Tomohiro Furukawa

IFMIF: overview of the validation activities

The Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF), an international collaboration under the Broader Approach Agreement between Japan Government and EURATOM, aims at allowing a rapid construction phase of IFMIF in due time with an understanding of the cost involved. The three main facilities of IFMIF (1) the Accelerator Facility, (2) the Target Facility and (3) the Test Facility are the subject of validation activities that include the construction of either full scale prototypes or smartly devised scaled down facilities that will allow a straightforward extrapolation to IFMIF needs. By July 2013, the engineering design activities of IFMIF matured with the delivery of an Intermediate IFMIF Engineering Design Report (IIEDR) supported by experimental results. The installation of a Linac of 1.125 MW (125 mA and 9 MeV) of deuterons started in March 2013 in Rokkasho (Japan). The worlds largest liquid Li test loop is running in Oarai (Japan) with an ambitious experimental programme for the years ahead. A full scale high flux test module that will house ~1000 small specimens developed jointly in Europe and Japan for the Fusion programme has been constructed by KIT (Karlsruhe) together with its He gas cooling loop. A full scale medium flux test module to carry out on-line creep measurement has been validated by CRPP (Villigen).

IFMIF/EVEDA lithium test loop: design and fabrication technology of target assembly as a key component

The engineering validation and engineering design activity (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeding as one of the ITER broader approach activities. In the concept of the IFMIF, two 40 MeV deuteron beams are injected into a liquid Li stream (Li target) flowing at a velocity of 15 m s−1. The EVEDA Li test loop (ELTL) is aimed at validating the hydraulic stability of the Li target at a velocity up to 20 m s−1 under a vacuum condition of 10−3 Pa as the most important issue. Construction of the ELTL, which is the largest liquid metal loop possessing 5.0 m3 Li for the fusion research ever, was completed in the O-arai Research & Development Center in the Japan Atomic Energy Agency on 22 November 2010. This paper presents the design and fabrication technology of a target assembly called integrated target assembly, in which the Li target is produced by a contraction nozzle along a concave channel. There are two concepts regarding the target assembly: the integrated target assembly and the bayonet target assembly. Both target assemblies are outlined in this paper, and then the newly proposed design of the integrated target assembly for the ELTL and its fabrication technology are given. The integrated target assembly was processed by a five-axis milling machine and the processing accuracy was measured by 3D measurement tools. Finally, methods applied for the validation of the stability of the Li target are introduced in this paper.

Evaluation Procedures for Irradiation Effects and Sodium Environmental Effects for the Structural Design of Japanese Fast Breeder Reactors

In the structural design of fast breeder reactors, irradiation effects and sodium environmental effects on structural materials have to be taken into account. In this paper, firstly, an evaluation procedure for irradiation effects on the mechanical properties of 316FR (FBR Grade 316 stainless steel), which is a newly developed stainless steel for the Japanese demonstration fast breeder reactor, is proposed. The procedure gives a limit of accumulated fast neutron fluence (E>0.1 MeV) as a function of temperature, so that the minimum tensile fracture elongation of 10 percent, which is the threshold for material to stay ductile, is maintained. Furthermore, the procedure determined a creep life reduction factor and a creep rate increase factor as a function of accumulated thermal neutron fluence (E<0.4 eV), within the limitation of the accumulated fast neutron fluence, to account for the creep life reduction and the increase of creep rate due to irradiation. Secondly, an evaluation procedure for sodium environmental effects on the integrity of 316FR and modified 9Cr-1Mo steel was proposed. It gave a corrosion allowance as a function of temperature, oxygen content, and service time, based on corrosion tests. It determined that no correction factors that correspond to sodium environment on design allowable stresses, etc., are needed, because no adverse effects of sodium on the mechanical properties of 316FR and modified 9Cr-1Mo steel were to be expected in the service conditions of FBRs. Both the procedures have been incorporated into the Japanese Elevated Temperature Structural Design Guide for Demonstration Fast Breeder Reactor.

Engineering Validation and Engineering Design of Lithium Target Facility in IFMIF/EVEDA Project

Abstract EVEDA Lithium Test Loop (ELTL) has been designed and constructed, has operated a liquid lithium flow test facility with the world’s highest flow rate and has succeeded in generating a 100-mm-wide and 25-mm-thick free-surface lithium flow along a concave back plate steadily at a high speed of 20 m/s at 300°C for the first time in the world. This result will greatly advance the development of an accelerator-based neutron source to high energy and high density, one of the key objectives of the fusion reactor materials development under the BA (Broader Approach) Activities. Recent related engineering validation and engineering design of the lithium facility has been evaluated.

