EUROfusion, the European consortium for fusion research, coordinates and funds Research and Development (R&D) activities as part of a long-term road map toward achieving commercial fusion energy, primarily through projects like ITER and its successor, DEMO. A significant technological challenge in this endeavor is the development of the breeding blanket (BB), which is crucial for power generation, neutron shielding, and tritium production. Currently, two primary BB concepts are under study in Europe: the water-cooled lithium lead (WCLL) and the helium-cooled pebble bed. While ITER will not feature a fully operational BB, it will incorporate specialized test blanket modules (TBMs) designed to evaluate BB technologies under fusion-like conditions. These TBMs will simulate the functionalities of the BB to provide invaluable experimental data in view of the design and operation of EU-DEMO, which is going to be a prototype power plant equipped with a fully operational BB. In this context, ENEA and Sapienza University of Rome focus their joint R&D activities on the WCLL BB concept, conducting thermal-hydraulic (TH) analyses using the RELAP5/Mod3.3 system code. Their collaboration encompasses various aspects of BB design, component validation, and evaluation of balance of plant (BoP) systems, aiming to assess design resilience, optimize performances, and validate safety measures across steady-state and transient scenarios for both ITER and DEMO. Particular emphasis is placed on the challenges presented by the pulsed operational mode, which is unique to tokamak fusion reactors and imposes demanding dynamic conditions on components and systems due to rapid power transitions between pulse and dwell phases. To support this effort, ENEA R.C. Brasimone is designing and building the W-HYDRA experimental platform to support the validation and the design optimization of systems and components for both ITER and DEMO facilities. Its main objectives include reproducing the TH conditions of fusion reactor subsystems, facilitating component testing, and advancing safety and performance standards under realistic operational scenarios. This paper provides a comprehensive overview of the RELAP5/Mod3.3 TH analyses conducted to support the design and the operational strategies of the ITER WCLL TBM and its associated water cooling system, as well as the DEMO WCLL BB and BoP. Given the significance of experimental results in the adoption of well-proven technology for a fusion-like environment, this work also details the extensive TH analyses performed with the RELAP5/Mod3.3 to design and optimize the W-HYDRA platform.

The EUropean DEMOnstration Power Plant (EU-DEMO) represents a milestone in nuclear fusion research, serving as crucial step towards the realization of commercial fusion energy production by bridging the gap between current research efforts and future industrial-scale deployment. A key component of the reactor is the Breeding Blanket (BB) that must perform several essential functions for the proper DEMO operation. Two primary concepts for BB have been proposed for DEMO: the Water Cooled Lithium Lead (WCLL) and the Helium Cooled Pebble Bed (HCPB). Both concepts are going to be tested under realistic fusion reactor conditions in ITER, in the form of Test Blanket Modules (TBMs). In this framework, at the ENEA R.C. Brasimone, the construction of an experimental infrastructure called W-HYDRA is ongoing. It is dedicated to the investigation of the water and lithium-lead technologies applied to the fusion research field. As part of the W-HYDRA infrastructure, Water Loop facility will investigate WCLL BB components, such as a First-Wall (FW) test section. The design characteristics and performance of the mock-up will be assessed to provide valuable experimental results in view of DEMO operation. The present paper is focused on the thermal-hydraulic numerical study of the WCLL BB FW test section within Water Loop facility using RELAP5/Mod3.3. Specifically, the study investigates expected operating transients (i.e., pulse-dwell and dwell-pulse transients) and accidental scenarios (i.e., Loss Of Feedwater Accident, LOFA), with the aim of supporting the design phase by providing preliminary results on the Water Loop facility and on the mock-up operation.

