• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4329-4337
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• 2022

Jet impinging device is designed for decay heat removal on horizontal fuel rods in a low temperature heating reactor. An experimental system with a fuel rod simulator is established and experiments are performed to evaluate water film covering capacity, within 0.0287-0.0444 kg/ms mass flow rate, 0-164.1 kW/m² heating flux and 13.8-91.4℃ feeding water temperature. An effective method to obtain the film coverage rate by infrared equipment is proposed. Water film flowing patterns are recoded and the film coverage rates at different circumference angles are measured. It is found the film coverage rate decreases with heating flux during single-phase convection, while increases after onset of nucleate boiling. Besides, film coverage rate is found affected by Marangoni effect and film accelerating effect, and surface wetting is significantly facilitated by bubble behavior. Based on the observed phenomenon and physical mechanism, dry-out depth and initial dry-out rate are proposed to evaluate film covering potential on a heating surface. A model to predict film coverage rate is proposed based on the data. The findings would have reliable guide and important implications for further evaluation and design of decay heat removal system of new reactors, and could be helpful for passive containment cooling research.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.655-668
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• 2023

In radiography, an antiscatter grid is a well-known device for eliminating unexpected x-ray scatter. We investigate a new stationary grid artifact suppression method based on a nonsubsampled contourlet transform (NSCT) incorporated with Gaussian band-pass filtering. The proposed method has an advantage that extracts the Moiré components while minimizing the loss of image information and apply the prior information of Moiré component positions in multi-decomposition sub-band images. We implemented the proposed algorithm and performed a simulation and an experiment to demonstrate its viability. We did this experiment using an x-ray tube (M-113T, Varian, focal spot size: 0.1 mm), a flat-panel detector (ROSE-M Sensor, Aspenstate, pixel dimension: 3032 × 3800 pixels, pixel size: 0.076 mm), and carbon graphite-interspaced grids (JPI Healthcare, 18 cm × 24 cm, line density: 103 LP/inch and 150 LP/inch, ratio: 5:1, focal distance: 65 cm). Our results indicate that the proposed method successfully suppressed grid artifacts by reducing them without either reducing the spatial resolution or causing negative side effects. Consequently, we anticipate that the proposed method can improve image acquisition in a stationary grid x-ray system as well as in extended x-ray imaging.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.109-118
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• 2023

In this paper, the evaluation method of fluidelastic instability (FEI) of newly designed steam generator tubing in pressurized water reactor (PWR) nuclear power plants is discussed. To obtain the parameters for prediction of the critical velocity of FEI for steam generator tubes, experimental research is carried out, and the design parameters are determined. Using CFD numerical simulation, the tube array scale of the model experiment is determined, and the experimental device is designed. In this paper, 7 groups of experiments with void fractions of 0% (water), 10%, 20%, 50%, 75%, 85% and 95% were carried out. The critical damping ration, fundamental frequency and critical velocity of FEI of tubes in flowing water were measured. Through calculation, the total mass and instability constant of the immersed tube are obtained. The critical damping ration measured in the experiment mainly included two-phase damping and viscous damping, which changed with the change in void fraction from 1.56% to 4.34%. This value can be used in the steam generator design described in this paper and is conservative. By introducing the multiplier of frequency and square root of total mass per unit length, it is found that the difference between the experimental results and the calculated results is less than 1%, which proves the rationality and feasibility of the calculation method of frequency and total mass per unit length in engineering design. Through calculation, the instability constant is greater than 4 when the void fraction is less than 75%, less than 4 when the void fraction exceeds 75% and only 3.04 when the void fraction is 95%.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.566-574
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• 2023

