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A RAMI (Reliability, Availability, Maintainability and Inspectability) assessment performed on the ITER Test Blanket Module (TBM) ancillary systems is presented. The assessment is aimed at evaluating design that may jeopardize the achievement of the overall 75% availability requirements for the TBM plant breakdown structure. The ancillary systems of the European Test Blanket Modules to be operated in ITER that have been analyzed in this assessment are the helium cooling systems (HCSs) of both the Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB), the lithium-lead loop of the HCLL TBM and the Tritium Extraction System (TES) of the HCPB TBM. RAMI performance was assessed by means of reliability block diagrams (RBDs) over a foreseen mission of 20 years, with operating cycles reflecting ITER schedule. In particular, a set of events leading to unavailability of the systems was initially defined by means of a failure mode and effect analysis. Then RBDs were implemented according to reliability-wise integration of the components included in such systems. Different RBDs were defined depending on the component being judged to impact normal operation or start-up operation mode. Systems analysis was performed exploiting a modular approach in order to elicit the relative contribution to overall availability so to support possible design improvement by highlighting critical sub-systems. In particular the unavailability of cooling, tritium extraction and trapping functions were assessed separately for HCLL and HCPB concepts. Finally the integrated impact of all ancillary systems on overall tokamak operational availability was estimated and the assessed systems, considering the current preliminary level of design development, appear able to meet the expected performance.

The validation of the key technologies relevant for a DEMO Breeding Blanket is one of the main objectives of the design and operation of the Test Blanket Systems (TBS) in ITER. In compliance with the main features and technical requirements of the parent breeding blanket concepts, the European TBM Project is developing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS, focusing in this phase on the design life cycle and on R&D activities in support of the design. The TBS ancillary systems are mainly circuits devoted to the removal of thermal power and to the extraction and recovery of the tritium generated in the Test Blanket Modules. They are: • The Helium Cooling System (HCS); • The Coolant Purification System (CPS); • The Tritium Extraction System (HCLL-TRS, HCPB-TES); • The Lead Lithium Loop. Their conceptual design was deeply analyzed during the ITER Conceptual Design Review (CDR) in 2015. The assessment of the CDR was overall positive and recommendations for improvements were made. The present design takes into account the main CDR recommendations as well as the implementation of the requirements related to ITER operation, safety principles application and physical space constrains.

The experimental facility THALLIUM was designed and installed at ENEA C.R. Brasimone to investigate the consequence of a HCLL-TBS In box LOCA, ensuring the geometrical relevance for HCLL-TBM. Within the framework of the contractual activities with Fusion for Energy, a first experimental campaign was carried out in HCLL-TBS relevant conditions. The experiments simulated a rupture in a cooling plate of the HCLL-TBM. Thereafter, a new experimental campaign was developed by installing a larger injection valve, with the aim of analyzing a worse accidental He injection. The experimental parameters that were varied during the first campaign were frozen, varying instead injection time and pressure. The pressure trend in different points of the facility and the effectiveness of the mitigation strategy are the most important outcomes of the experiments. A different behavior with respect to the first campaign was observed. In particular, the first pressure peaks are smaller as a percentage of the injection pressure and the pressure is increasing during the whole transient. The relief valve on the expansion tank drives the transient, confirming the importance of the design of this component to mitigate the effects of an In-box LOCA. The outcomes of this experimental investigation are not only supporting the design of HCLL-TBS but also provide important data in view of the conceptual design of HCLL breeding blanket for DEMO.

To analyze the impact of the magnetohydrodynamics (MHD) effect on the fast draining of a LiPb channel (lithium-lead eutectic, 15.7 at.% Li) for a liquid metal fusion blanket such as the water-cooled lithium-lead test blanket system of ITER or DEMO, an experimental campaign was carried out with the support of the Integrated European Lead Lithium LOop experimental facility (IELLLO), installed at the ENEA Brasimone research center, Italy. The experiments were carried out by measuring the drainage time of the internal permanent magnet pump channel, normally used to circulate the LiPb in the loop, with and without the magnetic field. Moreover, this paper proposes a new numerical methodology to study the time delay induced by the MHD by using the commercial software COMSOL Multiphysics. In this way, it was possible to evaluate the LiPb fraction present at each time step in the computational domain and to estimate the time necessary for the complete drainage of the channel. The level set method was used to describe the transient behavior of the MHD flow under low-Rm approximation. The developed code was compared with the experimental results and showed good agreement, and it constitutes the first step in model validation as a possible application to ITER and DEMO. The experimental and numerical analyses performed in this work can be used as a benchmark case for MHD code development.