One of the most promising designs for the breeding blanket is the water cooled lithium lead, extensive efforts have been carried out over the years to study the interaction between water and lithium-lead in case of a loss of coolant accident, both experimentally and numerically. This possible accident scenario requires a numerical tool able to simulate the complex phenomena involved. The SIMMER code is currently the best candidate since it is used to perform safety analysis for liquid metal nuclear fission reactors. Even though, different milestones have been reached for the adaptation of SIMMER code for the Lithium-Lead/water interaction, a comprehensive work on the lithium-lead equation of state (EOS) and thermo-physical properties has never been carried out. The investigation of the EOS and thermo-physical properties of SIMMER is conducted and the most updated data on lithium-lead is collected. Using the experimental data, the new SIMMER analytical equations for lithium-lead are obtained. The implementation of EOS and thermo-physical properties of lithium-lead in SIMMER enables the real fluid behavior of the liquid metal and they will represent more realistically the conditions occurring during an interaction between lithium-lead and water during nuclear fusion reactor accidents.

The Water Cooled Lithium Lead Breeding Blanket (WCLL BB) is a key candidate for the driver blanket of the European DEMO reactor, progressing toward its Conceptual phase by the end of 2027. To assess different water and lithium-lead technologies for the WCLL BB and Balance of Plant (BoP) systems, the Water-thermalHYDRAulic (W-HYDRA) experimental platform is under development at the ENEA Brasimone Research Centre. Among the facilities constituting the new W-HYDRA multipurpose infrastructure, STEAM is going to experimentally investigate the DEMO WCLL BoP thermal-hydraulics, focusing on the Steam Generator (SG) of the Primary Heat Transfer Systems (PHTS), to qualify its performances and suitability under its unconventional operation. The paper aims at supporting the thermal-hydraulic characterization of the Steam Generator mock-up during the sudden power variations typical of a pulsed fusion reactor. The analyzed selected scenario is the operational transient dwell-pulse-dwell, which determines high thermal cycling and correspondent high thermomechanical stresses on the primary side components. Two control logics with their relative drawbacks have been analyzed with a RELAP5/Mod3.3 1-D model, the first regulating the primary side average temperature, the second monitoring the minimum one. The comparison of the two systems highlighted that neither approach leads to hazardous conditions for the facility. However, while the average temperature controller is characterized by reduced thermal stresses on the components, the minimum temperature controller is characterized by higher thermal gradients for the primary loop. Both methodologies will be tested in the dedicated experimental campaign, aiming at yielding insights and evaluations concerning control strategies applicable to the DEMO reactor.

In the framework of the development of fusion energy, one of the most prominent technologies arising to address the issues of tritium breeding and power conversion is the Water-Cooled Lithium-Lead (WCLL). This technology utilizes a molten eutectic alloy of Lithium and Lead which circulates inside the Breeding Blanket (BB) and is irradiated with neutrons to produce tritium. Water is then circulated inside the system to cool the components. The simultaneous presence inside critical areas of the reactor of molten metal alloy and water, at high temperature and pressure, poses significant safety concerns. For this reason, adequate design and analysis techniques are required to ensure the ability of the system to survive and mitigate any possible damage in case of the in-box Loss of Coolant Accident (LOCA), the most critical postulated accidental scenario. With this aim in mind, a novel approach was implemented with the aim of coupling the SIMMER-III code and the ANSYS Mechanical code for the modelling of both the chemical and thermodynamical interactions between water and the alloy, and the resulting effects on the structures. This work presents the status of the coupling technique development and the results of the preliminary validation activities performed against experimental data provided by the LIFUS5 facility operating at ENEA Brasimone Research Centre. The resulting comparison between these data and the codes' predictions allows a careful evaluation of the errors introduced in each step of the chain. Moreover, it provides confidence in the capacity of the methodology to correctly predict the ability of the structures to withstand incidental loads without suffering extensive damage.This work aims at providing engineers with a usable and powerful tool that allows for the safety analysis of WCLL-based components during the early stages of the design phase. This would help save time, and effort and reduce the economic cost that might arise from any undetected issue propagating downstream the design process.

