Advanced Lead Fast Reactor European Demonstrator Research Articles

Thermal-Hydraulic Characterization and Numerical Modeling with RELAP5 Code of ATHENA Secondary Loop

Among the envisaged experimental infrastructures supporting Advanced Lead Fast Reactor European Demonstrator (ALFRED) reactor development, the FALCON [Fostering ALfred CONstruction] consortium identified the Advanced Thermo-Hydraulics Experiment for Nuclear Application (ATHENA) as one of the facilities to address the pool of thermal-hydraulic challenges and demonstrate the feasibility of the revised ALFRED configuration, along with the thermal-hydraulic performances of its main components. ATHENA is a large pool-type lead-cooled multipurpose experimental facility featuring a large-sized vessel (3.2-m diameter and 10-m height), conceived to host almost 800 tons of lead to test ALFRED-relevant scaled components. The test section to be installed in the main vessel includes an electrically heated core simulator, made of seven fuel assemblies, which delivers to the primary coolant a nominal thermal power of 2210 kW; a main coolant pump for lead circulation; and a countercurrent shell-and-tube main heat exchanger, of which the tube bundle is fed by pressurized water by a dedicated secondary circuit. This work presents the numerical model of ATHENA along with thermal-hydraulic characterization of the facility using the system code RELAP5/Mod3.3, properly modified to include the thermophysical properties of heavy liquid metals. After the characterization of the steady state representative of the Stage 3 foreseen for the ALFRED staged approach, results of a numerical sensitivity analysis aimed at defining the most suitable procedure for the shutdown transient of the facility are presented.

On the use of advanced nuclear cogeneration plant integrated into latent heat storage for district heating

There is a growing need to operate nuclear power plants (NPPs) flexibly, especially within the framework of cogeneration applications with high varying heat demand such as in a district heating (DH) system. In this regard, a thermal energy storage system (TESS) can be integrated into the NPP, which should be compatible with the operating conditions of NPP. The TESS considered in this study is capable of simultaneously being charged with constant power from NPP and transferring heat to the DH system with varying demand. For this purpose, thermal storage modules with two separate heat transfer fluids (HTFs) are presented, in which several cylindrical tubes of phase change material (PCM) are vertically located. The 300 MWth Advanced Lead Fast Reactor European Demonstrator (ALFRED) is used as the energy source of the cogeneration system. A comprehensive techno-economic analysis of the ALFRED and TESS has been performed considering several different cogeneration scenarios. Results demonstrate suitable operational flexibility in supplying about 130 MWth heat to the DH system and 95 MWe of electricity to the grid during winter with a total cogeneration efficiency of 75%.

Preliminary evaluation of ALFRED revised concept under station blackout

The Advanced Lead Fast Reactor European Demonstrator (ALFRED) was conceived in the framework of the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project since 2010. Recently, a revised concept of the Reactor Coolant System (RCS) is ongoing to solve some identified issues such as thermal stratification and others. To verify the improvements of the ALFRED design the authors developed a detailed nodalization scheme of the reactor using RELAP5-3D© code. This paper provides a summary of the main improvements introduced in the revised concept of ALFRED reactor and a detailed discussion of the thermal-hydraulic analysis performed with RELAP5-3D©. The nodalization scheme is described in detail together with transient boundary conditions and timing of events. The numerical activity demonstrates the improvements of the revised configuration of the reactor that avoid the establishment of relevant thermal stratification phenomena in both normal and accidental operations. The effect of two relevant parameters is also investigated, highlighting that safety conditions are maintained in the whole spectrum of the sensitivity analysis.

Coarse-mesh thermal-hydraulics and neutronics coupling for the ALFRED reactor

Among the six advanced reactor systems identified by the Generation IV International Forum, the lead fast reactor (LFR) has been considered as one of the most promising future nuclear power plants. Among the LFR designs, Ansaldo Nucleare, as coordinator of the Lead-cooled European Advanced DEmonstration Reactor project, proposes Advanced Lead Fast Reactor European Demonstrator (ALFRED) as LFR demonstrator, which is the study object of the present paper. A multiphysics model for the ALFRED core is developed and presented in this paper, which couples neutronics and thermal hydraulics. The first physics is solved with neutron transport Monte Carlo simulations, while temperatures and the lead density are updated with coarse-mesh-based finite volume method CFD runs. The proposed model focuses on the adoption of spatially non-uniform temperature distributions of materials to compute better on-the-fly estimations of nuclides cross sections and thence a more accurate neutron physics description.

