Multi-purpose hYbrid Research Reactor For High-tech Applications Research Articles

Preliminary studies of the MINERVA cryogenic supply system

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) will be composed of a 600 MeV LINear ACcelerator (LINAC) with a large number of cryomodules and a 100 MW thermal power subcritical nuclear reactor cooled by lead-bismuth. The MINERVA MYRRHA Isotopes productioN coupling the linEar acceleRator to the Versatile proton target fAcility) project is the first phase of the MYRRHA project composed of a 100 MeV, 4 mA continuous-wave proton LINAC. The MINERVA LINAC takes advantage of 60 superconducting radiofrequency (RF) cavities cooled at 2 K by saturated superfluid helium (He II). The MINERVA cryoplant will supply the required cryogenic cooling power of the LINAC through cryogenic lines and cryogenic valve boxes. This paper describes the LINAC cryogenic system supply architecture, the different operating modes, the heat loads, the cryoplant main requirements and the strategy to comply with high reliability requirements. The MINERVA cryoplant shall provide helium flows at 40 K for the thermal shielding, at 5 K for the cooling of the RF couplers and at 2 K for the 60 superconducting cavities installed in 30 cryomodules. The total equivalent refrigeration power of the MINERVA cryoplant is about 3.5 kW at 4.5 K. The cryoplant is composed of a warm compression station including sub-atmospheric volumetric compressors, helium gas storages and a refrigeration cold box with cold centrifugal compressors.

Design and Assessment of the New CIRCE-THETIS Facility for the Development of Heavy Liquid Metal Cooled Fast Reactors

Within the roadmap for the technological development of Generation IV reactors, the HORIZON2020 European Union–funded Partitioning And Transmuter Research Initiative in a Collaborative Innovation Action (PATRICIA) project was launched to support innovative solutions for the development of the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) accelerator-driven system concept and lead fast reactor technologies in general. The ENEA contributes to the project by involving the experimental infrastructures of the Brasimone Research Center (Italy). In particular, a large-scale pool-type facility named CIRColazione Eutettico (CIRCE), using lead-bismuth eutectic as the primary coolant and pressurized water as the secondary fluid, is under refurbishment with the implementation of a novel test section (TS) named Thermal-hydraulic HElical Tubes Innovative System (THETIS) to be installed in the CIRCE main vessel. The new TS will include a vertical mechanical pump for primary coolant circulation and a new prototypical helical coil steam generator (HCSG). This steam generator concept turns out to be very promising for nuclear power plants since the helical geometry is very compact and it assures high power removed, taking up a minimum amount of space. Accordingly, with the aims of the project, the experimental tests in CIRCE-THETIS will focus on (1) investigating the thermal-hydraulic behavior of the system in steady-state operation (forced circulation regime) during operational and accidental transients (postulated scenarios) and in a natural circulation regime considering as heat sink the HCSG (acting as a decay heat removal system) and the reactor vessel auxiliary cooling system in stand-alone or coupled operation and (2) characterizing the performance of the HCSG. The present work presents the layout of the CIRCE-THETIS facility at the end of the final design phase, describing in detail the main components of the TS, along with the instrumentation installed. Focus will be given to the HCSG mock-up, for which pretest analyses using the system thermal-hydraulic code RELAP5/Mod3.3 and a computational fluid dynamics code have been carried out to support the design of the component and to evaluate its thermal-hydraulic performance under the operative conditions foreseen during the experiments.

Liquid-Metal Thermal-Hydraulic Numerical Analyses in Support of the Upcoming CIRCE-THETIS Experimental Campaign

