• Nuclear Engineering and Technology
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• 제45권7호
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• pp.951-960
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• 2013

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that $UO_2-PuO_2$ (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the $UO_2$ fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of $H_2O$ moderated $UO_2-PuO_2$ (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1250-1258
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• 2017

The multigroup transport theory is the basis for many neutronics modules. A significant point of the cross-section (XS) generation procedure is the choice of the energy groups' boundaries in the XS libraries, which must be carefully selected as an unsuitable energy meshing can easily lead to inaccurate results. This decision can require considerable effort and is particularly difficult for the common user, especially if not well-versed in reactor physics. This work investigates a genetic algorithm-based tool which selects an appropriate XS energy structure (ES) specific for the considered problem, to be used for the condensation of a fine multigroup library. The procedure is accelerated by results storage and fitness calculation speedup and can be easily parallelized. The extension is applied to the coupled code SIMMER and tested on the European Sustainable Nuclear Industrial Initiative (ESNII+) Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID)-like reactor system with different fitness functions. The results show that, when the libraries are condensed based on the ESs suggested by the algorithm, the code actually returns the correct multiplication factor, in both reference and voided conditions. The computational effort reduction obtained by using the condensed library rather than the fine one is assessed and is much higher than the time required for the ES search.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1259-1268
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• 2017

The alpha eigenvalue problem in multigroup neutron diffusion is studied with particular attention to the theoretical analysis of the model. Contrary to previous literature results, the existence of eigenvalue and eigenflux clustering is investigated here without the simplification of a unique fissile isotope or a single emission spectrum. A discussion about the negative decay constants of the neutron precursors concentrations as potential eigenvalues is provided. An in-hour equation is derived by a perturbation approach recurring to the steady state adjoint and direct eigenvalue problems of the effective multiplication factor and is used to suggest proper detection criteria of flux clustering. In spite of the prior work, the in-hour equation results give a necessary and sufficient condition for the existence of the eigenvalue-eigenvector pair. A simplified asymptotic analysis is used to predict bands of accumulation of eigenvalues close to the negative decay constants of the precursors concentrations. The resolution of the problem in one-dimensional heterogeneous problems shows numerical evidence of the predicted clustering occurrences and also confirms previous theoretical analysis and numerical results.

• Nuclear Engineering and Technology
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• 제50권5호
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• 한국정보통신학회논문지
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• 제12권8호
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• pp.1488-1494
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• 2008

본 논문은 대역폭에 제약을 받지 않는 우주 통신용에 사용할 목적으로 도플러에 강인한 저 전력 비동기 FSK 수신기를 FPGA로 구현한 논문이다. 사용한 비동기 FSK 수신기는 심볼 검출을 하기 위해 16점 FFT를 이용하며 데이터의 주 속도는 10kbps이고 도플러에 강인하고 전력 효율과 신뢰성을 얻기 위해 디지털 회로로 설계된다. 또한 CORDIC 알고리듬을 이용하여 FFT 연산 시 사용되는 복소 승산을 가산기 및 천이기로 대체하여 저전력화 하였다. 설계 시스템의 검증을 하기 위해 먼저 Simulink로 시뮬레이션 하여 성능을 확인하고Xilinx사의 System Generator를 이용하여 FPGA 구현하여 성능을 비교 검증하였다. 결과적으로 Simulink 결과와 FPGA 구현 결과가 표6과 표7에 의해 잘 일치함을 확인하였다.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1241-1250
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• 2019

This study mainly focused on the neutronics modeling of bubbles in bubbly flow in boiling water reactors. The bubble, ring and homogenous models were used for radial void fraction distribution. Effect of the bubble and ring models on the infinite multiplication factor and two-group flux distribution was investigated by comparing with the homogenous model. Square pitch unit cell geometry was used in the calculations. In the bubble model, spherical and non-spherical bubbles at random positions, sizes and shapes were produced by Monte Carlo method. The results show that there are significant differences among the proposed models from the viewpoint of physical interaction mechanism. For the fully-developed bubbly flow, $k_{inf}$ is overestimated in the ring model by about $720{\pm}6pcm$ with respect to homogeneous model whereas underestimated in the bubble model by about $-65{\pm}9pcm$ with a standard deviation of 15 pcm. In addition, the ring model shows that the coolant must be separated into regions to properly represent the radial void distribution. Deviations in flux distributions principally occur in certain regions, such as corners. As a result, the bubble model in modeling the void fraction can be used in nuclear engineering calculations.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1479-1486
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• 2019

The purpose of the present study is to develop the 3D static and noise simulator based on Galerkin Finite Element Method (GFEM) using the unstructured hexahedral elements. The 3D, 2G neutron diffusion and noise equations are discretized using the unstructured hexahedral by considering the linear approximation of the shape function in each element. The validation of the static calculation is performed via comparison between calculated results and reported data for the VVER-1000 benchmark problem. A sensitivity analysis of the calculation to the element type (unstructured hexahedral or tetrahedron elements) is done. Finally, the neutron noise calculation is performed for the neutron noise source of type of variable strength using the Green function technique. It is shown that the error reduction in the static calculation is considerable when the unstructured tetrahedron elements are replaced with the hexahedral ones. Since the neutron flux distribution and neutron multiplication factor are appeared in the neutron noise equation, the more accurate calculation of these parameters leads to obtaining the neutron noise distribution with high accuracy. The investigation of the changes of the neutron noise distribution in axial direction of the reactor core shows that the 3D neutron noise analysis is required instead of 2D.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.764-769
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• 2022

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

• The Plant Pathology Journal
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• 제39권5호
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• pp.417-429
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• 2023

Ralstonia solanacearum species complex (RSSC) is a soil borne plant pathogen causing bacterial wilt on various important crops, including Solanaceae plants. The bacterial pathogens within the RSSC produce exopolysaccharide (EPS), a highly complicated nitrogencontaining heteropolymeric polysaccharide, as a major virulence factor. However, the biosynthetic pathway of the EPS in the RSSC has not been fully characterized. To identify genes in EPS production beyond the EPS biosynthetic gene operon, we selected the EPS-defective mutants of R. pseudosolanacearum strain SL341 from Tn5-inserted mutant pool. Among several EPSdefective mutants, we identified a mutant, SL341P4, with a Tn5-insertion in a gene encoding a putative NDP-sugar epimerase, a putative membrane protein with sugar-modifying moiety, in a reverse orientation to EPS biosynthesis gene cluster. This protein showed similar to other NDP-sugar epimerases involved in EPS biosynthesis in many phytopathogens. Mutation of the NDP-sugar epimerase gene reduced EPS production and biofilm formation in R. pseudosolanacearum. Additionally, the SL341P4 mutant exhibited reduced disease severity and incidence of bacterial wilt in tomato plants compared to the wild-type SL341 without alteration of bacterial multiplication. These results indicate that the NDP-sugar epimerase gene is required for EPS production and bacterial virulence in R. pseudosolanacearum.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.149-154
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• 1998

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute(KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons from spallation reactions are essentially required for operating the reactor in its steady state. furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance of the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases.