Operation and commissioning of IFMIF (International Fusion Materials Irradiation Facility) LIPAc injector

Okumura, Yoshikazu; Gobin, R.*; Knaster, J.*; Heidinger, R.*; Ayala, J.-M.*; Bolzon, B.*; Cara, P.*; Chauvin, N.*; Chel, S.*; Gex, D.*; et al.

Review of Scientific Instruments, 87(2), p.02A739_1 - 02A739_3, 2016/02

https://doi.org/10.1063/1.4936248

IFMIF is an accelerator based neutron facility having two set of linear accelerators each producing 125mA/CW deuterium ion beams (250mA in total) at 40MeV. The LIPAc (Linear IFMIF Prototype Accelerator) being developed in the IFMIF-EVEDA project consists of an injector, a RFQ accelerator, and a part of superconducting Linac, whose target is to demonstrate 125mA/CW deuterium ion beam acceleration up to 9MeV. The injector has been developed in CEA Saclay and already demonstrated 140mA/100keV deuterium beam. The injector was disassembled and delivered to the International Fusion Energy Research Center (IFERC) in Rokkasho, Japan, and the commissioning has started after its reassembly 2014; the first beam production has been achieved in November 2014. Up to now, 100keV/120mA/CW hydrogen ion beam has been produced with a low beam emittance of 0.2 .mm.mrad (rms, normalized).

Progress of the high current Prototype Accelerator for IFMIF/EVEDA

Okumura, Yoshikazu; Ayala, J.-M.*; Bolzon, B.*; Cara, P.*; Chauvin, N.*; Chel, S.*; Gex, D.*; Gobin, R.*; Harrault, F.*; Heidinger, R.*; et al.

Proceedings of 12th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.203 - 205, 2015/09

http://www.pasj.jp/web_publish/pasj2015/proceedings/PDF/FROL/FROL08.pdf

Under the framework of Broader Approach (BA) agreement between Japan and Euratom, IFMIF/EVEDA project was launched in 2007 to validate the key technologies to realize IFMIF. The most crucial technology to realize IFMIF is two set of linear accelerator each producing 125mA/CW deuterium ion beams up to 40MeV. The prototype accelerator, whose target is 125mA/CW deuterium ion beam acceleration up to 9MeV, is being developed in International Fusion Research Energy Center (IFERC) in Rokkasho, Japan. The injector developed in CEA Saclay was delivered in Rokkasho in 2014, and is under commissioning. Up to now, 100keV/120mA/CW hydrogen ion beams and 100keV/90mA/CW duty deuterium ion beams are successfully produced with a low beam emittance of 0.21 .mm.mrad (rms, normalized). Delivery of RFQ components will start in 2015, followed by the installation of RF power supplies in 2015.

Status quo of the injector for the IFMIF/EVEDA prototype accelerator

Shinto, Katsuhiro; Ichikawa, Masahiro; Takahashi, Yasuyuki*; Kubo, Takashi*; Tsutsumi, Kazuyoshi; Kikuchi, Takayuki; Kasugai, Atsushi; Sugimoto, Masayoshi; Gobin, R.*; Girardot, P.*; et al.

Proceedings of 11th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1009 - 1012, 2014/10

http://www.pasj.jp/web_publish/pasj2014/proceedings/PDF/SUP0/SUP025.pdf

The prototype accelerator is being developed as an engineering validation for the International Fusion Materials Irradiation Facility (IFMIF) equipped with an accelerator-driven-type neutron source for developing fusion reactor materials. This prototype accelerator is a deuteron linear accelerator consisting of an injector, an RFQ, a superconducting linac and their auxiliaries. It aims to produce a CW D beam with the energy and current of 9 MeV/125 mA. The injector test was completed at CEA/Saclay in 2012 for producing a CW H beam and a CW D beam with the energy and current of 100 keV/140 mA. After the beam test at CEA/Saclay, the injector was transported to the International Fusion Energy Research Centre (IFERC) located in Rokkasho, Aomori, Japan. In the end of 2013, installation of the injector was started at IFERC for the injector beam test beginning from summer 2014 in order to obtain better beam qualities to be satisfied with the injection and acceleration of the following accelerators. In this paper, some results of the injector beam test performed at CEA/Saclay and the status quo of the installation of the injector at IFERC are presented.

Corrosion behavior of Fe, Cr, Ni and their alloys in sodium molybdate melts

Nagai, Takayuki; Kikuchi, Kotaro*; Kano, Yoshiharu*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 12(1), p.76 - 88, 2013/03

https://doi.org/10.3327/taesj.J12.013

To understand the corrosion behavior of stainless steel and Ni-Cr alloys in sodium molybdate melts, the immersed corrosion tests of Fe, Cr, Ni, SUS316, 55Ni-45Cr, and Inconel 690 specimens were carried out in NaMoO and NaMoO melts at 750 C. The results of the immersed test, it was found that the corrosion rate decreased with an increase of chromium content in chemical composition of specimens. It was observed that CrO formed on the surface of Cr, 55Ni-45Cr, and Inconel 690 specimens by immersed in NaMoO melt, which corrosion rates were low in the melt, and we presumed that this CrO on the surface controlled the corrosion behavior of their material in the melt.

