• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1902-1912
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• 2024

This paper presents results from system-level modeling of a water-based reactor cavity cooling system using RELAP5-3D. The computational model is benchmarked with experimental data from a half-scale RCCS test facility at Argonne National Laboratory. The model prediction is first compared with a two-phase oscillatory baseline experimental case where mixed accuracy is obtained. The model shows reasonable prediction of mass flow rate, pressure, and temperature but significant overprediction of void fraction. The model prediction is then compared with a fault case where the inlet of the risers is gradually reduced using a throttling valve. As the valve is closed, the model is able to predict some major flow phenomena observed in the experiment such as the dampening of oscillations, the reintroduction of oscillations, as well as boiling, flashing, and geysering in the risers. However, the timeline of these events are not well captured by the model. The model is also used to investigate the evolution of flow regime in the chimney. This work highlights that the semi-empirical constitutive relations used in RELAP-3D could have a strong influence on the accuracy of the model in two-phase oscillatory flows.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1654-1666
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• 2024

The goal of this research is to propose a way to maximize small modular reactor (SMR) utilization to gain better market feasibility in support of carbon neutrality. For that purpose, a comprehensive tool was developed, combining off-design thermohydraulic models, economic objective models (levelized cost of electricity, annual profit), non-economic models (saved CO2), a parameter input sampling method (Latin hypercube sampling, LHS), and a multi-objective evolutionary algorithm (Non-dominated Sorting Algorithm-2, NSGA2 method) for optimizing a SMR-combined heat and power cycle (CHP) system design. Considering multiple objectives, it was shown that NSGA2+LHS method can find better optimal solution sets with similar computational costs compared to a conventional weighted sum (WS) method. Out of multiple multi-objective optimal design configurations for a 105 MWe design generation rating, a chosen reference SMR-CHP system resulted in its levelized cost of electricity (LCOE) below $60/MWh for various heat prices, showing economic competitiveness for energy market conditions similar to South Korea. Examined economic feasibility may vary significantly based on CHP heat prices, and extensive consideration of the regional heat market may be required for SMR-CHP regional optimization. Nonetheless, with reasonable heat market prices (e.g. district heating prices comparable to those in Europe and Korea), SMR can still become highly competitive in the energy market if coupled with a CHP system.

• 자원환경지질
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• 제42권6호
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• pp.609-618
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• 2009

지층 내에 발달한 고투수성 단층은 유체, 에너지, 그리고 용질이 이동하는데 있어서 중요한 역할을 하는 지질구조이다. 따라서 고투수성 단층 주변부에서는 지열수(혹은 온천), 지열 이상대, 그리고 금속 광상 등이 형성될 가능성이 크다. 단층의 구조에 따른 지하수 유동과 이에 따른 지층 내의 열적 상태를 확인하기 위해서 단층 구조가 다른 세 가지의 경우에 대해서 이차원 열-수리적 거동 모델링을 수행하였다. 모델링 결과로부터 세 가지 모든 단층 구조의 경우에서 단층의 투수율이 커지면 단층대에서의 지하수 용출 온도가 초기 온도 보다 높아지는 경향을 확인 할 수 있다. Peclet number 와 단층대에서의 용출온도의 상관관계 분석으로부터 상관계수($R^2$)가 0.98로 상당히 높은 것을 확인하였다. Peclet number가 1이상 일 때 단층대에서는 약 $32^{\circ}C$ 이상의 온도가 용출되며 이러한 경우에 단층대에서의 열흐름은 매질을 통한 전도 보다는 지하수에 의한 대류의 영향이 큰 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1053-1064
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• 2009

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1236-1249
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• 2021

Although there are still controversial opinions and uncertainty on application of SiC_f/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiC_f/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO₂-SiC was poor due to its low inpile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO₂-SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO₂-SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

• 한국수소및신에너지학회논문집
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• 제33권6호
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• pp.808-818
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• 2022

In this research paper, we investigated the cryogenic line chill down characteristics of liquefied natural gas (LNG). A numerical analysis model was established and verified so that it can calculate the precise cooling characteristics of cryogenic fluid for the stable and safe utilization especially such as LNG and liquid hydrogen. The numerical modeling was programmed by C++ as an one-dimensional homogeneous model. The thermohydraulic cooling process was simulated using mass, momentum, energy conservation equations and appropriate heat transfer correlations. In this process, the relevant heat transfer correlations for nuclear boiling, transition boiling, film boiling, and single-phase heat transfer that can predict the experimental results were implemented. To verify the numerical modeling, several cryogenic line chill down experiments using LNG were conducted at the Korea Institute of Machinery & Materials (KIMM) LNG and Cryogenic Technology Center.