European Lead-cooled System Research Articles

Monte Carlo methods for numerical simulations of the Lead-Cooled Fast Reactor

Closed nuclear fuel cycle is one of the most promising options for the efficient use of nuclear energy resources with full recycling of long-lived transuranic elements. However, it can be implemented only in fast breeder reactors of Generation IV. The paper shows our methodology applied in the analysis of the lead-cooled fast reactor equilibrium fuel cycle using the Continuous Energy Monte Carlo Burnup Code – MCB. The implementation of novel modules for nuclear transmutation trajectory folding allows us to trace the life cycle of crucial minor actinides from the beginning of the reactor’s life towards the state of adiabatic equilibrium. Changes in the mass contribution to gateway isotopes 242Pu, 243Am and 244Cm during 124.2 years were considered in the study. Numerical demonstration was performed for the reactor core designed within the European Lead-cooled System (ELSY) project and redefined in the follow-up Lead-cooled European Advanced Demonstration Reactor (LEADER) project.

Modeling representative Gen-IV molten fuel reactivity effects in the ZEPHYR ZPR - LFR analysis

© 2018 John Wiley & Sons, Ltd. Studies related to severe core accidents constitute a crucial element in the safety design of Gen-IV systems. A new experimental program, related to severe core accidents studies, is proposed for the zero-power experimental physics reactor (ZEPHYR) future reactor. The innovative program aims at studying reactivity effects at high temperature during degradation of Gen-IV cores by using critical facilities and surrogate models. The current study introduces the European lead-cooled system (ELSY) as an additional Gen-IV system into the representativity arsenal of the ZEPHYR, in addition to the sodium-cooled fast reactors. Furthermore, this study constitutes yet another step towards the ultimate goal of studying severe core accidents on a full core scale. The representation of the various systems is enabled by optimizing the content of plutonium oxide in the ZEPHYR fuel assembly. The study focuses on representing reactivity variation from 900°C at nominal state to 3000°C at a degraded state in both ELSY and Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) cores. The study utilizes the previously developed calculation scheme, which is based on the coupling of stochastic optimization process and Serpent 2 code for sensitivity analysis. Two covariance data are used: the ENDF 175 groups for ELSY and the Covariance Matrix Cadarache (COMAC) 33 groups for ASTRID. The effect of the energy group structure of the covariance data on the representativity process is found to be significant. The results for single degraded ELSY fuel assembly demonstrate high representativity factor (>0.95) for reactivity variation and for the criticality level. Also, it is shown that the finer energy group structure of the covariance matrices results in dramatic improvement in the representation level of reactivity variations.

The utilization of thorium in Generation IV reactors

The main goal of this paper is to show how thorium, as an alternative nuclear fuel, could be applied as fuel in a Generation IV reactor. The paper focuses on the multiplication factor, the produced 233U and delayed neutron fraction in infinite lattice models. For the investigations, simplified models of a fuel assembly of five design types of the six reactor concepts were elaborated. The MSR reactor type is out of scope of this paper due to the fact that it is designed for the utilization of thorium. Although the fissile isotope content was not increased to compensate the thorium caused multiplication factor decrease, the burnup calculations suggest that the designs of ESFR (European Sodium-Cooled Fast Reactor) and ELSY (European Lead-cooled System) are the most promising types according to the trend of the multiplication factor changes and the amount of produced fissionable 233U.

Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste.We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis includes burnup calculations in a two-batch fuel reloading pattern, assessment of the equilibrium fuel composition based on MOX with minor actinides, examination of power peaking with its evolution with burnup, and modeling of control rod operation for its influence on breeding gain and criticality.

Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Abstract The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

Parametric Study of Sloshing Effects in the Primary System of an Isolated Lead-Cooled Fast Reactor

Risks related to sloshing of liquid metal coolant due to seismic excitation need to be investigated. Sloshing effects on reactor performance include first, fluid-structure interaction and second, gas entrapment in the coolant with subsequent transport of void to the core region. While the first can hypothetically lead to structural damage or coolant spill, the second increases the risk of a reactivity insertion accident and/or local dryout of the fuel. A two-dimensional computational fluid dynamics study is carried out in order to obtain insights into the modes of sloshing depending on the parameters of seismic excitation. The applicability and performance of the numerical mesh and the Eulerian volume of fluid method used to track the free surface are evaluated by modeling a simple dam break experiment. Sloshing in the cold plenum free surface region of the European Lead-cooled SYstem (ELSY) conceptual pool-type lead-cooled fast reactor (LFR) is studied. Various sinusoidal excitations are used to imitate the seismic response at the reactor level. The goal is to identify the domain of frequencies and magnitudes of the seismic response that can lead to loads threatening the structural integrity and possible core voiding due to sloshing. A map of sloshing modes has been developed to characterize the sloshing response as a function of excitation parameters. Pressure forces on vertical walls and the lid have been calculated. Finally, insight into coolant voiding has been provided.

