Andrei Rineiski

Numerical Studies of Axial Fuel Shuffling

The concept of traveling wave reactor (TWR) applies the mechanism of self sustainable and propagation nuclear fission traveling waves in fertile media of 238U and 232Th to achieve very high fuel utilization. However, the long wave length of such fission traveling wave puts a limit on the applicability of the TWR concept. The axial fuel shuffling strategy is proposed based on the mechanism of asymptotic nuclear fission traveling wave, and is applied to a sodium-cooled fast reactor (SFR) loading metallic 238U fuel. The multi-group deterministic neutronic code ERANOS with JEFF3.1 data library is used as a basic tool to perform the neutronics and burn-up calculations. The calculations are firstly performed in a 1-D case for parametric understanding, and further extended to a 2-D R-Z case. The shuffling calculations for the 1-D and 2-D SFR model described in this paper brought about some interesting results. The results indicate that keff parabolically varies with the shuffling period, while the burn-up increases linearly. The highest burn-up achieved in 2-D case is 46at%. The power shape distortion in 2-D case is observed, and the power peaking factor is much higher than that in 1-D case, but it decreases with the shuffling period increasing.

COUPLE, A Time-Dependent Coupled Neutronics and Thermal-Hydraulics Code, and its Application to MSFR

Molten salt reactor (MSR) using liquid fuel is one of the Generation-IV candidate reactors. Its liquid fuel characteristics are fundamentally different from those of the conventional solid-fuel reactors, especially the much stronger neutronics and thermal hydraulics coupling is drawing significant attention. In this study, the fundamental thermal hydraulic model, neutronic model, and some auxiliary models were established for the liquid-fuel reactors, and a time-dependent coupled neutronics and thermal hydraulics code named COUPLE was developed to solve the mathematic models by the numerical method. After the code was verified, it was applied to the molten salt fast reactor (MSFR) to perform the steady state calculation. The distributions of the neutron fluxes, delayed neutron precursors, velocity, and temperature were obtained and presented. The results show that the liquid fuel flow affects the delayed neutron precursors significantly, while slightly influences the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank, which leads to the maximal temperature about 1100 K at the centre of the vortex. The results can provide some useful information for the reactor optimization.Copyright

Comparison of the Waste Transmutation Potential of Different Innovative Dedicated Systems and Impact on the Fuel Cycle

The performances of three different types of innovative transmutation systems have been investigated in order to assess in a comparative way their potential to manage nuclear waste arising in a geographical region, where different countries have different policies with respect to nuclear energy development, but share the objective of a common optimized waste management strategy in order to minimize the waste masses sent to a geological repository. The three systems are 1) a critical low conversion ratio fast reactor (LCFR); 2) an accelerator driven system (ADS) and 3) a hybrid fission-fusion system (FFH). In order to simplify the comparison, the three systems have been loaded with comparable fuels, in particular with the same Pu to Minor Actinides (MA) ratio. A waste management scenario study has been performed: the results show that, apart from the technological readiness of each single option, the performances, in terms e.g. of time needed to eliminate specific spent fuel inventories or in terms of reduction of decay heat and radiotoxicity in a deep geological repository, are rather comparable.

Reactivity Coefficients in BN-600 Core with Minor Actinides

In 1999, the IAEA has initiated a Coordinated Research Project on “Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects.” Three benchmark models representing different modifications of the BN-600 fast reactor have been sequentially established and analyzed, including a hybrid core with highly enriched uranium oxide and MOX fuel, a full MOX core with weapons-grade plutonium, and a MOX core with plutonium and minor actinides coming from spent nuclear fuel. The paper describes studies for the latter MOX core model. The benchmark results include core criticality at the beginning and end of the equilibrium fuel cycle, kinetics parameters, spatial distributions of power, and reactivity coefficients obtained by employing different computation tools and nuclear data. Sensitivity studies were performed to better understand in particular the influence of variations in different nuclear data libraries on the computed results. Transient simulations were done to investigate the consequences of employing a few different sets of power and reactivity coefficient distributions on the system behavior. The obtained results are analyzed in the paper.

Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

Recent developments extending the capabilities of the SIMMER-III [1, 2] code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmutater. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed.© 2006 ASME

Numerical Investigation of Corium Coolability in Core Catcher: Sensitivity to Modeling Parameters

To avoid settling of molten materials directly on the vessel wall in severe accident sequences, using of a ‘core catcher’ device in the lower plenum of sodium fast reactor designs is considered. The device is to collect, retain and cool the debris, created when the corium falls down and accumulates in the core catcher, while interacting with surrounding coolant. This Fuel-Coolant Interaction (FCI) leads to an energetic heat and mass transfer process and may threaten vessel integrity. For simulation of severe accidents, including FCI, the SIMMER code is employed at KIT. SIMMER is an advanced tool for CDA analysis of liquidmetal fast reactors (LMFRs) and other GEN-IV systems. It is a multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics code coupled with a fuel-pin model and a spaceand energy-dependent neutron kinetics model. However, the experience of SIMMER application to simulation of corium relocation and related FCI is limited. To verify the code applicability to FCI in a large system, an invessel model based on European Sodium Fast Reactor (ESFR) was established and calculated by the SIMMER code. In addition, a sensitivity analysis on some modeling parameters is also conducted to examine their impacts. The characteristics of the debris in the core catcher region, such as debris mass and composition are compared. Besides that, the pressure history in this region, the mass of boiling sodium vapor and average temperature of liquid sodium, which can be treated as FCI quantitative parameters, are also discussed. It is expected that the present study can provide some numerical experience of the SIMMER code in plant-scale corium relocation and related FCI simulation.

Decay Heat Benchmark for Uranium-Free Fuels With Minor Actinides

The decay heat (energy due to decay of unstable nuclei) is a small fraction of reactor power at nominal conditions, but after reactor shut-down it is the most important heat source. For taking this source into account in design and safety studies, recommendations are available for fuels of operating reactors, such as UOX and MOX. Fuels for EFIT (European Facility for Industrial Transmutation), unlike UOX and MOX, should contain a significant amount of Minor Actinides (MAs) that would influence decay heat. CEA, CIEMAT, ENEA, FZK (now KIT), PSI and SCK•CEN established a benchmark case and computed decay heat curves for MA-bearing fuels and a MOX-type fuel. The decay heat in the fuels with MAs is appreciably higher than in MOX, except for low burnup cases after short cooling times. This should be taken into account in the design of the decay heat removal system for EFIT. The obtained differences between the decay heat in MA-bearing and MOX fuels are supposed to be representative for the benchmark (or similar) conditions. More effort is needed to evaluate the uncertainties of the computed results.Copyright