• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.256-266
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• 1985

The loss of coolant flow accident is analyzed for the pressurized water reactor of Korea Nuclear Unit-1. The loss of coolant flow accident is classified into three types in accordance with its severity; partial loss of coolant flow, complete loss of coolant flow and pump locked rotor accident. Analysis has been carried out in three stages; system transient and average core analysis, DNBR calculation and hot spot analysis. The purpose of developing KTRAN is to simulate the transient fast. For the DNBR calculation, the thermal hydraulic codes, SCAN and COBRA IV-1, are adopted. And for the hot spot analysis, the fuel thermal transient code LTRAN is employed. This code system should be fast responding to the transient analysis. In case the transient occurs, severity comes within a couple of seconds. So response should be fast to accomodate the following sequence of the accident. Unfortunately this purpose could not be achieved by KTRAN. However, the calculated results are well comparable with FSAR results in range. Thereby, the effectiveness of KTRAN code analysis in this type of accident is proven.

• Journal of Radiation Protection and Research
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• v.30 no.2
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• pp.61-67
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• 2005

The criticality analysis method using HELIOS-MASTER code system, which is the nuclear core analysis code system, was developed for the spent fuel storage of SMART-P reactor. We generated the macroscopic cross section of the geometric model with HELIOS and estimated the criticality of the 3-dimensional model with MASTER for SMART-P spent fuel storage. The validity of criticality analysis method for SMART-P spent fuel storage with the HELIOS-MASTER code system by 3-D MCNP calculation was also verified. The result of the criticality analysis with the HELIOS-MASTER code system is more conservative than that with the MCNP and the accuracy of this result is within the range of an allowable error. Because HELIOS-MASTER can perform the 3-D depletion calculation lot a spent fuel storage, it will be useful to perform the criticality analysis including a burnup credit in future.

• Nuclear Engineering and Technology
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• v.11 no.1
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• pp.29-45
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• 1979

In this paper. nuclear characteristics of TRIGA Mark-III has been analyzed in detail for six different fuel options. Presently, 70 w/o enriched FLIP fuels are adopted for TRIGA core to improve fuel lifetime. However, such highly enriched fuels are not easily obtained due to nonproliferation treaty. This research examines the possible substitution for FLIP fuels with high density fuels without reducing the nuclear performance. This work will provide long-time plan for TRIGA operation.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.04a
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2005.11a
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• pp.167-168
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• 2005

The operational characteristics of the integrated three-phase flux-lock type superconducting fault current limiter (SFCL) were analyzed. The suggested three-phase SFCL consisted of a three-phase flux-lock reactor and three high-Tc superconducting (HTSC) elements. The former has three windings wound on an iron core, each of which has the same turn's ratio between coil 1 and coil 2. The latter are connected in series with coil 2 of each phase. The integrated three-phase flux-lock type SFCL showed the operational characteristics that the fault phase could affect the sound phase, which resulted in quenching the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the integrated three-phase flux-lock type SFCL according to the ground fault types were compared.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2422-2430
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• 2020

In order to improve calculational efficiency of the CAPP code in the analysis of the hexagonal reactor core, we have tried to implement a refined AFEN method with transverse gradient basis functions and interface flux moments in the hexagonal geometry. The numerical scheme for the refined AFEN method adopted here is the response matrix method that uses the interface partial currents as nodal unknowns instead of the interface fluxes used in the original AFEN method. Since the response matrix method is single-node based, it has good properties such as good calculational efficiency and parallel computing affinity. Because a refined AFEN method equivalent nonlinear FDM response matrix method tried first could not provide a numerically stable solution, a direct formulation of the refined AFEN response matrix were developed. To show the numerical performance of this response matrix method against the original AFEN method, the numerical error analyses were performed for several benchmark problems including the VVER-440 LWR benchmark problem and the MHTGR-350 HTGR benchmark problem. The results showed a more than three times speedup in computing time for the LWR and HTGR benchmark problems due to good convergence and excellent calculational efficiency of the refined AFEN response matrix method.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2006.12a
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• pp.112-113
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• 2006

The operational characteristics of the integrated three-phase flux-lock type superconducting fault current limiter (SFCL) were analyzed. The suggested three-phase SFCL consisted of a three-phase flux-lock reactor and three high-$T_c$ superconducting (HTSC) elements. The former has three windings wound on an iron core, each of which has the same turn's ratio between coil 1 and coil 2. The latter are connected in series with coil 2 of each phase. The integrated three-phase flux-lock type SFCL showed the operational characteristics that the fault phase could affect the sound phase, which resulted in quenching the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the integrated three-phase flux-lock type SFCL according to the ground fault types were compared.

• 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.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.20 no.3
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• pp.215-222
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• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.