1,721,179 research outputs found
On the relevance of the scattering anisotropy for a low order transport method
No full textThe present paper is concerned with the extension of the treatment of the scattering anisotropy developed in the AN-SPN (or simply AN) method up to the approximation of the second order. To assess such extension, two 3D examples are considered: the first one is an axial piece of a 17×17 PWR MOX assembly containing the heads of control rods; while the second example is related to a 3D reactor core containing UO2 and MOX homogenized assemblies. It is shown that the passage from the isotropic to the linearly anisotropic approximation of the differential scattering cross section turns out into a remarkable improvement of the results, as regards the multiplication factor as well as the flux distribution. For the kind of problems like those here considered, less important is, on the contrary, the extension to the second order of the anisotropy, although such an upgrade may be included in the calculation with only an almost negligible effort, owing to the irrelevant impact on the computing time
Solution of 3D linearly anisotropic scattering, fixed-source multigroup xyz reactor problems by the AN Boundary Element – Response Matrix method
Full text linkThis paper aims at extending the work performed by means of the AN Boundary Element – Response Matrix method in a previous paper, devoted to criticality problems, to non-homogeneous problems (i.e. the problems that involve a fixed external source, as in case of the accelerator driven systems). As in the preceding paper, the interest is given to the 3D, xyz multigroup systems in which linearly anisotropic scattering is allowed. Standard interface conditions have been adopted and no external intervention in order to improve the numerical results, such as some kind of discontinuity factors, has been introduced. The latter choice is motivated by the need of establishing a clean test of the performances of the AN method in itself. Results of 2D and 3D multigroup fixed source problems obtained with the method described in this paper are compared with those obtained by well assessed reference codes such as the MCNP Monte Carlo code and the PARTISN discrete ordinates code. Even in the absence of discontinuity factors and despite of the intrinsically approximate character of the AN method, the accuracy of the numerical results is more than satisfactory. Finally the computational time is in most cases much shorter than that required by the chosen reference code
Neutronic calculations for preliminary core design of SCW-SMR
Get PDFSerpent 2 particle transport code is used to develop the pre-conceptual neutronic design of the Supercritical Water Cooled SMR. After initial criticality and burnup calculations, the starting core design of (Schulenberg and Otic, 2021) is improved using predetermined criteria, such as burnup cycle length and power distribution, while also considering operational safety. In order to achieve higher reserve reactivity, several modifications are considered, including the introduction of alternative structural materials and fuel assembly wall type, moderation improvement by adjustment of moderator temperature and fuel assembly gap width, and selection of a suitable enrichment map. As a result of the introduced modifications, the burnup cycle length is increased to 26 months and an acceptable core power distribution is achieved. The improved core design can be used for further investigations, such as coupled calculations using neutronic and thermal–hydraulic codes and examinations targeting reactivity control during burnup
Analysis of an Autogenous Propellant Pressurization System for Nuclear Thermal Rocket
No full textSince the NERVA project, hydrogen has been the most proposed propellant for nuclear propulsion systems thanks to its overwhelming performance. However, its low density and the need to store it in cryogenic conditions, combined with the long duration of the foreseen interplanetary missions, led to the design of propulsion systems with enormous envelopes and the introduction of complex systems for propellant management. On the other hand, ammonia presents a clear loss of performance compared to hydrogen due to its higher molecular weight, limiting the spectrum of missions made accessible by the use of this propellant in a nuclear thermal propulsion system. However, the greater density of ammonia and its high vapor pressure at around 373 K allow for much more compact and less complex propellant management system configurations. This work proposes a new propellant management configuration for an ammonia-fueled nuclear thermal propulsion system for a class of missions involving cargo displacements between LEO and the lunar orbits of interest for future space programs. The suggested configuration maximizes the advantage deriving from the self-pressurization of ammonia by exploiting the thermal power lost by the nuclear reactor towards the vacuum space due to the escaping particles. In this layout a tank containing ammonia in saturated conditions is placed near the nuclear reactor and receives an input thermal power proportional to the dose of gamma rays and neutrons absorbed by the ammonia and the tank walls. This thermal power accelerates the vaporization process of the saturated ammonia, thus increasing the pressure in the tank. A pressure regulator valve exploits this overpressure to pressurize the ammonia propellant contained in a run tank to the level required by the mission by connecting the two ammonia volumes. The pressure achieved inside the run tank pushes the propellant with an adequate mass flow rate inside the nuclear reactor. The Homogeneous Equilibrium Model and the Two-Temperature Model describe the coupled dynamics of these two tanks, while a study of neutron and photon transport performed with the Monte Carlo code OpenMC provides the thermal power input received from the pressurizing ammonia tank. The developed analysis shows how this propellant management system can provide a constant mass flow to the nuclear reactor without using a turbopump assembly. This capacity depends on the relative distance between the tank and the nuclear fission reactor. Another advantage of the proposed concept concerns the reduction of the mass of the radiation shield: it does not have to protect the tanks from nuclear radiation, as happens for hydrogen-based projects, and therefore, its dimensions are governed exclusively by the radial size of the payload. For some missions, this equates to having a smaller shield size and, consequently, a lower mass
Going Beyond Counting First Authors in Author Co-citation Analysis
Full text linkThe present study examines one of the fundamental aspects of author co-citation analysis (ACA) - the way co-citation
counts are defined. Co-citation counting provides the data on which all subsequent statistical analyses and mappings
are based, and we compare ACA results based on two different types of co-citation counting - the traditional type that
only counts the first one among a cited work's authors on the one hand and a non-traditional type that takes into
account the first 5 authors of a cited work on the other hand. Results indicate that the picture produced through this non-traditional author co-citation counting contains more coherent author groups and is therefore considerably clearer. However, this picture represents fewer specialties in the research field being studied than that produced through the traditional first-author co-citation counting when the same number of top-ranked authors is selected and analyzed. Reasons for these effects are discussed
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