Water Cooled Lithium Lead Research Articles

"Exploring the role of lithium in Al2O3 tritium permeation barrier development: A crucial challenge for fusion reactor progress"

The development and application of robust tritium permeation barriers (TPBs) are crucial for a safe and economical fusion reactor operation because tritium permeation could cause serious safety problems, such as the brittleness of the structural material, fuel loss, and radioactive contamination. Coatings may effectively inhibit the permeation through the structural materials of tritium fuel generated inside the breeding blanket (BB) reducing the loss of lithium inventory contamination of the ancillary systems and the exposure of workers and the population. Al2O3 has been preliminary selected as the EU candidate for tritium permeation reduction of structural steels in the Water-cooled lithium–lead breeding blank (WCLL) concept, because of its very high permeation reduced factor (PRF). However, because of the chemical reactivity of lithium, several authors report the formation of Lithium aluminates after exposure to PbLi baths, based on different experimental techniques. In this contribution, we will show recent results about the quantification of the extent of the reaction and the thickness of the reacted layer and discuss the possible effects, that may happen under neutron irradiation based on the species produced (T and He) in the nuclear reactions (n, T) and (n, α) with 6Li and 7Li, with the Li atoms in the reacted layer To do that, the atomic density of produced T and He per second out of the transmutation reaction and the He/T ratio are calculated for diverse Li contents. Subsequently, we implant He into the coatings and characterize the desorption rate and the morphology. Finally, discussion and recommendations for further coating development will be presented.

The transient thermal ageing of Eurofer 97 by mitigated plasma disruptions

Plasma disruptions in a commercial-scale tokamak will impose high magnitude, short-duration thermal loads on its plasma-facing first wall. Despite a plethora of mitigation measures, these severe, off-normal events are likely to briefly expose the structural materials of the first wall to significant temperature excursions. Eurofer 97, the reference structural material for the plasma-facing first wall of the EU DEMO tokamak, is a reduced activation 9Cr steel with a normalised and tempered ferritic/martensitic microstructure. Repeated exposure to the thermal effects of disruptions over the operating lifetime of a reactor may promote a cumulative evolution of Eurofer 97′s microstructure, affecting key material properties crucial to first wall performance. This novel transient thermal degradation mechanism has been explored via a laser-based transient heating experiment, supported by time-dependent finite element thermal analysis studies of DEMO’s water-cooled lithium–lead first wall during a mitigated plasma disruption. Transient-affected samples of Eurofer 97 were characterised via scanning and transmission electron microscopy techniques (SEM/TEM), including electron backscatter diffraction (EBSD), energy-dispersive X-ray spectroscopy (EDX), and selective area electron diffraction (SAED). Microhardness and magnetisation testing data are also presented. A single 700 °C thermal transient was found sufficient to coarsen and partially recrystallise Eurofer 97′s tempered martensite sub-grains at the tungsten-Eurofer 97 interface. Further transient exposure at 700 °C resulted in the significant growth of equiaxed grains, the nucleation of intergranular Cr-rich M7C3 and M23C6 carbides, and the coarsening of V-rich and intra-granular Ta-rich MX precipitates. The microstructural effects of 850 °C transients are also reported. Notably, after 1,000 transients at 850 °C the hardness of Eurofer 97 was found to have decreased by 32 %.

W-HYDRA: a new experimental platform for the Water-Cooled Lead Lithium Breeding Blanket

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

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 MHD module development and validation: WCLL-TBM mock-up model

