• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.811-824
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• 1995

The CANDU element bowing is attributed to actions of both the thermally induced bending moments and the bending moment due to hydraulic drag and mechanical loads, where the bowing is defined as the lateral deflection of an element from the axial centerline. This paper consider only the thermally-induced bending moments which are generated both within the sheath and the fuel and sheath by an asymmetric temperature distribution with respect to the axis of an element The generalized and explicit analytical formula for the thermally-induced bending is presented in con-sideration of 1) bending of an empty tube treated by neglecting the fuel/sheath mechanical interaction and 2) fuel/sheath interaction due to the pellet and sheath temperature variations, where in each case the temperature asymmetries in sheath are modelled to be caused by the combined effects of (i) non-uniform coolant temperature due to imperfect coolant mixing, (ii) variable sheath/coolant heat transfer coefficient, (iii) asymmetric heat generation due to neutron flux gradients across an element and so as to inclusively cover the uniform temperature distributions within the fuel and sheath with respect to the axial centerline. As the results of the sensitivity calculations of the element bowing with the variations of the parameters in the formula, it is found that the element bowing is greatly affected relatively with the variations or changes of element length, sheath inside diameter, average coolant temperature and its variation factor, pellet/sheath mechanical interaction factor, neutron flux depression factor, pellet thermal expansion coefficient, pellet/sheath heat transfer coefficient in comparison with those of other parameters such as sheath thickness, film heat transfer coefficient, sheath thermal expansion coefficient and sheath and pellet thermal conductivities.

• 한국전산유체공학회:학술대회논문집
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• 2011.05a
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• pp.269-274
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• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.124-129
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• 1998

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.186-195
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• 1982

The steam line break accident for Kori Unit 1 is analyzed by a code SYSRAN which calculates nuclear power and heat flux using the point kinetics equation and the lumped-parameter model and calculates system transient using the mass and energy balance equation with the assumption of uniform reactor coolant system pressure. The 1.4 f $t^2$ steam line break accident is analyzed at EOL (End of Life), hot shutdown condition in which case the accident would be most severe. The steam discharge rate is assumed to follow the Moody critical flow model. The results reveal the peak heat flux of 38% of nominal full power value at 60 second after the accident initiates, which is higher than the FSAR result of 26%. Trends for the transient are in good agreement with FSAR results. A sensitivity study shows that this accident is most sensitive to the moderator density coefficient and the lower plenum mixing factor. The DNBR calculation under the assumption of $F_{{\Delta}H}$=3.66, which is used in the FSAR with all the control and the shutdown assemblies inserted except one B bank assembly and of Fz=1.55 shows that minimum DNBR reaches 1.62 at 60 second, indicating that the fuel failure is not anticipated to occur. The point kinetics equation, the lumped-parameter model and the system transient model which uses the mass and energy balance equation are verified to be effective to follow the system transient phenomena of the nuclear power plants.lear power plants.