The tritium technologies and transport modelling play an essential role for the achievement of the TBM Programme scientific objectives. Tritium technologies deployed in TBS (Test Blanket System) can be divided in processing systems and measurement devices. In the current design of the European HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) TBS, tritium extraction from flowing Pb-16Li and tritium concentration in He are the most relevant tritium technologies under development. The optimization and integration of real time tritium concentration measurement devices in TBS is a complex task needed to keep under control the tritium concentration in the TBSs and to support their correct operation. For HCLL and HCPB-TBS application, the tritium concentration measurement devices have to operate in flowing Pb-16Li and in He in a wide range of operating conditions. In parallel with the activities on tritium technologies, an important effort has been spent in the last few years in developing tritium transport modelling capabilities, which is one of the most relevant objectives of the European TBM Programme. This paper first of all provides an overview on the most relevant and updated R&D and design results achieved in the above mentioned areas within the European TBM Project. Recent results on tritium inventory and permeation rate in HCLL-TBS, predicted by a 1-D dynamic system simulation code developed within the Ecosimpro platform, are presented and discussed in the part dealing with the TBS tritium transport modelling development. Considerations on DEMO relevancy are also given, particularly for the tritium processing systems as the change of scale from TBM to a breeding blanket for DEMO may impact their feasibility and performance.

Six Test Blanket Systems (TBS) will be installed and operated in ITER in the framework of the ITER Test Blanket Module (TBM) Program. After a short description of each TBS and of the main operating parameters, this paper gives the main features of the various TBSs sub-systems, in particular, the cooling systems (helium, water and/or lithium-lead), the tritium system for extracting the tritium from the breeder, the coolant purification system, the tritium accountancy system, and several instrumentation & control systems. This paper addresses also the main services that are needed to operate a TBS such as the secondary water cooling system, the chilled water system to cool the helium circulators, the liquid and gaseous nitrogen used in the tritium extraction process, the compressed air needed to operate the various valves present on the circuits, and the electrical power supply for all the active electrical components such as pumps, circulators and heaters. Other important aspects are the required interfaces with other ITER plant systems such as those involving air and process gas detritiation, power supply, radiological and environmental monitoring, remote handling, measurements, and services.

The depiction of the nuclear responses of the ITER European Test Blanket Modules (TBMs), Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebbles Bed (HCPB) is presented in this work. Following important components update, and important methodological advances, the nuclear heat and the tritium production have been revisited, giving new estimations 10% higher than the previous evaluation for nuclear heat in both TBMs and to 15% higher for HCPB T production. This has an impact on the thermo-mechanical design of the TBM and the tritium handling. In addition, the Shutdown Dose Rates in the respective port interspace have been characterized in local approach. It shows a performance that could imply compatibility with planned in-situ maintenance activities when analysed in global approach, an improvement with respect to previous evaluations.

Six Test Blanket Systems (TBS) will be installed and operated in ITER in the framework of the ITER Test Blanket Module (TBM) Program. After a short description of each TBS and of the main operating parameters, this paper gives the main features of the various TBSs sub-systems, in particular, the cooling systems (helium, water and/or lithium-lead), the tritium system for extracting the tritium from the breeder, the coolant purification system, the tritium accountancy system, and several instrumentation & control systems. This paper addresses also the main services that are needed to operate a TBS such as the secondary water cooling system, the chilled water system to cool the helium circulators, the liquid and gaseous nitrogen used in the tritium extraction process, the compressed air needed to operate the various valves present on the circuits, and the electrical power supply for all the active electrical components such as pumps, circulators and heaters. Other important aspects are the required interfaces with other ITER plant systems such as those involving air and process gas detritiation, power supply, radiological and environmental monitoring, remote handling, measurements, and services.

The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to be used as a key component in fusion power devices. It provides tritium breeding and nuclear power conversion and extraction. Before its integration in a DEMO plant, a testing phase is required to gain data on both thermal and neutronic performances. For this purpose, a Test Blanket System is currently under conceptual design and will be integrated into dedicated ITER equatorial ports. Early insight on the safety performance of the Test Blanket System remains an essential element for integrating into ITER, and accident analysis is one of its critical components. Preliminary safety studies have been performed in the framework of Safety And Environment (SAE) work package of the EUROfusion consortium program to support the design and the integration of the Water-Cooled Lithium Lead Test Blanket System into ITER. At the present conceptual design phase, these studies are tailored to give insight on the effects of some parameters, such as postulated initiating event timing, plasma termination system or valves intervention, and possible safety provision implementation. A MELCOR model of the Water-Cooled Lithium Lead Test Blanket System has been developed, and two accident scenarios have been studied: an ex-vessel LOCA inside the port cell and a loss of flow accident because of pump seizure. The impact of mentioned parameters on accident evolution and consequences is investigated in support of the safety logic definition.