Magnetohydrodynamic (MHD) phenomena significantly impact the design of magnetic-confinement fusion reactors, particularly in the context of breeding blankets (BB) utilizing liquid metals (LMs) as working fluids. These effects arise from the interaction between the electro-conductive flowing metal and the magnetic field used for plasma confinement in the reactor chamber. Induced electrical currents in the liquid generate Lorentz forces, thereby altering flow behaviour in comparison to standard hydrodynamic conditions. For example, MHD effects modify velocity distribution and mass transport within ducts, amplify pressure losses, and influence heat transfer mechanisms. Accurate estimation of these impacts is crucial for the effective design of a liquid metal breeding blanket. While computational tools are essential for fusion-related physical analyses, no comprehensive MHD code currently exists for simulating all relevant phenomena in a liquid metal blanket. In this context, models predicting both distributed and concentrated MHD pressure drops have been integrated into the thermal-hydraulic system code RELAP5/Mod3.3. The Verification and Validation (V&V) process compares code results to direct numerical simulations and experimental data. For validation, RELAP5 recreates experimental results of a Water Cooled Lithium Lead test blanket module at magnetic field intensities ranging from Ha=500−3000, confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range.

Experimental investigation of MHD flows in a WCLL TBM mock-up

Experiments have been performed in the MEKKA laboratory of the Karlsruhe Institute of Technology to characterize the influence of a magnetic field on liquid metal flows in a scaled mock-up of the water-cooled lead-lithium test blanket module of ITER. The test section consists of eight breeder units that are fed and drained by long electrically coupled manifolds oriented along the poloidal direction. Pressure differences between several points of the mock-up have been recorded for various liquid metal flow rates and strengths of the imposed magnetic field. The main contributions to the total pressure drop in the test section have been identified as a function of characteristic flow parameters. For sufficiently strong magnetic fields, the experimental data shows that the non-dimensional pressure loss is practically independent of the flow rate, i.e. inertia forces become negligible. The experiments also confirm previous conclusions of magnetohydrodynamic experiments performed in a scaled-down mock-up of a helium-cooled lead-lithium blanket, namely that the largest pressure drop in the blanket module originates from the flow in the distributing and collecting manifolds. Moreover, the experimental study demonstrates that the current manifold design does not allow the flow to be uniformly distributed among all the breeder units.

Instrumentation integration in the European Test Blanket Modules

Amongst the four different Test Blanket Modules (TBM) that shall be tested in ITER, the European domestic agency (Fusion For Energy) participates in two concepts: the Water-Cooled Lithium Lead (WCLL) and the Helium-Cooled Ceramic Pebbles (HCCP). A progressive set of activities including implementation and integration of instrumentation has been conducted from preliminary design stage. The current work is mainly focused on the thermal and mechanical instrumentation, although other sensors are also considered. The integration of instrumentation in the TBMs is challenging; consequently, the number and position of the sensors should be carefully optimized. An important advance of the current work is the use of reconstruction algorithms to provide an estimation of the reconstruction ability of the different sensor sets. These algorithms comprise a series of mathematical tools which use model results and sensor measurements to infer different variables at positions where no in-situ measurement is available. Apart from the use of these algorithms, the final decision on sensor location has been conducted from a holistic point of view considering the different perspectives and integration challenges.

Assessment of the relevancy of ENEA Water Loop facility with respect to ITER WCLL TBS Water Cooling System by considering their thermal-hydraulic performances

The Water-Cooled Lead-Lithium (WCLL) is one of the two candidate concepts for the Breeding Blanket (BB) of DEMO. A Test Blanket Module (TBM) together with its Water Cooling System (WCS) is going to be installed and tested in the ITER reactor. The WCS acts as primary cooling circuit of the TBM module, and it is designed to reproduce the water thermodynamic conditions expected at the DEMO BB inlet.During last years, ENEA and the DIAEE of Sapienza University of Rome have carried out the conceptualization of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone. The W-HYDRA platform is composed by three individual facilities called: Water Loop, Steam, and LIFUS5/Mod4. Water Loop replicates the salient thermal-hydraulic features of the ITER WCLL WCS, and it is equipped with a test section placed inside a Vacuum Vessel (VV) to investigate mock-ups of the whole TBM or its individual parts.This paper assesses the relevancy of the WL facility with respect to ITER WCLL TBM System comparing their thermal-hydraulic performances during selected operational and accidental conditions. Two RELAP5/Mod3.3 models were developed, and the outcomes of the simulations showed a good agreement, thus demonstrating the effectiveness of WL facility to support the ITER TBM program.

