Conference Agenda

New small modular reactor technologies are being developed, having in common innovative compact design and a reliance on passive safety systems for enhanced safety. In the framework of the H2020 European project ELSMOR, a generic pressurised water compact design was defined for safety study purposes. A model for this study design has been built with the ASTEC V3.1.2 code and includes a passive heat removal system loop (PHRS) that should ensure the extraction of the reactor residual heat during an accident. The plant response during a station black-out scenario has been investigated when the safety system is active. While the system is able to extract the heat and keep the core covered, pressure in the loop is shown to be highly dependent on the modelling of the upper plenum area. During the PHRS operation, flow instabilities could be observed in the reactor primary loop. The mechanisms leading to the triggering and stopping of these oscillations in the calculated flow are analysed. The sensitivity of the global plant behaviour to different geometrical parameters of the safety system such as pipes diameter is also studied. As long as the PHRS can extract the residual heat, a very similar plant response is observed. The last investigated parameter is the passive loop filling ratio. This parameter is shown to have only a small impact on the heat removal capacity of the loop but can influence oscillatory flow development in the PHRS secondary loop.

Passive systems are being considered for advanced reactor designs owing to their enhanced reliability against an extended loss of onsite power. Particularly, the Safety Condenser (SACO) stands out because of its capacity of passively removing core decay heat through the steam generators by condensing steam inside a heat exchanger immersed in an external water pool. This work, embedded within the PASTELS European Project, presents experimental and numerical results on SACO performance at the PKL facility.

The data obtained in this study concerns a vertical straight-tube SACO immersed in a water pool. The SACO is connected to the secondary side of the PKL facility through charge and return lines. The P1.2 test consist of several secondary-side depressurization sequences aiming at studying the dependency of SACO power removal on secondary side pressure and nitrogen initial mass, pool water temperature and straight-tube liquid level. These results are then compared to the predictions of the CATHARE-3, ATHLET, TRACE and RELAP-5 system thermalhydraulic codes.

Experimental results on P1.2 experiments show the possibility to limit the SACO power removal by controlling the fill level in the SACO straight tubes, which is important for accident management purposes. Concerning simulation results, system thermalhydraulic codes generally overestimate depressurisation rates. In some cases, deviations can be ascribed to an inadequate modelling of a specific component (e.g. auxiliary heater, nitrogen injection), while in other cases they are related to the modelling of the pool and the difficulty of capturing the redistribution of nitrogen within the straight tubes along the transient.

Passive safety systems are increasingly utilized in prospective nuclear power plant designs. The low magnitude of the forces involved in such systems, combined with the uncertainty inherent in the factors affecting them, poses a problem in assessing their reliability compared to active counterparts.

The purpose of this paper is to investigate and apply a state-of-the-art technique in passive reliability assessment, known as the Reliability Methods of Passive Systems (RMPS) methodology, to the isolation condenser system (ICS) of the prospective BWRX-300 small modular reactor (SMR) design. The ICS is a safety system driven by natural circulation that provides emergency core cooling and pressure control for the BWRX-300. Using RMPS to analyze the effect of uncertainties in (a) the thermal characteristics of the fuel and (b) two-phase constitutive correlation factors on ICS operation, the reliability of natural circulation was quantified with a confidence of 95%, yielding an immeasurably small failure probability. Considering residual uncertainty, an engineering judgment assigned a failure probability of 1.00E-07. This finding was integrated into a fault tree analysis of the ICS using failure mode and effect analysis (FMEA) of system components, including insufficient natural circulation as a failure mode. Analysis of sequences leading to failure resulted in system unavailability being determined as 1.62E-07 for the case of all three loops initially available and 2.91E-05 for the case when only two loops are initially available. Sensitivity analysis of the natural circulation failure probability with respect to ICS system unavailability was also performed to investigate the robustness of the design.

A numerical simulation study of a Two-Phase Loop Thermosiphon (TPLT) in a novel Small Modular Integrated Heating Reactor (SMIHR) is conducted. First, a comparative analysis of TPLT performance with different design parameters is performed to determine the baseline parameters. Then, the phase change, natural circulation, and heat transfer characteristics of the TPLT under various operating conditions are investigated. The results revealed a complex relationship between these parameters and the performance of the TPLT. These insights provide valuable guidance for the design and optimization of TPLTs.

Computational Fluid Dynamics (CFD) is widely used for thermohydraulic problems, with the RANS (Reynolds-Averaged Navier-Stokes) approach being popular due to its fairly good quality/cost compromise. However, unsteady complex phenomena such as fluid structure interactions (FSI) or thermal fatigue cannot be predicted with a RANS or even (U)RANS (U: Unsteady) simulation. With the growth of computing resources, Large Eddy Simulation (LES) is increasingly used to model this kind of phenomena, even for high Reynolds number flows that require modeling at the walls (WM-LES: Wall-Modeled LES).

The present communication exhibits an update on validation test-cases computed using the EDF’s open-source code_saturne V8.0 software. These cases include fully turbulent pipe flow, an impinging jet on a heated plate, a wall mounted cube, a backward facing step and a 90 degrees bend pipe. They may be encountered in several location in the primary circuit and the validation on this non-exhaustive list of test-cases is crucial for every approach such as WM-LES or zonal and non-zonal hybrid RANS/LES before applying it on industrial applications. The simulations use the LES approach with standard numerical options, and the standard Smagorinsky sub-grid scale model. Comparisons are made on quantities such as the velocity field, the Reynolds stress tensor and the Nusselt number. WM-LES results show fairly good agreement with experimental and DNS data, particularly for the mean velocity field. The turbulence might be well predicted but remains a challenging issue when the wall is modeled with a standard wall function.

In the context of nuclear engineering, Flow-Induced Vibration (FIV) in steam generators can lead to mechanical damage, responsible for safety issues and significant maintenance cost in Nuclear Power Plants (NPPs). Before going towards FIV simulations with moving tubes, validating tube bundle simulations with fixed tubes is needed.

The present work is performed in the framework of the GO-VIKING Euratom European project (Gathering expertise On Vibration ImpaKt In Nuclear power Generation, https://go-viking.eu/). The open source CFD solver code_saturne (www.code-saturne.org) developed by EDF is utilized using massive parallel computing.

Wall-Resolved LES (WR-LES) is first revisited and validated for the flow around a cylinder at an incident Reynolds number of 3900 with 0% inlet turbulence and a periodic boundary condition in the spanwise direction. Several refinements and computational domains are used and conclusions are drawn to correctly predict drag and lift coefficient for this flow with the actual discretization scheme.

After simulating the flow around a single cylinder with a non-zero incident turbulence, the flow through a 3.5 x 5 tube bundle configuration with a pitch-to-diameter ration equal to 1.44 is studied. AMOVI CEA experimental data are used for comparisons. No periodic boundary conditions are employed in the span-wise direction, the full experiment test-section being represented with a no-slip boundary condition for all the walls. A particular attention is given to comparisons of the drag and lift spectra in time between the experiment and the CFD results.