In the framework of the activities coordinated by the EUROfusion consortium, the Water thermal-HYDRAulic (W-HYDRA) experimental platform is being built at the ENEA Brasimone Research Centre in order to support the development of the Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB). In particular, this infrastructure will make possible the installation and testing of prototypical mock-ups under relevant working conditions, such as the First Wall (FW), the manifold and the Steam Generator (SG). Moreover, it will represent an integral test facility for the investigation of phenomena characteristic of WCLL BB concept, such as the PbLi/water interaction. Finally, the collection of data coming from the different planned experimental campaigns will allow to qualify and validate numerical models and codes currently adopted for the design of components, as well as for the modelling of complex phenomena typical of the WCLL BB. In order to come to a definitive design of the different facilities constituting the experimental platform, several design analyses assessing the thermal, hydraulic and structural performances of the different facilities and components are necessary. The paper reports a highlight of the W-HYDRA platform with a general description of the facilities. Some of the most relevant design studies carried out so far are reported as well, highlighting their impact on the evolution of the design.

The Primary Heat Transfer System (PHTS) of the EU DEMO Water-Cooled Lithium-Lead (WCLL) Breeding Blanket might rely on Once-Through Steam Generators (OTSGs) technology, widely used in fission industry. ENEA is currently developing the STEAM facility, in order to characterize the behaviour of the EU DEMO OTSG. To this aim, a dedicated OTSG mock-up is being designed. The present work describes the Computational Fluid Dynamics (CFD) approach used to investigate the flow field within the inlet and outlet regions of the OTSG mock-up in steady-state conditions. One of the key objectives of the analysis is to assess the mass flow distribution within the OTSG primary bundle, as uniform distribution is crucial for achieving the desired performance. To this aim, the computational approach employs reliable models such as RANS k−ϵ as well as a fluid–fluid co-simulation routine. The aim is to reach a satisfactory self-consistency of the flow distribution within the OTSG primary bundle without the discretization of the full geometry. The CFD analysis predicts a maximum absolute deviation from the average mass flow rate value lower than 2%. This predicted behaviour is compared against a benchmark simulation that discretizes the whole primary bundle. The comparison displays an acceptable compatibility and a significant reduction in the required computational resources thanks to the implementation of a co-simulation routine.

DEMO power station aims to demonstrate the generation of hundred MWs of electrical power from fusion reactions, then transmitted from the Tokamak reactor to the grid through the Balance of Plant (BoP). The design approach for the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) leverages nuclear industry expertise but faces challenges due to DEMO pulsed operation and low-load periods. To assess the feasibility of these components, ENEA Experimental Engineering Division at Brasimone R.C. is designing STEAM, a facility investigating water technologies applied to the DEMO BB and BoP systems and components. STEAM is mainly composed of a primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328–295 °C) and a secondary two-phase (liquid/steam) loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238–300 °C). Experimental validation will reproduce steady-state and transient operation under DEMO-relevant conditions, including dedicated tests on the DEMO once through steam generator mock-up.STEAM objectives and description are presented in this paper, together with the RELAP5/Mod3.3 nodalization of the facility. The latter is used to, thermal-hydraulically characterize the facility behaviour. The outcomes of the steady-state qualification supported the optimization of the system layout appraising the performances of key components under the specified operating conditions.

While nuclear fusion reactors hold promise, their realization faces significant complexities, not only in the construction but also in managing the auxiliary systems required for safely operating plasmas. Pulsed plasma regime, integral to reactor normal operations, determines atypical operative conditions that can cause high thermal stresses on the components and generate instabilities. R&D activities are therefore required to characterize the components behavior in the extremely severe conditions of pressure and temperatures associated with fusion, especially during sudden power transitions. ENEA, as part of the EUROfusion consortium, is planning the construction of STEAM, an experimental facility aimed at qualifying the DEMO Steam Generator during pulse-dwell-pulse transitions within the W-HYDRA platform. Dedicated experimental campaigns will reproduce the low-power phase associated with the only material activation (dwell), the full-power phase (pulse) and the corresponding power transitions. This paper focuses on the low-power phase, presenting the performed RELAP5/Mod3.3 simulations, investigating the thermal-hydraulic performances of the system at 1 %, 5 % and 10 % of the nominal power, hence in conditions considerably different from fission standard practices. The simulation results are essential for defining regulation strategies for the SG test section and ensuring the correct operation of the entire system.

