Breeding Blanket Concepts Research Articles

Chemistry and relevant thermophysical properties of Pb84.3Li15.7 liquid solutions: Updated view and identification of new and coherent values

The liquid breeding blanket concept of a fusion reactor is based on the employment of the eutectic Lead-Lithium solution (LLE), where Lithium, enriched in 6Li, has the purpose of regenerating Tritium by reacting with neutrons, while Lead enhances the process acting as neutrons multiplier. Anyway, despite the huge interest on this binary system, LLE is not characterized yet by a fully consolidated description of its basic thermophysical properties and, in many cases, the spread of values reported so far in literature remains still significant. The purpose of this paper is hence to check again the many data available in literature, with a special attention to the ones more recently published, trying to explain and solve their inhomogeneity. Taking into account the general features of the Pb-Li interaction and particularly the not ideal behaviour of the Pb-Li liquid solutions, the following thermophysical properties of the LLE are dealt in detail: specific heat; density and volumetric thermal expansion coefficient; thermal conductivity and diffusivity; electrical resistivity; Sieverts' constant of Hydrogen. Based on the deep analysis of all the reported experimentation and through the adoption of several comparison criteria, the most trustable correlation is sought for each of the above properties; additionally, when a robust correlation couldn't be retrieved, a new, optimized, one has been proposed, capable also to foresee correctly the effect of small composition variations and to assure the internal coherence among linked properties.Specific heat and density values resulted at the end accurately described and not needing for additional experimentation; thermal properties and electrical resistivity can be evaluated with decent confidence, even if their uncertainty could be somehow reduced by future investigations; for Sieverts' constant it has not possible yet to identify a unique, trustable correlation, anyway the range of values correctly assumed up to 900K has been significantly restricted.

Design update of DEMO BOP for HCPB BB concept with an energy storage system

An option for the Intermediate Coupling Design (ICD) of the European Demonstration Fusion Reactor (DEMO) Balance of Plant (BOP) configuration for Helium-Cooled Pebble Bed (HCPB) Breeding Blanket (BB) concept consists of the Primary Heat Transport System (PHTS), the Intermediate Heat Transfer System (IHTS), using HITEC salt as coolant, equipped with a thermal Energy Storage System (ESS), and the Power Conversion System (PCS). The aforementioned BOP configuration is capable of producing electricity continuously during the pulse time (2 h) and dwell time (10 min) of DEMO plant operation.The present study reports on the current (2023) thermal system design of DEMO HCPB ICD BOP for 12 operational states defined by the DEMO Design Central Team (DCT) in 2021. It was shown that the PCS is able to accommodate all of the 12 operational points defined based on the uncertainties of the energy map, provided that an additional bypass line at the outlet of the Divertor Cassettes (DIV-CAS) Heat Exchanger (HX) towards the low-pressure part of the feedwater line (to the condenser in this investigation) is installed in the system. This line is activated only in the cases, where the feedwater flow to the low temperature DEMO heat sources HXs is needed to be higher than that necessary to remove BB power at the Steam Generators (SGs). In all the other cases, the bypass line remains unused. Additionally, perspective improvements of DEMO HCPB BOP regarding thermal ESS, as well as PCS are discussed.

Hyperfine interactions of paramagnetic radiation-induced defect centres in advanced ceramic breeder pebbles

Advanced ceramic breeder (ACB) pebbles consisting of 65 mol% lithium orthosilicate (Li4SiO4) and 35 mol% lithium metatitanate (Li2TiO3) are developed for tritium breeding in the European Union’s (EU) helium cooled pebble bed (HCPB) breeder blanket concept of the demonstration fusion power plant (DEMO). Electron-nuclear double resonance (ENDOR) spectroscopy is a powerful and widely used magnetic resonance technique that combines aspects of electron paramagnetic resonance (EPR) and nuclear magnetic resonance (NMR). In the present work, ENDOR spectroscopy was applied to investigate hyperfine (HF) interactions of paramagnetic radiation-induced defect centres (containing unpaired electrons) in the biphasic ACB pebbles after irradiation with 10 MeV accelerated electrons. To separate individual EPR signals of the formed and accumulated paramagnetic centres, isochronal annealing of the irradiated ACB pebbles followed by EPR spectra measurements at X- and Q-microwave frequency bands at room and low temperature were performed. Afterwards, X-band ENDOR spectra at low temperature were recorded for selected EPR signals to reveal HF interactions between the unpaired electron spins with the spins of magnetic 7Li and 1H isotope nuclei. The g-factor and HF coupling values determined from EPR and ENDOR spectra simulations provide novel insights into the local structure of paramagnetic hole- and electron-type centres in the irradiated ACB pebbles.

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.

Preliminary accident analyses of in-vessel LOCA for the chinese fusion test reactor HCCB blanket concept

For the helium-cooled ceramic breeder (HCCB) blanket concept improved in 2018 and the primary heat transfer system following China Fusion Engineering Test Reactor (CFETR) in 2019, an in-vessel loss of coolant accident (LOCA) has been investigated with the assumption of a break of the first wall (FW) coolant channels and a double-ended guillotine break of a helium cooling system (HCS) pipe in vacuum vessel (VV). The break leads to helium inject into the VV. After the accident occurs, the helium release leads a plasma disruption and pressure increase in the VV. This paper presents the analysis results of the in-vessel LOCA accidents in CFETR and gives the accident sequence compare with the break in the FW and header pipe. The accident analysis and influence compares can provide a reference for the iterative design.