ZrCo is a well-known tritium storage material and has been studied intensively in the literature. The most interesting properties with regards to the thermodynamics of the ZrCo-H system are the very low H2 partial pressure in equilibrium with ZrCoH3 at room temperature and the ease to reach sufficiently high temperature to completely release the stored H2. These properties motivate also to use ZrCo not as a simple storage, but rather as a concentrator of hydrogen isotopologues from inert gases like He. With such function, ZrCo getter beds are the reference solution adopted in the conceptual design of the tritium extraction system of the European Test Blanket Modules (TBM) to replace the cryogenic molecular sieve bed previously proposed. An experimental campaign was carried out on ZrCo in order to consolidate this choice. The results confirmed that ZrCo performs well as getter material but only substantially below the maximum loading capacity. They revealed that the dynamic thermo-mechanical respon...

This paper gives an overview of nuclear maintenance strategy to be followed for the European HCLL & HCPB Test Blanket Systems (TBS) to be installed in ITER. One of the several core documents to prepare in view of their licensing is their respective ‘Maintenance Plan’. This document is fundamental for ensuring sound performance and safety of the TBS during ITER’s operational phase and shall include, amongst others, relevant information on: maintenance organization, preventive and corrective maintenance task procedures, condition monitoring for key components, maintenance work planning, and a spare parts plan, just to mention some of the key topics. In compliance with the ITER Plant Maintenance policy, first steps have been taken aimed at defining nuclear maintenance strategy for some of the most relevant HCLL & HCPB TBS components, conducted by F4E in collaboration with industry. After a brief recall of maintenance strategy of the TBM Program (PBS-56), this paper analyses main features of EU HCLL & HCPB TBS maintainability and identifies, at their conceptual design phase, a preliminary list of maintenance tasks to be developed for their most representative components. In addition, the paper also presents the first nuclear maintenance studies conducted for replacement of the Q 2 Getter Beds, identifying some design recommendations for their sound maintainability.

One of the main open issues in the design of the HCLL Test Blanket System for nuclear fusion energy applications is the design and testing of the Tritium Extraction Unit (TEU). The lead-lithium alloy which, in the HCLL blanket project, is used both as tritium breeder and neutron multiplier, will generate tritium that must be extracted with an efficiency as high as possible in order to keep low the tritium concentration in the liquid metal, then minimizing tritium permeation into the main coolant, the consequent CPS load and, last but not least, tritium permeation into the Port Cell through the PbLi loop pipes. Gas-liquid contactors (GLC) and, more in detail, packed columns are the reference technology for TEU of European ITER HCLL-TBS. An experimental campaign on a GCL packed column was performed in TRIEX (TRItium Extraction) facility in ENEA Brasimone Research Centre. Starting from the experimental results a parametric analysis was performed in order determine the basic parameters of the gas extraction process and consequently the efficiency of this technology in different operative conditions. Experiments were conducted at a constant temperature of 450 °C varying the mass flow rate in the range 0,1–0,35 kg/s for lead lithium and 5–150 Nl/h for Ar stripping gas. The hydrogen partial pressure in lead lithium at extractor inlet was investigated in the range between 8.000–14.300 Pa. The experimental results have shown a tritium extraction efficiency in the range between 10 and 30% in good agreement with past experimental results obtained in MELODIE loop in CEA using similar packing material.

From 2014 to 2020, the Pre-Conceptual Design phase (PCD) of the EU DEMO has taken place. The activities in the PCD phase differ from past exercises in their strong Systems Engineering methodology, as well as for the pragmatic approach in their technology choices. The Helium Cooled Pebble Bed (HCPB) is one of the 2 candidates as driver blanket for the EU DEMO in the PCD phase. Several design iterations have been required during the PCD phase in order to adjust the design to the current demanding DEMO requirements, to the very challenging systems integration and to the need to keep near-term technologies. To this respect, the design has evolved to a so-called fuel-breeder pin architecture built in single-module segments. The pins are filled with a pebble bed of a ceramic breeder mixture of Li4SiO4 + 35 mol % Li2TiO3 (60% 6Li) and are embedded in prismatic blocks of Be12Ti acting as neutron multiplier. He gas at 8 MPa is used as coolant with a temperature window of 300–520 °C. This architecture has proven to achieve a large tritium breeding performance (≈1.20), a remarkably low plant circulating power (

The strategy for the development of the European DEMO Breeding Blanket (BB) and the TBM (Test Blanket Module) Programme have been recently realigned with the objective to mitigate the risks associated with the development of tritium breeding and power extraction technologies in a DEMO Fusion Plant. In such a context, a clear definition of the key technical requirements of the DEMO Breeding Blanket (BB) is essential not only to drive a consistent BB design but also to allow the TBM Programme to provide its contribution to the preliminary demonstration of compliance with those requirements. Such contribution is substantiated through all phases of the TBM Programme life cycle, from the TBS (Test Blanket System) conceptual design up to its installation in ITER, operation and final dismantling. After a recall of the status of development of the European BB DEMO and TBM Programme, the paper describes the strategy and the status of the activities for the preliminary demonstration of compliance with a selected group of DEMO BB requirements. Through the elements given in this paper it is highlighted the importance to have a continuous and strict connection between the DEMO BB development and the execution of the TBM Programme.

