• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.195-196
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• 2002

Studied is fretting wear behaviour of transversely vibrating tube which is supported by springs and dimples. This simulates the fuel rod fretting due to flow-induced vibration in a nuclear reactor. The contact between spacer grid springs and fuel cladding tubes arc brought into focus in this paper. From the mechanical viewpoint, a concave contact shape of spring is considered to perform a wider distribution of the contact stress. Sliding/impacting experiments are conducted in air at room temperature with the conditions of positive contact force and gap existence to accommodate the mechanical condition between the fuel rod and the grid spring during reactor operation. It is found that wear region is separated and wear volume becomes larger as the supporting condition becomes poorer. Spring and dimple cause similar wear.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• Tribology and Lubricants
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• v.22 no.3
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• pp.144-148
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• 2006

The experimental investigation was performed to find the associated changes in characteristics of fretting wear with various water temperatures. The fretting wear tests were carried out using the zirconium alloy tubes and the grids with increasing the water temperature. The tube materials in water of $20^{\circ}C,\;50^{\circ}C\;and\;80^{\circ}C$ were tested with the applied load of 20 N and the relative amplitude of $200{\mu}m$. The worn surfaces were observed by SEM, EDX analysis and 2D surface profiler. As the water temperature increased, the wear volume was decreased, but oxide layer was increased on the worn surface. The abrasive wear mechanism was observed at water temperature of $20^{\circ}C$ and adhesive wear mechanism occurred at water temperature of $50^{\circ}C,\;80^{\circ}C$. As the water temperature increased, surface micro-hardness was decreased, but wear depth and wear width were decreased due to increasing stick phenomenon. Stick regime occurred due to the formation of oxide layer on the worn surface with increasing water temperatures

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.2
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• pp.157-166
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• 2014

A light water reactor (LWR) fuel rod consists of zirconium alloy cladding tube and uranium dioxide pellets with a slight gap between them. The modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel behavior under irradiated conditions. Many researchers have been developing fuel performance codes based on finite element method (FE) to calculate temperature, stress and strain for multidimensional analysis. The gap conductance model for multi-dimension is difficult issue in terms of convergence and nonlinearity because gap conductance is function of gap thickness which depends on mechanical analysis at each iteration step. In this paper, virtual link gap element (VLG) has been proposed to resolve convergence issue and nonlinear characteristic of multidimensional gap conductance. In terms of calculation accuracy and convergence efficiency, the proposed VLG model has been evaluated for variable cases.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 1999.06a
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• pp.303-309
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• 1999

The fretting wear behaviour of Zircaloy-4 tube used as the fuel rod cladding in PWR nuclear power plants has been investigated at the different test environment, in light water and in air as a function of slip amplitude, normal load, test duration and frequency. Zircaloy-4 tubes were used for both of oscillating and stationary specimens. A fretting wear tester was designed to be suitable for this fretting test. The wear volume and specific wear rate of Zircaloy-4 tube in water were greater than those in air under various slip amplitude. It was found that delaminate debris and surface cracks were observed at low slip amplitude and high load in water Experimental results showed that the light water accelerated the wear of Zircaloy-4 tube at low slip amplitude in fretting.

• Tribology and Lubricants
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• v.15 no.1
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• pp.83-89
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• 1999

The fretting wear behaviour of Zircaloy-4 tube used as the fuel rod cladding in PWR nuclear power plants has been investigated at the different test environment, in light water and in air as a function of slip amplitude, normal load, test duration and frequency. Zircaloy-4 tubes were used for both of oscillating and stationary specimens. A fretting wear tester was designed to be suitable for this fretting test. The wear volume and specific wear rate of Zircaloy-4 tube in water was greater than those in air under various slip amplitude. Delaminates and surface cracks were observed at low slip amplitude and high load of fretting test in water, but the traces of adhesion and plowing were observed at and above 200 Um. The water accelerates the wear of Zircaloy-4 tube at lower slip amplitude in fretting.

• Journal of the Korean Society of Safety
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• v.37 no.4
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.6
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• pp.650-657
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• 2001

This paper explains the process of designing a shielded encircling remote field eddy current (RFEC) probe to inspect nuclear fuel cladding tubes and investigates resulting signal characteristics. To force electromagnetic energy from exciter coil to penetrate into the tube, exciter coil is shielded outside by laminations of iron insulated electrically from each other. Effects of shielding and the proper operating frequency are studied by the finite element analysis and the location for sensor coil is decided. However, numerically simulated signals using the designed probe do not clearly show the defect indication when the sensor passes a defect and the other indication appeared as the exciter passes the defect is affected by the shape of shielding structure, which demonstrates that the sensor is directly affected by exciter fields. For this reason, the sensor is also shielded outside and this shielding dramatically improves signal characteristics. Numerical modeling with the finally designed probe shows very similar signal characteristics to those of inner diameter RFEC probe. That is, phase signals show almost equal sensitivity to inner diameter and outer diameter defects and the linear relationship between phase signal strength and defect depth is observed.

• Nuclear Engineering and Technology
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• v.20 no.2
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• pp.88-104
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• 1988

The FUel Rod Analysis(FURA) code is developed using two-dimensional finite element methods for axisymmetric and plane stress analysis of fuel rod. It predicts the thermal and mechanical behavior of fuel rod during normal and load follow operations. To evaluate the exact temperature distribution and the inner gas pressure, the radial deformation of pellet and clad, the fission gas release are considered over the full-length of fuel rod. The thermal element equation is derived using Galerkin's techniques. The displacement element equation is derived using the principle of virtual works. The mechanical analysis can accommodate various components of strain: elastic, plastic, creep and thermal strain as well as strain due to swelling, relocation and densification. The 4-node quadratic isoparametric elements are adopted, and the geometric model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The pellet cracking and crack healing, pellet-cladding interaction are modelled. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behavior accurately and stably. The pellet and cladding model has been compared with both analytical solutions and experimental results. The observed and predicted results are in good agreement. The general behavior of fuel rod is calculated by axisymmetric system and the cladding behavior against radial crack is used by plane stress system. The sensitivity of strain aging of PWR fuel cladding tube due to load following is evaluated in terms of linear power, load cycle frequency and amplitude.