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.

A coupled neutronics and thermal-hydraulics simulation code is developed at JAEA. In the coupling simulation code, the 3-dimensional two-fluid ACE-3D code, which is the in-house code of JAEA, will be adopted to simulate thermal-hydraulics behavior inside nuclear reactor fuel assemblies. The ACE-3D code calculates void fraction distributions under operational conditions for use in neutron transport simulations. This research aims to validate ACE-3D using data from a two-phase flow experiment. For this purpose, a two-phase flow experiment was conducted in a 4×4 unheated fuel assembly. In the experiment, the time-averaged void fraction distribution was measured using a wire mesh sensor system under high temperatures (373 K-500 K) and high-pressure conditions of up to 2.6 MPa. The experimental results were analyzed, and the data were visualized to understand better the behavior and characteristics of the two-phase flow in the fuel assembly. A two-phase flow data set is being developed, covering a wide range of experimental conditions, including higher-pressure regions, which can be used for validating thermal-hydraulic codes. Finally, the ACE-3D thermal hydraulics code was applied to the two-phase flow experiment. The calculation results were then compared to the experimental ones, and the issues were identified for improving ACE-3D in future simulations.

CFD is considered to be a valuable tool for assessing and improving the performance and safety of nuclear reactors. Verifying or creating CFD models to predict reactor fluid dynamics is crucial for Gen-IV reactors, which are cooled by liquid metals. The thermal boundary layer in liquid metals is of a greater thickness than that of the momentum layer, leading common turbulence models to incorrect predictions: this highlights not only the necessity for the development of new models but also the creation of databases to validate them. Two main methods may be used to collect the data: perform experimental tests or conduct high-fidelity simulations. This study compares these two methods by assesing the results of a benchmark exercise proposed by the EGTHM. The reference system for the thermo-hydraulic exercise is the Advanced LFR European Demonstrator. Particle Image Velocimetry was employed to obtain high-resolution data on the flow around a triangular lattice.* Computational high-fidelity data are obtained via DNS using an original discretisation technique on a periodic domain of four subchannels. The numerical study also considers heat transfer, by setting a Prandtl number Pr=0.031 representative of LBE. Statistics of velocity, thermal fields and main turbulent flow features are presented and compared with experimental data. This approach allows not only the comparison of experimental and numerical results, but also the integration of one with the other where there are deficiencies, with the aim of providing the optimal dataset for the development of future turbulence and heat transfer models.

*Menezes et al., DOI: https://doi.org/10.1063/5.0154590

Spacer grids in PWR fuel assemblies are equipped with mixing vanes for inducing lateral coolant flow and improving thermal performance. Therefore, understanding how shape of mixing vanes has impacts on flow is important in spacer grid design . This paper presents the results of the PIV (Particle Image Velocimetry) experiment, which was performed on 5x5 rod bundles with spacer grids of two different designs. From the results, the shape effects of mixing vanes were evaluated. Additionally, the experimental results were compared to the results of CFD (Computational Fluid Dynamics) and the applicability of CFD in spacer grid design was investigated.

This study presents a novel application of Lagrangian Particle Tracking (LPT) combined with plenoptic imaging to perform three-dimensional flow measurements within a 6x6 nuclear fuel rod bundle. The experiments were conducted in the Shaking Bundle Facility (SBF), a full-scale experimental model of a nuclear fuel assembly featuring 36 acrylic rods arranged in a square lattice. The assembly replicates the rod-to-rod pitch and mechanical properties of prototypical fuel bundles, with spacer grids providing structural support and simulating realistic flow conditions. Para-cymene, a working fluid with the same refractive index as the surrogate rods, was used to achieve optical transparency and accurate flow measurements.
The experimental setup enables detailed characterization of mean flow fields, boundary layer profiles, and turbulence statistics. Flow data were captured using a plenoptic 3.0 imaging system capable of volumetric measurements in a 14x12x12mm domain, with particles tracked at a frequency of 2 kHz. Advanced Radial Basis Function (RBF) interpolation was applied to reconstruct smooth 3D velocity fields while enforcing incompressibility and boundary conditions.
Results are systematically compared to Direct Numerical Simulations (DNS), Large Eddy Simulations (LES), and prior experimental data, providing a robust benchmark for validating thermal-hydraulic models. This work offers a high-fidelity experimental dataset for improving the predictive accuracy of computational models used in nuclear reactor safety analysis and for understanding flow behaviors under reactor-like conditions.

The depressurization process in light water reactors is an important factor for nuclear safety, and there is a need to develop an analysis code for this transient phenomenon and its validation process. Corroborated experimental data are crucial for evaluating the thermal characteristics of transient boiling and its associated uncertainties. In particular, the spatiotemporal distribution of void fraction during the depressurization process remains undetermined. This study conducted a transient flow boiling experiment during a depressurization process with our test facility for 3D thermal hydraulics in light water reactors (SIRIUS-3D). The test section was a 5×5 rod bundle partially simulating the fuel assembly of an actual reactor. Five units of the subchannel void sensor, capable of measuring the local void fraction between electrodes at a high sampling rate, were installed along the axial direction in the test section’s heated region. We evaluated the multi-dimensional void behavior in a 5×5 rod bundle with a linear depressurization rate ranging from 0.5 to 2.0 MPa/s and an initial system pressure of 7.2 MPa. The rod-surface heat flux and inlet mass flux were set to 70 kW/m² and 750 kg/m²/s, respectively, for all cases. The development of the boiling flow during the depressurization process was summarized with the depressurization rate as a parameter. The void fraction growth rate and time-averaged void fraction were quantified. The spatial void fraction distribution was organized and discussed based on the average values obtained by dividing the regions according to the distance from the center of the bundle cross-section.

The United States has the largest operating fleet of nuclear reactors in the world. Operating cost reduction and power uprates are the two major topics that can further bolster the economic and technical sustainability of LWRs. Over the past decade, a large amount of two-phase flow and heat transfer data, especially for the post-CHF regime, has been collected through the NRC/PSU RBHT reflood test facility. The present work summarizes the on-going efforts in large-scale RBHT data compilation and the assessment of heat transfer modeling accuracy for the dispersed flow film boiling and the transition boiling regime under elevated pressure conditions (up to 60 psi). Various existing post-CHF heat transfer models were compared with the rigorously compiled large-scale experimental data sets, and their performances were evaluated. It was found that existing models carry with them large prediction uncertainties, which further leads to an overly conservative safety limit for existing LWRs. Therefore, it is recommended to develop new post-CHF heat transfer models with significantly reduced thermal hydraulic uncertainties that originate from advanced two-phase flow diagnostic and processing data techniques as well as more in-depth understanding of the underlying physics involved. The large-scale dataset is also found to be extremely useful for developing physic-integrated data-driven models that provide superior prediction accuracy with physically realistic projections.

In 2024, a novel separate-effect-test (SET) facility for two-phase critical (TPCF) flow studies was commissioned at LUT University’s Nuclear Engineering Laboratory. CRiticAl Flow Test facility (CRAFTY) utilizes a straight long tube for discharging subcooled water from an upstream pressure vessel to atmospheric pressure. Prior, a plethora of two-phase critical flow experiments have been carried out in the world. A preliminary literary survey found out that there is a lack of two-phase critical flow experiments utilizing very long length-to-diameter (L/D) ratio tubes (>200). In a postulated primary-to-secondary leak in a PWR, the L/D ratio of the tube can be upwards from 1000 depending on the steam generator design.

In CRAFTY, the L/D ratio and tube diameter can be conveniently changed with interchangeable discharge tubes. A discharge tube with an inner diameter of 13 mm and closely resembling the VVER-440 steam generator tube (inner diameter of 13.2 mm) was utilized in the tests conducted in 2024. The length-to-diameter ratio of the tube was 350 which is close to the half of an average length of the VVER-440 steam generator tube.

Altogether 12 discharge experiments with subcooling varying from 5 °C to 60 °C and upstream pressure from 5 MPa to 8 MPa were conducted. The nominal pressure difference between the primary and secondary circuit in VVER-440 is around 7 MPa.

This paper discusses the experiment results, introduces a simplified critical mass flux model utilizing a modified Jakob number, and presents some simulation results obtained with a system thermal hydraulic code.

