• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2007년도 제29회 추계학술대회논문집
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• pp.275-281
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• 2007

추진기관 노즐은 고온 고압의 연소가스를 화학에너지에서 운동에너지로 변환시켜 추력을 발생시킨다. 따라서 노즐 내부 벽면은 고온 고압의 연소가스에 노출되며, 특히 노즐 목에서는 최대 열하중을 받는 구간으로서 열구조적으로 안정성을 확보한 냉각 시스템 설계가 이루어져야 한다. 본 추진기관의 노즐은 수냉 방식으로서 열전달 효율을 높이기 위해 냉각 채널 구조로 되어 있다. 본 연구에서는 추진기관 노즐을 위한 냉각 채널 구조의 설계형상에 대해 개념 설계 및 유동 해석을 수행하고 공급압력 및 유량 변화에 따른 입/출구 사이의 압력 강하량을 예측하였다. 또한 압력 손실 및 설계 유량 공급을 위한 압력 조건에 대해서도 평가하였다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.873-879
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• 2024

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

• 한국유체기계학회 논문집
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• 제18권2호
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• pp.60-66
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• 2015

Conjugate heat transfer analysis was performed to investigate the flow and cooling performance of the high pressure turbine nozzle of gas turbine engine. The CHT code was verified by comparison between CFD results and experimental results of C3X vane. The combination of k-${\omega}$ based SST turbulence model and transition model was used to solve the flow and thermal field of the fluid zone and the material property of CMSX-4 was applied to the solid zone. The turbine nozzle has two internal cooling channels and each channel has a complex cooling configurations, such as the film cooling, jet impingement, pedestal and rib turbulator. The parabolic temperature profile was given to the inlet condition of the nozzle to simulate the combustor exit condition. The flow characteristics were analyzed by comparing with uncooled nozzle vane. The Mach number around the vane increased due to the increase of coolant mass flow flowed in the main flow passage. The maximum cooling effectiveness (91 %) at the vane surface is located in the middle of pressure side which is effected by the film cooling and the rib turbulrator. The region of the minimum cooling effectiveness (44.8 %) was positioned at the leading edge. And the results show that the TBC layer increases the average cooling effectiveness up to 18 %.

• 한국전산유체공학회지
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• 제20권1호
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• pp.84-91
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• 2015

SMART adopts, very unique facility, an FMHA to enhance the thermal and flow mixing capability in abnormal conditions of some steam generators or reactor coolant pumps. The FMHA is important for enhancing thermal mixing of the core inlet flow during a transient and even during accidents, and thus it is essential that the thermal mixing characteristics of flow of the FMHA be understood. Investigations for the mixing characteristics of the FMHA had been performed by using experimental and CFD methods in KAERI. In this study, the temperature distribution at the core inlet region is investigated for several abnormal conditions of steam generators using the commercial code, FLUENT 12. Simulations are carried out with two kinds of FMHA shapes, different mesh resolutions, turbulence models, and steam generator conditions. The CFD results show that the temperature deviation at the core inlet reduces greatly for all turbulence models and steam generator conditions tested here, and the effect of mesh refinement on the temperature distribution at the core inlet is negligible. Even though the uniformity of FMHA outlet hole flow increases the thermal mixing, the temperature deviation at the core inlet is within an acceptable range. We numerically confirmed that the FMHA applied in SMART has an excellent mixing capability and all simulation cases tested here satisfies the design requirement for FMHA thermal mixing capability.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.36-48
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• 2022

As a valid numerical method to obtain a high-resolution result of a flow field, computational fluid dynamics (CFD) have been widely used to study coolant flow and heat transfer characteristics in fuel rod bundles. However, the time-consuming, iterative calculation of Navier-Stokes equations makes CFD unsuitable for the scenarios that require efficient simulation such as sensitivity analysis and uncertainty quantification. To solve this problem, a reduced-order model (ROM) based on proper orthogonal decomposition (POD) and machine learning (ML) is proposed to simulate the flow field efficiently. Firstly, a validated CFD model to output the flow field data set of the rod bundle is established. Secondly, based on the POD method, the modes and corresponding coefficients of the flow field were extracted. Then, an deep feed-forward neural network, due to its efficiency in approximating arbitrary functions and its ability to handle high-dimensional and strong nonlinear problems, is selected to build a model that maps the non-linear relationship between the mode coefficients and the boundary conditions. A trained surrogate model for modes coefficients prediction is obtained after a certain number of training iterations. Finally, the flow field is reconstructed by combining the product of the POD basis and coefficients. Based on the test dataset, an evaluation of the ROM is carried out. The evaluation results show that the proposed POD-ROM accurately describe the flow status of the fluid field in rod bundles with high resolution in only a few milliseconds.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3910-3917
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• 2021

Reactor fuel's power distribution plays a vital role in designing the new generation thermionic Space Reactor Power Systems (SRPS). In this paper, the 1/12th SPACE-R's full reactor core was numerically analyzed with two kinds of different axial power distribution, to identify their impacts on thermal-hydraulic and thermoelectric characteristics. In the benchmark study, the maximum error between numerical results and existing data or design values ranged from 0.2 to 2.2%. Four main conclusions were obtained in the numerical analysis: a) The axial power distribution has less impact on coolant temperature. b) Axial power distribution influenced the emitter temperature distribution a lot, when the core power was cosine distributed, the maximum temperature of the emitter was 194 K higher than that of the uniform power distribution. c) Comparing to the cosine axial power distribution, the uniform axial power distribution would make the maximum temperature in each component of the reactor core much lower, reducing the requirements for core fuel material. d) Voltage and current distribution were similar to the axial electrode temperature distribution, and the axial power distribution has little effect on the output power.