Conference Agenda

The use of molten salt reactors (MSRs) presents a promising avenue for achieving energy independence and reducing reliance on fossil fuels. A key challenge in MSR development is enhancing heat exchanger efficiency while minimizing pressure drop and operational costs.
This study experimentally investigates the thermal-hydraulic performance of additively manufactured helically rifled tubing, a passive heat transfer enhancement (HTE) method that promotes flow mixing through centrifugal forces. The pressure drop characteristics of helically rifled tubes with different twist pitches (y = 3 and y= 6) were measured and compared to plain tubing, alongside computational fluid dynamics (CFD) simulations. Using Buckingham-Pi analysis, nine dimensionless groups were identified to characterize flow behavior and thermal performance. Preliminary experimental results indicate that existing helical rifling correlations significantly under predict frictional losses by 10%, highlighting the need for refined modeling approaches. A strong agreement was observed between simulations performed by Vulcanoff et al. and the experimental friction factor data for the helical rifling at the same geometric parameters (y=3). Geometric variation of the twist pitch from y=3 to y=6 suggests the dependence of the nondimensional parameter y/d_min diminishes at sufficiently high Reynolds numbers. Computational findings by Vulcanoff et al. suggest that helically rifled tubes improve heat transfer efficiency. Investigations are ongoing into the Nusselt number and thermal performance factor. This work provides a foundation for optimizing helical rifling geometries in MSRs, contributing to advancing advanced nuclear energy systems.

A new phenomenon, namely tangential oscillation of thermal stratification, can cause periodic temperature changes on the pipe wall and leads to material damage of thermal fatigue. At North China Electrical Power University, an experimental facility has been constructed and operated to investigate the initiation mechanism of the tangential oscillation of thermal stratification at pipeline elbows. In this study, experiments have been performed with variations of elbow-radiuses and inlet flow temperature. Results show temperature increase at the intrados side of the elbows, which indicates an angular shift of the thermal stratification at the elbow due to the centrifugal force. Thermocouples downstream of the elbow have captured temperature changes in the near-wall flow. The elbow-radius shows a clear influence on the locations of the high temperature region in the thermal stratification. In addition, temperature fluctuations have been calculated based on the measurement data. The location with the maximum temperature fluctuation can be found in the place, where the mean temperature reaches the maximum. Moreover, frequency spectra of the temperature data do not show any significant peak. Combined with the calculation results of Richardson-number, it can be understood that the thermal stratification is not stabile enough to keep the tangential oscillation downstream of the elbow. It leads back to the turbulent mixing enhancement at the elbow, which is clearly increased with decrease of elbow-radius. However, decrease of the elbow-radius leads to increase of the temperature fluctuation in the near-wall flow, which indicates a higher potential of thermal fatigue.

The Small Modular Reactors (SMRs), as an emerging nuclear energy technology, hold great promise for gradually replacing coal-fired power plants in the future due to its flexibility, high safety, and economic advantages, thereby contributing to the decarbonization of energy systems. In a domestically developed integrated SMR in China, the steam generator adopts a helical heat transfer tube design, with the outlet steam being saturated steam. Considering the compact size and high-level power density characteristics of SMRs, it is necessary to design an efficient and compact moisture separation component, to ensure that the quality of steam entering the turbine meets the required standards. This study conducted cold-state experiments and theoretical analyses on a designed swirl vane gas-liquid separator. Under conditions with different drainage step heights, the critical separation boundaries for both low-wetness and high-wetness scenarios were determined. For steady-state conditions, based on experimental data of the separation efficiency and pressure drop of guide vanes with different profile variation patterns, a predictive correlation was proposed. For unsteady conditions, a predictive model was developed to describe the transition from swirl annular flow to churn flow, and a flow regime map was constructed by integrating extensive experimental data with previous studies. These findings provide important theoretical support and experimental evidence for the optimization and performance enhancement of SMR steam-water separation systems.

