• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.281-290
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• 2003

The KN-12 spent nuclear fuel transport cask, which is a Type B(U) package designed to comply with the requirements of Korea Atomic Energy Act[1], IAEA Safety Standards Series No.TS-R-1[2] and US 10 CFR Part 71[3], is designed for carrying up to 12 PWR spent fuel assemblies in a basket structure. The cask has been licensed in accordance with Korea Atomic Energy Act and was fabricated in Korea in accordance with the requirements of ASME B&PV Sec.III, Div.3[4]. The cask must maintain thermal integrity in accordance with the related regulations and be evaluated to verify that the thermal performance of the cask complies with the regulatory requirements. The temperatures of the cask and components were determined by using finite elements methods with a numerical tool, safety tests using an 1/8 height slice model of the real cask were conducted to demonstrate verification of the numerical tool and methods, and heat transfer tests for normal transport conditions were performed as a fabrication acceptance test to demonstrate the heat transfer capability of the cask.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.769-779
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• 2023

This paper describes an hp-angular adaptivity algorithm in the discrete ordinates method for Boltzmann transport applications with strong angular effects. This adaptivity uses discontinuous finite element quadrature sets with different degrees, which updates both angular mesh and the degree of the underlying discontinuous finite element basis functions, allowing different angular local refinement to be applied in space. The regular and goal-based error metrics are considered in this algorithm to locate some regions to be refined. A mapping algorithm derived by moment conservation is developed to pass the angular solution between spatial regions with different quadrature sets. The proposed method is applied to some test problems that demonstrate the ability of this hp-angular adaptivity to resolve complex fluxes with relatively few angular unknowns. Results illustrate that a reduction to approximately 1/50 in quadrature ordinates for a given accuracy compared with uniform angular discretization. This method therefore offers a highly efficient angular adaptivity for investigating difficult particle transport problems.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.794-800
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• 2018

The proliferation resistance (PR) of Th/U and Th/Pu fuels used in CANada Deuterium Uranium for the deep geological repository was assessed by combining the multiattribute utility analysis proposed by Chirayath et al., 2015 with the transport model of radionuclides in the repository and comparing with that of the used natural U fuel case. It was found that there was no significant advantage for Th/U and Th/Pu fuels from the viewpoint of the PR in the repository. It was also found that the PR values for used nuclear fuels in the repository of Th/U, Th/Pu, and natural U was comparable with those for enrichment and reprocessing facilities in the pressurized water reactor (PWR) nuclear fuel cycle. On the other hand, the PR values considering the transport of radionuclides in the repository were found to be slightly smaller than those without their transport after the used nuclear fuels started dissolving after 1,000 years.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3347-3352
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• 2022

In 2013, the International Atomic Energy Agency (IAEA) changed the recommended maximum range of the Emergency Planning Zone (EPZ) to 30 km, and the Kori Nuclear Power Plant in Republic of Korea has also expanded the EPZ to 30 km, following the recommendation. As a result, metropolitan cities with a high population density are contained within the EPZ, and evacuating millions of people should be considered if the 30 km range of evacuation is to take place. This study proposes an evacuation plan using buses (public transportation) to transport people outside of the EPZ, quickly and efficiently. To verify the appropriate mode share ratio of buses that can guarantee the right of vulnerable road users and reduce traffic congestion, a model was built simulating the Kori Nuclear Power Plant in Ulsan Metropolitan City. The scenarios were established by changing the mode share ratio of buses and passenger cars by 10%. Considering a large-scale network analysis at the city level, a cell transmission model was applied to calculate the evacuation time in each scenario. The result shows that the optimal mode share ratio of buses is 40%, with a total evacuation time of 132 min, considering feasible bus fleets in Ulsan Metropolitan City.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2084-2094
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• 2021