Evaluation of the Fluid-Structure Interaction Effects in a Lead-Cooled Fast Reactor

The aim of the study is to investigate the structural effects induced by a beyond design basis earthquake on the main safety relevant structures and components of an isolated liquid metal reactor, such as the European Lead-cooled SYstem (ELSY) or ALFRED projects.An extensive R&D program related to heavy-metal cooled systems was recently carried out as Euratom projects of the 6th and 7th Framework Programmes, addressing many of the most important issues related to the viability of a lead-cooled fast reactor.The importance of seismic effects is mainly related to the high inertial forces of the primary coolant (liquid metal) and associated with the impact of the liquid waves on the reactor structures. The isolation devices considered for the design were represented by means of an iso-elastic approach. Moreover, the influence of isolator failure was also evaluated.The fluid-structure interaction and the sloshing phenomenon, characterized by hydrodynamic and impact forces, were numerically investigated, since an explicit analytical solution for structures of such complex geometry is not possible. Numerical calculations (i.e., dynamic nonlinear analyses) were carried out with appropriate finite element method codes and external coupling.A validation analysis was further performed to check the consistency and adequacy of the method used with respect to the American Society of Civil Engineers (ASCE) 4-98 rules.The accelerations propagated in the reactor building confirmed the favorable effect of the seismic isolation, even with 2% faulted isolators. The results indicated that the stress state, in the reactor internals, is not sufficient to impair their structural integrity, although there is localized plastic deformation.

A multi-physics time-dependent model for the Lead Fast Reactor single-channel analysis

This paper presents a multi-physics and time-dependent model for single-channel transient analysis of a Lead Fast Reactor (LFR). The work focuses on the coupling among the neutronic, the thermal-elastic and the fluid-dynamic phenomena in the considered reactor channel, benefiting from a finite-element scheme of analysis that is implemented in the same simulation environment (COMSOL Multiphysics®), and within a computational domain featured by a moving mesh. A purpose-made six-group neutron diffusion model is developed, which allows to take into account the local dependency of the neutron macroscopic cross-sections on the temperature and density fields. The potential of the multi-physics model to estimate the effective neutron multiplication factor, by means of a comparison with the Monte Carlo code SERPENT, is assessed. A special attention is given to the capability to implicitly catch the thermal-hydraulic and thermal-expansion feedbacks on reactivity, without relying on cross-sections corrective factors. The proposed multi-physics model is employed to investigate the active-core average conditions of the ELSY (European Lead cooled System) reactor, both in steady-state operation and during two transient scenarios. It is shown that the presented model represents a suitable simulation tool for a preliminary investigation of the LFR dynamics, and allows to simultaneously evaluate a wide set of the reactor channel parameters.

The impact of americium on transients in the European Lead-cooled SYstem ELSY loaded with nitride fuel

The Unprotected Loss-of-Flow transient (ULOF) and the Unprotected Transient-Over-Power (UTOP) have been evaluated for a modified design of the European Lead-cooled SYstem (ELSY) using nitride fuel. Additional americium has been added to the core in concentrations of 0%, 4% and 8%. During ULOF, fuel and clad asymptotic temperatures stabilize below limits. For UTOP, however, the maximum allowed cladding temperature limit is exceeded at EOC and introduction of a power penalty is necessary for Am transmutation. The power penalty is higher than that reported for SFR.

The impact of americium on the ULOF and UTOP transients of the European Lead-cooled SYstem (ELSY)

The Unprotected Loss-of-Flow (ULOF) transient and the Unprotected Transient-Over-Power (UTOP) have been evaluated for the European Lead-cooled SYstem (ELSY) for americium concentrations in the fuel between 0% and 10%. With americium in the core, the ELSY design exceeds the maximum allowed asymptotic temperature limit for fuel during UTOP, except for low Am contents. During ULOF, fuel and clad temperatures stabilize below the asymptotic temperature limits.