Liquid metal cooled reactors are among the design proposals accepted by the Generation IV International Forum for the fourth generation of nuclear power plants. During the last decade, many European Union (EU) projects started with the goal to pave the way for the development of this type of reactor. The present research work is performed in the framework of the EU Partitioning And Transmuter Research Initiative in a Collaborative Innovation Action (PATRICIA) project, supporting the development of the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA). Among the cornerstones of the project are experimental and numerical analyses involving the CIRColazione Eutettico (CIRCE) facility set at the ENEA Brasimone Research Centre. The upcoming experimental campaign will address the new CIRCE-THETIS configuration, including a new kind of heat exchanger named THETIS: a helical coil steam generator (SG) whose heat transfer capabilities and impact on pool thermal hydraulics are to be investigated. The main goal of the present work is to provide numerical support to experimentalists to help them set up the CIRCE-THETIS experimental facility, the operating conditions test matrix, and postulated transients, providing information about potential needed updates or dangerous conditions. This paper reports on pretest computational fluid dynamics and system thermal-hydraulic analyses performed at the University of Pisa in the frame of the PATRICIA project. The new SG was first addressed assessing its heat transfer capabilities. Sensitivity analyses were performed; among the outcomes, the excellent heat transfer capabilities of the SG were highlighted. The pool and the reactor vessel auxiliary cooling system component were later investigated reporting valuable information concerning both postulated steady-state and transient conditions. The performed analyses had a relevant impact on the design of the facility, suggesting updates based on numerical calculations. After the experimental campaign, posttest analyses will be performed to draw lessons from the observed phenomena. This will also provide room for improvement of the adopted numerical tools.

Combined Large-Eddy and Direct Numerical Simulations of a Planar Jet with Heated Co-Flow with Medium and Low Prandtl Fluids

In the present work, we have applied a combined dynamic large-eddy simulation (LES) and direct numerical simulation (DNS) approach for a three-dimensional planar jet in a turbulent forced convection regime (Re = 18000) with a heated co-flow. Results from LES are compared with Reynolds Averaged Navier-Stokes (RANS) simulations and experimental data. We have analyzed flow and heat transfer features for four values of the characteristic Prandtl numbers (Pr = 0.71, 0.2, 0.025, and 0.006), which are representatives of air, He-Xe gas mixture, Lead-Bismuth Eutectic (LBE), and sodium, respectively. The latter two low-Prandtl fluids have been considered because of their role as primary coolants in advanced fast pool-type reactor prototypes (such as the Multi-purpose Hybrid Research Reactor for High-tech Applications (MYRRHA) at SCK•CEN, Belgium). We have provided detailed insights into instantaneous and long-term time-averaged behavior of the velocity and temperature fields (the first- and second-order moments). Furthermore, we have analyzed profiles of characteristic velocity and temperature time scales and dissipation rates, as well as the power spectra of the streamwise velocity component and temperature at several characteristic locations. The mean temperature profiles demonstrated rather low sensitivity for various values of the Prandtl number. In contrast, profiles of the temperature standard deviation exhibited larger variations, decreasing in magnitude with lower Prandtl values. Here presented results of the high fidelity numerical simulations (dynamic LES/DNS) for the low-Prandtl working fluids can be used for further development, testing, and validation of the advanced RANS-type turbulence models.

A multi-stage approach of simulating turbulence-induced vibrations in a wire-wrapped tube bundle for fretting wear prediction

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a prototype of a Generation IV reactor that will be constructed in Mol, Belgium. It is a liquid metal fast reactor, using lead-bismuth eutectic (LBE) as a coolant. Flow-induced vibrations resulting in fretting wear could be of concern due to the high density of LBE. Additionally, a wire is wrapped around each fuel rod to preserve their mutual distance in the hexagonal array, which affects the flow pattern.In this research turbulence-induced vibrations are investigated. Large-eddy simulation is a sound technique to acquire turbulent loads, but it is computationally demanding to combine this in a two-way coupled fluid–structure interaction (FSI) simulation. However, because accurate prediction of the vibration of cylinders subject to liquid flow typically requires a two-way simulation due to the added-mass effect, a methodology consisting of multiple steps is proposed. First, the modal characteristics of the pin are determined in a fully coupled FSI simulation of a bare bundle. The material properties used in this model are modified to take into account the effect of the wire, which was determined experimentally beforehand. The resulting characteristics are then used to construct a realistic structural model, employed in a one-way simulation with turbulent loads from a large-eddy simulation of a wire-wrapped bundle.In this structural model contact with neighboring rods and the hexagonal duct is detected, providing realistic boundary conditions to the rod and allowing to extract contact forces. These forces are further post-processed together with the sliding displacement to calculate work rates, which is an important parameter for fretting wear prediction. Two pins of a 19-pin bundle were investigated, the central pin and a pin located in one of the corners. The simulation is repeated for multiple friction coefficients, and its impeding effect of friction on sliding motions was confirmed to moderate fretting.

