1,721,079 research outputs found
Monte Carlo analysis of dosimetric issues in space exploration
No full textThe Radiation protection is of paramount importance in the planning of human exploration activities in space. The related risks must be considered with respect to two aspects: devising a proper shielding and providing answers to the requirement of an effective dosimetry evaluation in astronaut’s activities. Both aspects have been considered using the Monte Carlo (MC) code MCNP 6.2 as the reference tool. As case study an application devised for the National Aeronautics and Space Administration (NASA) Artemis program has been chosen. The project aims to establish a sustainable human presence on the Moon, envisioning the realization of an outpost that will serve as a steppingstone for space exploration endeavors. A Class III shelter, in situ resource utilization (ISRU) built habitat for the Moon, has been designed through computational methods and topology optimization techniques, and analyzed in terms of radiation shielding performances and the strictly related structural behavior. The outpost must be able to withstand temperature variations, micrometeorite impacts, and the absence of a substantial atmosphere. Any solution studied to respect the constraints must devise robust and innovative materials and techniques to create habitats that have as goal the shielding from the Galactic Cosmic Rays (GCR) and from the solar flares to provide a safe and habitable environment at the time scales scheduled for the missions. Moreover, the outpost design must incorporate strategies for extracting and utilizing local re- sources. Overcoming such challenges will pave the way for the establishment of a sustainable human presence on the Moon and serve as a crucial leap for future space exploration missions
SCALE 6.1.3 Effective Heavy Reflector Cross Sections Sensitivity Analysis for a PWR GENIII Assembly/Reflector System
No full textThis paper presents the results of a sensitivity analysis on code and operational parameters for the evaluation of the effective cross sections of a central assembly/heavy reflector GENIII PWR system. The methodology used for the calculation is based on the T- NEWT control module of SCALE 6.1.3 package. In detail, the sensitivity analysis on the heavy reflector cross sections has been performed on the following set of code parameters: reflector zone computational meshes, Sn and Pm transport parameter, cross sections libraries and convergence parameters (eigenvalue and eigenfunction). It also has been carried out a series of calculations on a set of operational parameters such as: boron concentration (0 and 1300 ppm), operational condition (HZP and HFP) and reflector temperature (Tmax and Tmod)
Apparecchiatura per la produzione endogena di radioisotopi, particolarmente per diagnostica tomografica ad emissione di positroni
No full textApparecchiatura per la produzione endogena di radioisotopi, particolarmente per diagnostica tomografica ad emissione di positoni, caratterizzata dal fatto di comprendere:
- una camera da vuoto, la cui superficie interna è trattata per resistere all'implantazione ionica,
- una coppia di elettrodi posti all'interno di detta camera da vuoto,
- un banco capacitivo,
- mezzi di connessione di detto banco capacitivo a detta coppia di elettrodi per la generazione tra questi di una scarica elettrica, che provoca la formazione di plasma e crea le condizioni per lo sviluppo di reazioni nucleari che generano radioisotopi,
- un'induttanza complessiva del circuito elettrico equivalente di detta apparecchiatura non superiore a 50nH,
- mezzi comunicanti con detta camera da vuoto per la creazione di un vuoto non superiore a 10-6 torr,
- mezzi comunicanti con detta camera da vuoto per l'immissione in essa, dopo la creazione del vuoto, di almeno un gas di reazione ad una pressione in grado di garantire la formazione di plasma durante la fase di scarica ed il raggiungimento di condizioni di confinamento di detto plasma dell'ordine di 1015 keV. sec/cm3, e
- mezzi comunicanti con detta camera da vuoto per l'estrazione di gas
SCALE 6.1.3 Evaluation of the Heavy Reflector Effective Cross Sections for a GEN III PWR System and a Serpent 2.1.23 Model Comparison
No full textThis paper reports some results obtained using the methodology based on the T-NEWT control module of SCALE 6.1.3 modular code system for the calculation of Assembly Discontinuity Factors (ADFs) and homogenized two group condensed effective heavy reflector cross sections in the central and peripheral zones of a GEN III PWR core. A preliminary study has been carried out to verify the different quantitative neutronic behaviour between a heavy and a standard reflector. Afterwards, it analyzes the effect of different geometric and material assembly/reflectors models – homogeneous, homogenized slabs and heterogeneous – on the effective cross sections numerical results. The impact of the use of 2 or 8.5 assemblies coupled with the central reflector zone on the effective cross sections values is also investigated. The results of heavy central reflector effective cross sections data were then compared with those obtained on the peripheral region of a GEN III core. Finally, a cell data calculation has also been carried out on a 2x2 color-set to investigate a 2-D effect, in the reflector angle region, on the effective cross section results. The outcomes of the T-NEWT reference case are at last compared with those obtained with the Serpent 2.1.23 code