Numerical and experimental analysis of magneto-convective flows around pipes

Liquid metal flows exposed to intense magnetic fields play a fundamental role in the development of blankets for nuclear fusion reactors, where they serve to produce the plasma-fuel component tritium and transport the generated heat. When the liquid metal circulates in the blanket, it interacts with the plasma-confining magnetic field leading to the induction of electric currents. Electromagnetic forces and thermal buoyancy affect significantly velocity and pressure distributions. The resulting magneto-convective flow has peculiar features that have to be taken into account in order to evaluate heat transfer properties in the liquid metal. In the breeding zone of the water-cooled lead lithium blanket concept, the volumetric heat released in the liquid metal is removed by water-cooled circular pipes that represent obstacles for the liquid metal flow. In order to improve the understanding of the underlying physical phenomena and obtain a database for code validation, model experiments have been performed to investigate magneto-convective flow and heat transfer at two differentially heated parallel horizontal tubes immersed in a box filled with liquid metal. Among other properties, electric potential has been recorded on the surface of the test-section and compared with 3D numerical simulations. Computational results provide the basis for interpretation of the measured data, since they allow differentiating between flow-induced electric potential and thermoelectric effects.

Thermofluid-dynamic and thermal–structural assessment of the EU-DEMO WCLL “double bundle” Breeding Blanket concept left outboard segment

As part of the activities performed by the DEMO Central Team (DCT) to study alternative configurations of the Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB), to overcome the open issues that emerged at the end of the Pre-Conceptual Design phase, a new concept, the WCLL BB “double bundle” (db), was developed. This concept adopts an array of db tubes poloidally distributed inside the breeding zone, mimicking the arrangement inside heat exchangers. These are coaxial pipes the gap between which is filled with PbBi (or alternatively with gas and fins), which avoids direct contact between PbLi and water in case of in-box LOCA events. Similarly, the First Wall channel cooling water is separated from the PbLi. The use of PbBi as an inert fluid makes the Segment Box (SB) sizing less stringent: in the case of an in-box LOCA due to a break of the water cooling circuit inside the BB SB, the pressurized volume is only the PbBi one, instead of the whole SB volume as in the reference WCLL design configuration. Consequently, the total amount of steel inside the BB can be reduced, with benefits in terms of increased tritium breeding ratio. Within this framework, University of Palermo, in support of the DCT and under the EUROfusion action, preliminarily analysed the steady-state thermal, thermo-hydraulic and normal and off-normal thermo-mechanical performances of this concept, following a theoretical-numerical approach based on finite volume and finite element methods, and adopting the ANSYS CFX and ANSYS Mechanical codes. The analyses carried out allowed a preliminary optimization of the design, improving the thermal and thermo-mechanical performances of the WCLL-db BB, in compliance with the design requirements, while reducing the total steel amount. Models, assumptions and main results are herewith reported and critically discussed.

An integral methodological framework for the thermo-mechanical analysis and structural integrity assessment of the European TBM sets

Two European Test Blanket Modules (TBM) concepts are being developed by F4E, the Water-Cooled Lithium Lead (WCLL) and the Helium-Cooled Ceramic Pebbles concept (HCCP, in collaboration with ITER KOREA). For both concepts, the ESTEYCO Mechanics team is responsible for designing and analyzing the TBM-box, which serves as the plasma-facing module for tritium breeding, and the TBM Shield in the WCLL concept. The design-by-analysis activities for these TBM-Sets are particularly challenging due to several factors. These include diverse and complex loads such as thermal, electromagnetic, pressure, and seismic loads, intricate failure modes to be assessed, the use of a new structural material (EUROFER 97) with limited available information, and the need to work within tight compliance margins in a restricted design space. Over the years, ESTEYCO has developed advanced methodologies to virtually test various design solutions systematically, coherently, and efficiently, aligning with the applicable nuclear design code (RCC-MRx). This paper provides an overview of key developments in simulating the TBM-Set under a comprehensive set of loads and subsequently assessing its structural integrity in an exceptionally detailed and comprehensive manner. The authors have created novel automated tools that encompass the entire process. These tools address tasks like distributing nuclear thermal loads, simulating cooling system behavior under operational conditions, characterizing and distributing electromagnetic loads, implementing code assessment rules for numerous evaluation points, and developing robust visualization tools for result analysis and validation.

