Lead Fast Reactor Research Articles

Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

During the GUINEVERE FP6 European project (2006–2011), the zero-power VENUS water-moderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al2 O3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED

Among GEN IV projects for future nuclear power plants, lead-cooled fast reactors (LFRs) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant safety, and waste management. The novelty of this matter causes some open issues about coolant chemical aspects, structural aspects, monitoring instrumentation, etc. Particularly, hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e., source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal hydraulics and chemical conditions for the LFR core environment will be assumed, as the neutron flux will be studied extensively by the Monte Carlo transport code MCNPX (Monte Carlo N-Particles X-version). The core coolant’s high temperature drastically reduces the candidate instrumentation because only some kinds of fission chambers and self-powered neutron detectors can be operated in such an environment. This work aims at evaluating the capabilities of the available instrumentation (usually designed and tailored for sodium-cooled fast reactors) when exposed to the neutron spectrum derived from the Advanced Lead Fast Reactor European Demonstrator, a pool-type LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such a class of instrumentation does follow the power evolution, but is not completely suitable to detect the whole range of reactor power, due to excessive burnup, damages, or gamma interferences. Some improvements are possible to increase the signal-to-noise ratio by optimizing each instrument in the range of reactor power, so to get the best solution. The design of some new detectors is proposed here together with a possible approach for prototyping and testing them by a fast reactor.

Selecting benchmarks for reactor simulations: An application to a lead fast reactor

For several decades reactor design has been supported by computer codes for the investigation of reactor behavior under both steady state and transient conditions. The use of computer codes to simulate reactor behavior enables the investigation of various safety scenarios saving time and cost. There has been an increase in the development of in-house (local) codes by various research groups in recent times for preliminary design of specific or targeted nuclear reactor applications. These codes must be validated and calibrated against experimental benchmark data with their evolution and improvements. Given the large number of benchmarks available, selecting these benchmarks for reactor calculations and validation of simulation codes for specific or target applications can be rather tedious and difficult. In the past, the traditional approach based on expert judgement using information provided in various handbooks, has been used for the selection of these benchmarks. This approach has been criticized because it introduces a user bias into the selection process. This paper presents a method for selecting these benchmarks for reactor calculations for specific reactor applications based on the Total Monte Carlo (TMC) method. First, nuclear model parameters are randomly sampled within a given probability distribution and a large set of random nuclear data files are produced using the TALYS code system. These files are processed and used to analyze a target reactor system and a set of criticality benchmarks. Similarity between the target reactor system and one or several benchmarks is quantified using a similarity index. The method has been applied to the European Lead Cooled Reactor (ELECTRA) and a set of plutonium and lead sensitive criticality benchmarks using the effective multiplication factor (keff). From the study, strong similarity were observed in the keff between ELECTRA and some plutonium and lead sensitive criticality benchmarks. Also, for validation purposes, simulation results for a list of selected criticality benchmarks simulated with the MCNPX and SERPENT codes using different nuclear data libraries have been compared with experimentally measured benchmark keff values.

Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented.The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

The Lead Fast Reactor: An Opportunity for the Future?

As one of the six reactor concepts selected in the Generation IV Technology Roadmap, the lead fast reactor (LFR) has become very attractive in recent years, and its development is pursued in several countries [1,2]. The attractiveness of the LFR is justified by the intrinsic characteristics of the system, which are able to completely satisfy the goals of the Generation IV International Forum (GIF). The system’s enhanced safety is a particularly important characteristic, as safety is becoming one of the key criteria in the choice of a next-generation nuclear system. This paper briefly outlines the various on-going initiatives dedicated to the development of lead technology. The use of lead as a coolant permits a complete change of the design approach and provides several possibilities for innovation that will be briefly described here.

Helical coil thermal-hydraulic model for supercritical lead cooled fast reactor steam generators

This paper deals with the presentation of the developments of an in-house code created for the steady state thermal-hydraulic analysis of helical coil steam generators. These components are of particular interest in the nuclear field both for normal and emergency operation because they are able to remove thermal power with a very high degree of compactness. This quality is particularly attractive for small modular reactors where constraints related to the volume of components are a critical issue. The model is able to perform calculations to validate the design of these components, run fast sensitivity analyses and show the spatial behaviour of the most important parameters of the flow such as temperature, pressure, heat transfer coefficients and heat flux. In this paper the additional features of the model such as the possibility to simulate supercritical water and liquid metals are presented. The choices made for the selection of semi-empirical correlations available in literature for the estimation of the heat transfer coefficients in these conditions are described and justified. Finally, the predictive capability of the updated model is validated against the nominal data of the BREST reactor, which is a lead fast reactor that make use of supercritical water.

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.