Investigating the seepage characteristics of the leaching solution in the ore-bearing layer during the in situ leaching process can be useful for designing the process parameters for the uranium mining well. We prepared leaching solutions of four different viscosities and conducted experiments using a self-developed multifunctional uranium ore seepage test device. The effects of different viscosities of leaching solutions on the seepage characteristics of uranium-bearing sandstones were examined using seepage mechanics, physicochemical seepage theory, and dissolution erosion mechanism. Results indicated that while the seepage characteristics of various viscosities of leaching solutions were the same in rock samples with similar internal pore architectures, there were regular differences between the saturated and the unsaturated stages. In addition, the time required for the specimen to reach saturation varied with the viscosity of the leaching solution. The higher the viscosity of the solution, the slower the seepage flow from the unsaturated stage to the saturated stage. Furthermore, during the saturation stage, the seepage pressure of a leaching solution with a high viscosity was greater than that of a leaching solution with a low viscosity. However, the permeability coefficient of the high viscosity leaching solution was less than that of a low viscosity leaching solution.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2215-2221
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• 2023

In order to further meet the requirements of weight, volume, and dose minimization for new nuclear energy devices, the bare-bones multi-objective particle swarm optimization algorithm is used to automatically and iteratively optimize the design parameters of radiation shielding system material, thickness, and structure. The radiation shielding optimization program based on the bare-bones particle swarm optimization algorithm is developed and coupled into the reactor radiation shielding multi-objective intelligent optimization platform, and the code is verified by using the Savannah benchmark model. The material type and thickness of Savannah model were optimized by using the BBMOPSO algorithm to call the dose calculation code, the integrated optimized data showed that the weight decreased by 78.77%, the volume decreased by 23.10% and the dose rate decreased by 72.41% compared with the initial solution. The results show that the method can get the best radiation shielding solution that meets a lot of different goals. This shows that the method is both effective and feasible, and it makes up for the lack of manual optimization.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2572-2577
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• 2023

A field determination method for small D-T neutron source yield based on the oxygen prompt gamma rays was established. A neutron-gamma transport equation of the determination device was developed. Two yield field determination devices with a thickness of 20 mm and 50 mm were made. The count rates of the oxygen prompt gamma rays were calculated using three energy spectra processing approaches, which were the characteristic peak of 6.13 MeV, the overlapping peak of 6.92 MeV and 7.12 MeV, and the total energy area. The R-square of the calibration curve is better than 94% and the maximum error of the yield test is 5.21%, demonstrating that it is feasible to measure the yield of D-T neutron source by oxygen prompt gamma rays. Additionally, the results meet the requirements for field determination of the conventional D-T neutron source yield.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.745-752
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• 2024

A method for gamma/neutron event classification based on frequency-domain analysis for mixed radiation environments is proposed. In contrast to the traditional charge comparison method for pulse-shape discrimination, which requires baseline removal and pulse alignment, our method does not need any preprocessing of the digitized data, apart from removing saturated traces in sporadic pile-up scenarios. It also features the identification of neutron events in the detector's full energy range with a single device, from thermal neutrons to fast neutrons, including low-energy pulses, and still provides a superior figure-of-merit for classification. The proposed frequency-domain analysis consists of computing the fast Fourier transform of a triggered trace and integrating it through a simplified version of the transform magnitude components that distinguish the neutron features from those of the gamma photons. Owing to this simplification, the proposed method may be easily ported to a real-time embedded deployment based on Field-Programmable Gate Arrays or Digital Signal Processors. We target an off-the-shelf detector based on a small CLYC (Cs2LiYCl6:Ce) crystal coupled to a silicon photomultiplier with an integrated bias and preamplifier, aiming at lightweight embedded mixed radiation monitors and dosimeter applications.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.9-18
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• 2024

