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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

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04

Journal Articles

The CMMR program; BWR core degradation in the CMMR-3 test

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

Proceedings of International Conference on Dismantling Challenges; Industrial Reality, Prospects and Feedback Experience (DEM 2018) (Internet), 11 Pages, 2018/10

no abstracts in English

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

The CMMR program; BWR core degradation in the CMMR-1 and the CMMR-2 tests

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

Proceedings of 12th International Conference of the Croatian Nuclear Society; Nuclear Option for CO$$_{2}$$ Free Energy Generation (USB Flash Drive), p.109_1 - 109_15, 2018/06

no abstracts in English

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

A Simple method to create gamma-ray-source spectrum for passive gamma technique

Shiba, Tomooki; Maeda, Shigetaka; Sagara, Hiroshi*; Ishimi, Akihiro; Tomikawa, Hirofumi

Energy Procedia, 131, p.250 - 257, 2017/12

 Times Cited Count:0 Percentile:100

In the present paper, the $$gamma$$ ray source data was developed for the debris composition based on "best estimates", and the subsequent photon transportation calculation was performed to evaluate the leakage $$gamma$$ ray spectra according to the fuel debris. Since the creation of the line spectrum source requires a great deal, we have developed the relatively simple but accurate enough method to build up $$gamma$$ ray source, coupling of baseline spectra evaluated by ORIGEN2 code and several line spectra of interest. One of the advantages of the method is taking bremsstrahlung X rays into consideration by utilizing the bremsstrahlung libraries of ORIGEN2. The new $$gamma$$ ray source was used to calculate the detector response of HPGe detector and the results was compared as a benchmark with experimental measurement results of irradiated fuel pins. As the result, the simulated $$gamma$$ ray spectrum shape agreed well with the shape of $$gamma$$ ray spectrum obtained by the experiment.

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

 Times Cited Count:0 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

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

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:3 Percentile:54.39(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.

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

JAEA-Technology-2015-057.pdf:36.91MB

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

BB2013-2241.pdf:0.88MB

 Times Cited Count:11 Percentile:0.42

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:3 Percentile:61.74(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

JAEA-Data-Code-2014-012.pdf:82.93MB

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

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

 Times Cited Count:0 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.

JAEA Reports

Research for spectroscopy of fuel debris using superconducting phase transition edge sensor microcalorimeter; Measurement experiment and simulated calculation (Joint research)

Takasaki, Koji; Yasumune, Takashi; Onishi, Takashi; Nakamura, Keisuke; Ishimi, Akihiro; Ito, Chikara; Osaka, Masahiko; Ono, Masashi*; Hatakeyama, Shuichi*; Takahashi, Hiroyuki*; et al.

JAEA-Research 2013-043, 33 Pages, 2014/01

JAEA-Research-2013-043.pdf:13.81MB

In the Fukushima Daiichi Nuclear Power Plant, it is assumed that the core fuels melted partially or wholly, and the normal technique of accounting for a fuel assembly is not applicable. Therefore, it is necessary to develop the transparent and rational technique of accounting in the process of collection and storage of fuel debris. In this research, an application of the superconducting phase Transition Edge Sensor microcalorimeter (TES microcalorimeter) is studied for the accounting of nuclear materials in the fuel debris. It is expected that the detailed information of nuclear materials and fission products in fuel debris is obtained by using a high-resolution characteristic of TES microcalorimeter. In this report, the principle of TES microcalorimeter, the measurement experiment using TES in JAEA, and the simulated calculation using the EGS5 code system are summarized.

Journal Articles

Upgrading of X-ray CT technology for analyses of irradiated FBR MOX fuel

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:6 Percentile:47.47(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR); Post Irradiation Examination (PIE) toward investigation of the cause

Shinohara, Masanori; Motegi, Toshihiro; Saito, Kenji; Takada, Shoji; Ishimi, Akihiro; Katsuyama, Kozo

JAEA-Technology 2012-026, 21 Pages, 2012/08

JAEA-Technology-2012-026.pdf:2.31MB

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 outage by specifying the damaged part causing the event in the WRM. The one is a post irradiation examination using the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. The other is an experiment using a mock-up simulating the WRM fabricated by the fabricator. This report summarized the results of the PIE and the experimental investigation using the mock-up to reveal the cause of outage of WRM.

Journal Articles

Support system for training and education of future expert at PIE Hot Laboratories in Oarai JAEA; FEETS

Osaka, Masahiko; Donomae, Takako; Ichikawa, Shoichi; Sasaki, Shinji; Ishimi, Akihiro; Inoue, Toshihiko; Sekio, Yoshihiro; Miwa, Shuhei; Onishi, Takashi; Asaka, Takeo; et al.

Proceedings of 1st Asian Nuclear Fuel Conference (ANFC), 2 Pages, 2012/03

Support system for training and education of future expert in hot laboratories of Oarai-JAEA, named FEETS, is presented. The system has been established based on research results on both characterization of Oarai hot laboratory and user-needs. Various programs under FEETS are also introduced.

62 (Records 1-20 displayed on this page)