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.

RELAP5/Mod3.3 thermal-hydraulics characterization of the steam generator mock-up during operational transients in STEAM facility in support of the design of the DEMO WCLL BoP

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-thermal-HYDRAulic (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 thermo-mechanical 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.

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.

Challenges towards an acceleration in stellarator reactors engineering: The dual coolant lithium–lead breeding blanket helical-axis advanced stellarator case

Under the ambitious EUROfusion mission of “bringing the stellarator line to technological maturity”, the development of a Dual Coolant Lithium–Lead (DCLL) Breeding Blanket (BB) for a Helical-Axis Advanced Stellarator (HELIAS) started. The BB is the crucial systems to achieve tritium self-sufficiency in fusion power plants. The DCLL has potentialities to answer the challenges posed by complex HELIAS configuration, having liquid breeder and decoupled first wall (FW) and breeding zone circuits. Both characteristics are being empowered to simplify remote maintenance and integration of the BB segments. Hence, a “quasi-toroidal” BB segmentation has been proposed to minimize the magnetohydrodynamic pressure drop eschewing complex electrical insulating components while hindering maintenance. Such controversy motivated the use of a detached FW small panels manageable through ports. On this basis, a decoupled Capillary Porous System FW has been assessed under thermal-hydraulics and neutronics point of view. Moreover, ad-hoc tools have been developed for the parametrization of models and convergence toward a viable DCLL HELIAS design, speeding-up the coupling of computer-aided design modelling and analyses. Such novel strategies are described in view of a potential acceleration of the stellarator reactor engineering and by providing an overview of the stellarator challenges inside a global context.

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.

Embrittlement of WCLL blanket and its fracture mechanical assessment

In the European fusion programme, the Water Cooled Lithium Lead breeding blanket (WCLL BB) uses EUROFER as a structural material cooled with water at temperatures between 295 °C–328 °C and a pressure of 155 bar. The WCLL BB will be significantly irradiated (>2 dpa), while some parts will not receive significant heat loads, e.g. the sidewalls or the back-supporting structures. The irradiation, together with the irradiation temperature of EUROFER below 350 °C, produces a shift of the ductile-to-brittle-transition temperature (DBTT) to levels above room temperature at neutron doses, causing material damage as low as 2–3 dpa. Even though the DBTT does not reach the operating temperature level, brittle/non-ductile fracture is a concern during in-vessel maintenance when the BB temperature is below the DBTT. Two loading scenarios were identified as severe in this respect: (i) re-pressurization of the WCLL BB cooling loop after in-vessel maintenance, and (ii) dead weight loads during lifting of the BB segment. The embrittlement of the WCLL BB was investigated by quantifying the local DBTT shift in its parts based on current knowledge of the embrittlement behaviour of EUROFER under neutron irradiation. Therefore, a suitable, not overly conservative procedure was derived considering dpa damage and transmuted helium effects. The results demonstrate the ability to identify the 3D spread of the severely embrittled zones in the structure whose impact on the structural integrity was assessed considering the risk of brittle/non-ductile fracture. Thereby, the fracture mechanics approach established in nuclear codes was applied assuming its applicability to EUROFER. The embrittled zones in the first wall (FW) and its sidewalls pass the criteria when assessing the relatively low stresses resulting from the coolant pressure. The assessment was then continued considering stresses appearing in the FW during maintenance, in particular, when lifting the BB segment and transporting it out of the vacuum vessel. In this context, the maximum tolerable flaw sizes were determined in a parameter study considering designs of the FW with different cooling channel wall thicknesses.

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.

Neutronic Assessments towards a Novel First Wall Design for a Stellarator Fusion Reactor with Dual Coolant Lithium Lead Breeding Blanket

The Stellarator Power Plant Studies Prospective R&D Work Package in the Eurofusion Programme was settled to bring the stellarator engineering to maturity, so that stellarators and particularly the HELIAS (HELical-axis Advanced Stellarator) configuration could be a possible alternative to tokamaks. However, its complex geometry makes designing a Breeding Blanket (BB) that fully satisfies the requirements for such a HELIAS configuration, which is a difficult task. Taking advantage of the acquired experience in BB design for DEMO tokamak, CIEMAT is leading the development of a Dual Coolant Lithium Lead (DCLL) BB for a HELIAS configuration. To answer the specific HELIAS challenges, new and advanced solutions have been proposed, such as the use of fully detached First Wall (FW) based on liquid metal Capillary Porous Systems (CPS). The proposed solutions have been studied in a simplified 1D model that can help to estimate the relative variations in Tritium Breeding Ratio (TBR) and displacement per atom (dpa) to verify their effectiveness in simplifying the BB integration and improving the machine availability while keeping the main BB nuclear functions (i.e., tritium breeding, heat extraction and shielding). This preliminary study demonstrates that the use of FW CPS would drastically reduce the radiation damage received by the blanket by 29% in some of the selected configurations along with a small decrease of 4.9% in TBR. This could even be improved to just a 3.8% TBR reduction by using a graphite reflector. Such an impact on the TBR is considered affordable, and the results presented, although preliminary in essence, have shown the existence of margins for further development of the FW CPS concept for HELIAS, as they have been not found, at least to date, to be significant showstoppers for the use of this technological solution.

Advances, Challenges, and Future Perspectives of Microwave Reflectometry for Plasma Position and Shape Control on Future Nuclear Fusion Devices.