Pre-test analysis of accidental transients for ALFRED SGBT mock-up characterization

In the framework of the Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the CIRCE pool facility has been refurbished by ENEA to host HERO test section (Heavy liquid mEtal pRessurized water cOoled tubes) with the aim to test an innovative concept of the steam generator bayonet tubes proposed for ALFRED (Advanced Lead Fast Reactor European Demonstrator) and to provide experimental data for code validation. HERO consists of a bundle of seven bayonet tubes characterized by an active length of six meters (1:1 with the ALFRED steam generator tube length). The main purposes of the research were to investigate the thermal hydraulic behavior of the innovative concept and provide a set of experimental data aiming at validating STH (System Thermal-Hydraulic) code. The calculations were carried out adopting RELAP5-3D©, updating the nodalization scheme validated with the experimental data of CIRCE-ICE campaign. The experiment consists of a transition from forced to natural circulation in a loss of flow accidental scenario; in order to identify the initial conditions of the experiment, several full power conditions were simulated. Starting from the reference steady state conditions, five transient tests were simulated to evaluate the effect of the reduction of the secondary mass flow rate (reduced from the nominal value to simulate the activation of the DHR system) and the effect of the heat losses compensation. According with the calculation, HERO test section offers excellent thermal-hydraulic behavior ensuring a sufficient natural circulation conditions to remove the decay heat in the short term.

Design and Selection of Innovative Primary Circulation Pumps for GEN-IV Lead Fast Reactors

Although Lead-cooled Fast Reactor (LFR) is not a new concept, it continues to be an example of innovation in the nuclear field. Recently, there has been strong interest in liquid lead (Pb) or liquid lead–bismuth eutectic (LBE) both critical and subcritical systems in a relevant number of Countries, including studies performed in the frame of GENERATION-IV initiative. In this paper, the theoretical and computational findings for three different designs of Primary Circulation Pump (PCP) evolving liquid lead (namely the jet pump, the Archimedean pump and the blade pump) are presented with reference to the ALFRED (Advanced Lead Fast Reactor European Demonstrator) design. The pumps are first analyzed from the theoretical point of view and then modeled with a 3D CFD code. Required design performance of the pumps are approximatively around an effective head of 2 bar with a mass flow rate of 5000 kg/s. Taking into account the geometrical constraints of the reactor and the fluid dynamics characteristics of the molten lead, the maximum design velocity for molten lead fluid flow of 2 m/s may be exceeded giving rise to unacceptable erosion phenomena of the blade or rotating component of the primary pumping system. For this reason a deep investigation of non-conventional axial pumps has been performed. The results presented shows that the design of the jet pump looks like beyond the current technological feasibility while, once the mechanical challenges of the Archimedean (screw) pump and the fluid-dynamic issues of the blade pump will be addressed, both could represent viable solutions as PCP for ALFRED. Particularly, the blade pump shows the best performance in terms of pressure head generated in normal operation conditions as well as pressure drop in locked rotor conditions. Further optimizations (mainly for what the geometrical configuration is concerned) are still necessary.

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Application of Prompt Self-Powered Neutron Detectors to the Lead-Cooled Fast Reactor Demonstrator ALFRED: Validation of the Monte Carlo Model for Selected SPNDs

The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the industrial prototype European lead fast reactor proposed in lead-cooled European advanced demonstration reactor. It has a relatively low power (125 MWe) with a compact design to reduce the cost but maintaining its representativeness and it is cooled by pure lead. One of the open issues is linked to the neutron flux in-core monitoring system because of the harshness of the environment the detectors should be installed in, due to high temperatures, and the neutron-gamma radiation field levels. Monte Carlo simulation is a possible way of facing the problem, reproducing into a virtual world the reactor core, the surrounding environment and radiation interactions. In previous works, neutron spectra and gamma doses at possible detectors' locations in ALFRED were retrieved, with consideration on the applicability of each suitable device currently available. Fission chambers (FCs) were found to be exploited at reactor start-up and intermediate power range. Prompt self-powered neutron detectors (SPNDs) seemed to be the best solution to monitor the reactor full power, becoming the main research target: their effective applicability on field has to be demonstrated. SPND applications do not include reactor control purposes usually. Moreover, their irradiation experience involved thermal and epithermal neutron spectra monitoring, mainly. The lack of data when SPNDs sense fast neutron fluxes in terms of prompt-response pushed the authors to deepen the study in such direction. The work herein shows the mathematical approach based on Monte Carlo simulation of SPNDs by the Monte Carlo N-particle eXtended code (MCNPX), so as to study the capability of the code in reproducing real devices' signals while experimented on field. Such a verification turned out to be the preliminary stage for studying new concepts for SPNDs, in terms of sensitive materials and geometries, envisaging the possibility for designing, prototyping, and testing new devices in suitable fast neutron-flux facilities.

