• Nuclear Engineering and Technology
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• 제51권1호
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• pp.284-292
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• 2019

For the continuous operation of a nuclear reactor, burnt fuel needs to be replaced with fresh fuel, where appropriate (ex-vessel) fuel handling is required. Particularly for the Sodium-cooled Fast Reactor (SFR) refueling, its process has unique challenges due to liquid sodium coolant. The ex-vessel spent fuel transportation should concern several design features such as the radiation shielding, decay-heat removal, and inert space separated from air. This paper proposes a new design optimization methodology of cask shielding to transport the spent fuel assembly in a prototype SFR for the first time. The Particle Swarm Optimization (PSO) algorithm had been applied to design trade-offs between shielding and cask weight. The cask is designed as a double-cylinder structure to block an inert sodium region from the air-cooling space. The PSO process yielded the optimum shielding thickness of 26 cm, considering the weight as well. To confirm the shielding performance, the radiation dose of spent fuel removed at its peak burnup and after 1-year cooling was calculated. Two different fuel positions located during transportation were also investigated to consider a functional disorder in a cask drive system. This study concludes the current cask design in normal operations is satisfactory in accordance with regulatory rules.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.1140-1151
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• 2023

Fully Ceramic Microencapsulated (FCM) fuel is emerging advanced fuel material for the future nuclear reactors. The fuel pellet in the FCM fuel is composed of matrix and a large number of TRistructural-ISOtopic (TRISO) fuel particles which are randomly dispersed in the SiC matrix. The minimum layer thickness in a TRISO fuel particle is on the order of 10^-5 m, and the length of the FCM pellet is on the order of 10^-2 m. Hence, the heat transfer in the FCM pellet is a multi-scale phenomenon. In this study, three multi-scale heat conduction models including the Multi-region Layered (ML) model, Multi-region Non-layered (MN) model and Homogeneous model for FCM pellet were constructed. In the ML model, the random distributed TRISO fuel particles and coating layers are completely built. While the TRISO fuel particles with coating layers are homogenized in the MN model and the whole fuel pellet is taken as the homogenous material in the Homogeneous model. Taking the results by the ML model as the benchmark, the abilities of the MN model and Homogenous model to predict the maximum and average temperature were discussed. It was found that the MN model and the Homogenous model greatly underestimate the temperature of TRISO fuel particles. The reason is mainly that the conventional equivalent thermal conductivity (ETC) models do not take the internal heat source into account and are not suitable for the TRISO fuel particle. Then the improved ETCs considering internal heat source were derived. With the improved ETCs, the MN model is able to capture the peak temperature as well as the average temperature at a wide range of the linear powers (165 W/cm~ 415 W/cm) and the packing fractions (20%-50%). With the improved ETCs, the Homogenous model is better to predict the average temperature at different linear powers and packing fractions, and able to predict the peak temperature at high packing fractions (45%-50%).

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.90-95
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• 1997

In this work, linear dynamics of a circulating fluid-fueled subcritical reactor system with temperature feedback and external neutron source was modeled and examined. In a circulating fluid-fuel system, the stable region is slightly moved by a circulation fluid effect. The effects of subcriticality and temperature feedback coefficient on the reactor stability were tested by calculating frequency response of neutron density originated from reactivity perturbation or external source oscillation of system. The amplitude transfer function has a different shape near subcritical region due to the exponential term in the transfer function. The results of the study show that at a slightly subcritical region, low frequency oscillation in accelerator current or reactivity can be amplified depending on the temperature feedback. However, as the subcriticality increases, the oscillation becomes negligible regardless of the magnitude of the temperature feedback coefficient.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.908-917
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• 2020

Corrosion of structural materials presents a critical challenge in the use of lead-bismuth eutectic (LBE) as a nuclear coolant in an accelerator-driven system. By forming a protective layer on the steel surfaces, corrosion of steels in LBE cooled reactors can be mitigated. The amount of oxygen concentration required to create a continuous and stable oxide layer on steel surfaces is related to the oxidation process. So far, there is no oxidation experiment in fuel assemblies (FA), let alone specific oxidation detail information. This information can be, however, obtained by numerical simulation. In the present study, a new coupling method is developed to implement a coupling between the oxygen mass transfer model and the commercial computational fluid dynamics (CFD) software ANSYS-CFX. The coupling approach is verified. Using the coupling tool, we study the oxidation process of the FA and investigate the effects of different inlet parameters, such as temperature, flow rate on the mass transfer process.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4798-4808
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• 2022

The porous corrosion deposits (known as CRUD) adhered to the cladding have an important effect on the heat transfer from fuel rods to coolant in PWRs. The vapor film is the main constituent in the two-phase film boiling model. This paper presents a vapor film thickness correlation, associated with CRUD porosity, CRUD chimney density, CRUD particle size, CRUD thickness and heat flux. The dependences of the vapor film thickness on the various influential factors can be intuitively reflected from this vapor film thickness correlation. The temperature, pressure, and boric acid concentration distributions in CRUD can be well predicted using the two-phase film boiling model coupled with the vapor film thickness correlation. It suggests that the vapor thickness correlation can estimate the vapor film thickness more conveniently than the previously reported vapor thickness calculation methods.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.268-273
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• 2016

An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

• 한국유체기계학회 논문집
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• 제19권4호
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• pp.19-28
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• 2016

Objective of this study is to compare the inherent characteristics of natural convection flow inside the canister of spent fuel dry storage system with different backfill gases by utilizing computational fluid dynamics (CFD) code. Four working fluids were selected for comparison study. Helium currently used backfill gas for canister, air, nitrogen, and argon are frequently used as coolant in many heat transfer applications. The results indicate that helium has very distinct conductive behavior and show very weak natural convective flow compared to the others. Argon showed the strongest natural convective flow but also the worst coolability. Air and nitrogen showed similar characteristics to each other. However, due to difference in Prandtl number, nitrogen showed more effective natural convective flow. These results suggest that experimental validation for the nitrogen is needed to investigate the potential coolability other than currently commercially used helium.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.161-167
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• 1994

본 논문은 비상노심냉각계통을 필요로 하지 않는 압력관형 피동 수냉각로 개념을 제시한다. 여기서는 사고시 핵연료에서 생성되는 열을 감속재로 효과적으로 전달시키기 위해 금속 핵연료 매트릭스를 사용하는 핵연료 채널을 채택한다. 정상 운전시에는 보통의 냉각재가 핵연료를 냉각시키지만, 냉각재상실사고를 포함하여 정상적인 냉각계통의 작동이 이루어지지 않을 경우에는 피동 감속재냉각계통에 의해 핵연료가 냉각된다. 유한요소 코드를 이용한 해석 결과, 정상 상태 및 사고시 핵연료 온도를 허용 한도 이내로 유지할 수 있는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1914-1928
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• 2022

To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.