Water-Cooled Lithium Lead Breeding Blanket Research Articles

RELAP5-based thermal-hydraulic assessment of the STEAM facility for DEMO WCLL balance of plant analysis

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.

Analysis and optimization of the secondary circuit for the option WCLL BB direct coupling with small ESS of the EU DEMO power

The Primary Heat Transfer System (PHTS) of the European Demonstration fusion power plant (EU-DEMO) transfers heat from the reactor sources (Breeding Blanket (BB), Divertor and Limiters) to the secondary Power Conversion System (PCS). Two reference concepts of the EU-DEMO BB and the respective PHTS are currently under consideration: the Helium Cooled Pebble Bed (HCPB) BB with large Energy Storage System (ESS) and indirect coupling between PHTS and PCS, and the Water Cooled Lithium-Lead (WCLL) BB directly coupled to the PCS and with small ESS. According to the new EU-DEMO Energy Map there are significant uncertainties regarding the splitting of fusion power among different reactor sources. Because of these uncertainties 12 possible operational scenarios were formulated for each of the HCPB and WCLL options. In 2022 a design of the PCS circuit for the WCLL variant was developed, in which all the components were sized to be able to absorb the maximum possible thermal power of all reactor sources (“maximum of maximum” scenario). However, we observed that this PCS circuit could not reach the most problematic operational point A1-2, in which thermal power of BB and Divertor, was the lowest and the highest, respectively. In the present work we proposed a modified “maximum of maximum” PCS cycle for the WCLL BB option and we verified that it should be able to operate according to the A1-2 scenario. Operation of the considered PCS cycle during the plasma pulse and during the dwell phase was simulated using the GateCycle software.

Activated corrosion product contamination assessments of DEMO WCLL breeding blanket primary heat transport system

In water-cooled fusion reactors, the assessment of the primary system contamination is essential for waste management, machine availability, occupational radiation exposure, and radiological hazard determination. The primary cooling water is not only directly activated by the intense neutron field but is a contamination vector for a significant variety of gamma emitters with short to long decay half-lives. Corrosion products can be activated in those regions under neutron flux of the primary circuit and then released in the cooling water.In the EU-DEMO fusion power plant equipped with the Water-Cooled Lithium Lead Breeding Blanket (WCLL-BB) concept, the primary coolant undergoes intense neutron fields in the first wall and the breeding zone regions of the blanket. Activated Corrosion Products (ACPs) are then formed, released into the water, transported in the cooling loop and finally deposited onto the ex -vessel surfaces of the Primary Heat Transport System (PHTS), where working personnel are susceptible to being radiologically exposed.This work addresses the complete assessment of ACPs in the WCLL-BB PHTS of EU-DEMO. The simultaneous and multi-physical processes behind the ACP formation are tackled using the OSCAR-Fusion code, a comprehensive tool developed by the CEA (France) to assess contamination in fusion nuclear reactors. The whole system is modelled with zero-dimensional nodes with assigned geometrical, thermal-hydraulics, material and chemical parameters. Activation reaction rates integrated over the whole spectrum and calculated with MCNP are given to those regions exposed to the neutron flux. Results are provided in terms of mass and activity inventories of ACPs as deposit and inner oxide layers of components (pipes, heat exchangers, pumps...), ions in solution, particles in suspension, and filters and resins trapping.Mobilizable inventories such as ions, particles and deposits are important source terms in accidental scenario evolutions, while the whole activity inventory constitutes the main long-term gamma emitting source for dose rate maps determination in the tokamak building rooms housing the main PHTS equipment.

