Progress of fusion nuclear technologies in the broader approach framework

In the Broader Approach framework, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, the International Fusion Energy Research Center (IFERC) project, and the Satellite Tokamak project are implemented. In the IFMIF/EVEDA project, engineering design of IFMIF and engineering R&D include the construction and tests of an IFMIF prototype accelerator system up with a 9 MeV and CW deuteron beam, a liquid lithium test loop with free surface flow, and full scale irradiation test module including temperature control instrumentation. The commissioning of the EVEDA lithium test loop was completed in March 2011, and a lithium flow of ～5 m/s was obtained. As a part of the IFERC project, R&Ds on reduced activation ferritic/martensitic steels as blanket structural material, SiC_f/SiC composites as a flow channel insert material and/or alternative structural material, advanced tritium breeders and neutron multipliers, and tritium technology are carried out. At the beginning of 2011, the integrated DEMO design team was established among the IFERC project team and EU/JA home teams, where the design criteria, other design basis are discussed as an initial work. A high performance supercomputer with the peak performance of 1.3 Pflops is under installation at the Rokkasho BA site.     

Approach to the lifetime assessment of the bayonet back plate for IFMIF target

The IFMIF facility is aimed at the production of high flux (10¹⁸ n/m²/s) of 14 MeV neutrons to test the candidate Fusion materials under significant neutron damage, up to 50 dpa/year. The conceptual configuration of the IFMIF target, based on the bayonet back plate (BP), has been developed in the past years by several authors. The appropriate engineering design of the back plate, to be developed in the EVEDA (Engineering Validation and Engineering Design Activities) phase, would require a very high level of knowledge on the materials behaviour under irradiation, that will be acquired only after some years of IFMIF experimental activities. For this reason the back plate, which is primarily invested by the highest IFMIF neutron flux, has to be considered a sacrificial component. In spite of its systematic replacement, the engineering design has to be optimised and the lifetime analysis has to be made carefully, in order to credibly estimate the expected replacement frequency. Since the replacement time interval must be conservatively shorter than the back plate lifetime and, at each replacement, the facility has to be stopped for, at least, one week and subjected to risky and uncomfortable operations, it is necessary to perform a trustworthy analysis of the lifetime. To this purpose the various interconnections between the main damaging causes are discussed in order to evidence the most plausible reasons of back plate malfunctioning. Due to the lack of knowledge in some fields and the early stage of design, the analysis is only semi-quantitative. The analysis, which accounts for erosion/corrosion, hydraulic stability, neutron damage and thermo-mechanical stress as the main damaging causes, evidences also the research areas which deserve foremost attention during the EVEDA phase. The considered malfunctions are: lithium boiling, burning/piercing of the back plate, non-sufficient neutron flux, brittle rupture of the back plate, creep rupture, loss of tightness of the back plate sealing.     

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.     

The European contribution to the development and validation activities for the design of IFMIF lithium facility

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250 mA, energy of 40 MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility.The paper gives an overview of the status of the activities and of the main outcomes achieved so far.     

EPICS based low-level radio frequency control system in LIPAc

The IFMIF–EVEDA (International Fusion Materials Irradiation Facility – Engineering Validation and Engineering Design Activity) linear accelerator, known as Linear IFMIF Prototype Accelerator (LIPAc), will be a 9 MeV, 125 mA continuous wave (CW) deuteron accelerator prototype to validate the technical options of the accelerator design for IFMIF. The primary mission of such facility is to test and verify materials performance when subjected to extensive neutron irradiation of the type encountered in a fusion reactor to prepare for the design, construction, licensing and safe operation of a fusion demonstration reactor (DEMO). The radio frequency (RF) power system of IFMIF–EVEDA consists of 18 RF chains working at 175 MHz with three amplification stages each. The low-level radio frequency (LLRF) controls the amplitude and phase of the signal to be synchronized with the beam and it also controls the resonance frequency of the cavities. The system is based on a commercial compact peripheral component interconnect (cPCI) field programmable gate array (FPGA) board, provided by Lyrtech and controlled by a Windows host PC. For this purpose, it is mandatory to communicate the cPCI FPGA board from EPICS Channel Access [1]. A software architecture on EPICS framework in order to control and monitor the LLRF system is presented.     

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.     

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.     

Progress and achievements of the BA activities in Rokkasho

In Rokkasho Japan, the International Fusion Energy Research Center (IFERC) project and the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project are on going under the Broader Approach framework. The IFERC project consists of three sub-projects; a fusion demonstration reactor (DEMO) Design and R&D Coordination Center, a Computational Simulation Center (CSC), and an ITER Remote Experimentation Center (REC). DEMO Design activity has been conducted by the IFERC project team in Rokkasho and home teams in EU and JA. In the DEMO R&D activity, five R&D tasks mainly of the blanket materials are carried out intensively. A supercomputer with 1.23 Pflops of LINPAC performance has been installed in December 2011, the operation started in January 2012. Discussion of overall plan of REC has started in 2012 between EU and Japan. In the IFMIF/EVEDA project, an IFMIF prototype accelerator system up to 9 MeV with 125 mA CW deuteron beam will be installed and tested in Rokkasho. Major components of the accelerator are under development or fabrication in EU. The first component of the accelerator, an injector with an ECR ion source, will be delivered to Rokkasho in March 2013.     

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.     