In order to realistically simulate the thermal-hydraulic behavior and accident progression during a multiple failure accident, an integral effect test was performed to simulate an SBLOCA (Small break loss-of-coolant accident) with failure of safety injection in the ATLAS-CUBE test facility, which can simulate the thermal hydraulic interaction between the RCS (Reactor coolant system) and the containment. With the break at the cold leg, failure of the safety injection was assumed, whereas an accident management (AM) action was implemented to initiate the safety injection pumps (SIP). The test result confirmed the sufficient grace time during the multiple failure scenarios, including safety injection failure and loop seal clearing phenomena. The compartments acted as passive thermal sinks, effectively maintaining containment pressure below the set-point of spray injection, and ensuring long-term cooling without spray system operation. The test data in the ATLAS-CUBE facility was utilized to assess SPACE and CAP codes. The linked calculation of both codes was performed with considering the M/E (Mass and energy) transport and the P/T (Pressure and temperature) build-up in the containment. From comparing the test and calculation result, it was found that a higher pressure and temperature of the containment was predicted in the multi-volume of the containment in the CAP code calculation. The uniform temperature inside the containment in the single-volume case could overestimate the heat transfer at the passive heat sink and it affected a slower increase of the pressure and temperature of the containment.

Natural circulation imbalance or interruption (NCI) phenomena observed in C3.1 test of OECD-ATLAS3 project will be described in the present paper. When the primary forced flow is lost, the reactor core decay heat is generally removed through natural circulation (NC) convection: the flow is driven by the coolant density differences in the steam generators (SGs) as heat sink and in the reactor pressure vessel (RPV) as heat source.
During an asymmetric cooldown conditions between two steam generators, especially for the case of APR1400, Natural circulation imbalance between the two loops might occur. Moreover, natural circulation imbalance trend can be resulted in NCI phenomenon that means a complete loss of natural circulation flow in a certain loop (or loops) of the primary system. This NCI scenario may prevent the RCS to reach the entry conditions for the residual heat removal system (RHRS). Therefore, an understanding on the NCU is important.
C3.1 test consists of two separated tests named by Run1 and Run2 to compare an effect of primary coolant inventory. These two runs were performed on December 4th and 6th, 2023, respectively. Main objectives of the test were to investigate natural circulation imbalance and NCI phenomenon induced from an asymmetric cooldown transient operation, and to investigate an effect of related parameters such as the decay core power rate, injection flow rate of auxiliary feed water to unaffected SG, and discharge flow rate of ADV. In the paper, test results were compared with post-test calculation results.

Ensuring the safety of nuclear power plants is paramount, Integral Effect Test (IET) facilities have been utilized to verify the performance of reference reactors and assess the application of newly proposed technologies. Before the construction of IET facilities, System behavior analysis should be conducted to ensure that IET facilities can adequately represent reference reactors. In this study, Platform for Advanced TRaining and Integrated OPR1000 Thermal-hydraulic Test (PATRIOT), using refrigerant as the working fluid, was demonstrated to exhibit behavior similar to a reference reactor under station blackout (SBO) conditions through the utilization of system analysis codes. The PATRIOT, developed at UNIST based on the OPR-1000 design, operates with R134a refrigerant at 26.5 bar on the primary side and 13.5 bar on the secondary side. The MARS-KS code was used to analyze SBO behavior, and the R134a properties were generated within compatible pressure ranges for system analysis. The results were compared to the ATLAS, an IET facility developed by KAERI for APR-1400, which has similar design characteristics to OPR-1000. Compared to ATLAS, PATRIOT exhibited less pressure reduction and faster onset of dry-out phenomena, attributed to the lower latent heat and heat transfer of R134a. Despite these differences, the behavior of PATRIOT was similar to ATLAS, which demonstrated the feasibility of utilizing R134a in IET facilities. Therefore, It is confirmed that PATRIOT can simulate the reference reactor. Furthermore, considering the necessity of refrigerants for IET facilities to scale down, this study could contribute to the development and validation of refrigerant-based IET facilities.

In September 2024, SMART100 with a Passive Safety Injection System and Containment Pressure and Radioactivity Suppression System obtained standard design approval from the Korean regulatory agency. SMART-ITL built to evaluate the operating performance and safety of SMART100 was equipped with all passive safety systems except CPRSS. It is designed to simulate most accidents that can occur in SMART100, including transient accidents such as CLOF, SGTR, and FLB as well as SBLOCA. The role of the PSIS during SBLOCA is to supply coolant to the reactor for 72 hours without operator intervention, and its injection performance by gravity head was verified using SMART-ITL. The Passive Residual Heat Removal System operated in almost all accidents occurring in SMART100 is a natural circulation cooling system in which the condensation heat exchanger connected to the secondary side of the steam generator is contained in the Emergency Cooling Tank, and removes the core residual heat absorbed in the steam generator to the ECT. The heat removal performance of the PRHRS was verified through various types of accident simulation tests. This paper deals with the performance of the PSIS and the PRHRS confirmed from the thermal-hydraulic test results using SMART-ITL. In all individual accidents where the passive safety systems were activated, they performed sufficiently to bring the reactor coolant system to a safe shutdown.

A critical heat flux (CHF) occurring on a heat transfer surface under forced flow conditions has different mechanisms depending on the flow channel geometry and flow conditions. In the thermal design of reactor cores, the CHF is an important phenomenon, and it is essential to understand the CHF characteristics under actual flow conditions to improve the CHF prediction method. In this study, steady-state CHF experiments were conducted in forced convection boiling flow at low velocities under high-temperature and high-pressure conditions using a 2 × 2 heated rod bundle with a heated length of approximately 1.2 m. An optical fiber sensor inserted in a 0.5 mm diameter metal tube was mounted on the rod surface and captured the axial distribution of the rod surface temperature at a frequency of 100 Hz and a spatial resolution of 2.6 mm. The experimental results showed intermittent increases and decreases in the rod surface temperature at the top of the heated rod bundle section with stepwise increases in the rod bundle thermal power. This corresponds to repeated localized dry patch formation and rewetting. As the inlet subcooling decreased, the onset of the rod surface temperature increase shifted upstream and dry patches formed over a larger area in the flow direction. A slight increase in the thermal power of the rod bundle near the CHF expanded the area of dry patches or increased the frequency of their occurrence, leading to a transition to a continuous increase in rod surface temperature.

Subchannel analysis is the most competitive approach in thermal hydraulics analysis of rod bundles. It considers transport of mass, momentum and energy axially along the subchannel and laterally across the gaps between the subchannel. Turbulent mixing is an influential parameter for lateral exchange across the gaps which is caused due to velocity fluctuations in the axial direction. Several factors such as rod bundle geometry, coolant flowing properties, gap distance between the subchannels and eddy diffusivity play an essential role in the turbulent mixing parameter. A wide number of experimentally fitted empirical correlations are present to predict turbulent mixing parameter for different subchannel geometry with a significant average mean error among themselves. In 2018, Shen et.al. performed Computational Fluid Dynamics (CFD) for a square bare rod bundle and developed a correlation for square-square center subchannel interaction for the turbulent mixing parameter. In this paper, a similar CFD analysis is performed for a hexagonal bare rod bundle between two triangular center-center subchannel and a center-side subchannel using Reynolds Stress Model (RSM) for a range of Pitch to Diameter ratio varying from 1.1-1.5 for the Reynolds number in the range 8000 to 100000. A new correlation is being developed for turbulent mixing parameter using this numerically generated data. The developed correlation is then compared with the existing empirical correlations.

The void fraction is a key parameter for affecting the coolability and neutron-moderating performance of water-cooled nuclear reactor. More refined experimental data are required to develop multi-fluid dynamics models for determining the void fraction distribution. A Wire Mesh Sensor (WMS) with rod electrodes was developed to measure the cross-sectional distribution of void fraction in a 7 × 7 heated rod bundle with a diameter of 9.5 mm and pitched at 12.6 mm, and applied to a boiling two-phase flow experiment under atmospheric pressure conditions assuming at accident in pressurized water reactor (PWR). The sensor consists of 8-wire by 8-wire and 7-rod by 7-rod electrodes. Wire electrodes with a diameter of 0.2 mm are arranged in a horizontal and vertical crosswise between the rod bundles. For each measurement, the local void fraction in the subchannel center at 64 points were obtained from the wire by wire electrodes and 196 void fraction points near the rod surface were obtained from the wire to rod electrodes. The temporal resolution of the void fraction measurements was 2500 frames. The axial and radial power distribution of the heated rod bundle is uniform.

This study presents the development of a detailed experimental database for two-phase flow interfacial parameters in a 5×5 rod bundle channel featuring a spacer grid. The investigation aims to elucidate the spatial and transport characteristics of gas-liquid interfacial structures and the influence of spacer grids on two-phase flow dynamics.

A comprehensive experimental system was designed, incorporating flow visualization, four-sensor conductivity probe, and two-phase PIV (Particle Image Velocimetry) measurement technologies. Key interfacial parameters, including void fraction, bubble size, interfacial area concentration, and velocities of gas and liquid phases, were systematically measured and analyzed under various flow conditions.