Thermal and structural analysis of the European water-cooled lithium lead breeding blanket concept in the L-H mode transition

The design of a breeding blanket system (BB) which reliably fulfils the prescribed requirements is pivotal for the construction of the EU DEMO fusion reactor. For this reason, the BB conceptual design activities require a tighter interaction with plasma physics modelling, so to employ relevant boundary conditions to compare, and then down-select, the BB design alternatives. The preliminary thermo-mechanical study herein reported highlights the importance to assess in deep the BB performance during plasma operational transients. In particular, analyses of the BB thermal and mechanical transients during plasma ramp-up have been carried out, with focus on the L-H transition. Attention has been paid to the Water-Cooled Lithium Lead (WCLL) BB concept, one of the two driver blanket candidates for the EU DEMO project. Adopting DEMO-relevant plasma configurations, the fusion power and the radiative power time evolutions during the plasma ramp-up phase have been modelled under different assumptions using the code ASTRA. The obtained power transients have been used to calculate the time-dependent nuclear power density and radiative heat flux the WCLL BB is subjected to, in order to perform the corresponding transient thermal analysis. Then, the structural analysis has been carried out in order to verify the fulfilment of the RCC-MRx structural design criteria in the most critical time steps. The performed thermo-mechanical analysis has been carried out adopting the Abaqus FEM code. Results obtained are presented and critically discussed, highlighting their importance for the EU DEMO integrated design.

STEAM Experimental Facility: A Step Forward for the Development of the EU DEMO BoP Water Coolant Technology

Within the EUROfusion roadmap for the technological development of the European-DEMOnstration (EU-DEMO) reactor, a key point has been identified in the discontinuous operation (pulse-dwell-pulse) of the machine. Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) adopt technology and components commonly used in nuclear fission power plants, whose performances could be negatively affected by the above mentioned pulsation, as well as by low-load operation in the dwell phase. This makes mandatory a full assessment of the functional feasibility of such components through accurate design and validation. For this purpose, ENEA Experimental Engineering Division at Brasimone R.C. aims at realizing STEAM, a water operated facility forming part of the multipurpose experimental infrastructure Water cooled lithium lead -thermal-HYDRAulic (W-HYDRA), conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. The experimental validation has the two main objectives of reproducing the DEMO operational phases by means of steady-state and transient tests, as well as performing dedicated tests on the steam generator aiming at demonstrating its ability to perform as intended during the power phases of the machine. STEAM is mainly composed of primary and secondary water systems reproducing the thermodynamic conditions of the DEMO WCLL BB PHTS and power conversion system, respectively. The significance of the STEAM facility resides in its capacity to amass experimental data relevant for the advancement of fusion-related technologies. This capability is attributable to the comprehensive array of instruments with which the facility will be equipped and whose strategic location is described in this work. The operational phases of the STEAM facility at different power levels are presented, according to the requirements of the experiments. Furthermore, a preliminary analysis for the definition of the control strategy for the OTSG mock-up was performed. In particular, two different control strategies were identified and tested, both keeping the primary mass flow constant and regulating the feedwater mass flow to follow a temperature set-point in the primary loop. The obtained numerical results yielded preliminary feedback on the regulation capability of the DEMO steam generator mock-up during pulsed operation, showing that no relevant overtemperature jeopardized the facility integrity, thanks to the high system responsivity to rapid load variations.

Presentation and Preliminary Test of APRIL: A New Facility for the Characterization of Antipermeation Coatings for Fission and Fusion Applications

Antipermeation and anticorrosion coatings are being developed to reduce tritium permeation from liquid metal [LiPb of the water-cooled lithium-lead breeding blanket and Pb for lead-cooled fast reactors (LFRs)] to primary heat transfer systems. The facility APRIL (Alumina-coating for tritium Permeation Reduction for Innovative LFR) was designed and installed at ENEA Brasimone R.C. to characterize the permeation reduction factor (PRF) of the candidate coatings in static conditions. In the current configuration, APRIL is composed of three pipes, closed at one end, that simulate the heat exchangers of the ALFRED LFR. Two of the pipes are coated with 3 µm of alumina with pulsed laser deposition techniques, the reference method for a fission reactor. The third pipe is uncoated. During the tests, all the pipes are filled with pressurized steam at 100 bar and 480°C, the steam generator condition of the ALFRED LFR. The tests are made in the gas phase; indeed, the three pipes are installed in a chamber filled with helium with a known concentration of deuterium that simulates tritium. Deuterium permeates inside the pipes, allowing for the evaluation of the PRF by means of the ratio between the measured permeated flux in the uncoated pipe and in the coated ones. A first test with 0.5% of deuterium was carried out and the evaluated PRF was about 13.5.

