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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Nuclear Engineering and Technology
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Korean Nuclear Society
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Volume & Issues
Volume 44, Issue 8 - Dec 2012
Volume 44, Issue 7 - Oct 2012
Volume 44, Issue 6 - Aug 2012
Volume 44, Issue 5 - Jun 2012
Volume 44, Issue 4 - May 2012
Volume 44, Issue 3 - Apr 2012
Volume 44, Issue 2 - Mar 2012
Volume 44, Issue 1 - Feb 2012
Selecting the target year
ADVANCES IN MULTI-PHYSICS AND HIGH PERFORMANCE COMPUTING IN SUPPORT OF NUCLEAR REACTOR POWER SYSTEMS MODELING AND SIMULATION
Turinsky, Paul J. ;
Nuclear Engineering and Technology, volume 44, issue 2, 2012, Pages 103~122
DOI : 10.5516/NET.01.2012.500
Significant advances in computational performance have occurred over the past two decades, achieved not only by the introduction of more powerful processors but the incorporation of parallelism in computer hardware at all levels. Simultaneous with these hardware and associated system software advances have been advances in modeling physical phenomena and the numerical algorithms to allow their usage in simulation. This paper presents a review of the advances in computer performance, discusses the modeling and simulation capabilities required to address the multi-physics and multi-scale phenomena applicable to a nuclear reactor core simulator, and present examples of relevant physics simulation codes' performances on high performance computers.
PROSPECTS IN DETERMINISTIC THREE-DIMENSIONAL WHOLE-CORE TRANSPORT CALCULATIONS
Sanchez, Richard ;
Nuclear Engineering and Technology, volume 44, issue 2, 2012, Pages 113~150
DOI : 10.5516/NET.01.2012.501
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.
CHALLENGES AND PROSPECTS FOR WHOLE-CORE MONTE CARLO ANALYSIS
Martin, William R. ;
Nuclear Engineering and Technology, volume 44, issue 2, 2012, Pages 151~160
DOI : 10.5516/NET.01.2012.502
The advantages for using Monte Carlo methods to analyze full-core reactor configurations include essentially exact representation of geometry and physical phenomena that are important for reactor analysis. But this substantial advantage comes at a substantial cost because of the computational burden, both in terms of memory demand and computational time. This paper focuses on the challenges facing full-core Monte Carlo for keff calculations and the prospects for Monte Carlo becoming a routine tool for reactor analysis.
MCCARD: MONTE CARLO CODE FOR ADVANCED REACTOR DESIGN AND ANALYSIS
Shim, Hyung-Jin ; Han, Beom-Seok ; Jung, Jong-Sung ; Park, Ho-Jin ; Kim, Chang-Hyo ;
Nuclear Engineering and Technology, volume 44, issue 2, 2012, Pages 161~176
DOI : 10.5516/NET.01.2012.503
McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback,
theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.
FAST REACTOR PHYSICS AND COMPUTATIONAL METHODS
Yang, W.S. ;
Nuclear Engineering and Technology, volume 44, issue 2, 2012, Pages 177~198
DOI : 10.5516/NET.01.2012.504
This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.
PHYSICS OF AMERICIUM TRANSMUTATION
Wallenius, Janne ;
Nuclear Engineering and Technology, volume 44, issue 2, 2012, Pages 199~206
DOI : 10.5516/NET.01.2012.505
Using fast neutron Generation IV reactors, recycling of americium and curium may become feasible. The detrimental impact of americium on safety parameters has recently been quantified in terms of a power penalty for surviving a given set of transients in sodium fast reactors. In the present paper, a review of the physical reasons for the adverse effect of americium is provided, and different Gen-IV technologies are assessed with respect to their capability of hosting americium in the fuel.
ON SOME OUTSTANDING PROBLEMS IN NUCLEAR REACTOR ANALYSIS
Cho, Nam-Zin ;
Nuclear Engineering and Technology, volume 44, issue 2, 2012, Pages 207~224
DOI : 10.5516/NET.01.2012.506
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.