• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1826-1833
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• 2020

Administration of stable iodine has been considered a best measure to protect the thyroid from internal irradiation by radioiodine intake, and its efficacy on thyroid protection has been quantitatively evaluated in several simulation studies on the basis of simple iodine biokinetic models (i.e., three-compartment model). However, the new iodine biokinetic model adopted by the International Commission on Radiological Protection interprets and expresses the thyroid blocking phenomenon differently. Therefore, in this study, the new model was analyzed in terms of thyroid blocking and implemented to reassess the protective effects and to produce dosimetric data. The biokinetic model calculation was performed using computation modules developed by authors, and the results were compared with those of experimental data and prior simulation studies. The new model predicted protective effects that were generally consistent with those of experimental data, except for those in the range of stable iodine administration -72 h before radioiodine exposure. Additionally, the dosimetric data calculated in this study demonstrates a critical limitation of the three-compartment model in predicting bioassay functions, and indicated that dose assessment 1 d after exposure would result in a similar dose estimate irrespective of the administration time of stable iodine.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.446-454
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• 2021

Critical heat flux (CHF) has traditionally been evaluated using look-up tables or empirical correlations for nuclear power plants. However, under complex moving condition, it is necessary to reconsider the CHF characteristics since the conventional CHF prediction methods would no longer be applicable. In this paper, the additional forces caused by motions have been added to the annular film dryout (AFD) mechanistic model to investigate the effect of moving condition on CHF. Moreover, a theoretical model of the natural circulation loop with additional forces is established to reflect the natural circulation characteristics of the loop system. By coupling the system loop with the AFD mechanistic model, a CHF prediction program called NACOM for natural circulation loop under moving condition is developed. The effects of three operating conditions, namely stationary, inclination and rolling, on the CHF of the loop are then analyzed. It can be clearly seen that the moving condition has an adverse effect on the CHF in the natural circulation system. For the calculation parameters in this paper, the CHF can be reduced by 25% compared with the static value, which indicates that it is important to consider the effects of moving condition to retain adequate safety margin in subsequent thermal-hydraulic designs.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.579-588
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• 2022

In order to meet the needs of domestic reactor severe accident analysis program, a MIDAC (Module Invessel Degraded severe accident Analysis Code) is developed and maintained by Xi'an Jiaotong University. As the accuracy of the calculation results of the analysis program is of great significance for the formulation of severe accident mitigation measures, the article select three experiments to evaluate the updated severe accident models of MIDAC. Among them, QUENCH-06 is the international standard No.45, QUENCH-16 is a test for the analysis of air oxidation, and FROMA is an out-of-pile fuel rod melting experiment recently carried out by Xi'an Jiaotong University. The heating and melting model with lumped parameter method and the steam oxidation model with Cathcart-Pawel and Volchek-Zvonarev correlations combination in MIDAC could better meet the needs of severe accident analysis. Although the influence of nitrogen still need to be further improved, the air oxidation model with NUREG still has the ability to provide guiding significance for engineering practice.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.483-488
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• 1998

A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.361-371
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• 2000

The 123Te(p,n)123I, 124Te(p,n)124I and 124Te(p,2n)123I reactions, among the many reaction channels opened, are the major reactions under consideration from a diagnostic purpose because reaction residuals as the gamma emitters are used for most radiophamaceutical applications involving radioiodine. Based on the available experimental data, the absorption cross sections and elastic scattering angular distributions of the proton-induced nuclear reaction on Te isotopes below 60 MeV are calculated using the optical model code APMNK. The transmission coefficients of neutron, proton, deuteron, trition and alpha particles are calculated by CUNF code and are fed into the GNASH code. By adjusting level density parameters and the pair correction values of some reaction channels, as well as the composite nucleus state density constants of the pre-equilibrium model, the production cross sections and energy-angle correlated spectra of the secondary light particles, as well as production cross sections and energy distributions of heavy recoils and gamma rays are calculated by the statistical plus pre-equilibrium model code GNASH. The calculated results are analysed and compared with the experimental data taken from the EXFOR. The optimized global optical model parameters give overall agreement with the experimental data over both the entire energy range and all tellurium isotopes.

• 대한핵의학회:학술대회논문집
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• 1999.05a
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• pp.222-227
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• 1999

Re-188 is suitable for endovascular liquid-balloon brachytherapy for the prevention of restenosis after angioplasty. Re-188 was concentrated to 3700 MBq/ml and labeled with DTPA. According to dosimetric calculation, it took 420 seconds using Re-188 solution with concentration of 3700 MBq/ml to irradiate 17.6 Gy to the target at 1 mm from the balloon surface. Software was made to estimate the irradiation time. MIRD calculation with dynamic bladder model yielded the whole body dose of Re-188-DTPA as 0.005 mGy/MBq in case of balloon rupture and release of the whole amount into the blood.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.125-137
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• 2019

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.875-884
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• 2009

We present three nuclear/hydrogen-related R&D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA, with which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed a tool to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The dynamic equations for the distillation column of HI process are described with 4 material components involved in the HI process: H2O, HI, I2, H2. For the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. The improved Neumann model predicted a total pressure with 8.6% maximum relative deviation from the data and 2.5% mean relative deviation, and liquid-liquid-separation with 9.52% maximum relative deviation from the data.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.36-48
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• 2022

As a valid numerical method to obtain a high-resolution result of a flow field, computational fluid dynamics (CFD) have been widely used to study coolant flow and heat transfer characteristics in fuel rod bundles. However, the time-consuming, iterative calculation of Navier-Stokes equations makes CFD unsuitable for the scenarios that require efficient simulation such as sensitivity analysis and uncertainty quantification. To solve this problem, a reduced-order model (ROM) based on proper orthogonal decomposition (POD) and machine learning (ML) is proposed to simulate the flow field efficiently. Firstly, a validated CFD model to output the flow field data set of the rod bundle is established. Secondly, based on the POD method, the modes and corresponding coefficients of the flow field were extracted. Then, an deep feed-forward neural network, due to its efficiency in approximating arbitrary functions and its ability to handle high-dimensional and strong nonlinear problems, is selected to build a model that maps the non-linear relationship between the mode coefficients and the boundary conditions. A trained surrogate model for modes coefficients prediction is obtained after a certain number of training iterations. Finally, the flow field is reconstructed by combining the product of the POD basis and coefficients. Based on the test dataset, an evaluation of the ROM is carried out. The evaluation results show that the proposed POD-ROM accurately describe the flow status of the fluid field in rod bundles with high resolution in only a few milliseconds.