Modelling LOFA scenario for WCLL BB concept using MELCOR1.8.5 code

In the framework of the EUROfusion activities for the DEMO project, the evaluation of the Breeding Blanket (BB) performance and its ancillary systems is a compulsory activity. The European liquid BB concept, named Water Cooled Lithium Lead (WCLL), makes use of the liquid metal alloy PbLi as a tritium breeder/neutron multiplier and water as a coolant. A dedicated ancillary system guarantees the recirculation of the PbLi between the BBs and the tritium extraction/purification system. The BBs and the ancillary system must operate safely and reliably under the designed conditions, while accidental scenarios have to be identified for risk assessment and to define safety plans.In the present work an analysis of the In-Board Loss Of Flow Accident (IB LOFA) in the PbLi ancillary loop of the WCLL BB is conducted. This numerical analysis is performed by employing the fully integrated code MELCOR1.8.5, the release for fusion application. A LOFA scenario is considered by assuming the electromagnetic pump in the IB BB loop, which provides forced circulation to the PbLi, stops working. The nodalisation and numerical models are performed by applying geometries and design parameters using the 2020 WCLL PbLi design. The analysis is performed to develop a suitable numerical model, which simulates transients by studying several factors such as: tritium mobilization and transport inside the loop, fast plasma shutdown leading to pressure and temperature transients, and maximum temperatures inside the First Wall (FW).The preliminarily results demonstrate the maximum temperature inside the FW does not exceed the defined temperature thresholds for Tungsten and EUROFER (1000 °C and 500 °C, respectively). Furthermore, the pressure transient is appropriately accommodated within the ancillary loop. The initial tritium inventory in breeding modules is relocated and accrued in the tritium extraction system's free volume. This work is the first step in the safety analyses of the PbLi ancillary system aiming to support the R&D works carried out in the framework of the DEMO project.

Manufacturing of PAV-ONE, a Permeator against Vacuum Mock-Up with Niobium Membrane

The Permeator Against Vacuum (PAV) is one of the proposed technologies for the Tritium Extraction System of the WCLL BB (Water-Cooled Lithium-Lead Breeding Blanket) of the EU DEMO reactor. In this paper, the manufacturing of the first PAV mock-up with a niobium membrane with a cylindrical configuration is presented. This work aimed to demonstrate the possibility of manufacturing a relevant-size PAV to be later tested in the TRIEX-II facility. The adopted prototypical solutions are described in detail, starting with the methodology developed to join the Nb tubes with a 10CrMo9-10 (A182 F22) plate. Dedicated manufacturing and welding procedures, based on vacuum brazing with a nickel-based brazing alloy, were developed to solve the problem. This new kind of brazing was first analyzed to check the morphology of the joint and then tested to check its capability to withstand the TRIEX-II operative conditions. In parallel, the compatibility with a lithium-lead environment was analyzed by exposing samples of niobium and 10CrMo9-10 (A335 P22) to a flow of the eutectic alloy at 500 °C up to 4000 h. Finally, the PAV mock-up was installed in the TRIEX-II facility.