This paper provides an overview of the newly revised design and development strategy for the DEMO breeding blanket in Europe. This has been defined to take into account the input from the DEMO pre-conceptual design activities, the findings and recommendations of a thorough technical and programmatic assessment of the breeding blanket and the EU ITER Test Blanket Module (TBM) programs, conducted in 2017 by an independent expert panel. This work has led to the identification of (i) the most mature and technically sound breeding blanket concepts to be used as “driver” and “advanced” breeding blankets in DEMO, the latter to be installed and tested in a limited number of properly designed segments, potentially being more attractive for future fusion power plants; (ii) the remaining technical gaps and R&D priorities. A number of urgent steps that are required to better align and strengthen the EU TBM and DEMO Breeding Blanket Program as a whole and to aim at an efficient implementation of the work are described in this paper. These include a proposal to change the EU TBM options to be tested in ITER in order to obtain important and useful information from the two current breeders (solid and liquid) and coolants (helium and water) considered for DEMO.

This paper describes a project of a non-nuclear, 1:1 scale testing platform dedicated to tests, optimization and validation of integration and maintenance operations for the European TBM systems in the ITER Port Cell #16. This TBM platform is currently under construction in Centrum výzkumu Řež, Czech Republic. The facility is realized within the scope of the SUSEN project and its full operation is foreseen in the first half of 2016.

The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

Experimental facility THALLIUM was designed and installed at ENEA C.R. Brasimone to reproduce the geometry of the LLE (Lithium Lead Eutectic) loop of HCLL TBM. Within the framework of F4E-FPA-372, an experimental campaign was carried out in ITER relevant conditions. The experiments simulated a pipe rupture in a cooling plate of the HCLL TBM, injecting He from an injection valve connected to He-FUS3 loop. The aim was to study the release of high pressure He in LLE due to In- box LOCA. The fundamental phenomena to be observed are the pressure wave trend in the two pipes reproducing the pipe forest of the HCLL-TBM in ITER and the effectiveness of the mitigation strategy designed for HCLL-TBS. The main parameters that were varied during the experiments are: injection time, actuation of the isolation valves and opening pressure of the rupture disc. In the pipe forest mock-up, the pressure wave showed three distinct peaks, related to the water hammer, to the LLE oscillations in the expansion tank and to He which fills the loop. The isolation valves proved to have a deep influence on the pressure wave propagation also when they do not close, being able to damp the first pressure peaks.

Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives.

Licensing a pressurized nuclear equipment like the European Test Blanket Modules (TBM) Systems and, on the longer term, breeder blankets of a fusion demonstration reactor (DEMO), will require presenting to the Regulator and the Agreed Notified Body, along with design and safety analyses, supporting data like consolidated materials data and design limits, qualified fabrication procedures specifications and validated modeling tools that go often over today's state-of-the-art of nuclear industry. TBM systems feature indeed a newly developed structural material and advanced fabrication processes that were not referenced in any nuclear construction codes before, new type of functional materials, complex structures geometry and many interconnected sub-systems exchanging tritium by permeation or fluid mass transfer. For many years now, Europe has structured its development activities on TBM Systems toward the preparation of licensing. First tangible results are now arising: the EUROFER structural material has been introduced in the RCC-MRx nuclear code, supported by a database of several thousands of test records; TBM box fabrication procedure specifications are under standardization by industry in view of their qualification; a modeling tool for accurate simulation of tritium transport in TBM systems has been developed in view of refining conservative inventory data published in preliminary safety reports and optimizing waste management. Remaining challenges are identified and discussed.

Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown.

One of the main objectives of the experimental campaign on the Test Blanket Systems (TBS) in ITER is the demonstration of the efficient processing of the tritium generated in the Test Blanket Module (TBM). On the other side, efficient tritium processing in a TBS has deep implications on: (i) safe operation of TBS itself and whole ITER system; (ii) successful development and validation of tritium transport modelling codes; (iii) demonstration of DEMO relevancy of tritium processing technologies. This work describes various aspects of HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS activities related to TBS tritium management. After a short description of HCLL and HCPB blanket concepts and related TBS, the paper contains: 1. a presentation of the key tritium processing technologies in the current design baseline of the European TBS; 2. a discussion on the DEMO relevancy of some specific TBS tritium processing technologies; 3. an overview on the activities related to the tritium transport modelling tools that will be validated along the development of the TBM project, including experimental campaign in ITER, and used for supporting the DEMO Breeding Blanket design. These three items are connected each other since tritium-related data, generated through the experimental campaign in ITER and interpreted through suitable modelling tools, will be one of the most significant outcomes in support of the breeding blanket design for DEMO and beyond.

