Konstantin Mikityuk

FAST Code System: Review of Recent Developments and Near-Future Plans

The FAST code system is currently being developed and used at the Paul Scherrer Institut for static and transient analysis of the main Generation 4 fast-spectrum reactor concepts: sodium-, helium-, and gas-cooled fast reactors. The code system includes the ERANOS code system for static neutronics calculations, as well as coupled TRACE/PARCS/FRED for neutron kinetics, thermal hydraulic, and fuel transient analysis. The paper presents the status of the recent developments in neutronics (new 3D procedure for equilibrium cycle simulation and new transient cross section generation procedure), in thermal hydraulics and chemistry (equations-of-state for new coolants, two-phase flow models for sodium, and new model for oxide layer buildup in heavy-metal flow), and in fuel behavior (new model for the dispersed gas-cooled fast reactor fuel). Near-future plans for the further development of FAST are outlined. 2010 by ASME.

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

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

Development of a drift-flux model for heavy liquid metal/gas flow

The gas lift pump concept based on the bubbling of an inert gas into the primary reactor coolant to enhance natural circulation is currently considered in a number of PbBi-cooled reactor concepts. Thus, the analysis of available void fraction data and the development of two-phase heavy liquid metal/gas flow calculational models have become an important issue in the study of advanced nuclear reactor systems. In the absence of the detailed two-phase flow information needed to develop a flow regime map and the associated interfacial relations, drift-flux models have often been used in the thermal-hydraulic analysis of nuclear and other systems. Accordingly, we consider, in the current paper, the analysis of five sets of experimental data with different geometries, working fluids, flow rates and void fraction ranges, with a view to obtaining a best fit to the data in the form of a drift-flux model. The results of the analysis show that, for systems with flowing fluid, it is possible to represent the heavy liquid metal void fraction data in the form of a drift-flux correlation with a residual error of as low as 0.016, thus offering an improvement over existing void correlations.

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

Allegro: The European Gas Fast Reactor Demonstrator Project

In the Gas Fast Reactor development plan, ALLEGRO is the first necessary step towards the electricity generating prototype GFR. The ALLEGRO start of operation is planned by 2020. This needs to define all design options in 2010 and to start detailed design studies in 2013. ALLEGRO is a low power Gas Cooled Fast Reactor studied in the European framework. It is a loop type, non electricity generating reactor. Its power is about 80 MW. Several objectives are assigned to ALLEGRO. At first, it will demonstrate the viability of the GFR reactor system, no reactor of this type having been built in the past. Most of the GFR architecture, materials and components features are considered at reduced scale in ALLEGRO, excluding the energy conversion system. ALLEGRO will rely on the same safety options as the reactor system. In addition, the ALLEGRO core will allow the progressive qualification of the GFR ceramic fuel, with the possibility to load some ceramic carbide or nitride sub-assemblies in a first MOX core, with SiC/SiCf cladding and wrappers. When such unit test will be considered convincing enough, the diagrid and circuits are designed to accept full high temperature ceramic cores. The core neutrons can also be used to irradiate structural materials with fast neutron spectrum and in a large temperature range. The core can also include innovative irradiation fuel devices (samples or full bundles) for other reactor systems. Finally, branches on the main intermediate heat exchanger will allow the testing and validation of high temperature components and processes. The pre-conceptual design of ALLEGRO is shared between European partners through the GCFR 6th R&D Framework Program. After recalling the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: • Core design and neutron performances, • The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the MOX core, • Fuel handling principles and solutions, • System design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy (DHR) for depressurized accidents.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.

Fast Reactor Physics

A very simplified model is used to simulate the equilibrium fuel cycle in a sodium-cooled fast reactor, considering (for the sake of comparison) two feed fuels: natural uranium and natural thorium, assuming recycling of all actinides under conditions of constant power density and constant fuel mass. The balance of reaction rates, equilibrium fuel composition, neutron balance, main safety parameters, as well as radiotoxicity and decay heat level of the equilibrium fuel are presented. The paper is a shortened version of the lecture given at FJOH’2013 summer school (Mikityuk, Equilibrium closed fuel cycle, 2013, [1]).