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Journal Articles

Challenge next-generation nuclear system; Development of oxide dispersion strengthened ferritic steel

Otsuka, Satoshi; Kaito, Takeji

Enerugi Rebyu, 39(1), p.44 - 46, 2019/01

For performance improvement of next-generation nuclear system such as fast reactor, it has been expected to develop advanced material resistant to severe in-reactor environment (i.e. high-dose neutron irradiation at high-temperature). Japan Atomic Energy Agency (JAEA) has been developing Oxide Dispersion Strengthened (ODS) ferritic steel for long life fuel cladding tube of fast reactor. Application of ODS ferritic steel to fast reactor fuel can extend the fuel life time twice or more as long as the fuel with conventional cladding tube (i.e. modified SUS316), thus reducing fuel exchange frequency and fuel cost. It can be adaptable to high-temperature plant operation, which is favorable for improvement of power generation efficiency. This paper interprets the development of ODS ferritic steel cladding tube for sodium-cooled fast reactor, which has been led by JAEA for dozens of years.

Journal Articles

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Journal Articles

Application of nontransfer type plasma heating technology for core-material-relocation tests in boiling water reactor severe accident conditions

Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Journal of Nuclear Engineering and Radiation Science, 4(2), p.020901_1 - 020901_8, 2018/04

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mm (height)). An excellent perspective in terms of applicability of the non-transfer plasma heating to melting high melting-temperature materials such as ZrO$$_{2}$$ has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO$$_{2}$$ fuel Phebus-FPT tests. Furthermore, application of EPMA, SEM/EDX and X-ray CT led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the Severe Accident (SA) experimental study was obtained.

Journal Articles

Irradiation performance of sodium-bonded control rod for the fast breeder reactor

Sasaki, Shinji; Maeda, Koji; Furuya, Hirotaka*

Journal of Nuclear Science and Technology, 55(3), p.276 - 282, 2018/03

 Times Cited Count:1 Percentile:38.14(Nuclear Science & Technology)

Journal Articles

Distributions of density and fission products in the reaction product between irradiated MOX fuel and molten zircaloy-2

Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Furuya, Hirotaka*

Journal of Nuclear Science and Technology, 54(11), p.1274 - 1276, 2017/11

 Percentile:100(Nuclear Science & Technology)

Two- and three-dimensional images were obtained in the reaction product between zircaloy and MOX fuel by X-ray CT. In addition, the $$gamma$$-ray intensity distributions of two fission products (Cs-137 and Eu-154) and one neutron-activated nuclide (Co-60) were obtained in this specimen by $$gamma$$-ray measurements. The average values of the fuel density (about 10.5 g/cm$$^{3}$$) and the cladding density (about 6.55 g/cm$$^{3}$$) were obtained in the metallic phase region by evaluation of the density distributions on two-dimensional X-ray CT images. In addition, the distributions of the roughly crushed fuel pellet and the pores in the specimen could be clearly observed on the three-dimensional X-ray CT images. From the $$gamma$$-ray measurement, Cs-137 was observed on the unreacted fuel region and the region where pores exist in the metallic phase, and Eu-154 was widely distributed to all regions. On the other hand, Co-60 was confirmed only in the metallic phase region.

Journal Articles

Defect chemistry and basic properties of non-stoichiometric PuO$$_{2}$$

Kato, Masato; Nakamura, Hiroki; Watanabe, Masashi; Matsumoto, Taku; Machida, Masahiko

Defect and Diffusion Forum, 375, p.57 - 70, 2017/05

The basic properties of PuO$$_{2-x}$$ were reviewed, and the equilibrium defects in PuO$$_{2-x}$$ were evaluated from the experimental data of the oxygen potential and electrical conductivity as well as the Ab-initio calculation results. Consistency among various properties was confirmed, and the mechanistic models for thermal property representations were derived.