Numerical study of MHD mixed convection flow in the EU DEMO WCLL breeding blanket

The Water-Cooled Lithium Lead (WCLL) breeding blanket is a proposed candidate option for European DEMO nuclear fusion reactor. The utilization of PbLi as the predominant material occupying the bulk region of the breeding blanket has received significant attention. The interaction between Lorentz forces (Ha≈9800), intense buoyancy forces (Gr≈1.77×1010), and pressure-driven flow (Re≈236.77), plays a crucial role in the magnetohydrodynamics(MHD) and heat transfer phenomena of the region. To replicate the dynamics of PbLi, taking into account the impact of the strong magnetic field used for plasma confinement and the non-uniform heat source generated by nuclear heating, a 3-D DNS numerical simulation has been employed. Owing to the influence of buoyancy forces, several medium-sized circulations that are elongated in the direction of the magnetic field occupy the plenum region which is situated between the cooling pipes. Moreover, the electrical interaction between the liquid metal and the conducting wall gives rise to high-velocity flow jets, which exhibits peak velocities that are increased with the magnitude of the magnetic field. To characterize the heat transfer occurring in different regions of the system, local Nusselt and Grashof numbers have been determined. Optimization strategies for the BZ cooling system geometrical configurations are proposed based on the estimation of heat transfer and prediction of tritium permeation.

Thermomechanical and Thermofluid-Dynamic Coupled Analysis of the Top Cap Region of the Water-Cooled Lithium Lead Breeding Blanket for the EU DEMO Fusion Reactor

In the EU, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept is one of the candidates for the design of the DEMO reactor. From the past campaign of analysis emerged that the thermal-induced stress led to the failure in the verification of the RCC-MRx structural criteria. Hence, in this paper the classic conceptual design approach, based on a pure FEM thermal and structural analysis, is compared to a coupled thermofluid-dynamic/structural one. Even though the coupled approach requires tremendous modelling effort and computational burden, it surely allows determining the thermal field with a higher level of detail than the FEM analysis. Therefore, in this work, the focus is put on the impact of a more detailed thermal field on the DEMO WCLL BB global structural performances, focusing on the Top Cap region of its Central Outboard Blanket segment. The obtained results have allowed confirming the soundness of the design solution of the Top Cap region, except for concerns arising on the mass flow rate distribution. Moreover, results have shown that, globally, the pure FEM approach allows for obtaining more conservative results than the coupled one. This is a positive outcome in sight of the follow-up of the DEMO WCLL BB design, as it will be still possible adopting the pure FEM approach to quickly down-select design alternatives, using the most onerous coupled approach to finalise the most promising.

Exploratory study of the EU-DEMO Water-Cooled Lithium Lead breeding blanket behaviour in case of loss of cooling capability

Within the framework of the European Roadmap to the realization of fusion energy, a strong international cooperation is ongoing to develop a Breeding Blanket (BB) system for the EU-DEMO reactor. Although it is still to be decided whether the DEMO in-vessel components should perform any safety function, the pursuing of robust blanket concepts able to handle upset and accidental loading conditions has been always seen as good practice in fusion reactor engineering to enhance the inherent plant safety performances. Amongst the several classes of events that might challenge the BB structural integrity, the large Loss of Coolant Accident is one of the most relevant because it usually leads to a fast loss of cooling capability of the structures. Due to the characteristic of the tokamak assembly, the behaviour of each blanket segment during a sudden loss of cooling capability does not depend only upon distinguishing features of the component itself. In fact, the overall transient can be governed by conditions established in surrounding elements, like adjacent blanket segments and vacuum vessel, as well as by the plasma shutdown strategies adopted to protect the reactor.The scope of the activity herein presented is to make a preliminary assessment of the intrinsic capability of EU-DEMO tokamak architecture to cope with the loss of cooling in the Water-Cooled Lithium Lead (WCLL) BB concept. Evaluation of BB thermal field in short and medium term under simplified, yet conservative, assumptions was carried out for four transient scenarios with the aim of investigating the response of the structure in case of: a) fast or soft plasma shutdown, and b) different blanket cooling schemes. Moreover, the WCLL BB thermo-mechanical response in the most critical time steps has been assessed. The obtained results shall help for future decisions on safety systems/action to be implemented to cope with accidents.

