• 한국유체기계학회 논문집
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• 제18권6호
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• pp.5-11
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• 2015

In a pressurized water reactor (PWR), control rod assembly (CRA) falls into the guide tubes of a fuel assembly due to gravity for scram. Various theoretical approaches and numerical analyses have been performed because its shape is simple and its design was completely developed several decades ago. A control rod assembly for a sodium-cooled faster reactor (SFR) which is geometrically more complicated is being actively developed in Korea nowadays. Drop time and impact velocity of a CRA are important parameters with respect to reactivity insertion time and the mechanical robustness of a CRA and a guide duct. In this paper, computational method considering simultaneously the equation of motion for rigid body and the Navier-Stokes equations for fluid is suggested and verified by comparison with theoretical analysis results. Through this valuable CFD analysis method, drop time and impact velocity of initially designed SFR CRA are evaluated before performing scram tests with it.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.257-264
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• 2024

The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, which is the sodium-water reaction. Since sodium and water are separated only by a heat transfer tube with a thickness of only a few mm, there is inherently a risk of a sodium-water reaction (SWR) accident in the SFR. In this study, it is attempted to quantitatively evaluate the resistance of SWR accidents by replacing the shell and tube steam generator with printed circuit steam generator (PCSG) as a method to mitigate the SWR accident. To do this, a CATS-S (Compact Accident Tolerance Steam Generator-SWR) facility was designed and built. And for the quantitative evaluation of accident resistance, a methodology for measuring the impingement wastage rate was established. As a result of this research, the impingement wastage rate caused by SWR generated in a PCSG was measured first time. It was confirmed that the impingement wastage phenomenon was suppressed in the PCSG, and the accident resistance was higher than that of the SWR through comparison with the experimental results performed in the existing shell and tube steam generator. In conclusion, a PCSG is more resistant to impingement wastage as a result of the SWR accident than existing shell and tube steam generators, and it is estimated that a PCSG can mitigate SWR accidents, an inherent problem of SFR.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1874-1889
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• 2022

In this study, we evaluate the thermal-hydraulic performance and economics of Printed Circuit Heat Exchanger (PCHE) according to the channel types and associated shape variables for the design of recuperators with Sodium-cooled Fast Reactors (SFRs). To perform the evaluations with variables such as the Reynolds number, channel types, tube diameter, and shape variables, a code for the heat exchanger is developed and verified through a comparison with experimental results. Based on the code, the volume and pressure drop are calculated, and an economic assessment is conducted. The zigzag type, which has bending angle of 80° and a tube diameter of 1.9 mm, is the most economical channel type in a SFR using CO₂ as the working fluid. For a SFR using N₂, we recommend the airfoil type with vertical and horizontal numbers of 1.6 and 1.1, respectively. The airfoil type is superior when the mass flow rate is large because the operating cost changes significantly. When the mass flow rate is small, volume is a more important design parameter, therefore, the zigzag type is suitable. In addition, we conduct a sensitivity analysis based on the production cost of the PCHE to identify changes in optimal channel types.

• 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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• 제54권10호
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• pp.3928-3942
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• 2022

Thermal striping is a complex thermal-hydraulic phenomenon caused by fluid temperature fluctuations that can also cause high-cycle thermal fatigue to the structural wall of sodium-cooled fast reactors (SFRs). Numerical simulations using large-eddy simulation (LES) were performed to predict and evaluate the characteristics of the temperature fluctuations related to thermal striping in the upper internal structure (UIS) of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Specific monitoring points were established for the fluid region near the control rod driving mechanism (CRDM) guide tubes, CRDM guide tube walls, and UIS support plates, and the normalized mean and fluctuating temperatures were investigated at these points. It was found that the location of the maximum amplitude of the temperature fluctuations in the UIS was the lowest end of the inner wall of the CRDM guide tube, and the maximum value of the normalized fluctuating temperatures was 17.2%. The frequency of the maximum temperature fluctuation on the CRDM guide tube walls, which is an important factor in thermal striping, was also analyzed using the fast Fourier transform analysis. These results can be used for the structural integrity evaluation of the UIS in SFR.

• 기계저널
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• 제50권3호
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• pp.28-31
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• 2010

이 글에서는 제4세대(Generation-IV) 원자로시스템의 자원활용 측면에서 핵연료 주기와 관련하여 새롭게 부각되고 있는 소듐냉각고속로(SFR: Sodium-cooled Fast Reactor)의 개발 목적 및 설계 특성을 기술하고 원자로 구조관점에서 가압경수로(PWR)와 비교 설명한다.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.111-120
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• 2021

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it's very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

• 대한기계학회논문집A
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• 제41권5호
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• pp.375-380
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• 2017

본 논문은 한국원자력연구원에서 개발 중인 소듐냉각 고속로 핵연료의 연료봉 하단 마개에 있는 관통구멍과 마운팅 레일의 원기둥 형상과의 접촉부에 발생하는 접촉 손상을 예측하기 위한 가속시험 방법을 연구한 것이다. 가속시험 조건으로서 연료봉의 유체유발진동수 및 진폭을 유한요소 해석을 통하여 구하였다. 약 35000 시간의 연료봉 수명기간을 고려한 가속시험 시간을 결정하기 위해 일반 기계부품류의 신뢰성 평가 방법을 적용하였으며, 이때 가장 보수적인 형상 모수와 원자로 내에서의 연료봉 파손허용 개수 기준 및 연료봉 피복관 재료인 HT-9강의 피로수명 데이터를 이용하였다. 시편의 개수를 5개로 하였을 때, 최종적으로 계산된 가속 시험시간은 각 시편 당 16.5시간이었다. 가속시험 후 전체 시편에 어떠한 접촉손상도 관찰되지 않을 때 연료봉의 수명기간 중 $B_{0.004}$ 수명이 신뢰수준 99%로 보장되는 것으로 평가하였다.