• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1396-1408
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• 2020

Filtered containment venting system (FCVS) is one of the severe accident mitigation systems designed to release containment pressurization to maintain its integrity. The thermal-hydraulic behaviors in FCVSs are important since they affect the operation characteristics of the FCVS. In this study, a representative FCVS was modeled by RELAP5/Mod3.3 code, and the Station BlackOut (SBO) was chosen as an accident scenario. The thermal-hydraulic behaviors of an FCVS during long-term operation with two venting strategies (open-and-close strategy, open-and-non-close strategy) and the sensitivity analysis of important parameters were investigated. The results show that the FCVS can operate up to 250 h with a periodic open-and-close strategy during an SBO. Under the combined effects of steam condensation and water evaporation, the solution inventory in the FCVS increases during the venting phase and decreases during the intermission phase, showing a periodic pattern. Under this condition, the appropriate initial water level is 3-4 m; however, it should be adjusted according to the environment temperature. The FCVS can accommodate a decay heat power of 150-260 kW and may need to feed water for a higher decay heat power or drain water for a lower decay heat power during the late phase. The FCVS can function within an opening pressure range from 450 kPa to 500 kPa and a closing pressure range between 250 kPa and 350 kPa. When the open-and-non-close strategy is adopted, the solution inventory increases quickly in the early venting phase due to steam condensation and then decreases gradually due to the evaporation of water; drying-up may occur in the late venting phase. Decreasing the venting pipe diameter and increasing the initial water level can mitigate the evaporation of the scrubbing solution. These results are expected to provide useful references for the design and engineering application of FCVSs.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.109-118
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• 2023

In this paper, the evaluation method of fluidelastic instability (FEI) of newly designed steam generator tubing in pressurized water reactor (PWR) nuclear power plants is discussed. To obtain the parameters for prediction of the critical velocity of FEI for steam generator tubes, experimental research is carried out, and the design parameters are determined. Using CFD numerical simulation, the tube array scale of the model experiment is determined, and the experimental device is designed. In this paper, 7 groups of experiments with void fractions of 0% (water), 10%, 20%, 50%, 75%, 85% and 95% were carried out. The critical damping ration, fundamental frequency and critical velocity of FEI of tubes in flowing water were measured. Through calculation, the total mass and instability constant of the immersed tube are obtained. The critical damping ration measured in the experiment mainly included two-phase damping and viscous damping, which changed with the change in void fraction from 1.56% to 4.34%. This value can be used in the steam generator design described in this paper and is conservative. By introducing the multiplier of frequency and square root of total mass per unit length, it is found that the difference between the experimental results and the calculated results is less than 1%, which proves the rationality and feasibility of the calculation method of frequency and total mass per unit length in engineering design. Through calculation, the instability constant is greater than 4 when the void fraction is less than 75%, less than 4 when the void fraction exceeds 75% and only 3.04 when the void fraction is 95%.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.15-20
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• 1996

We present the result of physical studies for three integral-type burnable absorbers of gadolinia, erbia and IFBA, in the hexagonal fuel assembly. The analysis of nuclear characteristics for gadolinia and IFBA cases shows that the spectrum hardening of hexagonal fuel assembly compared to rectangular one leads to smaller reactivity hold-down worth(RHW) and less change of MTC in the negative direction per insertion of one burnable absorber rod. Erbia case, on the other hand, exhibits reversed trend in RHW and MTC due to the enhanced absorption of large resonance of Erbium at 0.5 eV It turns out to be that Erbia performs best in terms of minimizing the peak pin power and maintaining as more negative MTC as practically attainable during the entire operational phase, and IFBA provides the least residual reactivity penalty at EOC. Therefore, we take Erbium as the suitable burnable absorber and provide optimal designs of 60, 120, 180, 240 and 300 erbia-shimmed hexagonal fuel assemblies with regard to minimizing the peak pin power.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.127-137
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• 1997