Tritium Extraction from Lithium–Lead Eutectic Alloy: Experimental Characterization of a Permeator against Vacuum Mock-Up at 450 °C

Tritium extraction is one of the key open issues toward the development of the WCLL BB (Water-Cooled Lithium–Lead Breeding Blanket) of EU DEMO reactors, and different technologies have been proposed to address it. Among them, the Permeator Against Vacuum (PAV) has promising features, but it has never been tested in a relevant environment. This work presents the first experimental results ever obtained for a PAV mock-up. The experiments were carried out at ENEA Brasimone R.C. with the TRIEX-II facility on a mock-up characterized by a shell and tube configuration and using niobium as a membrane material. The experimental campaign was carried out with LiPb flowing at about 450 °C and 1.2 kg/s, while the hydrogen partial pressure was varied in the range 170–360 Pa. The characterization of the PAV performance was conducted by measuring the hydrogen partial pressure drop across the mock-up and the hydrogen permeated flux through a leak detector calibrated with an external hydrogen calibration cylinder. Moreover, the permeated flux was confirmed by a pressurization test performed measuring the pressure increase on the vacuum side of the PAV. The results constitute the first verification of the possibility to operate a PAV in flowing LiPb and to quantify its capabilities.

Design of a Test Section for the Experimental Investigation of the WCLL Manifold Hydraulic Features

A scaled-down test section representative of an Outboard Segment manifold of the Water-Cooled Lithium Lead Breeding Blanket for the European DEMO has been designed for installation and test in a high- mass flow branch of the W-HYDRA facility, under construction at the premises of ENEA Brasimone Research Center. The test section should confirm the flow repartition recently computed in the different breeding units on the full-scale manifold, validating at the same time the computational tools used for the design and analysis. The detailed objectives and requirements of the test section, as well as the scaling rationale and procedure adopted for its design, are presented in the paper. The final design of the test section is discussed. The preliminary analyses of the developed design are also presented and show that it is compliant with the initial objectives.

Design of the secondary circuit for the WCLL BB option of the EU DEMO power plant based on the new Energy Map

EU-DEMO will be a DEMOnstration Fusion power plant that is being designed to demonstrate the production of electricity from nuclear fusion at the level of a few hundred MW. The Primary Heat Transfer System (PHTS) transfers heat from the reactor to the Power Conversion System (PCS, the secondary circuit), responsible for converting the thermal power into electricity. According to the recent DEMO Energy Map data, In-Vessel components such as Breeding Blanket (BB), Divertor (DIV), Limiters, Vacuum Vessel and Auxiliaries feature uncertainties in generated thermal power. Some of them serve PCS as heat sources, namely: BB and DIV. Two reference concepts of the BB have been selected as a result of the DEMO pre-conceptual studies: the Helium Cooled Pebble Bed and the Water Cooled Lithium-Lead (WCLL). Because of the mentioned uncertainties regarding the distribution of energy released from fusion among the different reactor heat sources, several possible operational scenarios are considered. One of the challenges faced by the DEMO plant designers is the pulsed operation of the tokamak: plasma burn pulses (2 h) will be separated by dwell phases (10 min). To compensate drastic power reduction of the reactor heat sources during dwell, an Energy Storage System (ESS) of different sizes is included in considered DEMO plant configurations.Our work is focused on the design and analysis of the DEMO PCS circuit for the option WCLL BB with the small ESS. We create the GateCycle model of the “maximum of maximum” PCS cycle, in which the plant components are sized to incorporate the maximum power of all the reactor heat sources. Operation of the PCS circuit during the pulse (in the “Design” mode) and during dwell (in the “Off-design” mode with the fixed design of all the circuit components) is simulated, to verify the system performance and the potential feasibility of the considered PCS concept.