Journal Articles

Oxygen potentials, oxygen diffusion coefficients and defect equilibria of nonstoichiometric (U,Pu)O$$_{2pm x}$$

Kato, Masato; Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Akashi, Masatoshi

Journal of Nuclear Materials, 487, p.424 - 432, 2017/04

 Times Cited Count:1 Percentile:64.68(Materials Science, Multidisciplinary)

Oxygen potential of (U,Pu)O$$_{2pm x}$$ was evaluated based on defect chemistry using an updated experimental data set. The relationship between oxygen partial pressure and deviation $$x$$ in (U,Pu)O$$_{2pm x}$$ was analyzed, and equilibrium constants of defect formation were determined as functions of Pu content and temperature. Brouwer's diagrams were constructed using the determined equilibrium constants, and a relational equation to determine O/M ratio was derived as functions of O/M ratio, Pu content and temperature. In addition, relationship between oxygen potential and oxygen diffusion coefficients were described.

Journal Articles

Development of non-transfer type plasma heating technology to address CMR behavior during severe accident with BWR design conditions

Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Journal Articles

Application of FE-SEM to the measurement of U, Pu, Am in the irradiated MA-MOX fuel

Sasaki, Shinji; Tanno, Takashi; Maeda, Koji

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 6 Pages, 2017/00

During irradiation in a fast reactor, the microstructure change of the mixed oxide fuels and the changes of element distributions occur because of a radial temperature gradient. Therefore, it is important to study the irradiation behavior of MA-MOX for advancement of fast reactor fuels. In order to make detailed observations of microstructure and elemental analyses of MA-MOX, irradiated MA-MOX specimens were carried out PIE by using a FE-SEM equipped with WDX. Because fuel samples have high radio activities and emit alpha-particles, the instrument was modified. the instrument was installed in a lead shield box and the control unit was separately located outside the box. The microstructure changes were observed in irradiated MA-MOX specimen. The characteristic X-rays peaks were detected successfully. By measuring the intensities of characteristic X-rays, it was tried quantitative analysis of U, Pu, Am along radial direction of irradiated specimen.

Journal Articles

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.

Journal Articles

Fuel performance of hollow pellets for fast breeder reactors

Ishimi, Akihiro; Katsuyama, Kozo; Kihara, Yoshiyuki; *

Journal of Nuclear Science and Technology, 53(7), p.951 - 956, 2016/07

 Times Cited Count:1 Percentile:76.09(Nuclear Science & Technology)

Three fuel rods containing hollow MOX pellets of uranium and plutonium oxides were fabricated and irradiated at a high linear heat rate to burn-up of nearly 30,000 MWd/t in the experimental fast reactor, JOYO MK-II. After irradiation, one of the fuel rods pellets was examined by X-ray CT and conventional nondestructive and destructive methods. Swelling rate was evaluated by both dimensional change and radial density distribution. There were no differences between both types of results and it was concluded that swelling rate can be examined in detail by the X-ray CT technique by dismantling the assembly. In addition, the swelling rate of hollow pellets was nearly the same as values reported for the fuel rods containing solid pellets. LHR was higher in the examined fuel rod containing hollow pellets than in the conventional fuel rod containing solid pellets, but fission gas release rates for both fuel rods were nearly the same. Since it is possible to decrease the maximum temperature in the radial center of the hollow fuel pellets, they can be effectively utilized in reactor operation at higher LHR.

Journal Articles

Preparation for a new experimental program addressing core-material-relocation behavior during severe accident with BWR design conditions; Conduction of preparatory tests applying non-transfer-type plasma heating technology

Abe, Yuta; Sato, Ikken; Ishimi, Akihiro; Nakagiri, Toshio; Nagae, Yuji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using ca. 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.

Journal Articles

The Research of MOX fuels in Japan

Kato, Masato

Transactions of the American Nuclear Society, 114, p.987 - 988, 2016/06

In Japan, uranium and plutonium mixed oxide (MOX) has been developed as fuels of sodium-cooled fast reactors. The developing MOX fuels come in variety of O/M ratio, Pu content, minor actinide (MA) content and density. We have studied a science based fuel technology to evaluate fuel behaviors in fabrication process and irradiation condition of such various fuels. The technologies which are constructed based on experimental database can apply to mechanistic evaluation of fuel behaviors. To develop the science based fuel technology, many different varieties of basic properties have been investigated, and experimental database was constructed. And a mechanistic physical property model has been studied. The models contribute to describe various behaviors in fuel fabrication process and irradiation condition.

