• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2163-2173
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• 2024

This study introduces a transient analysis module developed for RAST-V and validates it using the Kalinin-3 benchmark problem. For the benchmark analysis, RAST-V standalone and STREAM/RAST-V calculations were performed. STREAM supplies the few-group constants and RAST-V conducts a 3D core simulation utilizing few-group cross-sectional data. To improve accuracy, the main solver was developed based on the advanced semi-analytic nodal method. To evaluate the computational capability of the transient analysis module in RAST-V, Kalinin-3 benchmark is employed. Kalinin-3 represents a coolant transient benchmark that offers experimental data during the deactivation of the Main Circulation Pumps. Consequently, the transient calculations reflected the changes in the reactor flow rate. Benchmark comprising steady-state and transient calculations. During the steady state, the STREAM/RAST-V combination demonstrated a 30 ppm root mean square difference from 0 to 128.50 EFPD. For the transient calculations, STREAM/RAST-V showed power differences within ±7 % over a range of 0-300 s. Axial offset differences were within ±3 %, and the RMS difference in radial power ranged within 2.596 % at both 0 and 300 s. Overall, this study effectively demonstrated the newly developed transient solver in RAST-V and validated it using the Kalinin-3 benchmark problem.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2498-2515
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• 2024

In this study, we propose a fully parallelized adaptive particle refinement (APR) algorithm for smoothed particle hydrodynamics (SPH) to construct a stable and efficient multi-resolution computing system for nuclear safety analysis. The APR technique, widely employed by SPH research groups to adjust local particle resolutions, currently operates on a serialized algorithm. However, this serialized approach diminishes the computational efficiency of the system, negating the advantages of acceleration achieved through high-performance computing devices. To address this drawback, we propose a fully parallelized APR algorithm designed to enhance both efficiency and computational accuracy, facilitated by a new adaptive smoothing length model. For model validation, we simulated both hydrostatic and hydrodynamic benchmark cases in 2D and 3D environments. The results demonstrate improved computational efficiency compared to the conventional SPH method and APR with a serialized algorithm, and the model's accuracy was confirmed, revealing favorable outcomes near the resolution interface. Through the analysis of jet breakup, we verified the performance and accuracy of the model, emphasizing its applicability in practical nuclear safety analysis.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.5
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• pp.502-509
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• 2019

Studies have been made on performance evaluation of expansion joint systems for an ordinary highway or road bridge. However little study has been made for runway connection bridges at airports. A study on performance evaluated from computer code analysis and shrinkage, extension, and compression repetition tests based on KS F 4425 is conducted to a newly developed expansion joint system which has been installed in a runway connection bridge at Incheon Airport Extension 2 Construction Site. The MIDAS computer code is used to analyze the performance before the manufacture of the mock-up of expansion joint system on the basis of design requirements. Tests based on the KS F 4425 of 2001 year-version are conducted for the mock-up. Domestic codes and standards to validate the performance of the expansion joint system in a connection bridge have been developed for a vehicle. However the expansion joint system tested in this study is installed in a runway connection bridge for an aircraft. Conservatively the heaviest one among airplanes departing and landing at Incheon Airport is assumed level-F $468.4kN/m^2$ and adopted for the tests and analyses in this study. KS F 4425 method is selected for the shrinkage, extension, and compression repetition tests. No remarkable problem was observed for the 2,500-cycle shrinkage and extension and two million-cycle repeatition load tests. The results of this study are expected to contribute to establishment of code and standard for the performance validation of an expansion joint system installed in a runway connection bridge for an aircraft by providing performance test results and computer analysis results based on finite element methods.

• Analytical Science and Technology
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• v.36 no.3
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• pp.105-112
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• 2023

Lycopene, a carotenoid hydrocarbon is known to have effects on reducing cardiovascular risk factors, blood lipids, and blood pressure. Thus, a lot of supplementary foods with lycopene in several dosage forms like soft capsule filled with liquid and hard capsule filled with powder are available in a market. Recently, however, our research group found that the lycopene assay in Supplementary Food Code of South Korea is only valid for oily lycopene preparation. Thus, here, we developed a simple method to determine lycopene in solid preparations for Supplementary Food Code of South Korea using saponification and liquid chromatography with an absorbance detector. The method was validated following Ministry of Food and Drug Safety guidelines. All validation parameters observed in this study were within acceptable criteria of the guidelines (selectivity, linearity of r² ≥ 0.991, lower limit of quantification of 0.0149 mg/mL, accuracy as recovery (R) between 92.70 and 97.18 %, repeatability as relative standard deviation (RSD) values of R between 0.85 and 1.59 %, and reproducibility as the RSD value of interlaboratory R of 3.70 %). Additionally, the practical sample applicability of the validated method was confirmed by accuracy between 98.81 and 101.59 % observed from its lycopene certified reference material (CRM) analyses. Therefore, the present method could contribute to fortify the supplementary food safety management system in South Korea.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.271-305
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• 2020