Supercritical Carbon Dioxide Applications for Energy Conversion Systems

In the present paper, the possibility of increasing the thermodynamic efficiency of an electric energy production plant, by using an advanced energy conversion system based on supercritical carbon dioxide (S-CO2) as working fluid, has been analyzed. Since the supercritical carbon dioxide cycles are being considered as a favorable candidate for the next generation of nuclear power plant energy conversion systems, a lead cooled fast reactor has been selected as reference in the present analyses. The main aim of the present study is to compare two different S-CO2 thermal cycles applied on the conversion system of a nuclear power plant. The reference Lead cooled Fast Reactor (LFR) used for the present analyses is the ALFRED reactor, which has a thermal power of 300 MW and it is considered the scaled down prototype of the industrial European Lead Fast Reactor (ELFR).Thermodynamic cycles selected for the present study are a Recompression Cycle and a Brayton Cycle with Regeneration. Each of them has been analyzed under several design conditions regarding the maximum pressure and the regeneration coefficient. Among different design conditions, the solution allowing the maximization of the overall efficiency has been identified. Thermodynamic analyses have been carried out with GateCycle™ v. 6.1.1, which is a General Electric software able to predict design and off-design performance of power plants.

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.

A multi-physics reduced order model for the analysis of Lead Fast Reactor single channel

In this work, a Reduced Basis method, with basis functions sampled by a Proper Orthogonal Decomposition technique, has been employed to develop a reduced order model of a multi-physics parametrized Lead-cooled Fast Reactor single-channel. Being the first time that a reduced order model is developed in this context, the work focused on a methodological approach and the coupling between the neutronics and the heat transfer, where the thermal feedbacks on neutronics are explicitly taken into account, in time-invariant settings. In order to address the potential of such approach, two different kinds of varying parameters have been considered, namely one related to a geometric quantity (i.e., the inner radius of the fuel pellet) and one related to a physical quantity (i.e., the inlet lead velocity). The capabilities of the presented reduced order model (ROM) have been tested and compared with a high-fidelity finite element model (upon which the ROM has been constructed) on different aspects. In particular, the comparison focused on the system reactivity prediction (with and without thermal feedbacks on neutronics), the neutron flux and temperature field reconstruction, and on the computational time. The outcomes provided by the reduced order model are in good agreement with the high-fidelity finite element ones, and a computational speed-up of at least three orders of magnitude is achieved as well.

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.

Addressing the heavy liquid metal – Water interaction issue in LBE system

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.

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.

An updated core design for the multi-purpose irradiation facility MYRRHA

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multi-purpose research facility currently being developed at SCK•CEN. It will be able to work in both critical and sub-critical modes and, cooled by lead-bismuth eutectic. It will play a key role in the development of the Pb-alloy technology needed for the lead fast reactor GEN IV concept. MYRRHA will demonstrate the Accelerator Driven System (ADS) full concept by coupling a proton accelerator, a spallation target and a sub-critical reactor at a reasonable power level to allow operation feedback. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Recently, a new core design with a longer active core has been proposed. This paper presents the neutronic analyses for this design improvement. The analyses have been done using the MCNP/X code and the in-house developed ALEPH2 depletion code.

RELAP5 and SIMMER-III code assessment on CIRCE decay heat removal experiments

In the frame of THINS Project of the 7th Framework EU Program on Nuclear Fission Safety, some experiments were carried out on the large scale LBE-cooled CIRCE facility at the ENEA/Brasimone Research Center to investigate relevant safety aspects associated with the removal of decay heat through heat exchangers (HXs) immersed in the primary circuit of a pool-type lead fast reactor (LFR), under loss of heat sink (LOHS) accidental conditions. The start-up and operation of this decay heat removal (DHR) system relies on natural convection on the primary side and then might be affected by coolant mixing and temperature stratification phenomena occurring in the LBE pool.The main objectives of the CIRCE experimental campaign were to verify the behavior of the DHR system under representative accidental conditions and provide a valuable database for the assessment of both CFD and system codes. The reproduced accidental conditions refer to a station blackout scenario, namely a protected LOHS and loss of flow (LOF) transient.In this paper the results of 1D RELAP5 and 2D SIMMER-III simulations are compared with the experimental data of more representative DHR transients T-4 and T-5 in order to verify the capability of these codes to reproduce both forced and natural convection conditions observed in the primary circuit and the right operation of the DHR system for decay heat removal.Both codes are able to reproduce the stationary conditions and with some uncertainties the transition to natural convection conditions until the end of the transient phase. The trend of temperature stratification in the LBE pool during both forced and natural convection conditions is rather well reproduced by the codes. The largest uncertainties of the simulation concern the DHR heat exchanger behavior, which showed large instabilities in its operation during the tests.

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.

SEISMIC ISOLATION OF LEAD-COOLED REACTORS: THE EUROPEAN PROJECT SILER

SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the 7th Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the 6th Framework Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the nonisolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.

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.