Jiri Krepel

Dynamics of Molten Salt Reactors

Abstract The dynamics of the molten salt reactor (MSR), one of the Generation IV International Forum concepts, was studied. The graphite-moderated channel-type MSR was selected for numerical simulation. MSR, a liquid-fueled reactor, has specific dynamics with two physical peculiarities: The delayed neutron precursors are drifted by the fuel flow, and the fission energy is released directly into the coolant. Presently, there are few accessible numerical codes appropriate for MSR simulation; therefore, the DYN1D-MSR and DYN3D-MSR codes were developed based on the light water reactor dynamics code DYN3D. These allow calculation of one-dimensional and full three-dimensional transient neutronics in combination with parallel channel-type thermal hydraulics. The codes were validated with experimental results of the Molten Salt Reactor Experiment from Oak Ridge National Laboratory and applied to several transients typical for a liquid fuel system. Those transients were initiated by reactivity insertion, by cold or overfueled slugs, by the fuel pump start-up or shutdown, or by the blockage of selected fuel channels. In these considered transients, the response of MSR is characterized by the immediate change of the fuel temperature relative to the temperature at that power level. This causes fast insertion of feedback reactivity, which is negative for power-related temperature increase. On the other hand, the graphite response is slower, and its feedback coefficient depends on the core size and geometry. The addition of erbium to the graphite can ensure negative feedback and inherent safety features also for big low leakage cores. The DYN1D-MSR and DYN3D-MSR codes have been shown to be effective tools for MSR dynamics studies. The MSR response to the majority of transients is considered acceptable within safety margins as long as the graphite feedback coefficient is negative. A transient that is possibly an exception is a local channel blockage.

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.

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 and Verification of Dynamics Code for Molten Salt Reactors

To perform transient analysis for Molten Salt Reactors (MSR), the reactor dynamics code DYN3D developed in FZR was modified for MSR applications. The MSR as a liquid fuel system can serve as a thorium breeder and also as an actinide burner. The specifics of the reactor dynamics of MSR consist in the fact, that there is direct influence of the fuel velocity to the reactivity, which is caused by the delayed neutrons precursors drift. This drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit. This leads to a reactivity loss due to the fuel flow acceleration or to the reactivity increase in the case of deceleration. For the first analyses, a 1D modified version DYN1D-MSR of the code has been developed. By means of the DYN1D-MSR, several transients typical for the liquid fuel system were analyzed. Transients due to the overcooling of fuel at the core inlet, due to the reactivity insertion, and the fuel pump trip have been considered. The results of all transient studies have shown that the dynamic behavior of MSR is stable when the coefficients of thermal feedback are negative. For studying space-dependent effects like e.g. local blockages of fuel channels, a 3D code version DYN3D-MSR will be developed. The nodal expansion method used in DYN3D for hexagonal fuel element geometry of VVER can be applied considering MSR design with hexagonal graphite channels.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.

Coupled 3-D Neutronics/Thermal-Hydraulics Optimization Study For Improving The Response Of A 3600 Mw(Thermal) Sfr Core To An Unprotected Loss-Of-Flow Accident

The sodium-cooled fast reactor (SFR), as a fast-neutron spectrum system, is characterized by several performance advantages. In particular, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. However, the SFR has one dominating neutronics drawback, namely, there is generally a positive reactivity effect when there is voiding of the sodium coolant in the core. Furthermore, this effect becomes even stronger in the equilibrium closed fuel cycle. Considering that in a hypothetical SFR unprotected loss-of-flow (ULOF) accident scenario, i.e., flow rundown without SCRAM, sodium boiling can be anticipated to occur, it is crucial to assess the corresponding impact of the positive sodium void effect. An optimization study for improving the safety characteristics of a large [3600-MW(thermal)] SFR has currently been conducted in the above context. The dynamic core response to a reference ULOF scenario is investigated with the use of a coupled three-dimensional neutronics/thermal-hydraulics PARCS/TRACE model. The starting point of the study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the sodium void effect, the core has been modified by introducing an upper sodium plenum, along with a boron layer above it. Furthermore, the original core height-to-diameter ratio is reduced. In comparison to the reference ESFR core behavior, certain improvements are achieved, thanks to the static neutronics optimization carried out. However, these changes are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting during the considered ULOF event. Thermal-hydraulics optimization has thus been considered necessary, in order to (a) prevent sodium flow blockage in the fuel channel and (b) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. The corresponding measures taken are (a) the introduction of an innovative wrapper design, which features small openings in each side surface of the fuel assembly, and (b) replacement of the original upper sodium plenum by an extended fission gas plenum. Following implementation of these thermal-hydraulics-related design changes, one arrives at a final configuration of the SFR core, in which, for the selected accident scenario, a new “steady state” involving stable sodium boiling is found to be achievable, with melting of neither cladding nor fuel. Such a satisfactory behavior has been confirmed not only for the beginning-of-life state of the core but also for the equilibrium closed fuel cycle.