Conference Agenda

The lead-cooled fast reactor (LFR) is one of the promising Gen-IV designs which is being actively developed by several commercial vendors. One of the safety aspects of interest for LFR designs is the potential burst of a fuel pin.

During a fuel pin failure event, a jet of gaseous fission products is ejected into the coolant subchannels adjacent to the failed pin. The jet has a relatively high momentum due to the pre-pressurized nature of fuel pins. The question of interest is whether the gas jet and subsequent gas bubble formation in coolant subchannels could potentially thermally blanket adjacent fuel pins, leading to them failing. Hence, a potential chain failure propagation across the core is imaginable.

The aim of the present work is to experimentally investigate the bubble formation, location and behavior.

The experimental setup used for the investigation consists of a liquid metal loop equipped with high-resolution measurement techniques. First, experiments were conducted using a single sub-channel test section, combined with high-speed imaging. This experimental campaign was used to gain novel first general insights into the behavior of a buoyant gas jet in low Prandtl, high-density liquid. A second test section was then built, allowing for a multichannel observation of the phenomenon. The paper will include the results of the two experimental campaigns and the conclusions that could be drawn concerning the potential occurrence of fuel pin failure propagation in LFR cores.

After the SGTR (steam generator tube rupture) accident in the LFRs, the high-pressure water on the secondary side enters the primary side and is heated to generate a large number of bubbles, which may hinder the flow of LBE in the reactor core, cause heat transfer deterioration, and threaten the nuclear safety. The behavior of bubbles in the fluid phase is obviously affected by the size of the flow channel. However, there have been relatively few systematic studies on the influence of channel size on bubble distribution characteristics in LBE. The drift flux model is one of the most successful models to predict the distribution of void fraction in gas-liquid two-phase flow. In this paper, based on the upward flow experiment of gas-LBE two-phase flow in circular pipes with various hydraulic diameters, the phase distribution parameters were measured, and the influence of channel size on the parameters of drift flow model was studied.

After the occurrence of a steam generator tube rupture (SGTR) accident in lead-bismuth fast reactors, the interface evolution and transport of bubbles can lead to bubble aggregation, thereby affecting core safety. The interfacial area transport equation (IATE) is an important method for predicting IAC and has significant applications in system analysis codes. In this study, a nitrogen-liquid lead-bismuth metal two-phase flow experiment was conducted in a vertical circular tube channel, and the local interfacial area concentration (IAC) was measured. The measurement data reflect the radial distribution and axial development characteristics of IAC and reveal the evolution and transport characteristics of interfaces. Additionally, this study reviewed the available IAC prediction models including IAC correlations and IATE. However, most of these prediction models have been not developed for the gas-liquid metal two-phase flow, the experiment database was used to verify the applicability of these models in the liquid lead-bismuth metal fluid. The verification shows that the IAC correlations cannot give good predictions of the IAC in liquid lead-bismuth two-phase flow, while the IATE could have a better prediction result, but there is still some difference from the experimental measurement IAC.

During a steam generator tube rupture (SGTR) accident in a lead-cooled fast reactor (LFR), vapor entering the core can induce power excursion and threaten reactor safety. Accurately predicting bubble transport in LFR during SGTR is crucial for its safety assessment. This paper uses a neural network (NN) to predict the bubble distribution within the Europe Lead Cooling System primary system during SGTR accidents. The NN-based model uses one-hot encoding to accommodate heterogeneous inputs and implements a modified Softmax function to avoid non-physical outputs. The method of deep ensembles then quantifies the prediction model uncertainties. The prediction model can accurately predict bubble distributions at three different locations. A relatively large ensemble size is required to converge the ensemble mean, while the convergence of ensemble standard deviation may suffer from outlier samples. Ensemble predictions at different locations tended to be negatively correlated, which usually became weak near extreme values (0 and 1).

The steam generator heat transfer tube rupture (SGTR) accident can lead to violent interactions between lead-bismuth liquid metal (LBE) and water in lead-cooled fast reactors, which can seriously threaten the safety of the core. In this paper, high parameter experiments and refined numerical simulations are used to investigate the lead-bismuth liquid metal-water interaction mechanism. Thermocouples and pressure sensors were used to capture the fluctuation behavior in temperature and pressure in the experiments. This complex and opaque internal interaction is modeled by constructing dynamic boundary conditions of multiphase and multiphysics processes. We demonstrated the existence of three stepwise sequential interaction mechanisms. Moreover, special phenomena such as vapor film wrapping around the core of the jet and secondary penetration have been discovered. his study provides new insights into the interaction between LBE and water and offers important reference for developing mitigation strategies for SGTR.

This study investigates the optimization characteristics of a secondary-side passive residual heat removal system (PRHRS) for enhancing inherent safety performance in lead-bismuth cooled fast reactors (LFRs). Using the fully implicit NUSOL-LMR code with fluid-structure coupling, analyses demonstrate that the PRHRS activates secondary-side natural circulation during unprotected transient overpower (UTOP) and unprotected loss of heat sink (ULOHS) accidents, reducing core temperature rise by 12.3% (p<0.01) while maintaining fuel temperatures 231 K below melting thresholds. The system synergizes with inherent reactivity feedback (coolant density feedback: −1.22 pcm/°C, Doppler feedback: −0.663 pcm/°C) to suppress coolant solidification risks. Under UTOP conditions, PRHRS caps peak cladding temperatures at 1,147 K (52% below safety limits), whereas during ULOHS, it sustains decay heat removal via secondary-side passive flow (2.3% rated capacity). Results conclusively show that integrating the passive system significantly enhances inherent safety under extreme accidents, substantially mitigating potential risks. These findings provide critical insights for optimizing safety designs in forced-circulation LFRs.