The Water Cooled Lithium Lead (WCLL) blanket is based on the eutectic liquid alloy Pb16Li as breeder material and neutron multiplier, and pressurised water as coolant. The liquid breeder flows at few mm/s in the blanket module while the pressurised water is circulated inside double-wall tubes. In spite of the adoption of double-wall tubes for the coolant, the probability of a water large leak because of a tube rupture accident cannot be considered negligible. As a consequence, the Pb16Li/water interaction due to a large break in one or more cooling tubes still remains one of the biggest concerns for this blanket concept. This paper reports the results of three experimental tests on Pb16Li/water interaction carried out at ENEA-Brasimone operating the LIFUS 5 facility. Water was injected into the reaction tank, containing Pb16Li at 330 C, at a pressure of 155 bar with different values of sub-cooling and with different free volumes in the reaction system. In addition, post test analyses with SIMMER III code are presented in order to compare the pressure evolution measured during the experiments with that calculated by the code. © 2013 Elsevier B.V.

The instrumentation of the HCLL and HCPB Test Blanket System is fundamental in ensuring that ITER safety and operational requirements are satisfied as well as in enabling the scientific mission of the TBM program. It carries out three essential functions: (i) safety, intended as compliance with ITER requirements toward public and workers protection; (ii) system control, intended as compliance with ITER operational requirements and investment protection; and (iii) scientific mission, intended as validating technology and predictive tools for blanket concepts relevant to fusion energy systems. This paper describes the strategy for instrumentation development by providing details of the following five steps to be implemented in procured activities in the short to mid-term (3–4 years): (i) provide mapping of sensors requirements based on critical review of preliminary design data; (ii) develop functional specifications for TBS sensors based on the analysis of operative conditions in the various ITER buildings in which they are located; (iii) assess availability of commercial sensors against developed specifications; (iv) develop prototypes when no available solution is identified; and (v) perform single effect tests for the most critical solicitations and post-test examination of commercial products and prototypes. Examples of technology assessment in two technical areas are included to reinforce and complement the strategy description.

The experimental facility IELLLO (Integrated European Lead Lithium LOop) was designed and installed at the ENEA Brasimone Research Centre to support the design of the HCLL TBM (Helium Cooled Lithium Lead Test Blanket Module). This work presents the results of the experimental campaign carried out within the framework of F4E-FPA-372 and which had three main objectives. First, to produce new experimental data for flowing LLE (Lead-Lithium Eutectic) for an analysis of the loop and the characterization of its main components. Then, to evaluate performances of commercial instrumentation as available instrumentation is not designed for use in LLE. Lastly, to use the data for validation of the model developed with the system code RELAP5-3D. The data collected could prove helpful to analyze the behavior of the LLE loop of ITER and DEMO in accidental conditions. The results show that the regenerative countercurrent heat exchanger has an efficiency ranging from 70 to 85%, mainly depending on the LLE mass flow rate. It was verified that the air cooler has the capability to keep the cold part of the loop at 623 K, even in the most demanding situation (700 rpm and maximum temperature of the hot part). The instrumentation tested showed good accuracy, with the exception of the turbine flow meter. Nevertheless, specific limitations in the upper operative temperatures were found for the LLE direct contact pressure transducer. RELAP5-3D simulations fit very well the associated experimental results achieved.

An extensive thermal-hydraulic experimental campaign was conducted on He-FUS3 helium loop facility to support the conceptual design of HCLL and HCPB Test Blanket System. The experiments were divided into three distinct phases. The first one was dedicated to the evaluation of the new ATEKO Turbo Circulator (TC) performances, identifying its operating limits in terms of supplied helium mass flow as a function of rotational speed, cold by-pass opening and loop pressure. The outcomes were compared with the manufacturer theoretical performance map and with a RELAP5-3D pre-test computation. In the second phase, experiments were carried out to analyze the facility dynamic response in hot conditions and to characterize its main components (TC, heaters, economizer, cooling system and valves). The wide amount of collected data will serve for the development and validation of a numerical model of the facility at TBS conditions. For the third phase, the tests were designed to investigate He-FUS3 behavior in accidental conditions representative of LOFAs and LOCAs scenarios.

The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

The design of the Test Blanket Modules for ITER and the breeding blanket for DEMO requires robust and accurate modeling tools. Transport phenomena through the blanket tritium cycle are complex and involve a large number of physical properties and parameters, many of which have not been determined yet with a level of accuracy adequate for design optimization. Similarly, the use of simplified models with experimentally determined lumped coefficients allows satisfactory predictions only in very limited range of operative conditions, strongly reducing their potential to be relevant to the DEMO design. Within the European Union fusion program a road map to develop such modeling tools has been defined with the purpose of supporting the design of the ITER Tritium Blanket System and to exploit the TBM experimental testing for extrapolation to DEMO. The roadmap includes the development of the simulation tools as well as the supporting validation and verification experiments that must be carried out in parallel. This paper gives an overview of the state of the art of tritium modeling tools for blanket design, proposes a structure of the tritium modeling tools in order to facilitate their development and identifies a realistic work plan to achieve their final delivery.