Major improvements in the WCLL TBM design towards next review gates

The European domestic agency (Fusion for Energy) is developing the WCLL-TBS (Water Cooled Lithium Lead Test Blanket System) concept for its implementation and testing in ITER. The system is based on the use of the eutectic Pb–15.7Li enriched at 90 % in 6Li as tritium breeder and tritium carrier. Water is used as coolant with current target nominal inlet/outlet temperatures of 295/328 °C and 15.5 MPa pressure (similar to PWR conditions). The Test Blanket Module –TBM– is the in-vessel component where tritium generation will take place and whose structural material is EUROFER 97. Its design has been significantly improved in the last years. These design improvements have focused on three major areas: (i) overcoming some weaknesses of the previous design in the rear manifold area associated to the stiffening rod concept, (ii) strengthening the link with manufacturing developments and (iii) optimizing the amount of EUROFER 97 in the TBM box, so that the design converges to the limits imposed by the ITER project requirements. The work presented in this paper describes the latest design evolutions achieved in the WCLL TBM design with respect to that presented at CDR and, in particular, the major re-design of the TBM in the manifold area based on the ‘crossing vertical’ stiffening plate concept proposed and developed by the ESTEYCO Mechanics design team in collaboration with F4E.

Maintenance study on WCLL-TBS ancillary systems in process room

The Water-Cooled Lithium Lead Test Blanket System (WCLL-TBS) is one of the European TBS candidates for being installed and operated in ITER. The ongoing activities toward the Preliminary Design Review of the WCLL-TBS Ancillary Systems include the reallocation of their units in the reserved space, also in the view of the proper execution of the maintenance activities. Starting from the available documentation (System Design Description, Process Flow Diagrams, Process and Instrumentation Diagrams), detailed 3D CAD models have been developed for two Ancillary Systems to be installed in the Process room: the Tritium Extraction System (TES) and the Tritium Accountancy System (TAS). In the Process room (PR), the largest part of the TES process units is placed inside a Glove-box, as enclosure to avoid the contamination of the room in case of radioactive release due to failure of units. Based on the expected radioactivity content, all the TAS units are placed inside the Glove-box. Maintainability studies have been performed, focusing on the preventive (i.e. planned) and corrective (i.e. required by failures) maintenance tasks on the units installed inside the Glove-box. They were based on the 3D CAD models and supported using Virtual Reality as intuitive and immersive human-computer interface. They provided insights for the iterative improvement and consolidation of the Systems design and information for the evaluation of their availability.

Optimized Water Distillation Layout for Detritiation Purpose

Tritium permeation constitutes a key issue for the future EU-DEMO, especially in the Breeding Blanket (BB) where fusion energy must be delivered to the Primary Heat Transport System (PHTS) and where tritium must be bred. Currently, the mitigation strategy of the tritium permeation from BB into primary coolant is based on the adoption of anti-permeation barriers and on the operation of the Coolant Purification System (CPS). This system must ensure a tritium removal rate from the primary coolant equal to the BB permeation rate at a target tritium-specific activity inside the PHTS. In the case of the Water-Cooled Lithium Lead (WCLL) BB, water distillation was selected as the most promising technology for the primary coolant detritiation due to its intrinsic simplicity and safety. Nevertheless, power consumption was recognized as a relevant concern. For this reason, the present work aims at investigating possibilities to reduce power consumption of the water CPS implementing Heat Pump-Assisted Distillation (HPAD) concepts. To do this, a review of the HPADs developed in the chemical industry was carried out, and the best options for the water CPS were identified based on qualitative considerations. Then, a quantitatively assessment of the best solution in terms of power consumption and tritium inventory was performed with the commercial numerical tool Aspen Plus. Finally, the Mechanical Vapor Recompression (MVR) concept was recognized as the most promising solution, ensuring a power saving of around 80% while keeping a limited tritium inventory.