JAEA Reports

Development of field emission SEM to observe high radioactive irradiated fuels

Isozaki, Misaki; Sasaki, Shinji; Maeda, Koji; Katsuyama, Kozo

JAEA-Technology 2015-058, 28 Pages, 2016/03


During irradiation in the fast reactor "JOYO", the changes of fuel structures with the formation of central void occur in the uranium-plutonium mixed oxide fuels (MOX fuels) because of radial temperature gradient. The changes of element (U, Pu, and so on) distributions along radial direction proceed from these changes. Therefore, it is important to study the changes of fuel structures of the minute area in fuel pellet and the changes of element distribution behavior for development of fast reactor fuels. In order to make detailed observations of microstructure and elemental analyses of fuel samples, a field emission scanning electron microscope (FE-SEM) equipped with a wavelength-dispersive X-ray spectrometer (WDS) and an energy-dispersive X-ray spectrometer (EDS) were installed in Fuel Monitoring Facility (FMF). The samples of this FE-SEM are very high radioactivity because the samples contain the nuclear fuel elements (U, Pu, etc.), the fission products (Cs, Rh, etc.) and activation product (Co, Mn etc.). Owing to this, it is necessary to prevent leakage of radioactive materials (particularly, U, Pu is need tight accountancy in law) and to protect operators from radiation. In this installation of FE-SEM, it is selected JSM-7001F (made by JEOL) for base model. The notable modified points were as follows. (1) To protect operators from radiation, lead shields was installed around FE-SEM. (2) To prevent leakage of radioactive materials, the instrument was attached to a remote control air-tight sample transfer unit between a shielded hot cell and the FE-SEM and the instrument was fixing rigid structure without vibration damper. (3) The design and manufacture the lead shields with consideration of instrument maintainability. This paper was described the summary of FE-SEM, the notable modified points, the ways of FE-SEM installation, the result of performance test.

JAEA Reports

Development of BDI behavior evaluation method in the fast reactor fuel assembly; Improvement of out-of-pile bundle compression test technology

Higashiuchi, Atsushi; Ishimi, Akihiro; Katsuyama, Kozo; Uwaba, Tomoyuki; Ichikawa, Shoichi

JAEA-Technology 2015-057, 72 Pages, 2016/03


Bundle-duct interaction (BDI) in fast reactors (FRs) is one of the limiting factors for burnup. To study the high performance fuel for FR fuel, it is important to establish the method to predict accurately the BDI behavior for the fuel assembly of large-diameter fuel pins. Therefore, it was adopted a new method that the bundle compression test apparatus is placed outside the cell, the bundle specimen is put in the airtight container for contamination prevention, and the bundle specimen is carried in the cell for internal observation by X-ray CT examination apparatus. From the result of this test, it was confirmed that the new method of out-of-pile bundle compression test is carried out as it was before. The results of this test are available to study integrity assessment of fast reactor fuel, validation of the BDI analysis code and substantiation of the safety design guidelines of fast reactor. In addition, it is possible to reflect in the BDI behavior evaluation for "ASTRID".

Journal Articles

Research program for the evaluation of fission product and actinide release behaviour, focusing on their chemical forms

Miwa, Shuhei; Yamashita, Shinichiro; Ishimi, Akihiro; Osaka, Masahiko; Amaya, Masaki; Tanaka, Kosuke; Nagase, Fumihisa

Energy Procedia, 71, p.168 - 181, 2015/05


 Times Cited Count:8 Percentile:0.64

A basic study towards enhanced safety management of irradiated fuels and materials from a severe accident is underway utilizing JAEA's hot laboratory complex in Oarai. The present study that consists of three basic research programs is aimed at contributing to building enhanced safety management measures (including radioactive decontamination, evaluation measurements, safekeeping, treatment and disposal) of irradiated fuels and materials from the severe accident. In this paper, not only the overview of activities of individual research programs but also the several preliminary results were shown together with future plans.

Journal Articles

Radial density distribution in irradiated FBR MOX fuel pellets

Ishimi, Akihiro; Katsuyama, Kozo; Nakamura, Hirofumi; Asaka, Takeo; Furuya, Hirotaka

Nuclear Technology, 189(3), p.312 - 317, 2015/03

 Times Cited Count:2 Percentile:66.76(Nuclear Science & Technology)

A high resolution X-ray CT technique was developed, which made it possible to obtain fine X-ray CT images of an irradiated fuel assembly. In addition, the density distributions in the irradiated MOX fuel pellet could be continually measured, using the relationship between the densities and CT values. These results were compared to the one obtained by metallographical method. As results, it was found that the relative change of radial density distributions in the irradiated fuel pellet can be measured more accurately by the X-ray CT technique than by the metallographical examination.