Microstructural investigation and identification of intermetallic σ-phase in solution annealed 316L-type austenitic stainless steel

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is an accelerator driven system, currently under development at SCK CEN in Mol, Belgium. This nuclear system will use liquid Lead-Bismuth eutectic alloy as a spallation target for fast neutron production and as coolant. The ideal structural material for a liquid metal cooled reactor should be unsusceptible to both liquid metal embrittlement and liquid metal corrosion, while possessing high toughness. Nuclear grade austenitic stainless steels similar to AISI 316L have therefore been chosen as the main candidate structural materials for MYRRHA. In the framework of the qualification of those candidates, a specific batch of this material has shown unexpectedly poor mechanical properties, which triggered the need of in-depth microstructural analysis. The behaviour was attributed to the unexpected and undesired presence of intermetallic σ-phase. The σ-phase was identified with a high confidence by combining the data for chemical composition from energy-dispersive X-ray spectroscopy and crystallographic information from electron backscatter diffraction by comparing simulated Kikuchi diffraction patterns with experimentally recorded ones. At first sight the optical appearance of σ-phase resembles δ-ferrite islands, which results in the risk of overlooking this when only classical material qualification methods are used. When left undetected, testing this material including the brittle σ-phase in a liquid metal environment, in combination with miniature mechanical test specimens, could lead to misinterpretation of embrittlement of the austenitic matrix.

Numerical Analysis of the CIRCE-HERO PLOFA Scenarios

The present work deals with simulations carried out at the University of Pisa by using the System Thermal Hydraulics code RELAP5/Mod3.3 to support the experimental campaign conducted at the ENEA (Energia Nucleare ed Energie Alternative) Brasimone Research Centre on the CIRColazione Eutettico—Heavy liquid mEtal pRessurized water cOoled tubes (CIRCE-HERO) facility. CIRCE is an integral effect pool type facility dedicated to the study of innovative nuclear systems and cooled by heavy liquid metal, while HERO is a heat exchanger heavy liquid metal/ pressurized cooling water system hosted inside the CIRCE facility. Beside the H2020 project Multi-Purpose Hybrid Research Reactor for High-Tech Applications (MYRRHA) Research and Transmutation Endeavour (MYRTE), a series of experiments were performed with the CIRCE-HERO facility, for both nominal steady-state settings and accidental scenarios. In this framework, the RELAP5/Mod3.3 code was used to simulate the experimental tests assessing the heat losses of the facility and analyzing the thermal hydraulics phenomena occurring during the postulated Protected Loss Of Flow Accident (PLOFA). The modified version Mod. 3.3 of the source code RELAP5 was developed by the University of Pisa to include the updated thermo–physical properties and convective heat transfer correlations suitable for heavy liquid metals. After reproducing the facility through an accurate nodalization, boundary conditions were applied according to the experiments. Then, the PLOFA scenarios were reproduced by implementing the information obtained by the experimental campaign. Sensitivity analyses of the main parameters affecting the thermofluidynamics of the Lead-Bismuth Eutectic (LBE) were carried out. In the simulated scenario, the LBE mass flow rate strongly depends on the injected argon flow time trend. The numerical results are in agreement with the experimental data, however further investigations are planned to analyze the complex phenomena involved.