Main Nuclear Responses of the DEMO Tokamak with Different In-Vessel Component Configurations

Research and development of the DEMOnstration power plant (DEMO) breeder blanket (BB) has been performed in recent years based on a predefined DEMO tritium breeding ratio (TBR) requirement, which determines a loss of wall surface due to non-breeding in-vessel components (IVCs) which consume plasma-facing wall surface and do not contribute to the breeding of tritium. The integration of different IVCs, such as plasma limiters, neutral beam injectors, electron cyclotron launchers and diagnostic systems, requires cut-outs in the BB, resulting in a loss of the breeder blanket volume, TBR and power generation, respectively. The neutronic analyses presented here have the goal of providing an assessment of the TBR losses associated with each IVC. Previously performed studies on this topic were carried out with simplified, homogenized BB geometry models. To address the effect of the detailed heterogeneous structure of the BBs on the TBR losses due to the inclusion of the IVCs in the tokamak, a series of blanket geometry models were developed for integration in the latest DEMO base model. The assessment was performed for both types of BBs currently developed within the EUROfusion project, the helium-cooled pebble bed (HCPB) and water-cooled lead–lithium (WCLL) concepts, and for the water-cooled lead and ceramic breeder (WLCB) hybrid BB concept. The neutronic simulations were performed using the MCNP6.2 Monte Carlo code with the Joint Evaluated Fission and Fusion File (JEFF) 3.3 data library. For each BB concept, a 22.5° toroidal sector of the DEMO tokamak was developed to assess the TBR and nuclear power generation in the breeder blankets. For the geometry models with the breeder blanket space filled only with blankets without considering IVCs, the results of the TBR calculations were 1.173, 1.150 and 1.140 for the HCPB, WCLL and WLCB BB concepts, respectively. The TBR impact of all IVCs and the losses of the power generation were estimated as a superposition of the individual effects.

Overview of tritium management in WCLL test blanket system of ITER

A key aspect for the achievement of the fuel self-sufficiency of future fusion power plants is the management of tritium in the different stages of the reactor fuel cycle. In particular, one of the greatest challenges for the success of liquid metal breeder concepts such as the Water-Cooled Lithium-Lead (WCLL) is the extraction of tritium from the liquid LiPb eutectic alloy (15.7 at.%Li), which fulfils the main functions of tritium breeder, tritium carrier and neutron multiplier, and the processing in a form suitable for an accurate tritium accountancy. Within this frame, significant advancements have been carried out in the last years as regards the WCLL Test Blanket System (TBS) of ITER. Tritium is mainly produced inside the LiPb and extracted through the Tritium Extraction Unit (TEU). The reference technology for the TEU of the WCLL-TBS is the Gas-Liquid Contactor (GLC) in the packed-column configuration. Tritium is then concentrated in ZrCo getter beds before being sent to the accountancy system where different technologies for both static and dynamic accountancy are adopted in order to provide precise and reliable tritium measurements, which are an essential need in view of an application to the European DEMO reactor. In this paper, the present status and the design solutions foreseen for tritium management in WCLL-TBS are presented and discussed. The implementation and optimization of the conceptual design of the lithium-lead loop (LLP), the Tritium Extraction System (TES) and the Tritium Accountancy System (TAS) have been performed in order to manage tritium concentration in LiPb and in the ancillary systems.

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.

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.