Management of actinide waste inventories in nuclear phase-out scenarios

The improvement of the "radiological cleanliness" of nuclear energy is a primary goal in the development of advanced reactors and fuel cycles. The multiple recycling of actinides in advanced nuclear systems with fast neutron spectra represents a key option for reducing the potential hazard from high-level waste, especially when the fuel cycle is fully closed. Such strategies, however, involve large inventories of radiotoxic, transuranic (TRU) nuclides in the nuclear park, both in-pile and out-of-pile. The management of these inventories with the help of actinide burners is likely to become an important issue, if nuclear energy systems are eventually phased out, i.e. replaced by other types of energy systems. The present paper compares phase-out scenarios for two transmutation strategies involving fast reactors (FRs) and accelerator-driven systems (ADSs), respectively, operating in symbiosis with conventional light water reactors (LWRs). Particular objectives are to evaluate and compare the TRU reduction performance of the systems as a function of the phase-out time and to determine the appropriate phase-out length for different phase-out criteria. In this connection, an interesting aspect concerns the continuous optimisation of the fuel cycle to counterbalance the reactivity decrease due to the depletion of the fissile isotopes in the fuel. It will be shown that both FRs and ADSs can achieve the goal, provided that the phase-out operation can be continued for about a hundred years. 2008 Elsevier Ltd. All rights reserved.

Standardisation des outils de calcul pour les ADS et leur application à différents scénarios de transmutation des déchets

Marco Cometto

The management of radioactive wastes from the nuclear fuel cycle has become an important issue in the development of future, more sustainable nuclear energy systems. Partitioning and transmutation (P&T) of actinides and some long-lived fission products could reduce the mass and radiotoxicity of highlevel wastes and possibly ease repository licensing requirements. Influenced by political and technological developments, an increasing number of countries employing nuclear power have become interested in P&T technology which involves the development of new types of critical and/or sub-critical reactors and very efficient fuel cycle strategies. The present thesis is closely connected to two P&T related projects of the OECD Nuclear Energy Agency in Paris, a numerical benchmark exercise for an accelerator-driven minor actinides burner and a calculational system study on the role of accelerator-driven systems (ADS) and advanced fast reactors (FR) in advanced nuclear fuel cycles. The author co-coordinated the benchmark exercise and performed the comparative analysis of the different "fuel cycle schemes" investigated in the system study. The benchmark exercise contributed to a greater understanding of the physical phenomena characteristic of the transmutation of actinides in a closed fuel cycle, and allowed to test and improve the tools used for modeling ADSs and corresponding advanced fuel cycles. This work thus effectively led to the development of an efficient and more complete calculational scheme, which has permitted an in-depth and reliable analysis of different transmutation scenarios to be carried out. To quantitatively assess the advantages and drawbacks of transmutation, we considered a limited number of "base" transmutation scenarios which constitute an envelope for various possible transmutation strategies employing fast spectrum systems. We then evaluated, for an equilibrium situation, the main aspects which characterise the different transmutation strategies, such as the nuclear park composition, the mass and toxicity of the waste, the need for natural uranium, the in-pile and out-pile mass inventories and, finally, the fuel cycle requirements. Thus, this study has allowed for the reliable comparison of performances of the three main options in the nuclear waste management, viz. open cycle, recycling of plutonium alone and recycling of all the actinides, as well as an evaluation of the advantages and disadvantages of ADSs in comparison to advanced critical reactors in the management of minor actinides or transuranics. The last part of the dissertation is dedicated to a phase-out study for two schemes employing either ADSs or critical reactors. It has been shown thereby that advanced systems, optimised for an equilibrium fuel composition, can indeed be used to reduce the inventory of transuranics. A comparison of ADS and FR performances has been made in this context, and the relative benefits of transmutation have been quantified in each case taking into consideration, more realistically, the residual mass inventory after the shutdown of all nuclear installations.

EPFL2003

Analysis of Advanced Sodium-cooled Fast Reactor Core Designs with Improved Safety Characteristics

Kaichao Sun

In order to meet the steadily increasing worldwide energy demand, nuclear power is expected to continue playing a key role in electricity production. Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years' nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it – after the initial loading – to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor-years of practical operation. As a fast-neutron-spectrum system, 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. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback, viz. there is generally a positive reactivity effect when sodium coolant is removed from the core. Furthermore, this so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed improvements address both neutronics and thermal-hydraulics aspects. Furthermore, emphasis has been placed on not only the beginning-of-life (BOL) state of the core, but also on the beginning of closed equilibrium fuel cycle (BEC) state. An important context for the current thesis is the 7th European Framework Program's Collaborative Project for a European Sodium Fast Reactor (CP-ESFR), the reference 3600 MWth ESFR core being the starting point for the conducted research. The principally employed computational tools belong to the so-called FAST code system, viz. the fast-reactor neutronics code ERANOS, the fuel cycle simulating procedure EQL3D, the spatial kinetics code PARCS and the system thermal-hydraulics code TRACE. The research has been carried out in essentially three successive phases. The first phase has involved achieving a clearer understanding of the principal phenomena contributing to the SFR void effect. Decomposition and analysis of sodium void reactivity have been carried out, while considering different fuel cycle states for the core. Furthermore, the spatial distribution of void reactivity importance, in both axial and radial directions, is investigated. For the reactivity decomposition, two methods – based on neutron balance considerations and on perturbation theory, respectively – have been applied. The sodium void reactivity of the reference ESFR core has been, accordingly, decomposed reaction-wise, cross-section-wise, isotope-wise and energy-group-wise. Effectively, the neutron balance based method allows an in-depth understanding of the "consequences" of sodium voidage, while the perturbation theory based method provides a complementary understanding of the "causes". The second phase of the research has addressed optimization of the reference ESFR core design from the neutronics viewpoint. Four options oriented towards either the leakage component or the spectral effect have been considered in detail, viz. introducing an upper sodium plenum and boron layer, varying the core height-to-diameter (H/D) ratio, introducing moderator pins into the fuel assemblies, and modifying the initially loaded plutonium content. The sensitivity of the principal safety and performance parameters, viz. void reactivity, Doppler constant, nominal reactivity and breeding gain, has been evaluated with respect to each of the options. Finally, two synthesis core concepts – representing different selected combinations of the optimization options – have been proposed. The third and last phase has been to test the proposed optimized designs in terms of the dynamic core response to a representative ULOF accident, in which sodium boiling occurs. Such a hypothetical transient is of prime importance for SFR safety demonstration, since it may lead to a severe accident situation, where both cladding and fuel melt. As compared to the response of the reference ESFR core, certain improvements are indeed achieved with the static neutronics optimization carried out. However, these are found, in themselves, to be insufficient as regards the prevention of cladding and fuel melting. Thermal-hydraulic optimization has thus been considered necessary to: 1) prevent sodium flow blockage in the fuel channels and 2) avoid boiling instabilities caused by the vaporization/condensation process in the upper sodium plenum. Following implementation of appropriate 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 shown to be achievable, with melting of neither cladding nor fuel. Such satisfactory behavior has been confirmed not only for the beginning-of-life core state, but also for the equilibrium closed fuel cycle.

