Types d'armes nucléaires

Résumé

Les deux grands types d'armes nucléaires se distinguent par leur fonctionnement : Armes à fission ou « bombes A » ; Armes à fusion, bombes thermonucléaires ou « bombes H ». Dans ces deux grandes familles, des armes plus spécialisées ont été conçues en fonction d'effets spéciaux recherchés : la plus connue est la bombe à neutrons. Principe de fonctionnement Explosion par fission ou « bombes A » right|thumb|Les premières armes nucléaires, tel Fat Man, étaient très difficiles à transporter et leur mise à feu était complexe. Bombe A Une explosion nucléaire par fission pure est la première étape, à la fois historiquement, et dans la conception des bombes à étage. Les bombes à fission (sans étage) furent les premières à être développées et sont communément appelées « bombes atomiques ». Elles se fondent sur le principe de la fission nucléaire et utilisent des éléments fissiles comme l'uranium 235 et le plutonium 239. Pour obtenir une explosion nucléaire, il est n

Source officielle

https://fr.wikipedia.org/wiki/Types_d'armes_nucl%C3%A9aires

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This thesis presents a systematic study on the merits and limitations on using pin-by-pin resolution and transport theory based approaches for nuclear core design calculations. Starting from the lattice codes and an optimal cross section generation scheme, it compares different methods, transport approximations, and spatial discretizations used in pin-by-pin homogenized codes. It is in the interest of nuclear power plant operators to employ more heterogeneous core loadings in order to improve the fuel utilization and decrease the amount of spent fuel. This necessarily increases the requirements on the accuracy of the computation tools used for the core design and safety analysis. One possibility is employing 3D core solvers with higher spatial resolution, e.g. pin-cell wise. The comparison of several lattice codes indicates that already the proper generation of diffusion coefficients and higher-order scattering moments for pin-cell geometry is not straightforward. Out of three available lattice codes, two generated unphysical diffusion coefficient when using the inscatter approximation, while the last code was not able to provide the higher scattering moments. Several few-assembly test cases with either high neutron leakage effects or MOX/uranium interfaces showed that the quality of the results of diffusion and SP3 solvers depends critically on the choice of the diffusion coefficient: in the outscatter approximation, it can cause major deviations, while the inscatter seems overall more adequate. Regarding the transport solvers, results obtained with the SN solver DORT showed very good performance for MOX/uranium interfaces, but major deviations occurred for problems with large power gradients. On the other hand, the MOC solver nTRACER showed in all cases small average error, but 2 - 3 % error around the interface of different assemblies. A study on spatial discretization indicated that the finite difference method applied on pin-cells does not properly capture the big flux changes between MOX and uranium fuel, while the nodal expansion method is more accurate but too slow. It was suggested to use the finite difference method with finer mesh in the outer assembly pin-cells, which increases the required computation time by only 50 % and decreases the pin power errors below 1 % with respect to lattice code results. Due to some problems which were observed with the available diffusion/SP3 solvers, a new SP3 solver was implemented in the DORT-TD platform. Several core tests showed that the SP3 pin-by-pin solver can significantly outperform the state-of-the-art nodal solver SIMULATE-5, in particular for reactor cores with inserted control rods. Finally, the pin-by-pin solvers were coupled to a depletion solver. For that, an accurate and fast interpolation routine had to be implemented. The obtained results of full-cycle depletion with complex core loading showed very good performance in comparison to a heterogeneous transport-based fine-group calculations.

EPFL2018

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A standalone decay heat removal device for the Gas-cooled Fast Reactor for intermediate to atmospheric pressure conditions

Rakesh Chawla, Aaron Simon Epiney, Konstantin Mikityuk

This paper reports a design study for a Brayton cycle machine, which would constitute a dedicated, standalone decay heat removal (DHR) device for the Generation IV Gas-cooled Fast Reactor (GFR). In comparison to the DHR reference strategy developed by the French Commissariat a l'Energie Atomique during the GFR pre-conceptual design phase (which was completed at the end of 2007), the salient feature of this alternative device would be to combine the energetic autonomy of the natural convection process - which is foreseen for operation at high and medium pressures - with the efficiency of the forced convection process which is foreseen for operation down to very low pressures.

Elsevier2012

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Using re-calibrated design S-N curves for fatigue assessment of road bridges

Alain Nussbaumer

Fatigue design of road bridges according to Eurocode standards can be based on damage accumulation, provided that information about loading spectrum is available. Characteristic values of fatigue resistance and load effects as well as partial safety factors are introduced in the design equation in order to achieve a target reliability level. Partial safety factors are used to obtain the fatigue design S-N curves from the characteristic S-N curves. New probabilistic framework proposed by the authors for calibration of fatigue partial safety factors allows for a redefinition of Eurocode-based design S-N curves. The new framework improves the confidence in the reliability level represented by the design S-N curves. In this paper a real case study showing the impact of re-calibration of design S-N curves on fatigue design of a road bridge is presented. Importance of choice of design S-N curves is assessed both in terms of allowable traffic volume and of bridge fatigue design life.

Wilhelm Ernst & Sohn2017

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EPFL2018