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The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295–328 °C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330 °C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100 mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out. © 2019 Elsevier B.V.

The Water-Cooled Lithium Lead breeding blanket concept foresees the eutectic lithium-lead (Pb-15.7Li) alloy being cooled by pressurized sub-cooled water (temperature 295–328 °C; pressure 15.5 MPa) flowing in double wall tubes. Therefore, the interaction between the Pb-15.7Li and water (e.g. tube rupture) represents one of the main safety concerns for the design and safety analysis. Available LIFUS5/Mod2 experimental data are employed to assess the performances of thermal-hydraulic and thermo-mechanic codes. Thermal-hydraulic simulations, by SIMMER-III code, are focused on the prediction of the thermodynamic interaction among the fluids. ABAQUS Finite Element code, used for the design activities, is adotped to perform the thermo-mechanic simulations, calculating the stress and strain fields of LIFUS5/Mod2 main vessel during the experiments. Code results are compared with the experimental data and the outcomes from the analyses are discussed, in order to derive conclusions on the code assessment.

In the European and international framework, ENEA coordinates the Italian fusion program, supported by linked third parties as universities, research institutes and industries. In this context, the Experimental Engineering Division (FSN-ING) is involved in experimental and numerical activities related to Breeding Blanket (BB) and Divertor technologies development. This paper describes the scientific works presently ongoing at Brasimone R.C. enveloped in LLE, lithium, helium, tritium and pressurized water technologies, characterization of structural materials, analysis of materials corrosion rate and development and qualification of anti-permeation/corrosion barrier. The experimental activities conducted for the investigation of safety-relevant scenarios as In-Box LOCA (LLE-water interaction in the WCLL or shock waves propagation generated by helium injection in LLE in the HCLL/DCLL-BBs) are here reported. Finally, new activities have been planned to support the Divertor Tokamak Test divertor characterization, the large-scale LLE-water interaction and a LLE components validation in relevant scale for WCLL-BB.

The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (Water-Cooled Lead-Lithium Breeding Blanket) that will be tested in the Test Blanket Module of the ITER reactor. In this accident, a tube rupture in the coolant circuit would cause a potential disruptive exothermic chemical reaction between the water (the primary coolant) and the Lead-Lithium. This paper presents the development of numerical tools for the analysis of the in-box LOCA consequences in the breeding unit of the blanket. In this work a simplified computational model of the breeding unit was created by SIMMER-IV which includes the lead lithium and water systems and the internal solid structures. The simulation is set up at the foreseen nominal working conditions of the units and involves a small rupture of the cooling tube, with the consequent propagation of high pressure waves on the breeder side and the fast evolution of the chemical reaction. It is shown here that the code is capable of correctly predicting both these events, from a qualitative point of view, thus providing a highly valuable insight of the whole scenario. These preliminary results will provide a crucial base for further development of this computational tool and the simulation of the whole Test Blanket Module.

The European liquid Test Blanket Module (TBM) concept named Water Cooled Lithium Lead makes use of the liquid metal alloy PbLi as tritium breeder/neutron multiplier and water as coolant. A dedicated PbLi ancillary system feeds the detritiated liquid metal alloy to the TBM and returns the PbLi enriched with tritium. The object of this study is developing a numerical model able to simulate the thermo-hydraulic behavior of the PbLi ancillary system under both normal operation and accidental scenario, with the aim to investigate the propagation of pressure waves generated by an in-box LOCA. The accident consists of a tube rupture of the coolant circuit inside the TBM, leading to the interaction between PbLi and water. As a consequence, a pressure increase occurs inside the breeder zone. The accident might affect the ancillary systems due to the propagation of the pressure waves generated inside the TBM. The PbLi loop is modelled by means of the thermo-hydraulics system code RELAP5/Mod3.3 and an instantaneous pressure peak is applied inside the control volume simulating the TBM soon after reaching the steady state conditions inside the whole loop. The propagation of the pressure wave inside the circuit and the hydraulic behavior of the individual components is studied. The present paper represents a first step towards the investigation of the PbLi system and its safety features with the aim to support the R&D activities carried out in the framework of the ITER project.

