• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.10
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.493-498
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• 2002

The letdown system of pressurized water reactor (PWR) nuclear fewer plants had experienced instabilities in letdown system due to unacceptable flow characteristics of control valves. The Korean Standard Nuclear Power Plants (KSNPs) have three flow paths in parallel for letdown new control. Each flow path consists of two offices and one isolation valve. This study evaluates the effect of orifice arrangement and valve stroke time of letdown isolation valve on the system transients because sudden flow changes due to valve actuation can generate high pressure peaks in letdown line. A pressure transient analysis has been preformed to evaluate the impact of dynamic transients. This analysis uses MMS which is a simulation code developed by EPRI based on the method of characteristics. The result shows that the pressure peak is reduced in the continuous arrangement but negligible. Additionally, it shows that the stroke time of linear type flog valve greater than 15 seconds can give more stable performance.

• Analytical Science and Technology
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• v.17 no.5
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• pp.438-442
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• 2004

To determine the contents of $^{99}Tc$ in the spent PWR (pressurized water reactor) nuclear fuels by ICP-MS (inductively coupled plasma-mass spectrometry), a technetium separation method using an extraction chromatographic resin (TEVA Spec resin) has been established. $^{99}Tc$ was separated from a spent PWR nuclear fuel solution by this separation procedure and its concentration was determined by ICP-MS. The result agrees well with the value calculated by the program ORIGEN 2 and also the value measured by AG MP-1 resin/ICP-MS method described in our previous paper. It can be concluded that the present separation procedure is superior to the AG MP-1 resin procedure with respect to the time required for technetium separation as well as the efficiency of decontamination from other radioactive nuclides.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.40 no.10
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• pp.901-909
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• 2012

This paper describes characteristics of solid-state $NaBH_4$ hydrogen generation and supply system for fuel cell UAV. Flow rate and pressure of the generated hydrogen were dramatically changed during $NaBH_4$ decomposition using acid. Hydrogen supply was stabilized by a self-pressurized reactor, and hydrogen stabilization method was introduced. For hydrogen generation in below zero-temperature, hydrochloric acid was diluted by propylene glycol-water mixtures. Solid-state $NaBH_4$hydrogen generation and supply system was designed. Basic operation experiments was performed to reveal the characteristics of this hydrogen generation system.

• Nuclear Engineering and Technology
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• v.12 no.2
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• pp.127-134
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• 1980

Shielding problems associated with neutron streaming through the reactor vessel cavity of pressurized water reactors are discussed to a certain extent with the actual examples in the currently operating reactors. Various remedial techniques are proposed herein to mitigate the tedious neutron streaming phenomena including piling up in heaps of temporary boron-containing bags and the installation of permanent shield structure making use of a certain refractory materials. In conclusion, optimum cavity shielding design concepts are presented with special emphasis on such major factors as the identification of major neutron streaming path, selection of necessary shielding materials with acceptable constraints, detailed design characteristics and physical configuration as well as the formulation of dependable mathematical tools to predict the final outcome of each design concept proposed in the context.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.344-363
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• 1999

A multi-dimensional thermal-hydraulic system analysis code, MARS 1.3.1, has been developed in order to have the realistic analysis capability of two-phase thermal-hydraulic transients for pressurized water reactor (PWR) plants. As the backbones for the MARS code, the RELAP5/MOD3.2.1.2 and COBRA-TF codes were adopted in order to take advantages of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the MARS code, all the functional modules of the two codes were unified into a single code first. Then, the source codes were converted into the standard Fortran 90, and then they were restructured using a modular data structure based on "derived type variables" and a new "dynamic memory allocation" scheme. In addition, the Windows features were implemented to improve user friendliness. This paper presents the developmental work of the MARS version 1.3.1 including the hydrodynamic model unification, the heat structure coupling, the code restructuring and modernization, and their verifications.their verifications.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• Transactions of the Korean Society of Mechanical Engineers
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• v.8 no.5
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• pp.416-424
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• 1984

A numerical and experimental investigation was carried out on the single-phase natural circulation system. This study is concerned with the multiloop system which is relevant to the primary system of the pressurized water reactor. For numerical analysis, five time-dependent governing equations were derived using the one-dimensional lumped parameter model. These equations were discretized by the space-time integration technique, and a simplified computer program, SIMFARS, was developed to solve those discretized equations. Experiments were performed for two purposes-one is to validate the developed code, and the other is to understand the qualitative behavior of the natural circulation loop. Comparison of the computational results with experiments, and several experimental and numerical results are presented in this article.

• Journal of computational fluids engineering
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• v.20 no.1
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• pp.39-46
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• 2015

In this research, we studied the pressure drop affecting on the internal surface roughness and the curvature radius of a U-tube, which is used for the cooling system in PWR(Pressurized Water Reactor). Using ANSYS-FLUENT, a commercial code based on CFD(Computational Fluid Dynamics) technique, we compared a Moody chart with the Darcy friction factor changed by a range of various surface roughness and Reynolds numbers of a straight pipe model. We studied the effect giving variation about a range of various surface roughness and the curvature radius of the full scale U-tube model. The material of the heat transfer tube is Inconel 690 used in the steam generator. We compared the velocity distribution of selected 4 locations, and derived the correlation between the surface roughness and the pressure drop for the U-tube of each representative curvature radius using the linear regression method.

• Journal of the Korean Society for Precision Engineering
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• v.32 no.9
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• pp.815-824
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• 2015

In this study, we conducted the approximate multi-objective optimization of gap sizes of pressurized-water reactor (PWR) annular fuels. To determine the contacting tendency of the inner-outer gaps between the annular fuel pellets and cladding, thermoelastic-plastic-creep (TEPC)analysis of PWR annular fuels was performed, using in-house FE code. For the efficient heat transfer at certain levels of stress, we investigated the tensile, compressive hoop stress and temperature, and optimized the gap sizes using the non-dominant sorting genetic algorithm (NSGA-II). For this, response surface models of objective and constraint functions were generated, using central composite (CCD) and D-optimal design. The accuracy of approximate models was evaluated through $R^2$ value. The obtained optimal solutions by NSGA-II were verified through the TEPC analysis, and we compared the obtained optimum solutions and generated errors from the CCD and D-optimal design. We observed that optimum solutions differ, according to design of experiments (DOE) method.