• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3551-3566
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• 2022

The STELLA program was launched to support the PGSFR development in 2012 and for the 2nd stage, the STELLA-2 facility was designed to investigate the integral effect of safety systems including the comprehensive interaction among PHTS, IHTS and DHRS. In STELLA-2, the long-term transient behavior after accidents can be observed and the overall safety aspect can also be evaluated. In this paper, the basic design concept from engineering basis to specific design is described. The design was aimed to meet similarity criteria and requirements based on various non-dimensional numbers and the result satisfied the key features to explain the reasoning of safety evaluation. The result of this study was used to construct the facility and the experiment is on-going. In general, the final design meets the similarity criteria of the multidimensional physics inside the reactor pool. And also, for the conservation of natural circulation phenomena, the design meets the similarity requirements of geometry and thermo-dynamic behavior.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• 유체기계공업학회:학술대회논문집
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• 2006.08a
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• pp.489-492
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• 2006

The present work was performed to investigate the thermal and hydraulic characteristics of flow inside the plain and turbulator flat tubes for the automobile application. The pressure drop and heat transfer coefficient at laminar, transition and turbulent regimes were studied experimentally and numerically. The flow transition was confirmed by flow visualization and quantitative data. It is proposed equations for the friction and heat transfer coefficient in the fully developed laminar flow inside rectangular tube as function of aspect ratio.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.745-750
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• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• 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.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• 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 Korean Tunnelling and Underground Space Association
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• v.25 no.2
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• pp.175-186
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• 2023

The buffer materials in disposal hole are exposed to the decay heat from spent nuclear fuels and groundwater inflow through adjacent rockmass. Since understanding of thermal-hydro-mechanical-chemical (T-H-M-C) interaction in buffer material is crucial for predicting their long-term performance and safety of disposal repository, it is necessary to investigate the heating-hydration characteristics and consequent T-H-M-C behavior of the buffer materials under disposal conditions considering geochemical factors. In response, the Korea Atomic Energy Research Institute developed a laboratory-scale 'Lab.THMC' experiment system, which characterizes the T-H-M behavior of buffer materials under different geochemical conditions by analyzing heating-hydration process and stress changes. This technical report introduces the detail design of the Lab.THMC system, summarizes preliminary experimental results, and outlines future research plans.