• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4125-4133
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• 2023

The transportation of spent nuclear fuel is an important process that involves road and sea transport from an interim storage facility to storage and final disposal sites. As spent nuclear fuel poses a significant risk, carefully evaluating its vibration and shock characteristics under normal transport conditions is essential. In this regard, full-scale multi-modal transport tests (MMTT) have been conducted domestically and internationally. In this paper, we discuss the process of developing a multi-body dynamics (MBD) model to analytically simulate conditions that cannot be considered in tests. The MBD model is based on the KORAD-21 transportation system was validated using the Korean MMTT results from 2020 to 2021. This paper summarizes the details of the development and verification of the MBD model for the KORAD-21 transportation system under normal transport test conditions. This approach can be applicable to various transportation scenarios and systems, and the results of this study will help to ensure that nuclear fuel transportation is conducted safely and effectively.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.381-390
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• 2009

This article provides selects of outstanding problems in computational neutron transport, with some suggested approaches thereto, as follows: i) ray effect in discrete ordinates method, ii) diffusion synthetic acceleration in strongly heterogeneous problems, iii) method of characteristics extension to three-dimensional geometry, iv) fission source and $k_{eff}$ convergence in Monte Carlo, v) depletion in Monte Carlo, vi) nuclear data evaluation, and vii) uncertainty estimation, including covariance data.

• 방사성폐기물학회지
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• 제20권3호
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• pp.321-338
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• 2022

The transport of radionuclides at oceanic scales can be assessed using a Lagrangian model. In this review an application of such a model to the Atlantic, Indian and Pacific oceans is described. The transport model, which is fed with water currents provided by global ocean circulation models, includes advection by three-dimensional currents, turbulent mixing, radioactive decay and adsorption/release of radionuclides between water and bed sediments. Adsorption/release processes are described by means of a dynamic model based upon kinetic transfer coefficients. A stochastic method is used to solve turbulent mixing, decay and water/sediment interactions. The main results of these oceanic radionuclide transport studies are summarized in this paper. Particularly, the potential leakage of ¹³⁷Cs from dumped nuclear wastes in the north Atlantic region was studied. Furthermore, hypothetical accidents, similar in magnitude to the Fukushima accident, were simulated for nuclear power plants located around the Indian Ocean coastlines. Finally, the transport of radionuclides resulting from the release of stored water, which was used to cool reactors after the Fukushima accident, was analyzed in the Pacific Ocean.

• Nuclear Engineering and Technology
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• 제44권2호
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• pp.207-224
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• 2012

This article discusses selects of some outstanding problems in nuclear reactor analysis, with proposed approaches thereto and numerical test results, as follows: i) multi-group approximation in the transport equation, ii) homogenization based on isolated single-assembly calculation, and iii) critical spectrum in Monte Carlo depletion.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2625-2634
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• 2022

Discretization errors are extremely challenging conundrums of discrete ordinates calculations for radiation transport problems with void regions. In previous work, we have presented a multi-collision source method (MCS) to overcome discretization errors, but the efficiency needs to be improved. This paper proposes a goal-oriented algorithm for the MCS method to adaptively determine the partitioning of the geometry and dynamically change the angular quadrature in remaining iterations. The importance factor based on the adjoint transport calculation obtains the response function to get a problem-dependent, goal-oriented spatial decomposition. The difference in the scalar fluxes from one high-order quadrature set to a lower one provides the error estimation as a driving force behind the dynamic quadrature. The goal-oriented algorithm allows optimizing by using ray-tracing technology or high-order quadrature sets in the first few iterations and arranging the integration order of the remaining iterations from high to low. The algorithm has been implemented in the 3D transport code ARES and was tested on the Kobayashi benchmarks. The numerical results show a reduction in computation time on these problems for the same desired level of accuracy as compared to the standard ARES code, and it has clear advantages over the traditional MCS method in solving radiation transport problems with reflective boundary conditions.

• 방사성폐기물학회지
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• 제19권1호
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• pp.61-74
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• 2021

A paradigm shift in the government's energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant's units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2221-2228
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• 2021

One-dimensional oxygen transport relation is indispensable to study the oxygen distribution in the LBE-cooled system with an oxygen control device. In this paper, a numerical research is carried out to study the oxygen transport characteristics in a gas-phase oxygen control device, including the static case and dynamic case. The model of static oxygen control is based on the two-phase VOF model and the results agree well with the theoretical expectation. The model of dynamic oxygen control is simplified and the gas-liquid interface is treated as a free surface boundary with a constant oxygen concentration. The influences of the inlet and interface oxygen concentration, mass flow rate, temperature, and the inlet pipe location on the mass transfer characteristics are discussed. Based on the results, an oxygen mass transport relation considering the temperature dependence and velocity dependence separately is obtained. The relation can be used in a one-dimensional system analysis code to predict the oxygen provided by the oxygen control device, which is an important part of the integral oxygen mass transfer models.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.353-362
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• 2014

The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling) expert working group. The "Fast Total Monte Carlo" method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on $k_{\infty}$, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.

• Nuclear Engineering and Technology
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• 제34권3호
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• 방사성폐기물학회지
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• 제9권3호
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• pp.161-167
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• 2011

사용후핵연료 운반용기 표면온도가 $85^{\circ}C$를 초과할 경우, 대인용 보호막(Personnel Barrier) 또는 운반용 덮개(Transport Hood)를 설치하여 운반 중 운반용기 표면에 사람이 직접 접근할 수 없도록 하여야 한다. 운반용 덮개가 설치된 경우, 열적 안전성 평가의 한 가지 경우인 정상조건 열해석 시, 외부환경 경계조건(환경온도 및 외부복사온도)으로 적용하기 위해서 운반용 덮개 내부 열 환경 조건(내부 공기온도 및 운반용 덮개 표면온도)을 계산해야 한다. 따라서 본 연구에서는 운반용 덮개 내부 공기온도 및 표면온도를 계산하기 위한 해석적 방법 및 열전달 특성에 대한 분석을 수행하였고 CFD 해석 결과와 비교를 통해 타당성을 검증하였다.