• 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.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.349-351
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• 2003

To ensure the safety of the currently operating nuclear power plant, the Periodic Safety Review program has been conducted. In PSR program, the cumulative behavior of the radionuclide that might be released from the power plant is addressed. The Cs-137 in soil around Kori nuclear power plant was investigated. The soil sample was analyzed and compared with the reference area. The model calculation explained the depth profile of Cs-137.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.859-866
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• 2015

The aim of this work is to investigate the transport mechanism of radioactive nuclides through the reverse osmosis (RO) membrane and to estimate its effectiveness for nuclide separation from radioactive liquid waste. An analytical model is developed to simulate the RO separation, and a series of experiments are set up to confirm its estimated separation behavior. The model is based on the extended Nernst-Plank equation, which handles the convective flux, diffusive flux, and electromigration flux under electroneutrality and zero electric current conditions. The distribution coefficient which arises due to ion interactions with the membrane material and the electric potential jump at the membrane interface are included as boundary conditions in solving the equation. A high Peclet approximation is adopted to simplify the calculation, but the effect of concentration polarization is included for a more accurate prediction of separation. Cobalt and cesium are specifically selected for the experiments in order to check the separation mechanism from liquid waste composed of various radioactive nuclides and nonradioactive substances, and the results are compared with the estimated cobalt and cesium rejections of the RO membrane using the model. Experimental and calculated results are shown to be in excellent agreement. The proposed model will be very useful for the prediction of separation behavior of various radioactive nuclides by the RO membrane.

• Nuclear Medicine and Molecular Imaging
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• v.42 no.4
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• pp.261-266
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• 2008

We established a model to calculate radioactive waste from sewage disposal tank of hospitals to optimize the number of patients receiving inpatient radioiodine therapy within the safety guideline in our country. According to this model and calculation of radioactivity concentration using the number of patients per week, the treatment dose of radioiodine, the capacity and the number of sewage tanks and the daily amount of water waste per patient, estimated concentration of radioactivity in sewage waste upon disposal from disposal tanks after longterm retention were within the safety guideline (30 Bq/L) in all the hospitals examined. In addition to the fact that we could increase the number of patients in two thirds of hospitals, we found that the daily amount of waste water was the most important variable to allow the increase of the number of patients within the safety margin of disposed radioactivity. We propose that saving the water amount be led to increase the number of patients and they allow two patients in an already furnished hospital inpatient room to meet the increasing need of inpatient radioiodine treatment for thyroid cancer.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.310-323
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• 2023

The k-eigenvalue neutronics/thermal-hydraulics coupling calculation is a key issue for reactor design and analysis. Jacobian-free Newton-Krylov (JFNK) method, featured with super-linear convergence rate and high efficiency, has been attracting more and more attention to solve the multi-physics coupling problem. However, it may converge to the high-order eigenmode because of the multiple solutions nature of the k-eigenvalue form of multi-physics coupling issue. Based on our previous work, a modified JFNK with a line search method is proposed in this work, which can find the fundamental eigenmode together with thermal-hydraulics feedback in a wide range of initial values. In detail, the existing modified JFNK method is combined with the line search strategy, so that the intermediate iterative solution can avoid a sudden divergence and be adjusted into a convergence basin smoothly. Two simplified 2-D homogeneous reactor models, a PWR model, and an HTR model, are utilized to evaluate the performance of the newly proposed JFNK method. The results show that the performance of this proposed JFNK is more robust than the existing JFNK-based methods.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.546-558
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• 2015

Chord length sampling method in Monte Carlo simulations is a method used to model spherical particles with random sampling technique in a stochastic media. It has received attention due to the high calculation efficiency as well as user convenience; however, a technical issue regarding boundary effect has been noted. In this study, after analyzing the distribution characteristics of spherical particles using an explicit method, an alternative chord length sampling method is proposed. In addition, for modeling in finite media, a correction method of the boundary effect is proposed. Using the proposed method, sample probability distributions and relative errors were estimated and compared with those calculated by the explicit method. The results show that the reconstruction ability and modeling accuracy of the particle probability distribution with the proposed method were considerably high. Also, from the local packing fraction results, the proposed method can successfully solve the boundary effect problem. It is expected that the proposed method can contribute to the increasing of the modeling accuracy in stochastic media.

• Journal of Radiation Protection and Research
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• v.41 no.2
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• pp.123-131
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• 2016

Background: With the increase in the number of particle accelerator facilities under either operation or construction, the accurate calculation using Monte Carlo codes become more important in the shielding design and radiation safety evaluation of accelerator facilities. Materials and Methods: The calculations with different physics models were applied in both of cases: using only physics model and using the mix and match method of MCNPX code. The issued conditions were the interactions of 600 MeV proton and $290MeV{\cdot}n^{-1}$ oxygen with a carbon target. Both of cross-section libraries, JENDL High Energy File 2007 (JENDL/HE-2007) and LA150, were tested in this calculation. In the case of oxygen ion interactions, the calculation results using LAQGSM physics model and JENDL/HE-2007 library were compared with D. Satoh's experimental data. Other Monte Carlo calculations using PHITS and FLUKA codes were also carried out for further benchmarking study. Results and Discussion: It was clearly found that the physics models, especially intra-nuclear cascade model, gave a great effect to determine proton-induced secondary neutron spectrum in MCNPX code. The variety of physics models related to heavy ion interactions did not make big difference on the secondary particle productions. Conclusion: The variations of secondary neutron spectra and particle transports depending on various physics models in MCNPX code were studied and the result of this study can be used for the shielding design and radiation safety evaluation.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.12 no.4
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• pp.261-271
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• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.17 no.11
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• pp.1156-1166
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• 2004

Antimony profiles by MeV implantation are measured by secondary ion mass spectrometry (SIMS) and spreading resistance (SR). The moments of SIMS and simulated profiles are calculated and compared for the exact range in MeV energy. SRIM, DUPEX, ICECREM, and TSUPREM4 simulation programs are used for the calculation of range 1D, 2D. SRIM is a Monte Carlo simulation program and different inter-atomic potentials can be used for the calculation of nuclear stopping power cross-section (Sn) and range moments. Nevertheless, the range parameters were not influenced from nuclear stopping power in MeV. Through the modification of electronic stopping power cross-section (Se), the results of simulation are remarkably improved and matched very well with SIMS data. The values of electronic stopping power are optimized for Sb high energy implantation. For the electrical activation, Sb implanted samples are annealed under $N_2$ and $O_2$ ambient. Finally, Oxidation retard diffusion(ORD) effect of Sb implanted sample are demonstrated by SR measurements and ICECREM simulation.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1209-1217
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• 2019

Due to the limitation of the computers, the conventional homogenization method is based on many assumptions and approximations, and some tough problems such as energy spectrum and boundary condition are faced. To deal with those problems, the Monte Carlo global homogenization is adopted. The Reactor Monte Carlo code RMC is used to study the global homogenization method, and the one-step global homogenization method is proposed. The superimposed mesh geometry is also used to divide the physical models, leading to better geometric flexibility. A set of multigroup homogenization cross sections is online generated for each mesh under the real neutron energy spectrum and boundary condition, the cross sections are adjusted by the superhomogenization method, and no leakage correction is required. During the process of superhomogenization, the author-developed reactor core program NLSP3 is used for global calculation, so the global flux distribution and equivalent homogenization cross sections could be solved simultaneously. Meanwhile, the calculated homogenization cross section could accurately reconstruct the non-homogenization flux distribution and could also be used for fine calculation. This one-step global homogenization method was tested by a PWR assembly and a small reactor model, and the results show the validity.