STREAM
Go to STREAM 2D Page
Go to STREAM 3D Page
A neutron transport analysis code, STREAM (Steady state and Transient REactor Analysis code with Method of Characteristics), has been developed to perform a whole LWR core calculation with the direct transport analysis method and the two-step method. Numerous advanced features, especially resonance treatment methods, have been developed and implemented in the STREAM code for higher accuracy and performance. STREAM with the advanced methods has order of ~100 pcm accuracy in LWR analyses. STREAM has capabilities to analyze the whole LWR core through the two-step (with PARCS or RAST-K v2) method and direct transport method (2-D and 3-D).
RAST-K
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Go to RAST-K v3 Page
The RAST-K v2 code is a recently developed three-dimensional two-group PWR core analysis code by UNIST CORE lab. It aims to be used by utilities to perform in-core fuel management studies, core design calculations, load follow simulation and transient analysis in neutronics. The goal to develop the RAST-K was to introduce some new methods in the developments in the past decade for better performance and more general utilizations. The new kernels are more sophisticated than those used in the previous version as RAST-K. The RAST-K v2 code is capable of performing steady-state, quasi transient and transient calculations by solving the two-group three-dimensional neutron diffusion equation in eigenvalue or fixed source modes. From the RAST-K v3, it is possible to perform pin by pin calculation based on SP3neutron transport equation for whole core analysis. The non-linear iteration scheme is used as the basis calculating scheme with optimized computational flow.
MCS
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A Monte Carlo code MCS has being developed at Ulsan National Institute of Science and Technology (UNIST) since 2013. The target of MCS is to solve complex whole core problems like BEAVRS. MCS can treat the 3D whole core geometry with universe and lattice, and the neutron physics with probability-table, free-gas treatment, S(a,b) and Doppler Broadening Rejection Correction.
GREAPMC
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GREAPMC (GPU-optimized REActor Physics Monte Carlo) is a specialized Monte Carlo code designed for the efficient criticality simulations of pressurized water reactors (PWRs) using GPU acceleration. The primary objective behind developing GREAPMC is to conduct comprehensive whole-core calculations within feasible timeframes. Leveraging state-of-the-art GPU acceleration techniques, GREAPMC enables the execution of large-scale reactor physics simulations for full-scale PWRs within practical time constraints and resource limitations.GREAPMC (GPU-optimized REActor Physics Monte Carlo) is a specialized Monte Carlo code designed for the efficient criticality simulations of pressurized water reactors (PWRs) using GPU acceleration. The primary objective behind developing GREAPMC is to conduct comprehensive whole-core calculations within feasible timeframes. Leveraging state-of-the-art GPU acceleration techniques, GREAPMC enables the execution of large-scale reactor physics simulations for full-scale PWRs within practical time constraints and resource limitations.
STREAM-SNF
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A radiation source term capability has been implemented in STREAM to perform SNF characterization, cask dose rate analysis, for waste management, radiological safety and burnup credit applications.
RXSP
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The accuracy of the nuclear cross section data is a prerequisite for the accuracy of reactor neutron transport calculations, i.e. MC method. RXSP is a nuclear Cross Section Processing code being originally developed by REAL group, Department of Engineering Physics, Tsinghua University, which is mainly intended to reactor analysis. The current version is RXSP-Beta 2.0 released in August,2013 domestically in China Mainland. The Beta3.0 version is being developed jointly by UNIST and Tsinghua University(per the agreement of Prof. Lee and Prof. WANG).
MPCORE
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Nowadays multi-physics simulation attracts a lot of attention from nuclear researchers worldwide since it is able to produce more realistic results in terms of reactor core safety margins against critical core conditions. The analysis of non-quantified uncertainties on account of multi-physics phenomena involves the coupled modeling of neutron kinetics, coolant thermal-hydraulics and nuclear fuel performance using the numerical integration methods with built-in precision and accuracy control. A new reactor core multi-physics system has been developed to meet the control precision criterion and to facilitate the transparency of the coupling procedure using the external loose coupling approach. The new code implements an adaptive time step to achieve a solution of a prescribed tolerance, the restart capability to maintain sustainability of numerical simulation, the random sampling method for uncertainty quantification, and the lossy compression algorithm for output data size optimization. The present configuration of the multi-physics system addresses the two-step core neutronics approach with a method-of-characteristic cross-section code and a nodal diffusion solver aided by a pin-by-pin power reconstruction module.