• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2534-2546
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• 2023

Gas-cooled space reactor, which adopts He-Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He-Xe gas mixture in 2×2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%-50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He-Xe gas cooled space nuclear reactor.

• 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.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1825-1834
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• 2022

Performing high-fidelity computational fluid dynamics (HF-CFD) to predict the flow and heat transfer state of the coolant in the reactor core is expensive, especially in scenarios that require extensive parameter search, such as uncertainty analysis and design optimization. This work investigated the performance of utilizing a multi-fidelity reduced-order model (MF-ROM) in PWR rod bundles simulation. Firstly, basis vectors and basis vector coefficients of high-fidelity and low-fidelity CFD results are extracted separately by the proper orthogonal decomposition (POD) approach. Secondly, a surrogate model is trained to map the relationship between the extracted coefficients from different fidelity results. In the prediction stage, the coefficients of the low-fidelity data under the new operating conditions are extracted by using the obtained POD basis vectors. Then, the trained surrogate model uses the low-fidelity coefficients to regress the high-fidelity coefficients. The predicted high-fidelity data is reconstructed from the product of extracted basis vectors and the regression coefficients. The effectiveness of the MF-ROM is evaluated on a flow and heat transfer problem in PWR fuel rod bundles. Two data-driven algorithms, the Kriging and artificial neural network (ANN), are trained as surrogate models for the MF-ROM to reconstruct the complex flow and heat transfer field downstream of the mixing vanes. The results show good agreements between the data reconstructed with the trained MF-ROM and the high-fidelity CFD simulation result, while the former only requires to taken the computational burden of low-fidelity simulation. The results also show that the performance of the ANN model is slightly better than the Kriging model when using a high number of POD basis vectors for regression. Moreover, the result presented in this paper demonstrates the suitability of the proposed MF-ROM for high-fidelity fixed value initialization to accelerate complex simulation.

• Nuclear Engineering and Technology
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• 제2권1호
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• pp.19-25
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• 1970

이 연구는 다봉속내에서의 인접유로간의 냉각재 혼합류를 실험적으로 다룬 것이다. 실험은 19봉속내의 사각형 유로와 삼각형 유로간의 혼합류를 단상 유동과 공기-물 이상 유동에 물질 전달량을 측정하여 얻고 있다. 실험결과는 단상 유동에서 낮은 혼합률을, 공기-물 이상 유동에서 큰 혼합률을 얻고 있으며 공기-물 유동에서의 혼합률은 공기 체적률의 증가에 따라 감소되고 있음을 나타내고 있다. 공기-물 이상 유동에서의 높은 혼합률은 공기류에 의한 충분한 교란효과 때문인 것 같다.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.89-94
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• 2014

The 4th generation SFR is being designed with a milestone of construction by 2028. It is important to understand the subchannel flow characteristics in fuel assembly through the experimental investigations and to estimate the calculation uncertainties for insuring the confidence of the design code calculation results. The friction coefficient and the mixing coefficient are selected as primary parameters. The two parameters are related to the flow distribution and diffusion. To identify the flow distribution, an iso-kinetic method was developed based on the previous study. For the mixing parameters, a wire mesh system and a laser induced fluorescence methods were developed in parallel. The measuring systems were adopted on 37 rod bundle test geometry, which was developed based on the Euler number scaling. A scaling method for a design of experimental facility and the experimental identification techniques for the flow distribution and mixing parameters were developed based on the measurement requirement.

• 한국소음진동공학회논문집
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• 제22권7호
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• pp.619-625
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• 2012

Since the fuel is always exposed to turbulent flow, the grid strap shows flow induced vibration characteristics that impact on the nuclear fuel soundness. The dynamic behavior of grids in nuclear fuels is quite complex, since two pairs of spring and dimple support are contacted with rods by friction in the limited space. This paper focuses on investigation of the grid strap(test fuel strap, TFS) vibration in one cell. TFS consists of a single spring and double dimples. To identify the grid strap vibration, modal analysis of the strap is performed using finite element method(FEM). Modal testing on a $5{\times}5$ grid structure without rods is performed. The modal testing results are compared to analytic results. In addition, random test considering rod effect is performed about a $5{\times}5$ grid with rods under real contact condition in the air. Finally, the strap vibration of a $5{\times}5$ fuel bundle in investigation of flow induced vibration(INFINIT) facility is measured in real fluid velocity condition without heating. It is shown that modal frequencies from the test are almost equal to those peak frequencies in the INFINIT test.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1218-1230
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• 2019

