• Journal of Energy Engineering
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• v.21 no.4
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• pp.419-426
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• 2012

A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.306-319
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• 2020

Characterizations of Excavation Damage Zone (EDZ), which is hydro-mechanical degrading the host rock, are the important issues on the geological repository for the spent nuclear fuel. In the DECOVALEX 2019 project, Task G aimed to model the fractured rock numerically, describe the hydro-mechanical behavior of EDZ, and predict the change of the hydraulic factor during the lifetime of the geological repository. Task G prepared two-dimensional fractured rock model to compare the characteristics of each simulation tools in Work Package 1, validated the extended three-dimensional model using the TAS04 in-situ interference tests from Äspö Hard Rock Laboratory in Work Package 2, and applied the thermal and glacial loads to monitor the long-term hydro-mechanical response on the fractured rock in Work Package 3. Each modelling team adopted both Finite Element Method (FEM) and Discrete Element Method (DEM) to simulate the hydro-mechanical behavior of the fracture rock, and added the various approaches to describe the EDZ and fracture geometry which are appropriate to each simulation method. Therefore, this research can introduce a variety of numerical approaches and considerations to model the geological repository for the spent nuclear fuel in the crystalline fractured rock.

• Explosives and Blasting
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• v.41 no.3
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• pp.38-61
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• 2023

An Excavation Damaged Zone(EDZ) caused by blasting impact changes rock properties, in situ stress distribution, etc., and its effects are noticeable at around a radioactive waste repository located at deep underground. In particular, the increase in permeability due to the formation of cracks may significantly increase the amount of groundwater inflow and the possibility of radioactive nuclide outflow. In this study, FLAC2D and FLAC3D were used to analyze the mechanical and thermal behaviors for three categories: a)No EDZ, b)Uniform EDZ, and c)Random EDZ. It was found that the tunnel displacement in the Random EDZ case was 423% higher than that in the No EDZ case and was 16% higher than that in the Uniform EDZ case. Tunnel inflow in the Random EDZ was also 17.3% and 10.8% higher than that in the No EDZ and the Uniform EDZ case, respectively. The permeability around the tunnel was increased by up to 10 times in the corner of the tunnel wall and roof due to the stress redistribution after excavation. From the computer simulation, it was found that the permeability around the tunnel wall was partially increased but the overall tunnel inflow was decreased with increase of stress ratio. Mechanical analysis using FLAC 3D showed similar results. Slight difference between 2D and 3D could be explained with the development of plastic zone during the advance of tunnel excavation in 3D.

• Journal of Korean Tunnelling and Underground Space Association
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• v.21 no.5
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• pp.587-609
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• 2019

Short-and long-term stabilities of bentonite, favored material as buffer in geological repositories for high-level waste were reviewed in this paper in addition to alternative design concepts of buffer to mitigate the thermal load from decay heat of SF (Spent Fuel) and further increase the disposal efficiency. It is generally reported that the irreversible changes in structure, hydraulic behavior, and swelling capacity are produced due to temperature increase and vapor flow between $150{\sim}250^{\circ}C$. Provided that the maximum temperature of bentonite is less than $150^{\circ}C$, however, the effects of temperature on the material, structural, and mineralogical stability seems to be minor. The maximum temperature in disposal system will constrain and determine the amount of waste to be disposed per unit area and be regarded as an important design parameter influencing the availability of disposal site. Thus, it is necessary to identify the effects of high temperature on the performance of buffer and allow for the thermal constraint greater than $100^{\circ}C$. In addition, the development of high-performance EBS (Engineered Barrier System) such as composite bentonite buffer mixed with graphite or silica and multi-layered buffer (i.e., highly thermal-conductive layer or insulating layer) should be taken into account to enhance the disposal efficiency in parallel with the development of multilayer repository. This will contribute to increase of reliability and securing the acceptance of the people with regard to a high-level waste disposal.

• Journal of Energy Engineering
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• v.25 no.4
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• pp.124-132
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• 2016

In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.125-135
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• 1993

The safe and efficient operation of nuclear plants is recognized to be accomplished through the application of plant automation using digital technology, which is one of main targets of the next generation nuclear plants. For plant level automation, it is first required that each major subsystem be digitalized, and the steam generator water level control system is discussed in this study. The transfer functions between inputs and the level are derived by employing the thermal hydraulic model of the steam generator and are applied to the analysis of the current three-element control system. Since the control scheme in this study includes the steam generator itself as a process plant, the system order is high and the numerical instability arises in digitalizing. Together with this, the unreliability of the feedwater feedback signal at low power level leads to the proposal of a two-element control system with a proper digital controller. The digital PI controller developed for this system has the initial power adaptive gain and integration time constant. And it makes the overall system response satisfy the stability and other necessary control specifications simultaneously. Since the two-element control system using this controller depends on the initial power only, it is simple to define and it shows a similar level response behavior to that of its corresponding analog system.

• Journal of Energy Engineering
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• v.27 no.4
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• pp.27-35
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• 2018

The passive containment cooling system (PCCS) has been designed to remove the released decay heat during the accident by means of the condensation heat transfer phenomenon to guarantee the safety of the nuclear power plant. The heat removal performance of the PCCS is mainly governed by the condensation heat transfer of the steam-air mixture. In this study, the heat removal performance of the PCCS was evaluated by using the MARS-KS code with a new empirical correlation for steam condensation in the presence of a noncondensable gas. A new empirical correlation implemented into the MARS-KS code was developed as a function of parameters that affect the condensation heat transfer coefficient, such as the pressure, the wall subcooling, the noncondensable gas mass fraction and the aspect ratio of the condenser tube. The empirical correlation was applied to the MARS-KS code to replace the default Colburn-Hougen model. The various thermal-hydraulic parameters during the operation of the PCCS follonwing a large-break loss-of-coolant-accident were analyzed. The transient pressure behavior inside the containment from the MARS-KS with the empirical correlation was compared with calculated with the Colburn-Hougen model.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.91-101
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• 1993

To predict accurately the thermal hydraulic behavior of light water reactors during normal or abnormal operation, the accurate estimation of the void distribution is required. Up to date, many techniques for predicting void fraction of two-phase flow systems have been suggested. Among these techniques, the drift-flux model is widely used because of its exact calculation ability and simplicity. However, to get more accurate prediction of void fraction using drift-flux model, slip and flow regime effects must be considered more properly In the drift-flux method, these two effects are accounted for by two drift-flux parameters ; $C_{o}$ and (equation omitted). At earlier stage, $C_{o}$ is measured in a circular tube. In this study, $C_{o}$ is experimentally determined by measuring local void fraction and vapor velocity distribution in a rectangular subchannel having 4 heating rods which simulates nuclear subchannels. The measurements are peformed with two-electrical conductivity probes which are known to be adequate for measuring local parameters. The experiments are performed at low flow rate and the system pressure less than 3 atmo spheric pressure. In this experiment, (equation omitted), is not measured, but quoted from well-known empirical correlation to formulate $C_{o}$. Finally, $C_{o}$ is expressed as a function of channel averaged void fraction. fraction.