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JAEA Reports

Analysis of deposits inside the reactor at Fukushima Daiichi Nuclear Power Station in JFY 2017-2018; The Subsidy programs "Project of Decommissioning and Contaminated Water Management in the FY2016 Supplementary Budget, (Development of Technologies for Grasping and Analyzing Properties of Fuel Debris)

Nakayoshi, Akira; Mitsugi, Takeshi; Sasaki, Shinji; Maeda, Koji

JAEA-Data/Code 2021-011, 279 Pages, 2022/03

JAEA-Data-Code-2021-011.pdf:37.76MB

At the TEPCO's Fukushima Daiichi Nuclear Power Station (1F), an investigation inside the reactors has been carried out, and R&D has been made on methods of fuel debris retrieval and storage after retrieval. In order to carry out the decommissioning work safely and steadily, understanding characteristics of fuel debris in the reactors is required. Therefore, in the development of technologies for grasping and analyzing properties of fuel debris project, the characteristics of simulated fuel debris, such as hardness, drying behavior, etc., of fuel debris for design of removal and storage, have been investigated and estimated, and provided to other projects conducting the decommissioning work. As part of this project, U-containing particles in samples (e.g., deposit on the investigation equipment, sediment in the reactors, etc.) obtained during the internal investigation of the reactors of 1F units 1 to 3 were analyzed. This report summarized the results of FE-SEM/WDX, FE-SEM/EDS, STEM/EDS, and TEM analysis, which were extracted from all analysis results obtained, as a database for the evaluation of the generation mechanism of U-containing particles. The analyses were performed at the JAEA Oarai Research and Development Institute and Nippon Nuclear Fuel Development Co., LTD.

Journal Articles

Review of Fukushima Daiichi Nuclear Power Station debris endstate location in OECD/NEA preparatory study on analysis of fuel debris (PreADES) project

Nakayoshi, Akira; Rempe, J. L.*; Barrachin, M.*; Bottomley, D.; Jacquemain, D.*; Journeau, C.*; Krasnov, V.; Lind, T.*; Lee, R.*; Marksberry, D.*; et al.

Nuclear Engineering and Design, 369, p.110857_1 - 110857_15, 2020/12

 Times Cited Count:0 Percentile:0.33(Nuclear Science & Technology)

Much is still not known about the end-state of core materials in each of the units at Fukushima Daiichi Nuclear Power Station (Daiichi) that were operating on March 11, 2011. The Nuclear Energy Agency of the Organization for Economic Development has launched the Preparatory Study on Analysis of Fuel Debris (PreADES) project as a first step to reduce some of these uncertainties. As part of the PreADES Task 1, relevant information was reviewed to confirm the accuracy of graphical depictions of the debris endstates at the damaged Daiichi units, which provides a basis for suggesting future debris examinations. Two activities have been completed within the PreADES Task 1. First, relevant knowledge from severe accidents at the Three Mile Island Unit 2 and the Chernobyl Nuclear Power Plant Unit 4 was reviewed, along with results from prototypic tests and hot cell examinations, to glean insights that may inform future decommissioning activities at Daiichi. Second, the current debris endstate diagrams for the damaged reactors at Daiichi were reviewed to confirm that they incorporate relevant knowledge from plant observations and from severe accident code analyses of the BSAF (Benchmark Study of the Accident at Daiichi Nuclear Power Station) 1 and 2 projects. This paper highlights Task 1 insights, which have the potential to not only inform future Decontamination and Decommissioning activities at Daiichi, but also provide important perspectives for severe accident analyses and management, particularly regarding the long term management of a damaged nuclear site following a severe accident.

Journal Articles

Leaching behavior of prototypical Corium samples; A Step to understand the interactions between the fuel debris and water at the Fukushima Daiichi reactors

Nakayoshi, Akira; Jegou, C.*; De Windt, L.*; Perrin, S.*; Washiya, Tadahiro

Nuclear Engineering and Design, 360, p.110522_1 - 110522_18, 2020/04

 Times Cited Count:5 Percentile:84.6(Nuclear Science & Technology)

Journal Articles

Material characterization of the VULCANO corium concrete interaction test with concrete representative of Fukushima Daiichi Nuclear Plants

Brissonneau, L.*; Ikeuchi, Hirotomo; Piluso, P.*; Gousseau, J.*; David, C.*; Testud, V.*; Roger, J.*; Bouyer, V.*; Kitagaki, Toru; Nakayoshi, Akira; et al.