An effect of coupled thermo-hydro-mechanical and chemical (THMC) behavior is an essential part of the performance and safety assessment of geological disposal systems for high-level radioactive waste and spent nuclear fuel. Furthermore, numerical models and modeling techniques are necessary to analyze and predict the coupled THMC behavior in the disposal systems. However, phenomena associated with the coupled THMC behavior are nonlinear, and the constitutive relationships between them are not well known. Therefore, it is challenging to develop numerical models and modeling techniques to analyze and predict the coupled THMC behavior in the geological disposal systems. It is also difficult to verify and validate the development of the models and techniques because it requires expensive laboratory tests and in-situ experiments that need to be performed for a long time. DECOVALEX was initiated in 1992 to efficiently develop numerical models and modeling techniques and validate the developed models and techniques against the lab and in-situ experiments. In Korea, Korea Atomic Energy Research Institute has participated in DECOVALEX-2011, DECOVALEX-2015, and DECOVALEX-2019 since 2008. In this study, all tasks in the three DECOVALEX projects were introduced to the researcher in the field of rock mechanics and geotechnical engineering in Korea.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1273-1279
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• 2016

Production of iodine-131 by neutron activation of tellurium in tellurium dioxide ($TeO_2$) material requires a target that meets the safety requirements. In a radiopharmaceutical production unit, a new lid for a can was designed, which permits tight sealing of the target by using tungsten inert gaswelding. The leakage rate of all prepared targets was assessed using a helium mass spectrometer. The accepted leakage rate is ${\leq}10^{-4}mbr.L/s$, according to the approved safety report related to iodine-131 production in the TRIGA Mark II research reactor (TRIGA: Training, Research, Isotopes, General Atomics). To confirm the resistance of the new design to the irradiation conditions in the TRIGA Mark II research reactor's central thimble, a study of heat effect on the sealed targets for 7 hours in an oven was conducted and the leakage rates were evaluated. The results show that the tightness of the targets is ensured up to $600^{\circ}C$ with the appearance of deformations on lids beyond $450^{\circ}C$. The study of heat transfer through the target was conducted by adopting a one-dimensional approximation, under consideration of the three transfer modes-convection, conduction, and radiation. The quantities of heat generated by gamma and neutron heating were calculated by a validated computational model for the neutronic simulation of the TRIGA Mark II research reactor using the Monte Carlo N-Particle transport code. Using the heat transfer equations according to the three modes of heat transfer, the thermal study of I-131 production by irradiation of the target in the central thimble showed that the temperatures of materials do not exceed the corresponding melting points. To validate this new design, several targets have been irradiated in the central thimble according to a preplanned irradiation program, going from4 hours of irradiation at a power level of 0.5MWup to 35 hours (7 h/d for 5 days a week) at 1.5MW. The results showthat the irradiated targets are tight because no iodine-131 was released in the atmosphere of the reactor building and in the reactor cooling water of the primary circuit.

• Journal of the Korea Society for Simulation
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• v.27 no.1
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• pp.101-109
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• 2018

Field experiments as well as numerical analyses with finite element analysis codes are two valuable and complemental ways to understand the structural response under explosive blast load. However, there seems to be only limited information available about finite element analysis and experimental validation on the response of structural components under internal explosions. For complementary use of the two ways, the numerical analyses should be validated with field experiments by comparing their results. In this paper, a small-scaled reinforced concrete building with a room is employed for experimental investigations. An amount of TNT is detonated at the center of the room. Pressure at three different sites in the room, displacement of centers of two walls, and damage patterns of four walls are measured and compared to results from numerical analyses. The experimental results are much similar to the numerical analyses results. The finite element analysis code ANSYS AUTODYN is employed to numerically analyze both pressure distribution inside the room and response of walls subjected to blast pressure. The feasibility and validity of the numerical analysis on the reponses of structural components under internal explosions are discussed in terms of structural damage assessment, and evaluated as the same damage in the analysis and the experiments.

• Journal of Radiation Protection and Research
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• v.25 no.2
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• pp.75-80
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• 2000

The predicted results of INDAC code were compared with measured $^3H$ concentrations in air and pine-needle around the Wolsung site. The optimal sets of input data to INDAC were in addition selected by comparing the measured values with the predicted values of INDAC based on various conditions such as the release modes of effluents into the environment, the classification of wind classes, and the consideration of terrain. The predicted $^3H$ concentrations in air and pine-needle were shown to have good agreement with measured values, although there are some limitations such as uncertainties in measured values, complex topology around the site, and the land-sea breeze effects. The assumption on the $^3H$ behavior in vegetables or plants that the ratio of $^3H$ concentration in plant water to $^3H$ concentration in atmospheric water is 1/2 was shown to be conservative in terms of the audit calculation performed by the regulator. It was also found that data sets based on mixed mode and no terrain data were not appropriate for the audit calculation ensuring the compliance with regulations. Thus, if the mixed mode is considered as the release mode of effluents into the environment, meteorological data measured at 58 m height and terrain data should be used to evaluate the atmospheric dispersion factor.