[en] The decommissioning of the nuclear plants is one of the main issues related to the use of nuclear power as energy source. Radiological and technical problems are related to this topics and consist, not only in how to remove and to manage the reactors spent fuel elements and the process liquids, but also in how to disassemble all the reactor's activated and/or contaminated components and systems. After a short review of the present law regulations and of the main technological activities related to the decommissioning, new decommissioning oriented reactor design criteria are proposed. (author)

[en] The realization of nuclear power plants based on fusion principles is expected to be, at the moment, very expensive. As a result the expected cost of electricity (COE) of fusion power plants is much higher than the COE of fission and fossil power plants. Thus it is necessary to study new solutions for fusion power plant designs to reduce the COE. An interesting solution for the first generation of fusion plants is to produce a surplus of tritium for commercial purposes. The present paper is concerned with the study of whether such a tritium surplus production can improve the plant economic balance, so that the COE is reduced, and to what extent. The result was that such a production allows a considerable reduction of COE and seems to be a good direction for development for the first generation of fusion power plants. To give an example, for a reference inertial confinement fusion (ICF) power plant the rise of the plant net tritium breeding ratio (TBR_n) from 1 to 1.2 would allow, in the conservative estimate of a tritium market price (C_T) of 5 M$/kg, a COE reduction of about 20%. In the estimate of a TBR_n rise from 1 to 1.3 and of a C_T value of 10 M$/kg, COE reduction could be more than 50%! In conclusion, the present paper points out the influence of TBR increase on COE reduction. Such a conclusion, which holds true for every fusion plant, is much more valid for ICF plants in which it is possible to reach higher TBR values and to use tritium extraction systems easily. Thus, considering the relevant economic advantages, a commercial tritium surplus production should not be disregarded for first generation fusion power plant designs, in particular for ICF plant designs

[en] The present work tries to solve the problem of realizing a parametric conceptual CAD model of a modular reactor for future inertial fusion power plants. The choice of a modular structure seems to be a good solution for efficiency and economic requirements. On the other hand, the realization of a parametric-variational CAD model is very useful to optimize nuclear and mechanical parameters and to permit the shift from the conceptual to the final model. First, geometric solutions for a modular reactor are analysed; the most interesting is that of a 20-face regular polyhedron (icosahedron). The subdivision of each face into six equal triangles consents to obtain a mono-modular reactor with 120 modules (called 'ICO120'). This solution should be easy, efficient and cheap. Secondly, the work proposes a conceptual CAD model of the ICO120 reactor in which special attention is put on the parametrization. Starting from such parametric model it will be possible to develop and optimize icosahedral reactors with different features, sizes and performances

[en] Monte Carlo modelling of an irradiation facility, for boron neutron capture therapy (BNCT) application, using a set of advanced type, accelerator based, ³H(d,n)⁴He (D-T) fusion neutron source device is presented. Some general issues concerning the design of a proper irradiation beam shaping assembly, based on very hard energy neutron source spectrum, are reviewed. The facility here proposed, which represents an interesting solution compared to the much more investigated Li or Be based accelerator driven neutron source could fulfil all the medical and safety requirements to be used by an hospital environment

[en] A nuclear fusion plant either of the magnetic or the inertial confinement type must be able to satisfy both the energetic and the economic balance conditions as any other energetic power system. The inertial confinement fusion (ICF) research is currently directed both towards the laser beam fusion and the accelerated particle beam fusion. The choice between the two drivers is not yet defined because the energetic parameters which characterize these systems may be perhaps improved, but in a barely predictable way. In this work, starting from the conceptual design of a possible inertial fusion nuclear power plant, the energy balance is evaluated, depending on efficiency values assumed by considering different plant sections. Then the thermonuclear energy gain in the fusion chamber versus the recycling energy fraction and the driver efficiency is evaluated

No abstract available

[en] The idea to couple the treatment planning system (TPS) to the information on the real boron distribution in the patient acquired by positron emission tomography (PET) is the main added value of the new methodology set-up at DIMNP (Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione) of University of Pisa, in collaboration with the JRC (Joint Research Centre) at Petten (NL). This methodology has been implemented in a new TPS, called Boron Distribution Treatment Planning System (BDTPS), which takes into account the actual boron distribution in the patient's organ, as opposed to other TPSs used in BNCT that assume an ideal uniform boron distribution. BDTPS is based on the Monte Carlo technique and has been experimentally validated comparing the computed main parameters (thermal neutron flux, boron dose, etc.) to those measured during the irradiation of an ad hoc designed phantom (HEterogeneous BOron phantoM, HEBOM). The results are also in good agreement with those obtained by the standard TPS SERA and by reference calculations carried out using an analytical model with the MCNP code. In this paper, the methodology followed for both the experimental and the computational validation of BDTPS is described

[en] Present standard treatment planning (TP) for glioblastoma multiforme (GBM - a kind of brain tumor), used in all boron neutron capture therapy (BNCT) trials, requires the construction (based on CT and/or MRI images) of a 3D model of the patient head, in which several regions, corresponding to different anatomical structures, are identified. The model is then employed by a computer code to simulate radiation transport in human tissues. The assumption is always made that considering a single value of boron concentration for each specific region will not lead to significant errors in dose computation. The concentration values are estimated 'indirectly', on the basis of previous experience and blood sample analysis. This paper describes an original approach, with the introduction of data on the in vivo boron distribution, acquired by a positron emission tomography (PET) scan after labeling the BPA (borono-phenylalanine) with the positron emitter ¹⁸F. The feasibility of this approach was first tested with good results using the code CARONTE. Now a complete TPS is under development. The main features of the first version of this code are described and the results of a preliminary study are presented. Significant differences in dose computation arise when the two different approaches ('standard' and 'PET-based') are applied to the TP of the same GBM case

[en] The experience acquired at the Instituto di Impianti Nucleari of the University of Pisa on computer code assessment in small LOCA is presented. From the observation of the smoothed behaviour of thermal hydraulic variables during a small break LOCA, a coarse nodalization was assumed to perform both pre-and post-test analysis in this field. This approach has been used in calculations related to International Standard Problems (ISP) and LOBI program tests: ISP 4 (SEMISCALE TEST S-02-06); ISP 9 (LOFT TEST L3-1); ISP 11 (LOFT TESTS L3-6/L8-1); and Shake-down test SD-SL-03 of LOBI program. The experience gained allows to draw some criteria about the use of RELAP 4 in small break LOCA calculations and to emphasize the need of improvements in this respect

[en] International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) Breeding Blanket (BB) will be different from the design defined by the released version named 'ITER-98' from the size as well as the neutron wall loading (NWL) point of view. In fact global dimensions and output power of ITER-FEAT have been approximately reduced by a factor of 3. The BB under study corresponds to the main solutions of ITER-98 project , that is: breeder in tube, consisting of pellets of a Lithium compound; neutron multiplier made of a pebble bed of Beryllium; cooling plates (using pressurized water as coolant) as basic cell boundaries; stainless steel as structural material. This paper deals with the neutronic as well as the thermal mechanical analyses of the envisaged BB in order to assess its performances in terms of Tritium production and thermal mechanical resistance. The code MCNP version 4B has been used for the neutronic analyses while the thermal-mechanical analyses have been performed by means of the FEM codes MARC and ANSYS. The preliminary analyses demonstrated that the proposed BB verifies the temperature constraints for the Breeder (temperature contained in the 'thermal window') as well as for the multiplier (Be temperature lower than 500 deg. C), but the configuration of the breeder and multiplier in the basic cell have to be modified in order to obtain a Tritium Breeder Ratio larger than 1