Results reveal distinct distributions of void fractions transitioning from "core-peak" to "gap-peak" patterns as liquid velocity increases, driven by enhanced turbulent mixing. Spacer grids significantly disrupt flow characteristics, causing bubble breakup and coalescence, with effects extending approximately 20 hydraulic diameters downstream. Existing drift and fluctuation velocity models underpredict the impact of spacer grids, highlighting the need for model optimization.

This work provides critical insights into the complex behavior of two-phase flow in rod bundle channels, offering validated datasets to enhance computational models for reactor thermal-hydraulics and guiding the design of spacer grid structures for improved flow stability.

The boundary layer forms the main thermal resistance for heat transfer between the coolant and the fuel rod. Therefore, the structures of the velocity boundary layer greatly affect the thermal hydraulic performance of the fuel rods. The present study performed experimental investigation on effects of pulsating flow on evolution of the velocity boundary layer in a 5×5 rod bundle channel. The Time Resolved Particle Image Velocimetry (TR-PIV) technique is used to directly measure the velocity distributions near the rod surface under different flow conditions. The velocity boundary layer is reconstructed from the measured velocity. The dimensionless velocity distribution over the surface of the fuel rod is obtained by fitting the experimental data to the Spalding formula. The structure of the boundary layer and flow characteristics are analyzed and compared quantitatively. The experimental results indicate that the perturbation introduced by the pulsating flow can disrupt the development of the boundary layer and significantly reduce the thickness of the inner layer of the boundary layer. The larger the amplitude and the smaller the period, the greater the perturbation introduced by the pulsating flow.

Critical heat flux (CHF) can potentially cause catastrophic incidents in numerous thermal industries. At low-pressure conditions, due to high surface tension, the vapour bubbles grow in bigger sizes compared to the high-pressure conditions and may locally accumulate on the heated wall. Due to this, a local dry patch is formed on the heated wall causing a sharp rise in the wall temperature which is referred as DNB-type CHF. Therefore, CHF occurrence is the most crucial factor for ensuring the safe operation of thermal systems that experience coolant phase change. The present work predicts the subcooled flow boiling characteristics and CHF under low-pressure conditions in hexagonal fuel sub-assembly with plate-type spacer. In fuel assembly, spacer grids support fuel rods, reduce flow-induced vibrations, and increase coolant mixing. A WHFP model is employed with the EMF model to simulate low-pressure conditions. The Tolubinsky and Kostanchuk correlation for bubble departure diameter is modified to incorporate the bigger vapor bubble sizes that occur in low-pressure conditions. The current methodology demonstrates strong consistency when validated against the experimental data available for low-pressure conditions. The numerical analysis is further extended to investigate the influence of the spacer on subcooled flow boiling characteristics and the occurrence of CHF. The results show that the spacer acts as blockage, resulting in an increased pressure drop in the spacer's region and inducing a secondary flow within subchannels. In fuel sub-assembly with spacer, coolant velocity was found to be maximum at the spacer's position.

Accurate prediction of two-phase parameters in pressurized water reactors (PWRs) is crucial for the safety analysis of nuclear reactor cores. The refined subchannel model can enhance the spatial resolution of traditional subchannel codes by a factor of four. The ATHAS-H subchannel code, based on the refined subchannel model, has already completed the development of a single-phase flow and heat transfer calculation model. This study represents a continuation of previous work, developing a two-phase flow and heat transfer model for ATHAS-H based on a homogeneous flow model with slip ratio. The code was validated using experimental data from two different types of mixing grid crossflow experiments, the CE5×5 subcooled boiling experiment, and the PSBT bundle void fraction experiment. The validation included directional crossflow, subchannel outlet temperature, rod wall temperature, and void fraction. The results indicate that the ATHAS-H calculations are in good agreement with the experimental data. ATHAS-H can accurately reflect the non-uniformity of local parameters within subchannels caused by mixing grids and non-uniform power distribution in the rod bundle. This study demonstrates the advantages of high-precision subchannel code ATHAS-H in improving the accuracy of two-phase parameter predictions in PWRs. This capability lays a solid foundation for high-precision analysis of critical heat flux (CHF).

In LOCA, there are various uncertain variables that must be considered. These uncertainties determination have been developed by using simple random sampling method. Here, fuel burn-up must be considered and also thermal conductivity and random variables must be considered to derive staistic evaluation results. In this study, the developed SUPER code was used to perform SRS evaluation considering LODA optimization and uncertainty due to fuel burn-up. While considering how to apply the fuel burn up effect and thermal conductivity using the FRAPCON correlation, we introduce a method of fully automated evaluation using SUPER code for uncertainty calculation and optimization.

In this study, variables that should be considered fuel burn up condition, thermal conductivity, and uncertainty were selected to compare and review the PCT evaluation according to fuel burn up and the uncertainty distribution according to fuel burn up, and the most conservative evaluation results were derived. The evaluation results confirmed that the thermal conductivity and SRS statistical distribution results were limited aroung fuel burn up 30 MWD/kgU.In this study, 124 SRS(Simple Random Sampling) calculation is carred out by SUPER code.

However, the different burn up 7 cases between 0 MWD/kgU and 60 MWD/kgU are estimated.

Throught the sensitivity study, some conclusions are introduced as below:

1) Under various fuel burn up condition, in each case, 124 SRS calculations are carried out and PCT statistical distribution and 95/95 percent/accuracy results are introduced.

2) In high burn up conditions, PCT results are decreased by FQ burn down.

It is crucial to understand the behaviour of non-condensable gases (NCGs) in light water reactors (LWR) coolant systems, as their presence could exacerbate accidents and transients by interfering with heat transfer and flow paths, especially during long-term post-accident reactor cooling. The potential sources of NCGs in the reactor coolant system have been thoroughly investigated. However, despite the significant role that NCGs play in the reactor coolant system, there is a relative scarcity of published works addressing the details of the release and dissolution of NCGs. This paper presents previous research efforts on the release and dissolution of NCGs, covering experiments and modelling.

The release and dissolution of NCGs is an intricate phenomenon. When simulating the dissolution and release of NCGs, it is crucial to consider various physical aspects. These include the transport equation for dissolved gas content, release and dissolution rates, conservation equations for the gas phase, and the equations of state for a mixture of two components, where one component is water that may exist in liquid and vapour forms. This paper discusses the gaps in modelling NCG release and dissolution.

Additionally, the paper provides insights into the ongoing NCGDENSE project of LUT University of Finland which is funded by SAFER2028 (National Nuclear Safety and Waste Management Research Programme 2023-2028), which focuses on studying the release and dissolution of NCGs through analytical, experimental, and numerical methods. A separate-effect-test (SET) facility will be constructed to serve as a platform for direct NCG release and dissolution measurements.

A Loss of Coolant Accident (LOCA) might occur if there is a rupture in any of the piping systems linked to the reactor vessel. This rupture may lead to the continuous and uncontrolled loss of reactor coolant into containment. In the absence of an adequate emergency cooling water source, the subsequent increase in fuel temperature could cause damage to the fuel and the release of fission products. The use of best estimate codes and methodologies for simulating LOCA can offer detailed insights into the actual plant response, essential for evaluating the effectiveness of an emergency core cooling system. The TRACE thermal-hydraulics code was specifically developed to simulate transient scenarios in LWRs, including LOCA. This code was applied to simulate the full spectrum LOCA in several locations for BWR-6. The analysis was done using the actual plant configuration and operating conditions available from the core follow simulator. An advanced hot channel methodology was specially developed for these simulations. LOCA analysis for a specific BWR-6 proves the plant compliance to the applicable safety criteria and confirm the TRACE BWR-6 LOCA methodology applicability for the full spectrum LOCA analysis. In addition, this study helps to understand better the LOCA phenomenology as well as the plant response to LOCA transient.

When exposed to extreme conditions accompanied by loss of coolant accident (LOCA), the fuel rods experience swelling or, in severe cases, burst of fuel clad, in accordance with the heat up due to the loss of cooling performance during the accident. In such extreme conditions, the multiple deformation of fuel rods impairs coolable geometry, imposing further degradation of cooling performance with flow blockage. Nevertheless, the system code analysis has conventionally focused on the behavior of single hot pin, by which the details of its surroundings were lumped as averaged assembly-scale conditions. Thus, using the conventional modeling scheme, it is difficult to consider the effect of flow blockage accompanied by the deformation of individual fuel rods surrounding the hot pin of interest. Therefore, in this study, LBLOCA analysis has been performed on APR1400 plant using different modeling scheme for the reactor core by additionally modeling the individual fuel rods surrounding the hot pin in the subchannel-scale level. The effect of flow restriction with multiple deformation of fuel rods has been evaluated, using the thermal-hydraulic system code, MARS-KS. As a result, the clad expansion resulted in about 14% volume reduction in maximum within the subchannel where the hot pin was located. Despite of small deformation as such, the PCT of hot pin increased about 36K during reflood.