Water distillation for coolant purification system of DEMO water-cooled lithium lead breeding blanket

During the DEMO pre-concept design phase, a review of existing and under development technologies suitable for the Coolant Purification System (CPS) of the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) was carried out and two tritium recovery strategies were individuated, named on-line and off-line. The on-line concept relies on the continuous by-pass and treatment of a certain primary coolant fraction into CPS whereas the off-line scheme foresees the discharge of the whole primary water inventory after one year of DEMO operation, that will be treated in an external facility inside the tritium plant. Among these, the on-line strategy has been recently recommended as the most suitable for lowering the tritium inventory within BB and Primary Heat Transfer System (PHTS). Furthermore, such solution does not need high decontamination factor since treated water is not discharged in the environment, but it is redirected again into PHTS. In such condition, Water Distillation (WD) could be a promising technology with intrinsic simplicity and safety. The present work deals with the feasibility assessment of WD for DEMO WCLL BB CPS, investigating its capability to meet water CPS duty. A parametric study has been preliminary carried out, analysing the effect of selected parameters (i.e., tritium concentration at CPS inlet, tritium concentration in the enriched stream, efficiency of anti-permeation barriers, and operative conditions) on the WD sizing and operation. Then, a preliminary reference design has been individuated and verified with a dynamic model developed in MATLAB environment. Finally, the process flow diagram of the system has been derived and a preliminary sizing of auxiliary components is provided.

Design of a Prototypical Mock-Up for the Experimental Investigation of WCLL First-Wall Performances

A large research effort is currently ongoing within the framework of the EUROfusion consortium for the study and design of a water-cooled lithium–lead (WCLL) breeding blanket (BB). This concept will be tested in ITER through the installation of a test blanket module (TBM) and it is one of the two candidates adopted as driver BBs in DEMO. In this framework, at the ENEA research centre of Brasimone, the realization of the experimental platform, W-HYDRA, is envisaged. The platform is dedicated to the support of the development of WCLL BB and ITER TBM and the investigation of the DEMO balance of plants. One of the most important experimental infrastructures is the water-loop facility, the aim of which is to provide water at a high pressure and temperature (PWR conditions), with a sufficient mass-flow rate and power for the experimental testing of BB and TBM components. The facility will be equipped with a vacuum chamber and an electron beam gun for the reproduction of high surface heat flux on plasma-facing components. In the present work, the design of a prototypical mock-up (MU) of the WCLL BB first wall is described. The MU is used to investigate the thermal, hydraulic and structural behavior of the current first-wall design under relevant heat loads at the expected operational conditions. The delineation of the main experimental test’s features and the instrumentation needed is assessed in the paper. A preliminary CFD calculation on the prototypical MU and the computational results are also presented.

Modeling the Transport of Activated Corrosion Products in the WCLL PbLi Loop for ITER and the EU DEMO With the GETTHEM Code

This work presents the results of the implementation in the GEneral Tokamak THErmal-hydraulic Model of available models for the generation and transport of any dispersed material in the PbLi eutectic mixture circulating in the Water-Cooled Lithium-Lead Breeding Blanket of the EU DEMO fusion reactor, with main focus on Activated Corrosion Products as solid suspension. A simple test case is used to show that the distribution of the concentration of activated corrosion products at any point of the PbLi loop, both in the water-cooled lithium-lead breeding blanket and in the related ITER Test Blanket System, can be determined by the model. The results obtained with this tool can be useful not only for radiological safety purposes, but also because activated corrosion products may affect the PbLi flow itself and the efficiency of the tritium removal system, with consequences on the achievable Tritium Breeding Ratio. A rigorous verification of the model is also performed.

Transient analysis of a locked rotor/shaft seizure accident involving the EU-DEMO WCLL Breeding Blanket primary cooling circuits

The EU-DEMO Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) main subsystems to be cooled are the Breeder Zone (BZ) and the First Wall (FW). Each subsystem will be equipped with an independent Primary Heat Transfer System (PHTS). Within the framework of the EUROfusion Work Package Breeding Blanket research program, several accidents belonging to the category of “Decrease in Coolant System Flow Rate” were studied. The activity was aimed at evaluating the blanket and primary cooling systems thermal-hydraulic performances during such transient conditions. A complete model including the BB and related PHTS circuits has been developed at Sapienza University of Rome. A modified version of RELAP5/Mod3.3 system code has been used to perform the calculations. The simulation results showed that a locked rotor/shaft seizure of a BZ or a FW main coolant pump is the most challenging scenario. BZ and FW system behavior has been analyzed following this initiating event with the goal of the design improvement and to individuate the need for preventive measures. The influence of loss of off-site power on the accident evolution has also been investigated. Moreover, management strategies have been proposed for different reactor components. Calculations demonstrate that the current blanket and PHTS design is appropriate to cope with these kinds of accident scenarios.

