Carlo Fiorina

Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of- core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities.

ANALYSIS OF THE MSFR CORE NEUTRONICS ADOPTING DIFFERENT NEUTRON TRANSPORT MODELS

The Molten Salt Fast Reactor (MSFR) has recently been chosen as the reference circulating-fuel Molten Salt Reactor design in the framework of the Generation IV International Forum. Different from most of the Molten Salt Reactor designs developed and proposed in the past, the MSFR is featured by a fast neutron spectrum. On one hand, this choice leads to a simplified core (no graphite is present) and to improved characteristics in terms of breeding and/or transuranic burning. On the other hand, the removal of graphite leads to a significantly different behavior in terms of both core neutronics and dynamics. In view of the ongoing development of dedicated tools for the MSFR transient simulation, it is useful: 1) to accurately determine the core feedback coefficients, for lumped models and to get an insight into the reactor safety features; and 2) to analyze the degree of approximation related to the use of deterministic multi-group transport and diffusion approaches, in case of multi-dimensional models. The present paper compares the results obtained by means of the deterministic code ERANOS 2.2N with those obtained through the Monte Carlo code PSG2/SERPENT, for different fuel cycle strategies. In particular, the comparison is based on the capability to reproduce nominal reactivity, feedback coefficients, spectra and flux profiles. In the light of these results, the capability of a simple few-group diffusion model to deal with the MSFR neutronics is preliminarily assessed. Such model has been set up by means of the general-purpose finiteelement COMSOL Multiphysics software. It is of interest for a subsequent development of multi-physics models able to reproduce the peculiar MSFR transient behavior. NOMENCLATURE

A preliminary study of the MSFR dynamics

In the last years, increasing interest has been focused on an innovative concept of Molten Salt Reactor (MSR) characterized by a fast neutron spectrum that combines the favorable characteristics of MSRs adopting molten salt fluorides both as fuel and coolant with those ones of fast neutron reactors. As a matter of fact, the Molten Salt Fast Reactor (MSFR) has been recognized as a long term alternative to solid-fuelled fast neutron systems and has been identified as reference Gen-IV configuration. Although considerable studies have been carried out for the analysis of the graphite-moderated MSR dynamics, the adoption of a fast spectrum configuration without graphite in the core is expected to notably modify the dynamic behavior of the system, thus requiring further investigation.In this paper, a preliminary analysis of the MSFR dynamics is performed allowing for the evaluation of the impact of some safety parameters (e.g., feedback coefficients and delayed neutron fraction) on the system behavior for different fuel cycle strategies. For this purpose, a simplified non-linear one-dimensional model of the primary circuit has been developed and the dynamic response of the system has been investigated with reference to different significant transient initiators, namely: unprotected transient overpower, unprotected loss of flow, and unprotected loss of heat sink.These analyses are thought to give a basic understanding of the MSFR dynamics, as well as significant indications in terms of the system safety parameters.Copyright

Preliminary Analysis of the MSFR Fuel Cycle Using Modified-EQL3D Procedure

In the last few years, use of thorium in fast reactors has been gaining increasing attention due to its favorable features in terms of waste minimization and resource availability. In particular, use of thorium in a closed cycle leads to a limited build-up of transuranic isotopes, and to a lower radio-toxicity generation. Unfortunately, U233 breeding is always accompanied by the production of undesirable U232, which represents a serious radiological hazard to workers during fuel reprocessing and, in particular, fabrication. A possible way to partially overcome such difficulty would be to avoid fuel pin fabrication through the development of liquid-fuelled reactors, among which the most promising ones are the Molten Salt Reactors. The present paper investigates the fuel cycle performances of the reference GEN-IV Molten Salt Fast Reactor (MSFR) in terms of isotope evolution, radio-toxicity generation and safety parameters, considering different fuel cycle strategies. Main results are also compared with those obtained for the GEN-IV ELSY Lead Fast Reactor. Calculations are performed by means of the EQL3D procedure developed at the Paul Scherrer Institut (Switzerland) for the analysis of equilibrium fuel cycles in fast reactors. In order to take into account the peculiarities of the MSFR design, a modified version of the procedure is proposed and adopted to model on-line reprocessing and the presence of blankets.Copyright

Thermo-Hydrodynamics of Internally Heated Molten Salts for Innovative Nuclear Reactors