Despite their importance and ubiquity, high void fraction two-phase flows are notoriously difficult to probe optically. With a significant number of dense scattering bubbles, images become rapidly corrupted by intensely scattered photons, resulting in occlusion, loss of definition, and errors in size and motion estimation. Information lost through these detrimental effects can be recovered if scattered photons are excluded entirely. However, this corrupting light appears femtoseconds to picoseconds after initial illumination, depending on bubble sizes. Achieving such short measurement times is not possible by conventional electronic means. Through non-linear optical phenomena, including the Optical Kerr Effect, images can be acquired with picosecond and sub-picosecond exposure times and gated in 33 femtosecond intervals.
Using semiconductor and liquid crystal-mediated Kerr gating, static and dynamic bubbly flow scenes have been temporally decomposed, capturing only useful imaging light. In doing so, the definition of bubbles is preserved, reducing uncertainty in size determination, even in optically thick scenes. Furthermore, as the spatial scattering profile is time-dependent, experimentally capturing the complete scattering history allows bubble size measurement. Applying this technique to a newly built, high void fraction bubble column, proof of concept is shown for simultaneous tracking and measurement of size and deformation. With continued development, ultrafast imaging may significantly improve multiphase imaging and provide insight into ultra-high-speed transient events.

In order to further decarbonize energy uses, numerous compact and small-scale nuclear systems are being developed worldwide. In this context, there is a growing interest in conducting research on two-phase flows in mini-channels, as understanding the behavior of these flows is essential for optimizing the heat transfer processes and enhancing the efficiency of these advanced nuclear systems. Among the thermal-hydraulic issues identified in this field is the frictional pressure drop under two-phase conditions. The present study addresses those issues by means of an experimental investigation of the two-phase frictional pressure drop which was carried out at a millimetric-scale. Those laboratory experiments were conducted using a set of two mini-channels of different length, with an inner diameter of 1.38 mm and arranged horizontally, under controlled conditions to measure the pressure drop as a function of imposed phasic flow rates. Demineralized water was used as the working liquid. In order to simulate an adiabatic two-phase flow, air was injected within the liquid at the inlet of the mini-channels. Phasic mass flow rates were imposed up to 9 g/s and 0.14 g/s, yielding maximum Reynolds numbers of 21,000 and 8,000, respectively for the liquid and gas phases. A dimensionless two-phase pressure drop was calculated from the acquired data and compared with the most recommended model for pressure drop in mini-channels to date. This model has proven incapable of reproducing the experimental data. This highlights the need to improve the predictability of pressure drop models in mini-channels.

This study presents a miniaturized impedance sensor probe designed for two-phase flow measurements, with significant implications for nuclear reactor systems. Positioned between Electrical Resistance Tomography (ERT) and optical/resistive needle probes, the sensor is based on closely spaced parallel needle electrodes to accurately measure local flow parameters, including void fraction, interfacial velocity, and bubble size. Its design allows for high spatial resolution without requiring interaction with the bubble surface, a key advantage over conventional local intrusive probes.

Numerical simulations to assess the electric field distribution across various electrode configurations were employed for two critical purposes. First, assess the impact of bubbles passing outside the primary measurement zone (inter-electrode area), which is essential for understanding how off-axis flow phenomena affect measurement accuracy. Second, the simulations provided insights into the signal processing requirements by generating simulated sensor outputs. These outputs were a key aspect for refining the algorithms used to extract key flow parameters from the sensor data.

Experimental validation, including high-speed imaging and comparisons with resistive probes, confirmed the sensor's capability to detect smaller bubbles and continuously track two-phase flow changes in real-time. The combination of its robust design, high spatial resolution and temporal resolution makes this sensor a promising alternative to existing technologies, suited for applications in nuclear reactor coolant monitoring, where precise control over multiphase flows is essential for ensuring system safety and performance.

Aerosols play a critical role in the transport of radioactive products within nuclear reactors. During severe accident scenarios, high-temperature and high-pressure coolant sprays can lead to complex temperature distributions within the containment, influencing the thermophoretic transport, evaporation, condensation, and coalescence of aerosols. Aerosol measurement technologies have evolved significantly over the years, leading to the development of diverse methodologies, including single-point/full-field, sampling/in-situ, and intrusive/non-intrusive approaches. For instance, laser-based particle visualization has been widely employed to study aerosol particle dynamics. However, temperature field measurements during aerosol transport remain rarely reported in the literature. This study introduces a novel non-intrusive method capable of simultaneously capturing the temperature field during aerosol transport and visualizing the motion of aerosol particles. The Background Oriented Schlieren (BOS) method is employed to obtain the temperature fields of aerosol particle flow through a controlled temperature gradient in a visualized channel. Experimental results demonstrate that this method can accurately obtain the velocity and temperature fields within the measurement domain, with uncertainties less than 5% for the temperature field. Additionally, this study quantitatively analyzed the influence of aerosol introduction on BOS measurements through comparative experiments. The results indicate that light scattering caused by the aerosol particles has no significant effect on the BOS measurement outcomes.

CATHARE is the French thermal-hydraulic code used for nuclear reactors safety analysis. Its reflooding module has been extensively validated for Large Break Loss-Of-Coolant Accident (LB-LOCA) scenarios. However, as Intermediate Break LOCA (IB-LOCA) studies are becoming more and more frequently, the validation of the CATHARE reflooding module needs to be extended to higher pressures than those encountered during LB-LOCA core reflooding.

The PERICLES experimental program at CEA primarily aims to improve the understanding of core thermal-hydraulics during the reflooding phase of a Pressurized Water Reactor (PWR). The test section consists of an insulated stainless-steel shroud, containing 368 full-length electrically heated fuel rod simulators (FRSs) and 25 stainless steel guide tubes arranged in a 17 × 17 geometry. The PERICLES facility has been considerably used for validating the CATHARE code. The experimental program includes 15 high-pressure (10 to 60 bar) reflooding tests which have not previously been used for CATHARE validation.

This paper first introduces the CATHARE code, the reflooding module, and its adaptation to pressures above 6 bar. Then, it presents a comparison between the PERICLES experimental data and the CATHARE computation results. The CATHARE code demonstrates a good prediction of the reflooding tests. Activating the reflooding module of the CATHARE code allows better predictions compared to results without the module's activation.

This paper will conclude with a discussion on the limitations of the presented validation and the need of further experimental data on high-pressure reflooding.

As the third stage of the large break loss of coolant accident, whether the core can achieve effective cooling in the reflooding stage is the most important stage to prevent the large break accident from developing into a serious accident. The motion and heat transfer behavior of the droplets play an important role in the development of the re-submergence stage and are the key factors limiting the peak cladding temperature ( PCT ). The measurement of droplet parameters can provide necessary data support for the development of relevant mechanism models and the safety analysis of reactors under water loss accidents.
Conductivity probe is an effective means to measure droplet parameters. In this paper, the droplet characteristic signals and processing methods under different flow patterns are analyzed from the perspective of algorithm. The deviation of droplet measurement by three-sensor conductivity probe under different droplet size, droplet velocity and probe offset was studied by visualization technology. The relationship between uncertainty and different physical quantities was obtained. This investigation can provide a reference for droplet measurement using conductivity probes under re-submergence.

Accurately detecting coolant level and state is essential for ensuring nuclear reactor core safety, playing a critical role in early incident detection and severe accident prevention. However, level information alone is insufficient to fully evaluate the heat transfer performance and flow conditions of the coolant, necessitating additional void fraction measurement. To complement this, this study developed a horizontal pipe experimental system simulating air-water stratified flow and designed a fiber optic sensor-based device to measure the liquid fraction, and validated its concept through experiments. The device is integrated into the air-water stratified flow system, designed to minimize flow disturbances, and to detect the liquid level by utilizing differences in heat transfer characteristics between the media. This enables the calculation of local flow velocity and liquid fraction. Preliminary operation tests demonstrated stable performance under water flow rates of up to 5 L/min and air flow rates of up to 50 L/min. Experiments varying the power applied to the heating wire revealed distinct heat transfer characteristics, which were also observed under cooling conditions. Additionally, the sensor was able to measure interface movement in various flow environments, particularly confirming a tendency for the interface to be detected in regions with rapid temperature gradient changes. The system, leveraging the high spatial resolution of the fiber optic sensor, provides reliable data while validating the measurement method. Future research will construct a steam injection environment to enable phase detection and steam quality measurement in multiphase flow conditions similar to nuclear systems, further enhancing its practical applications.

In a reactivity-initiated accident, the reactor power might increase exponentially, following an escalation period. The larger is the insertion of reactivity, the shorter is the period.

