• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.12 no.2
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• pp.97-107
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• 2002

The continuous circular cylindrical shells are widely used for the high performance structures of aircraft, spacecraft, missile, nuclear fuel rod shell and so on. In this paper, a method for the vibrational analysis of the continuous circular cylindrical shells with the clamped-clamped supports at two end edges is developed by using the modal expansion method. Forces and/or moments acting on the shell surface are expressed in terms of the Dirac Delta Function. Frequency equation of the continuous shell is also derided by the application of the equilibrium of forces and the continuity of displacements at the boundary. Natural frequencies of the continuous shell are calculated numerically with mathematica 3.0 and they are compared with FEM results from the ANSYS 5.3 to improve the reliability of analytic solutions. Mode shares obtained by the FEM are Presented in this paper.

• Nuclear Engineering and Technology
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• v.2 no.1
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• pp.19-25
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• 1970

A study was conducted on the coolant mixing between water flowing in two adjacent subchannels. Measurements were made of the quantity of mass transferred between a larger rectangular channel and a smaller triangular channel in a 19-rod fuel bundle under the conditions of single phase flow and air-water two-phase flow. The results of the experiments showed that the low mixing rate appears in single phase flow, and high mixing rate was measured in air-water two-phase flow Mixing rate decreases with the increasing of air void fraction during the air-water flow. It seems that the high mixing rate in the air-water flow was caused due to adequate agitation of the chaotic air void.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.153-162
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• 1980

The investigation of the fuel cladding temperature behavior and heat transfer mechanism during the reflooding phase of a LOCA plays an important role in performance evaluation of ECCS and safety analysis of water reactors. Reflooding experiments were performed with horizontal and vertical flow channels to investigate the effect of coolant flow channel orientation on rewetting process. Emphasis was mainly placed on the CANDU reactor which has horizontal pressure tubes in core, and the results were compared with those of vertical channel. Also to investigate the rewetting process visually, the experiments by using a rod in annulus and a quartz tube heated outside were performed. It can be concluded that the rewetting velocity in horizontal flow channel is clearly affected by flow stratification, however, the average rewetting velocity is similar to those in vertical flow channel for same conditions.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.45-57
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• 1990

A steady-state margin comparison study was performed between analog and digital protection systems. The systems compared are the thermal overpower and overtemperature delta T system of Westinghouse, and Core Protection Calculator System of Combustion Engineering, Inc. No dynamic offset was considered to eliminate the margin differences by different safety analysis methodologies. The result shows that the digital protection system has about 30% more rated power margin than the analog system in protecting against the fuel rod centerline melting. The digital protection system is shown to have almost same margin with the analog protection system in preventing the DNB at EOC (End of Cycle) even if the digital protection system has about 10% more margin at BOC(Beginning of Cycle).

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.398-406
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• 2015

In this study, R-123 flow boiling experiments were carried out to investigate the effects of nanoparticle deposition on heater surfaces on flow critical heat flux (CHF) and boiling heat transfer. It is known that CHF enhancement by nanoparticles results from porous structures that are very similar to layers of Chalk River unidentified deposit formed on nuclear fuel rod surfaces during the reactor operation period. Although previous studies have investigated the surface effects through surface modifications, most studies are limited to pool boiling conditions, and therefore, the effects of porous surfaces on flow boiling heat transfer are still unclear. In addition, there have been only few reports on suppression of wetting for decoupled approaches of reasoning. In this study, bare and $Al_2O_3$ nanoparticle-coated surfaces were prepared for the study experiments. The CHF of each surface was measured with different mass fluxes of $1,600kg/m^2s$, $1,800kg/m^2s$, $2,100kg/m^2s$, $2,400kg/m^2s$, and $2,600kg/m^2s$. The nanoparticle-coated tube showed CHF enhancement up to 17% at a mass flux of $2,400kg/m^2s$ compared with the bare tube. The factors for CHF enhancement are related to the enhanced rewetting process derived from capillary action through porous structures built-up by nanoparticles while suppressing relative wettability effects between two sample surfaces as a highly wettable R-123 refrigerant was used as a working fluid.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4066-4076
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• 2023

