MPCORE

Download     MPCORE_CODE   MPCORE_BEAVRS

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 Choi, 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/10.1016/j.pnucene.2020.103441 (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)

Introduction

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.

Features

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 dimesional fuel performance codes FRAPCON and FRAPTRAN

Multi-physics Analysis

– Reactor core depletion, transient and accidents simulation
– Dynamic 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

Application