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

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.

Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

Gaëtan Girardin

During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. It is in this context that the Generation IV international forum (GIF) launched the initiative, in 2000, to collaborate on the research and development (R&D) efforts needed for the next generation, i.e. Generation IV, of nuclear reactors. These advanced systems will be ideally deployed beyond the year 2030, following the Generation III or III+ nuclear power plants, which are mainly based on light water technology and are currently entering deployment. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle. Thus, apart from improvements in safety, they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials (fuel pellets of mixed uranium-plutonium carbide within a plate-type, honeycomb structure made of SiC). It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for the previous generation of sodium-cooled reactors. The validation of ERANOS for GFR applications was, therefore, the first task to be realized in the present research. This has entailed a systematic reanalysis of the GFR-relevant, integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970's. Thus, during the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1, in order to derive a reference computational scheme to be used later for the GFR analysis. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement – within 1σ measurement uncertainty – with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal-hydraulics analysis, the CEA code COPERNIC was used. This design work was followed by the study of an appropriate CA implementation scheme (number of CAs and corresponding positions within the core). Detailed neutronics studies revealed the existence of large CA interaction effects, so-called shadowing/anti-shadowing effects leading to an amplification/reduction of the CA worth. The interactions between the absorber pins within a CA, and between the CAs themselves, were investigated in detail, with the goal to optimize the CA efficiency, in terms of the absorber fraction and minimization of the associated heterogeneity effects. The proposed CA pattern consists of 54 absorber pins placed in a triangular lattice. Each absorber pin is a stainless-steel tube filled with highly enriched 10B boron carbide pellets. As a result of the detailed investigations, the absorber pin diameter could be chosen such as to minimize the pin-to-pin influence within the assembly. In particular, a central part of the CA was designed without any absorber pins (zone filled with stagnant helium). A final reduction of the heterogeneity effect (difference between homogeneous and heterogeneous treatments) to 13% was achieved through this feature. Of special importance, the neutronics investigations performed for the reference GFR core ("2004-Core"), especially those related to the CA interactions, have directly contributed to a new core design ("2007-Core"), with the height-to-diameter ratio having been increased to 0.6, compared to 0.3 for the reference core. During the third phase, detailed coupled, 3D neutron-kinetics (NK) and 1D thermal-hydraulics (TH) models were developed for the GFR core, the aim being to arrive at an in-depth understanding of the 3D core behavior during CA driven transients, especially from the viewpoint of spatial effects. The coupled models were developed using the PARCS code for the 3D NK and the TRACE code for the 1D TH modeling. Particular attention was paid to have each individual fuel sub-assembly and CA represented, in order to allow the analysis of local deformations of the 3D distributions of power and safety related parameters, such as the coolant, cladding and fuel temperatures. The validation of the coupled full-core models was performed against reference ERANOS-VARIANT calculations. In particular, the CA worth and reactivity feedbacks were benchmarked, the discrepancies being shown to be relatively low (

EPFL2009

Improvement of the Decay Heat Removal Characteristics of the Generation IV Gas-cooled Fast Reactor

Aaron Simon Epiney

Gas cooling in nuclear power plants (NPPs) has a long history, the corresponding reactor types developed in France, the UK and the US having been thermal neutron-spectrum systems using graphite as the moderator. The majority of NPPs worldwide, however, are currently light water reactors, using ordinary water as both coolant and moderator. These NPPs – of the so-called second generation – will soon need replacement, and a third generation is now being made available, offering increased safety while still based on light water technology. For the longer-term future, viz. beyond the year 2030, R&D is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Since the very beginning of the international cooperation on Generation IV, viz. the year 2000, the main research interest in Europe as regards the advanced fast-spectrum systems needed for achieving complete fuel cycle closure, has been for the Sodium-cooled Fast Reactor (SFR). However, the Gas-cooled Fast Reactor (GFR) is currently considered as the main back-up solution. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high-pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For depressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be powered either by the power grid or by batteries for at least 24 hours. The specific contributions of the present research – aimed at achieving enhanced passivity of the DHR system for the GFR – are design and analysis related to (1) the injection of heavy gas into the primary circuit after a LOCA, to enable natural convection cooling at an intermediate-pressure level, and (2) an autonomous Brayton loop to evacuate decay heat at low primary pressure in case of a loss of the guard-containment pressure. Both these developments reduce the dependence on blower power availability considerably. First, the thermal-hydraulic codes used in the study – TRACE and CATHARE – are validated for gas cooling. The validation includes benchmark comparisons between the codes, serving to identify the sensitivity of the results to the different modeling assumptions. The parameters found to be the most sensitive in this analysis, such as heat transfer and friction models, are then validated via a detailed re-analysis of earlier PSI (EIR, at the time) gas-loop experiments conducted in the 1970s. Conclusions and recommendations on the models to be used for transient analysis are derived. In general, it has been shown that the agreement, between experiments and the correlations for heat transfer and friction used in TRACE and CATHARE, is quite satisfactory. The thus validated codes are then used in the two detailed, DHR improvement studies carried out. The first improvement of the reference DHR strategy is the heavy gas injection. Assuming a DHR blower failure after a LOCA, the helium pressure in the guard containment is not high enough to evacuate the decay heat by natural convection. To improve the natural convection, the effects of injecting different heavy gases (N2, CO2, Ar and a N2/He mixture) into the primary circuit were analyzed, in order to address the possibility of dealing with DHR-blower failure while accepting an intermediate back-up pressure in the guard containment. Furthermore, different injection locations and injection mass flows were considered, and the sensitivity to the number of available DHR loops and LOCA break-sizes was also addressed. It has been found that injecting the heavy gas in the vicinity of the core could lead to overcooling problems. For an injection point sufficiently far from the core, however, both CO2 and N2 are found to be able to cool the core satisfactorily in natural convection. N2 is proposed as the reference, due to possible chemical problems with CO2. The second proposition for DHR improvement is related to the possibility of a simultaneous guard-containment failure, i.e. a loss-of-back-up-pressure (LOBP) combined with a blower failure after a LOCA. In this case the natural convection, even with heavy gas injection, is no longer strong enough to evacuate the decay heat. To address this issue, the possibility of decay heat removal via use of a dedicated autonomous Brayton cycle – as a standalone DHR loop – has been investigated. First, an analytical Brayton cycle model has been set up, so as to identify convenient machine design points and to study the machine's off-design behavior. Two machine designs have then been drawn up: one for helium in order to provide a reference for understanding the Brayton loop behavior in a generic sense, and the other for nitrogen which is the envisaged gas to be injected after a LOCA. Both, the design of the proposed devices and their validation are discussed. Finally, a detailed transient analysis, involving usage of both heavy-gas injection and the Brayton device (i.e. of the complete, proposed DHR system), is presented. This serves to illustrate the effectiveness of the new strategy for the highly hypothetical worst-case scenario of sequential failures following a LOCA.

EPFL2010

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