Under such conditions, critical heat flux (CHF) limits cannot be described using models and correlations derived from and validated against steady-state experiments.

In this work, we present experimental results of transient CHF conducted on surfaces with different finish and wettability in subcooled (10, 50 and 75K) flow boiling conditions at atmospheric pressure.

The results confirm that, for slow transients, the transient CHF approaches the steady state value, which depends on surface finish. However, for fast transients, the CHF values are found to be independent of the surface finish and mostly increase with decreasing period. This observation suggests that the triggering mechanism of the boiling crisis in transient conditions may be different from the one under steady power inputs. It also undermines the rationale of models and correlations that aims at estimating the transient CHF on a certain surface starting from the steady-state CHF values.

Accurate measurement of wall shear stress and near-wall velocity profiles is critical for understanding the thermal and hydraulic performance of pressurized water reactor (PWR) fuel bundles. This study introduces an innovative experimental setup that employs Molecular Tagging Velocimetry (MTV) for direct measurement of flow velocity and gradients fields within a surrogate PWR bundle. The system integrates high-power optical fibers for laser light delivery and a borescopic imaging system embedded within the bundle rods, minimizing distortions and enabling local, high-resolution measurements.

Custom-designed optics ensure efficient laser coupling and delivery through optical fibers, achieving over 85% transmission efficiency. A custom borescopic system, paired with refractive index-matched (RIM) materials, minimizes imaging distortions caused by material interfaces. Preliminary results demonstrate the system’s capability to capture high-resolution flow patterns with a spatial resolution of approximately 10 m/pixel. A small-scale 3×3 rod bundle prototype with an instrumented central rod has been developed and tested under controlled flow conditions, validating the imaging and laser delivery systems.

This work lays the foundation for implementing MTV techniques to measure velocity gradients and wall shear stress in a realistic reactor geometry. By overcoming optical and spatial limitations, this setup provides a pathway for precise experimental data to support advanced numerical simulations. Future efforts will focus on deploying this methodology
in larger-scale and more complex PWR surrogate bundles.

The reliability of the primary cooling circuit in a pressurized water reactor (PWR) is crucial for safe operation. Debris fretting, caused by solid particles in the coolant, can damage fuel rods and lead to the leakage of fission products into the primary circuit coolant. Optimized filters in the fuel assembly bottom nozzle (BNO) can capture debris while minimizing pressure loss and reducing clogging risk.

To investigate the cooling flow through the bottom nozzle and filter, Magnetic Resonance Velocimetry (MRV) was employed using a 3 Tesla magnetic resonance imaging (MRI) scanner. The study focused on a simplified bottom nozzle, filter, and the leading edge of a 5x5 fuel rod bundle. MRV provided high-resolution measurements of 3D velocity vectors and 3D Reynolds stress tensors, without requiring optical access to the complex filter structure. The pressure drop across the filter was measured separately. Wire-like particles were introduced sequentially, enabling precise tracking of their positions and analysis of their impact on flow.

At a Reynolds number (Re) of 50,250, and with up to 100 particles, the filter test resembled standard conditions. A clogging scenario was created by introducing 240 additional particles at Re = 20,000. Using MRI data, the clogged filter’s geometry was reconstructed for CFD implementation. These CFD-grade measurements provide unique experimental data for validating particle motion and clogging models. Time-averaged velocity and Reynolds stress tensor data provide critical insights into how particles and agglomerations influence flow through the filter and around fuel rods, informing design improvements for enhanced reactor safety and efficiency.

Both direct simulation heater and indirect simulation heater rods have an extensive history of being used for a variety of nuclear reactor thermal-hydraulic testing including rod bundle CHF measurement, natural circulation cooling, reflood quenching analysis, flow induced vibration, and experimental observation of different thermal hydraulic phenomena. Despite this, recent studies show that there are a number of challenges associated with the use of indirect heaters, including the introduction of major measurement uncertainty and non-conservatism in CHF prediction, the non-prototypical and non-conservative peak cladding temperatures during blowdown, reflood quenching, and other transient heat transfer temperature measurements during both heat up and quenching processes. Most of these issues can be directly correlated to the physical composition of an indirect heater. Because the internal heating element is surrounded by highly conductive boron nitride or magnesium oxide, heat loss in the axial, lateral, or circumferential directions will be substantially large in case of any local heat transfer transient and/or deterioration events, causing the inaccuracies and non-conservatism observed in many experimental tests. This paper further details the challenges and non-conservatism of indirect simulation heaters, including experimental and simulation based examples. The non-conservative measurements were also verified with different modeling computations. Comparatively, the performance of direct simulation heaters is assessed in a similar manner, with results confirming that the use of indirect heaters poses a great risk to safety analysis and accurate thermal hydraulic analysis.

Valve leakage failures are a common equipment problem in nuclear power plants. Such failures not only lead to economic losses, but also cause radioactive leaks in serious cases, resulting in major safety hazards. Currently, experimental data on valve leakage is very scarce, and the diagnosis of many types of valves is very complex, so it is necessary to supplement experimental data on valve leakage and select the best diagnostic algorithm.

This paper describes a novel experimental system designed to simulate Reactor Coolant Capacity Control System (RCV) upper loading line valve leakage faults, to verify the reasonableness of the experimental setup by analysing the experimental data and to explore the impact of leakage on the system performance. Within the realm of algorithmic analysis, this study evaluates the diagnostic efficacy of various algorithms, including logistic regression, random forest, and support vector machine (SVM) models. The empirical findings of the study reveal that the Random Forest algorithm exhibits the most superior diagnostic precision, achieving a remarkable accuracy of 99.82% in the detection of valve leakage incidents. Algorithms such as Support Vector Machines (SVMs) and Simple Bayes have unsatisfactory performance metrics, with diagnostic accuracies not exceeding the 95% threshold. Therefore, the latter algorithms are considered less suitable for diagnosing valve leakage faults.

This research not only enriches the experimental data but also offers a valuable reference for the selection of appropriate diagnostic algorithms. The in-depth investigation of valve leakage faults can serve as a robust safeguard for the secure operation of nuclear power plants.

One of state-of-the-art pressurized light-water small modular reactors, an innovative small modular reactor (i-SMR), is being developed in Republic of Korea. A steel containment vessel (CV) is adopted not only to prevent release of radioactivity material but also to reduce pressure and temperature of the reactor module. As a newly suggested passive safety system (PSS), a passive emergency core cooling system (PECCS) prevents water level reduction of reactor coolant system (RCS) using natural circulation without additional injection of coolant. The emergency depressurization valve (EDV) and emergency recirculation valve (ERV) which are installed on the wall of reactor vessel (RV) play as the natural circulation flow paths between RV and CV. The pressurized steam from the RV through the EDV is condensed in the CV by heat transfer on heat exchanger of passive containment cooling system (PCCS). The condensed water recirculates to the RV through the ERV. The level of condensed water is important physical variable because the difference of water levels between CV and RV determines recirculation flow rate.
Integral effect test facility is under basic design to conduct thermal-hydraulic validation tests to support the process of standard design approval (SDA) for i-SMR. The basic design of PCCS heat exchanger considering condensation heat transfer was completed and the scaling distortion in integral effect test facility against the prototype was evaluated. Four trains of PCCS heat exchanger with thermocouples to measure fluid and outer wall temperatures will be installed in the integral effect test facility.

The integrated full natural circulation SMR has a high degree of integration, high intrinsic safety, flexible arrangement and can be applied in various scenarios. However, integrated full natural circulation SMR also incorporates some innovative designs, such as the elimination of the Main Circulation Pumps (MCPs), the adoption of helical-coiled tube heat exchanger, the application of the passive safety design, and so on.

Due to these new characteristics of the integrated full natural circulation SMR, the existing system analysis code cannot be directly applied to the integrated full natural circulation SMRs, and the corresponding system analysis code still needs to be developed. RELAP5 code is a widely used system analysis code internationally and has been successfully applied to some SMRs. The RELAP5 code has been improved or added the relevant models that are required for the analysis of integrated full natural circulation SMR by authors.

The improved RELAP5 code is validated by an Integral Effect Test conducted by SNERDI. The comparison results of natural circulation tests at various power levels are shown in this paper. The results show that the improved RELAP5 code compares well with the test results, with a temperature difference of about 5 degrees. Additional test cases will be performed and further validation and evaluation will be conducted in the future.

