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The Recommended Publication for Citing
  1. Alexey Cherezov, Jaerim Jang, Deokjung Lee*, “A PCA Compression Method for Reactor Core Transient Multi-Physics Simulation”, Prog. Nucl. Energy, 28:103441, https://doi.org/10.1016/j.pnucene.2020.103441 (2020)
  2. Alexey Cherezov, Jinsu Park, Hanjoo Kim, Jiwon Choe, Deokjung Lee*, “A Multi-physics Adaptive Time Step Coupling Algorithm for Light-Water Reactor Core Transient and Accident Simulation”, Energies, 13: 6374 https://doi.org/10.3390/en13236374 (2020)
  3. Cherezov A., Kim H., Park J., Lee D. Fuel Rod Analysis Programming Interface for a Loosely Coupled Multiphysics System. American Nuclear Society Winter Meeting, November 16 - 19, 2020, vol. 123, pp. 1331-1334
  4. Alexey Cherezov, Hanjoo Kim, Jinsu Park, Deokjung Lee*, “MPCORE Code for OPR-1000 Transient Multiphysics Simulation with Adaptive Step Size Control”, ANS Summer Meeting, USA, Jun 8-11 (2020)
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.

Constituent Modules
Two-group cross-section library calculated by code STREAM
Two-group nodal diffusion code RAST-K 2.0 with pin-by-pin power reconstruction
– Homogeneous two-phase coolant T/H code CTH1D
– One dimensional fuel performance codes FRAPCON and FRAPTRAN

Multi-physics Analysis
Reactor core depletion, transient and accidents simulation
Dynamical pellet-to-cladding gap heat transfer
– Fuel swelling, densification, thermal expansion and relocation
Cladding creep, elastic and plastic deformations
Cladding hydrogen pickup and corrosion
Pellet-cladding mechanical interaction and cladding ballooning models

Coupling interface
External loose coupling algorithm for interchangeable modules
Damped Picard iterations with Gauss-Seidel acceleration
Adaptive time step based on the step-doubling approach
Time step rejection and restart capability for robustness improvement

Output Data Processing
High resolution multi-physics data
Storage in HDF5 format
PCA compression algorithm

Uncertainty Quantification
Error propagation by random sampling
Nuclear data and core parameters uncertainties