Delta-form-based methods for solving high order spatial discretization schemes are introduced into the reactor S_N transport equation. Due to the nature of the delta-form, the final numerical accuracy only depends on the residuals on the right side of the discrete equations and have nothing to do with the parts on the left side. Therefore, various high order spatial discretization methods can be easily adopted for only the transport term on the right side of the discrete equations. Then the simplest step or other robust schemes can be adopted to discretize the increment on the left hand side to ensure the good iterative convergence. The delta-form framework makes the sweeping and iterative strategies of various high order spatial discretization methods be completely the same with those of the traditional S_N codes, only by adding the residuals into the source terms. In this paper, the flux limiter method and weighted essentially non-oscillatory scheme are used for the verification purpose to only show the advantages of the introduction of delta-form-based solving methods and other high order spatial discretization methods can be also easily extended to solve the S_N transport equations. Numerical solutions indicate the correctness and effectiveness of delta-form-based solving method.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.485-498
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• 2020

A diffusion synthetic acceleration (DSA) technique for the S_N transport equation discretized with the linear discontinuous expansion method with subcell balance (LDEM-SCB) on unstructured tetrahedral meshes is presented. The LDEM-SCB scheme solves the transport equation with the discrete ordinates method by using the subcell balances and linear discontinuous expansion of the flux. Discretized DSA equations are derived by consistently discretizing the continuous diffusion equation with the LDEM-SCB method, however, the discretized diffusion equations are not fully consistent with the discretized transport equations. In addition, a fine mesh rebalance (FMR) method is devised to accelerate the discretized diffusion equation coupled with the preconditioned conjugate gradient (CG) method. The DSA method is applied to various test problems to show its effectiveness in speeding up the iterative convergence of the transport equation. The results show that the DSA method gives small spectral radii for the tetrahedral meshes having various minimum aspect ratios even in highly scattering dominant mediums for the homogeneous test problems. The numerical tests for the homogeneous and heterogeneous problems show that DSA with FMR (with preconditioned CG) gives significantly higher speedups and robustness than the one with the Gauss-Seidel-like iteration.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.39-44
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• 1996

A nodal transport method based on the F$_{N}$ method is developed for the transport calculation in x- y geometry and tested for benchmark problems. Using transverse integration, the two-dimensional transport equation is converted to one-dimensional equations for x, y-directions and the one-dimensional equations are integrated over azimuthal angle. With proper approximations for the transverse leakage, the one-dimensional equations are discretized by using the F$_{N}$ method without truncation error. At present, isotropic approximation of the transverse with a quadratic or flat shape in spatial variable is tested.ted.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.113-150
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• 2012

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1189-1198
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• 2017

Computer-aided design (CAD)-based Monte Carlo radiation transport is of value to the nuclear engineering community for its ability to conduct transport on high-fidelity models of nuclear systems, but it is more computationally expensive than native geometry representations. This work describes the adaptation of a rendering data structure, the signed distance field, as a geometric query tool for accelerating CAD-based transport in the direct-accelerated geometry Monte Carlo toolkit. Demonstrations of its effectiveness are shown for several problems. The beginnings of a predictive model for the data structure's utilization based on various problem parameters is also introduced.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3766-3777
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• 2022

It is essential to ensure the safety of spent nuclear fuel (SNF) transport cask in drop situation that is included in transport accident scenarios. The safety of the drop situation is affected by the impact absorption performance of impact limiters. Therefore, when designing an impact limiter, the uncertainty in the material properties that affect the impact absorption performance must be considered. In this study, the material properties of the wood inside the impact limiter were selected as the variables for a parametric study. The sensitivity analysis of the drop response of the SNF transport cask with impact limiter was performed. The minimum wood strength required to prevent a direct collision between the cask and floor was derived from the analysis results. In addition, the plastic strain response was analyzed and strain-based evaluation was performed. Based on this result, the critical values of wood properties that change the impact dynamic characteristics were investigated. Finally, the optimal material properties of wood were obtained to secure the structural safety of the SNF transport cask. The results of this study can contribute to the development of SNF transport cask, thereby ensuring safety in transport accident conditions.