In the design of the Water Cooled Lithium Lead Breeding Blanket, a critical concern is the in-box Loss of Coolant Accident (LOCA). Extensive research, both experimental and numerical, has been devoted to understanding the interaction between Lithium-Lead and Water and its implications for the overall system. To simulate this complex scenario, the SIMMER code, originally developed for Liquid Metals, was chosen. However, SIMMER alone cannot handle complex pipeline geometries. In contrast, RELAP5, a thermal-hydraulic system code, is well suited for these kinds of geometries and it is intensively used in the nuclear industry. A coupling methodology was developed, combining the strengths of both codes. The simulation involved a geometry resembling the ITER TBM BU test section using SIMMER-IV for nodalization, while RELAP5/Mod3.3 modeled the piping system of the new Integral Test Facility under construction at ENEA CR Brasimone, called LIFUS5/Mod4. This work shows that the coupling between the two codes achieved high synchronization in steady-state conditions and successfully replicated a fast transient in-box LOCA scenario. The results show the pressure propagation along the entire Lithium-Lead loop generated by the accidental scenario. The simulation outcomes, though preliminary from a technical point of view shows that the pressure wave is dampened due to the overall system compressibility. This coupling tool proves invaluable for supporting facility design and safety assessments of the WCLL breeding blanket design.

•SIMMER-III code validation for WCLL in-box LOCA assessment•Lead-lithium eutectic and water chemical interaction•Numerical Analysis of Experimental Test E5.2•Execution of the Series E experimental campaign employing the facility, LIFUS5/Mod3Research activities are continuing between the University of Pisa and ENEA Brasimone Research Center to better understand the phenomena and processes that occur during a postulated in-box LOCA in the Water-Cooled lead-lithium Breeding Blanket (WCLL BB) and during the system's safety response. Various activities are also going on to strengthen the reliability of numerical tools and validate computer models, codes, and procedures for their implementations.

The WCLL Breeding Blanket of DEMO and the Test Blanket Module (TBM) of ITER require accurate R&D activities, i.e., concept validation at a relevant scale and safety demonstrations. In view of this, the strategic objective of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone of ENEA, is twofold: to conduct R&D activities for the WCLL BB to validate design performances and to increase the technical maturity level for selection and validation phases, as well as to support the ITER WCLL Test Blanket System program. Basically, the Water Loop facility will have the capability to investigate the design features and performances of scaled-down or portions of breeding blanket components, as well as full-scale TBM mock-ups. It is a large-/medium-scale water coolant plant that will provide water coolant at high pressure and temperature. It is composed by single-phase primary (designed at 18.5 MPa and 350 °C) and secondary (designed at 2.5 MPa and 220 °C) systems thermally connected with a two-phase tertiary loop acting as an ultimate heat sink (designed at 6 bar and 80 °C). The primary loop has two main sources of power: an electrical heater up to about 1 MWe, installed in the cold side, downstream of the pump and upstream of the test section, and an electron beam gun acting as a heat flux generator. The WL has unique features and is designed as a multi-purpose facility capable of being coupled with the LIFUS5/Mod4 facility to study PbLi/water reaction at a large scale. This paper presents the status of the Water Loop facility, highlighting objectives, design features, and the analyses performed.

