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Onishi, Takashi; Maeda, Koji; Katsuyama, Kozo
Journal of Nuclear Science and Technology, 58(4), p.383 - 398, 2021/04
Times Cited Count:10 Percentile:73.46(Nuclear Science & Technology)Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.
Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08
Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.
Ishimi, Akihiro; Katsuyama, Kozo; Furuya, Hirotaka*
Journal of Nuclear Science and Technology, 56(11), p.981 - 987, 2019/07
Times Cited Count:3 Percentile:28.63(Nuclear Science & Technology)Both high and low density MOX fuel pellets of uranium and plutonium oxides were irradiated in the experimental fast reactor JOYO. After irradiation, these fuel pellets were examined by X-ray CT and their irradiation behavior was evaluated for formation of the central void. In particular, the central void size and temperature of fuel pellets at the beginning and end of irradiation were analyzed. The central voids in the low density fuel pellets were bigger than those of the high density fuel pellets at the same linear heating rate (LHR), and the threshold LHR and temperature at which the central voids were formed were lower than those of the high density fuel pellets. It was understood from these results that the irradiation behaviors of high and low density fuel pellets were different.
Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Furuya, Hirotaka*
Journal of Nuclear Science and Technology, 54(11), p.1274 - 1276, 2017/11
Times Cited Count:1 Percentile:9.95(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 -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 -ray measurements. The average values of the fuel density (about 10.5 g/cm) and the cladding density (about 6.55 g/cm) 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 -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.
Kihara, Yoshiyuki; Tanaka, Kosuke; Koyama, Shinichi; Yoshimochi, Hiroshi; Seki, Takayuki; Katsuyama, Kozo
NEA/NSC/R(2017)3, p.341 - 350, 2017/11
In order to investigate the effect of the addition of americium to MOX fuels on the irradiation behaviour, the "Am-1" program is being conducted at the JAEA. The Am-1 program consists of two short-term irradiation tests of 10-min and 24-h irradiation periods, and a steady-state irradiation test. The short-term irradiation tests and their post irradiation examinations (PIEs) have been successfully completed. To date, the data for PIE of the Am-MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation have been obtained and reported. In this paper, the results obtained from the Am-1 program are reviewed and detailed descriptions of the fabrication and inspection techniques for the Am-MOX fuels prepared for the program are provided. PIE data for the Am-MOX fuels at the initial stage of irradiation have been accumulated. In this paper, unpublished PIE data for the Am-MOX fuels are also presented.
Ishikawa, Takashi; Onishi, Takashi; Hirosawa, Takashi; Tanaka, Kosuke; Katsuyama, Kozo
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 10 Pages, 2017/00
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.
Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 2(4), p.041008_1 - 041008_5, 2016/10
Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.
Ishimi, Akihiro; Katsuyama, Kozo; Kihara, Yoshiyuki;
Journal of Nuclear Science and Technology, 53(7), p.951 - 956, 2016/07
Times Cited Count:4 Percentile:34.69(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.
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.
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".
Tanaka, Kosuke; Sasaki, Shinji; Katsuyama, Kozo; Koyama, Shinichi
Transactions of the American Nuclear Society, 113(1), p.619 - 621, 2015/10
In order to evaluate the microstructural change behavior of Am-MOX fuels at the initial stage of irradiation, detailed investigations using image analysis were performed on X-ray Computed Tomography (X-ray CT) images and on ceramographs from fuels irradiated in both B11 and B14.
Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.
Ishimi, Akihiro; Katsuyama, Kozo; Nakamura, Hirofumi; Asaka, Takeo; Furuya, Hirotaka
Nuclear Technology, 189(3), p.312 - 317, 2015/03
Times Cited Count:4 Percentile:31.62(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.
Uwaba, Tomoyuki; Ito, Masahiro*; Nemoto, Junichi*; Ichikawa, Shoichi; Katsuyama, Kozo
Journal of Nuclear Materials, 452(1-3), p.552 - 556, 2014/09
Times Cited Count:1 Percentile:8.38(Materials Science, Multidisciplinary)The BAMBOO code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle-duct interaction (BDI) condition. The pin diameters of were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype FBR and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT)images and local parameters of bundle deformation were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms.
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.
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
Times Cited Count:1 Percentile:9.23(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 NbSn 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 NbSn cable is induced by the difference in the thermal expansion coefficients between the NbSn cable and stainless steel jacket. The strands bending in the NbSn cable of the He-inlet is anticipated since there is the compressive residual strain and a gap between the NbSn 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 NbSn 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 NbSn cable of the He-inlet are reported and discussed.
Uwaba, Tomoyuki; Ichikawa, Shoichi; Katsuyama, Kozo
JAEA-Research 2013-039, 25 Pages, 2014/02
Bundle-Duct Interaction (BDI) in core fuel subassemblies in FBRs is a limiting factor for fuel burnup. Thus, BDI is an important evaluation item in the upgraded core of the Monju prototype FBR and the demonstration FBR studied in the FaCT project because the fuel subassemblies are to be used to high burnup condition. Since fuel subassemblies of these FBRs consists of large diameter fuel pins, the out-of-pile bundle compression test with large diameter pins was performed to evaluate their BDI bundle. In the compression test, bundle cross-sectional images (CT images) were obtained by using the X-ray computer tomography. The CT images were numerically analyzed to evaluate the deformation of pin bundles due to BDI. The evaluation results revealed that deformation of large diameter pin bundles are controlled by pin bowing and cladding oval-distortion the same as in the case of currently used small diameter pin bundles.
Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Nagamine, Tsuyoshi; Asaka, Takeo; Furuya, Hirotaka
Journal of Nuclear Science and Technology, 49(12), p.1144 - 1155, 2012/12
Times Cited Count:8 Percentile:51.26(Nuclear Science & Technology)no abstracts in English
Shinohara, Masanori; Motegi, Toshihiro; Saito, Kenji; Haga, Hiroyuki; Sasaki, Shinji; Katsuyama, Kozo; Takada, Kiyoshi*; Higashimura, Keisuke*; Fujii, Junichi*; Ukai, Takayuki*; et al.
JAEA-Technology 2012-032, 29 Pages, 2012/11
An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of HTGR. Then, two experimental investigations were carried out to reveal the cause of the malfunction by specifying the damaged part causing the event in the WRM. One is an experiment using a mock-up sample test which strength degradation on assembly accuracy and heat cycle to specify the damaged part in the WRM. The other is a destructive test in FMF to specify the damaged part in the WRM. This report summarized the results of the destructive test and the experimental investigation using the mock-up to reveal the cause of malfunction of WRM.