Annotated record of the detailed examination of Mn deposits from SAFARI II cruise stations

The cores and dredges described at this site were taken on the SAFARI II cruise from 21 July until 1 September 1981 by the MusÈum National d'Histoire Naturelle from the R/V Marion Dufresne. A total of 65 cores, dredges and camera stations were recovered and are available at MNHN for sampling and study.

Geometric Integrability of the Camassa-Holm Equation. II

It is known that the Camassa–Holm (CH) equation describes pseudo-spherical surfaces and that therefore its integrability properties can be studied by geometrical means. In particular, the CH equation admits nonlocal symmetries of “pseudo-potential type”: the standard quadratic pseudo-potential associated with the geodesics of the pseudo-spherical surfaces determined by (generic) solutions to CH, allows us to construct a covering π of the equation manifold of CH on which nonlocal symmetries can be explicitly calculated. In this article, we present the Lie algebra of (first-order) nonlocal π-symmetries for the CH equation, and we show that this algebra contains a semidirect sum of the loop algebra over sl(2,R) and the centerless Virasoro algebra. As applications, we compute explicit solutions, we construct a Darboux transformation for the CH equation, and we recover its recursion operator. We also extend our results to the associated Camassa–Holm equation introduced by J. Schiff.

Study of the II Network as the Compound Slot Equivalent Circuit Model

A combination of Method of Moments (MoM) and compound slot Equivalent Circuit Model for linear array design is presented in this document. From the S Matrix of the single element, the more suitable network for its characterization is analyzed and selected. Then according to the radiation requirements of the desired array, the elements are designed and then properly connected by means of Forward Matching Procedure (FMP), which takes into account impedance matters in order to keep the input matched at the designing frequency. Comparison between HFSS simulations and MoM-FMP results are also presented. First part of this work was introduced in (1)(2) but a summary is included here to make the understanding easier.

Methodologies for an improved prediction of the isotopic content in high burnup samples. Application to Vandellos-II reactor core

Fuel cycles are designed with the aim of obtaining the highest amount of energy possible. Since higher burnup values are reached, it is necessary to improve our disposal designs, traditionally based on the conservative assumption that they contain fresh fuel. The criticality calculations involved must consider burnup by making the most of the experimental and computational capabilities developed, respectively, to measure and predict the isotopic content of the spent nuclear fuel. These high burnup scenarios encourage a review of the computational tools to find out possible weaknesses in the nuclear data libraries, in the methodologies applied and their applicability range. Experimental measurements of the spent nuclear fuel provide the perfect framework to benchmark the most well-known and established codes, both in the industry and academic research activity. For the present paper, SCALE 6.0/TRITON and MONTEBURNS 2.0 have been chosen to follow the isotopic content of four samples irradiated in the Spanish Vandellós-II pressurized water reactor up to burnup values ranging from 40 GWd/MTU to 75 GWd/MTU. By comparison with the experimental data reported for these samples, we can probe the applicability of these codes to deal with high burnup problems. We have developed new computational tools within MONTENBURNS 2.0. They make possible to handle an irradiation history that includes geometrical and positional changes of the samples within the reactor core. This paper describes the irradiation scenario against which the mentioned codes and our capabilities are to be benchmarked.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.