Stand-alone core sensitivity and uncertainty analysis of ALFRED from Monte Carlo simulations

The aim of this paper is the sensitivity and uncertainty analysis of a Lead-Cooled Fast Reactor (LFR) based on Monte Carlo simulation of sizes up to 2000. The methodology developed in this work considers the uncertainty of sensitivities and uncertainty of output variables due to a single-input-variable variation. The Advanced Lead fast Reactor European Demonstrator (ALFRED) is analyzed to determine the behavior of the essential parameters due to effects of mass flow and temperature of liquid lead. The ALFRED core mathematical model developed in this work is fully transient, which takes into account the heat transfer in an annular fuel pellet design, the thermo-fluid in the core, and the neutronic processes, which are modeled with point kinetic with feedback fuel temperature and expansion effects. The sensitivity evaluated in terms of the relative standard deviation (RSD) showed that for 10% change in the core inlet flow, the response in thermal power change is 0.58%, and for 2.5% change in the inlet lead temperature is 1.87%. The regression analysis with mass flow rate as the predictor variable showed statistically valid cubic correlations for neutron flux and linear relationship neutron flux as a function of the lead temperature. No statistically valid correlation was observed for the reactivity as a function of the mass flow rate and for the lead temperature. These correlations are useful for the study, analysis, and design of any LFR.

Design by theoretical and CFD analyses of a multi-blade screw pump evolving liquid lead for a Generation IV LFR

Lead-cooled fast reactor (LFR) has both a long history and a penchant of innovation. With early work related to its use for submarine propulsion dating to the 1950s, Russian scientists pioneered the development of reactors cooled by heavy liquid metals (HLM). More recently, there has been substantial interest in both critical and subcritical reactors cooled by lead (Pb) or lead–bismuth eutectic (LBE), not only in Russia, but also in Europe, Asia, and the USA. The growing knowledge of the thermal-fluid-dynamic properties of these fluids and the choice of the LFR as one of the six reactor types selected by Generation IV International Forum (GIF) for further research and development has fostered the exploration of new geometries and new concepts aimed at optimizing the key components that will be adopted in the Advanced Lead Fast Reactor European Demonstrator (ALFRED), the 300MWt pool-type reactor aimed at proving the feasibility of the design concept adopted for the European Lead-cooled Fast Reactor (ELFR). In this paper, a theoretical and computational analysis is presented of a multi-blade screw pump evolving liquid Lead as primary pump for the adopted reference conceptual design of ALFRED. The pump is at first analyzed at design operating conditions from the theoretical point of view to determine the optimal geometry according to the velocity triangles and then modeled with a 3D CFD code (ANSYS CFX). The choice of a 3D simulation is dictated by the need to perform a detailed spatial simulation taking into account the peculiar geometry of the pump as well as the boundary layers and turbulence effects of the flow, which are typically tri-dimensional. The use of liquid Lead impacts significantly the fluid dynamic design of the pump because of the key requirement to avoid any erosion affects. These effects have a major impact on the performance, reliability and lifespan of the pump. Albeit some erosion-related issues remain to be fully addressed, the results of this analysis show that a multi-blade screw pump could be a viable option for ALFRED from a thermo-fluid-dynamic point of view.

Petri-net based modelling approach for ALFRED reactor operation and control system design

In this work, the Petri-net modelling approach applied to the control system design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is presented, paying particular attention to the startup procedure. The reactor startup is the operational transient in which all the systems of the plant are brought from the cold shutdown condition to the full power mode, close to load-frequency control. In this phase, the several control actions to be taken need to be properly coordinated. To this end, the operational sequence which constitutes the reactor startup procedure has been described by adopting the Petri-nets approach, i.e., a useful formalism for the modelling and the analysis of Discrete Event Systems. Thanks to this quantitative representation, it is possible to easily derive the corresponding control scheme. In addition, the Petri-nets approach has been also exploited for the two-level control system architecture, namely a master system coordinates the operation of the plant by sending suitable signals to the slave system, in which feedback controllers are implemented. As a major outcome of this work, the procedure for the reactor startup and the transition to the full power mode has been simulated in order to assess the control system performance.

