Wound contact layer (WCL) dressings are intended to protect tissue during the healing process. A randomised controlled trial was undertaken to compare 2 such dressings. Outpatients with acute wounds were randomly allocated to treatment with either a soft silicone‐coated WCL (intervention group, n = 59) or a lipidocolloid‐impregnated WCL (control group, n = 62). At the first dressing removal (day 3), 89.8% of patients in the intervention group experienced non‐painful dressing removal (defined as a pain rating <30 mm on a 100 mm visual analogue scale), compared with 73.6% of patients in the control group (P = .017) (per protocol population). At day 21, wounds were considered as healed in 66.1% of patients in the intervention group compared with 43.5% in the control group (P = .012) (intention‐to‐treat population). Both dressings were well tolerated and rated highly in terms of in‐use characteristics, although the soft silicone‐coated WCL was rated significantly higher than the lipidocolloid‐impregnated WCL in terms of its ability to remain in place (P= .016). The results indicate that the soft silicone‐coated WCL is suitable for the management of acute wounds as it can minimise dressing‐associated pain and support healing.
EDF is involved with CEA and AREVA in a common effort for the development of the future nuclear reactor generations. The studies, currently performed by the partners, concentrate on the design of Sodium Fast Reactor types that may include different kinds of innovative circuits and components as compared to the SPX (Super PheniX) plant design. Based on previous knowledge on SG developed at EDF, with Sodium as hot fluid, and with the help of more recent methods of modeling using the Modelica libraries, a new model for the simulation of steam generator has been developed in order to help the designers of the heat exchangers to meet the requirements for a Sodium Fast Reactor plant design. The paper will present the current status of the model and a comparison of the results with those of the actual SuperPhenix steam generator database.
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.
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.
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