Preliminary evaluation of structural response of ELSY reactor in the after shutdown condition

Abstract The paper is aimed at preliminary evaluating the load bearing capability of a Generation IV metal cooled reactor (LFR), like the European Lead-cooled System (ELSY), with the intent to support the viability of lead technology for use in a future commercial power plant with waste transmutation capability. Two relevant safety aspects were investigated: the fluid–structure interaction, induced by the propagation of seismic loading (in the event of safe shutdown earthquake), and the high thermal loads due to the residual decay heat (influenced also by the burn up level). The first issue depends on the presence of a liquid free surface that, especially, in SSE condition allows for fluid motion (“sloshing”); the forming and the impact of lead waves could impair the integrity of reactor structures. The second aspect deals with the thermal effects induced by the irradiated core heat source after the reactor shutdown. In order to evaluate the ELSY reactor structural response, induced by both high temperature and hydrodynamic loads, appropriate 3-D finite element models were set up and implemented in MARC code. In addition the input decay power after shutdown has been evaluated by means of MCNPX and Origen2.2 codes. The obtained preliminary results provide information to make possible an upgrading of the reactor design, however with no significant modification of their functionality.

Conceptual core design study for a high-flux LFR demonstrator

A preconceptual core design study for a pool-type Lead-cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed in the frame of an I-NERI between the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL), based on the European Lead-cooled System (ELSY) reference concept. A demonstration reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. DEMO specifications as a nuclear power facility demonstrating ELSY main features and performance, besides validating design methodology and tools, have been defined. Suitable design parameters have been set to meet the foremost objective of reaching a high fast neutron flux while respecting all technological constraints. Preliminary thermal-hydraulic analyses have been carried out to verify safety limits were not exceeded.

Lead-cooled system design and challenges in the frame of Generation IV International Forum

The Generation IV International Forum (GIF) Technology Roadmap identified the Lead-cooled Fast Reactor (LFR) as a technology well suited for electricity generation, hydrogen production and actinide management in a closed fuel cycle. One of the most important features of the LFR is the fact that lead is a relatively inert coolant, a feature that conveys significant advantages in terms of safety, system simplification, and the consequent potential for economic performance. In 2004, the GIF LFR Provisional System Steering Committee was organized and began to develop the LFR System Research Plan. The committee selected two pool-type reactor concepts as candidates for international cooperation and joint development in the GIF framework: these are the Small Secure Transportable Autonomous Reactor (SSTAR); and the European Lead-cooled System (ELSY). The high boiling point (1745 °C) of lead has a beneficial impact to the safety of the system, whereas its high melting point (327.4 °C) requires new engineering strategies, especially for In-Service-Inspection and refuelling. Lead, especially at high temperatures, is also relatively corrosive towards structural materials. This necessitates that coolant purity and the level of dissolved oxygen be carefully controlled, in addition to the proper selection of structural materials. For the GIF LFR concepts, lead has been chosen as the coolant rather than Lead–Bismuth Eutectic primarily because of its greatly reduced generation of the alpha-emitting 210Po isotope formed in the coolant. This results in significantly reduced levels of radioactive contamination of the coolant while minimizing the effect of decay power in the coolant from such contaminants; an additional consideration is the desire to eliminate dependence on bismuth which might be a limited resource. This paper provides an overview of the historical development of the LFR, a summary of the advantages and challenges associated with heavy liquid metal coolants, and an update of the current status of development of LFR concepts under consideration. The main characteristics of the SSTAR and ELSY systems are summarized, and the current status of design of each system is presented. Because of the significant recent efforts in the ELSY system design, greater emphasis is placed on the ELSY plant, with focus on the technological development and design provisions intended to overcome or alleviate recognized drawbacks to the use of heavy liquid metal coolants. In the case of the SSTAR system for which development has proceeded more slowly, a more limited summary is provided. It is noted that both systems share many of the same research needs and objectives thus providing a strong basis for international collaboration.

European lead fast reactor—ELSY

The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the EU-FP6-ELSY project. The ELSY (European Lead-cooled System) reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, while fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative Decay Heat Removal (DHR) systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been conceived in order to minimize pressure drops and, as a consequence, to allow decay heat removal by natural circulation. Moreover two redundant, diverse and passive operated DHR systems have been developed and adopted. The paper presents the overall work performed so far.

