Completion of IFMIF/EVEDA lithium test loop construction

The EVEDA Li test loop (ELTL) successfully completed its construction and installation of a total of 2.5-ton Li in the frame work of the IFMIF/EVEDA as one of the ITER-BA. Design for the ELTL had been done from March 2009 to December 2009 in large part, and then the construction was started on November 2009 in the O-arai site of the Japan Atomic Energy Agency and completed on the middle of November 2010 after passing an authority inspection by a fire department in O-arai town. Subsequently, the 2.5-ton Li was installed to the ELTL by using a glove box in the form of ingots which is 240 mm long and 125 mm in diameter. The nitrogen concentration in the 2.5-ton Li was found to be 127 wppm. During the installation, the oxygen concentration and the humidity in the glove box were almost kept less than 20 wppm, and any large contamination by air was prevented during the handling of Li.     

Improvement of IFMIF/EVEDA bayonet concept back-plate design

In the frame of the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International Fusion Materials Irradiation Facility (IFMIF) project, a supporting lithium loop has been designed and is currently under construction at Oarai (Japan) with the main objective to test several technological solutions to be adopted in the future IFMIF plant. Among these, the lithium target system represents one of the most critical components as it will be exposed to high-energy intense neutron flux and consequently to severe irradiation damage rates (up to 60 dpa/fpy). For this reason, it must be designed for periodic replacement. The solution proposed by ENEA is based on the so-called back-plate bayonet concept which consists of a replaceable element that can be inserted to and removed from the permanent structure of the target assembly by means of a sliding-skate mechanism. Recently, the design of the bayonet back-plate has been revised and some important modifications have been introduced in order to improve its functionality and optimize its features in terms of compactness, robustness and remote maintainability. Several design solutions have been conceived to achieve better performance including smaller overall dimensions, sealing load reduction, gasket retention system improvement, positioning and centering effectiveness and optimized detachment mechanism. Moreover, a new variable-curvature geometry for the lithium channel profile has been calculated using an analytic approach based on the simplified Navier–Stokes equations in order to avoid the fluid dynamic instabilities evidenced in the old profile. In this paper, the new design features of the back-plate are presented, along with the main outcomes obtained from the engineering assessment performed so far.     

Radioactive waste assessment for the IFMIF/EVEDA facility

During the IFMIF/EVEDA phase, a 125 mA and 9 MeV deuterons prototype accelerator will be designed and tested for the final IFMIF project. During operation of the accelerator deuteron losses will occur in several components leading to material activation induced by deuteron and/or by secondary neutrons, depending on its location. This work is focused on a first radioactive waste assessment at the end of the operational life of this facility. The radioactive wastes generation will be evaluated, focusing on the beam dump and main accelerator components. Following the current approach to the back-end of the activated materials, they will be categorized according to radiological complexity of operations and final management routes. For the calculations, MCUNED and ACAB codes were used together with TENDL-2010 and EAF-2007 data libraries, respectively.     

Safety managements of the linear IFMIF/EVEDA prototype accelerator

On the Linear IFMIF/EVEDA Prototype Accelerator (LIPAc), the validation up to 9 MeV deuteron beam with 125 mA continuous wave is planned in Rokkasho, Aomori, Japan. Since the deuteron beam power exceeds 1 MW, safety issue related to γ-ray and neutron production is critical. To establish the safety management indispensable to reduce radiation exposure for personnel and activation of accelerator equipment, Personnel Protection System (PPS) of LIPAc control system, which works together with Radiation Monitoring System and Access Control System, was developed for LIPAc. The management of access to the accelerator vault by PPS and the beam duty management of PPS are presented in details.     

Current status of the technology development on lithium safety handling under IFMIF/EVEDA

Studies for establishing technology for the safe handling of lithium was performed in the Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Material Irradiation Facility (IFMIF). This research comprises four tasks: (a) extinguishing lithium files, (b) chemical reactions of lithium on the event of a leak, (c) lithium removal from the components, and (d) chemical analysis of impurities in lithium. Tasks (a) and (b), related to functions on the event of a lithium leak, involve selection of the material suitable for extinguishing lithium fires and assessment of corrosive effects of leaked lithium on materials at high temperature, respectively. Task (c) involves evaluation of methods for the replacement and/or decommissioning of the lithium components. Task (d) constitutes the development of high-precision techniques for the determination of impurities in lithium, particularly the dominating corrosive impurity—dissolved nitrogen. Experimental results addressing the objectives of each of the tasks are described in this communication.     

Basic design guideline for the preliminary engineering design of PIE facilities in IFMIF/EVEDA

The international fusion materials irradiation facility (IFMIF) is an accelerator-based intense 14 MeV neutron source for testing fusion reactor materials. Under broader approach (BA) agreement between EURATOM and Japan, the engineering validation and engineering design activity (EVEDA) were started from 2007. The IFMIF needs the post irradiation examination (PIE) facilities to generate a materials irradiation database for the design and licensing of fusion DEMO reactors. In this study we examined and discussed about the safety such as remote handling, hot cell design, and the equipments and apparatus of hot cells, and we summarized a basic design guideline for the preliminary engineering design of the PIE facilities.     