The recent revision of Emergency Core Cooling System (ECCS) acceptance criteria, which incorporates Design Extension Conditions (DECs) and addresses high-burnup fuel safety concerns, has intensified the need for more accurate loss-of-coolant accident (LOCA) analyses. Previously, thermo-mechanical and thermal-hydraulic behaviors were evaluated separately, resulting in conservative estimates that limited insights into actual coupled phenomena. Implementing an integrated multi-physics approach now enables simultaneous characterization of these behaviors, leading to more realistic analyses and enhanced safety margins. In response to this need, the Korea Atomic Energy Research Institute (KAERI) developed the ICARUS facility to simulate fuel cladding behavior from the post-blowdown stage through the reflood phase of a large-break LOCA (LBLOCA). A Zircaloy-4 cladding and heater assembly, combined with controlled boundary conditions, replicates the thermo-mechanical and thermal-hydraulic environment of a reflood scenario. Real-time measurements of cladding surface temperature, deformation, subchannel fluid temperature, and water level are carried out. By varying heater power, internal cladding pressure, and the reflood initiation time, this study systematically evaluates the coupled phenomena, thereby offering critical insights into the multi-physics behavior of nuclear fuel cladding under LBLOCA conditions. By integrating thermo-mechanical and thermal-hydraulic analyses, this work moves beyond conservative assumptions and provides a more realistic understanding of cladding behavior under accident conditions.

The Molten Salt Reactor Experiment (MSRE) was a key nuclear project in the 1960s that demonstrated the viability of molten salt as a coolant and fuel. In molten salt reactors, understanding complex thermohydraulic behavior is essential for optimizing performance and safety. However, building a full-scale experimental model is often impractical due to high costs and the reactor’s large size. A scaled-down model provides an efficient and cost-effective approach to studying critical aspects of fluid flow, heat transfer, and pressure distribution while capturing key physical phenomena. This study presents the design and computational fluid dynamics (CFD) validation of a 1/5 scaled-down mock-up of the MSRE. The scaled-down model was developed to replicate the geometry of the original MSRE while maintaining fluid behavior, such as Reynolds number. The study also provides a detailed explanation of the scaling laws used to ensure that the down-scaled model accurately reflects the behavior of the full-scale system.

Given the reduced size, the model cannot replicate every detail, such as all the surfaces and channels exposed to molten salt within the reactor core, making validation crucial. CFD simulations were performed using the scaled model to analyze fluid flow and pressure characteristics. The results of the simulations were compared to experimental data from the original full-scale MSRE. This comparison confirmed the accuracy of the scaled mock-up and its reliability for predicting the thermohydraulic behavior of molten salt reactors, making it a valuable tool for further research.

Differentially-heated cavity natural convection is an important phenomenon relevant to the design of thermal energy storage systems, concentrated solar power receivers, building-integrated photovoltaic systems, and nuclear reactor passive safety systems. Particle Image Velocimetry (PIV) measurements are implemented to study the natural convection behavior of molten nitrate salt in a differentially-heated cavity for Rayleigh numbers up to 10⁹ and Prandtl numbers from 22 to 30. A low melting point salt mixture, NaNO₃-KNO₃-LiNO₃-CaNO₃, is selected as a working fluid to provide operating temperatures from 100°C to 500°C. The experimental methodology for PIV measurements in a heated molten salt test section with a transparent optical window is presented along with preliminary test data. 2-D planar PIV measurements of the flow field in molten nitrate salt are compared to measurements of the flow field in water, with matching Rayleigh numbers and Prandtl numbers from 2 to 7. Quantification of measurement uncertainties is described and compared to alternative flow measurement techniques.

Natural circulation flow is widely adopted in Small Modular Reactors (SMR) to simplify system design and achieve passive heat removal. The characteristics of low-velocity natural circulation flow are affected by both forced and natural convection. Since the amount of natural circulation flow is dependent on pressure losses, such as the wall friction, understanding this mixed convection flow is essential for the system design and safety evaluation of SMR. However, most of the previous studies were conducted using air as a working fluid or under low-temperature water conditions. Therefore, this experimental study investigated the wall friction factor in high-temperature water flows. The wall friction factor was measured at a vertical annular channel under natural circulation flow conditions. Experiments were performed under Re of 690-5,020, and Gr of 10⁵-5.5×10⁷ at the heated channel. The gaps of the concentric annular channels were 2.9, 5, and 7 mm, respectively. Evaluation of the existing model showed that the forced convection wall friction model underpredicts the present experimental data under low-velocity and larger gap conditions. Under these conditions, secondary flow within the channel prevented development of flow. Accordingly, this caused continuous changes in the velocity profile, increased viscous dissipation, and greater pressure loss. To predict accurately the increased wall friction factor in mixed convection flow, a new model was developed based on the present experimental data. The model accounts for secondary flow induced by buoyancy and radial temperature gradients within the channel. The developed model demonstrated good prediction under low-velocity mixed convection flow conditions.

In this study, we proposed scaling criteria for wall boundary conditions of thin plates with volumetric heat generation based on analyses of free convection-conduction conjugate heat transfer. Unlike uniform wall temperature(UWT) or uniform heat flux(UHF) boundaries, nuclear fuels typically involve volumetric heat generation. To examine the effect of conjugate heat transfer, the parameter for scaling criteria of wall thermal boundary conditions was analytically derived using the perturbation method. To quantify this parameter, the governing equations were numerically solved using the Runge-Kutta method for free convective flow and the finite volume method for solid conduction. The results showed that when the axial conduction-to-convection ratio—defined as half the plate thickness divided by the product of the modified Biot number and plate length—is greater than 0.5, the solution converges to the UWT solution. Conversely, when this ratio is less than 0.01, the solution aligns more closely with the UHF solution. For plates with the same volumetric heat generation, the peak temperature is highest under UHF condition and lowest under UWT condition. Therefore, the scaling criteria for the wall boundary condition proposed in this study can make a significant contribution to the thermal design of nuclear fuels. Furthermore, the scaling criteria were validated against experimental data for both air-cooled and water-cooled free convective plates with volumetric heat generation. According to the scaling criteria proposed in this study, the air-cooled test data were closer to the UWT condition, whereas the water-cooled test data were closer to the UHF condition.

Natural convection in an inclined channel with downward facing heater were investigate to validate the performace of PECS(passive ex-vessel corium retaining and cooling system), the Korean core catcher developed for the exporting APR1000 reactor. PECS has inclined channels under the structure , and the channels are heated from the top surface when the molten corium drops on the structure by a severe accident. The VPEX(Variable PECS experimental facility)was designed and built to examine the phenomena in the PECS channel experimentally. The VPEX has the heating block made of the carbon-steel with stainless steel coating, which is the same as the PECS structure. And the heat flux distribution over the channel were given by CFD calculation considering the corium behavior on the PECS. The tests were performed for various parameters such as the heat flux, the inlet subcooling, the channel shape, and the pressure. The results shows that the PECS has sufficient cooling capability even with the 175% of the expected heat flux. Also, the behavior of natural convection in the PECS channel were calculated using 1-D code, NCir, and the results were compared with the tests. The calculation results vary by the two-phase friction model and the void fraction model, however, corresponds to the experiments well for certain models.

This study investigates heat transfer enhancement in two-phase natural circulation loops, focusing on the role of superficial velocity in Small Modular Reactors (SMRs). Natural circulation, driven by buoyancy forces, is a key passive cooling mechanism in SMRs, where compact designs necessitate efficient heat removal. Using the MARS-KS code, this research simulates flow behavior, void fraction, mass flow rate, and heat transfer coefficients under two-phase conditions, specifically analyzing the NuScale SMR design, reduced to 1/10th of the original size for experimental feasibility. The study examines the impact of the minimum and maximum superficial velocity ranges, representing variations in the cooling channel gap. The results demonstrate that optimizing superficial velocity enhances thermal efficiency by improving convective boiling and phase change dynamics, while maintaining system stability. Higher velocities lead to better heat transfer performance in the evaporator, riser, and condenser, with the increased flow velocity fostering more efficient heat dissipation. These findings indicate that controlling the cooling channel gap can optimize flow velocity and, consequently, heat transfer. This research provides a foundation for future experimental studies on cooling channel height control, which will further investigate the influence of gap adjustments on heat transfer. The results contribute to the development of more efficient and reliable passive cooling strategies in SMRs, advancing reactor safety, performance, and sustainability. By optimizing natural circulation and refining reactor designs, this study supports the ongoing efforts to enhance the safety, efficiency, and long-term stability of next-generation nuclear energy systems.