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.

Overview of progress on water cooled ceramic breeder blanket in Japan

National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual fusion environment. QST has led the development of the WCCB TBM Concept in the program since the signing of the TBM Arrangement in 2014. The preliminary design of test blanket system (TBS) is on-going. Recently, the status of the preliminary design of TBS was assessed between QST and the ITER Organization (IO). So, the status of the latest TBS design and the DEMO relevancy is presented.

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.

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.

PbLi/Water Reaction: Experimental Campaign and Modeling Advancements in WPBB EUROfusion Project

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes.

Exploratory Thermo-Mechanical Assessment of the Bottom Cap Region of the EU DEMO Water-Cooled Lead Lithium Central Outboard Blanket Segment

The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two BB concept candidates to be selected as the driver blanket for the EU DEMO fusion reactor. In this regard, the development of a sound architecture of the WCLL Central Outboard Blanket (COB) Segment, ensuring the fulfilment of the thermal and structural design requirements, is one of the main goals of the EUROfusion consortium. To this purpose, an exploratory research campaign has been launched to preliminarily investigate the thermo-mechanical performances of the Bottom Cap (BC) region of the WCLL COB segment because of its peculiarities making its design different from the other regions. The assessment has been carried out considering the nominal BB operating conditions, the Normal Operation (NO) scenario, as well as a steady-state scenario derived from the in-box LOCA accident, the Over-Pressurization (OP) scenario. Starting from the reference geometric layout of the WCLL COB BC region, a first set of analyses has been launched in order to evaluate its structural performances under a previously calculated thermal field and to select potential geometric improvements. Then, the analysis of a complete BC region was conducted from both the thermal and structural standpoints, evaluating its structural behaviour in compliance with the RCC-MRx code. Finally, after some iterations and geometric updates, a promising geometric layout of the BC region has been obtained even though some criticalities still persist in the internal Stiffening Plates and First Wall. However, the obtained results clearly showed that the proposed layout is worthy to be further assessed to achieve a robust enough configuration. The work has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted Ansys commercial FEM code.

Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities

The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.

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.

The European DEMO Helium Cooled Pebble Bed Breeding Blanket: Design Status at the Conclusion of the Pre-Concept Design Phase

Significant design efforts were undertaken during the Pre-Concept Design (PCD) phase of the European DEMO program to optimize the helium cooled pebble bed (HCPB) breeding blanket. A gate review was conducted for the entire European DEMO program at the conclusion of the PCD phase. This article presents a summary of the design evolution and the rationale behind the HCPB breeding blanket concept for the European DEMO. The main performance metrics, including nuclear, thermal hydraulics, thermal mechanical, and tritium permeation behaviors, are reported. These figures demonstrate that the HCPB breeding blanket is a highly effective tritium-breeding and robust driver blanket concept for the European DEMO. In addition, three alternative concepts of interest were explored. Furthermore, this article outlines the upcoming design and R&D activities for the HCPB breeding blanket during the Concept Design phase (2021–2027).

Deuterium permeation behavior of zirconium oxide coating after exposure to solid breeder pebbles

Tritium permeation barrier has been developed using ceramic coatings to mitigate tritium leakage through structural materials in most of the fusion reactor blanket concepts. In a solid breeder blanket concept, corrosion of the coating is a serious issue in addition to tritium permeation; however, the effects of corrosion on tritium permeation have been rarely researched. Here, we report deuterium permeation behaviors of zirconium oxide coatings prepared by metal organic decomposition after exposure to two kinds of solid breeder pebbles with different compositions, followed by deuterium permeation measurements. The deuterium permeation flux varied with the pebble composition: the coated sample after exposure to the pebbles of 30 mol% lithium metatitanate and 70 mol% lithium orthosilicate was smaller than that after exposure to the pebbles of 20 mol% lithium metatitanate and 80 mol% lithium orthosilicate. Because the thickness of the corrosion layer formed by the pebbles with 30 mol% lithium metatitanate and 70 mol% lithium orthosilicate was larger, the composition of the pebbles affected the thickness of the corrosion layer and the permeation behaviors rather than coating integrity.