The objective of the ITER TBM Program is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. Such information is essential to design and predict the performance of DEMO and future fusion reactors. It foresees to test six mock-ups of breeding blankets, called Test Blanket Module (TBM), in three dedicated ITER equatorial ports from the beginning of the ITER operation. The TBM and its associated ancillary systems, including cooling system and tritium extraction system, forms the Test Blanket System (TBS) that will be fully integrated in the ITER machine and buildings. This paper describes the main features of the six TBSs that are presently planned for installation and operation in ITER, the main interfaces with other ITER systems and the main aspects of the TBM Program management.

The knowledge of the tritium transport parameters in lead lithium is fundamental for the design of the HCLL (helium cooled lead lithium) blanket. In fact, the inventory of tritium in fusion reactors blankets and the permeation of tritium into the blanket coolant, with the consequent leaks toward the environment, are strongly depending on its solubility and diffusivity in the lead alloy PbLi. Several experiments, devoted to investigate the function linking the tritium solubilised in lead lithium with the corresponding tritium partial pressure at equilibrium, were carried out in the past, but significant uncertainties still remain. A detailed analysis of the past experimental works is carried out in this paper with the aim to investigate the main problems occurred in the facilities used to measure the tritium solubility in PbLi that caused such a big spread in the achieved results. On the basis of this analysis, a new a multipurpose laboratory scale apparatus has been designed. The apparatus is able to measure the tritium solubility and diffusivity in PbLi in the range of temperature 300–550 °C and it will be operated with hydrogen partial pressure in the range 10 2 –10 4 Pa. The facility can work with desorption and absorption technique. Moreover, the apparatus has been designed to allow the testing of H/D concentration sensors in Pb–15.7Li in operative conditions relevant to the HCLL–TBM and the characterisation of hydrogen permeation barrier.

The coolant purification system (CPS) together with the tritium extraction system (TES) and helium cooling system (HCS) are the principal auxiliary circuits of helium-cooled-lithiium–lead (HCLL) and helium-cooled-pebble-bed (HCPB) test blanket modules (TBMs). To extract heat from TBMs, Helium is used as primary coolant. CPS is used to extract tritium from the helium primary circuit as well as to guarantee removal of impurities which could interact with structural material. The reference process proposed for CPS is composed of 3 main successive steps. Step 1 consists in oxidation of Q 2 and CO to Q 2 O and CO 2 using a copper oxide bed ( Q represents either: H, D or T). Step 2 is dedicated to the removal of water which is adsorbed together with CO 2 on molecular sieve bed. Step 3 will remove residual impurities using a heated getter. Based on the operating conditions of CPS (pressure, flowrate, temperature) and on an estimation of the impurities foreseen, this paper presents a design of the oxidising bed which fulfils all requirements in terms of efficiency and lifespan. The design is obtained using a phenomenological approach taking into account competition between oxidation of CO and Q 2 on the metal oxide. The model was implemented in matlab software. A column of 0.41 m large and 2 m long containing 480 kg of CuO is proposed to assure complete oxidation of Q 2 for 16 months long.

Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the feasibility of any PbLi based tritium breeding blanket (BB). Particularly in DEMO, high tritium extraction efficiency will be required in order to keep low the tritium concentration in the Pb-16Li loop. This is essential to minimize tritium release into the environment and tritium permeation from BB into the primary cooling system. In addition, the tritium extraction process needs to be highly reliable in order not to impact negatively on the operation of the whole fusion reactor, ITER or DEMO.In the present paper, a critical review of the main candidate technologies for tritium extraction from Pb-16Li, particularly gas liquid contactors and vacuum permeators, is accomplished. The intrinsic limits and possible advantages of these technologies are presented and discussed, in the light of considerations coming directly from mathematical models describing their behaviour as well as from the experimental results so fa...

The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium–Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, already available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated.

The HCPB (Helium Cooled Pebble Bed) and HCLL (Helium Cooled Lithium Lead) Test Blanket Modules (TBMs), developed in EU to be tested in ITER, adopt helium at 80 bar as primary coolant. This paper contains a conceptual design of the TBMs Coolant Purification System (CPS) based on the need to remove permeated tritium and gas impurities. The following steps have been considered: identification of CPS design requirements; review of the purification systems developed for Helium Cooled Fission Reactors and proposed for DEMO Fusion Reactor; selection of the most promising technologies for CPS; indications on instrumentation and procedures for tritium balance. The proposed solution is a three-stage process constituted by an oxidiser to convert Q 2 and CO to Q 2 O and CO 2 , an adsorption step, performed on molecular sieve at room temperature to remove Q 2 O and CO 2 , and a final step performed on a heated getter to remove residual impurities.

