Conference Agenda

In the case of a LOCA in the PWR, during the long-term cooling stage, boric acid in the coolant may be carried with steam or entrained droplets due to steam discharge, which may affect the reactivity of the core. There are a number of research that make it possible to determine the value of droplet entrainment leaving the reactor. However, the loss of boric acid with steam is related to the solubility at steam. It is necessary to study dissolution process of boric acid in steam under different parameters.

This paper conducts an experimental study of the distribution coefficient of boric acid between the steam and liquid phases of the solvent. The test device of the solubility of boric acid with steam has been established, including a heated test section, steam separator, and steam condenser. The experiments are conducted under a certain pressure of 0.1-0.4MPa, with an initial boric acid solution concentration range of 1000-10000 ppm. The concentration of boric acid in the discharged steam and the remained solution are obtained. The effects of pressure, temperature and initial concentration on the solubility of boric acid in steam are summarized. The empirical relationship of distribution coefficient of boric acid in the steam phase and liquid phase are obtained by fitting the experimental data.

This study investigates the flow characteristics in a 3x3 Helical Cruciform Fuel (HCF) bundle using Magnetic Resonance Velocimetry (MRV) experiments and Computational Fluid Dynamics (CFD) analysis. The HCF bundle, designed to enhance thermal-hydraulic performance in nuclear reactors, features a unique geometric configuration with helically twisted cruciform-shaped fuel elements. To validate the numerical predictions, experimental measurements were conducted using MRV technology, which provides three-dimensional velocity field data without intrusive flow disturbance.

The experimental facility consisted of a full-scale 3x3 HCF bundle model operating at high Reynolds numbers exceeding 10,000. MRV measurements captured the complex flow structures, including secondary flows and vortex formations in the sub-channels. The CFD analysis employed the Reynolds stress model with a refined mesh containing approximately 2.5 million elements, validated through rigorous mesh sensitivity studies.

The study revealed distinct flow patterns characterized by enhanced mixing due to the helical geometry. Secondary flows were particularly pronounced in corner sub-channels, exhibiting higher tangential velocities compared to interior sub-channels. These findings provide crucial validation data for CFD methodologies in nuclear fuel bundle analysis and contribute to understanding the thermal-hydraulic behavior of advanced fuel designs.

The validated CFD model can serve as a reliable tool for future HCF bundle optimization studies and thermal-hydraulic characteristic analyses, potentially leading to improved nuclear reactor fuel efficiency and performance.

Thermal fatigue cracking may initiate at a T-junction where high and low temperature fluids flow in. In this study, the flow structure and fluid temperature fluctuations in a branch pipe of a T-junction were investigated under flow patterns where the main pipe flow penetrates into the branch pipe. The test section consists of a horizontal main pipe with an inner diameter of 150 mm and a vertical branch pipe with an inner diameter of 50 mm. A 45º elbow was installed upstream on the branch pipe side in order to study its effect on the penetration flow pattern and temperature fluctuations. To simulate penetration flow, the experiment was conducted under conditions where the inlet flow velocity on the branch pipe side was much smaller than the inlet flow velocity on the main pipe side. The flow pattern was visualized using the tracer method. The flow in the branch pipe was classified into three flow patterns: no penetration; entrained penetration; and impinged penetration. These patterns depended on the momentum ratio of the main and branch pipes, regardless of the presence of the elbow. The maximum penetration depth into the branch pipe increased when the upstream elbow was installed. Fluid temperature distribution along the branch pipe was measured with eight sheathed thermocouples. The fluid temperature fluctuations also increased, especially in the range of relatively small momentum ratios where the hot mainstream intermittently penetrated into the branch pipe.

Fatigue and defect tolerance assessments of high integrity PWR pressure boundary components require transient temperature fields to be defined to produce thermal stress predictions. These temperatures are often produced using a heat transfer coefficient and bulk temperature boundary condition approach. This can be imprecise and inaccurate for components with complex flows and geometries. An alternative approach is to directly predict the temperatures using conjugate heat transfer CFD, where the solid temperature field is predicted directly and simultaneously with the adjacent flow. This approach removes the uncertainties of using an intermediate model to transfer the information, but since it requires flow predictions at all times, the computational cost is impractical for the large number of multi-hour plant transients that need to be considered.

The cost of the CFD solution can be reduced by using an 'infrequent updates' approach, where the flow-field is considered to change slowly and is 'frozen' for intervals where only the thermal fields are solved. This is cheaper to calculate and can use longer time steps. The flow is solved in brief update intervals throughout. This method has been applied to the assessment of a large number of transients for the feedwater nozzles for the steam generators at the Sizewell B PWR in the UK.

The setup considerations and accuracy of the infrequent updates approach are discussed, as well as the effects of finite domain size and buoyancy driven flow instabilities.

Hydroaccumulators are designed to provide fast water injection into the vessel to ensure proper core cooling. One of its main function is that during LOCA cases, when no HPSI pumps are available, the hydroaccumulator supplies cooling water until LPSI pumps can operate and maintain the long term cooling. In VVER-440 reactors the hydroaccumulator empties at higher pressure than the LPSI can operate. The core should survive that injection hiatus. In that sense the primary pressure when hydroaccumulator injection terminates is very important

In PMK-2 facility several experiments were conducted utilizing various ECCS configurations and these tests were later calculated using RELAP5 thermal-hydraulic system code. During these post-test calculations a difference between the measured and calculated hydroaccumulator injection characteristic and terminating pressure value was noticed.

The hydroaccumulator is pressurized using nitrogen gas. Under pressure, some of this gas gets dissolved into the coolant in the hydroaccumulator water. During injection, the dissolved gas is reintroduced into the gas dome increasing its pressure. The RELAP5 system code does not consider this effect, leading to the observed differences. This is an unconservative deviation since the code predicts lower pressures at HA emptying thus shorter injection hiatus.

To address the phenomena experiments were performed. Using the PMK-2 facility more than 70 separate effect tests were conducted using one of the hydroaccumulator vessels. The tests were done at several different pressure levels and coolant temperatures. RELAP5 post-test calculations were carried out for each test, and the effect of introducing additional gas into the vessel was studied.

A significant percentage of reactors in operation, under construction or recently commissioned are VVER reactors. In parallel, there is a growing interest in analyzing the behavior of the Advanced Technology Fuels (ATF) under development in this type of nuclear reactor. Among the new ATF designs, the FeCrAl and Cr-coated claddings are the most promising evolutionary options. In addition to a relatively high level of technology readiness, these evolutionary cladding materials offer improved oxidation and hydride resistance, as well as improved mechanical strength. All these properties are essential in accident sequences where high core temperatures are reached. In the present study, core damage sequences have been analyzed with a model of a VVER-1000 reactor for the thermal-hydraulic code TRACE. In addition, an in-house version of the TRACE5P6 system code for FeCrAl cladding has been developed by NFQ and UPM.The selected sequences are SBO and LOCA sequences. The results show that the core damage temperature for the Zircaloy cladding cases is reached well before that for the FeCrAl and Cr coating cladding cases. The performance of these new cladding materials provides additional time for recovery of the safety systems responsible for core cooling and replenishment of the reactor coolant system inventory.