External reactor vessel cooling (ERVC) plays a vital role in preventing failure. Nucleate boiling on the curved lower head of the RPV is a key heat removal mechanism. While current models focus on coolability limits, early cooling through nucleate boiling is equally important. Nucleate boiling models rely on accurate heat flux partitioning, with single-phase heat transfer being a key contributor. However, the correlations used in system analysis codes are not suited to the curved geometry of the RPV's lower head. Understanding single-phase heat transfer on downward-facing curved surfaces is essential for developing accurate nucleate boiling models. This study addresses this gap by developing heat transfer correlations for single-phase flow on curved, downward-facing surfaces under constant heat flux. Experimental measurements and CFD simulations were used to develop two correlations: one for forced convection (Ri < 0.1) and another for mixed convection (0.1 < Ri < 10), within the ranges 1,000 < Re < 13,000, 2.56 < Pr < 4.36, and 0.001 < Ri < 10. The study highlights that in forced convection, curvature affects boundary layer development and heat transfer. For mixed convection, buoyancy effects are captured through the Buoyancy number (Bo), with correlations showing how buoyancy transitions from impairment to enhancement of heat transfer. These findings are vital for improving system codes used in nuclear safety analysis, particularly for predicting heat transfer during ERVC and ensuring the integrity of the RPV in severe accident conditions.

The study of boiling flow is crucial for designing safety features of chemical and nuclear plants. The boiling behavior and bubble dynamics depend on various system parameters such as pressure, subcooling, mass flux, heat flux, and flow geometry. While a previous flow boiling experiment by the authors’ laboratory has revealed valuable parametric effects in a decent operational range, its setup alone is insufficient to reveal the influence from the channel width. A new dataset of upward flow boiling is therefore collected in a narrower annular test section. Compared to the previous configuration, this new channel has a larger inner diameter of 25.4 mm and the same outer diameter of 38.10 mm, with a heated inner rod of 3-m long. Multi-sensor conductivity probes are adopted measuring void fraction, gas velocity, and interfacial area concentration following a two-group description. Traversing mechanisms are employed allowing the probes to scan across the flow area, and a dedicated sensor pattern is designed and validated to minimize near-wall blind zones for the narrow channel. In addition, high-speed visualization is conducted recording axial flow evolution, and the Onset of Nucleate Boiling and Point of Net Vapor Generation are identified and recorded. Parametric studies are also presented investigating the flow field dependence on systematic boundary conditions. This work presents valuable new experimental data on narrow-channel flow boiling.

Due to the fuel self-shielding effect, reactor irradiation, and component arrangement, plate-type fuel elements exhibit significant transverse non-uniformity in heat generation. Consequently, the internal thermal-hydraulic characteristics of coolant channels may differ from those of conventional channels. To address this, experimental studies on single-phase flow and heat transfer in narrow rectangular channels with transverse non-uniform heating were performed using PIV visualization technology. Analysis of the flow and heat transfer characteristics in narrow rectangular channels revealed the following: Time-averaged velocity field analysis demonstrated that low-velocity regions in laminar flow regimes are more prominent compared to turbulent flow regimes, while both laminar and turbulent regimes exhibit distinct stratification phenomena in transitional flow regions. In laminar regimes, non-uniform heating reduces the nominal boundary layer thickness , with no observed correlation between velocity gradient and heating power magnitude. In turbulent regimes, non-uniform heating increases the nominal boundary layer thickness, and a decreasing trend in velocity gradient is observed with increasing non-uniform heating power.

Under the conditions of a reactor accident, heat transfer at the wall can be hindered, leading to the risk of boiling crisis. The calculation of the near-wall void fraction is crucial for predicting the boiling crisis. In narrow rectangular channels, large-scale interfaces exist near the wall due to geometric constraints. This paper is based on an improved two-phase multi-scale interface model that considers the interfacial transfer of concentration, momentum, heat, and mass for bubbles of different scales. The model is embedded within the Eulerian two-fluid model in Fluent. The results from the modified model were compared with experimental data, validating the accuracy of the model.

Flow reversal in narrow coolant channels is a critical phenomenon for the safety of research reactors, especially those designed with a downward nominal flow. During a loss of forced flow event, the downward movement of the coolant may temporarily halt before transitioning to an upward natural circulation. Fuel damage can result if dryout occurs and the fuel or cladding temperatures exceed safe limits. This study provides an in-depth examination of flow reversal in narrow rectangular channels by analyzing experimental results and an analytical approach. The analytical approach utilizes the calculated steady-state system pressure drop versus flow curves for different heat flux values. A review of the relevant literature was conducted, and selected experimental data were utilized to benchmark against the flow reversal limits predicted by the analytical approach. The experimental data is selected from tests involving flow reversal in a narrow rectangular channel. The findings were compared with successful flow reversal test data and predicted dryout power under dryout conditions. Also, the study examined the effects of inlet liquid temperature, system pressure, and localized pressure drops. The analytical approach provides physical insights into the flow reversal phenomenon. The approach may be used in conjunction with thermal-hydraulic analysis software, strengthening the confidence in the software predictions.

The present study presents a numerical investigation into the fluid flow and heat transfer performance of a straight mini-channel with an additional inter-connected mixing area in a heat sink plate. The influence of the dimension of the inter-connected area on the thermal-hydraulic performance was examined. Three different sizes of the inter-connected area, defined in terms of aspect ratio (AR), were studied to understand their effect on the thermal-hydraulic performance. Water was used as the coolant, flowing in a single-phase regime under turbulent conditions at Reynolds numbers ranging from 1000 to 4000. A constant heat flux of 10 kW/m2 was applied to both surfaces of the cooling plate. The grid independence test showed a deviation of less than 2% for the friction factor and Nusselt number, while the GCI results indicated that the deviation of the friction factor and Nusselt number was less than 2% within an asymptotic range around 1. The results demonstrated that the aspect ratio of the inter-connected area has an impact on the thermal and hydraulic performance. Both the friction factor and Nusselt number decreased with an increase in the size of the inter-connected area. Furthermore, this study revealed that the interconnection zone created two stationary circulation zones, which influenced the velocity and temperature contours. Finally, a new correlation was developed to explain the relationship between the friction factor and Nusselt number in terms of the Reynolds number and the aspect ratio of the inter-connected area.

Steam Generator Tube Rupture (SGTR) is a critical safety event in nuclear power plants, particularly in Pressurized Water Reactors (PWRs). An SGTR occurs when one or more tubes within the steam generator fail, allowing radioactive coolant from the primary circuit to leak into the secondary side, potentially contaminating the secondary steam and elevating radiation levels outside the reactor containment. This study investigates the thermal-hydraulic response of the PKL facility under a Multi-SGTR (MSGTR) scenario. Conducted within the framework of the OECD/NEA ETHARINUS project, Test J5.1 aims to evaluate the system's performance during an MSGTR event. Two experimental runs were executed: in the first, three out of four steam generators (SGs) were assumed to have ruptured, while the second run assumed failure in all four SGs at the PKL facility. The study presents the outcomes of blind simulations for both scenarios, emphasizing the differences in operational sequences due to varying depressurization and cooldown strategies. In Run 1, depressurization was initiated via the intact SG, followed by activation of the pressurizer (PZR) relief valve on the primary side. In Run 2, only the PZR relief valve was used for depressurization. Both runs extended to 30,000 seconds, during which primary and secondary pressures equalized.

This paper presents an experimental investigation of thermal radiation effects in PANDA facility (PSI, Switzerland), focusing on gas mixture atmospheres with different steam content. Thermal radiation influences the containment atmosphere temperature and buoyancy and, therefore, has an impact on the hydrogen distribution during a postulated accident. The experiments were conducted as part of the ongoing efforts to understand the role of radiative heat transfer in containment thermal hydraulics. In this paper, we analyze the experimental results of two new tests of the so-called P1A2 series performed within the OECD/NEA PANDA project.

The test atmosphere initially consisted of an air and steam mixture at 110°C, with steam concentrations ranging from nearly zero to high values (60%). A stratification layer of 50% nominal helium was created in the upper 2 meters to isolate with best efforts the effects of radiative heat transfer from convective mixing. For the compression of the gas mixture, helium was injected at 10 g/s from the top over a period of 1200 seconds. The subsequent gas thermal behavior and concentration distribution were recorded during the compression phase and an 1800-second decay phase. The results demonstrate that the magnitude of the temperatures is a strong function of the initial steam content, with higher temperatures for lower steam content. Additionally, the experiments confirmed that thermal radiation has a major impact on temperature homogenization during the decay phase, with faster homogenization occurring in atmospheres with higher steam content. These findings have profound implications for CFD calculations in the presence of steam.

