Decommissioning activities at Ningyo-toge Environmental Engineering Center

Kihara, Yoshiyuki

RANDEC Nyusu, (112), 1 Pages, 2019/07

http://www.randec.or.jp/publish/randecnews.html

In December 2016, we announced our new concept of the "Uranium and Environmental Research Platform". To move forward with this concept, we believe it is important to explain the R&D plan and its outcomes in an open venue and to ensure the reliability and transparency of the project through two-way communication with local communities. We established discussion meetings to receive opinions and recommendations from local citizens and experts. As a result of the meetings from June 2017 to March 2019, opinions and recommendations were shown that it is appropriate to conduct R&D shown in the concept of "Uranium and Environmental Research Platform" in order to reduce risks and environmental conservation efforts. In addition, we recognized that it is important to discuss through two-way communication with local communities again. We will continue activities to consider problems for Back-end projects, while utilizing experiences accumulated for more than 60 years through Front-end projects.

Development of advanced PIE technique for fuel debris using X-Ray computer tomography

Katsuyama, Kozo; Ishimi, Akihiro; Tanaka, Kosuke; Kihara, Yoshiyuki

Proceedings of 53rd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2016) (Internet), 5 Pages, 2016/11

Following the Fukushima Daiichi Nuclear Power Plant accident, a feasibility study on the application of X-ray CT technique for observation of the inner condition of the fuel debris was initiated. First, a preliminary test was performed using a dummy specimen of irradiated fuel pellets, which was heated to 2373 K. As a result, we obtained high resolution X-ray CT images in which the small pieces of fuel pellets could be clearly distinguished from one another. Analyzing these X-ray CT images enables us to know the density distribution of the fuel debris.

Development of PRW welding technology for 9Cr-ODS cladding tube

Seki, Masayuki; Kihara, Yoshiyuki; Kaito, Takeji; Tsukada, Tatsuya*; Motoki, Kazuhiko*; Hirako, Kazuhito*

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Oxide Dispersion Strengthened (ODS) steel has been developed as an advanced fuel cladding tube for sodium cooled fast reactors in Japan Atomic Energy Agency (JAEA) to attain the target burn up of 150 GWd/t in the bundle average because of its excellent swelling resistance and high mechanical strength in high temperature. If conventional TIG welding is applied to the ODS welding, it is difficult to obtain necessary mechanical strength at the weld zone because of the formation of porosity. It is formed by the argon bubbles which initially dissolve in the matrix and grow up under the high temperature during welding. Therefore JAEA has been conducted the development of pressurized resistance welding (PRW) technology for ODS cladding tube, which is one of the solid state welding methods. This paper describes in the development of PRW technology, an ultrasonic test method for detecting weld defects, the result of machine strength measurement examination in weld part and the result of fuel pin irradiation examination using nuclear reactor.

Development of oxygen-to-metal ratio of MOX pellet adjustment technology for the simplified MOX pellet fabrication method in the FaCT project

Takano, Tatsuo; Sudo, Katsuo; Takeuchi, Kentaro; Kihara, Yoshiyuki; Kato, Masato

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12

Development of high burn-up fuels is essential to improve economy of the fast reactor fuel cycle. Increase of fuel burn-up is known to cause fuel-cladding chemical interaction (FCCI) and it mainly determines a lifetime of fuel pin. In order to extend a lifetime of fuel pin by mitigating FCCI, development of low oxygen-to-metal (O/M) MOX fuel has been carried out in plutonium fuel development center of JAEA. MOX fuel needs adjustment of the O/M ratio to less than 1.97 for high burn-up of 150 GWd/t. Therefore, O/M adjustment technology is one of the main subjects in development of a simplified MOX pellet fabrication method which has been advanced in the FaCT (Fast reactor Cycle Technology development) project. In previous work, changes in O/M ratio of MOX pellet during heat treatment were calculated from measurement results of oxygen potentials. On the basis of above calculation, heating tests were carried out to prepare low O/M ratio MOX pellets on a laboratory scale. The O/M ratios obtained in the heating tests were well consistent with calculation results. In the present study, a kilogram MOX scale furnace to adjust O/M ratio of MOX pellets for targeted value has been developed as next step.

