Conference Agenda

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.