The direct vessel injection technology is gradually adopted in new pressurized water reactors because of its advantages of simplifying the design of safety injection system and improving economic benefits. This experiment focuses on the modified DVI (direct vessel injection) safety injection system of the HPR1000. An improved linear scaling method was employed to model the prototype, and relevant experiments were conducted on a 1:8.5 scale visualization test section. Through experimentation, phenomena such as CCFL (counter-current flow limitation) and bypass flow near the break were observed during the refilling and reflooding stages within the annular cavity under a large break loss-of-coolant accident (LBLOCA). The study investigated the impact of various break locations, injection heights, and the presence or absence of guiding structures on the bypass effect in the DVI safety injection system. Additionally, comparisons were made between DVI safety injection and cold-leg injection. The research findings reveal that direct bypass flow dominates the safety injection bypass during the refilling and reflooding stages of an LBLOCA. The closer the cold-leg break location is to the DVI nozzle, the significant increase in bypass flow at the break location. Different DVI injection heights affect the spread of the liquid film, thereby influencing the proportion of bypass flow. The installation of guiding devices can effectively reduce the proportion of safety injection bypass flow. The data results from this study provide crucial insights for the optimization and innovation of the modified safety injection system in HPR1000.

In marine environments, the internal components of in-vessel pressurizers, such as control rod guide tubes and electric heating rods, are susceptible to structural damage due to fluid slamming, especially under severe ocean conditions. Therefore, it is crucial to study the structural reliability of the pressurizer.In order to analyze this issue,this study employs a six-degree-of-freedom mobile platform to input various motion excitations. Strain gauges were attached to different locations of the internal components to measure strain responses,and this study conducts a comprehensive time-frequency domain analysis of the structural responses caused by fluid slamming under various operating conditions using the Fast Fourier Transform (FFT) and Continuous Wavelet Transform (CWT). The research results show that the strain on the first layer of the control rod guide tubes in the direction of motion is significantly greater than that at other locations, but it remains below the material's yield limit, and it has little influence on the equipment. In coupled motions, the location of maximum strain is determined by the motion with the largest amplitude, and the strain on internal components located in directions with smaller motion amplitudes is influenced simultaneously by this motion and the motion with the largest amplitude.This study provides insights into the structural response of pressurizer internal components under fluid slamming in marine environments.

To reduce computational expenditures, thermal radiation has often been neglected in system or CFD code simulations of containment flows involving infrared-absorbing water vapor. However, large-scale experiments at PANDA facility, along with related benchmarks using CFD tools have demonstrated that radiative heat transfer is significant, even at very low steam content. Accordingly, the present work focuses on exploring the effects of thermal radiation within a smaller containment environment.

Two tests with different steam volume concentrations (0.1 and 2.5%) were conducted in the medium-scale LINX facility, which is a vessel of 2-meter diameter and 4-meter height (1/10 PANDA drywell volume). Steam/Air mixture was compressed by injecting Helium from the top at a mass flow rate of 1g/s for a duration of 1200 seconds. This led to the formation of a helium layer at the top, which pushed down the steam/air mixture, creating a high-temperature bubble. Additionally, CFD simulations of both tests were performed using ANSYS Fluent, employing the k-ω SST turbulence model and the P1 model to incorporate the thermal radiation. The steam absorptivity was treated with the Weighted Sum of Gray Gases Model (WSGGM).

The experiments showed that 0.1% steam test yielded a significantly higher peak temperature compared to the 2.5% steam test. CFD simulations without inclusion of thermal radiation highly overestimated the temperature profiles. Meanwhile, a good match with experimental data was achieved using the P1 model. Overall, these results highlight the importance of considering thermal radiation when modeling naturally driven flows in steam-containing atmospheres, even at smaller scales.

Understanding the coolant flow behaviour in the primary heat transport system (PHTS) is crucial for reactor safety analysis of accident scenarios. Databases from rigorously designed experiments are necessary to support the modeling of complex coolant flow behaviours. This study focused on analyzing the two-phase flow distribution in an important component of a typical CANDU-type PHTS, namely the header-feeder system, under flow conditions relevant to large break loss-of-coolant accidents (LB-LOCA). Experiments were conducted on a 1:3 scaled Header Test Facility that replicates the piping configuration representative of a CANDU PHTS (specifically the Advanced CANDU Reactor, ACR-1000) header/feeder system using air-water as the working fluids to simulate steam-water. The experiments simulated a scenario where a large break occurs at the inlet header and the emergency core cooling system fails to activate. Flow conditions were varied using a range of water and air flow rates, simulating different levels of a LOCA. A total of 116 wire mesh sensors were installed along the header and feeder pipes, measuring various two-phase flow parameters, including instantaneous void fraction, bubble size and interfacial velocity distribution in the flow channel. Experimental data revealed that the void fraction and air flow rate in each feeder depend not only on the initial break condition and local phenomena in the headers, but also on the location where the feeder is connected to the header and the elevation of the feeder pipe.

In molten salt reactor (MSR) systems, tritium is generated in significantly larger quantities when compared to other reactors. Due to tritium’s high permeability and solubility, tritium must be removed from the molten salts to reduce the associated radiological risk. Gas sparging is used for the online separation and removal of fission products within FLiBe where noble gas is bubbled through the molten salt to remove dissolvable fission gasses. To advance gas sparging technology, we are doing experiments to develop a detailed understanding of gas sparging behavior and create benchmark datasets for Multiphysics Object Oriented Simulation Environment (MOOSE) models. To do this, we have designed and built the Modular Separate Effects Test Facility (MSEFT) - Tritium Removal Investigation of Transport Interactions Using Mass-transfer (TRITIUM) flow loop to observe and measure localized bubbling behavior and integral gas concentrations using surrogate fluids. The Normalized Dissolved Oxygen Concentration (NDOC) within the surrogate fluid will be reported on for different prototypical conditions. The NDOC is used to characterize the liquid-gas mass transfer coefficients of sparging bubbles within water at various glycerol weight percentages used to match relevant non-dimensional numbers of Reynolds, Sherwood, and Weber. The NDOC data will be combined with measurements of relevant local bubble dynamics including average bubble diameter and velocimetry taken with high-speed shadowgraphy and Particle Image Velocity systems. These combined measurements will be useful to inform future design improvements for different MSRs’ gas sparging components.

In prior studies, we introduced an optimized prototypical natural circulation water-based reactor cavity cooling system (RCCS) design for a pebble-bed generic FHR based on 1D modeling and in later work performed more detailed 2D transient performance analyses. The prototypical design was an initial step to facilitate experimentation. Experiments in a university laboratory necessitate downscaling of the prototypical RCCS while maintaining key non-dimensional parameters such as Grashof number, particularly in systems involving natural circulation. The design of the RCCS, core, and the radiated power significantly affect the non-dimensional parameters in the RCCS. Tradeoffs exist between the non-dimensional parameters as a combination of design parameters may yield ideal scaling for one parameter but result in unacceptable scaling for other parameters. In this work, we systematically study the dependence of the relative scaling of the non-dimensional parameters compared to the prototypical case as a function of design and physics parameters using an iteratively refined base case design. These non-dimensional parameters considered herein include Grashof, Reynolds, Nusselt, Prandtl, Rayleigh, Biot, Stanton, Froude, and Richardson numbers. An idealized, downscaled design was obtained based on these analyses and a practical experimental setup was subsequently designed.

Kairos Power recently obtained NRC approval for construction of its high-temperature fluoride-salt cooled pebble bed reactor, Hermes, and is presently working towards submission of its operating license. As part of this activity, Kairos is using the KP-SAM systems code to perform safety analysis of the reactor under transient conditions. As part of the verification & validation (V&V) of this code, Kairos has built and operated a reduced-scale integral effects test (IET) facility which leverages hierarchical two-tiered scaling (H2TS) to simulate a loss of forced circulation (LOFC) in the reactor in which a pump failure leads to the onset of natural circulation. This facility employs a surrogate heat transfer fluid which simultaneously matches several dimensionless numbers – such as the Prandtl number – of molten salt at a significantly reduced temperature and scale, enabling rapid testing with high-fidelity instrumentation. This paper discusses the scaling, testing campaign, results, and future plans for the IET.

We present selected integral effect test results conducted in the STELLA 2 facility on the decay heat removal systems (DHRS), along with numerical analyses with MARS LMR system thermal hydraulics code. STELLA 2 is a large scale sodium test platform modeled after the Korean sodium cooled fast reactor, PGSFR, with a 1/5 scale reduction in length. The facility aims to evaluate the overall plant dynamics and safety aspects of the reactor under long term transient conditions. With equivalent conservation of the reactor components’ shapes and layouts, reactor transients can be investigated while maintaining the integral behavior and interactions of the prototypic reactor’s heat transport systems with minimal distortions. The facility has four individual DHRS loops, comprising two active and two passive loops differentiated by their cooling mechanisms for ultimate heat sinks. The selected tests focus on scenarios where the primary system after reactor shutdown relies solely on two DHRS loops, postulating a loss of offsite power condition. Overall, the system code successfully captured the general transient behaviors in the individual loops, although variations were observed in the peak temperature and the time to reach it, compared to the experimental data. It is believed that discrepancies in local temperature distributions, particularly in larger volumes, are attributed to the system code’s limitations in discretization method and relations.