• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3344-3351
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• 2021

The Rare Isotope Science Project was launched in 2011 in Korea toward constructing the Rare isotope Accelerator complex for ON line experiments (RAON). RAON will house several experimental systems, including the Muon Spin Rotation/Relaxation/Resonance (μSR) facility in High Energy Experimental Building B. This facility will use 600-MeV protons with a maximum current of 660 pμA and beam power of 400 kW. The key μSR features will facilitate projects related to condensed-matter and nuclear physics. Typical experiments require a few million surface muons fully spin-polarized opposite to their momentum for application to small samples. Here, we describe the design of a muon transport beam line for delivering the requisite muon numbers and the electromagnetic-component specifications in the μSR facility. We determine the beam-line configuration via beam-optics calculations and the transmission efficiency via single-particle tracking simulations. The electromagnet properties, including fringe field effects, are applied for each component in the calculations. The designed surface-muon beamline is 17.3 m long, consisting of 2 solenoids, 2 dipoles affording 70° deflection, 9 quadrupoles, and a Wien filter to eliminate contaminant positrons. The average incident-muon flux and spin rotation angle are estimated as 5.2 × 10⁶ μ⁺/s and 45°, respectively.

• Journal of Radiation Industry
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• v.6 no.1
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• pp.41-47
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• 2012

A field tracer experiment using a radioisotope was conducted to analyze the transport and dispersion characteristics of pollutants in the coastal area near a Wolsung. A rod float including GPS was released to track the paths of radioisotope. NaI detector was installed to measure the released radioisotope from the boat, and measurements were performed with the real time. The measured tracer data by a field experiment can be used as the basic data for understanding the transport characteristics of pollutants and verifying numerical models near the offshore.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.2018-2025
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• 2019

This study systematically compared two hybrid deterministic/Monte Carlo transport codes, ADVANTG/MCNP and MAVRIC, in solving a difficult shielding problem for a real-world spent fuel storage cask. Both hybrid codes were developed based on the consistent adjoint driven importance sampling (CADIS) methodology but with different implementations. The dose rate distributions on the cask surface were of primary interest and their predicted results were compared with each other and with a straightforward MCNP calculation as a baseline case. Forward-Weighted CADIS was applied for optimization toward uniform statistical uncertainties for all tallies on the cask surface. Both ADVANTG/MCNP and MAVRIC achieved substantial improvements in overall computational efficiencies, especially for gamma-ray transport. Compared with the continuous-energy ADVANTG/MCNP calculations, the coarse-group MAVRIC calculations underestimated the neutron dose rates on the cask's side surface by an approximate factor of two and slightly overestimated the dose rates on the cask's top and side surfaces for fuel gamma and hardware gamma sources because of the impact of multigroup approximation. The fine-group MAVRIC calculations improved to a certain extent and the addition of continuous-energy treatment to the Monte Carlo code in the latest MAVRIC sequence greatly reduced these discrepancies. For the two continuous-energy calculations of ADVANTG/MCNP and MAVRIC, a remaining difference of approximately 30% between the neutron dose rates on the cask's side surface resulted from inconsistent use of thermal scattering treatment of hydrogen in concrete.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4137-4141
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• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.50-53
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• 2022

To investigate the validity of the finite element analysis program to assess structural integrity of a spent nuclear fuel transport cask subjected to extreme impact loads, structural integrity of the cask for the case of an aircraft engine collision is evaluated using three FE analysis programs: Autodyn, Speed and ABAQUS explicit version. As a result of all analyses, it is confirmed that no penetration occurred in the cask wall. Even though the different programs are used, it is identified that there are insignificant differences in the FE analysis variables such as von Mises effective stress and equivalent plastic strain among the programs.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.272-272
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• 2001

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.273-274
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• 2019

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2019.05a
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• pp.197-198
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• 2019

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.179-180
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• 2018