CFD/Monte-Carlo neutron transport coupling scheme, application to TRIGA reactor

Development of a new steady-state thermal hydraulic and safety analysis code, OpenTHY, for a TRIGA MARK II research reactor

This work presents the development of a new powerful computer code OpenTHY for thermal hydraulic safety analysis of TRIGA type research reactor in the case of one-phase flow around the fuel pins. This code is used for preparing a very detailed thermal-hydraulic model for the Moroccan reactor to study the efficiency of core cooling in natural convection mode and to ensure that the reactor is operating in a safely manner respecting all safety limits. The OpenTHY code has also the possibility to take into account the presence of the Zirconium rod that is inserted into an annular fuel rod, which is essential to avoid cladding failure due to the hydrogen overpressure. The fuel element was discretized into multiple axial sections, with each section further subdivided radially into various subdivisions, covering four distinct regions: cladding, gap, active fuel, and zirconium rod. The active part was divided radially into a significant number of subdivisions allowing for detailed spatial distribution of the power peaking factors in both axial and radial directions. Then, we calculated different thermal-hydraulic safety parameters using the single channel model and specific correlations related to heat transfer coefficients and critical heat flux to determine the temperature profiles of coolant and fuel element in different axial and radial locations, the critical heat flux (CHF) and the departure from nucleate boiling ratio (DNBR) in each axial subdivision of the hottest channel of the core. Additional core configurations were evaluated and the results obtained by the OpenTHY code were validated through a comparison with experimental measures of temperature inside two instrumented fuel elements at various power levels. This study presents also the results calculated by PARET/ANL code. From these output data, the safety of the reactor is ensured in the conditions of natural convection-cooling and the maximum temperature profiles at the cladding outer surface and at the inner surface of the active fuel rod remain largely far from safety limits prescribed in TRIGA safety analysis report (SAR). In addition, the MDNBR presents a value widely higher than the design limit of 1.3 which guarantees the safe operation of the 2MW TRIGA MARK II under normal conditions.

An Efficient Scheme for Coupling OpenMC and FLUENT with Adaptive Load Balancing

This paper develops a multi-physics interface code MC-FLUENT to couple the Monte Carlo code OpenMC with the commercial computational fluid dynamics code ANSYS FLUENT. The implementations and parallel performances of block Gauss–Seidel-type and block Jacobi-type Picard iterative algorithms have been investigated. In addition, this paper introduces two adaptive load-balancing algorithms into the neutronics and thermal-hydraulics coupled simulation to reduce the time cost of computation. Considering that the different scalability of OpenMC and FLUENT limits the performance of block Gauss–Seidel algorithm, an adaptive load-balancing algorithm that can increase the number of nodes dynamically is proposed to improve its efficiency. Moreover, with the natural parallelism of block Jacobi algorithm, another adaptive load-balancing algorithm is proposed to improve its performance. A 3 x 3 PWR fuel pin model and a 1000 MWt ABR metallic benchmark core were used to compare the performances of the two algorithms and verify the effectiveness of the two adaptive load-balancing algorithms. The results show that the adaptive load-balancing algorithms proposed in this paper can greatly improve the computing efficiency of block Jacobi algorithm and improve the performance of block Gauss–Seidel algorithm when the number of nodes is large. In addition, the adaptive load-balancing algorithms are especially effective when a case demands different computational power of OpenMC and FLUENT.

Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations

Multiphysics coupling of neutronics/thermal-hydraulics models is essential for accurate modeling of nuclear reactor systems with physics feedback. In this work, SCALE/TRACE coupling is used for neutronic analysis and spent fuel validation of BWR assemblies, which have strong coolant feedback. 3D axial power profiles with coolant feedback are captured in these advanced simulations. The methodology is applied to two BWR assemblies (2F2DN23/SF98 and 2F2D1/F6), discharged from the Fukushima Daini-2 unit. Coupling is performed externally, where the SCALE/T5-DEPL module transfers axial power data in all axial nodes to TRACE, which in turn calculates the coolant density and temperature for each of these nodes. Within a burnup step, the data exchange process is repeated until convergence of all coupling parameters (axial power, coolant density, and coolant temperature) is observed. Analysis of axial power, criticality, and coolant properties at the assembly level is used to verify the coupling process. The 2F2D1/F6 benchmark seems to have insignificant void feedback compared to 2F2DN23/SF98 case, which experiences large power changes during operation. Spent fuel isotopic data are used to validate the coupling methodology, which demonstrated good results for uranium isotopes and satisfactory results for other actinides. This work has a major challenge of lack of documented data to build the coupled models (boundary conditions, control rod history, spatial location in the core, etc.), which encourages more advanced methods to approximate such missing data to achieve better modeling and simulation results.