• 대한방사선방어학회:학술대회논문집
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• 대한방사선방어학회 2010년도 춘계 학술발표회 및 심포지엄
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• pp.96-97
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• 2010

• 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.

• 한국항만학회지
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• 제13권1호
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• pp.147-154
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• 1999

A Numerical model is developed in order to predict cross-shore beach profile change. In this model it is assumed that sediment transport is generated by waves(bed load transport suspended load transport) and undertow which is defined as offshore directional steady flow in the surf zone. In addition wave tank experiments which reproduce storm-surge were performed. By comparing resulting profile of calculation with experiments, the applicability of this method is verified.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.681-688
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• 2020

The telegrapher's theory was used to develop a new formulation for the neutron noise equation. Telegrapher's equation is supposed to demonstrate a more realistic approximation for neutron transport phenomena, especially in comparison to the diffusion theory. The physics behind such equation implies that the signal propagation speed is finite, instead of the infinite as in the case of ordinary diffusion. This paper presents the theory and results of the development of a new method for calculation of the neutron noise using the telegrapher's equation as its basis. In order to investigate the differences and strengths of the new method against the diffusion based neutron noise, a comparison was done between the behaviors of two methods. The neutron noise based on SN transport considered as a precision measuring point. The Green's function technique was used to calculate the neutron noise based on telegrapher's and diffusion methods as well as the transport. The amplitude and phase of Green's function associated with the properties of the medium and frequency of the noise source were obtained and their behavior was compared to the results of the transport. It was observed, the differences in some cases might be considerable. The effective speed of propagation for the noise perturbations were evaluated accordingly, resulting in considerable deviations in some cases.

• Nuclear Engineering and Technology
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• 제40권4호
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• pp.319-326
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• 2008

To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1995년도 가을 학술발표회 논문집
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• pp.76-83
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• 1995

Estimation of blasting behavior of explosives is prerequisite in the numerical analysis of blasting works. In this study, blasting pressure is estimated by the finite difference method using the Flux-Corrected Transport Algorithm. To formulate the behavior of blasting gas, the mass conservation equation, the moment conservation equation, the energy conservation equation and the ideal gas state equation are used. The simplified species conservation equation is included to simulate the behavior of reacting explosives. To verify the calculation, the Sod's shock tube problem, the strong shock problem and the reacting problem we used. Numerical results show that the shock wave can be captured by means of the FCT algorithm in the reacting and nonreacting states.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.920-927
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• 2017

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

• Nuclear Engineering and Technology
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• 제46권2호
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• pp.263-272
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• 2014

Scattering source calculations using conventional spherical harmonic expansion may require lots of computation time to treat full-coupled three-dimensional photon-electron transport in a highly anisotropic scattering medium where their scattering cross sections should be expanded with very high order (e.g., $P_7$ or higher) Legendre expansions. In this paper, we introduce a modified scattering kernel approach to avoid the unnecessarily repeated calculations involved with the scattering source calculation, and used it with parallel computing to effectively reduce the computation time. Its computational efficiency was tested for three-dimensional full-coupled photon-electron transport problems using our computer program which solves the multi-group discrete ordinates transport equation by using the discontinuous finite element method with unstructured tetrahedral meshes for complicated geometrical problems. The numerical tests show that we can improve speed up to 17~42 times for the elapsed time per iteration using the modified scattering kernel, not only in the single CPU calculation but also in the parallel computing with several CPUs.

• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2748-2755
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• 2024

This paper presents two new methods for variance reduction for shielding calculation in Monte Carlo radiation transport. One method is CADIS-NEE, which combines Consistent Adjoint Driven Importance Sampling (CADIS) and next-event estimator (NEE) methods to increase the calculation efficiency of tallies at points. The other is CADIS-deterministic transport (DXTRAN), which combines CADIS and DXTRAN to obtain higher performance than using CADIS and DXTRAN separately. The combination processes are derived and implemented in the hybrid Monte-Carlo-Deterministic particle-transport code NECP-MCX. Various problems are tested to demonstrate the effectiveness of the two methods. According to the results, the two combination methods have higher efficiency than using CADIS, NEE or DXTRAN separately. In a long-distance photon-transport problem, CADIS-NEE converges faster than NEE and the figure of merit (FOM) of CADIS-NEE is 75.6 times of NEE. In a labyrinthine problem, CADIS-DXTRAN's FOM surpasses that of DXTRAN and CADIS by a factor of 45.3 and 17.7, respectively. Therefore, it is advisable to employ these two novel methods selectively in appropriate scenarios to reduce variance.

• 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.