Numerical and experimental analysis of a planar jet with heated co-flow at medium and low Prandtl-number values

In the present work, we combine experiments and numerical simulations of a planar jet with heated co-flow with medium (air) and low-Prandtl (He-Xe gas mixture) fluids. Jets are recognized as representative test cases to be investigated in large components of pool-type liquid metal-cooled nuclear systems, like the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA), currently under design at SCK•CEN. The present planar jet configuration mimics a closed wind tunnel that is designed and operated at VKI to generate an experimental database for velocity and temperature fields of a turbulent forced-convection flow regime. The performed experiments combine the Particle Imaging Velocimetry (PIV) (in characteristic planes) and thermocouple (single point) measurements. In parallel with experiments, comprehensive numerical simulations have been performed within the RANS modeling framework. Next to the standard eddy-viscosity based two-equation k-ε model, an extended variant based on the low-Reynolds elliptic relaxation concept (so-called ζ-f model) has been applied too. To investigate the low-Prandtl effects on the heat transfer, series of the turbulent heat transfer models have been applied, ranging from a conventional constant turbulent Prandtl number to a more elaborate kθ-εθ model. The combination of the low-Reynolds ζ-f and kθ-εθ models was explored for the first time in the content of nuclear engineering applications. The focus of the numerical studies is to address in details the effects of low-Prandtl fluid in the strongly forced convection flow (central planar cold jet) in presence of a strong shear (hot co-flow). We demonstrate the importance of the proper specification of the inlet boundary conditions in numerical simulations to mimic correctly experimentally observed asymmetrical distributions of the cross-wise profiles of stream-wise velocity, turbulent kinetic energy and temperature. Finally, the minor differences in results between the assumed constant turbulent Prandtl number and more advanced kθ-εθ model of the turbulent heat flux confirmed the overly dominant mechanisms of the strong convection and molecular diffusion in the present configuration.

Nuclear data analyses for improving the safety of advanced lead-cooled reactors

A target accuracy assessment of the effective neutron multiplication factor, keff, for MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) lead-bismuth cooled fast reactor has been performed with JEFF-3.3 and ENDF/B-VIII.0 state-of-the-art nuclear data libraries and the SUMMON system. Uncertainties in keff due to uncertainties in nuclear data have been assessed against the target accuracies provided by SG-26 of the WPEC of OECD/NEA in 2008 for LFR. Results show that keff target accuracy is still exceeded by more than a factor of two using the latest nuclear data evaluations released in 2018. Consequently, nuclear data assimilation has been carried out using criticality experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of MYRRHA. The results from this work show that the level of accuracy needed in nuclear data cannot be obtained using only differential experiments, but the combination of experimental covariance data and integral experiments together with Generalised Least Squares technique can provide adjusted nuclear data capable of predicting reactor properties with lower uncertainty and consistent with differential data.

Coupled system thermal-hydraulic/CFD analysis of a protected loss of flow transient in the MYRRHA reactor

Within the MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, the Belgian Nuclear Research Centre SCK•CEN is developing and designing a flexible irradiation facility, configured as an accelerator driven system (ADS) with a MOX fueled reactor core, able to operate in both critical and sub-critical modes. The system design features a compact pool-type primary cooling system operating with molten Lead-Bismuth Eutectic (LBE). With regard to the thermal-hydraulic design and safety assessment of the installation, a major challenge is represented by the complex coolant flow field characterizing the large open regions of the primary vessel, namely the cold and hot plenum, with the presence of pronounced three-dimensional phenomena that may impact the evolution of accidental transients such as loss of flow events. In order to have a realistic representation of such effects, a coupled system thermal-hydraulic/CFD model of the MYRRHA reactor is developed and presented in this paper. The proposed multi-scale methodology, which couples the 1D system code RELAP5-3D to the CFD code FLUENT, is based on domain decomposition and a novel implicit numerical scheme is developed. The coupled reactor-scale computational model is applied in this work to the analysis of a postulated protected loss of flow (PLOF) accident, and preliminary validated against RELAP5-3D stand-alone solution data. The results of the analysis were found in agreement, demonstrating the capability of the tool to perform integral simulation taking into account 3D flows and local phenomena.