A sub-channel solver, named ${\underline{S}}teady$ and ${\underline{T}}ransient$ ${\underline{A}}nalyzer$ for ${\underline{R}}eactor$ ${\underline{T}}hermal$ hydraulics (START), has been developed using the homogenous model for two-phase conditions of light water reactors. The code is developed as a fast and accurate TH-solver for coupled and multi-physics calculations. START has been validated against the NUPEC PWR Sub-channel and Bundle Test (PSBT) database. Tests like single-channel quality and void-fraction for steady state, outlet fluid temperature for steady state, rod-bundle quality and void-fraction for both steady state and transient conditions have been analyzed and compared with experimental values. Results reveal a good accuracy of solution for both steady state and transient scenarios. Axially different values for turbulent mixing coefficient are used based on different grid-spacer types. This provides better results as compared to using a single value of turbulent mixing coefficient. Code-to-code evaluation of PSBT results by the START code compares well with other industrial codes. The START code has been parallelized with the OpenMP algorithm and its numerical performance is evaluated with a large whole PWR core. Scaling study of START shows a good parallel performance.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4597-4606
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• 2023

Thermal analysis of the CANDU spent fuel dry storage canister is very important to ensure the integrity of the spent fuel. The analyses have been conducted using a conservative approach, with a particular focus on the peak cladding temperature (PCT) of the fuel rods in the canister. In this study, we have performed a realistic thermal analysis using a computational fluid dynamics (CFD) code. The canister contains 9 fuel bundle baskets. A detailed analysis of even a single basket requires significant computational resources. To overcome this challenge, we replaced each basket with an equivalent heat conductor (EHC), of which effective thermal conductivity (ETC) is developed from the results of detailed CFD calculations of a fuel bundle basket. Then, we investigated the effects of some conservative models, ultimately aiming at a realistic analysis. The results revealed: (i) The influence of convective heat transfer in the basket cannot be ignored, but it's less significant than expected. (ii) Modeling of the lifting rod is crucial, as it plays a decisive role in axial heat transfer at the center of the canister and significantly reduces the PCT. (iii) Convection within the canister is very important, as it not only reduces the PCT but also shifts its location upwards.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.345-354
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• 1994

수치 해석을 통하여 이차유동을 포함한 핵연료봉 주위의 난류 유동장을 예측하였다. 등방성 난류와 점성계수 모델과 이차 유동을 모사하기 위해 단순화된 대수응력모델을 사용하여 난류 운동 에너지 (k)와 난류 에너지 소멸률($\varepsilon$)의 이 방정식 모델과 운동량 방정식을 유한 차분법으로 풀어 유동장내의 평균속도, 이차유동, 난류 운동 에너지, 난류 응력 분포 등을 구하였다. 수치해석 예측치를 실험데이타와 비교하여 만족할만한 결과를 얻었고 유동장내에서 이차유동의 영향을 확인하였다. 즉 이차유동이 절대 크기는 작더라도 대류 효과에 의해 큰 영향을 미치는 것을 본 연구를 통해 알 수 있었다.

• 설비공학논문집
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• 제28권5호
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• pp.186-194
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• 2016

As a turbulence-enhancing device, a mixing vane installed at a spacer grid of the fuel assembly plays a role in improving the convective heat transfer by generating either swirl flow in the subchannels or cross flow between fuel rod gaps. Therefore, both configuration and arrangement pattern of a mixing vane are important factors that determine the performance of a mixing vane. In this study, in order to examine the flow distribution features inside $5{\times}5$ fuel assembly with swirl-type mixing vanes used in benchmark calculation of OECD/NEA, simulations were conducted with commercial CFD software ANSYS CFX R.14. Predicted results were compared to data measured from MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, the effect of swirl-type mixing vanes on flow pattern inside the fuel assembly was described.