Scale 6.1 Evaluation of the Effective Cross Sections of a LWR Assembly-Reflector Model with Application to the NEA TMI-1 PWR Benchmark
Full text linkThis paper reports some results obtained using the methodology based on the T-NEWT control module of SCALE 6.1 for the calculation of the effective, two-group cross sections for the 2-D assembly-reflector geometry of TMI-1 PWR given in NEA/NSC/DOC (2013)7 benchmark [1]. The calculation of the effective two-group cross sections of the reflector zone is performed both for a homogenized geometry and for the exact 2-D one. The effect on the results of a variation of the boron concentration in the moderator zones is also investigated. The outcomes of the reference case are then compared with those obtained from a calculation performed with the Monte Carlo code SERPENT and those (MCNP5, DRAGON) presented in the available literature for the same benchmark problem. The differences in the numerical values obtained from the various codes are also discussed
Pressurized Water Small Modular Reactor (SMR), Design Basis Accident Analysis using the ASTEC code
Full text linkAccording to classifications adopted by the IAEA, small reactors are characterized by an equivalent electric output of less than 300 MW while medium-sized reactors by an equivalent electric power between 300 and 700 MW. Pressurized water small and medium sized reactors (SMR) generally adopt an integral layout of the primary circuit with in-vessel location of steam generators and control rod drives; one compact modular loop-type design features reduced length piping, an integral reactor cooling system accommodating all main and auxiliary systems within a leak-tight pressure boundary, and leak restriction devices.
In this paper, a description is given of the development of the modelling and noding of the primary loop, secondary loop passive core cooling system and containment for a SMR, based on the available data of the SPES3-IRIS integral test facility.
SPES3-IRIS is under construction at SIET laboratories in Piacenza (Italy), simulating with 1:100 volume scale and 1:1 height scale, the primary, secondary, containment and safety systems typical of the IRIS small modular reactor.
Three ASTEC code modules were adopted: the ICARE module to predict the in-vessel phenomena, the CESAR module to compute two-phase thermal–hydraulics in the Reactor Cooling System (RCS) and for the control and safety systems, and the CPA module to evaluate thermal–hydraulic and aerosol behaviour in the reactor containment.
The SMRs as well as the advanced nuclear water-cooled reactors rely on containment behaviour to achieve some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts. Thus, to simulate correctly the main phenomena involved during an accident scenario, the coupling between primary circuit and containment has to be reproduced accurately. Furthermore, given that the containment plays a fundamental role during every accident scenario, it has to be taken into account just as a real safety system.
The worst design basis event for the SMR was analysed, and the calculated results were compared with those obtained by the University of Zagreb in collaboration with Westinghouse using the coupled codes RELAP-GOTHIC. The aim of this work is to evaluate the applicability of ASTEC coupled modules in the safety analyses of the new reactor systems with strong interaction between primary system and containment
Shape, Structure and Material Compliance with Radiation Protection Requirements for Extraplanetary Modules
No full textThis research aims to explore a design solution for an innovative extraplanetary module that combines architectural design, structure and radiation protection for sustaining human life on Mars.
The WATER (Water shielded Architectural Tree for Extraplanetary Resiliency) module is designed in order to increment the use of local resources (In Situ Resources Utilization) and robotic fabrication techniques for remote construction before human arrival on Mars.
The key element of the design is the water that can be extracted from the substrate of the Martian regolith. Water plays an essential role in both in supporting life and protecting humans inside the habitat. Because of the reduced gravity and the fine atmosphere, the major load that a structure has to withstand on Mars is the internal pressurization. To balance that load and have a more efficient foundation system, the structure needs to be covered by a thick layer of water that is also extremely important for shielding against the harmful cosmic radiation.
In fact, it is well known that a major threat to extraplanetary exploration is given by high energy cosmic particles and gamma fluxes. This work deals with the radiation protection constraints that should be considered for the WATER module, designed as an optimized possible long term habitat for Mars. The main materials considered for the module are the Martian regolith and, with respect to radiation shielding, the water that will be driven to fill the layer between the external and internal surfaces that will sustain the exposed external structures.
The simulations, carried out with a standard Monte Carlo code like MCNPX and MCNP6, that is able to directly analyze the mesh geometries coming from the WATER module structural Finite Element model, define the optimal conditions in terms of shielding thickness and layer’s material composition.