Neutron-induced nuclear data for the MYRRHA fast spectrum facility

The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) concept is a flexible experimental lead-bismuth cooled and mixed-oxide (MOX) fueled fast spectrum facility designed to operate both in sub-critical (accelerator driven) and critical modes. One of the key issues for the safe operation of the reactor is the uncertainty assessment during the design works. The main objective of the European project CHANDA (solving CHAllenges in Nuclear DAta) Work Package 10 is to improve MYRRHA relevant nuclear data in order to reduce the reactor parameter uncertainties derived from them. In order to achieve this goal, several tasks have been undertaken. First, a sensitivity study of MYRRHA integral parameters, such as energy dependent cross sections, fission spectra and neutron multiplicities, to nuclear data has been conducted resulting in a list of MYRRHA relevant quantities (nuclides and reactions). On the second task, an analysis of the existing experimental data and evaluations for the quantities included in the list has been carried out. In this framework, the impact on the multiplication factor of quantities from different nuclear data libraries for different nuclides, reactions and energy regions has been investigated on the MYRRHA MOX critical core model. As the next step, new experiments and evaluations will be performed in order to improve existing nuclear data libraries.

A preliminary stability analysis of MYRRHA Primary Heat Exchanger two-phase tube bundle

The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, started at SCK·CEN since 1999, aims at the construction of a pool-type sub-critical Accelerator-Driven System (ADS) which could also operate as a critical reactor. The primary system, enclosed in the primary vessel, is filled with Lead Bismuth Eutectic (LBE) which acts as primary coolant. The power is then delivered through four heat exchangers to four secondary loops. The secondary cooling fluid is two-phase water operating at relatively low pressure (16bar). Four aero-condensers act as heat sinks, since MYRRHA design does not foresee any electricity generation.The MYRRHA Primary Heat eXchangers (PHXs) cover a role of fundamental importance in normal operation and accidental conditions, being part of the primary and secondary cooling system and of the Decay Heat Removal (DHR) system. It is thus highly relevant to understand the PHXs behavior under all the potential working conditions.In particular, the stability of the PHXs must be guaranteed under all operating conditions.System code models play an important role in understanding and predicting the behavior of the reactor in all conditions, from steady state to operational and accidental transients, and simulating all the postulated scenarios.A solid PHX design requires a complete assessment of two-phase flow instabilities in the secondary system water tube bundle and the potential implementation of a suitable stabilizing device (orifice) to reduce the impact of the perturbations along the channel. The stability assessment should take in consideration all the possible reactor operational power levels in order to prove the stable behavior under all operational conditions.The tube bundle stability assessment has been carried out by following a similar procedure used for BWR fuel channels, through a specific RELAP5-3D model representing the PHX and able to evaluate the propagation of a density wave in the tube length.A series of suitable boundary conditions, on both primary and secondary side, and perturbation triggers have been foreseen into the model, so to discover all kind of unstable behavior and to dimension the needed orifice to guarantee the flow stability in all operating conditions.The PHX stability analysis is initially performed on the original tube bundle without the adoption of any stabilizing devices, in order to check the natural behavior of the system. The possible adoption and design of an orifice is then conducted on the basis of this preliminary study. The system response against the various types of instabilities, before the introduction of an orifice, is not completely satisfactory: a stable flow is found within certain specific system parameters ranges. After the introduction of a suitable orifice, the system behavior becomes stable under all operating conditions against all types of two-phase flow instabilities.

Comparison of MYRRHA RELAP5 MOD 3.3 and RELAP5-3D Models on Steady State and PLOF Transient

The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project aims at the construction of a pool-type subcritical accelerator-driven system that could also operate as a critical reactor. The primary system, enclosed in the primary vessel, is filled with lead-bismuth eutectic (LBE). The secondary cooling fluid is two-phase water operating at relatively low pressure (16 bars). Four aerocondensers act as heat sinks.The code version used for the development of the MYRRHA models is RELAP5 MOD 3.3; this version has been properly modified to allow the use of LBE as a fluid.Since the RELAP5-3D code is already equipped with LBE as working fluid, RELAP5-3D has recently been acquired by SCK•CEN in anticipation of the licensing process.The first important action taken consisted of comparing the two codes by running the existing MYRRHA model input deck, developed for RELAP5 MOD 3.3, on RELAP5-3D.From the steady-state comparative analysis, it appears clear how the two code versions are using different physical models since the steady-state predictions show several differences. Several code issues have been found, mainly about LBE physical properties, initial noncondensable gas computation at LBE free surface level, and LBE heat transfer coefficient correlations.For what concerns the transient analysis, the protected loss-of-flow (PLOF) accident has been taken as reference. Also, in PLOF conditions the mass flow rates and temperature distributions are affected by physical properties and heat transfer model differences.

