• Journal of the Korean Society of Radiology
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• v.11 no.7
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• pp.571-578
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• 2017

Bubbles are generated by the boiling of the cooling water when an accident occurs in the reactor and then in order to measure the void fraction, the Optical Fiber Probe(OFP) and optical camera are used in thermal hydraulic safety research. However, such an optical method is not suitable for measuring the void fraction in a $17{\times}17$ array of fuel rods due to the geometrical limitations. This study was conducted as a preliminary study using x-ray system and various phantoms before applying to rod bundles. Through radiographic and tomographic experiments, the tube voltage of the x-ray generator was 130 kVp and the tube current was 1 mA. In addition, it is possible to measure the hole of 1mm in size visually through the bubble resolution phantom, and it is confirmed that the contrast is relatively decreased in the inside of the freon in the case of the contrast evaluation using the road phantom. However, we could obtain good image without distortion when reconstructing the image. Bubble generation phantom experiments were used to confirm the flow direction of the bubbles and to acquire tomography images. The image J tool was used to measure the void fraction of 18 % for a single tomography image. This study has carried out previous researches for the measurement of the bubble rate around the nuclear fuel and could be used as a basic research for continuous research.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

• 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.

• Journal of the Korean Society of Marine Environment & Safety
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• v.20 no.3
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• pp.304-311
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• 2014

In this study, to apply hydrogen energy to ship engine and to generate effective hydrogen production, we investigated the effects of high temperature $H_2SO_4$ feed rate and cooling water rate to pump parts with fixed frequency needed to reciprocate motion and a simulation was conducted at each condition. In the fixed frequency and cooling water inlet flow rate of 0.5 Hz and 3.9 kg/s, we changed the high temperature $H_2SO_4$ flow rate to 47.46 kg/s (it is 105 % of 45.2 kg/s), 49.72 kg/s (110 %), and 51.98 kg/s (115 %). Also, at 0.5 Hz and 45.2 kg/s of frequency and high temperature $H_2SO_4$ flow, the thermal hydraulic analysis was performed at the condition of 95 % (3.705 kg/s), 90 % (3.51 kg/s), and 85 % (3.315 kg/s). In overall simulation cases, the physical properties of materials are more influential to the temperature increase in the pump part rather than the changes on the feed rate of high temperature $H_2SO_4$ and cooling water. A continuous operation of pump was also capable even if the excess feed of high temperature $H_2SO_4$ of about 15 % or the less feed of cooling water of about 15 % were performed, respectively. When the increasing feed of high temperature $H_2SO_4$ of up to 5 %, 10 %, and 15 % were compared with base flow (45.2 kg/s), the deviation of time period rose to a certain temperature and ranged from 0 to 4.5 s in the same position (same material). In case of cooling water, the deviation of time period rose to a certain temperature and ranged from 0 to 5.9 s according to the decreasing feed changes of cooling water at 5 %, 10 %, and 15 % compared to a base flow (3.9 kg/s). Finally, the additional researches related to the two different materials (Teflon and STS for Pitch and End-plate), which are concerned about the effects of temperature changes to the parts contacting different materials, are needed, and we have a plan to conduct a follow-up study.

• 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.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.54-70
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• 2017

The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility-the advanced core-cooling mechanism experiment (ACME)-was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups-a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break-were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative effects on the passive core cooling performance caused by nitrogen injection during the SBLOCA transient.

• Journal of Energy Engineering
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• v.25 no.2
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• pp.1-20
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• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

• Journal of Environmental Impact Assessment
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• v.28 no.6
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• pp.507-524
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• 2019

The purpose of this study was to validate the EFDC model for the weir pool of Gangjeong-Goryong Weir located in Nakdong River, and evaluate the effect of flow control and phosphate loading reduction on the water quality and algae biomass by group (Diatom, Green, Cyanobacteria). As a result of model validation using 2018 experimental data,the time series of water level and vertical distribution of water temperature, DO, organic matter, nitrogen, and phosphorus time series were properly simulated. Seasonal fluctuations of algae biomass by group were adequately reproduced, but the deviations between measured and simulated values were significant in some periods. As a result of scenario simulations to control the water level and flow rate, the thermal stratification was resolved as the water level was lowered and the flow rate increased. The flow velocity at which the water temperature stratification was resolved was about 0.1 m/s, which is consistent with the previous study results of Baekje Weir in Geum River. Simulations of the 2Q flow scenario showed that Chl-a decreased by 8.7% and the cell density of diatom and green algae declined. The cell density of cyanobacteria increased, however, because the high concentrations of cyanobacteria in the upstream boundary conditions directly affected downstream due to increased flow velocity. In the scenario simulation of reducing the influent phosphate load concentration (average 0.056 mg/L) to 50%, Chl-a decreased by 13.6%.The results suggest that the upstream algae concentration and phosphorus load reduction should be considered simultaneously with hydraulic control to prevent algal overgrowth of Gangjeong-Goryong Weir.

• Tunnel and Underground Space
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• v.21 no.5
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• pp.394-405
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• 2011

In this paper, we applied coupled non-isothermal, multiphase fluid flow and geomechanical numerical modeling using TOUGH-FLAC coupled analysis to study the complex thermodynamic and geomechanical performance of underground lined rock caverns (LRC) for compressed air energy storage (CAES). Mechanical stress in concrete linings as well as pressure and temperature within a storage cavern were examined during initial and long-term operation of the storage cavern for CAES. Our geomechanical analysis showed that effective stresses could decrease due to air penetration pressure, and tangential tensile stress could develop in the linings as a result of the air pressure exerted on the inner surface of the lining, which would result in tensile fracturing. According to the simulation in which the tensile tangential stresses resulted in radial cracks, increment of linings' permeability and air leakage though the linings, tensile fracturing occurred at the top and at the side wall of the cavern, and the permeability could increase to $5.0{\times}10^{-13}m^2$ from initially prescribed $10{\times}10^{-20}m^2$. However, this air leakage was minor (about 0.02% of the daily air injection rate) and did not significantly impact the overall storage pressure that was kept constant thanks to sufficiently air tight surrounding rocks, which supports the validity of the concrete-lined underground caverns for CAES.

• 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.