• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3930-3943
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• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1487-1503
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• 2019

The methods and performance of a pin-level nuclear reactor core thermal-hydraulics (T/H) code ESCOT employing the drift-flux model are presented. This code aims at providing an accurate yet fast core thermal-hydraulics solution capability to high-fidelity multiphysics core analysis systems targeting massively parallel computing platforms. The four equation drift-flux model is adopted for two-phase calculations, and numerical solutions are obtained by applying the Finite Volume Method (FVM) and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE)-like algorithm in a staggered grid system. Constitutive models involving turbulent mixing, pressure drop, and vapor generation are employed to simulate key phenomena in subchannel-scale analyses. ESCOT is parallelized by a domain decomposition scheme that involves both radial and axial decomposition to enable highly parallelized execution. The ESCOT solutions are validated through the applications to various experiments which include CNEN $4{\times}4$, Weiss et al. two assemblies, PNNL $2{\times}6$, RPI $2{\times}2$ air-water, and PSBT covering single/two-phase and unheated/heated conditions. The parameters of interest for validation include various flow characteristics such as turbulent mixing, spacer grid pressure drop, cross-flow, reverse flow, buoyancy effect, void drift, and bubble generation. For all the validation tests, ESCOT shows good agreements with measured data in the extent comparable to those of other subchannel-scale codes: COBRA-TF, MATRA and/or CUPID. The execution performance is examined with a mini-sized whole core consisting of 89 fuel assemblies and for an OPR1000 core. It turns out that it is about 1.5 times faster than a subchannel code based on the two-fluid three field model and the axial domain decomposition scheme works as well as the radial one yielding a steady-state solution for the OPR1000 core within 30 s with 104 processors.

• 한국의류산업학회지
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• 제11권4호
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• pp.661-666
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• 2009

The purpose of this study is to analyze the dyeability of polyester(PET) fabric by thermal bonding with low melting component of bicomponent fiber and to describe the change of physical properties of thermal bonded PET fabrics. The PET fabrics were prepared with regular PET fiber as warp and bicomponent fiber as weft. The bicomponent fiber of sheath-core type was composed with a regular PET core and low melting PET sheath. The thermal bonding of PET fabric was carried out in pin tenter($195^{\circ}C$) for 60 seconds. In this study, we investigated the dye ability and fastness of the dyed PET fabric. Dye ability of E-type dyestuff is higher than S-type dyestuff. In the case of E- type dyestuff, the saturated dyeing time was 10minutes at $130^{\circ}C$. The washing fastness and light fastness were excellent as 4-5grade.

• Nuclear Engineering and Technology
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• 제34권3호
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• pp.250-258
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• 2002

This Paper is to apply the well-established coarse mesh finite difference(CMFD) method to the method of characteristics(MOC) transport calculation as an acceleration scheme. The CMFD problem is first formulated at the pin-cell level with the multi-group structure To solve the cell- based multi-group CMFD problem efficiently, a two-group CMFD formulation is also derived from the multi-group CMFD formulation. The performance of the CMFD acceleration is examined for three test problems with different sizes including a realistic quarter core PWR problem. The CMFD formulation provides a significant reduction in the number of ray tracings and thus only about 9 ray tracing iterations are enough for the realistic problem. In computing time, the CMFD accelerated case is about two or three times faster than the coarse-mesh rebalancing(CMR) accelerated case.

• 한국산업융합학회 논문집
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• 제23권4_2호
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• pp.661-667
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• 2020

The optimal shape modeling of core parts through 3D modeling and structural analysis for the development of small and medium-sized ships. The goal is to improve the efficient structure of the hydraulic system for controlling the rudder among the core steering parts in small and medium-sized ships. Through 3D modeling and structural analysis, a new concept of tiller parts and a double-acting hydraulic cylinder control system were proposed and operational structural stability was evaluated. Structural analysis of the three different tiller designs that can be replaceable onto existing fishing vessels was conducted to derive the final shapes. The emphasis was placed on evaluating the structural stability of the key drive components, the tiller, pin, and cylinder rodin the maximum torque condition of the hydraulic cylinder.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1871-1885
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• 2019

The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic P_N scattering, and the multi-group cross-section generation model was performed on the k_eff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a k_eff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.803-810
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• 2022

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.751-757
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• 2016

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.237-246
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• 1994

KMRR 노심의 열수력 설계 및 안전 해석을 위한 고유량 CHF 상관식을 개발하였다. 상관식 개발에는 fin이 부착된 경우와 부착되지 않은 경우의 가열봉에 대한 단일봉 CHF 실험 자료를 사용하였다. 상관식은 Reynolds 수, 열역학 평형 건도, 액상과 기상의 밀도비 및 등가수력 직경비 등의 무차원 변수로 이루어져 있으며, fin이 있는 경우와 없는 경우의 가열 둘레비를 사용하여 fin의 영향을 고려하였다. 이처럼 단일봉에 대하여 개발된 상관식의 KMRR 핵연료 집합체에 대한 적용 타당성을 보이기 위하여 비가열면 및 축방향 비균일 출력 분포의 영향을 논의하였다. CHF 실험 자료 분석 결과, 상관식 한계 DNBR은 1.44로 결정되었다.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.