Europe has developed two reference tritium breeder blankets concepts that will be tested in ITER under the form of Test Blanket Modules: (i) the Helium-Cooled Lithium–Lead which uses the liquid Pb–15.7Li as both breeder and neutron multiplier, (ii) the Helium-Cooled Pebble-Bed with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. An extensive development program has been carried-out over the last decade combining experimental and numerical simulations aimed at identifying and quantifying physical phenomena occurring in breeder blankets and at optimizing their design and technologies accordingly. On this basis, sound guidelines for the design and technological choices of Test Blanket Modules can be derived. In addition, regulatory and ITER project requirements, which prefigure the future DEMO blanket ones, are now integrated in the development and qualification program of the Test Blanket Modules. Their scope and implication are analyzed.

Tritium as one of the two fuel components for fusion power plays a special role in any fusion device. Due to its volatile character, radioactivity and easy incorporation as HTO it needs to be contr...

In this paper, we present computations linking the tritium release rate to the characteristics of lithium–lead and helium cooling circuits. Impacting component performances are evaluated such as tritium permeation towards the He coolant in the blanket modules, lithium–lead circulation rate, tritium extraction unit efficiency, tritium permeation in steam generator, helium coolant leak rate, helium purification unit maximum flow rate and efficiency. Safety considerations are also taken into account. A finite element model (FEM) for tritium permeation was developed considering various phenomena such as tritium transport by convection and diffusion in lithium–lead, MHD effects on liquid metal flows, tritium permeation in structures with temperature gradients. Other sub-system performances, like He leak rate and efficiency of tritium extraction systems, are discussed via an engineering approach. The results show that a reasonable compromise among the various requirements can be found, leading to technologically achievable requirements for tritium permeation barriers, tritium extraction systems both from Pb–17Li and He, and leak rates from the He cooling system.

The reduction of tritium permeation from the Pb–17Li, or plasma, into the coolant is of crucial importance in order to reduce the radiological hazard in the steam generator vault as well as in the turbine/condenser area and to optimise the tritium balance in the reactor. The use of aluminium rich coatings has been selected as reference solution for the water cooled lithium lead (WCLL) blanket in order to produce reliable tritium permeation barriers (TPB). TPB qualification activities performed in the past allowed the selection of two reference deposition techniques, the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. On the basis of the results obtained in the past with the Corelli I–II devices, a new apparatus named Vivaldi was designed to perform comparative tests on two hollow cylindrical specimens in the same operating conditions. The performance of alumina coating on EUROFER 97 steel has been tested in gas and liquid metal phase. The obtained results in terms of permeated fluxes and permeation reduction factors (PRF) are herein presented and discussed. A post experiment examination of coatings was performed by use of optical and SEM microscopy.

The interaction between pressurised water and liquid Pb–17Li is a topical issue for the WCLL blanket concept and needs to be studied in detail because of the potentially significant effect on the reliability and safety of the blanket. Particularly, it seems important from the design point of view to deeply investigate the consequences of coolant micro-leaks in the blanket module generated by micro-cracks with a size in the order of 10−3 mm2. This kind of interaction has been studied during an extensive experimental campaign on the RELA loops in the ENEA Research Centre at Brasimone. The results showed a reduction in the liquid metal flow-rate in the circuit and a deterioration of the heat exchange properties between breeder and coolant due to the formation and growth of solid reaction products generated by the chemical reaction between lithium and water. The molar ratio between the recovered hydrogen recovered and water injected was in the range 0.3–0.5, confirming that, in the test conditions, the main solid reaction product is lithium hydroxide (LiOH). The possible impact of these results on the TBM-ITER design is also presented and discussed.

He-FUS3 is a helium facility designed and realized by ENEA in order to test the thermal-mechanical properties of prototypical breeding blanket module assemblies of a DEMO reactor. The actual facility has been upgraded with a high performance turbo circulator and a water heat exchanger integrating the pre-existent air cooler. In addition, a new test section located in the loop hot zone has been settled down with the objective of investigating safety relevant transient conditions of "In-TBM" LOCA scenarios. A RELAP5-3D© model has been developed to perform a set of preliminary simulations on the new He-FUS3 layout. Both cold and hot stationary conditions have been analyzed evaluating the turbo circulator performances for a wide range of helium flow rate. Outcomes have shown that RELAP5-3D© is an effective tool in reproducing the most significant phenomena of He-FUS3 system, providing relevant insight supporting future experimental campaigns. The post-test analysis phase will be, of course, fundamental for the qualification of a consistent numerical model. © 2015 Elsevier B.V. All rights reserved.

This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations. © 2015 Elsevier B.V.

Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant TBM and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. © 2015, American Nuclear Society. All rights reserved.