Evaluation of gamma-ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

The engineering validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of gamma-rays and ³He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated. It is found that the neutron and photon flux are decreased by 10⁴-order by employing the local shields using concrete and polyethylene around beam dump, and the photon energy can be suppressed in the low energy.     

Development of the IFMIF Tritium Release Test Module in the EVEDA phase

This paper presents the engineering design of the IFMIF (International Fusion Materials Irradiation Facility) Tritium Release Test Module (TRTM). The objectives of the TRTM are: (i) in-situ measurements of the tritium released from lithium ceramics and beryllium pebble beds during irradiation, (ii) studying the chemical compatibility between lithium ceramics and structural materials under irradiation, and (iii) performing post irradiation examinations for the irradiated materials. The TRTM has eight rigs which are arranged in two rows (2 × 4) perpendicular to the beam axis and enclosed by a structural container. Each rig includes one capsule that contains lithium ceramic or beryllium pebbles for irradiation. Neutrons reflectors are implemented at different locations to reflect the scattered neutrons back to the active region aiming to improve the tritium production. The TRTM is required to provide irradiation temperature range of 400–900 °C for the capsules filled with lithium ceramics and 300–700 °C for the ones packed with beryllium. The engineering design of the TRTM components such as container, rigs, capsules, pebble beds, neutrons reflectors, and purge gas and coolant tubes are presented. In addition a test matrix for the irradiation campaign is proposed.     

Technical analysis of an early fusion neutron source based on the enhancement of the IFMIF/EVEDA accelerator prototype

In the framework of the Engineering Design and Engineering Validation Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA), three major prototypes have been designed and are being manufactured, commissioned and operated which are firstly the Accelerator Prototype (LIPAc) at Rokkasho, fully representative of the IFMIF low energy (9 MeV) accelerator stage, secondly the EVEDA Lithium Test Loop (ELTL) at Oarai, and thirdly critical components of the High Flux Test Modules to be tested in the helium cooling loop (HELOKA-LP) at Karlsruhe. The present paper analyses possibilities from a technical point of view, for combining, modifying, and enhancing, at limited cost, selected components of the prototypes towards the realisation of an early reduced-flux neutron source, able nonetheless to start the testing of candidate DEMO materials and realising by this a first step towards the construction and operation of a complete IFMIF plant.Various options of deuteron beam parameters, such as energy, current and shape are analysed with respect to their technical challenges and the neutron yield resulting from the nuclear reaction with the Li target. Related requirements for the liquid Li target with respect to jet parameters are evaluated and the neutron mapping in the high flux region is presented underlying an analysis of the available volume for testing reduced activation ferritic martensitic (RAFM) steels at relevant structural damage levels.     

SiC/SiC composite heater for IFMIF

As the part of Japan/EU Broader Approach (BA) program for fusion, International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project is on-going. Where, High Flux Test Module (HFTM) design aiming up to 1000 °C has been intensively investigated. Innovative SiC/SiC heater is currently potential and promising option. The weak temperature dependence of the SiC/SiC up to 1000 °C was interpreted to suggest the important role of grain boundaries and fiber–matrix interphase. To control the heater properties, variety of fabrication conditions, mainly on SiC nano-powders and green-sheet, were selected and tested. The heater performance was reasonably controlled and the results of heater performance and underlying microstructure behavior are provided.     

Computational modelling of the IFMIF lithium target

The present work is devoted to the computational modelling of the process of beam action on a lithium target. The aim of the investigation is to determine the maximum values of temperature and pressure as well as general pattern of the process. The analysis is based on the compressible Euler equations with the stiffened gas equation of state with parameters corresponding to lithium. The energy influx allocation caused by the beam interaction with the target is described by the source term in the energy balance law. The formulated problem is solved numerically by a high-resolution Godunov-type method. The obtained results show a moderate rise in the lithium temperature and relatively large pressure variations.     

IFMIF specifications from the users point of view

This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the users requirements and top level specifications for the facility. Special attention is given to the different roadmaps of fusion pathway towards power plants, of materials R&D and of facilities and their interactions. The materials development and validation activities on structural materials, blanket functional materials and non-metallic materials are analyzed and specific objectives and requirements to be implemented in IFMIF are proposed. Emphasis is made in additional potential validation activities that can be developed in IFMIF for ITER TBM qualification as well as for DEMO-oriented mock-up testing.     

Conceptual design of the IFMIF Start-Up monitoring module

The preliminary engineering design of the test facilities, including the various test modules to be used in the IFMIF plant is a part of the IFMIF/EVEDA (Engineering Validation and Engineering Design Activities) project from the Broader Approach to fusion.One presents the current status of the conceptual development of the IFMIF Start-Up Monitoring Module, a dedicated device used in the IFMIF test cell during the commissioning phase of the installation, in order to completely characterise the irradiation conditions behind the target on which the beam of deuterons will be focused. This STUMM embarks a lot of instrumentation to precisely characterise the neutron field, the nuclear heating and the temperatures in the test cell.One briefly describes the measuring instruments (including a specific radiation flux monitor under development), the possible layouts and the possible positioning. One also defines which types of measurements are expected by this especially dedicated commissioning module.     