A critical problem in the Water-Cooled Lead-Lithium Breeding Blanket system (WCLL-BB) is the possible interaction between the water and the Lithium-Lead eutectic alloy - which act respectively as primary coolant and as breeder/neutron multiplier - due to a postulated rupture of the coolant circuit in the Breeding Unit of the BB. This scenario involves a complex multiphase interaction together with an exothermal chemical reaction between the two fluids with the production of hydrogen. The PbLi/water chemical reaction was implemented in SIMMER-IV code by the University of Pisa and, consequently, a coupling methodology was successfully developed between SIMMER-IV and RELAP5/Mod3.3 codes, in order to overcome SIMMER-IV unsuitability in the simulations of complex pipelines. This paper presents an application of the coupling methodology to the simulation of experimental tests, recently performed at ENEA inside the experimental campaign carried out with the LIFUS5/Mod3 facility at the ENEA Brasimone Research Centre. The injection line of the facility is simulated by RELAP5/Mod3.3, whilst the reaction vessel is simulated with SIMMER-IV. Results of different simulations are presented and compared against experimental data, providing both qualitative and quantitative evaluations of the performance of the coupling methodology in the prediction of the chemical and thermal-hydraulic phenomena involved in the experiments, such as the fast pressurisation of the injection line and the pressurisation of the reaction vessel, the energy release due to the chemical reaction and the propagation of pressure waves inside the reaction vessel.

Recent R&D activities in nuclear fusion have identified the DEMO reactor as the ITER successor, aiming at demonstrating the technical feasibility of fusion plants, along with their commercial exploitation. However, the pulsed operation of the machine causes an “unconventional” operation of the system, posing unique challenges to the functional feasibility of the steam generator, for which it is necessary to define and qualify a reference configuration for DEMO. In order to facilitate the transitions between different operational regimes, the Once Through Steam Generator (OTSG) is considered to be a suitable choice for the DEMO primary heat transfer systems, being characterized by lower thermal inertia with respect to the most common U-tube steam generators. In this framework, the ENEA has undertaken construction of the STEAM facility at Brasimone R.C., aiming at characterizing the behavior of the DEMO OTSG and related water coolant systems in steady-state and transient conditions. A dedicated OTSG mock-up has been conceived and designed, adopting a scaling procedure, keeping the height 1:1 of the DEMO OTSGs. The conceptual design has been supported by RELAP5/Mod3.3 thermal-hydraulic calculations. CFD and FEM codes have been used for fluid-dynamic analyses and mechanical stress analyses, respectively, in specific parts of the component.

In the framework of the development of fusion energy, one of the most prominent technologies arising to address the issues of tritium breeding and power conversion is the Water-Cooled Lithium-Lead (WCLL). This technology utilizes a molten eutectic alloy of Lithium and Lead which circulates inside the Breeding Blankets (BB) and is irradiated with neutrons to produce tritium. Water is then circulated inside the system to cool the components. The simultaneous presence inside critical areas of the reactor of molten metal alloys and water, at high temperature and pressure, poses significant safety concerns. For this reason, adequate design and analysis techniques are required to ensure the ability of the system to survive and mitigate any possible damage in case of the in-box Loss Of Coolant Accident (LOCA), the most critical postulated accidental scenario. This work introduces a new methodology for the integral safety analysis of WCLL components, with a particular focus on the WCLL Breeding Blankets, which is based on a fully automated code-chain technique. Its goal is to couple the calculations performed in the fluid domain by the SIMMER-III code, which models the chemical and thermodynamical interactions between the water and the alloy, and the structural simulations performed by the ANSYS code on the mechanical components. The entire process is validated against experimental data provided by the LIFUS5 facility operating at ENEA Brasimone Research Centre. The resulting comparison between these data and codes’ predictions allows a careful evaluation of the errors introduced in each step of the chain. Moreover, it provides confidence in the capacity of the methodology to correctly predict the ability of the structures to withstand incidental loads without suffering extensive damage.

