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• Evaluation of local void effect with deterministic calculation is very challenging. • An original 2-steps deterministic scheme is proposed using APOLLO3® code. • The subgroup selfshielding method gives the most accurate void effect estimation. • A combination of finestructure and Tone's selfshielding methods is calibrated. • A Local Flux Volume normalization of SPH equivalence coefficients is proposed. In some countries, as France, plutonium is used in industrial Pressurized Water Reactors (PWRs), using Mixed Oxides (MOX) fuels. To avoid a reactivity increase during a loss of coolant accident, the plutonium content in a PWR is limited at 12 wt%. This limit, that have been determined in the early 1990 s using a very conservative approach, is a strong constraint on the possibility to stabilize the plutonium inventory. Thus, further studies are now conducted to refine the plutonium limit content in the fuel. One interesting case is the local moderator voiding, which could for instance occurs at the beginning of a core depressurization accident. To study this phenomenon, a 2-steps neutronic deterministic scheme is proposed in this paper, using the APOLLO3® code. It is shown that the choice of the selfshielding method to use at the lattice step has a strong influence on the precision of the calculation. It is shown that, as expected, the subgroup formalism is the more precise to cope with both nominal and voided configurations. However, methods relying on this formalism are numerically costly. The use of a method relying on the finestructure formalism allow to conserve a satisfying estimation of the void effect on the reactivity or the spatial distribution of energy-integrated reaction rates. Unfortunately, we show that those good results are due to compensation of large energetic biases in the nominal case. A combination of Tone's and finestructure selfshielding methods is presented. This method permit to strongly improve per-energetic-groups precision in the voided configuration compared to the method solely relying on the finestructure formalism, while drastically decreasing the numerical cost and duration compared to the subgroup method. The necessity to perform environed lattice calculation is demonstrated in the case of the voiding of a single assembly. A Local Flux Volume normalization of SPH coefficients is proposed to run equivalence calculation on this heterogeneous geometry. A method of characteristic solver is used for the flux calculation at the lattice step, while a S 8 transport calculation is performed at the core level, using pin-by-pin discretization and a 26-groups energy mesh. Finally, the precision of the calculation scheme is assessed by comparison to Monte Carlo calculation ran with TRIPOLI–4® on a two-dimensional clusters of assemblies. [ABSTRACT FROM AUTHOR]

• Flow and rupture study of liquid film on corrugated plate wall are reviewed. • Chaotic analysis of film thickness series will be a prospect research direction. • Film rupture mechanism cannot be explained by surface wave disturbance model. • More study will be focused on rupture phenomenon at the corrugated plate corner. Steam generated by steam generator in a nuclear power plant must firstly be subjected to steam-water separation treatment before entering the turbine. The research on flow and rupture characteristics of the liquid film on the wall surface of corrugated plate has important and practical significance for improving the separation efficiency. Since the 1940s, researchers from the United States, France, Japan and other countries have carried out related researches. With the rapid development of steam-water separation technology, many scholars have carried out a large number of experiments, theoretical study and numerical calculations on this phenomenon. Based on the published papers, the research progress and significant achievements of the flow and rupture characteristics of the liquid film on corrugated plate wall are reviewed. Shortcomings of past research and prospects for future research are summarized and proposed, respectively, which provides a reference for improvement of separation efficiency of the corrugated plate dryer. [ABSTRACT FROM AUTHOR]

Abstract: Although MOX fuel coming from reprocessed depleted uranium fuels has been used as a recycling strategy by countries like France and Japan it is not a common policy in the 30 countries that uses nuclear power, nowadays it seems to be a more direct alternative to reduce the depleted fuel interim storage. Previously, the spent fuel pools of Laguna Verde Nuclear Power plant were redesigned to host the total operating life depleted fuel under its original nominal power condition, however the plant has been up-rated to 120% of its original nominal power increasing the number of depleted fuel forecasted. This new situation makes necessary the analysis of alternatives, being one of them recycling. The current paper assesses the viability of using MOX fuel in the up-rated Power Plant; the design of the boiling water reactor MOX fuel addresses the two main constraints of its use: shutdown margin and reactor stability. Fuel design proposed sets the appropriate MOX enrichment and the maximum MOX fuel batch reload that does not imply any modification to the reactor control systems to avoid an extra economical cost due to its use. [Copyright &y& Elsevier]