Conference Agenda

This article presents key two-phase thermal-hydraulics research activities conducted at the POSEIDON and TOTEM platforms, located at CEA IRESNE in Cadarache. The facilities' capabilities support advancements in nuclear thermal-hydraulics research, with applications in reactor safety and innovation. Current studies include steam generator clogging in Pressurized Water Reactors (PWR) at the COLENTEC facility, aimed at enhancing maintenance and predicting clogging behavior. Passive safety systems development for Small Modular Reactors (SMR) is a significant focus, with tests at the EVEREST and EXOCET facilities evaluating natural circulation cooling efficiency. Research on compact steam generators at the MAGIC-3 and BICHE facilities targets performance improvements for space-efficient reactor designs. Additional studies on fuel cladding corrosion under PWR conditions at the CORAIL and CIRENE test loops contribute to more resilient cladding material development. Upcoming research will involve two-phase natural convection flows at the ANUBIS test rig to understand passive cooling in advanced reactors. The platform will also study subcooled boiling under PWR conditions at the DIOGEN facility to optimize heat transfer and safety margins. Lastly, in the continuity of studies for ASTRID demonstrator project, the PLATEAU and OLYMPE experimental loops could support R&D for Advanced Modular Reactors (AMR), including Sodium-cooled Fast neutron Reactors (SFR) and Molten Salt Reactors (MSR).

This study investigates the effect of condensation on the primary depressurization during a loss of coolant accident (LOCA) scenario in an integral small modular reactor (SMR). Two tests were conducted in an integral test facility, one with the passive residual heat removal (PRHR) system and break valve activated, and the other with only the break valve activated. Results show that condensation on the helical heat exchanger (HX) tubes has a significant impact on the primary system’s depressurization rate, which is found to be more important than the effect of the break itself. It is also observed that condensation water can compensate for the coolant loss caused by the break, leading to a slower decrease in the coolant level. The study highlights the importance of considering the effect of condensation in SMR LOCA accidents and suggests further research in this area.

Condensation is a key phenomenon for passive safety systems such as passive containment cooling system during an accident. Accordingly, numerical analysis tools are required to be sufficiently verified and validated for the development of a passive safety system. However, it is also challenging to predict condensation heat transfer precisely because various variables such as temperatures of wall and bulk fluids, geometric parameters, and non-condensable gas fraction affect the phenomena. In this study, we conducted condensation experiment and compared the numerical results from two one-dimensional system analysis codes with the experimental data. The condensation test facility was designed for simulating the condensation phenomenon in the small modular reactor passive safety system. Input models for the MARS-KS and SPACE codes were developed based on experimental facility geometric parameters and test conditions. Both MARS-KS and SPACE showed good agreements with experiment. However, SPACE code provides numerical option to choose model to calculate condensation. In other words, SPACE enables more detailed modeling than MARS-KS to choose an appropriate condensation model for the specific case. Accordingly, we investigated which model shows best agreement with the experiments and which model does not. These results suggest that selecting an appropriate condensation model according to the specific conditions of the condensation can enhance the accuracy of predictions.

Release of hydrogen in the containment of a nuclear power plant, during a postulated beyond design basic accident is a safety concern because explosive mixtures could form and damage components or even threaten containment integrity. The validation of computational tools against experimental data which a representative of postulated accident phenomena is an intermediate step aiming at enhancing the confidence in the code predictive capability. In this paper we present the experimental results of a series of experiments performed in the thermal-hydraulics PANDA facility investigating the erosion of a stratified containment atmosphere rich in helium (used to simulate hydrogen) by a vertical jet from a pipe, after interacting with a horizontal disk. For these experiments were used two PANDA interconnected vessels each of 4 m diameter and 8 m height (total volume 183.3 m³). The helium-rich layer was created in one vessel at the elevations 6 to 8 m. The jet was created by injecting steam from a vertical pipe with 20 cm exit diameter and 4 m elevation. The horizontal disk had a diameter of 20 cm, and it was installed at 5 m elevation. The experimental measurements include gas mixture temperature using thermocouples and concentration using mass spectrometer, and flow velocities using PIV. The tests with horizontal flow obstruction showed that decreasing the jet Reynolds number by a factor of two tripled the helium layer erosion time. On the other hand, changing the initial jet buoyancy does not have an appreciable effect on the overall helium layer erosion time.

The plate-type fuel assembly is widely utilized in nuclear research reactors and consists of several fuel plates and support plates. The fuel plate consists of fuel foil and metal cladding. The coolant is segmented into independent water gaps by the fuel plates and support plates. Due to the disturbances caused by the inlet structure of the plate-type fuel assembly, the flow velocity in each water gap is inconsistent. The significant differences in flow velocity between water gaps can lead to complex flow-induced vibrations in the fuel plates, potentially compromising structural stability. This study employs self-developed measurement technology to conduct detailed experimental research on the flow-induced vibration behavior of a simulated plate-type fuel assembly using strain gauges and eddy current sensors. The experimental results indicate significant differences in the deformation and vibration behaviors of the fuel plates along the axial direction. The deformation and vibration behaviors among the fuel plates also vary. The deformation at the inlet of the internal fuel plate is notably large. The deformation and amplitude at the entrance of the support plate are also notably large. However, the deformation at the outlet of the external fuel plate is larger. At low flow velocity, the amplitude in the middle axial region of the fuel plate is relatively large. At high flow velocity, the amplitude in the inlet region of the fuel plate is larger. The flow-induced vibrations at various positions of the plate-type fuel assembly do not exhibit a dominant frequency within the experimental flow velocity range.

The last Advanced Gas Reactor (AGR-5/6/7) experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory from February 2018 to July 2020, accumulating 360.9 effective full power days. Since fuel temperatures could not be measured directly—because contact between a thermocouple and the fuel could lead to unwanted particle failures—the ABAQUS-based finite element heat transfer code was used to predict daily fuel temperatures over the entire irradiation period. Accurate determination of calculated temperature uncertainties is crucial in interpretation of fuel irradiation performance to ensure achievement of the AGR program objectives. Thermal model parameters with high sensitivity and/or large uncertainty were identified for quantification of the calculated temperature uncertainty. Propagation of model parameter uncertainty and sensitivity was then used to quantify the overall uncertainty of calculated temperatures. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish the sufficiency of the time-dependent first-order (linear) expansion terms in constructing the temperature response surface. Since heat produced in the fuel compacts is transferred through the gas gaps surrounding the compacts and graphite holder, uncertainty in the gap widths is a dominant factor in fuel temperature uncertainty. For all AGR-5/6/7 capsules, an error in capsule design allowed the graphite holders more lateral movement within the capsule shell than intended, resulting in a nonuniform gas gap around the capsule circumference that impacted fuel temperatures. This paper focuses on quantification of gap width uncertainties and the corresponding fuel temperature uncertainties during irradiation of the AGR-5/6/7 experiment.