Investigation of the void reactivity effect in large-size Lead Fast Reactors

An aspect of great relevance in Lead Fast Reactors (LFRs) is the actual void reactivity evaluation. The purpose of this work is double: to inquire into the physical problem, and to evaluate the impact of different approaches and numerical methods on the calculation of the critical reactor parameters involved. Thus, results concerning void effect contributors have been evaluated through a cross-checked analysis performed by means of both a deterministic and a stochastic code. The field of investigation that has been assumed consists in the reference configuration of the 600MWe European Lead-cooled SYstem (ELSY), under development within the 6th and 7th EURATOM Framework Programmes. Calculations have been carried out on a 1500MWth MOX-fuelled core, composed by wrapper-less square Fuel Assemblies (FAs) with pins on a square lattice. The ERANOS (European Reactor Analysis Optimized System) deterministic code ver. 2.1 and the MCNP Monte Carlo code ver. 4c have been employed in conjunction with the JEFF-3.1 nuclear data library to assess the void reactivity variation and its breakdown into the most relevant nuclides, using both the neutron balance equation method and perturbation theory. Results have shown a very good agreement between ERANOS and MCNP outcomes: the huge reactivity worth determined by the core active region voiding (approximately 5000pcm) is due to the predominant contributions of even isotopes – among which 238U plays a major role, being responsible for roughly 4300pcm – as a consequence of their fission cross-section high sensitivity to spectral hardening (threshold reactions), despite their modest contribution to the total fissions.

Preliminary Study on a Plant Control System Structure for a GEN-IV Fast Reactor

AbstractIn this paper, the design of an innovative liquid metal-cooled fast nuclear reactor is considered, with specific reference to the European Lead-cooled SYstem (ELSY) project, complying with the Generation IV International Forum (GIF) requirements. A lumped-parameter dynamic model, based on mass and energy conservation equations, has been developed for the ELSY plant secondary system. This includes, in particular, the lead-water steam generators, which have a main role in the feedback between the primary and secondary system and whose operation has to satisfy, beside common requirements on steam pressure and temperature, additional physical constraints, like a lower limit of the feedwater temperature to avoid lead freezing in any operating condition.Different control strategies have been set up for different operating conditions, such as the normal mode and the alternate mode; these are characterised by different plant subsystem configurations and can be compared to the well-known coordinated mode and turbine-following mode for conventional thermal power plants, respectively.The model and the control loops and feedforward actions have been implemented in the MATLAB-Simulink® environment, with model parameters derived from design data. The control system reliability has been verified by simulating most common operational transients.

Preliminary evaluation of seismic isolation effects in a Generation IV reactor

This paper deals with the treatment of methods for a preliminary safety assessment of next generation nuclear power plant (of Liquid Metal cooled Fast Reactor types) structures. As an example, it was considered the European Lead-cooled System (ELSY) international project response under a reference design basis earthquake (Safe Shutdown Earthquake-SSE) excitation, taking into account also the main isolating device effects in order to increase the reactor safety. This study is intended to analyze the structural effects of a reference SSE by means of an appropriate dynamic FEM code, for possible geometries of both not isolated and isolated ELSY containment building foundation cases. A great attention should be focused too on the arisen hydrodynamic forces and coupling effects between fluid and structures (sloshing phenomenon) that may impair the structures resistance and/or operating capabilities. The obtained numerical results were critically analysed with the intent also to contribute to a step of the safety optimization of the mentioned systems.

Thermal-hydraulics analyses of ELSY lead fast reactor with open square core option

