• Nuclear Engineering and Technology
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• 제48권1호
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Nuclear Engineering and Technology
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• 제53권7호
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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.

• Transactions on Electrical and Electronic Materials
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• 제9권6호
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• pp.265-272
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• 2008

In this paper, AC losses of long wire Bi-2223 tapes with different twist pitch of superconducting core were fabricated, measured and analyzed. These samples produced by a powder-in-tube method are multi-filamentary tape with Ag matrix. Also, it's produced by non-twist. The critical current measurement was carried out under the environment in Liquid nitrogen and in zero field by 4-prob method. And the Magnetic measurement was carried out under the environment of applied time-varying transport current by transport method. From experiment, the susceptibility measurements were conducted while cooling in a magnetic field. Flux loss measurements were conducted as a function of ramping rate, frequency and field direction. The AC flux loss increases as the twist-pitch of the tapes decreased, in agreement with the Norris Equation. Neutron-diffraction measurements have been carried out investigate the crystal structure, magnetic structures, and magnetic phase transitions in Bi-2223([Bi, Pb]:Sr:Ca:Cu:O).

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.103-108
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• 1995

By applying the P$_1$ and P$_2$ equations to the operator form of a synthetic acceleration, we derive the p,-acceleration (diffusion synthetic acceleration: DSA) and P$_2$-acceleration (p$_2$SA) schemes in one dimensional slab geometry. We Fourier-analyze the derived acceleration schemes with the discrete-ordinates transport equation and showed that the DSA outperforms the P$_2$SA. These results confirm that one cannot simply assume that replacement of the DSA with a higher order approximation will lead to a better acceleration performance.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.319-328
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• 1992

일차원 구에서 유한요소법의 Galerkin formulation이 일차형태의 단일 에너지 중성자 수송방정식의 적분법에 적용되었다. 구분적으로 1차 혹은 2차인 Lagrange 다항식들이 선형대수 방정식들의 집합을 만들기 위해 적분법에 있는 각의존 중성자속(angular flux)에 대하여 활용되었다. 수치해석이 균질구에서의 임계문제와 비균질구에서의 scalar flux 분포에 대해서 행해졌다. 공간과 각에 대하여 연속적인 유한요소를 사용한 균질구에서의 임계문제에 대한 유한요소법의 결과들은 이론적인 해들자 비교되었다. 비균질 문제에서는 각자 공간에 대하여 불연속 유한요소를 사용하여 구한 scalar flux 분포는 ANISN code에 의한 계산결과와 잘 일치하였다.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.223-229
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• 2020

The present work proposes a solution to the static Boltzmann transport equation approximated by the simplified P₃ (SP₃) on angular, and the analytic coarse mesh finite difference (ACMFD) for spatial variables. Multi-group SP₃-ACMFD equations in 3D rectangular geometry are solved using the GMRES solution technique. As the core time dependent analysis necessitates the solution of an eigenvalue problem for an initial condition, this work is hence devoted to development and verification of the proposed static SP₃-ACMFD solver. A 3D multi-group static diffusion solver is also developed as a byproduct of this work to assess the improvement achieved using the SP₃ technique. Static results are then compared against transport benchmarks to assess the proximity of SP₃-ACMFD solutions to their full transport peers. Results prove that the approach can be considered as an acceptable interim approximation with outputs superior to the diffusion method, close to the transport results, and with the computational costs less than the full transport approach. The work would be further generalized to time dependent solutions in Part II.

• Nuclear Engineering and Technology
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• 제3권3호
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• pp.121-128
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• 1971

중성자속의 각분포를 unified modal-nodal 방법으로 전개하였다. 몇몇 표준 nodal 해석법이나 modal 해석법은 이 방법의 특수한 경우임을 밝혔다. 중성자의 발생과 산란이 등방성인 경우에 단일에너지 수송방정식을 modal-nodal moment 형으로 도출하여 고유치를 구하였다. 중성자가 생성되지 않는 매질 내에서의 역확산거리를 근사적으로 계산한 결과 표준방법으로 한 것보다 더 정화하였다.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1079-1087
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• 2021

NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

• Nuclear Engineering and Technology
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• 제41권3호
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• pp.279-286
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• 2009

Improving over a previous study [1], this paper provides a Monte Carlo method for the heat conduction analysis of problems with complicated geometry (such as a pebble with dispersed fuel particles). The method is based on the theoretical results of asymptotic analysis of neutron transport equation. The improved method uses an appropriate boundary layer correction (with extrapolation thickness) and a scaling factor, rendering the problem more diffusive and thus obtaining a heat conduction solution. Monte Carlo results are obtained for the randomly distributed fuel particles of a pebble, providing realistic temperature distributions (showing the kernel and graphite-matrix temperatures distinctly). The volumetric analytic solution commonly used in the literature is shown to predict lower temperatures than those of the Monte Carlo results provided in this paper.

• Nuclear Engineering and Technology
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• 제8권4호
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• pp.209-217
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• 1976

슬레브형로내의 중성자수송에 관한 적분 방정식을 유한요소법으로 기술했다. Hermite 내삽에 의한 1차 및 3차 다항식을 사용하여 얻은 수식으로 글레브형로의 임계치와 Milne 문제의 근사해를 계산하고 해석해와의 비교를 통하여 유한 요소법의 유용성을 논의했다.