EPFL2012

Development and application of an advanced fuel model for the safety analysis of the generation IV gas-cooled fast reactor

Petr Petkevich

Until about the year 2030, current-day nuclear power plants (NPPs) will be replaced by so-called Gen-III or Gen-III+ units, which are mainly based on light water reactor technology. The principal new features are increased safety and improved economical effectiveness. However, these systems use the same fuel forms and are based on the same fuel cycle. Beyond 2030, the interest is likely to shift towards fourth generation NPPs, which offer the possibility of complete fuel cycle closure. Generation-IV reactor concepts include both thermal and fast systems, and involve a wide range of fuel forms and compositions. The present research has been focused on the development of a thermo-mechanical model for the innovative fuel design of the Generation-IV Gas-cooled Fast Reactor (GFR). The principal distinctive feature of the fuel is that the fuel pellets are arranged within plates which enclose an inner honeycomb structure. Apart from the geometry, the usage of new materials is foreseen. Thus, the fuel pellets are of mixed uranium-plutonium carbide, and the cladding is bulk or fiber-reinforced SiC. The setting up of an appropriate materials database was thus the very first task which had to be carried out in the current work. The main purpose of the currently developed model is to provide reliable data, in the context of transient analysis, for the calculation of the principal neutronic feedbacks in the GFR core, viz. the fuel temperature for the Doppler effect and the fuel plate deformation for the axial core expansion effect. None of the available fuel modeling codes is suitable for a realistic simulation of the GFR fuel, as the inner honeycomb structure cannot be explicitly taken into account. The development work has been carried out largely in the context of PSI's generic code system for fast reactor safety analysis, FAST. Thereby, it has mainly involved extension of the thermo-mechanical code FRED, developed originally for the modeling of traditional rodded fuel. Within the FAST system, FRED is coupled to the TRACE code for the thermal-hydraulic modeling, so that the present work has comprised not only the development of a 2D FRED model for the plate-type GFR fuel, but also the implementation of corresponding changes in TRACE for ensuring appropriate information exchange between the two codes. The 2D thermo-mechanical model has been developed with certain assumptions. Since no experimental data exist for this fuel type, benchmarking of the new simulation tool was carried out by building up a detailed 3D model using the finite-elements code ANSYS. The 3D model has, moreover, been employed for conducting certain supplementary studies to obtain an in-depth understanding of the thermal and mechanical behavior of the fuel. It was found how the complex, multi-dimensional, heat transfer in the plate-type fuel accounts for the discrepancies between results of 2D and 1D simulations. Furthermore, it was shown that, under certain conditions, the temperature field can be well predicted by the 1D model with slight modifications of the solution algorithm. Other insights have been obtained from the detailed mechanical analysis. Thus, it has been shown that, during operation, cusping occurs at the pellet periphery which results in an unfavorable concentration of stresses both in pellet and cladding. Several alternative ways to optimize the fuel design and to avoid, or at least minimize, this effect have been proposed. As mentioned, the new fuel model is intended for usage in GFR transient analysis. In order to quantify the impact of the current model developments, a range of hypothetical accident events have been analyzed using the FAST code system, with and without usage of the new fuel model. It has been shown that the pure geometry effects on the temperatures are quite significant. However, for the specific honeycomb structure geometry considered, these are somewhat mitigated by the fuel and cladding expansion and the corresponding decrease of the axial fuel-cladding gas gaps. Accounting for the deformations thus, in this case, brings the 2D model results closer to those of a 1D treatment. The evolution of reactivity feedbacks and the reactor power were evaluated using a point-kinetics approximation and hence were driven by the average temperature change rate (dT/dt) rather than by the absolute temperature values. Correspondingly, the results from the two models are not drastically different. The more significant discrepancies have been obtained for the transients with similar magnitudes of the Doppler and axial core expansion effects (core overcooling, loss of heat sink). For the transient-overpower transient, which is mainly determined by the fuel temperature, very similar results are obtained for total reactivity and reactor power. In brief, the present research has resulted in a flexible and easy-to-use simulation tool for carrying out reliable transient analysis for the Generation IV GFR with its innovative plate-type fuel, the implemented methodology combining an explicit accounting of the fuel inner structure with acceptable computing time.

EPFL2008