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents' acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristics are analyzed through BALI experiments with similar condition. In this study, computational fluid dynamics (CFD) via Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) for BALI experiments were performed. Steady-state CFD analysis was performed on three turbulence models, and SST k-ω model was in good agreement with the experimental measurement temperature within the maximum error range of 1.9%. LES CFD analysis was performed based on the RANS analysis results and it was confirmed that the temperature and wall heat flux for depth was consistent within an error range of 1.0% with BALI experiment. The LES CFD analysis results were compared with those of the Lagrangian-based solver. LES matched the temperature distribution better than SOPHIA, but SOPHIA calculated the position of boundary between stratified layer and convective layer more accurately. On the other hand, Lagrangian-based solver predicted several small eddy behaviors of the convective layer and LES predicted large vortex behavior. The vibration characteristics near the cooling part of the BALI experimental device were confirmed through Fast Fourier Transform (FFT) investigation. It was found that the power spectral density for pressure at least 10 times higher near the side cooling than near the top cooling.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1679-1686
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• 2024

Graphite material plays an important role in nuclear reactors especially the high-temperature gas-cooled reactors (HTGRs) by its outstanding comprehensive nuclear properties. The structural integrity of graphite pebble fuel elements is the first barrier to core safety under any circumstances. The correct knowledge of the stiffness coefficient of the graphite pebble fuel element inside the reactor's core is significant to ensure the valid design and inherent safety. In this research, a vertical extrusion device was set up to measure the stiffness coefficient of the graphite pebble fuel element by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. The stiffness coefficient equations of graphite pebble fuel elements at different temperatures are given (in a helium atmosphere). The result first provides the data on the high-temperature stiffness coefficient of pebbles in helium gas. The result will be helpful for the engineering safety analysis of pebble-bed nuclear reactors.

• 수산해양기술연구
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• 제40권3호
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• pp.206-213
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• 2004

종횡비, 다각형 모양에 따른 평판과 범포의 유체역학적 특성을 규명하고자 직사각형, 사다리꼴 모양으로 모형 평판과 범포를 제작하고 회류수조에서 양 ${\cdot}$ 항력 실험을 수행하였다. 그 결과를 요약하면 다음과 같다. 1. 삼각형 평판의 경우, 종횡비가 1 이하인 모형에서는 38${\sim}$42$^{\circ}$에서 최대 $C_L$이 1.23${\sim}$1.32, 1.5 이상인 모형에서는 20${\sim}$50$^{\circ}$에서 $C_L$이 약 0.85 전후였다. 역삼각형 평판의 경우, 종횡비가 1 이하인 모형에서는 영가가 36${\sim}$38$^{\circ}$에서 최대 $C_L$이 1.46${\sim}$1.56, 1.5 이상인 모형에서는 22${\sim}$26$^{\circ}$에서 1.05${\sim}$1.21 정도였다. 같은 삼각형 평판 모형에서는 전자의 모형이 후자보다 $C_L$이 작게, 양항비도 작게 나타났다. 2. 삼각형 범포의 경우, 종횡비가 1 이하인 모형에서는 영각 46${\sim}$48$^{\circ}$에서 최대 $C_L$이 1.67${\sim}$1.77, 1.5 이상인 모형에서는 20${\sim}$50$^{\circ}$에서 $C_L$이 약 1.1 전후였다. 역삼각형 범포의 경우, 종횡비가 1 이하인 모형에서는 영각 28${\sim}$32$^{\circ}$에서 최대 $C_L$이 1.44${\sim}$1.68, 1.5 이상인 모형에서는 18${\sim}$24$^{\circ}$에서 10.3${\sim}$1.18 정도였다. 같은 삼각형 범포 모형에서는 전자의 모형이 후자보다 $C_L$은 크게, 양항비는 작게 나타났다. 3. 모형에서 물의 유체력을 많이 받을 수 있는 곳에서 만곡꼭지점이 만들어지며, 삼각형 모형에서는 종횡비가 클수록, 역삼각형 모형에서는 작을수록 만곡꼭지점의 위치도 컸다. 4. 만곡도는 전 모형에서 종횡비가 클수록 컸으며, 삼각형 모형에서는 영각이 클수록 컸고 역삼각형 모형에서는 작을수록 컸다.