• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3580-3596
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• 2021

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.834-841
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• 2022

System thermal-hydraulic (STH) code is adopted for nuclear safety analysis. The critical flow model (CFM) is significant for the accuracy of STH simulation. To overcome the defects of current CFMs (low precision or long calculation time), a CFM based on a genetic neural network (GNN) has been developed in this work. To build a powerful model, besides the critical mass flux, the critical pressure and critical quality were also considered in this model, which was seldom considered before. Comparing with the traditional homogeneous equilibrium model (HEM) and the Moody model, the GNN model can predict the critical mass flux with a higher accuracy (approximately 80% of results are within the ±20% error limit); comparing with the Leung model and the Shannak model for critical pressure prediction, the GNN model achieved the best results (more than 80% prediction results within the ±20% error limit). For the critical quality, similar precision is achieved. The GNN-based CFM in this work is meaningful for the STH code CFM development.

• Journal of the Korean Geotechnical Society
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• v.39 no.11
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• pp.53-62
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• 2023

The characteristics of unfrozen water content in frozen soils significantly impact the thermal, hydraulic, and mechanical behavior of the ground. A thorough analysis of the unfrozen water content characteristics of the target subsoil material is crucial for evaluating the stability of frozen ground. This study conducted indoor experiments to measure the freezing point and unfrozen water content of sandy soil while considering pore water salinity. Utilizing the experimental data, we introduced a novel empirical model to conveniently estimate the unfrozen water retention curve. Furthermore, the validity of the unfrozen water retention curve was assessed by comparing the experimental data with the results of a simulation model that utilized the proposed empirical model as input data.

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.32-45
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• 2023

Implementing Severe Accident Management (SAM) strategies is crucial for enhancing a nuclear power plant's resilience and safety against severe accidents conditions represented in the analysis of Station Blackout (SBO) event. Among these critical approaches, the In-Vessel Retention (IVR) through External Reactor Vessel Cooling (IVR-ERVC) strategy plays a key role in preventing vessel failure. This work is designed to evaluate the efficacy of the IVR strategy for a high-power density reactor APR1400. The APR1400's plant is represented and simulated under steady-state and transient conditions for a station blackout (SBO) accident scenario using the computer code, ASYST. The APR1400's thermal-hydraulic response is analyzed to assess its performance as it progresses toward a severe accident scenario during an extended SBO. The effectiveness of emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs) are systematically examined to assess their ability to mitigate the accident. A group of associated key phenomena selected based on Phenomenon Identification and Ranking Tables (PIRT) and uncertain parameters are identified accordingly and then propagated within DAKOTA Uncertainty Quantification (UQ) framework until a statistically representative sample is obtained and hence determine the uncertainty bands of key system parameters. The Systems Engineering methodology is applied to direct the progression of work, ensuring systematic and efficient execution.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3330-3334
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• 2024

Void drift between subchannels in a rod bundle is a crucial phenomenon affecting the calculation accuracy of thermal-hydraulic parameters in SMRs. It holds significant importance in enhancing the precision of safety analysis for SMRs. Existing research on experiment and model of void drift between rod bundle subchannels is relatively rare, and the accuracy of model calculations requires improvement. In this study, experiments on gas-liquid two-phase non-equilibrium flow were conducted to measure the redistribution of two-phase flow induced by void drift in a 1 × 2 rod bundle. The experiment results indicated that in bubby flow regime with void fraction less than 0.3, the void diffusion coefficient showed little variation with changes in void fraction. However, in slug flow and annular flow regimes with void fraction exceeding 0.3, the void diffusion coefficient significantly increased with an increase in void fraction. Furthermore, a new void drift model was developed and validated based on a subchannel code. The overall predicted uncertainty for the outlet void fraction in the rod bundle benchmark was less than 13%.

• 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.

• Geomechanics and Engineering
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• v.16 no.4
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• pp.363-373
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• 2018

The evaluation of Thermo-Hydro-Mechanical (THM) coupling behavior is important for the development of underground space for various purposes. For a high-level radioactive waste repository excavated in a deep underground rock mass, the accurate prediction of the complex THM behavior is essential for the long-term safety and stability assessment. In order to develop reliable THM analysis techniques effectively, an international cooperation project, Development of Coupled models and their Validation against Experiments (DECOVALEX), was carried out. In DECOVALEX-2015 Task B2, the in situ THM experiment that was conducted at Horonobe Underground Research Laboratory(URL) by Japan Atomic Energy Agency (JAEA), was modeled by the research teams from the participating countries. In this study, a THM coupling technique that combined TOUGH2 and FLAC3D was developed and applied to the THM analysis for the in situ experiment, in which rock, buffer, backfill, sand, and heater were installed. With the assistance of an artificial neural network, the boundary conditions for the experiment could be adequately implemented in the modeling. The thermal, hydraulic, and mechanical results from the modeling were compared with the measurements from the in situ THM experiment. The predicted buffer temperature from the THM modelling was about $10^{\circ}C$ higher than measurement near by the overpack. At the other locations far from the overpack, modelling predicted slightly lower temperature than measurement. Even though the magnitude of pressure from the modeling was different from the measurements, the general trends of the variation with time were found to be similar.

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.200-208
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• 1986

The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.