Characteristics of surface oscillation on high speed liquid Li jet

In the International Fusion Materials Irradiation Facility (IFMIF), high speed liquid lithium (Li) wall jet will be used as target irradiated by two deuteron beams of 125 mA at 40 MeV. To obtain knowledge of Li flow behavior, we have been studying on the surface wave characteristics experimentally using the liquid metal Li circulation loop at Osaka University. In this present study, the characteristic of surface oscillation on high speed liquid Li jet were examined. The free surface oscillation of Li flow was measured by an electro-contact probe apparatus, which detects electric contacts between a probe tip and Li surface. It was installed at 175 mm and 15 mm downstream from the nozzle exit to see influence of the initial growth of surface waves. The wave height of free surface waves was obtained from contact signal. While at 15 mm region the flow surface is very smooth covered with small waves in amplitude, the surface waves are developed sufficiently at the 175 mm. In the case of the velocity of 15 m/s, the maximum wave height reaches 4.8 mm. Heat deposition was estimated on the target back-plate with using the present statistical wave data.     

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.     

Start-up phase of the HELOKA-LP low pressure helium test facility for IFMIF irradiation modules

As part of the Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) [1], it is foreseen to design and test a 1:1 scale prototype of the IFMIF High Flux Test Module (HFTM) [2]. The module has been designed to be cooled by a low pressure helium gas flowing through minichannels to remove the nuclear heat.The Helium Loop Karlsruhe-Low Pressure (HELOKA-LP) has been designed to provide coolant at 1:1 HFTM operational conditions: massflow 12–120 g/s, inlet pressure 0.3–0.6 MPa, inlet temperature RT – 250 °C. A secondary objective is to use the experience gained with HELOKA-LP for the planning of the IFMIF helium cooling system.The facility has been put into operation in 2009, and has since then been in a test and optimization phase. It was proven, that the above mentioned requirements for the facility are achieved. The paper describes the process layout and components of the facility. The performance is characterized by the results of several steady state and transient benchmark tests. Typical start-up and transition times relevant for the operation mode in the IFMIF irradiation campaigns are obtained. Additionally first results on the impurity ingress and the cooling gas chemistry are described.     

Design and analysis of the IFMIF–EVEDA beam dump cooling system

The IFMIF–EVEDA beam dump is designed to stop a 9 MeV, 125 mA continuous wave deuteron beam that deposits along its surface a total of 1.125 MW. The beam dump design is based on a 2.5 m long copper cone whose inner surface absorbs the beam. This piece is cooled by water flowing at high velocity through the annular channel formed between it and a second piece (shroud) made of four truncated cones of slightly different slopes.In this paper the beam dump cooling system will be briefly described, and the relevant 1D and 3D results will be presented paying especial attention to the computational fluid dynamics results.     

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.     

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.     

Design of a beam dump for the IFMIF-EVEDA accelerator

The IFMIF-EVEDA accelerator will be a 9 MeV, 125 mA cw deuteron accelerator prototype for verifying the validity of the accelerator design for IFMIF. A beam stop will be used for the RFQ and DTL commissioning as well as for the EVEDA accelerator tests. Therefore, this component must be designed to stop 5 MeV and 9 MeV deuteron beams with a maximum power of 1.13 MW.The first step of the design is the beam-facing material selection. The criteria used for this selection are low neutron production, low activation and good thermomechanical behavior. In this paper, the mechanical analysis and radioprotection calculations that have led to the choice of the main beam dump parameters will be described.The present design is based on a conical beam stop (2.5 m length, 30 cm diameter, and 3.5 mm thickness) made of copper plus a cylindrical 0.5 m long beam scraper. The cooling system is based on an axial high velocity flow of water. This design is compliant with the mechanical design rules during full power stationary operation of the accelerator. The radioprotection calculations performed demonstrate that, with an adequate local shielding, doses during beam on/off phases are below the limits.     

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.     

Current status of the engineering design of the test modules for the IFMIF

Under Broader Approach (BA) Agreement between EURATOM and Japan, IFMIF/EVEDA (International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities) has been performed since the middle of 2007. IFMIF presents three main facilities (the Accelerator Facility, Li Target Facility and Test Facilities). A previous design of IFMIF was summarized in the comprehensive design report [1]. The present EVEDA phase aims at producing a detailed, complete and fully integrated engineering design of IFMIF. The delivery of the “Intermediate IFMIF Engineering Design Report” is foreseen mid-2013. The goal of IFMIF is to obtain the indispensable design database to allow the design and licensing of DEMO and ensuring commercial reactors thanks to the materials data set obtained from planned evaluation tests such irradiations in high flux test modules (HFTM-vertical rig, HFTM-horizontal rig), medium flux test modules (creep fatigue test module, tritium release test module, liquid breeder validation module) and low flux test modules of IFMIF. In addition, the Startup Monitoring Module will be used for IFMIF commissioning. This paper is summarizing the overall current progress status of the engineering and conceptual design of the test modules in IFMIF/EVEDA.     

Erosion of lithium coatings on TZM molybdenum and graphite during high-flux plasma bombardment

The rate at which Li films will erode under plasma bombardment in the NSTX-U divertor is currently unknown. It is important to characterize this erosion rate so that the coatings can be replenished before they are completely depleted. An empirical formula for the Li erosion rate as a function of deuterium ion flux, incident ion energy, and Li temperature was developed based on existing theoretical and experimental work. These predictions were tested on the Magnum-PSI linear plasma device capable of ion fluxes >10²⁴ m⁻² s⁻¹, ion energies of 20 eV and Li temperatures >800 °C. Li-coated graphite and TZM molybdenum samples were exposed to a series of plasma pulses during which neutral Li radiation was measured with a fast camera. The total Li erosion rate was inferred from measurements of Li-I emission. The measured erosion rates are significantly lower than the predictions of the empirical formula. Strong evidence of fast Li diffusion into graphite substrates was also observed.