The Water-Cooled Lithium–Lead (WCLL) is one of the most promising technologies for power conversion and tritium production in future fusion-powered reactors; it will be implemented in one of the Test Breeding Modules (TBM) inside the ITER reactor and the DEMO EU reactor. However, the simultaneous presence in the system of high-temperature PbLi and high-pressure water poses significant safety issues in the event of an in-box LOCA (Loss Of Coolant Accident). For this reason, a complete understanding of the system response is crucial to avoid extensive damage in such a scenario. This paper describes the status and design features of the LIFUS5/Mod4 facility, an experimental plant that is currently being designed and constructed at ENEA CR Brasimone in the framework of the FP9 EUROfusion Horizon Europe to address these issues. This facility aims at being representative of the geometry and operational conditions of the Test Breeding System (TBS) to allow the precise reproduction of its behavior under simulated incidental scenarios. For this reason, peculiar design choices have been made, which will be extensively discussed throughout this work and which will allow the generation of high-quality data useful for the TBS development. Moreover, the facility is expected to become a test stand for the implementation of different safety functions, to identify the best accident-mitigation strategy. Possible upgrade plans for the facility are described as well, with the chance for it to become a fully functional test stand for any component of the TBS in their operative conditions.

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes.

DEMO is a nuclear fusion power station that has, among others, the objective of demonstrating the possibility to produce several hundred MWs of electrical power from fusion reaction by the middle of this century, increasing the production of carbonneutral electricity. In particular, the Balance of Plant of DEMO has the key role to demonstrate the feasibility of delivering the power produced within the Tokamak reactor to the grid.

The LIFUS5 facility is a separate effect test facility aimed at investigating the heavy liquid metal-water interaction. It has been designed and constructed to withstand high pressure and temperature (i.e. up to 200 bar and up to 500 degrees C) and to record the fast pressure transients in heavy liquid metal melt. This type of transient is typical of a Steam Generator Tube Rupture event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system, and should be investigated from the safety point of view because it could potentially induce, beyond the damaging of the internal structures (HX tube bundle, above core structures, Fuel Assembly, Control Rods, etc.) several negative effects on the operation of the reactor. These include an insertion of positive reactivity into the system or reduced cooling efficiency due to steam dragging into the core. It will also have an effect on the chemistry control of the cooling. All these effects compromise the safety and the reliability of the system. The twenty years experimental programme and the different LIFUS5 configurations are presented in the paper, as well as the experimental campaigns and test matrix characterizing the heavy liquid metal-water interaction phenomena and data for codes validation.

The Breeding Blanket is a necessary component to close the nuclear fusion reactor fuel cycle. amongst the most promising conceptual design, there is the Water Cooled Lithium Lead Breeding Blanket, with water as coolant and eutectic Lithium-Lead as neutron multiplier and breeder. The possible interaction between water and Lithium-Lead poses a main safety concern and prompted the scientific community to develop a numerical analysis tool able to simulate such a complex interaction. The SIMMER-III code was modified by UNIPI to simulate the chemical interaction between water and Lithium-Lead, furthermore also a coupling methodology between SIMMER-III and RELAP5/Mod3.3 was developed. The coupling tool employs SIMMER-III code to simulate the zone of Lithium-Lead interacting with water, whilst the RELAP5 code is used to simulate the water pipelines. The LIFUS5/Mod3 facility at the ENEA Brasimone Research Centre was designed to perform reliable experimental activities on the interaction between water and Lithium-Lead. In this facility water at high pressure is injected inside a reaction vessel, where the thermodynamic and chemical interaction between water and Lithium-Lead occurs. The experimental activities are divided in different tests matrix, the Series D and Series E tests. The two series differ in the amount of water injected during the transient. In series D the mass of water is predetermined whilst in series E water was injected continuously for a pre-fixed interval of time and the total injected mass was estimated a posteriori. This work shows the results of the coupling tool applied to Series E. The comparison between the experimental and numerical results is performed by identifying and characterizing the phenomena involved in the interaction. Furthermore, the overall performance of the coupled codes in the simulation of the phenomena is presented here.