• Proceedings of the Materials Research Society of Korea Conference
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• 2002.11a
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• pp.193-193
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.249.1-249
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• 1999

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.85-86
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• 2017

• Journal of the Korean Society of Visualization
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• v.18 no.3
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2756-2766
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• 2024

Small rod-controlled pressurized water reactors (PWR) are the ideal energy source for vessel propulsion, benefiting from their high reactivity control efficiency. Since the control rods (CRs) increase the complexity of reactivity control, this paper seeks to study the performance of CRs in small rod-controlled PWRs to extend the lifetime and reduce power offset due to CRs. This study investigates CR grouping, move-in/out strategies, and axially non-uniform design effects on core neutron physics metrics. These metrics include axial offset (AO), core lifetime (CL), fuel utilization (FU), and radial power peaking factor (R-PPF). To simulate the movement of the CRs, a "Critical-CR-burnup" function was developed in OpenMC. In CR designs, the CRs are grouped into three banks to study the simultaneous and prioritized move-in/out strategies. The results show CL extension from 590 effective full power days (EFPDs) to 638-698 EFPDs. A lower-worth prioritized strategy minimizes AO and the extremum values decrease from -0.69 and + 0.81 to -0.28 and + 0.51. Although an axially non-uniform CR design can improve AO at the beginning of cycle (BOC), considering the overall CR worth change is crucial, as a significant decrease can adversely impact axial power distribution during the middle of cycle (MOC).

• Journal of Welding and Joining
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• v.30 no.1
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• pp.59-63
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• 2012

The spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water and maintains geometry from external impact load and cyclic stress by the vibration of nuclear fuel rod, it is necessary to have sufficient strength against dynamic external load and fatigue strength. In this study, the mechanical properties and fatigue characteristics of laser beam welded zircaloy thin sheet are examined. The material used in this study is a zirconium alloy with 0.66 mm of thickness. The fatigue strength under cyclic load was evaluated at stress ratio R=0.1. S-N curves are presented with statistical testing method recommend by JSME- S002 and compared with S-N curves at R.T. and $315^{\circ}C$. As a result of the experimental approach, the design guide of fatigue strength is proposed and the results obtained from this study are expected to be useful data for spacer gird design.

• Journal of Mechanical Science and Technology
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• v.14 no.8
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• pp.816-822
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• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.

• Journal of the Korean Society for Precision Engineering
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• v.33 no.4
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• pp.279-286
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• 2016

To enhance the safety of nuclear reactors after the Fukushima accident, researchers are developing various types of accident tolerant fuel (ATF) to increase the coping time and reduce the generation of hydrogen by oxidation. Coated cladding, an ATF concept, can be a promising technology in view of its commercialization. We applied 3D printing technology to the fabrication of coated cladding as well as of coated pellets. Direct metal tooling (DMT) in 3D printing technologies can create a coated layer on the tubular cladding surface, which maintains stability during corrosion, creep, and wear in the reactor. A 3D laser coating apparatus was built, and parameter studies were carried out. To coat pellets with erbium using this apparatus, we undertook preliminary experiments involving metal pellets. The adhesion test showed that the coated layer can be maintained at near fracture strength.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.33-38
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• 1997

Criticality experiments with fixed neutron poison plates for water moderated and reflected low enriched(2.35 and 4.31 wt%) UO$_2$fuel rod clusters were evaluated to validate calculation techniques employed in analyzing fuel shipping and storage systems having steel, boral, or cadmium shield. Measurements were obtained for both the 2.35 wt% and the 4.31 wt% enriched rods in square pitched, water flooded lattices. The critical experiments with the 2.35 wt% enriched rods consists of three 20$\chi$ 16 or 20$\chi$ 17 fuel cluster. Critical separation were used in the experiments with the 4.31 wt% enriched fuel rods. In the experiments, the poison plates were placed on both sides of the centrally located fuel cluster. Critical separation between the three sub-critical fuel clusters were then measured for varying plate thicknesses and distances of the plates to the center fuel cluster. Calculations were performed for thirty eight critical configuration using KENO-V. a and MCNP. All of the results were within 1.23% in $\Delta$k when individually compared with the critical value of 1.0. Discrepancies of the code results are probably due to uncertainties in experiments and/or analytical modeling experiments. In general, MCNP predictions were observed to be in best agreement with the experiments.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.13-22
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• 2004

This study presents the thermal analyses of a spent fuel dry storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 15 $^{\circ}C$ under the normal condition. The off-normal condition has an environmental temperature of 38 $^{\circ}C$. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The significant thermal design feature of the storage cask is the air flow path used to remove the decay heat from the spent fuel. Natural circulation of the air inside the cask allows the concrete and fuel cladding temperatures to be maintained below the allowable values. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal and off-normal conditions.