Globally, advanced reactors are adopting helical tube steam generators to reduce volume compared to traditional U-tube designs in large light water reactors. These steam generators feature a once-through operation and a shared header for multiple heat transfer tubes but often encounter issues associated with two-phase flow instabilities such as Density Wave Oscillations (DWO). Such instabilities are critical in boiling water reactor cores and pressurized water reactor steam generators, causing significant flow and pressure oscillations, which can potentially degrade the integrity of a system.

This study aims to address this gap by comparing experimental results on two-phase flow instabilities in helical tubes from previous research works with the predictions obtained from Korean nuclear safety codes MARS-KS and SPACE. The objective is to assess whether the current helical tube thermo-hydraulic models in these codes adequately reflect observed physical behaviors or if there are significant discrepancies that require model enhancements. This analysis intends to provide insights into the dynamics of two-phase flow in helical steam generators and help improve the predictive accuracy of safety codes, thereby enhancing the reliability and safety of advanced nuclear reactor.

In nuclear power systems, the insertion of a throttling device at the inlet of the heat exchange tubes in steam generators enhances the flow resistance within the single-phase flow region of the tubes, thereby mitigating the risk of flow instability within the steam generator. This study proposed a novel gear-type throttling device designed specifically for steam generators. Various gear-type throttling prototypes with differing gear heights were designed and fabricated for experiment and numerical analysis .Through a systematic experimental testing and numerical simulations, the resistance characteristics of the throttling device with different structural parameters were obtained in a wide range of flowing conditions, . The results reveal that the resistance coefficient of the innovative gear-type throttling device can fit for different operational requirements in steam generators. The resistance coefficient exhibits significant sensitivity in gear height and width. Additionally, the resistance coefficient for throttling devices with varying gear heights remains relatively stable across different Reynolds numbers.A mathematical relationship was established to correlate multiple structural parameters and the resistance coefficient. This work is valuable for the design optimization and validation of next-generation steam generators in the nuclear energy system.

The function of reactor core residual heat removal system is to extract heat from the core and reactor coolant systems during the shutdown of the power plant. According to the principle of simplified configuration, improving safety and economy, possible reactor core residual heat removal systems are studied, and the configuration of recommended optimization solution is fixed.The optimized configuration is based on two completely independent series; This configuration can be used for core residual heat removal function after connecting the primary circuit, and for containment spray function. the analysis of single failure and system capacity for this optimized configuration are performed. The analysis of single meets the safety requirements, except that the exemption criterion is used for the mode of residual heat removal, which can make sure the nuclear power plant is brought to the safe state.The capacity of key equipment including pumps and heat exchanger is analyzed,which is in making full use of the equipment of original some reactor, considering the equipment design, the shutdown time of plant, the design limit of containment and layout space comprehensively, the actual shutdown time and the pressure and temperature of containment are calculated, finally the capacity of equipment can meet the mode of residual heat removal and the mode of containment spray. The optimized reactor core residual heat removal scheme, not only improves the safety of the power plant, but also improves the economics, which has the great significance to the subsequent improvement of the market competitiveness of power plants.

Due in great part to the earthquake and ensuing tsunami, the Great East Japan Earthquake of 2011 seriously damaged the Fukushima Daiichi Nuclear Power Plant and resulted in a major accident. The malfunction of the cooling water level measuring system was one element causing this accident. Differential pressure gauge monitoring of the reactor pressure vessel (RPV) water level was used at the time. But the temperature of the reference side piping surged greatly during the severe accident, which caused the water on the reference surface to evaporate. It is not possible to precisely identify the real drop in water level since this evaporation lowered the differential pressure between the water level inside the reactor containment vessel and the reference piping.

In order to find a solution to this issue, we are working on a new water level meter that is capable of functioning even in severe accidents. Through the utilization of an ultrasonic transducer (TDX), this apparatus enables real-time measurements to be taken from outside the containment vessel of the reactor. The concept of measurement originates from the disparity in the reflection coefficients of ultrasonic waves that travel through metal that is in contact with water as opposed to air.

The results of experiments measuring the reflection coefficients of metal walls in contact with air and water using a small water level measuring device are reported in this work. Comparisons with hypothetical values computed with the ultrasonic wave propagation simulator SWAN21 verified the validity of the experimental results.