Development of the ALFRED reactor full power mode control system

In this paper, the full power mode control scheme for the pool-type Advanced Lead Fast Reactor European Demonstrator (ALFRED) is presented. The design process of a nuclear power plant control system is a multi-step process, whose final result is the implementation of the feedback controllers. In particular, the control system configuration has been finalized after having considered different strategies to govern the variables of interest (e.g., power produced in the core, coolant temperature, pressure, etc.). In a previous work (Ponciroli et al., 2014b), the most relevant issues regarding both the controlled outputs (e.g., the lead temperature in the cold leg) and the control inputs (e.g., the feedwater inlet temperature and the lead mass flow rate) have been allowed for. Based on these preliminary analyses, the control strategy has been further developed, and the full power mode control scheme for the ALFRED reactor has been implemented. As a major outcome, it has been demonstrated how a decentralized control scheme based on a battery of PI controllers allows obtaining satisfactory performance. Indeed, starting from the linearization of the dedicated object-oriented model, the level of decoupling among the control loops has been verified, confirming the validity of the proposed solution based on the outcomes of the Relative Gain Array (RGA) method. This result has been achieved thanks to a proper tuning of the controller parameters, which allows damping the undesired influence due to the other control loops. Finally, as for the design of the actuators, a quantitative approach based on the outcomes of the system linear stability analysis has been proposed, in order to define the maximum safety value for the extraction speed of the control rods.

Control approach to the load frequency regulation of a Generation IV Lead-cooled Fast Reactor

One of the most pressing issues in the study of the power generation and distribution is the characterization of the grid behavior, whether a relevant fraction of the connected power plants relies on Renewable Energy Sources. Indeed, because of the discontinuous power supply and the limited presence of energy accumulators, concerning power imbalances may take place on the grid. The power plants ensuring high reliability performance should be ready to feed the loads when the Renewable Energy Sources are not available. In order to ensure the grid stability and the sustainability of nuclear energy, the possibility of operating Generation-IV nuclear reactors in a flexible way should be considered, i.e., the Nuclear Power Plants should adjust the mechanical power produced so as to comply with the sudden grid frequency variations. In the present work, this opportunity is assessed for the Lead-cooled Fast Reactors, adopting the Advanced Lead Fast Reactor European Demonstrator (ALFRED) as a representative of Lead-cooled Fast Reactor technology. For this reactor concept, because of the large thermal inertia that characterizes the system, the adoption of the “reactor-follows-turbine” scheme (currently employed in the Pressurized Water Reactors) is not feasible. An alternative solution is proposed, i.e., the set-point for the thermal power produced in the core is kept constant at the nominal value (or slowly variable), and the set-point for the mechanical power available to the alternator is adjusted according to the load demands. In order to assess the performance of the developed control scheme, two case studies are simulated. In the first one, a frequency profile of the synchronous Continental European grid is provided. In the second simulation, the possibility of performing the islanding operational transient is evaluated. Such an operational transient needs the connected power plants to perform relevant power variations in few seconds in order to deal with the abrupt change in the grid powered configuration. The simulation outcomes show that the proposed control scheme allows achieving prompt mechanical power variations. As for the quantitative results, in both simulations the frequency value is maintained below the upper safety threshold (50.8Hz). On the other hand, the control scheme complies with the technological constraints set for the safe operation of the primary circuit of the ALFRED reactor. In particular, the Balance of Plant pressure-induced feedbacks on the primary circuit dynamics are effectively dampened, i.e., the temperature in the cold leg is effectively maintained close to its set-point (i.e., 400°C) and the plant is almost insensitive to the variations of the grid power demands.

A passive decay heat removal system for the lead cooled fast reactor demonstrator “Alfred”

The generation IV lead cooled fast reactors are of particular interest for the Italian research. At present, one significant European project in progress is LEADER (Lead cooled European Advanced DEmonstrator Reactor) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, ALFRED (Advanced Lead Fast Reactor European Demonstrator). The demonstrator provides technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. In a previous paper, the Authors proposed the EBBSG (External Boiling Bayonet Steam Generator) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. In the present paper, the Authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the system is investigated through a Matlab–Simulink model. The results are satisfactory since, according to simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

Numerical investigation on a jet pump evolving liquid lead for GEN-IV reactors

Heavy-liquid metals, such as lead and lead–bismuth eutectic, are promising candidates as coolant for advanced GEN-IV fast reactors as well as for Accelerator-Driven Systems. The advancing knowledge of the thermal-hydraulic behavior of these fluids leads to explore new geometries and new concepts aimed at optimizing the key components of a GEN-IV reactor for these fluids. In this paper, a theoretical and computational analysis is presented of a jet pump evolving liquid lead as primary pump for ALFRED (Advanced Lead Fast Reactor European Demonstrator). The jet pump is modeled with a 3D CFD code (FLUENT) and at design operating conditions. The analysis shows that a jet pump could be a viable solution for ALFRED (at least from a thermal-hydraulic point of view), albeit some technological issues remain to be fully addressed.