The problem of heat transfer in pipe flow has been extensively investigated in the past. Many different models have been proposed and adopted to predict the velocity profile, the eddy diffusivity, the temperature distributions, the friction factor and the heat transfer coefficient (Kays et al., 2004; Schlichting & Gersten, 2000). However, the majority of such studies give a description of the problem for non-internally heated fluids. Models regarding fluids with internal heat generation have been developed more than 50 years ago (Kinney & Sparrow, 1966; Poppendiek, 1954; Siegel & Sparrow, 1959), giving in most cases a partial treatment of the problem in terms of boundary conditions and heat source distribution, and relying on a turbulent flow treatment that does not seem fully satisfactory in the light of more recent investigations (Churchill, 1997; 2002; Kays, 1994; Zagarola & Smits, 1997). Internally heated fluids are of great interest in the current development of Molten Salt Reactors (MSR) (LeBlanc, 2010), included as one of the six innovative nuclear reactors selected by the Generation IV International Forum (GIF-IV, 2002) for a more sustainable version of nuclear power. MSRs are circulating fuel reactors (Nicolino et al., 2008), which employ a non-classical (fluid-type) fuel constituted by a molten halide (fluoride or chloride) salt mixture playing the distinctive role of both heat source and coolant. By adopting classical correlations for the Nusselt number (e.g., Dittus-Boelter), the heat transfer coefficient of the MSR fuel can be overestimated by a non-negligible amount (Di Marcello et al., 2008). In the case of thermal-neutron-spectrum (graphite-moderated) MSRs (LeBlanc, 2010), this has significant consequences on the core temperature predictions and on the reactor dynamic behaviour (Luzzi et al., 2011). Such influence of the heat source within the fluid cannot be neglected, and thus required proper investigation. The present chapter deals with this critical issue, first summarizing the main modelling efforts carried out by the authors (Di Marcello et al., 2010; Luzzi et al., 2010) to investigate the thermo-hydrodynamics of internally heated fluids, and then focusing on the heat transfer coefficient prediction that is relevant for analysing the molten salt behaviour encountered in MSRs.

A Time-Dependent Solver for Coupled Neutron-Transport Thermal-Mechanics Calculations and Simulation of a Godiva Prompt-Critical Burst

The present paper describes a first step taken at the Paul Scherrer Institut in the development of a new multi-physics platform for reactor analysis. Such platform is based on the finite-volume software OpenFOAM and aims at a tightly coupled description of neutron transport, thermal mechanics and fluid dynamics. For this purpose, a steady-state 3-D discrete ordinates/thermal-mechanics solver was first developed in collaboration with the Politecnico di Milano. The present work briefly discusses such solver and its preliminary validation, which will be described in detail in parallel publications. It then focuses on its extension to time-dependent simulations. The solver is first tested by simulating different step-wise reactivity insertions in a critical configuration constituted by an infinite slab of highly enriched uranium. Subsequently, a super-prompt-critical power burst in the Godiva reactor has been simulated. Godiva was a spherical assembly of highly enriched uranium built and operated at the Los Alamos National Laboratory (US) during the Fifties. A prompt-critical transient in such system configures as a quick power excursion (up to ∼10 GW), which causes a temperature rise, and a subsequent reactivity reduction via expansion of the sphere. The overall transient lasts for few fractions of a millisecond. Results obtained with the newly developed model have been compared to experimental results, showing a relatively good agreement.Copyright

Molten Salt Reactor With Simplified Fuel Recycling and Delayed Carrier Salt Cleaning

The Molten Salt Reactor (MSR) represents an old concept, but its properties are qualifying it for the advanced utilization: inherent safety, excellent neutron economy, possibility of continuous or batch reprocessing without fuel fabrication. The aim of this paper is to characterize the MSR unique fuel cycle advantages in different neutron spectra using the results of ERANOS-based EQL3D and ECCO-MATLAB based EQL0D procedures. It also focuses on the low production of higher actinides in the Th-U cycle and based on the results, it proposes a simplified in situ recycling of the fuel and the delayed ex situ carrier salt cleaning or direct disposal by vitrification.Copyright

Paul Scherrer Institute’s Studies on Advanced Molten Salt Reactor Fuel Cycle Options

Nuclear reactors operated with liquid fuel may have several remarkable advantages and features. The most developed reactor system in this category is the molten salt reactor (MSR). It represents an old concept, but its properties qualify it for advanced utilization; these include inherent safety, excellent neutron economy, continuous or batch reprocessing possibilities without fuel fabrication. The aim of this study is to characterize the MSR physics, highlighting its unique fuel cycle advantages in several different neutron spectra by using the ERANOS-based EQL3D procedure and ECCO-MATLAB- or Serpent-MATLAB-based EQL0D procedures.

The Proto-Earth Georeactor: A Thorium Reactor?

Georeactors have been suspected to occur in natural uranium deposits [1]. They have, for example, been found in the Earth’s crust in Oklo, Gabon [2]. The feasibility of one or more massive nuclear fission reactors deeper inside the Earth was also proposed.

A Procedure for the Pre-Conceptual Design of Fast Reactors and Application to a Gas-Cooled Sub-Critical Transmuter

A C++ procedure has been developed for the design and optimization of Fast Reactor (FR) cores. It couples the ERANOS based EQL3D procedure developed at PSI for FR equilibrium fuel cycle analysis with a dedicated MATLAB script that evaluates the thermal-hydraulic characteristics of the reactor core. It is conceived to investigate reactors with both standard pins and annular pins. The procedure accepts as input the physical properties of the system, as well as a set of target core parameters presently consisting of core power, maximum fuel burnup, multiplication factor, inner pin diameter (for annular pins) or maximum pressure losses (for standard pins), and core height. It gives as a result a core design fulfilling these design objectives and meeting the constraints on maximum fuel and clad temperatures. In case of annular pins, it also equalizes the temperature rise inside and outside of the core average pin. The procedure considers the possibility of two-zone cores and adjusts the fuel composition in the two zones to achieve an optimal radial power distribution. Finally, it can evaluate safety parameters and fuel cycle characteristics both at beginning-of-life and at equilibrium. As a test case, the procedure has been used for the pre-conceptual design of a sub-critical Gas Fast Reactor core employing inert-matrix sphere-pac fuel and annular pins with SiC cladding.Copyright