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).


Go to RAST-K v2 Page

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.


Go to MCS Page

 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.



 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.


Go to RXSP Page

 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).


Go to MPCORE Page

 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.