CTF and FLOCAL Thermal Hydraulics Validations and Verifications within a Multiscale and Multiphysics Software Development

Simulation codes allow one to reduce the high conservativism in nuclear reactor design improving the reliability and sustainability associated with nuclear power. Full-core coupled reactor physics at the rod level are not provided by most simulation codes. This has led in the UK to the development of a multiscale and multiphysics software development focused on LWRS. In terms of the thermal hydraulics, simulation codes suitable for this multiscale and multiphysics software development include the subchannel code CTF and the thermal hydraulics module FLOCAL of the nodal code DYN3D. In this journal article, CTF and FLOCAL thermal hydraulics validations and verifications within the multiscale and multiphysics software development have been performed to evaluate the accuracy and methodology available to obtain thermal hydraulics at the rod level in both simulation codes. These validations and verifications have proved that CTF is a highly accurate subchannel code for thermal hydraulics. In addition, these verifications have proved that CTF provides a wide range of crossflow and turbulent mixing methods, while FLOCAL in general provides the simplified no-crossflow method as the rest of the methods were only tested during its implementation into DYN3D.

Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark – re-connection of an isolated loop

The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre “Kurchatov Institute” (with the code ATHLET/BIPR-VVER), ÚJV Řež (with the code RELAP5-3D©) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

The reactor dynamics code DYN3D

The article provides an overview on the code DYN3D which is a three-dimensional core model for steady-state, dynamic and depletion calculations in reactor cores with quadratic or hexagonal fuel assembly geometry being developed by the Helmholtz-Zentrum Dresden-Rossendorf for more than 20 years. The current paper gives an overview on the basic DYN3D models and the available code couplings. The verification and validation status is shortly outlined. The paper concludes with the current developments of the DYN3D code. For more detailed information the reader is referred to the publications cited in the corresponding chapters.