Pressure resistance welding is usually used to seal the connection between the cladding tube and the end plug made of zirconium alloy. The seal welded joint has a direct effect on the service performance of the fuel rod cladding structure. In this paper, the pressure resistance welded joints of zirconium alloy tube-plug structure were obtained by thermal-mechanical simulation experiments. The microstructure and microhardness of the joints were both analyzed. The effect of processing parameters on the microstructure was studied in detail. The results showed that there was no β-Zr phase observed in the joint, and no obvious element segregation. There were different types of Widmanstätten structure in the thermo-mechanically affected zone (TMAZ) and heat affected zone (HAZ) of the cladding tube and the end plug joint because of the low cooling rate. Some part of the grains in the joint grew up due to overheating. Its size was about 2.8 times that of the base metal grains. Due to the high dislocation density and texture evolution, the microhardnesses of TMAZ and HAZ were both significantly higher than that of the base metal, and the microhardness of the TMAZ was the highest. With the increasing of welding temperature, the proportion of recrystallization in TMAZ decreased, which was caused by the increasing of strain rate and dislocation annihilation.

• Journal of the Korean Society of Radiology
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• v.11 no.7
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• pp.571-578
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• 2017

Bubbles are generated by the boiling of the cooling water when an accident occurs in the reactor and then in order to measure the void fraction, the Optical Fiber Probe(OFP) and optical camera are used in thermal hydraulic safety research. However, such an optical method is not suitable for measuring the void fraction in a $17{\times}17$ array of fuel rods due to the geometrical limitations. This study was conducted as a preliminary study using x-ray system and various phantoms before applying to rod bundles. Through radiographic and tomographic experiments, the tube voltage of the x-ray generator was 130 kVp and the tube current was 1 mA. In addition, it is possible to measure the hole of 1mm in size visually through the bubble resolution phantom, and it is confirmed that the contrast is relatively decreased in the inside of the freon in the case of the contrast evaluation using the road phantom. However, we could obtain good image without distortion when reconstructing the image. Bubble generation phantom experiments were used to confirm the flow direction of the bubbles and to acquire tomography images. The image J tool was used to measure the void fraction of 18 % for a single tomography image. This study has carried out previous researches for the measurement of the bubble rate around the nuclear fuel and could be used as a basic research for continuous research.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.37-46
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• 2019

As candidates for accident-tolerant fuels, ceramic microcell fuels, which are distinguished by their peculiar microstructures, are being developed; these fuels have $UO_2$ grains surrounded by cell walls. They contribute to nuclear fuel safety by retention of fission products within the $UO_2$ pellet, reducing rod pressure and incidence of SCC failure. Cesium, a hazardous fission product in terms of amount and radioactivity, can be captured by chemical reactions with ceramic cell materials. The capture-ability of cesium therefore depends on the thermodynamics of the capturing reaction. Conversely, compositional design of cell materials should be based on thermodynamic predictions. This study proposes thermodynamic calculations to evaluate the cesium capture-ability of three ceramic microcell compositions: Si-Ti-O, Si-Cr-O and Si-Al-O. Prior to the calculations, the chemical and physical states of the cesium and the cell materials were defined. Then, the reactivity was evaluated by calculating the cesium potential (${\Delta}G_{Cs}$) and oxygen potential (${\Delta}G_{O_2}$) under simulated LWR circumstances of normal operation. Based on the results, cesium capture is expected to be spontaneous in all cell compositions, providing a basis for the compositional design of ceramic microcell fuels as well as a facile way for evaluating cesium capture.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.3
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• pp.281-287
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• 2011

We consider the uncertainty in the elastic buckling formula for a thin tube. We take into account the measurement uncertainty of Young's modulus and Poisson's ratio and the tolerance of the tube thickness and diameter. Elastic buckling must be prohibited for a thin tube such as a nuclear fuel rod that must satisfy a self-stand criterion. Since the predicted critical buckling pressure overestimated that found in the experiment, the determination of the minimum safety factor is crucial. The uncertainty in each parameter (i.e., Young's modulus, Poisson's ratio, thickness, and diameter) is mutually independent, so the safety factor is evaluated as the sum of the inverse of each uncertainty. We found that the thickness variation greatly affects the uncertainty. The minimum safety factor of a thin tube of Zirconium alloy is evaluated as 1.547 for a thickness of 0.87 mm and 3.487 for a thickness of 0.254 mm.