JAEA Reports

X-ray CT basic data about inspection of irradiated fuel assembly

Ishimi, Akihiro; Tachi, Yoshiaki; Katsuyama, Kozo; Misawa, Susumu*

JAEA-Data/Code 2014-012, 72 Pages, 2014/08


The Fuels Monitoring Section (FMS) of Japan Atomic Energy Agency (JAEA) has carried out examination of the fuel assemblies irradiated at Experimental Fast Reactor Joyo to verify about deformation and damage using X-ray computed tomography (CT) technique. This technique can observe deformation and internal information of the irradiated fuel assembly without dismantling and thus can apply to inspections of the irradiated fuel assembly in Fukushima Daiichi Nuclear Power Plant (1F). In order to obtain X-ray CT basic data for 1F fuel assembly inspection, the simulated specimens were made and the X-ray CT examinations were performed in the Fuels Monitoring Facility (FMF). This paper compiled the data about the X-ray CT examination of the simulated specimens.

Journal Articles

Distribution of radioactive nuclides of boring core samples extracted from concrete structures of reactor buildings in the Fukushima Daiichi Nuclear Power Plant

Maeda, Koji; Sasaki, Shinji; Kumai, Misaki; Sato, Isamu; Suto, Mitsuo; Osaka, Masahiko; Goto, Tetsuo*; Sakai, Hitoshi*; Chigira, Takayuki*; Murata, Hirotoshi*

Journal of Nuclear Science and Technology, 51(7-8), p.1006 - 1023, 2014/07

 Times Cited Count:5 Percentile:43.1(Nuclear Science & Technology)

Since the start of the severe accident at the Fukushima Daiichi Nuclear Power Plant in March 2011, concrete surfaces within the reactor buildings have been exposed to radioactive contaminants. Released radiation sources still remain too high to permit entry into some areas of the RBs to allow the damage to be assessed and to allow carrying out the restoration of lost safety functions, decommissioning activities, etc. In order to clarify the situation of this contamination in the RBs, 18 samples were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. Decontamination tests on the sample of Unit 2 were conducted to reduce the levels of radioactivity present near the sample surface. As a result of the tests, the level of radioactivity of the sample was reduced with the removal of 97% of the contamination present near the sample surface.

Journal Articles

Investigation of strand bending in the He-inlet during reaction heat treatment for ITER TF Coils

Hemmi, Tsutomu; Matsui, Kunihiro; Kajitani, Hideki; Okuno, Kiyoshi; Koizumi, Norikiyo; Ishimi, Akihiro; Katsuyama, Kozo

IEEE Transactions on Applied Superconductivity, 24(3), p.4802704_1 - 4802704_4, 2014/06

 Percentile:100(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency (JAEA), as Japan Domestic Agency, has responsibility to procure nine ITER Toroidal Field (TF) coils. The TF coil winding consists of a Nb$$_{3}$$Sn Cable-In-Conduit conductor, a pair of joints and a He-inlet. The current capacity of 68 kA is required at the magnetic field of 7 T around the He-inlet region in the TF coil winding. During reaction heat-treatment, the compressive residual strain in Nb$$_{3}$$Sn cable is induced by the difference in the thermal expansion coefficients between the Nb$$_{3}$$Sn cable and stainless steel jacket. The strands bending in the Nb$$_{3}$$Sn cable of the He-inlet is anticipated since there is the compressive residual strain and a gap between the Nb$$_{3}$$Sn cable and the He-inlet to introduce SHE flow. If the strand is bent, the variation of mechanical behaviors, such as the elongation of He-inlet during the reaction heat-treatment and the thermally induced residual strain on the jacket around the He-inlet, are expected. To investigate the strands bending in the Nb$$_{3}$$Sn cable of the He-inlet, the following items are performed; (1) elongation measurement during reaction heat-treatment, (2) residual longitudinal strain measurement using strain gauges by sample cuttings, (3) nondestructive inspection on the cable and strands using high resolution X-ray CT, Detail of test results and investigation of the strands bending in the Nb$$_{3}$$Sn cable of the He-inlet are reported and discussed.

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