• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Journal of the Korean Society for Precision Engineering
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• v.20 no.9
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• pp.205-213
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• 2003

This study describes the mathematical equation of drawdown test model and introduces the experimental test apparatus and procedure to estimate the desorption rate ratio of a filter. The characteristics of a hydraulic filtration system of drawdown test were demonstrated by numerical simulation for various properties of filters and operation conditions. Experiments for three kinds of fuel filters were conducted according to the proposed test method. And the test results of desorption rate ratio were compared with those values anticipated in precedent multipass filtration tests. Experimental results revealed the validation of drawdown test method proposed in this study. Domestic fuel filter yielded high desorption rate ratio comparing with other foreign products, which means that the Beta ratio decreases a lot during the test. The results also showed that filtration system model could be developed including desorption rate ratio to estimate the variable Beta ratio in service life.

• Journal of Korean Society of Water and Wastewater
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• v.25 no.5
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• pp.635-642
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• 2011

Hydraulic efficiency was a vital component in evaluating the disinfection capability of clearwell. Current practice evaluates these system based on the tracer test only. In this paper, CFD(Computational Fluid Dynamics) was applied on the clearwell for alternating or supplementing the tracer test. The baffle factor derived from the CFD modeling closely matched the values obtained from full scale tracer testing. And, for suggesting proper numerical model in clearwell; the turbulence model, discretization scheme, convergence criteria were investigated through separate simulation runs. The model validation was conducted by comparing the simulated data with experimental data. In the turbulence model, the realizable ${\kappa}-{\varepsilon}$ model and the standard ${\kappa}-{\varepsilon}$ model were found to be more appropriate than RNG ${\kappa}-{\varepsilon}$ model. The residuals of convergence criteria should be used as not $10^{-3}$ but $10^{-4}$ or $10^{-5}$. In discretization scheme, the difference of simulated values in 1st, 2nd, 3rd upwind scheme was found to be insignificant. Moreover, the result of this study suggest that CFD modeling can be a reliable alternative to tracer testing for evaluating the hydraulic efficiency.

• Journal of Drive and Control
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• v.16 no.1
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• pp.7-13
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• 2019

The objective of this study was to present a damped pneumatic suspension, a bike fork suspension, which can adapt itself to incoming road excitations is presented in this paper. It consists of a hydraulic damper and a pneumatic spring in parallel with a linear spring. The study also proposed a variable and switchable orifice, in the hydraulic damper, to select appropriate damping property. Hydraulic-pneumatic circuit model for the bike fork suspension was established based on AMESim, in order to predict its performance. In addition, elastic-damping characteristics of the fork such as spring constant and viscous damping coefficient were computed and compared, for validation, with those evaluated by experiment using the universal test machine. Through simulation analysis and test, it was established that the hydraulic-pneumatic circuit model is effective and practical for development of future MTB suspensions.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3182-3195
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• 2021

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.

• Journal of the Korean Society of Safety
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• v.36 no.2
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.776-792
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• 2021

Conventional integral effect test facilities were constructed to enable the precise observation of thermal-hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal-hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to the cost and large scale of the facilities. Various nuclear technologies converging 4th industrial revolution technologies such as artificial intelligence, drone, and 3D printing, are being developed to improve plant management strategies. Additionally, new conceptual passive safety systems are being developed to enhance reactor safety. A new integral effect test facility having a noticeable scaling ratio, i.e., the (UNIST reactor innovation loop (URI-LO), is designed and constructed to improve the technical quality of these technologies by performance and feasibility tests. In particular, the URI-LO, which is constructed using a transparent material, enables better visualization and provides physical insights on multidimensional phenomena inside the reactor system. The facility design based on three-level approach is qualitatively validated with preliminary analyses, and its functionality as a test facility is confirmed through a series of experiments. The design feature, design validation, functionality test, and future utilization of the URI-LO are introduced.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.34 no.3
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• pp.723-728
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• 2014

Hydraulic experiments are performed in order to verify the swash effect of seashore structures installed to prevent scouring. However, a great deal of investment and time are required for producing the test apparatus and seashore structure used to perform the hydraulic experiment. The swash effect can be predicted, however, by using a numerical model and validation can be done based on comparisons of the numerical model and hydraulic experiment analysis results, thereby saving the cost and time required for producing the test apparatus and seashore structure. Taking a polyhedral rubble mound structure as the subject, this study performed a comparative analysis of wave run-up and run-down height of the numerical model interpretative results and the hydraulic experiment results, and validated the interpretative simulation wave test modeling technique. The study also predicted the swash effect by using the numerical interpretation approach method, whereby the volume ratio and friction area of the rubble mound were varied for different results.