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.

Experimental and numerical investigation of liquid-metal free-surface flows in spallation targets

Accelerator Driven Systems (ADS) are extensively investigated for the transmutation of high-level nuclear waste within many worldwide research programs. The first advanced design of an ADS system is currently developed in SCK•CEN, Mol, Belgium: the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA). Many European research programs support the design of MYRRHA. In the framework of the Euratom project ‘Thermal Hydraulics of Innovative nuclear Systems (THINS)’ a liquid-metal free-surface experiment is performed at the Karlsruhe Liquid Metal Laboratory (KALLA) of Karlsruhe Institute of Technology (KIT). The experiment investigates a full-scale model of the concentric free-surface spallation target of MYRRHA using Lead Bismuth Eutectic (LBE) as coolant. In parallel, numerical free surface models are developed and tested which are reviewed in the article. A volume-of-fluid method, a moving mesh model, a free surface model combining the Level-Set method with Large-Eddy Simulation model and a smoothed-particle hydrodynamics approach are investigated. Verification of the tested models is based on the experimental results obtained within the THINS project and on previous water experiments performed at the University Catholic de Louvain (UCL) within the Euratom project ‘EUROpean Research Programme for the TRANSmutation of High Level Nuclear Waste in Accelerator Driven System (EUROTRANS)’. The design of the target enables a high fluid velocity and a stable surface at the beam entry. The purpose of this paper is to present an overview of both experimental and numerical results obtained for free surface target characterization. Without entering in technical details, the status, the major achievements and lessons for the future with respect to model development are described as well as some applications, which were carried out within the work package ‘multi-phase flow’ of THINS.

E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant.The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing.The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE.The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

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.

MYRRHA: A multipurpose nuclear research facility

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback.As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

MYRRHA – A multi-purpose fast spectrum research reactor

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental Accelerator-Driven System (ADS) currently under development at SCK⋅CEN and will replace the Material Testing Reactor (MTR) BR2. The MYRRHA facility is currently being developed with the aid of the FP7-project “Central Design Team” and will be as a flexible irradiation facility, able to work in both subcritical and critical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications, and Si-doping.MYRRHA will also demonstrate the full concept of Accelerator Driven Systems by coupling the requisite three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow for the study of efficient transmutation of high-level nuclear waste. Since MYRRHA is based on the heavy liquid metal technology, Lead–Bismuth Eutectic, it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) technology. Further, in critical mode, MYRRHA will play the role of European Technology Pilot Plant in the path forward for LFR.In this paper we present the historical perspectives, international and high profile membership within the consortium of the MYRRHA project and the rationale for the design choices are presented. Also, the latest configuration of the reactor system is described together with the different irradiation capabilities. More specifically, the possibilities and performances for fuel irradiations are presented in detail.

Multi-purpose hYbrid Research Reactor for High-tech Applications a multipurpose fast spectrum research reactor

Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) is the flexible experimental accelerator-driven system in development at SCK•CEN. The MYRRHA facility, currently developed with the aid of the FP7-project ‘Central Design Team’ is conceived as a flexible irradiation facility. MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications, and Si-doping. MYRRHA will also demonstrate the accelerator-driven system full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. Because MYRRHA is based on the heavy liquid metal technology, lead–bismuth eutectic, it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) Technology, and in critical mode, MYRRHA will play the role of the European Technology Pilot Plant in the roadmap for LFR. Copyright © 2011 John Wiley & Sons, Ltd.