The in-box Loss of Coolant Accident (LOCA) scenario is considered as one of the most affecting safety concerns for the Water-Cooled Lead Lithium Breeding Blanket (WCLL-BB) modules of the DEMOnstration (DEMO) reactor, which is sequentially followed by a multi-phase multi-component physical and chemical interaction. Therefore, the transient behavior of such accidents has to be carefully investigated during the design phase of the plant, to evaluate the consequences and to adopt the necessary mitigating countermeasures. This also requires a numerical predictive tool, which is capable to model such transients and predict the relevant phenomena under an operational condition and the connected safety parameters i.e. system pressure, temperature, chemical products mass, and volume fractions of all the existing components. Consequently, the SIMMER-III code was firstly improved at the University of Pisa by implementing the chemical reaction between PbLi eutectic alloy and water. In addition to this, an experimental campaign and a test-matrix have been recently designed according to the LIFUS5/Mod3 facility to perform a series of experiments and code post-test analyses. In the present work, the experimental data of the first LIFUS5/Mod3 test is used for the validation of the chemical model implemented in SIMMER-III through a comprehensive sensitivity study. The applied methodology for the code validation is based on a three-step procedure including qualitative analysis, quantitative analysis and the results from sensitivity analyses. The qualitative accuracy evaluation is performed through a systematic comparison between experimental and calculated time trends based on the engineering analysis, the resulting sequence of main events and the identification of phenomenological windows and of relevant thermo-hydraulic aspects. Afterwards, the accuracy of the code prediction is evaluated from a quantitative point of view by means of selected, widely used, figures of merit. Finally, the results from the sensitivity cases are analysed and quantified, to determine the effects of the most influencing code input options and transient parameters. Furthermore, the analysis is followed by applying the Fast Fourier Transform Method (FFTM) to the experimental signals and all the sensitivity calculations. The comparison shows a very good agreement for pressure transient between the experimental and numerical data, while for the temperature and the hydrogen production the results fall into acceptable criteria, which means that the code is reliable in capturing and predicting the transient values but not perfectly match with the experimental signals.

The in-box LOCA (Loss of Coolant Accident) is a relevant safety issue for the design of the WCLL BB. Research activities are ongoing to master phenomena and processes occurring during the postulated accident, to enhance the predictive capability and reliability of numerical tools, and to validate computer models, codes and procedures for their applications. Current status of knowledge requires the availability of qualified and reliable experimental data to support these activities. In view of this, the new separate effect test facility LIFUS5/Mod3 has been commissioned and the Series D experimental campaign is in progress. The tests are focused on the generation of reliable experimental data for the validation of the modified version SIMMER codes for fusion application. Moreover, data will be also used to investigate the dynamic effects of energy release on the structures and to provide relevant feedbacks for the follow up experimental campaigns. The experimental data and results of Test D1.1 are reported and critically discussed, focusing on pressures, temperatures, amount of injected water, and hydrogen production quantification.

Water-cooled lithium-lead breeding blanket is considered a candidate option for European DEMO nuclear fusion reactor. ENEA and the linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is based on horizontal (i.e. radial-toroidal) water-cooling tubes in the breeding zone, and on lithium lead flowing in radial-poloidal direction. This design choice is driven by the rationale to have a modular design, where a basic geometry is repeated along the poloidal direction. The modules are connected with a back supporting structure, designed to withstand thermal and mechanical loads due to normal operation and selected postulated accidents. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium lead system. Rationale and features of current status of water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanics, thermo-hydraulics and neutronics analyses. Open issues and areas of research and development needs are finally pointed out.

Three of the four breeder blanket concepts currently under investigation for the European DEMO Reactor use the eutectic Pb-16Li as breeder material. Those are the Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL) blankets. Moreover, the WCLL is one of the blanket concepts that will be qualified in the ITER reactor, therefore the development and design of lead lithium loops and auxiliary systems is essential. The main functional requirements that Pb-16Li systems have to fulfill are: • to circulate the liquid Pb-16Li through the blanket and ancillaries; • to extract the tritium produced inside the breeder modules from Pb-16Li; • to control Pb-16Li chemistry and to remove accumulated impurities; The present work aims to describe the activities performed in order to achieve the following objectives: i) design and integration of the Pb-16Li loops inside the tokamak building, ii) development and characterization of antipermeation and anticorrosion coatings on structures in contact with Pb-16Li, iii) development and design of an activation products removal system, iv) design of a chemistry control system for Pb-16Li loops, v) performing magnetohydrodynamic analyses taking into account the impact on heat transfer and tritium transport in breeding blankets and performing safety analyses of water/Pb-16Li interaction due to LOCA inside the WCLL blanket.

