaluated result, a high-temperature creep-fatigue structural test was performed. The test model had a circumferential through wall defect at the center of the model. The defect front of the test model was investigated after the $100^{th}$ cycle of the testing by utilizing a metallurgical inspection technique with an optical microscope, after which the test result was compared with the evaluation result. This study shows how creep-fatigue crack initiation for a high-temperature structure can be predicted with conservatism per the RCC-MR A16 guide.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1310-1319
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• 2024

The reactor cavity cooling system (RCCS) is a passive reactor safety system commonly present in the designs of High-Temperature Gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel by means of natural convection and radiation. It is one of the factors responsible for ensuring that the reactor does not melt down under any plausible accident scenario. For the simulation of accident scenarios, which are transient phenomena unfolding over a span of up to several days, intermediate fidelity methods and system codes must be employed to limit the models' execution time. These models can quantify radiation heat transfer well, but heat transfer caused by natural convection must be quantified with the use of correlations for the heat transfer coefficient. It is difficult to obtain reliable correlations for HTGR RCCS heat transfer coefficients experimentally due to such a system's size. They could, however, be obtained from high-fidelity steady-state simulations of RCCSs. The Rayleigh number in RCCSs is too high for using a Direct Numerical Simulation (DNS) technique; thus, a Reynolds-Averaged Navier-Stokes (RANS) approach must be employed. There are many RANS models, each performing best under different geometry and fluid flow conditions. To find the most suitable one for simulating an RCCS, the RANS models need to be validated. This work benchmarks various RANS models against three experiments performed on the HTTR RCCS Mockup by the Japanese Atomic Energy Agency (JAEA) in 1993. This facility is a 1/6 scale model of a vessel cooling system (VCS) for the High Temperature Engineering Test Reactor (HTTR), which is operated by JAEA. Multiple RANS models were evaluated on a simplified 2d-axisymmetric geometry. They were found to reproduce the experimental temperature profiles with errors of up to 22% for the lowest temperature benchmark and 15% for the higher temperature benchmarks. The results highlight that the pragmatic turbulence models need to be validated for high Rayleigh natural convection-driven flows and improved accordingly, more publicly available experimental data of RCCS resembling experiments is needed and indicate that a 2d-axisymmetric geometry approximation is likely insufficient to capture all the relevant phenomena in RCCS simulations.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.57-64
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• 2010

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established the gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype to be tested in the loop. In this study, as part of the high-temperature structural-integrity evaluation of the PHE prototype, which is scheduled to be tested in the gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal expansion analysis of the PHE prototype. The results obtained in this study will be used to design the performance test setup for the PHE prototype.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.48-53
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• 2011

A PHE (Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. In this study, as a part of the evaluation on the high-temperature structural integrity of the small-scale PHE prototype, we carried out macroscopic high-temperature structural analysis of the small-scale PHE prototype under the gas loop test conditions considering the pipeline stiffness.

• 한국막학회:학술대회논문집
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• 한국막학회 2003년도 추계 총회 및 학술발표회
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• pp.25-33
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• 2003

원자력 발전의 고온 가스로(high temperature gas-cooled reactor, HTGR)의 냉각제로 사용되는 He가스의 폐열에너지를 이용하여 물을 분해해서 수소를 생산하는 “열화학적 수소제조 IS프로세스”에서의 분리막 기술의 응용에 대해 정리하였다. 고온 원자력 열에너지를 이용한 열화학적 수소 제조법은 실현 가능한 단계까지 왔다고 생각되며, 아직 연구 개발 과제가 많이 남아 있지만, 미래의 청정에너지 중의 하나인 수소를 대량 생산할 수 있는 가능성을 갖고 있다.