Equilibrium concentrations of major elements in an underground repository with a capacity of 100,000 drums have been simulated using the geochemical computer code (EQMOD). The simulation has been carried out at the conditions of pH 12 to 13.5, and Eh 520 and -520 mV. Solubilities of magnesium and calcium decrease with the increase of pH. The solubility of iron increases with pH at Eh -520 mV of reducing environment while it almost entirely exists as the precipitate of Fe(OH)$_3$(s) at Eh 520 mV of oxidizing environment. All of cobalt and nickel are predicted to be dissolved in the liquid phase regardless of pH since the solubility limit is greater than the total concentration. In the case of cesium and strontium, all forms of both ions are present in the liquid phase because they have negligible sorption capacity on cement and large solubility under disposal atmosphere. And thus the total concentration determines the equilibrium concentration. Adsorbed amount of iodide and carbonate are dependent on adsorption capacity and adsorption equilibrium constant. Especially, the calcite turns out to be a solubility-limiting phase on the carbonate system. In order to validate the model, the equilibrium concentrations measured for a number of systems which consist of iron, cement, synthetic groundwater and radionuclides are compared with those predicted by the model. The concentrations between the model and the experiment of nonadsorptive elements cesium, strontium, cobalt nickel and iron, are well agreed. It indicates that the assumptions and the thermodynamic data in this work are valid. Using the adsorption equilibrium constant as a free parameter, the experimental data of iodide and carbonate have been fitted to the model. The model is in a good agreement with the experimental data of the iodide system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

• Journal of Powder Materials
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• v.16 no.4
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• pp.243-248
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• 2009

Carbon-coated Cu nanopowders with core/shell structure have been successfully fabricated by pulsed wire evaporation (PWE) method, in which a mixed gas of Ar/$CH_4$ (10 vol.%) was used as an ambient gas. The characterization of the samples was carried out using x-ray diffraction (XRD), scanning electron microscope (SEM), and high resolution transmission electron microscope (HRTEM). It was found that the nanoparticles show a spherical morphology with the size ranging of 10-40 nm and are covered with graphite layers of 2-4 nm. When oxygen-passivated Cu nanopowders were annealed under flowing argon gas (600 and 800$^{\circ}C$), the crystallinity of $Cu_2O$ phase and the particle size gradually increased. On the other hand, carbon-coated Cu nanopowders remained similar to as-prepared case with no additional oxide or carbide phases even after the annealing, indicating that the metal nanoparticles are well protected by the carbon-coating layers.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• The Korean Journal of Nuclear Medicine
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• v.24 no.1
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• pp.101-107
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• 1990

To develop $^{131}I-labelled$ m-iodobeneylguanidine $(^{131}I-MIBG)$, various experiments such as synthesis of MIBG, establishment of labelling conditions, determination of radiochemical purity, and examination of stability were carried out. 1) m-Iodobenzylguanidine (MIBG) sulfate was synthesized with a total yield of 62.4% by the condensation of m-iodobenzylamine hydrochloride with cyanamide via MIBG bicarbonate. Its physical properties, IR, $^1H-NMR$, and elemental analysis data were nearly identical to those of literature. 2) Freeze-dried or vacuum-dried kit vials were prepared from the mixture so as to contain MIBG (2 mg), ascorbic acid (10 mg), copper (II) sulfate (0.14 mg), and tin (II) sulfate (0.5 mg) per vial. Copper ( I ) catalyzed radioiodination of MIBG was carried out using kit vials and 0.01 M $H_2SO_4$ as solvent at $100^{\circ}C$ for 30 min under nitrogen atmosphere (optimal conditions). Labelling yield was 98% and radiochemical purity was 99.5%, respectively. 3) Solid-phase radioiodination of MIBG was carried out at $155^{\circ}C$ for 30 min using the prepared vials to contain MIBG (2 mg) and ammonium sulfate (10 mg). Duplicate reactions under the same conditions showed labelling yield of 95% and radiochemical purity of 99.5%. 4) $^{131}I-MIBG$ prepared either by catalytic or by solid-phase exchange method showed radio-chemical purity of 99% even after 3 days storing at room temperature.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.1-13
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• 2020

At present, the Department of Energy (DOE) in Unite State are directing the efforts of developing accident tolerant fuel (ATF) technology. As the first barrier of nuclear fuel system, the material selection of fuel rod cladding for ATFs is a basic but very significant issue for the development of this concept. The advanced cladding is attractive for providing much stronger oxidation resistance and better in-pile behavior under sever accident conditions (such as SBO, LOCA) for giving more coping time and, of course, at least an equivalent performance under normal condition. In recent years, many researches on in-plie or out-pile physical properties of some suggested cladding materials have been conducted to solve this material selection problem. Base on published literatures, this paper introduced relevant research backgrounds, objectives, research institutions and their progresses on several main potential claddings include triplex SiC, FeCrAl and MAX phase material Ti3SiC2. The physical properties of these claddings for their application in ATF area are also reviewed in thermohydraulic and mechanical view for better understanding and simulating the behaviors of these new claddings. While most of important data are available from publications, there are still many relevant properties are lacking for the evaluations.