• Journal of Energy Engineering
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• v.12 no.4
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• pp.259-266
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• 2003

Load reduction ring (LRR) was installed on the ABB-Atom sparger to reduce the oscillatory loadings due to the air bubble clouds in the water pool in case of safety relief system operations. In order to investigate the effect of LRR on the pressure field, a numerical simulation on the behavior of air bubble clouds discharging into a water pool through a ABB-Atom sparser without LRR was performed by using a commercial thermal hydraulic analysis code, FLUENT 4.5. Among the multi-phase models contained in the code, the VOF (Volume Of Fluid) model was used to simulate the interface of water, air and steam flows. By comparing the analysis results with the previous ones, the load reduction ring has an effect on reducing the oscillatory loads at the wall. It also includes the effect of air mass and inlet boundary conditions of the pipe on the pressure oscillations at the wall.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.359-381
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• 2020

Within the framework of DECOVALEX-2019 Task D, full-scale engineered barriers experiment (FEBEX) at Grimsel Test Site was numerically simulated to investigate an applicability of implemented Barcelona basic model (BBM) into TOUGH2-MP/FLAC3D simulator, which was developed for the prediction of the coupled thermo-hydro-mechanical behavior of bentonite buffer. And the calculated heater power, temperature, relative humidity, total stress, saturation, water content and dry density were compared with in situ data monitored in the various sections. In general, the calculated heater power and temperature provided a fairly good agreement with experimental observations, however, the difference between power of heater #1 and that of heater #2 could not captured in the numerical analysis. It is necessary to consider lamprophyre with low thermal conductivity around heater #1 and non-simplified installation progresses of bentonite blocks in the tunnel for better modeling results. The evolutions and distributions of relative humidity were well reproduced, but hydraulic model needs to be modified because the re-saturation process was relatively fast near the heaters. In case of stress evolutions due to the thermal and hydraulic expansions, the computed stress was in good agreement with the data. But, the stress is slightly higher than the measured in situ data at the early stage of the operation, because gap between rock mass and bentonite blocks have not been considered in the numerical simulations. The calculated distribution of saturation, water content, and dry density along the radial distance showed good agreement with the observations after the first and final dismantling. The calculated dry density near the center of the FEBEX tunnel and heaters were overestimated compared with the observations. As a result, the saturation and water content were underestimated with the measurements. Therefore, numerical model of permeability is needed to modify for the production of better numerical results. It will be possible to produce the better analysis results and more realistically predict the coupled THM behavior in the bentonite blocks by performing the additional studies and modifying the numerical model based on the results of this study.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.131-140
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• 2015

This research studied a scaling analysis for the selection of proper heat generation at tube for 1/4-scale storage vaults. First of all, the temperature field and velocity distribution of an original scale storage vault were analyzed and then numerical analysis of a 1/4-scale storage vault was performed to compare each model. The proper heat generation for a 1/4-scale storage vault, at which the temperature and velocity field of a 1/4-scale storage vault showed the best agreement with that of the original storage vault, was evaluated with proposed dimensionless parameters. The behavior of temperature and velocity of fluid in the 1/4-scale case were most similar to those of the original scale, using a heat flux 1.3 times higher than that seen in the original scale, which was approximately 190 W.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.243-251
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• 1992

Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD 2 and /MOD3 codes with the test of SB-CL-18 of the LSIF (Large Scale Test Facility）. The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery including the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in かe base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs wiか the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing.

• Transactions of the Korean hydrogen and new energy society
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• v.19 no.5
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• pp.367-376
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• 2008

To obtain the data of the pressure loss and differential pressure at the inside of the stack that was composed of 126 cells with 7,500 cm2 electrode area, 75kW molten carbonate fuel cell system has been operated. Computational fluid dynamics was applied to estimate reactions and thermal fluid behavior inside of the stack that was adopted with internal manifold type separator. The pressure loss coefficient K showed 72.29 to 84.01 in anode and 6.34 to 8.75 in cathode at low part of cells at the inside of 75 kW MCFC stack respectively. Meanwhile, the pressure loss coefficient of the higher part of cells at the interior of the stack showed 15.36 and 56.44 in anode and cathode respectively. These results mean that there is no big total pressure difference between anode and cathode at the inner part of 75 kW MCFC stack. This result will be reflected in 250kW MCFC system design.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.613-624
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• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.358-373
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• 1995

The Heat Transport system loop stability of CANDU-6 reactor using the CANFLEX fuel bundle was studied. The Thermal-hydraulic behavior of CANFLEX fuel bundle is similar to the conventional 37-element fuel bundle since the reactor power and the frictional pressure drop through the fuel channel is almost the same each other, Mounter the CANFLEX fuel bundle gives higher critical channel power and more homogeneous enthalpy distributions in the subchannels than 37-element fuel bundle. The SOPHT modelling or the CANFLEX fuel bundle and the Reactor outlet Header(ROH) interconnection line was made and the stability analysis response of Wolsong-1 reactor with CANFLEX fuel bundle was obtained. Without the ROH interconnection line the Heat Transport system loop using 43-element fuel bundle is unstable like the current 37-element fuel bundle. With the ROH interconnection line, however, the Heat Transport system is stable within $\pm$1％ of nominal flow. In the Heat Transport system loop stability point of view for Wolsong-1 plant therefore, the CANFLEX fuel loading is considered to be acceptable.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.137-164
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• 2018

The guarantee of the performance of the engineered barriers in a geological repository is very important for the long-term safety of disposal as well as the efficient design of the repository. Therefore, the performance of the engineered barriers under repository condition should be demonstrated by in-situ experiments conducted in an underground research laboratory. This article provides a review of the major in-situ experiments that have been carried out over the past several decades at underground research laboratories around the world to validate the performance of engineered barriers of a repository, as well as their results. In-situ experiments to study the coupled thermal-hydraulic-mechanical behavior of the engineered barrier system used to simulate the post-closure performance of the repository are analyzed as a priority. In addition, in-situ experiments to investigate the performance of the buffer material under a real repository environment have been reviewed. State-of-the art in-situ validations of the buffer-concrete interaction, and the installation of the buffer, backfill and plug, as well as characterization of the near-field rock and the corrosion of the canister materials are, also performed.

• Journal of the Korean Society of Safety
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• v.36 no.3
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• pp.74-80
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• 2021

Korea Atomic Energy Research Institute (KAERI) has been operating an integral effects test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), according to APR1400 for transient experimental and design basis accident simulation. Moreover, based on the experimental data, the domestic standard problem (DSP) program has been conducted in Korea to validate system codes. Recently, through DSP-05, the performance of the passive auxiliary feedwater system (PAFS) in the event of multiple steam generator tube rupture (MSGTR) has been analyzed. However, some errors exist in the reference input model distributed for DSP-05. Furthermore, the calculation results of the heat loss correlation for the secondary system presented in the technical report of the reference indicate that a large difference is present in heat loss from the target value. Thus, in this study, the reference model is corrected using the geometric information from the design report and drawings of ATLAS. Additionally, a new heat loss correlation is suggested by fitting the results of the heat loss tests. Herein, MSGTR-PAFS accident analysis is performed using MARS-KS 1.5 with the improved model. The steady-state calculation results do not significantly differ from the experimental values, and the overall physical behavior of the transient state is properly predicted. Particularly, the predicted operating time of PAFS is similar to the experimental results obtained by the modified model. Furthermore, the operating time of PAFS varies according to the heat loss of the secondary system, and the sensitivity analysis results for the heat loss of the secondary system are presented.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1066-1080
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• 2024

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.