Simulation Model of a Passive Decay Heat Removal System for Lead-Cooled Fast Reactors

The generation IV lead cooled fast reactors are of particular interest for the Italian research: several influential companies (Ansaldo Nucleare, ENEA) are involved in these important European R&D projects. At present, one significant European project in progress is lead cooled European advanced demonstrator reactor (LEADER) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, advanced lead fast reactor European demonstrator (ALFRED). The demonstrator has to include technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. The authors have addressed the issue of bayonet-tube steam generators proposing the external boiling bayonet steam generator (EBBSG) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. This is possible thanks to an external feed-water boiling, based on the known Loeffler scheme, coupled to the bayonet tube concept. In the present paper, the authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the proposed DHR system is investigated through a Matlab-Simulink model. The results are satisfactory since, according to the simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

Application of the TRANSURANUS code for the fuel pin design process of the ALFRED reactor

The Advanced Lead Fast Reactor European Demonstrator (ALFRED) is the 300 MWth pool-type reactor aimed at proving the feasibility of the design concept adopted for the European Lead-cooled Fast Reactor (ELFR) of Generation-IV, whose preliminary design has been proposed within the LEADER (Lead Cooled European Advanced Demonstration Reactor) EURATOM Project.In the frame of investigating different options to optimize the conceptual fuel pin design of ALFRED, the fuel performance analysis of the reactor core is carried out by means of the TRANSURANUS code, which is presented in this paper. Results of the average and the hottest reactor conditions are discussed concerning both fuel and cladding performance on the basis of indicative design limits. A sensitivity analysis is performed in order to assess the impact of models and parameters affected by a larger uncertainty, and to identify a “worst case” scenario. This strategy allows identifying some design parameters which could be modified to improve the fuel pin behaviour, in order to gain benefits mainly in terms of maximum fuel temperature. The results of the present work are useful for giving feedback to the conceptual design of the ALFRED reactor and, in perspective, to improve the safety-by-design characteristics of the LFR systems.

A preliminary approach to the ALFRED reactor control strategy

In this paper, a preliminary approach to the definition of a suitable control strategy for the Advanced Lead Fast Reactor European Demonstrator (ALFRED), developed within the European 7th Framework Program, has been undertaken. The Generation IV reactors offer new challenges for what concerns the nuclear power plant control since several constraints both on primary and secondary loops have to be faced, differently from the conventional Light Water Reactors. A simulator of the ALFRED plant has been developed in a previous work (Ponciroli et al., 2014) with the main purpose of studying the system free dynamics and stability features in a control-oriented perspective. Based on the outcomes of these investigations, in the present work, the possibility of adopting decentralized control schemes has been investigated. Accordingly, Single Input Single Output control laws have been applied directly to the selected couples of input–output variables, which have been identified first on the basis of the preliminary plant dynamics analyses, and then confirmed by the indications of the Relative Gain Array method. Afterwards, two different control schemes have been studied depending on the number of available inputs, and then implemented and compared in order to evaluate the effect of each control action on the associated potential control strategy effectiveness. As a last step, the ALFRED control system has been finalized. The regulator design has been set up based on a simultaneous feedforward-feedback scheme incorporating four closed feedback loops. A controlled power reduction and a controlled overpower transient have been simulated in order to assess the performance of the two proposed control schemes. Results show that both the adopted control strategies can assure an efficient control of the thermal power while guaranteeing an effective control of lead and steam temperatures as well. In addition, some non-negligible differences between the two schemes have been observed and discussed in the simulation results of control and controlled variables.

Stability analyses for the European LFR demonstrator

The development of next-generation Lead-cooled Fast Reactors (LFRs) up to industrial deployment requires the realization of a demonstration plant intended to prove the viability of technology as well as the overall system behavior. The conceptual design of a low-cost scaled-down LFR named ALFRED (Advanced Lead Fast Reactor European Demonstrator) has been undertaken within the European Commission's 7th Framework Programme's LEADER (Lead-cooled European Advanced Demonstration Reactor) project. A simulation tool has been developed in MATLAB® based on a lumped-parameter approach, aimed at investigating the reactor stability and identifying potential regions of instability through relatively straightforward analyses. As a major outcome, ALFRED has turned out to be inherently stable on the entire power range independently of both the fuel burn-up and the value of the coolant density coefficient, which should reach unrealistic high values (nearly 6pcmK−1) to make the reactor unstable. The results of this study have provided the reactor designers with important feedbacks leading to the finalization of the primary system configuration.