Hydro-thermo analysis of Li flow over IFMIF geometry

A thermo-hydraulic analysis of high-speed free surface Li flow over a concave plate (IFMIF geometry) is performed. Simulations are done for bulk velocities between 10 and 20 m/s using ANSYS Fluent. A pre-computed heat source was imposed at the center of the curved section to simulate the interaction of a dual deuteron beam with the Li jet. LES and k-ϵ models were used for turbulence modeling and Volume of Fluid and Level Set methods were used to model the free surface flow. Results reported are the variation of temperature, pressure and velocities across the Li jet at various locations along the curved region. Safety margins before Li starts boiling are also predicted. All cases predicted smooth surfaces without any waves.     

Nuclear analysis of the IFMIF European lithium target assembly system

In the framework of the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International Fusion Materials Irradiation Facility (IFMIF) project, ENEA was in charge of the design of the European version of the target assembly (TA) system which employs a removable bayonet backplate (BP) concept. With the aim of assessing the nuclear behaviour of the system and supplying the necessary input data to the thermomechanical analysis, coupled neutron-gamma transport calculations have been carried out for the whole TA + BP system, using the MCNP5 1.6 Monte Carlo transport code integrated with the McDeLicious-11 neutron source code provided by KIT. Neutron activation calculations have been performed by means of the EASY-2010 activation system in order to provide radioactive inventories useful for thermomechanical analysis and safety purposes. This paper summarizes the results obtained by the neutronic and activation calculations for the most irradiated components of the TA, such as backplate, frame, nozzle and target chamber.     

Thickness distribution of high-speed free-surface lithium flow simulating IFMIF target

International Fusion Materials Irradiation Facility (IFMIF) employs liquid lithium as D⁺ beams target. The liquid Li target is formed as flat plane free-surface flow by a nozzle and flows at high speed around 15 m/s. This paper focuses on flatness of the liquid Li target. The Li flow experiment was conducted in Osaka University Li Loop with a test section which was 1/2.5 scaled model of IFMIF. The thickness of the Li flow was measured and obtained by a contact method which was developed for the measurement. Analytical study on Kelvin wake and numerical calculation on the wake near the side wall of the flow channel were also conducted and compared with the experimental results. As a result, the positions of the wake crest obtained from both of the experiment and numerical calculation assuming contact angle 140° agreed well with the iso-phase line of the analytical model. The generation of the wake likely depends on wettability between Li and the structural material which is 304SS in the present study.     

Present status of the Liquid Breeder Validation Module for IFMIF

One of the objectives of IFMIF (International Fusion Materials Irradiation Facility), as stated in its specifications, is the validation of breeder blanket concepts for DEMO design. The so-called Liquid Breeder Validation Module (LBVM) will be used in IFMIF to perform experiments under irradiation on functional materials related to liquid breeder concepts for future fusion reactors. This module, not considered in previous IFMIF design phases, is currently under design by CIEMAT in the framework of the IFMIF/EVEDA project.In this paper, the present status of the design of the LBVM is presented.     

Development and validation status of the IFMIF High Flux Test Module

The development of the IFMIF (International Fusion Material Irradiation Facility) High Flux Test Module in the EVEDA (Engineering Validation and Engineering Design Activities) phase up to 2013 includes conceptual design, engineering analyses, as well as design and engineering validation by building of prototypes and their testing. The High Flux Test Module is the device to facilitate the irradiation of SSTT samples of RAFM steels at temperatures 250–550 °C and up to an accumulated irradiation damage of 150 dpa. The requirements, the current design and the performance of the module are discussed, and the development process is outlined.     

Tritium removal by Y hot trap for purification of IFMIF Li target

The use of Li is expected for a flowing target in the International Fusion Materials Irradiation Facility (IFMIF) for high-intensity neutron generation. Tritium (T) of by-product is removed by a Y hot trap placed in a by-path position of the flowing Li target. Since the equilibrium pressure of hydrogen isotopes is extremely low, the removal rate of H isotopes is affected by coexisting other impurities such as N, C and O dissolved in Li. In the present study, the removal rate of H isotopes in Li is experimentally investigated under the three different methods or conditions: (i) measuring gravimetrically in a static Li + Y system with the supply of low-concentration H₂ in Ar purge, (ii) counting the change of T radioactivity in a static Li + Y system after neutron irradiation and (iii) measuring the change of the H₂ concentration in a stirred Li + Y system with the supply of low-concentration H₂ in Ar purge. There was no corrosive action in the Li-Y interface observed after contact at 500 °C for 100 h. The present results are extended to design a Y hot trap for the IFMIF-EVEDA Li loop in order to remove H isotopes less than a specified concentration.