Compatibility of niobium, vanadium and P22 steel in high temperature flowing LiPb

The LiPb loops of the WCLL BB (Water Cooled Lithium-Lead Breeding Blanket) have the primary aim of routing tritium-rich LiPb from the Breeding Blanket to the Tritium Extraction System of DEMO Reactor. Besides, the loops maintain LiPb in the desired conditions and purify it from corrosion and activated products and from helium generated by the neutron transmutation of lithium. For some of these tasks the loops need special functional materials. Among them, niobium and vanadium are widely considered for the tritium extraction technologies for their high permeability to tritium. In addition, the loops will be made of structural materials capable of withstanding the operative conditions and the aggressive environment. The current reference material for the piping is the ASTM A335 Grade P22. This paper investigates the compatibility at 500 °C of niobium, vanadium and P22 with LiPb flowing at 0.5 m/s. This velocity is the maximum foreseen in the LiPb loops, while the temperature was chosen following a conservative approach, as the maximum expected temperature is 450 °C in the BB and 330 °C in the LiPb loops. The specimens of the three materials were exposed for 1,000, 2,000 and 4,000 h in IELLLO LiPb facility at ENEA Brasimone Research centre. After the exposure, the specimens were examined by SEM-EDX, both in cross-section and surface view, XRD analyses were performed in order to investigate potential corrosion compounds formed. Weight loss analyses were also performed on niobium specimens, after chemical cleaning from adherent LiPb. The analyses highlighted the impact of corrosion products and their interaction with refractory metals (Ni and Fe) and P22 steel (Cr, after redeposition).

Development and application of an alternative modelling approach for the thermo-mechanical analysis of a DEMO water-cooled lithium lead breeding blanket segment

In the frame of the EUROfusion research activities devoted to the design of the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead BB (WCLL) concept is one of the candidates currently assessed in EU. To this end, an intense research campaign is ongoing to develop a robust geometric configuration for the WCLL BB Central Outboard Segment (COB). Since the current reference design of the WCLL COB segment is not mature enough to allow a full thermal-hydraulic assessment, an alternative procedure aimed at obtaining a thermal field for the whole segment without performing its complete thermal-hydraulic analysis is presented and applied in this work. The scope of the work is to obtain a thermal field for the whole WCLL COB segment detailed enough to ensure a reliable prediction of the secondary stress spatial distribution, which is pivotal for the segment structural analysis.The procedure is articulated in three steps. Firstly, a thermal analysis of the equatorial region of the WCLL COB is performed, in order to calculate the thermal field in the most loaded region of the segment. Then, a multi-region interpolation is carried out in order to obtain a set of polynomial functions of the radial and toroidal variable, accurately representing the thermal field of the assessed equatorial BB slice. Finally, the calculated functions are applied to the whole COB to impose a realistic thermal field to the segment for the prediction of the stress, strain and displacement fields.Since the mesh for the analysis of the whole segment is expected to be coarser than that used for the assessment of its equatorial region, in this paper a mesh independence assessment of the interpolating functions is reported. In particular, their predictive power both in terms of temperature and secondary stress (i. e. thermal-induced stress) is assessed taking into account a decreasing level of detail of the adopted spatial discretization grid.The obtained results, widely reported and discussed hereafter together with the models and the adopted assumptions, show a great predictive power of the adopted polynomial functions, with a good mesh independence of the calculated temperature and stress values.