Development of Parallel Coupling System between Three-Dimensional Nodal Kinetic Code ENTREE and Two-Fluid Plant Simulator TRAC/BF1
スポンサーリンク
概要
- 論文の詳細を見る
The high-speed three-dimensional neutron kinetic code ENTRÉE was developed based on the polynomial and semi-analytical nonlinear iterative nodal methods (PNLM and SANLM) with also introducing the discontinuity factor. In order to enhance the efficiency of transient calculation, the nonlinear correction-coupling coefficients are intermittently updated based on the changing rate of core state variables. By giving the analytical form for two-node problem matrix elements, the additional computing time in SANLM was minimized. A fast algorithm was developed for the multi table macro-cross section rebuilding process. The reactivity component model was implemented based on the variation of the neutron production and destruction terms. The code was coupled with the two-fluid thermal hydraulic plant simulator TRAC/BF1 through PVM or MPI protocols. Two codes are executed in parallel with exchanging the feedback parameters explicitly. Based on the LMW PWR transient benchmark, it was shown that both PNLM and SANLM spend less than 20% excess computing time in comparison with the coarse mesh finite difference method (CFDM). The implementation of the discontinuity factor was verified based on the DVP problem. Adequacy and parallel efficiency of the coupling system TRAC/BF1-ENTRÉE was demonstrated based on the BWR cold water injection transient proposed by NEA/CRP.
- 一般社団法人 日本原子力学会の論文
- 2000-10-25
著者
-
Ninokata Hisashi
Research Laboratory For Nuclear Reactors Tokyo Institute Of Technology
-
Ninokata Hisashi
Research Lab. For Nuclear Reactors Tokyo Institute Of Technology
-
Hotta Akitoshi
In-core Fuel Management Department Tepco Systems Corporation
-
BARATTA Anthony
Nuclear Engineering Department, The Pennsylvania State University
-
Baratta Anthony
Nuclear Engineering Department The Pennsylvania State University
関連論文
- Numerical Studies on Dynamic Behavior of Air-Water Cross Flow Between Two Circular Interconnected Channels
- Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel
- Computational Fluid Dynamics Study of Liquid Droplet Impingement Erosion in the Inner Wall of a Bent Pipe
- PRA-Based SMA : the First Tool toward a Risk-Informed Approach to the Seismic Design of the IRIS
- Prediction of Release Rate of Burnt Sodium as Aerosol
- Proposal of an Effective Initial Guess Scheme for Higher Harmonic Mode Calculation of Boiling Water Reactor Cores
- Analysis of an Out-of-pile Experiment for Materials Redistribution under Core Disruptive Accident Condition of Fast Breeder Reactors
- Prediction of the Equilibrium Two-Phase Flow Distributions in Inter-Connected Subchannel Systems
- Analytical Study on Detailed Void Distributions Inside BWR Fuel Bundle under Turbine Trip Event Considering Time-Dependent Pin Power Distributions
- Concept of Erbium Doped Uranium Oxide Fuel Cycle in Light Water Reactors
- Second International Symposium on Global Environment and Nuclear Energy Systems
- Numerical Method for Simulation of Fluid Flow and Heat Transfer in Geometrically Disturbed Rod Bundles
- Numerical Study on Observed Decay Ratio of Coupled Neutronic-Thermal Hydraulic Instability in Ringhals Unit 1 under Random Noise Excitation
- An Algorithm for Attenuation of Turbulence in Particulate Flow Linked to the Fluid-dynamic Code COMMIX-M
- Development of Parallel Coupling System between Three-Dimensional Nodal Kinetic Code ENTREE and Two-Fluid Plant Simulator TRAC/BF1
- The Multi-Fluid Multi-Phase Subchannel Analysis Code KAMUI for Subassembly Accident Analysis of an LMFR
- Nonlinear Iterative Nodal Method Applied to Neutron Flux Modal Analysis