Comparison Among Monte Carlo Based Burnup Codes Applied to the GFR Demonstrator ALLEGRO

This paper aims to compare three Monte Carlo (MC) burnup based codes, i.e. MCNP6, Monteburns and Serpent on a future prototype reactor, named ALLEGRO, based on Gas cooled Fast Reactor (GFR) technology. GFR reactors are one of the proposed Generation-IV fast reactors; ALLEGRO facility is scheduled to be built in Europe as a GFR demonstrator, so its deepened simulation can help in its future development. The present study follows other researches already performed and aims to exhibit the different approaches in burnup calculations applied to a gas cooled fast reactors, i.e. this paper would like to show and to compare some results concerning nuclear parameters as keff and flux spectra, as well as the mass inventories versus burnup for some nuclides evaluated with different Monte Carlo codes. From obtained results, it seems to exist some differences in evaluation of nuclear parameters, mainly in effective multiplication factor and in mass inventories. The remaining differences are mainly related to calculation time: indeed between the fastest, that is SERPENT, and the slowest, that are MCNP6 and MONTEBURNS, the differences are about one order of magnitude. As far as precision is concerned, it was considered the standard only for effective multiplication factor and it seems that all codes are in good agreement

Partial Redesign of an Accelerator Driven System Target for Optimizing the Heat Removal and Minimizing the Pressure Drops

Accelerator Driven Systems (ADS) seem to be a good solution for safe nuclear waste transmutation. One of the most important challenges for this kind of machine is the target design, particularly for what concerning the target cooling system. In order to optimize this component a CFD-based approach has been chosen. After the definition of a reference design (Be target cooled by He), some parameters have been varied in order to optimize the thermal-fluid-dynamic features. The final optimized target design has an increased security margin for what regarding Be melting and reduces the maximum coolant velocity (and consequently even more the pressure drops)

Assessment of LWR-HTR-GCFR Integrated Cycle

Preliminary analyses already performed showed that innovative GCRs, both thermal and fast, are very promising candidate to reach the Gen-IV sustainability goal. The integrated LWR-HTR-GCFR basically aims at closing the current nuclear fuel cycle: in principle, thanks to the unique characteristics of Helium coolant reactors, LWR SNF along with DU become valuable material to produce energy. Additionally, burning HMs of LWR SNF means not only a drastic reduction in the demand but also a remarkable decrease in the long-term radiotoxic component of nuclear waste to be geologically stored. This paper focuses on the analyses of the LWR-HTR-GCFR cycle performed by the University of Pisa in the frame of the EU PUMA project (6th FP). Starting from a brief outline of the main characteristics of HTR and GCFR concepts and of the advantages of linking LWR, HTR and GCFR in a symbiotic way, this paper shows the integrated cycle involving a typical LWR (1000 ), a PBMR (400 ) and a GCFR-"E" (2400 ). Additionally, a brief overview of the main technological constraints concerning (Pu+MA)-based advanced fuels is given, in order to explain and justify the choices made in the framework of the considered cycle. Thereafter, calculations performed and results obtained are described

Assessment of a 2D CFD model for a single phase natural circulation loop

The use of passive safety systems are more and more diffused in many technological fields. Natural circulation is probably one of the main phenomenon applied in this kind of systems: indeed, as known, by means of gravity and buoyancy forces, the fluids can circulate without any external power sources. In this paper a preliminary analysis (also by comparisons between experimental tests and numerical simulations) of a natural circulation based loop (namely a natural circulation based facility installed at University of Genova) is presented. Starting from some experimental results, the data deriving from CFD loop simulations (both in steady and in unsteady conditions) are used for a first preliminary validation, mainly in order to have a computational tool reliable and able to computationally simulate motion inversions related phenomena. The physical inversions phenomena are very well reproduced also by the a simplified numerical 2D model of the loop, and the physical considerations related to the temperature and velocity fluctuations during the transient simulations, are in agreement with the well-known observations formulated by Welander on the basis of a simple point source analysis scheme

Assessment of Accident-Tolerant Fuel with FeCrAl Cladding Behavior Using MELCOR 2.2 Based on the Results of the QUENCH-19 Experiment