Physical properties and irradiation behavior analysis of Np- and Am-bearing MOX fuels

Kato, Masato; Maeda, Koji; Ozawa, Takayuki; Kashimura, Motoaki; Kihara, Yoshiyuki

Journal of Nuclear Science and Technology, 48(4), p.646 - 653, 2011/04

https://doi.org/10.3327/jnst.48.646

Physical properties and irradiation behavior of minor actinide-bearing MOX were evaluated for the development of advanced fast reactor fuels. The physical properties of the fuels were described as functions of minor actinide content and oxygen-to-metal (/) ratio, and the effect of minor actinide addition into MOX on those properties was found to be negligibly small. The irradiation tests of fuel pellets having / ratios of 1.98 or 1.96 were carried out at high linear heat rate of about 430W/cm. The redistribution of actinide element and oxygen were analyzed by using the evaluated properties, and maximum temperatures of the pellets were estimated. The maximum temperature of the pellets of /=1.96 was estimated to reach 2680K which was 130K higher than that of /=1.98 pellets. The maximum temperature of the pellet was lower as compared with its melting temperature of higher than 3000K. In the results of post-irradiation examination, no trace of melting was observed.

Innovative powder production and granulation for advanced MOX fuel fabrication

Kurita, Tsutomu; Kato, Yoshiyuki; Yoshimoto, Katsunobu; Suzuki, Masahiro; Kihara, Yoshiyuki; Fujii, Kanichi

Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.94 - 102, 2009/09

With regard to advanced MOX fuel fabrication, a new concept in which one vessel especially designed to meet microwave de-nitration is utilized also for crushing and for granulation, without organic lubricant nor powder transfer across the processes, was introduced for innovative MOX powder production. In order to realize this concept, two attempts were made: A specially designed three blade impeller coupled with auxiliary blade. A uniquely shaped mixing blade coupled with an auxiliary blade having auto-orbital hybrid rotation. The mixing blade promotes the growth of particles, whereas the auxiliary blade suppresses the overgrowth by chopping larger particles. These granulators use a little water as binder. As a result, major diameter of granule 400-1000 micron and flow-ability 82-85 was obtained with fine WO model powder. Therefore, a prospect to satisfy both desirable powder properties and simplified nuclear material production was opened, as well as improvement of working efficiency and cut down on costs.

The Phase state at high temperatures in the MOX-SiO system

Nakamichi, Shinya; Kato, Masato; Sunaoshi, Takeo*; Uchida, Teppei; Morimoto, Kyoichi; Kashimura, Motoaki; Kihara, Yoshiyuki

Journal of Nuclear Materials, 389(1), p.191 - 196, 2009/05

https://doi.org/10.1016/j.jnucmat.2009.01.028

Japan Atomic Energy Agency researchers have developed mixed oxide (MOX) fuels containing minor actinides (MA). These fuels were irradiated for ten minutes in the FBR Joyo in some short-term irradiation tests. The Si-condensed phases were observed at the center of the pellets in the post irradiation examination. Si impurities came to be mixed into the raw materials in the ball milling process, because Si rubber was used as the lining of the milling pot. Content of Si in the pellets was within the specification of the fuel. It is important to investigate the Si state in MOX at high temperatures like the reactor operating temperature of the fuel to evaluate irradiation behavior. In the present work, MOX specimens with mixed SiO impurity were prepared. The ratio of MOX to SiO was controlled at a mol fraction of 3 to 1. The specimens were first heated at 1973K in atmospheres of three different oxygen partial pressures to adjust the O/M ratio. Then these specimens were sealed in a tungsten capsule, and heated at 2273K or 2673K. Compounds consisting of Pu and Si were observed at grain boundaries of the MOX matrix in specimens after heat treatment. These compounds were not observed in grain interior and MOX matrix was not affected significantly by Si impurity. These compounds tended to form in specimens with low O/M ratio and in specimens heated at higher temperatures.

