Conference Agenda

Argonne is participating in the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)”. The NACIE loop includes a fuel pin simulator section which is a hexagonal array of 19 wire-wrapped, electrically heated pins and uses lead-bismuth eutectic as a working fluid. Argonne’s work on the CRP includes CFD simulations with the NekRS and Cardinal codes. Both LES and RANS turbulence models were used in NekRS coupled via Cardinal to a solid conduction model to account for conjugate heat transfer. Via comparisons against experimental measurements from the NACIE tests, these benchmark simulations are being performed to expand the validation basis of these codes. The objective of this paper is to present recent progress on NACIE test simulations which cover both forced and mixed convection conditions with uniform and skewed heating profiles. Results from simulations will be compared to available experimental data.

CFD becomes more and more important in nuclear reactor safety considerations. For multiphase flows in the related medium and large scales the Euler-Euler approach is most frequently used and often the only feasible one. In many flow situations, the involved interfaces cover a wide range of scales leading to different coexisting morphologies. Established simulation methods differ for the different interfacial scales. Large interfaces are represented in a resolved manner usually basing on the one fluid approach, e.g. Volume of Fluid (VOF) or Level Set. Unresolved (dispersed) flows are modelled using the two- or multi-fluid approach. A simulation method that requires less knowledge about the flow in advance would be desirable and should allow describing both interfacial structures – resolved and unresolved – in a single computational domain.

The morphology adaptive multifield two-fluid model MultiMorph, which is developed at HZDR based on the software from the OpenFOAM Foundation, is able to handle unresolved and resolved interfacial structures coexisting in the computational domain with the same set of equations. An interfacial drag formulation for large interfacial structures is used to describe them in a VOF-like manner, while the usual closure models are applied for the unresolved phases. In addition, MultiMorph allows to simulate transitions between the morphologies. This concerns both empirical transitions such as entrainment and detrainment as well as transitions resulting from a change in the size of the numerical mesh within the domain. The basic framework including the handling of transitions between the morphologies will be presented.

Blockage in a reactor fuel assembly is considered to be one of the most important accidents that should be analyzed in detail. A variety of factors can contribute to the occurrence of such an accident, among which are fuel element bending or local deformation and swelling of the cladding. The reduction of coolant flow area can cause local heat transfer deterioration and temperature augmentation, which can further lead to dry-out and possible loss of fuel assembly integrity. It is challenging to evaluate the consequence of flow blockage accident due to lack of knowledge about local flow parameters as well as their response mechanism, especially when two-phase flows are concerned. Moreover, due to negligible influence on global mass flow, it is difficult to detect the accident through protection system. Owing to the availability of advanced computer systems, simulation using either system or CFD codes has become an important tool in assisting the analysis of these local phenomena and evaluation of their impact on the safe operation of nuclear reactors. This study presents a CFD study of three vertical pipe bubbly flow cases, one empty pipe, one with a ring obstacle and one with a baffle obstacle, and both obstacles block a half of the flow area. The focus is put on analyzing the effect of blockage on the velocity and turbulence field as well as the development of bubble size. Different turbulence models and mechanisms leading to bubble coalescence and breakup are discussed and evaluated with the aid of experimental data.

During severe nuclear accidents, hydrogen generated in the reactor core can mix with air to form potentially explosive mixtures. The severity of such explosions depends on the mixture composition and the combustion regime. In highly turbulent conditions, combustion is significantly accelerated, and simulations have shown that heat loss has minimal impact on flame propagation due to its slower rate relative to combustion. However, in the case of slower deflagration, heat loss is expected to play a more significant role in flame evolution.

This study focuses on modeling the effects of conductive and radiative heat losses in a slow hydrogen-air-steam combustion scenario during the HD-22 experiment at the THAI experimental facility. Unsteady Reynolds-Averaged Navier-Stokes (RANS) simulations were conducted using computational fluid dynamics software OpenFOAM software and combustion model flameFoam. Heat loss was modeled through conductive heat transfer to an isothermal wall and the P1 model for radiative transfer.

Simulation results showed good agreement with experimental data, indicating that including heat loss mechanisms slightly delayed the completion of combustion and slighlty reduced the maximum overpressure, as well as slowed down the vertical flame propagation. Results show that the conductive and radiative heat loss contribute similarly to the total heat loss, emphasizing radiative heat loss importance in modeling slow combustion.

Overall, the study highlights the critical role of heat loss, particularly radiative heat transfer, in accurately simulating slow hydrogen explosions.

The pressure drop of PWR fuel assemblies has an importance in the core itself (global flowrate, flow distribution, hydrodynamic forces) but also in the storage pools and in the different cells or casks in which it may be stored. The behavior in nominal conditions is well known and has been largely experimentally and numerically investigated, however the characteristics at much lower Reynolds numbers are less studied.

The first objective of this paper is to discuss the adequate turbulence models for those low turbulence situations. Available experiments provide useful data of pressure drops and velocity of different elements in a large range of Reynolds numbers (device made of a typical fuel bundle (17x8) with 4 mixing and supporting grids, at full scale). CFD computations with RANS, URANS, Detached Eddy Simulations and Large Eddy Simulation turbulence models are performed and compared with measurements. The modeling of the wall friction is also discussed. The computational meshes are obtained from an automatic process and are based on polyhedrons.

The second objective is the influence of the domain modeled in such quasi-periodic configurations and the associated boundary conditions. The different periodicity conditions are particularly investigated on 2x2 simplified models, with the transverse effect of mixing vanes.

The commercial Ansys-Fluent and Star-CCM+ codes are both used and conclusions are drawn on the two objectives mentioned above.

Finally, this paper concludes on the recommended setup in order to provide industrial values of pressure drops of the components of the fuel assembly in the different off-reactor situations.

The Forschungs-Neutronenquelle Heinz Meier-Leibnitz (FRM II) is actively contributing to global efforts to reduce the use of Highly Enriched Uranium (HEU) in civilian nuclear applications. A key step in this initiative is converting the current fuel system to a high-density Low Enriched Uranium (LEU) fuel. In 2023, a feasibility study demonstrated the scientific viability of converting the FRM II to U-10Mo LEU, while maintaining FRM II scientific performance. Due to the changed plate design, many LEU designs exhibit lower pressure drops through the fuel element. To minimize the impact to the FRM II reactor, we aim to have a similar pressure drop across the fuel element in the forward flow direction than for today’s HEU core and a lower pressure drop required in reverse to mitigate impacts during transient scenarios. To meet these demands, the previously published design incorporated a flow restrictor downstream of the fuel element to increase the pressure drop. In the current study, an alternative approach is explored by thickening the cladding at the end of the fuel plates. Various cladding thicknesses are evaluated, and the resulting pressure drops are computed using Computational Fluid Dynamics (CFD). Additionally, the influence of rounded fuel plate caps is quantified.