To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental conditions must be assessed. While the dependences of the behavior of single physical parameters can be investigated in single- or separate-effect experiments, and more complex phenomena can be investigated using integral-effect tests, the behavior of an entire system as complex as a nuclear power plant core must be investigated using computer code modeling. One of the most commonly used computer codes for the assessment of severe accidents is MELCOR 2.2. In version 18019, the authors enabled the modeling of the behavior of the nuclear fuel with FeCrAl cladding (namely, alloy B136Y3) for the first time, using the GOX model. The ability of this model to reasonably accurately predict the behavior of FeCrAl cladding in accident conditions with quenching was verified in this work by modeling the QUENCH-19 experiment carried out in the Karlsruhe Institute of Technology on the QUENCH experimental device and by subsequent comparison of the MELCOR calculation results with the experiment. This article proves that the GOX model can be used to evaluate the behavior of FeCrAl cladding and that the results can be considered conservative

GCFR Coupled Neutronic and Thermal-Fluid-Dynamics Analyses for a Core Containing Minor Actinides

Problems about future energy availability, climate changes, and air quality seem to play an important role in energy production. While current reactor generations provide a guaranteed and economical energy production, new nuclear power plant generation would increase the ways and purposes in which nuclear energy can be used. To explore these new technological applications, several governments, industries, and research communities decided to contribute to the next reactor generation, called "Generation IV." Among the six Gen-IV reactor designs, the Gas Cooled Fast Reactor (GCFR) uses a direct-cycle helium turbine for electricity generation and for a -free thermochemical production of hydrogen. Additionally, the use of a fast spectrum allows actinides transmutation, minimizing the production of long-lived radioactive waste in an integrated fuel cycle. This paper presents an analysis of GCFR fuel cycle optimization and of a thermal-hydraulic of a GCFR-prototype under steady-state and transient conditions. The fuel cycle optimization was performed to assess the capability of the GCFR to transmute MAs, while the thermal-hydraulic analysis was performed to investigate the reactor and the safety systems behavior during a LOFA. Preliminary results show that limited quantities of MA are not affecting significantly the thermal-fluid-dynamics behavior of a GCFR core

Development of a Reduced Order Model for Fuel Burnup Analysis

Fuel burnup analysis requires a high computational cost for full core calculations, due to the amount of the information processed for the total reaction rates in many burnup regions. Indeed, they reach the order of millions or more by a subdivision into radial and axial regions in a pin-by-pin description. In addition, if multi-physics approaches are adopted to consider the effects of temperature and density fields on fuel consumption, the computational load grows further. In this way, the need to find a compromise between computational cost and solution accuracy is a crucial issue in burnup analysis. To overcome this problem, the present work aims to develop a methodological approach to implement a Reduced Order Model (ROM), based on Proper Orthogonal Decomposition (POD), in fuel burnup analysis. We verify the approach on 4 years of burnup of the TMI-1 unit cell benchmark, by reconstructing fuel materials and burnup matrices over time with different levels of approximation. The results show that the modeling approach is able to reproduce reactivity and nuclide densities over time, where the accuracy increases with the number of basis functions employed

A useful observable for estimating keff in fast subcritical systems

The neutron multiplication factor k-eff is a key quantity to characterize subcritical neutron multiplying devices and for understanting their physical behaviour, being related to the fundamental eigenvalue of Boltzmann transport equation. Both the maximum available power-and all quantities related to it, like, e.g. the effectiveness in burning nuclear wastes-as well as reactor kinetics and dynamics depend on k-eff. Nevertheless , k e f f is not directly measurable and its determination results from the solution of an inverse problem: minimizing model dependence of the solution for k-eff then becomes a critical issue, relevant both for practical and theoretical reasons

Development of CFD models and pre-test calculations for thermal-hydraulics and freezing experiments on Lead coolant

Heavy Liquid Metals (HLM) are objects of interest in the nuclear research sector because of their optimal thermal and neutronic properties; the development and the validation of models allowing to predict their behaviour are fundamental for the future development of the Generation IV energy systems. An experimental facility named SESAME-stand, is planned to be operated at Research Centre Rez (RC-Rez) under the framework of SESAME project. The aim of the facility is to study the solidification of Lead under GEN-IV Heavy Liquid Metal pool type nuclear reactors relevant conditions and to provide a database for the benchmarking and validation of numerical models. Corresponding CFD models are developed using commercial software and are used for the pre-test assessment and to support the experimental work. The aim of this paper is to describe the CFD models, explain how they are tested and used in order to define a valuable experimental matrix that will be needed in order to run the facility itself. First of all, the facility is introduced together with the range of foreseen investigations. The numerical models are then presented. Emphasis is given to the geometrical and physical assumptions. Different approaches of modelling are compared and discussed. Results from the pre-test simulations are illustrated. Encountered challenges and their relevance with regard to the experimental matrix and setup are commented