Search

A simplified Fusion-Fission Hybrid System (FFHS) deterministic calculation model has been developed in order to study the coupling be- tween a fusion machine and a subcritical fission system. Monte-Carlo (MC) codes are very flexible and they have to be considered as refer- ence calculation tools for complex systems like FFHS. On the other hand, deterministic codes can provide, in shorter calculation time respect to MC codes, parametric and sensitivity analysis, also by exploiting General Perturbation Theory (GPT) methodologies, which can usefully support the global FFHS theoretical analysis. This work describes the comparison between the results obtained for a simplified FFHS model by using a MC code (MCNP6.1, MC N-Particle) and a deterministic code (ERANOS). In particular, in this preliminary study, integral pa- rameters like keff, thermal power, neutron fluxes and some key reaction rate profiles, will be compared. If the comparison between the two mentioned codes will give encouraging results further, more refined, studies will be performed in order to consider deterministic codes as a powerful workhorse for the FFHS analysis, including time dependent analysis devoted to transient and reactivity monitoring issues.

Based on the heuristically generalized perturbation theory (HGPT) adapted to subcritical systems [1], a procedure for online operational monitoring of the subcriticality level of a hybrid fusion-fission system in which an “external” neutron source is generated in a magnetically confined tokamak-type plasma is described. This procedure [2], commonly referred to as power control-based subcriticality monitoring (PCSM), consists of compensating of slow and small movements of a specialized control rod in the nuclear fission zone, previously calibrated using a standard procedure, with equally slow and small alterations of the fusion neutron source. PCSM is verified by solving the multi-group transport equation for the direct flux, as well as for the importance function, associated with normalized fission power [3], and a technique is proposed for modifying the fusion neutron source strength based on plasma compression/expansion. Plasma and confining magnetic field adjustments, needed to implement the PCSM, are estimated with due account of the 0-D-plasma power balance.

Uncertainties estimation is essential for a sound assessment and a clear definition of safety margins, which in turn are crucial for licensing approval of new nuclear reactors by safety authorities. Moreover, uncertainties are also matter for target accuracy requirements (TARs), whose achievement could disclose new or enhanced opportunities. The concept of target accuracy requirements has been over the years one of the returning focuses of Massimo Salvatores, who just recently relaunched an update exercise within the OECD/NEA Working Party on International Nuclear Data Evaluation Co-operation (WPEC) aiming at providing new estimates having as reference the advanced systems meant for tomorrow. This work presents an evaluation of the current accuracies on the main core integral parameters of the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED), as a starting point to discuss which would be the target values that might introduce beneficial improvements in the design. Despite the use of one of the most recent nuclear data libraries (ENDF/B-VIII.0), and the broad margins set in the design of ALFRED according to its mission as demonstrator, room for improvement is found on the multiplication factor. Then, moving from the target accuracies on the integral parameters, an inverse uncertainty problem is solved to derive the corresponding target accuracies on nuclear data, meeting which would guarantee the achievement for the former ones. The inverse problem is solved introducing three different sets of cost parameters, reflecting possible challenges related to the execution of the differential experiments requested to achieve the aimed reductions of the accuracies. By this, the resulting target accuracies on nuclear data here presented are deemed useful to obtain information for the high-priority lists driving the selection of isotopes and reactions for future refinement experiments.

The EURATOM FP7 project FREYA was launched in 2011 to support the design and licensing of lead-cooled reactors, by choosing as reference the MYRRHA and ALFRED fast spectrum systems. During the five-year project, a number of critical experiments were conducted in the VENUS-F zero-power reactor with the main aim to validate the neutronic codes adopted for the core design. Five critical core configurations were assembled in VENUS-F: the start-up core, the mock-up dedicated to ALFRED and three layouts simulating the MYRRHA reactor conditions. The ERANOS code system, coupled with the JEFF-3.1 nuclear data library, was one of the tools adopted for the design and characterisation of the experiments. In this paper, the results obtained with the ERANOS deterministic code are compared with the experimental measurements of integral and local parameters, such as: the core reactivity and kinetic parameters, the lead void reactivity worth, the axial and radial distributions of fission rates for the nuclides of major interest, the spectral indexes of important actinides ( e.g. , U 238 , Pu 239 , Np 237 , Am 241 ) respect to U 235 . The comparison between measurements and simulations shows a general good agreement, with some discrepancies for few local parameters that are here discussed.

In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R&D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

After Fukushima events, the Italian nuclear program has been redefined leaving space only to activities related to Generation IV nuclear systems. Accordingly with this renewed national scenario, TAPIRO fast reactor facility is gaining a relatively major strategic role. A program is in fact being proposed to host in TAPIRO benchmark experimental activities related to the development of Lead fast reactor and Accelerator Driven Systems. A first step of this program would consist on the validation of neutronic codes, cross section data and reactor models to be adopted for its analysis. Along this line in this work the results of a simulation study has been made relevant to the measurements performed in the SCK-CEN/ENEA experimental campaign carried out in the 1980-1986 period. The calculations have been made using the Monte Carlo MCNPX 2.7.0 Code. In this article the main results are presented and discussed, with particular emphasis on the uncertainties, relevant both to nuclear data and the model layout. The results of this simulation study indicate in particular that TAPIRO's MCNPX model is adequate for the optimization of set-ups of perspective neutron irradiation experiments, this allowing cuts in costs and development time. © 2014 Elsevier B.V.