• Journal of Mechanical Science and Technology
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• 제14권9호
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• pp.985-996
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• 2000

The issue of quench is related to safety operation of large-scale superconducting magnet system fabricated by cable-in-conduit conductor. A numerical method is presented to simulate the thermal hydraulic quench characteristics in the superconducting Tokamak magnet system, One-dimensional fluid dynamic equations for supercritical helium and the equation of heat conduction for the conduit are used to describe the thermal hydraulic characteristics in the cable-in-conduit conductor. The high heat transfer approximation between supercritical helium and superconducting strands is taken into account due to strong heating induced flow of supercritical helium. The fully implicit time integration of upwind scheme for finite volume method is utilized to discretize the equations on the staggered mesh. The scheme of a new adaptive mesh is proposed for the moving boundary problem and the time term is discretized by the-implicit scheme. It remarkably reduces the CPU time by local linearization of coefficient and the compressible storage of the large sparse matrix of discretized equations. The discretized equations are solved by the IMSL. The numerical implement is discussed in detail. The validation of this method is demonstrated by comparison of the numerical results with those of the SARUMAN and the QUENCHER and experimental measurements.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.322-326
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• 2008

In this paper, we have studied the images which have been reconstructed by using combination of images acquired by the variation of operating frequency. When inner images have been reconstructed, they have been superposed by the surface state effect. In this case, the images of the phase object can be enhanced by the contrast of inner images. There is a kind of specimen, one is a reference block having 1/4T, 1/2T, 3/4T side drilled holes as main run piping material of the steam generator in nuclear power plants. It has been shown that the two results of defect shapes have better than before in this processing and phase contrast grow about twice. And we have constructed the acoustic microscope by using a quadrature detector that enables to acquire the amplitude and phase of the reflected signal simultaneously. Further more we have studied the reconstruction method of the amplitude and phase images, the enhancement method of the defect images' contrast.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.45-53
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• 2019

Thermal-hydraulic passive safety systems (PSSs) are incorporated into many advanced reactor designs on the bases of simplicity, economics and inherent safety nature. Several factors among which are the critical parameters (CPs) that influence failure and reliability of thermal-hydraulic (t-h) passive systems are now being explored. For simplicity, it is assumed in most reliability analyses that the CPs are independent whereas in practice this assumption is not always valid. There is need to critically examine the dependency influence of the CPs on reliability of the t-h passive systems at design stage and in operation to guarantee safety/better performance. In this paper, two multivariate analysis methods (covariance and conditional subjective probability density function) were presented and applied to a simple PSS. The methods followed a generalized procedure for evaluating t-h reliability based on dependency consideration. A passively water-cooled steam generator was used to demonstrate the dependency of the identified key CPs using the methods. The results obtained from the methods are in agreement and justified the need to consider the dependency of CPs in t-h reliability. For dependable t-h reliability, it is advisable to adopt all possible CPs and apply suitable multivariate method in dependency consideration of CPs among other factors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.149-154
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• 1998

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute(KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons from spallation reactions are essentially required for operating the reactor in its steady state. furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance of the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2332-2342
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• 2024

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2381-2384
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• 2008

The heat exchange between the Borehole Heat Exchanger(BHE) and the surrounding ground depends directly on ground thermal conductivity k at the certain site. The k is thus a key parameter in designing BHE and coupled geothermal heat pump systems. Currently, although a thermal hydraulic Response Test(TRT) is mostly used in practice, the thermal hydraulic TRT needs additional power and is generally time-consuming. A new, simple wireless probe for hi-speed k determination was introduced in this paper. This technique using a wireless probe is less time-consuming and requires no external source of energy for measurement and predicts local thermal properties by measuring soil temperatures along the depth. Measured temperature data along the depth was analyzed. As a result, the electronic wireless probe can replace the conventional hydraulic TRT method after carrying out the additional research on a lot of local heat flow, etc.

• 한국지반공학회논문집
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• 제37권12호
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• pp.25-31
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• 2021

지반의 열-수리 현상에 대한 수치해석에서 경계조건은 해석결과의 정확도에 중요한 역할을 한다. 본 연구에서는 지반과 대기의 상호작용을 고려한 열-수리 경계조건을 제시하였다. 지면의 에너지 평형은 태양복사, 지구복사, 바람에 의한 대류, 수분 증발에 대한 잠열 그리고 지중으로의 열전도로 구성된다. 각각의 열흐름에 대한 방정식을 제시하고, 불포화 지반의 열-수리 현상에 대한 해석프로그램과 연계하여 수치해석을 수행하였다. 울산기상대에서 관측된 기상데이터를 이용한 수리-열적 해석에서 실측된 지표면 온도와 수치해석 결과가 매우 유사하였다. 낮시간의 수분 증발에 의한 잠재열은 비포장 지면의 온도를 낮추며, 야간시간에는 지면조건의 영향이 감소한 열적평형 상태에 도달하였다. 지면의 온도변화는 지중으로 깊어질수록 열확산으로 감소하였다. 지표면의 온도가 주요 관심사인 수치해석에서는 지반과 대기의 열-수리적 상호작용을 고려한 수치해석을 수행해야 한다.