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The water-cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant(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. The current design is based on DEMO 2015specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach isemployed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by meansof radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidaldirection. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operationand selected postulated initiating events. Water and lithium–lead manifolds are designed and integrated with a consistentprimary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium–leadsystem. Rationale and features of the single-module-segment water-cooled lithium–lead breeding blanket design arediscussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration withthe primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areasof research, and development needs are finally pointed out.

The water-cooled lithium-lead breeding blanket is a candidate option for the European Demonstration Power Plant (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. The current design is based on DEMO 2015 specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial-toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium-lead flows in a radial-poloidal direction. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. 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 the single-module-segment water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium

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 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). Research activities are ongoing to master the phenomena and processes that occur 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. Following these objectives, ENEA designed and built the new separate effects test facility LIFUS5/Mod3. Two experimental campaigns (Series D and Series E) were executed by injecting water at high pressure into a pool of PbLi in WCLL-BB-relevant parameter ranges. The obtained experimental data were used to check the capabilities of the RELAP5 system code to reproduce the pressure transient of a water system, to validate the chemical model of PbLi/water reactions implemented in the modified version of SIMMER codes for fusion application, to investigate the dynamic effects of energy release on the structures, and to provide relevant feedback for the follow-up experimental campaigns. This work presents the experimental data and the numerical simulations of Test E4.1. The results of the test are presented and critically discussed. The code simulations highlight that SIMMER code is able to reproduce the phenomena connected to PbLi/water interaction, and the relevant test parameters are in agreement with the acquired experimental signals. Moreover, the results obtained by the first approach to SIMMER-RELAP5 code-coupling demonstrate its capability of and strength for predicting the transient scenario in complex geometries, considering multiple physical phenomena and minimizing the computational cost.

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.

In the framework of the European Union MAXSIMA project, the safety of the steam generator (SG) adopted in the primary loop of the Heavy Liquid Metal Fast Reactor has been studied investigating the consequences and damage propagation of a SG tube rupture event and characterizing leak rates from typical cracks. Instrumentation able to promptly detect the presence of a crack in the SG tubes may be used to prevent its further propagation, which would lead to a full rupture of the tube. Application of the leak-before-break concept is relevant for improving the safety of a reactor system and decreasing the probability of a pipe break event. In this framework, a new experimental campaign (Test Series C) has been carried out in the LIFUS5/Mod3 facility, installed at ENEA Centro Ricerche Brasimone, in order to characterize and to correlate the leak rate through typical cracks occurring in the pressurized tubes with signals detected by proper transducers. Test C1.3_60 was executed injecting water at about 20bars and 200°C into lead-bismuth eutectic alloy. The injection was performed through a laser microholed plate 60 μm in diameter. Analysis of the thermohydraulic data permitted characterization of the leakage through typical cracks that can occur in the pressurized tubes of the SG. Analysis of the data acquired by microphones and accelerometers highlighted that it is possible to correlate the signals to the leakage and the rate of release.

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.