• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.176-177
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.148-148
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• 2003

• Nuclear Engineering and Technology
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• 제36권3호
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• pp.248-262
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• 2004

Turbulence models are separately assessed for a three dimensional thermal-hydraulic analysis of the integral reactor SMART. Seven models (mixing length, k-l, standard $k-{\epsilon},\;k-{\epsilon}-f{\mu},\;k-{\epsilon}-v2$, RRSM, and ERRSM) are investigated for flat plate channel flow, rotating channel flow, and square sectioned U-bend duct flow. The results of these models are compared to the DNS data and experiment data. The results are assessed in terms of many aspects such as economical efficiency, accuracy, theorization, and applicability. The standard $k-{\epsilon}$ model (high Reynolds model), the $k-{\epsilon}-v2$ model, and the ERRSM (low Reynolds models) are selected from the assessment results. The standard $k-{\epsilon}$ model using small grid numbers predicts the channel flow with higher accuracy in comparison with the other eddy viscosity models in the logarithmic layer. The elliptic-relaxation type models, $k-{\epsilon}-v2$, and ERRSM have the advantage of application to complex geometries and show good prediction for near wall flows.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.219-222
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• 2013

Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7-28) ${\times}10^{19}n/cm^2$ (E>0.1MeV) at $250^{\circ}C$, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at $250^{\circ}C$ did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1%) at fluences of (0.7~28) ${\times}10^{19}n/cm^2$ (E>0.1MeV).

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• 정보처리학회논문지:컴퓨터 및 통신 시스템
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• 제2권12호
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• pp.561-568
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• 2013

원전 계측제어계통의 제어기기에 대한 사이버공격은 원자로 정지 또는 방사선 누출 등의 심각한 재앙을 유발할 가능성이 있다. 그러므로 원전 규제기관 및 산업표준의 요건에 따라 심층방호 특성을 고려한 사이버보안 대책설계가 원전 계측제어계통 설계에 요구된다. 본 논문에서는 사이버공격에 대한 원전 계측제어계통의 심층방호 대처 설계시 필수적으로 요구되는 단방향통신 구조를 구현하기 위한 설계방법을 제안한다. 제안한 단방향 통신 구조 구현방법은 통신보드 내에 구현되었다. 단방향 통신을 지원하는 통신보드는 시험을 통하여 심층방호 설계에 필요한 단방향 통신 구조를 제공하며, 사이버공격에 강인한 특성을 갖는 것으로 분석되었다. 이러한 사이버공격에 대비한 심층방호 대처 설계 및 단방향 통신을 지원하는 통신구조는 일체형원자로(SMART) 계측제어계통 설계에 적용되었다.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1015-1024
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• 2009

The presence of a non-condensable gas can considerably reduce the level of condensation heat transfer. The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of the System-integrated Modular Advanced ReacTor (SMART) and the Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR). This study examined the capability of the Multi-dimensional Analysis of Reactor Safety (MARS) code to predict condensation heat transfer in a vertical tube containing a non-condensable gas. Five experiments were simulated to evaluate the MARS code. The results of the simulations showed that the MARS code overestimated the condensation heat transfer coefficient compared to the experimental data. In particular, in small-diameter cases, the MARS predictions showed significant differences from the measured data, and the condensation heat transfer coefficient behavior along the tube did not match the experimental data. A new method for calculating condensation heat transfer coefficient was incorporated in MARS that considers the interfacial shear stress as well as flow condition determination criterion. The predictions were improved by using the new condensation model.

• 대한전기학회논문지:시스템및제어부문D
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• 제52권8호
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• pp.452-461
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• 2003

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indicator system and its actual implementation in the existing nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indicator. The hysteresis of reed switches is one of the important factors in a repeat accuracy of control rod position indicator as well. This paper investigates efficiency of the magnetic flux concentrator and the hysteresis using FEM and verified differences in physicals characteristics by comparing the results of FEM and those of the experiment. As a result, it is shown that the characteristics of prototype control rod position indicator have a good agreement with the results of FEM.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1254-1266
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• 2024

This study investigates the application of supercritical carbon dioxide (S-CO₂) direct-cycle micro modular reactors (MMRs) in primary frequency control (PFC), which is a scenario characterized by significant load fluctuations that has received less attention compared to secondary load-following. Using a modified GAMMA + code and a deep neural network-based turbomachinery off-design model, the authors conducted an analysis to assess the behavior of the reactor core and fluid system under different PFC scenarios. The results indicate that the acceptable range for sudden relative electricity output (REO) fluctuations is approximately 20%p which aligns with the performance of combined-cycle gas turbines (CCGTs) and open-cycle gas turbines (OCGTs). In S-CO₂ direct-cycle MMRs, the control of the core operates passively within the operational range by managing coolant density through inventory control. However, when PFC exceeds 35%p, system control failure is observed, suggesting the need for improved control strategies. These findings affirm the potential of S-CO₂ direct-cycle MMRs in PFC operations, representing an advancement in the management of grid fluctuations while ensuring reliable and carbon-free power generation.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3315-3329
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• 2024

Small Modular Reactors (SMRs) adopt the Helically Coiled Once-Through Steam Generators (OTSG) extensively for its compactness and higher heat transfer efficiency. As a heat exchanger between the primary side (reactor coolant system) and the secondary side (feedwater and steam system) of nuclear steam supply system, the inlet/outlet conditions both of shell side and tube side of OTSGs have significant impacts on overall system response. Considering the flexible operation of SMRs and heat application by extracting steam, a simulation tool for accurate prediction of the OTSG dynamic behaviors would be required for optimizing design and control. In this study, the OTSG dynamic simulation model has been developed. Mathematical governing equation has been derived by using moving boundary approach and a simulation module has been developed by using Modelica Language. The developed module has been compared with publicly available experimental results and benchmarked with MARS-KS calculation results. Also, it has been incorporated into the integrated SMR model (i.e., reactor core, primary side, secondary side) and dynamic behaviors with reactivity feedback and heat balancing have been investigated. In both of steady-state and transient conditions, it shows the promising accuracy.