Several nuclear reactor designs rely on passive containment cooling systems. The so-called containment wall condenser relies on natural circulation loops to extract heat from high-temperature steam in the containment to a water tank at ambient pressure. In such passive systems, phase changes can happen and cause flow instabilities in the cooling loop. The flashing-induced instability occurs when the heated fluid in the riser suddenly vaporizes due to a hydrostatic pressure decrease. This instability causes periodic flow peaks, which are of major concern but whose characteristics have not been studied quantitatively.This paper presents two analytical models that predict the flashing frequency and a maximum flow amplitude from geometry and basic operating parameters such as power level and reservoir temperature. The expressions are derived from a physical analysis and do not involve any calibration constants. The flashing frequency appears to be driven by the power level, the inlet temperature and the riser pipe geometry. For the amplitude, the maximum flow rate can be expressed in a Froude number that depends only on the total pressure losses. These models are validated against PASI experiments and system-scale simulations with the CATHARE 3 code, both performed as part of the European Commission funded PASTELS project. Additional data from numerous experimental studies in the literature are used to extend the validity range of the frequency model.Successfully validated against experimental data and additional simulations, these models provide an explicit relationship between oscillations characteristics and design parameters, making them valuable tools for nuclear engineers.

Two-phase flow head losses across tube support plates in U-bend tube steam generators are classically evaluated from single phase head losses duly modified with a correction factor to account for two-phase effects. This communication reports upon a series of experiments to feed head loss models for clogged tube support plates under two-phase flow. By using a classical approach, a single phase head loss model is used together with a two-phase flow multiplier. The experiments cover a large span of clogging levels, up to the unrealistic limit case of 95% of clogging. Singular head loss due to the tube support plate is measured under single-phase flow for clean and clogged plates. The head loss factor is hence evaluated and compared to the prediction of some available models. Then the same experiment is repeated under two-phase flow condition. By combining the previously measured single-phase factor with the two-phase flow measurements, recommendations for predicting two-phase flow multiplier with existing models are given.

After a period of several years of operation, steam generators can be affected by fouling and clogging. Fouling means that deposits of sludge accumulate on tubes or tube support plates (TSP). That results in a reduction of heat exchange capabilities and can be modelled by means of a fouling factor. Clogging is a reduction of flow free area due to an accumulation of sludge in the space between TSP and tubes. The increase of the clogging ratio results in an increase of the overall TSP pressure loss coefficient.The link between the clogging ratio and the overall TSP pressure loss coefficient is the most important aspect of our capability to accurately calculate the thermal-hydraulics of clogged steam generators. The aim of the paper is to detail the experimental approach chosen by EDF and AREVA NP to address the calculation uncertainties.The calculation method is classically based on the computation of a single-phase (liquid-only) pressure loss coefficient, which is multiplied by a two-phase flow factor. Both parameters are well documented and can be derived on the basis of state of the art methods such as IDEL'CIK diagrams and CHISHOLM formula.The experimental approach consists of a validation of the correlations by performing tests on a mock-up section with an upward flow throughout a vertical array of tubes. A mixture of water and vapour refrigerant R116 is used to represent two-phase flows. The tube bundle is composed of a 25 tubes array in a square arrangement. The overall height of the mock-up is 2 m. Eight test TSPs were manufactured, considering eight different clogging configurations: six plates with a typical clogging profile at six clogging ratios (0, 44%, 58%, 72%, 86%, 95%), and two plates with a clogging ratio of 72% associated with two different clogging profiles (large bending radius profile and rectangular profile).A series of tests were performed in 2009 in single-phase flow conditions. Two-phase flow tests with a mixture of liquid water and vapour refrigerant R116 will be performed in 2010. The paper illustrates the main results obtained during the single-phase tests performed in 2009.

During operation, sludge steadily appears at a slow pace on the secondary side of nuclear power plant steam generators. This leads to clogging of the tube bundle support plates, and consequently to a change in the thermal-hydraulic flow conditions. The circulation ratio of a steam generator is defined as the ratio between the total flowrate circulating in the riser and the steam flowrate at the outlet of the steam generator. This is a good indicator of the hydraulic pressure losses in the circulation loop. In particular, the increase in hydraulic resistance due to the tube support plate clogging leads to a drop in this parameter. For this reason, in order to check that clogging does not reach too high a level, the circulation ratio is regularly evaluated on steam generators of French nuclear power plants, and then compared to established safety limits. The purpose of this paper is to present an accurate method to determine the circulation ratio of a steam generator based on temperature measurements taken around the wall of the steam generator. This method consists of carrying out a thermal balance of the flow circulating in the downcomer. In order to accomplish this, the temperature of the water circulating in the downcomer is evaluated using thermocouple belts put on the external wall of the appliance. However, additional hypotheses in the calculation method are considered in order to take into account for the heat transfer between hot water inside the downcomer and the sensors. The steam generator circulation loop and the clogging of the tube support plates are presented in §1. Then §2 and §3 describe in detail the method and the associated hypotheses as well as the required instrumentation. Finally, §4 presents an application of this method to real cases of clogged steam generators.

During the normal cycle of a pressurized water reactor, boron concentration is reduced in the core until fuel burns up. A stretch out of the normal cycle is however possible afterwards, provided primary coolant temperature is reduced. In those stretch out periods, nuclear operators want to keep constant thermal power exchanged in the steam generator, in order to preserve its performances. Under that constraint, the required reduction in primary coolant temperature involves both a decrease of secondary cooling system pressure and an increase of tube bundle vibrations. Since neither pressure nor vibrations should exceed some given thresholds in order to preserve component integrity, the reduction of primary coolant temperature has to be limited. Nuclear plant operators thereafter need an operating diagram, i.e. a diagram that provides minimum allowed primary coolant temperature versus power rate. In that context, we propose a method to derive such a diagram, by combining, on the one hand a code for simulating primary and secondary fluid flows in steam generators and, on the other hand, a software that allows one to predict fluid elastic tube bundle instabilities. That method allows one to take into account both tube fouling and plugging. It is now used by French utility “Electricite´ De France”, in order to check or supplement the analysis that are provided by steam generator manufacturers.