Feasibility study on particle flowability evaluation in simplified MOX pellet fabrication process using vibrating tube method

Ishii, Katsunori; Suzuki, Masahiro; Yamamoto, Takuma; Kihara, Yoshiyuki; Yasuda, Masatoshi*; Matsusaka, Shuji*

Funtai Kogakkai-Shi, 45(5), p.290 - 296, 2008/05

http://ci.nii.ac.jp/naid/10021172223

Particle flowability test using a vibrating tube with a tapered end was carried out to evaluate the applicability of this method to the analysis of MOX particles for nuclear fuel pellet fabrication. In this experiment, zirconia particles were used as model particles, which are less than several hundred micrometers in diameter. The sample particles were prepared by mixing the particles with different sizes, whose shape and surface state were changed by a mechanical treatment. The experimental results showed that the removal of smaller particles is more effective than the mechanical surface treatment to improve the flowability. Furthermore, it was found that the vibrating tube method can identify the difference of the particle flowability more easily and clearly than the method based on Carr's flowability index.

FUJI, an initial sintering comparison test for pelletized-, sphere-pac- and vipac-fast breeder reactor mixed oxide fuel

Bart, G.*; Bakker, K.*; Hellwig, C.*; Kihara, Yoshiyuki; Ozawa, Takayuki; Wallin, H.*; Shigetome, Yoshiaki

Journal of Nuclear Science and Technology, 44(3), p.329 - 336, 2007/03

https://doi.org/10.3327/jnst.44.329

Within the FUJI collaboration program among Japan Atomic Energy Agency (JAEA, former JNC), Paul Scherrer Institute (PSI, Switzerland) and Nuclear Research and Consultancy Group (NRG, the Netherlands) the production paths of plutonium and neptunium mixed oxide- (sphere-pac- and vipac-) particle fuels (20wt% Pu and 5wt% Np) are tested as well as the initial sintering and power-to-melt behavior under simulated FBR conditions. The various fuel forms were produced at PSI under the support of JNC, the irradiations were accomplished at High Flux Reactor (HFR) in Petten, the post irradiation examinations are being achieved mainly at NRG and the fuel modelling being performed at JNC and PSI. The present paper reviews mainly the project planning, fuel behaviour- pre-calculations and the fuel- and fuel segment- production. A short overview of the irradiation conditions and ceramographic post irradiation examination analyses is also given.

Report of the collaboration project for research and development of sphere-pac fuel among JNC-PSI-NRG (II); Irradiation tests and post irradiation examinations (Joint research)

Nakamura, Masahiro; Ozawa, Takayuki; Morihira, Masayuki; Kihara, Yoshiyuki

JAEA-Research 2006-028, 146 Pages, 2006/03

JAEA-Research-2006-028.pdf:37.95MB

The collaboration project concerning sphere-pac fuel among JNC (Japan Nuclear Cycle Development Institute, now Japan Atomic Energy Agency), Swiss PSI (Paul Scherrer Institut) and Dutch NRG (Nuclear Research and Consultancy Group) was performed from 1996 till 2005. The target of this project is comparative irradiation tests of sphere-pac fuel in the HFR (High Flux Reactor) in Petten in the Netherlands with pellet fuel and vipac fuel. Total 16 fuel segments (8 pins) containing 5%Np-MOX sphere-pac segments were irradiated. No fuel failure was occurred. Restructuring of sphere-pac fuel was quickly progressed in early stage of irradiation, and formation of the central hole was almost completed within 48 hours steady state irradiation. According to the results of the power-to-melt test, the power to melt linear heat rates were estimated as 60kW/m for the sphere-pac fuel and as 73kW/m for the pellet fuel under HFR irradiation conditions. Irradiation behaviors of the vipac fuel and Np-MOX sphere-pac fuel were basically similar to that of the MOX sphere-pac fuel. However, the central hole of the Np-MOX sphere-pac fuel was larger than that of MOX sphere-pac fuel in the restructuring test. It suggests that the thermal conductivity of Np-MOX fuel is smaller than that of MOX fuel.