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RISS 인기검색어
- 검색키워드
- 저자 : PAUL J. TURINSKY (검색결과 5 건)
- 무료
- 기관 내 무료
- 유료
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Evolution of Nuclear Fuel Management and Reactor Operational Aid Tools
PAUL J. TURINSKY,PAUL M. KELLER,HANY S. ABDEL-KHALIK 한국원자력학회 2005 Nuclear Engineering and Technology Vol.37 No.1
In this paper are reviewed the current status of nuclear fuel management and reactor operational aid tools. In addition, we indicate deficiencies in current capabilities and what future research is judged warranted. For the nuclear fuel management review the focus is on light water reactors and the utilization of stochastic optimization methods applied to the lattice, fuel bundle, core loading pattern, and for BWRs the control rod pattern/core flow design decision making problems. Significant progress in addressing separately each of these design problems on a single cycle basis is noted; however, the outstanding challenge of addressing the integrated design problem over multiple cycles under conditions of uncertainty remains to be addressed. For the reactor operational aid tools review the focus is on core simulators, used to both process core instrumentation signals and as an operator aid to predict future core behaviors under various operational strategies. After briefly reviewing the current status of capabilities, a more in depth review of adaptive core simulation capabilities, where core simulator input data are adjusted within their known uncertainties to improved agreement between prediction and measurement, is presented. This is done in support of the belief that further development of adaptive core simulation capabilities is required to further significantly advance the utility of core simulators in support of reactor operational aid tools.
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PAUL J. TURINSKY 한국원자력학회 2012 Nuclear Engineering and Technology Vol.44 No.2
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.
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EVOLUTION OF NUCLEAR FUEL MANAGEMENT AND REACTOR OPERATIONAL AID TOOLS
TURINSKY PAUL J.,KELLER PAUL M.,ABDEL-KHALIK HANY S. Korean Nuclear Society 2005 Nuclear Engineering and Technology Vol.37 No.1
In this paper are reviewed the current status of nuclear fuel management and reactor operational aid tools. In addition, we indicate deficiencies in current capabilities and what future research is judged warranted. For the nuclear fuel management review the focus is on light water reactors and the utilization of stochastic optimization methods applied to the lattice, fuel bundle, core loading pattern, and for BWRs the control rod pattern/core flow design decision making problems. Significant progress in addressing separately each of these design problems on a single cycle basis is noted; however, the outstanding challenge of addressing the integrated design problem over multiple cycles under conditions of uncertainty remains to be addressed. For the reactor operational aid tools review the focus is on core simulators, used to both process core instrumentation signals and as an operator aid to predict future core behaviors under various operational strategies. After briefly reviewing the current status of capabilities, a more in depth review of adaptive core simulation capabilities, where core simulator input data are adjusted within their known uncertainties to improved agreement between prediction and measurement, is presented. This is done in support of the belief that further development of adaptive core simulation capabilities is required to further significantly advance the utility of core simulators in support of reactor operational aid tools.
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Turinsky, Paul J. Korean Nuclear Society 2012 Nuclear Engineering and Technology Vol.44 No.2
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.
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Surrogate based model calibration for pressurized water reactor physics calculations
Bassam A. Khuwaileh,PAUL J. TURINSKY 한국원자력학회 2017 Nuclear Engineering and Technology Vol.49 No.6
In this work, a scalable algorithm for model calibration in nuclear engineering applications is presentedand tested. The algorithm relies on the construction of surrogate models to replace the original modelwithin the region of interest. These surrogate models can be constructed efficiently via reduced ordermodeling and subspace analysis. Once constructed, these surrogate models can be used to performcomputationally expensive mathematical analyses. This work proposes a surrogate based model calibrationalgorithm. The proposed algorithm is used to calibrate various neutronics and thermal-hydraulicsparameters. The virtual environment for reactor applications-core simulator (VERA-CS) is used tosimulate a three-dimensional core depletion problem. The proposed algorithm is then used to constructa reduced order model (a surrogate) which is then used in a Bayesian approach to calibrate the neutronicsand thermal-hydraulics parameters. The algorithm is tested and the benefits of data assimilationand calibration are highlighted in an uncertainty quantification study and requantification after thecalibration process. Results showed that the proposed algorithm could help to reduce the uncertainty inkey reactor attributes based on experimental and operational data.
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