• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.232-241
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• 2002

A safety assessment is carried out for the near-surface radioactive waste disposal in the reference engineered vault facility. The analysis is mainly divided into two parts. One deals with the release and transport of radionuclide in the vault and unsaturated zone. The other deals with the transport of radionuclide in the saturated zone and radiological impacts to a human group under well drinking water scenario. The parameters for source-term, geosphere and biosphere models are mainly obtained from the site specific data. The results show that the annual effective doses are dominated by long lived, mobile radionuclides and their associated daughters. And it is found that the total effective dose for drinking water is far below the general criteria of regulatory limit for radioactive waste disposal facility.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2268-2276
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• 2021

Spent nuclear fuel, which can degrade during long-term storage, must be transported intact in normal transport conditions. In this regard, many studies, including those involving Multi-Modal Transportation Test (MMTT) campaigns, have been conducted. In order to transport the spent fuel safely, a tie-down structure for supporting and transporting a cask containing the spent fuel is essential. To ensure its structural integrity, a method for finding an optimum conceptual design for the tie-down structure is presented. An optimized transportation test model of a tie-down structure for the KORAD-21 metal cask is derived based on the proposed optimization approach, and the transportation test model is manufactured by redesigning the optimized model to enable its producibility. The topology optimization approach presented in this paper can be used to obtain optimum conceptual designs of tie-down structures developed in the future.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.734-748
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• 2023

A comprehensive investigation of the Ronen method is performed in homogeneous and heterogeneous slab problems from the Sood benchmark, considering isotropic and linearly-anisotropic problems. Three finite differences implementations are exercised and compared. The results are compared to reference solutions using one and two energy groups. The validation is performed for the criticality eigenvalue and the fundamental neutron flux distribution. The results demonstrate the significantly improved accuracy achievable by the Ronen method using a broad set of problems. For standard convergence tolerances, the maximal deviation in criticality eigenvalue is less than ten pcm, and the maximal deviation in the spatial distribution of the flux is less than 2%, always located near sharp interfaces or vacuum boundaries.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.275-289
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• 2020

In 2017, multimodal transportation tests for evaluating road, sea, and rail transport were performed by research institutes in the US, Spain, and the Republic of Korea. In this study, acceleration and strain data determined through road and sea tests were analyzed. It was investigated whether the load generated for each transport mode was amplified or attenuated according to the load transfer path. From the results, it was confirmed that the load transfer characteristics differed according to the transportation mode and loading path. The effects of strain determined through each test on the structural integrity of the spent nuclear fuel were also investigated. It was found that the magnitude of the measured strain had a negligible effect on the structural integrity of the spent nuclear fuel, considering its fatigue strength. The results for the acceleration and strain data analyses obtained in this study will be useful for scheduled domestic transportation tests under normal transport conditions.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1358-1367
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• 2017

The data integration with modeled predictions (DIMP) model is a promising inverse radiation transport method for solving the special nuclear material (SNM) holdup problem. Unlike previous methods, DIMP is a completely passive nondestructive assay technique that requires no initial assumptions regarding the source distribution or active measurement time. DIMP predicts the most probable source location and distribution through Bayesian inference and quasi-Newtonian optimization of predicted detector responses (using the adjoint transport solution) with measured responses. DIMP performs well with forward hemispherical collimation and unshielded measurements, but several considerations are required when using narrow-view collimated detectors. DIMP converged well to the correct source distribution as the number of synthetic responses increased. DIMP also performed well for the first experimental validation exercise after applying a collimation factor, and sufficiently reducing the source search volume's extent to prevent the optimizer from getting stuck in local minima. DIMP's simple point detector response function (DRF) is being improved to address coplanar false positive/negative responses, and an angular DRF is being considered for integration with the next version of DIMP to account for highly collimated responses. Overall, DIMP shows promise for solving the SNM holdup inverse problem, especially once an improved optimization algorithm is implemented.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.127-131
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• 2022

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.320-326
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• 1997

The nuclide transport in the one-dimensional porous medium is considered as a first step in developing a decay chain transport in multidimensional inhomogeneous media. A method of solving conventional advection-dispersion equation with decay chain of arbitrary length by using the method of characteristics (MOC) is introduced. In specific cases where the advection are dominant rather than dispersion, the method is known to be useful : one of the most distinctive advantages in applying the model is that the MU minimizes the numerical dispersion, which is distinguished in such common numerical schemes as finite element method and finite difference method. The suggested model is considered to be effective through several illustrations for the case that decay chain of arbitrary length is involved during transport which is difficult to solve by standard numerical solutions if the medium becomes more complicated.

• Nuclear Engineering and Technology
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• v.44 no.1
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• pp.79-88
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• 2012

Migration tests were performed with conservative tracers in a fractured zone that had a single fracture of about 2.5 m distance at the KURT. To interpret the migration of the tracers in the fractured rock, a solute transport model was developed. A two dimensional variable aperture channel model was adopted to describe the fractured path and hydrology, and a particle tracking method was used for solute transport. The simulation tried not only to develop a migration model of solutes for open flow environments but also to produce ideas for a better understanding of solute behaviours in indefinable fracture zones by comparing them to experimental results. The results of our simulations and experiments are described as elution and breakthrough curves, and are quantified by momentum analysis. The main retardation mechanism of nonsorbing tracers, including matrixdiffusion, was investigated.

• Journal of Radiation Industry
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• v.8 no.1
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• pp.59-63
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• 2014

Proton Induced X-ray Emission (PIXE) is a MeV ion beam analysis method for use with particle accelerators. PIXE uses low-energy charged particles as an excitation mechanism to generate characteristic x-ray emission from each element in a target. In PIXE analysis, the beam current used is from a few nA to several tens of nA. Chosun University (Cyclotron Research Center) designed a $50{\mu}A$ beam line from the 13 MeV cyclotron for use with a PIXE analysis system, as well as performing beam transport line optimization research. In this study, the beam line operation conditions for the optimization process of beam transport and beam characteristics are shown.