Clinical findings and prognosis of interference injuries to the palmar aspect of the forelimbs in Standardbred racehorses: A study on 74 cases

Information on interference injuries in racehorses is lacking

International Standard Problem on containment thermal-hydraulics ISP47 - STEP1 - Results from the MISTRA exercise

International audienceThe understanding of hydrogen distribution during severe accidents in a nuclear reactor containment is still an open issue. Several containment thermal-hydraulics international standard (ISP) have been conducted to address this topic. However the predictions made by the available Lumped Parameter or CFD computer codes were generally not satisfactory. Therefore a new exercise was launched in 1999 using new state-of-the-art experimental facilities TOSQAN, MISTRA and ThAI that included sophisticated 3D instrumentation and well controled boundary conditions. Predictive capabilities of important and still uncertain phenomena such as wall condensation, natural circulation and gas stratification are assessed. In addition, comparison between LP and CFD codes and assessment of the capability of CFD codes to deal with scaling effects are performed. This article reports on the part of the exercise which concerns the MISTRA facility includind experimental results and blind benchmark exercises

European research on issues concerning hydrogen behaviour in containment within the SARNET network of excellence

The paper describes European research on issues concerning hydrogen behaviour in containment within the SARNET network of excellence. The hydrogen combustion and associated risk mitigation have been studied, concentrating on the formation of combustible gas mixtures in containments, its local gas composition and potential combustion modes, including the reaction kinetics inside catalytic recombiners. The potential hydrogen distribution in different parts of the containment has been investigated, with the objective of assessing the formation of combustible gas mixtures taking into account the effect of mitigation systems such as sprays or recombiners. Due to its influence on hydrogen distribution, steam condensation is also being investigated

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

International audienceThe safety of gas-cooled reactors (High Temperature Reactors HTR, Very High Temperature Reactors VHTR or Gas cooled Fast Reactors GFR) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal-hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modelling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderation coefficient),critical position of control rods, reactivity insertion aspects For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M-ARCTURUS thermal-hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal-hydraulic exercise. Examples of containment thermal-hydraulics calculations for fast reactor design (GFR) are also detailed

EU-ERCOSAM PROJECT, scaling from nuclear power plant to experiments

International audienceIn case of a severe accident in a light water nuclear reactor, hydrogen would be produced duringreactor core degradation and released into the reactor building. The stratification of the released hydrogen in the reactor containment could lead to local pockets of gas mixtures of high hydrogen concentration and, in case ofcombustion, to high pressure loads which might challenge the containment structural integrity.The objectives of ERCOSAM and SAMARA projects, co-funded by the European Union and the Russia, are to investigate hydrogen concentration build-up and break-up due to safety components operations, as sprays, coolers and Passive Auto-catalytic Recombiners (PARs).For this purpose, various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT) are considered. This paper describes the work performed in framework of the workpackage WP1 of the ERCOSAM project and presents theadopted methodology to scale down the real plant calculations results, provided by the projects partners, to the experimental facilities