Analysis of test matrix and design status of test modules of IFMIF

Wakai, Eiichi; Kikuchi, Takayuki; Yokomine, Takehiko*; Yamamoto, Michiyoshi; Soldaini, M.*; Polato, A.*

Fusion Science and Technology, 62(1), p.246 - 251, 2012/07

https://doi.org/10.13182/FST12-A14142

Status of Japanese design and validation activities of test facilities in IFMIF/EVEDA

Wakai, Eiichi; Kikuchi, Takayuki; Kogawara, Takafumi; Kimura, Haruyuki; Yokomine, Takehiko*; Kimura, Akihiko*; Nogami, Shuhei*; Kurishita, Hiroaki*; Saito, Masahiro*; Nishimura, Arata*; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 6 Pages, 2011/03

Japanese activities of test facilities in IFMIF-EVEDA (International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities) project have three subjects of engineering design of post irradiation examination (PIE) facilities, small specimen test technique (SSTT), and engineering design of high flux test module (HFTM), and this paper is summarized about present status. Functional analysis and design of 2-D and 3-D models of PIE facility were performed. In HFTM, as materials of heater, W-3Re alloy and/or SiC/SiC composite were selected in the points of high temperature materials, fabrication technology and some suitable properties such as resistance of thermal shock, high temperature re-crystallization, ductility, resistance of irradiation degradation, and low-activation. In SSTT, a test machine of fracture toughness was designed and developed for small specimens with 10 mm square, and it had high accuracy controllability for stress and displacement.

Development of mechanical seal structure for reuse of re-instrumentation device used in fuel irradiation test

Inoue, Shuichi; Yamaura, Takayuki; Saito, Takashi; Ishikawa, Kazuyoshi; Kikuchi, Taiji; Sozawa, Shizuo; Tsuchiya, Kunihiko

JAEA-Technology 2009-076, 33 Pages, 2010/03

JAEA-Technology-2009-076.pdf:18.29MB

In Japan Material Testing Reactor (JMTR), a lot of experiments of fuel irradiation with the power ramping tests have been performed by using the shroud irradiation facility and the Boiling Water Capsule (BOCA). The fuel samples used in these tests were welded to re-instrumentation devices such as thermocouples and FP gas pressures. In this development, the mechanical connection method as "mechanical seal structure", that enables the re-use of re-instrumentation devices, was adopted in order to improve the utilization efficiency of the device. The test samples with mechanical seal structure were fabricated and the confirmatory tests such as He leakage test, thermal cycle test, autoclave test, etc. were carried out. The test samples with the mechanical seal structure showed an excellent result in various confirmatory tests, and the prospect are bright for the re-use of re-instrumentation devices with the mechanical seal structure.

Corrosion behavior of SUS316 clad in alkalis molybdate melt

Nagai, Takayuki; Kikuchi, Kotaro*; Kano, Yoshiharu*; Fukushima, Mineo

Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(4), p.370 - 379, 2008/12

https://doi.org/10.3327/taesj.J07.050

Study of a new pyrochemical processing using alkaline molybdate melt has been carrying out as a candidate reprocessing process for spent oxide fuels. In our previous study, we had confirmed that melting of UO pellet into a molten NaMoO-MoO and electrolytic recovering of UO from the melt included uranyl ions. In this report, the corrosion behavior of the SUS316 cladding tube in molten NaMoO-MoO was evaluated, and we confirmed the dissolution of SUS316 elements into the melt when the amount of additional MoO was excess. And, it is necessary to understand an appropriate amount of additional MoO since the result of this study.

Basis data for surveillance test on core support graphite structures for the High Temperature Engineering Test Reactor (HTTR)

Sumita, Junya; Shibata, Taiju; Kikuchi, Takayuki; Ishihara, Masahiro; Iyoku, Tatsuo; Sawa, Kazuhiro; Fujimoto, Nozomu

JAEA-Data/Code 2007-001, 57 Pages, 2007/02

JAEA-Data-Code-2007-001.pdf:3.0MB

Visual inspection by a TV camera and material properties measurement by surveillance test on core support graphite structures are planned for the High Temperature Engineering Test Reactor (HTTR) to confirm their structural integrity and characteristics. The surveillance test is aimed to investigate the change of material and mechanical properties by aging effects such as fast neutron irradiation and oxidation. The obtained data will be used not only for evaluating the structural integrity of the core support graphite structure of the HTTR but also for design data to advanced Very High Temperature Reactor (VHTR) discussed at generation IV international forum. This report describes the material properties and installed position of surveillance specimens in the HTTR in order to carry out the surveillance test.

Post Irradiation Examination for The FUGEN High Burn-up MOX Fuel Assembly (III) Final Report

Ikusawa,Yoshihisa; Kikuchi, Keiichi; Ozawa, Takayuki; Nakazawa, Hiroaki; Isozaki,Takao*; Nagayama, Masahiro*

JNC TN8410 2005-012, 113 Pages, 2005/08

JNC-TN8410-2005-012.pdf:15.2MB

The E09 fuel assembly was irradiated in the FUGEN from February 1990 to January 1997. The fuel assembly was the highest burn-up assembly in FUGEN and the pellet peak burn-up reached about 48 GWd/t. The E09 fuel assembly was transported to Japan Atomic Energy Research Institute (JAERI) Tokai in 2001. Post Irradiation Examinations (PIE) were started in July 2001, and all PIE items were completed by March 2005. The irradiation behavior of E09 MOX fuel was evaluated from the result of PIE. The major results are as follows; The integrity of E09 fuel assembly and fuel rods was confirmed. The corrosion behavior of ATR MOX fuel cladding was similar to that of LWR-UO2 fuel cladding. The central void was observed in outer ring samples irradiated with the maximum linear power over 45kW/m. A porous fine structure, similar to the rim structure seen in LWR-UO pellet, was observed in the circumferential region of MOX pellet and around the plutonium-rich spots. The MOX fuel properties irradiated up to ~48 GWd/t, which are pellet swelling, thermal conductivity, pellet melting temperature and diffusivity of fission gas, were similar to LWR-UO fuel properties. These results will be used for CANDU-OPTION program, which is one of Russian surplus weapon plutonium disposition programs with AECL in Canada, and available for LWR plutonium recycle program in Japan.