As output of the analysis, expositions and doses, that the inhabitants of these future architecture should bear, have been obtained.
The final shielding configuration is integrated in the Finite Element model of the project for the structural analysis. The results prove that the water content, subjected to the Martian gravity, helps reducing the tensile stresses inside the structure due to the internal pressurization
Dismantling of the graphite pile of Latina NPP: Characterization and handling/removal equipment for single brick or multi-bricks
Full text linkThis work describes the issues related the dismantling of graphite piles of the 1st generation gas cooled reactor of Latina NPP (Italy). The retrieval of the graphite is a strategic matter for the decommissioning of this type of plant: in this study were described and analysed the current approaches used to access the core and to perform the remote and dry extraction of graphite bricks from the top. Based on these data, the removal of the graphite of Latina NPP will be planned; the extraction of the graphite will be carried out layer by layer by means of a dedicated remote controlled handling systems. This equipment will be duly designed according to the nuclear, physical and mechanical constraints of the graphite piles in core. In doing that the issues regarding the irradiated graphite have been also analysed by FEM code, especially those related to the core geometry and to the proposed technique of hooking the graphite bricks by a ‘gripper’ tool inside the axial channel. Data on fresh nuclear grade and irradiated graphite, used for the numerical simulations, were obtained by means of experimental tests, which were carried out on samples extracted from the reactor, and from theoretical models. The results obtained could support the final design of proper lifting and gripper tools and handling equipment, for single brick or multi-bricks, and to implement waste management strategy for the graphite
Exploratory Studies of Small Modular Reactors Using the ASTEC Code
Full text linkNuclear safety has been one of the major issues studied since the inception of the nuclear industry. Establishing and maintaining core cooling and ensuring containment integrity are two main goals that nuclear safety must guarantee. Improvement in these safety systems has generally involved the development of suitable Passive Containment Cooling Systems (PCCSs). This kind of safety approach poses significant issues for computational and analysis methods since the vessel and containment are strongly coupled and the system response is based on the interaction between the two. This is the case of Small Modular Reactors (SMRs), which adopt a completely passive safety approach, and the integral design eliminates the large coolant loop piping, which in turn eliminates large Loss-Of-Coolant-Accidents (LOCAs) as well as the individual component pressure vessels and supports. For these reasons, no severe accident analyses have yet been conducted on this type of plant. Nevertheless, it is useful to investigate the possible consequences of a multiple failure scenario in these advanced systems. In order to perform these analyses, starting from the available data of the SPES3 experimental facility, a SMR model was developed for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. The facility based on the IRIS reactor (International Reactor Innovative and Secure) design reproduces the primary, secondary and containment systems with a 1:100 volume scale, full elevation and prototypic fluid and thermal hydraulic conditions. The IRIS reactor is a SMR developed by an international consortium led by Westinghouse/BNFL, which includes universities, national laboratories, commercial companies and utilities. The Design Basis Accident (DBA) Direct Vessel Injection (DVI) line double-ended guillotine break was the reference transient that allowed matching the trend of the main physical parameters predicted by the ASTEC code model with those computed by the well-established best estimate coupled codes RELAP-GOTHIC. In this paper, a multiple system failure scenario was reproduced, to investigate and evaluate the in-vessel phase phenomena and the effectiveness of the passive mitigation measures. The results of the calculations confirmed the good performance of the IRIS system during the DBA accident, and showed for the first time how the ASTEC code can reproduce well the behaviour of this non- prototypic system
Monte Carlo benchmark of the experimental evaluation of the activation processes in an electron linear accelerator for radiotherapy applications
Full text linkSeveral kinds of isotopes are generated during radiotherapy treatments with high-energy electron sources due to the onset of many nuclear reactions. These isotopes are often unstable, can appear both in the device and in the treatment chamber materials and, as a consequence of the decay process, involving also gamma-ray emissions, some additional dose is given to the patient and to the radiotherapy unit staff. These effects have been experimentally monitored with a LaBr detector for gamma spectrometry. Then the measurement setup and data have been benchmarked through Monte Carlo (MC) simulations, with the MCNPX code, aiming to evaluate all kinds of activation, due to both photons and photoneutrons. All the MC activation estimates have been parameterized with respect to the 187W produced in the primary collimator of the accelerator. The simulation results obtained with MCNPX have shown a good agreement with the experimental measurements. The results suggest a possible general approach to perform the activation analysis by coupling the experimental spectrometric measurements with MC calculations to properly identify photopeaks and source components
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