The In-Box Loss Of Coolant (LOCA) postulated accident is considered a major concern for the safety connected with the development of EU-DEMO fusion reactor. Relating to the renewed interest in the Water-Cooled Lithium Led blanket concept, an innovative experimental campaign is under development at ENEA Brasimone laboratories aiming at investigating the consequences related to the In-Box LOCA applied to the WCLL breeding blanket. In this frame, a new coupling tool between the SIMMER-III (modified version to implement the PbLi/water chemical interaction) and the RELAP5/Mod3.3 codes (modified version to implement PbLi thermo-physical properties) has been developed together with its preliminary application to simple test cases with water as working fluid. The coupling procedure can be defined as a “two-way”, “non-overlapping”, “online” technique aiming at investigating multi-physics and multi-scales phenomena in support of the development of fusion reactor technologies.

A major safety issue in the Water-Cooled Lead-Lithium Breeding Blanket (WCLL-BB) system foreseen for fusion reactor is the interaction concerning the primary coolant (water) and the neutron multiplier (PbLi), due to a hypothetical tube rupture in the coolant circuit. This scenario involves an exothermic chemical reaction between PbLi and water with the production of hydrogen, in addition to critical interactions in a complex multiphase system in non-thermal equilibrium. In recent years the PbLi/water reaction was successfully implemented in the SIMMER-III code and validated against data from the LIFUS5/Mod3 experimental campaign. However, due to limitations of SIMMER-III, this work was restricted to the prediction of the phenomena inside the vessel, neglecting the simulation of the injection line. Nevertheless, since the injection line may actually have an important effect on the development of the transient, the simulation of the whole facility would be highly desirable. Indeed, the University of Pisa recently developed a coupling methodology between the SIMMER-III and RELAP5/Mod3.3 codes and applied it to simple single-phase cases. In this paper the complete simulation of the LIFUS5/Mod3 facility is presented, with the injection line modelled through RELAP5. Furthermore, all the complex aspects of the phenomena inside the reaction tank were included: the multiphase system and the interaction between water and PbLi with the chemical reaction and the production of hydrogen were modelled by SIMMER. Preliminary results are presented, showing that the coupling methodology can be effectively employed for the prediction of the chemical and thermal-hydraulic behaviour of complex loop experimental facilities.

The In-box LOCA for the WCLL-BB is recognized as a Design Basis Accident (DBA) and is of substantial interest to the DEMOnestration reactor design, therefore, the transient response of such an accident must be carefully investigated and addressed to ensure the safe operation and integrity of the whole system. In this way, the LIFUS5/Mod3 test facility (constructed at ENEA Brasimone Research Center) has been upgraded in the period 2018–2020 to perform a series of tests. The first set of the tests were named Series D, which are characterized by injecting specified amounts of pressurized water into the Lithium lead liquid bulk. The experimental campaign aimed at validating and qualifying the SIMMER-III code as a reliable numerical tool for the safety studies of the WCLL BB. In parallel with performing the tests, SIMMER-III Verification and Validation (V&V) was conducted according to a standard code validation procedure. SIMMER-III is a two-dimensional, multiphase, multicomponent, Eulerian, fluid-dynamics code which was firstly developed at the Japan Nuclear Cycle Development Institute (JNC). An adopted version of the original SIMMER-III code so-called Ver.3 F Mod.0.1 (which was developed at University of Pisa), was employed for the analyses. The V&V activity was successfully completed and documented as the technical reports within the past numerical analytical and experimental activities for the first three tests (D1.1, D1.2 and D1.5). In this article, the experimental data of the Tests D1.1, D1.2 and D1.5 are used for the SIMMER-III code results comparison. A qualitative analysis of the results obtained is reported according to the time trends for the most relevant parameters. The results show that the SIMMER-III code acceptably predicts the transient and the accuracies of the relevant test parameters are in agreement with the acquired experimental signals. Although that the primarily validation results are highly promising but further code assessment, development and validation are essential to approach such a qualified system code, which is suitable for fusion safety applications. The present validation work has been successfully followed by its ongoing experimental and numerical activities as a multilateral EUROfusion project.