Among the next generation of nuclear reactors, well-known as Gen-IV, the LFR (Lead Fast Reactor) is one of the most promising advanced reactor able to comply the principles of sustainability, economics, safety and proliferation resistance. The ELSY project (European Lead-cooled SYstem), funded by the 6th European Framework Programme, aims at investigating the technical/economical feasibility of a high power lead fast reactor with waste transmutation capability. Several innovative design solutions have been proposed at system level and some of them regard the core region with open-assemblies that represents the reference option. To support the design phase and the safety assessment of ELSY, the thermal-hydraulic code RELAP5, modified to treat heavy liquid metals, has been taken into account. The paper deals with the development of the ELSY thermal-hydraulic and point kinetic model for RELAP5 focusing the attention at core assembly level to verify that the temperatures at nominal power conditions stay within the safety limits both in Beginning-of-Cycle (BOC) and in End-of-Cycle (EOC) conditions. Moreover, in order to have a first evaluation of the system behavior in accidental conditions, an Unprotected Loss-of-Flow Accident (ULOF) simulation at BOC has been analysed and discussed. List of acronyms ADS Accelerator Driven System BOC Beginning-of-Cycle DEC Design Extension Condition EFIT European Facility for Industrial Transmutation ELSY European Lead-cooled SYstem EOC End-of-Cycle FA Fuel Assembly FP6 6th Framework Programme Gen-IV Generation IV (four) GESA Gepulste ElektronenStrahlanlage (Pulsed Electron Beam Facility) HX Heat eXchanger IC Isolation Condenser LFR Lead Fast Reactor LMFR Liquid Metal Fast Reactor PDS-XADS Preliminary Design Studies on eXperimental ADS RVACS Reactor Vessel Air Cooling System ULOF Unprotected Loss-of-Flow W-DHR Water Decay Heat Removal

Inter fuel-assembly thermal-hydraulics for the ELSY square open reactor core design

The lead-cooled reactor is one of the six proposed innovative reactor types by the Generation IV International Forum (GIF). In Europe, the lead-cooled reactor design is known as the European Lead-cooled System (ELSY), which is a 600 MWe medium size fast reactor. The reference design of the ELSY core foresees square open (wrapper-less) fuel-assemblies with a staggered arrangement. In this design, the fuel rods in a fuel-assembly are separated by 3.4 mm. The gap between fuel rods of neighboring fuel-assemblies is 5.5 mm. In other words, the reference gap size between fuel-assemblies is larger than the gap between fuel rods within a fuel-assembly. This article discusses the involved inter fuel-assembly thermal-hydraulics between neighboring fuel-assemblies in the ELSY core. For this purpose as a starting point a validated Reynolds Averaged Navier Stokes (RANS)-based Computational Fluid Dynamics (CFD) approach is adopted. Moreover, bare fuel rods are considered in the present analyses that serve as a step towards inclusion of a spacer grid when its design is fixed. As the next step, the fuel-assemblies are numerically arranged with different gap sizes of 2.1 mm and 3.4 mm in order to analyze the influence of gap size on the inter fuel-assembly thermal-hydraulics. As a final step, analyses on the influence of different power levels of neighboring fuel-assemblies in the ELSY core are presented based on the reference ELSY core design. These inter-fuel assembly thermal hydraulic analyses lead to a conservative Nusselt number correlation for calculating maximum surface temperature of bare fuel rods that are located in the gap region between neighboring fuel-assemblies having different power levels. Such correlations, when implemented, will improve the applicability of system codes.

Preliminary core characterization of a Generation IV lead fast reactor DEMO: Goals, design rationales and options

An Italian effort has been initiated under a cooperation between ENEA and CIRTEN for the investigation of a concept for a small size Generation IV Lead-cooled Fast Reactor (LFR) demonstration project (DEMO) which could provide significant support to the European Lead-cooled System (ELSY), by validating lead technology and the overall system behaviour. A demonstration reactor is expected to prove the viability of technology for use in a future commercial power plant, construction and operation, with the purpose of assessing the general strategy to use, to the largest extent. Thus, the first step towards the DEMO conceptual design has been the specification of the issues of ELSY – assumed as reference LFR – to investigate/validate, the objectives to reach, tools and methodology. As a result, a set of general requirements that must be met have been first identified and further scope analyses have been carried out in order to determine suitable core configurations. Finally, static neutronics analyses have been performed using the ERANOS2.1 deterministic code in conjunction with the JEFF3.1 nuclear data library to accomplish a preliminary core characterization; the mutual compatibility of both objectives and technological constraints has been confirmed a posteriori by thermal–hydraulic estimations. Results have shown that a 250 MWth lead-cooled MOX-fueled DEMO may be a viable concept: in the case of 480 °C reference coolant outlet temperature, 2 m s −1 average coolant velocity and 42 cm active core extension, high neutron flux values (around 2.5 times the ELSY ones) have been obtained (peak flux: 6.3 × 10 15 cm −2 s −1 with corresponding fast fraction of approximately 15.8%, i.e., some 1 × 10 15 cm −2 s −1) thanks to the significant power densities reached.