Providing energy from fusion and finding ways to scale up the fusion process to commercial proportions in an efficient, economical, and environmentally benign way is one of the grand challenges for engineering. Controlling the burning plasma in real-time is one of the critical issues that need to be addressed. Plasma Position Reflectometry (PPR) is expected to have an important role in next-generation fusion machines, such as DEMO, as a diagnostic to monitor the position and shape of the plasma continuously, complementing magnetic diagnostics. The reflectometry diagnostic uses radar science methods in the microwave and millimetre wave frequency ranges and is envisaged to measure the radial edge density profile at several poloidal angles providing data for the feedback control of the plasma position and shape. While significant steps have already been given to accomplish that goal, with proof of concept tested first in ASDEX-Upgrade and afterward in COMPASS, important, ground-breaking work is still ongoing. The Divertor Test Tokamak (DTT) facility presents itself as the appropriate future fusion device to implement, develop, and test a PPR system, thus contributing to building a knowledge database in plasma position reflectometry required for its application in DEMO. At DEMO, the PPR diagnostic's in-vessel antennas and waveguides, as well as the magnetic diagnostics, may be exposed to neutron irradiation fluences 5 to 50 times greater than those experienced by ITER. In the event of failure of either the magnetic or microwave diagnostics, the equilibrium control of the DEMO plasma may be jeopardized. It is, therefore, imperative to ensure that these systems are designed in such a way that they can be replaced if necessary. To perform reflectometry measurements at the 16 envisaged poloidal locations in DEMO, plasma-facing antennas and waveguides are needed to route the microwaves between the plasma through the DEMO upper ports (UPs) to the diagnostic hall. The main integration approach for this diagnostic is to incorporate these groups of antennas and waveguides into a diagnostics slim cassette (DSC), which is a dedicated complete poloidal segment specifically designed to be integrated with the water-cooled lithium lead (WCLL) breeding blanket system. This contribution presents the multiple engineering and physics challenges addressed while designing reflectometry diagnostics using radio science techniques. Namely, short-range dedicated radars for plasma position and shape control in future fusion experiments, the advances enabled by the designs for ITER and DEMO, and the future perspectives. One key development is in electronics, aiming at an advanced compact coherent fast frequency sweeping RF back-end [23-100 GHz in few μs] that is being developed at IPFN-IST using commercial Monolithic Microwave Integrated Circuits (MMIC). The compactness of this back-end design is crucial for the successful integration of many measurement channels in the reduced space available in future fusion machines. Prototype tests of these devices are foreseen to be performed in current nuclear fusion machines.

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.

Development of a thermal-hydraulic model of the EU-DEMO Water Cooled Lithium Lead Breeding Blanket Primary Heat Transport System

The EUROfusion consortium is developing the project of a DEMOnstration Fusion Reactor (EU-DEMO) which would follow ITER in the pathway towards the quest for the exploitation of fusion energy. EU-DEMO has been conceived to deliver net electric power to the grid. Therefore, proper critical evaluations of the tokamak cooling and power conversion systems are needed because they play a pivotal role in the design and licencing of the overall plant. The EU-DEMO reactor will be based on the tokamak concept and, as such, it is supposed to undergo a pulsed duty cycle under normal conditions, which might challenge the qualified lifetime of the main equipment inducing undue thermal and mechanical cycling. Moreover, the EU-DEMO plasma control strategy postulates the possible occurrence of planned and off-normal plasma overpower transients that might jeopardise the structural integrity of the plasma facing components. It is, therefore, of paramount importance to have appropriate tools to reproduce the thermal-hydraulic behaviour of tokamak cooling systems during major operational and accidental scenarios in a realistic and reliable way. In this context, University of Palermo in cooperation with EUROfusion has developed a finite volume model of the Primary Heat Transport System (PHTS) feeding the EU-DEMO Water Cooled Lithium Lead Breeding Blanket (WCLL BB). The activity has been led following a theoretical–computational approach based on the adoption of the TRACE thermal-hydraulic system code. Particular attention has been paid to capturing all the main geometrical, hydraulic and heat transfer features characterising both in-vessel and ex-vessel components. Preliminary analyses have also been carried out to check the code’s predictive potential in fusion relevant applications. Models, assumptions, and outcomes of the analyses are herewith reported and critically discussed.

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.

Tritium Transport in WCLL Outboard Breeding Blanket of EU-DEMO Reactor Under Pulsed Plasma Operation

The accurate prediction of the tritium inventory and permeation fluxes in the water-cooled lithium-lead (WCLL) breeding blanket of the European DEMO fusion reactor is a key aspect for future thermonuclear power plant licensing. In this context, a tritium transport analysis is essential for the evaluation of the tritium retention in LiPb (15.7 at. % Li) and in the structural components and tritium permeation fluxes into the cooling water. This study presents a COMSOL Multiphysics model of a portion of the minimum functional unit of the outboard equatorial module of the WCLL. The neutronics analysis, performed with the MCNP Monte Carlo transport code, allows for the assessment of the tritium generation rate and the volumetric power deposition in the lithium-lead, both of which are used as an input for tritium transport modeling. Moreover, the buoyancy forces and the magnetohydrodynamic effect are also included. In order to take into account a pulsed operation for the DEMO reactor, a suitable algorithm was developed. The results show peak velocities in LiPb up to 24.1 mm with a maximum temperature at the first wall (FW) of 441°C. Most of the tritium is concentrated between the FW and the baffle plate, with partial pressures between 0.15 and 0.4 Pa considering Aiello’s correlation and in-house measurements of Sieverts’ constant on the HyPer-QuarCh II (HQ-II) device, whereas from 70 to 180 Pa adopting Reiter’s correlation. This has an impact also on the inventory in the Eurofer and in permeation fluxes to the water pipes. For Reiter’s correlation, the inventories in the Eurofer are about one order of magnitude lower with respect to LiPb, and the inventory in water is five orders of magnitude higher with respect to HQ-II and Aiello’s correlation.

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.

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.

Lithium-Lead/water interaction: LIFUS5/Mod3 series E tests analysed by SIMMER-III coupled with RELAP5

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.