Single- and Two-Phase Flow Modeling for Coupled Neutronics/Thermal-Hydraulics Transient Analysis of Advanced Sodium-Cooled Fast Reactors

Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major R&D related issue. The current doctoral research aims at the development of a computational tool for the in-depth understanding of SFR core behavior during accidental transients, particularly those including boiling of the coolant. An accurate modeling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. Models for the representation of sodium two-phase flow are not present in most of the thermal-hydraulics codes used currently, and these have specifically been focused upon. The particular contributions of the present research are: (1) implementation of sodium two-phase flow models into the thermal-hydraulics code TRACE, which forms part of the FAST code system at PSI and, as such, can easily be coupled to the spatial neutron kinetics code PARCS; (2) validation of the TRACE sodium single- and two-phase flow modeling using out-of-pile sodium boiling experiments; (3) validation of the coupled TRACE/PARCS code system on the basis of experimental reactor data; and (4) application of the developed new tool for the core behavior analysis of an advanced SFR during a transient with boiling onset. The extension of the TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. A review has first been performed of the models available in the open literature for the representation of the interfacial and wall-to-fluid transfer mechanisms. The different correlations have then been implemented as options in the extended TRACE code. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the liquid film after dryout onset. The validation of the extended TRACE code has been achieved through the successful simulation of out-of-pile experiments. A review of available sodium boiling test data has first been carried out, and complementary tests have then been selected to assess the quality of the different physical models. These tests, performed in the 1980s, include the study of the pressure drop and cooling limits under quasi steady-state conditions, as well as the simulation of a loss-of-flow transient. Sensitivity analyses, using the specifically implemented correlations, have enabled one to identify the most pertinent physical parameters and to define, for each, the most appropriate model. Usage of the set of models thus selected has demonstrated the capacity of the extended TRACE code to predict, with satisfactory accuracy, the main thermal-hydraulics characteristics such as the single- and two-phase pressure drop and heat transfer, as also the characteristic quantities describing the sodium two-phase flow, e.g. boiling inception, void fraction evolution and expansion of the boiling region, pressure evolution, as well as coolant and clad temperatures. The natural convection test conducted in 2009 in the Phenix reactor has been used to validate TRACE single-phase sodium flow modeling. This represents the first international benchmark exercise conducted on the basis of actual SFR experimental data. Data from the Phenix test have additionally been used as basis for the validation of the FAST code system as a whole. Analyses based on a point-kinetics TRACE model and on coupled TRACE/PARCS 3D-kinetics modeling have enabled an in-depth understanding of the transient behavior of a sodium-cooled fast reactor core, as well as the identification of potential improvements in the FAST code system. The experimental power evolution could be satisfactorily reproduced within the measurement uncertainties with both models, and the detailed analysis of the core neutronics has enabled one to define the most important reactivity feedbacks taking place during the considered transient. In the final part of the thesis, the developed tool has been applied to the simulation of a hypothetical, unprotected loss-of-flow event for one of the European SFR (ESFR) core concepts. This study has demonstrated the new calculational tool's capability to adequately simulate the core response through the modeling of single- and two-phase sodium flow, coupled to 3D neutron kinetics. Thereby, the space-dependent reactivity feedbacks, such as the void and Doppler effects, have been shown to be modeled accurately. The resulting analysis has shown the ability of the TRACE/PARCS modeling to predict the expansion of the boiling region and calculate the resulting feedbacks, as well as to predict the interactions between parallel boiling channels. This first-of-a-kind study has provided detailed results for the thermal-hydraulics and neutronics parameters during the pre-severe phase of the simulated accident, thus allowing a comprehensive understanding of the core behavior during such transients. In brief, the present research has led to the development of a key calculational tool for SFR safety analysis. A first application of the new tool has demonstrated its potential for usage in SFR design optimization aimed at enhanced safety.

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

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

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