Conference Agenda

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.