In view of the ITER conceptual design review, the design of the ancillary systems of the European test blanket systems presented in [1] has been updated and made consistent with the ITER requirements for the present design phase. Europe is developing two concepts of TBM, the helium cooled lithium lead (HCLL) and the helium cooled pebble bed (HCPB) one, having in common the cooling media, pressurized helium at 8 MPa [2]. TBS, namely helium cooling system (HCS), coolant purification system (CPS), lead lithium loop and tritium extraction/removal system (TES-TRS) have the purpose to cool down the TBM and to remove tritium to be driven to TEP from breeder and coolant. These systems are placed in port cell 16 (PC#16), chemical and volume control system (CVCS) area and tritium building. Starting from the pre-conceptual design developed in the past, more mature technical interfaces with the ITER facility have been consolidated and iterative design activities were performed to comply with design requirements/specifications requested by IO to conclude the conceptual design phase. In this paper the present status of design of the TBS is presented together with the preliminary integration in ITER areas. © 2015 Elsevier B.V.

The ITER project aims at building a fusion device with the general goal of demonstrating the scientific and technical feasibility of fusion power. The testing of Tritium Breeder Blanket concepts is one of the ITER missions and has been recognized as an essential milestone in the development of a future fusion reactor ensuring tritium self-sufficiency, extraction of high grade heat and electricity production. Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The strategy for the development of the instrumentation of the HCLL and HCPB Test Blanket Systems, which include the TBMs and their Ancillary Systems, is briefly recalled in this paper, along with the overview of the requirements coming from the harsh operational environment and the main challenges related to the integration with the complex design of the TBS components.

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of hydrogen and its isotopes in materials. We address the problem of tritium transport in Helium Cooled Lead-Lithium (HCLL) DEMO blanket from lead lithium breeder through different heat transfer surfaces to the environment by developing a computational code (FUS-TPC). The main features of the code are briefly described and a parametric study is performed in order to identify the most influencing parameters in terms of tritium releases into the environment and of tritium inventories. The results showed that the results are strongly affected by the Lithium and the efficiency of permeation barriers

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of atomic hydrogen in materials in which hydrogen and its isotopes are present. In this work the problem of tritium transport from lead–lithium breeder through different heat transfer surfaces to the environment has been studied and analyzed by means of a computational code. The code (FUS-TPC) is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. A simulation, using the code, was performed by adopting the configuration of the European configuration of the Helium Cooled Lead Lithium (HCLL) blanket for DEMO.

In the helium cooled lithium lead (HCLL) and water cooled lithium lead (WCLL) blanket concepts for DEMO correct and reliable management of tritium is of basic importance, both for safety and fuel cycle reasons. To develop a sensor for measurements of hydrogen (and its isotopes) concentration in liquid Pb-17Li, a permeable capsule of niobium was chosen. Different simulations with a mathematical model have been performed, and then the sensor was designed, constructed and tested. The first experimental results in gas phase showed a permeating flux much lower than the predicted one, probably due to the formation of an oxide layer on the capsule surface or to the formation of niobium hydrides. To solve this problem different solutions are presented.

Since several years ENEA is involved in experimental activities concerning the interaction between molten lithium lead alloy, in eutectic composition, and pressurised water in conditions relevant for DEMO fusion reactor, in order to predict the behaviour of a WCLL blanket module in case of an in blanket LOCA. In this ambit LIFUS 5 apparatus was designed and constructed to carry out the experimental campaign on water large leaks into liquid Pb–17Li, with the final aim to provide data for the validation of the mathematical modelling of the related phenomena. After two tests performed to qualify all mechanical components and the data acquisition system, LIFUS 5 was operated at the ENEA C.R. Brasimone for tests nos. 3 and 4. Water was injected into the reaction tank at a pressure of 155 bar with two different values of sub-cooling. The initial liquid metal temperature was fixed to 330 °C. The first pressure peak due to the water vaporization and jet expansion was clearly recognized together with the subsequent pressure increase due to further water injection and hydrogen generation. In the performed experiments the maximum pressure peak, as detected in both reaction and expansion vessels, never overcame the value of the injected water. Moreover, in these two tests a significant temperature increase in the reaction vessel occurred, strictly connected to the amount of injected water.

The determination of the radioactive inventory and of the contact dose rates in the different ITER Test Blanket Modules systems is carried out, both for Helium-Cooled Lithium-Lead (HCLL) concept and the Helium-Cooled Pebble-Bed (HCPB) concept. The evaluations have been carried out by means of the MICROSHIELD code, starting from the data on the neutron-induced radioactivity in the blanket materials, completely available for both the blanket modules. The possible sources of radioactive material in all the systems have been individuated and their contributes estimated. In general, for both HCLL and HCPB systems, radioactivity inventory and contact dose rates turn out to be quite moderate. No particular radioactive safety concern should arise for the examined components