European DEMO Fusion Reactor: Design and Integration of the Breeding Blanket Feeding Pipes

This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the same time, the size of the upper port is constrained by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. The BB feeding pipes inside the vertical port need to be removed prior to BB maintenance, as they obstruct the removal kinematics. Since they are connected to the BB segments on the top and far from their vertical support on the bottom, the pipes need to be sufficiently flexible to allow for the thermal expansion of the BB segments and the pipes themselves. The optimization and verification of these BB pipes inside the upper port design are critical aspects in the development of DEMO. This article presents the chosen pipe configuration for both BB concepts considered for DEMO (helium- and water-cooled) and their structural verification for some of the most relevant thermal conditions. A 3D model of the pipes forest, both for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) concepts, has been developed and integrated inside the DEMO Upper Port (UP), Upper Port Ring Channel, and Upper Port Annex (UPA). A preliminary structural analysis of the pipeline was carried out to check the structural integrity of the pipes, their flexibility against the thermal load, their internal pressure, and the deflection induced by the thermal expansion of the BB segments. The results showed that the secondary stress on the hot leg of the HCPB pipeline was above the limit, suggesting future improvements in its shape to increase the flexibility. Moreover, the WCLL concept did not have a critical point in terms of the secondary stress on the pipeline, since the thicknesses and the diameters of these pipes were smaller than the HCPB ones.

Tritium transport in the vacuum vessel pressure suppression system for helium cooled pebble bed

In the frame of the safety studies for the EU-DEMO reactor, attention is paid to the hydrogen concentration in the vacuum vessel and connected volumes since it would lead to a possible hazard of releasing tritium and activated dust. The risk of explosion cannot be excluded a priori if H2 stockpiles. For this reason, in both water (WCLL) and helium (HCPB) cooled breeding blanket concepts of EU-DEMO, the problem is under investigation with a cross-reference between the available technologies in fission (such as the Passive Autocatalytic Recombiners – PAR) and fusion application. In particular, the recent analyses pointed out the implementation of the PARs into the Vacuum Vessel Pressure Suppression System or linked systems.This paper evaluates the Hydrogen behavior (main mobilized tritium source term) for the Helium-Cooled Pebble Bed (HCPB) VVPSS concept. The analyses preliminary investigate the stratification of the hydrogen mass inventory inside the PSS. In particular, a MELCOR 1.8.6 model of the PSS, based on past activities aimed at dust transport and thermohydraulic analyses, is adopted. The paper also introduces the applicability of PAR technology in the operation range of fusion devices, analyzing the problem of the recombination rate due to the dilution of Hydrogen after a Helium blowdown.

3D MHD analysis of prototypical manifold for liquid metal blankets

The water-cooled lead lithium (WCLL) and the dual-cooled lead lithium (DCLL) are two of the breeding blanket (BB) concepts that the EUROfusion consortium is pursuing in the framework of the development of the fusion reactor industrial demonstrator DEMO. Both involve the use of a liquid metal (LM) as working fluid, the lead-lithium eutectic alloy (PbLi), due to its excellent thermal properties and the possibility to serve as both the blanket coolant and tritium breeder and carrier. Unfortunately, due to the high electrical conductivity of LMs, their motion is influenced by the magnetic field used in the reactor to confine the plasma, generating a complex phenomenology which is studied by magnetohydrodynamics (MHD). In this work, a representative prototypical manifold of a BB bottom feeder is investigated for different configurations with the custom phiFoam solver, capable of simulating unsteady, incompressible and isothermal MHD flow. The aim of this study is to investigate which configuration minimizes the flow imbalance in the manifold for the WCLL or in the poloidal breeding zone channels for the DCLL. The distribution of the flow rate among the channels is strongly influenced by the position of the feeding pipe (FP) and by the development of the MHD internal layer near the expansion, which generates important jets close to the lower plate and the upper one, where the channels are attached. The channel aligned with the FP is the one carrying most of the flow, from 55% to 82%, while in the more distant one the flow is almost stagnant, carrying from 17% to 6% of the total flow rate. The total pressure loss is also estimated and its functional dependence on the manifold configuration is discussed.

Effects of atmospheric gases and compression load on effective thermal conductivity of Li2TiO3 pebble bed

Tritium breeder material in the solid-type of breeding blanket concepts is usually used in a pebble bed form because of the efficiency and convenience of tritium extraction. The pebble bed is continuously heated by the exothermal reaction between neutron and lithium-6 during the operation of a fusion reactor. Thus, a compression pressure is applied on the pebble bed by the thermal expansion of pebbles and container structure. Therefore, the accurate thermal properties of the pebble bed are one of the important design parameters to ensure a function and stability of the breeding blanket. This study aims at the investigation of the change on effective thermal conductivity of Li2TiO3 pebble beds under the operation conditions of breeding blanket.The measurement of effective thermal conductivities of Li2TiO3 pebble beds has been performed by laser flash technique in several atmosphere gases, such as helium, nitrogen, and air, under the statically compression load controlled by specially design pneumatic device incorporated to the laser flash apparatus. Also, in the case of helium atmosphere, the effects of hydrogen concentration from 0.1 wt.% to 1.0 wt.% on the effective thermal conductivity were also investigated. The effective thermal conductivity of the Li2TiO3 pebble bed was strongly dependent on a kind of atmosphere gases, especially its thermal conductivity, in the relatively low temperature. However, the concentration of hydrogen in the helium atmosphere did not significantly affect the effective thermal conductivity of the pebble bed. Applied axial compression load has influenced on the effective thermal conductivity of the Li2TiO3 pebble bed, especially at high temperature. However, the obvious change of effective thermal conductivity with the value of compression load was not observed.