• 한국경영과학회:학술대회논문집
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• 한국경영과학회/대한산업공학회 2005년도 춘계공동학술대회 발표논문
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• pp.106-111
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• 2005

In nuclear power plant, the source of the energy is generated from the nuclear fuel rod. Given a certain level of consumption, the burnt fuel rod should be removed and replaced by a new(fresh) one. The burnt fuel is approximately one third of the whole fuel rods. Currently, this operation is done manually using paper documents and verbal communication and consumes a lot of operation time. In this study, we develop an computerized operation process of nuclear fuel rod replacement procedure based on the ERP(Enterprise Resource planning) methodology.

• Tribology and Lubricants
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• 제17권6호
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• pp.453-458
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• 2001

The fretting wear of a nuclear fuel rod it a dangerous phenomenon. In this study, TiN coating was used to reduce the fretting wear of Zircaloy-4 tube, a nuclear fuel rod cladding material. TiN coating is probably one of the molt frequently and successfully used PVD coatings for the mitigation of fretting wear. The fretting tester was designed and manufactured for this experiment. The number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. The results of this research showed that wear volume was improved 1.3∼3.2 times with TiN coating. The worn surfaces were observed by SEM. Wear mechanism at lower slip amplitude was the brittle cracks and rupture of TiN coating. However, adhesive and abrasive wear were mainly observed on most surfaces at higher slip amplitude.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.1819-1824
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• 2004

The common method to improve heat transfer in Nuclear fuel rod bundle is install a mixing vane in space grid. The previous split mixing vane is guides cooling water to swirl flow in sub-channel of fuel assembly. But, this swirl flow decade rapidly after mixing vane and the effect of enhancing the heat transfer vanish behind this short region. The large scale secondary vortex flow was generated by rearranging the inclined angle direction of mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about 35 $D_H$ after the spacer grid and the streamwise vorticity in subchannel with LSVF mixing vane sustain two times more than that in subchannel with split mixing vane. The turbulent kinetic energy and the Reynolds stresses generated by the mixing vanes have nearly same scales but maintain twice more than previous type.

• 설비공학논문집
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• 제18권10호
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• pp.811-818
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• 2006

Mixing vanes have been installed in the space grid of nuclear fuel rod bundle to improve turbulent heat transfer. Split mixing vanes induce the vortex flow in the cooling water to swirl in sub-channel of fuel assembly. But, The swirling flow decays rapidly so that the heat transfer enhancing effect limited to short length after the mixing vane. In the present study, the large scale vortex flow (LSVF) is generated by rearranging the mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about $35D_h$ after the spacer grid. The streamwise vorticity generated by LSVF sustain two times more than that split mixing vane.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.1592-1597
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• 2004

Experimental studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel rod bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Experimental studies were carried out at Reynolds Number 60,000 with hydraulic condition. Normal variations of mean velocity and turbulent intensity in the rod bundle subchannel were measured by the 2-color LDV measurement system. The turbulence generated by split mixing vanes has small length scales so that they maintain only about 10DH after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up $25D_{H}$ after the spacer grid.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.331-332
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• 2005

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• 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.

• Nuclear Engineering and Technology
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• 제43권5호
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• pp.447-458
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• 2011

This paper proposes an approximated contact force model to identify the nonlinear behavior of a fuel rod with gap supports; also, the numerical prediction of interfacial forces in the mechanical contact of fuel rods with gap supports is studied. The Newmark integration method requires the current status of the contact force, but the contact force is not given a priori. Taylor's expansion can be used to predict the unknown contact force; therefore, it should be guaranteed that the first derivative of the contact force is continuous. This work proposes a continuous and differentiable contact force model with the ability to estimate the current state of the contact force. An approximated convex and differentiable potential function for the contact force is described, and a variational formulation is also provided. A numerical example that considers the particularly stiff supports has been studied, and a fuel rod with hardening supports was also examined for a realistic simulation. An approximated proper solution can be obtained using the results, and abrupt changes from the contacting state to non-contacting state, or vice versa, can be relieved. It can also be seen that not only the external force but also the developed contact force affects the response.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).