Engineering Design of Contact-Type Liquid Level Sensor for Measuring Thickness Variation of Liquid Lithium Jet in IFMIF/EVEDA Lithium Test Loop

Abstract In the Engineering Validation and Engineering Design Activities (EVEDA) on the International Fusion Materials Irradiation Facility (IFMIF), hydraulic stability of a liquid Li jet simulating the IFMIF Li target is planned to be validated in the EVEDA Li Test Loop (ELTL). This paper presents the engineering design of a contact-type liquid level sensor for use in ELTL. The sensor is going to be utilized to measure variation of jet thickness in the validation test on hydraulic stability of the Li jet, which is one of the major key tests to be performed in ELTL. A fundamental requirement for the engineering specification of the sensor is to ensure the position accuracy of the measurement probe against the pressure load of approximately 0.1 MPa between the inside and the outside of the test chamber. The calculation result on structural strength of the sensor with a Nastran code showed that the maximum displacement was 0.65 mm and that the sensor has adequate strength against the pressure load. The calculation result on the sensor temperature with an ABAQUS code showed that the probe tip’s temperature can be heated up to approximately the operation temperature with no heaters installed on the sensor.

Experimental Study on Fire-Extinguishing of Lithium

Fire-extinguishing behavior of four fire extinguishants, dry sand, pearlite, Natrex-L and Natrex-M on burning lithium was examined. Temperature and flame increase in chemical reaction between lithium and silicon, which is the major element in the fire extinguishants, were observed for dry sand and pearlite. For Natrex-L, temperature increase was not observed visually, although flame was slightly increased when it was applied to the burning lithium. The effect of lithium pool depth on the fire-extinguishing performance of Natrex-L was investigated on the definite area of the lithium combustion surface because the density of Natrex-L was larger than that of liquid lithium. It was found that the amount (thickness) of fire extinguishant necessary for fire-extinguishing increased as the depth increased. In this experimental condition (combustion area: 270cm2 , lithium depth: 1–2cm), the minimum thickness of the fire extinguishant was 1.5 times the depth of the lithium pool.Copyright

Basic Experiment on Lithium Removal Technique

As a preliminary investigation into the establishment of a lithium removal technique for the components used at the International Fusion Materials Irradiation Facility (IFMIF), experiments were performed on the dissolution of lithium in three solvents: ethanol, pure water, and ethanol–water. In these experiments, hemispherical lithium was immersed in the solvents at constant temperatures, and the degree of dissolution was measured continuously from the height of the sample. From the obtained data, the average dissolution rate in the solvents at each testing temperature (10–90 °C) and the amount of hydrogen generated by the chemical reaction were calculated. The average dissolution rates in ethanol, pure water, and ethanol–water at 30 °C were 0.01, 1.6, and 0.43 mm/min, respectively. Although the average dissolution rate increased with the testing temperature in the low-temperature range (10–50 °C) for all solvents, this increase was saturated in the high-temperature range (50–90 °C) in experiments with pure water and ethanol–water as solvents. The volume of gas collected during each experiment was in good agreement with the volume of hydrogen assumed to be generated from the chemical reaction of lithium with the solvents.Copyright

Experimental Study on Lithium Leakage Behavior

In the Japanese engineering validation activities on the target facility of the International Fusion Materials Irradiation Facility, two studies on the advanced lithium leak detection system and the fire-extinguishing behavior of lithium were performed.In the former study, the fluorescence was distinguishable from the background using the difference of decay curves from the results of the lithium fluorescence measurement under two lithium concentration (256 ng/L and 4.58 ng/L) in argon. This result shows the basic applicability of the laser based lithium measuring technique as the lithium detector could be confirmed.In the latter study, fire-extinguishing behavior of two kinds of carbon base fire extionguishishants — MITEX and GRAPHEX — to lithium was examined. The both extinguishants showed excellent fire-extinguishing performance, and there were equivalent to that of Natlex-L obtained in the previous study. Since the bulk densities of the both carbon based extinguishants were about equal to the density of lithium, the subsidence of the fire-extinguishings which was observed in the Natlex-L was not observed. In fact, it was considered that no dependency of lithium pool depth on the required thickness of the fire extinguishants was observed for the carbon based extinguishants.Copyright

Material Improvement of Hydrogen Isotope Permeation Monitor for Liquid Lithium

Abstract Titanium plates coated by iron were fabricated electrochemically as a candidate hydrogen isotope permeation window for liquid lithium. Contacting the window to liquid lithium containing deuterium between 673 and 873 K, its permeation behavior and resistance to degradation were investigated. The iron-coated window showed less permeability than the bare titanium window as well as good agreement with the theoretical value. Furthermore, the iron layer at the vacuum side surface strongly improved oxidation resistance. Decrease of permeability for 2 hours air exposure could be recovered within 30 hours in the case of the coated window, while that of the bare titanium window could not. The coated permeation window is promising regarding high permeability and antidegradation properties.

Current Status of Design and Construction of IFMIF/EVEDA Lithium Test Loop

The International Fusion Materials Irradiation Facility (IFMIF) is a D+ -Li neutron source aimed at producing an intense high energy neutron flux (2 MW/m2 ) for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDA) of IFMIF started on July 2007. Regarding to the lithium (Li) target facility, design and construction of EVEDA Li Test Loop is a major activity and is in progress. This paper presents the current status of the design and construction of EVEDA Li Test Loop. The EVEDA Li Test Loop consists of a main loop system and a purification loop system. The detail design was started at the early 2009. Fabrication of the loop was started at middle of 2009, and completion is planned at the end of Feb. 2011. Currently, the system diagram of the EVEDA Li Test Loop is finished to be defined. The diagram and function of major components in the main loop system and the purification loop system are described in this paper.Copyright