Journal of Nuclear Materials, 528, p.151860_1 - 151860_18, 2020/01

 Times Cited Count:5 Percentile:89.87(Materials Science, Multidisciplinary)

Journal Articles

Current situation of OECD/NEA, Preparatory Study on Analysis of Fuel debris (PreADES) project

Nakayoshi, Akira; Journeau, C.*; Rempe, J.*; Barrachin, M.*; Bottomley, D.; Nauchi, Y.*; Song, J. H.*

Proceedings of 2019 International Workshop on Post-Fukushima Challenges on Severe Accident Mitigation and Research Collaboration (SAMRC 2019) (USB Flash Drive), 6 Pages, 2019/11

Journal Articles

Knowledge obtained from dismantling of large-scale MCCI experiment products for decommissioning of Fukushima Daiichi Nuclear Power Station

Nakayoshi, Akira; Ikeuchi, Hirotomo; Kitagaki, Toru; Washiya, Tadahiro; Bouyer, V.*; Journeau, C.*; Piluso, P.*; Excoffier, E.*; David, C.*; Testud, V.*

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

Journal Articles

Large scale VULCANO molten core concrete interaction test considering Fukushima Daiichi condition

Bouyer, V.*; Journeau, C.*; Haquet, J. F.*; Piluso, P.*; Nakayoshi, Akira; Ikeuchi, Hirotomo; Washiya, Tadahiro; Kitagaki, Toru

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 13 Pages, 2019/03

Journal Articles

Current situation of OECD/NEA, Preparatory Study on Analysis of Fuel debris (PreADES) project

Nakayoshi, Akira; Bottomley, D.; Washiya, Tadahiro

Proceedings of 56th Annual Meeting on Hot Laboratories and Remote Handling (HOTLAB 2019) (Internet), 3 Pages, 2019/00

Journal Articles

Prediction of the drying behavior of debris in Fukushima Daiichi Nuclear Power Station for dry storage

Nakayoshi, Akira; Suzuki, Seiya; Okamura, Nobuo; Watanabe, Masayuki; Koizumi, Kenji

Journal of Nuclear Science and Technology, 55(10), p.1119 - 1129, 2018/10

 Times Cited Count:1 Percentile:17.98(Nuclear Science & Technology)

Journal Articles

Investigation of a LiCl-KCl-UCl$$_{3}$$ system using a combination of X-ray diffraction and differential thermal analyses

Nakayoshi, Akira; Kitawaki, Shinichi; Fukushima, Mineo; Murakami, Tsuyoshi*; Kurata, Masaki

Journal of Nuclear Materials, 441(1-3), p.468 - 472, 2013/10

 Times Cited Count:12 Percentile:70.63(Materials Science, Multidisciplinary)

Electrorefining is one of the main steps of pyroreprocessing where spent nuclear fuels are recycled. Electrorefining is conducted in a molten salt of LiCl-KCl eutectic (59:41 mol%) containing actinide chlorides (AnCl$$_{3}$$) at 773 K. In order to operate and maintain the electrorefiner, it is necessary to accumulate fundamental data on LiCl-KCl-AnCl$$_{3}$$ salt such as the melting point. In this study, based on X-ray diffraction and differential thermal analysis, a partial phase diagram of (LiCl-KCl)eut.-UCl$$_{3}$$ pseudo-binary system and partial phase diagram of LiCl-KCl-UCl$$_{3}$$ system were developed, which UCl$$_{3}$$ concentration was up to 20 mol%.

Journal Articles

Electrorefining test of U-Pu-Zr alloy fuel prepared pyrometallurgically from MOX

Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Sakamura, Yoshiharu*; Murakami, Tsuyoshi*; Akiyama, Naoyuki*

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/09

In the FaCT project, the metal fuel cycle including metal fuel fast reactor and pyrochemical reprocessing has been being developed. JAEA and CRIEPI have continued a collaborative study on pyrochemical reprocessing. In the pyrochemical reprocessing, actinides in the spent fuels dissolve anodically in the LiCl-KCl, and U is collected selectively on a solid cathode, Pu and MA are recovered simultaneously in a liquid Cd cathode. In the previous electrorefining tests, at the anode Zr was allowed to dissolve into the electrolyte salt together with U, Pu and MA. The Zr co-dissolution may cause some problems. In this study, through the anode dissolution test of U-Pu-Zr alloy fuel, the controlling the dissolution of the Zr and the improvement of dissolution ratio of U, Pu were studied. The U-Pu alloy was prepared from MOX pellets by using the electrochemical reduction method. U-Pu-Zr ternary alloy was produced by alloying the obtained U-Pu alloy and prepared U-Zr alloy. U-Pu-Zr ternary alloy was immersed into electrolyte salt, and electrolysis test was carried out.

Journal Articles

Anodic behaviour of a metallic U-Pu-Zr alloy during electrorefining process

Murakami, Tsuyoshi*; Sakamura, Yoshiharu*; Akiyama, Naoyuki*; Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo

Journal of Nuclear Materials, 414(2), p.194 - 199, 2011/07

 Times Cited Count:13 Percentile:74.14(Materials Science, Multidisciplinary)

An electrorefining is one of the main steps of pyrochemical reprocessing of spent metallic fuels (U-Zr, U-Pu-Zr). The electrorefining is carried out dissolving a portion of Zr together with actinides to accomplish a high dissolution ratio of actinides. However, the electrorefining with Zr co-dissolution should bring some practical problems in the pyrochemical reprocessing. Therefore, electrorefining tests of non-irradiated U-Pu-Zr alloy were performed with minimizing the amount of Zr dissolved in LiCl-KCl-(U, Pu, Am)Cl$$_{3}$$ melts at 773 K. The tests were performed both by potentiostatic electrolysis at -1.0 V (Ag$$^{+}$$/Ag) that was more negative than the Zr dissolution potential and by galvanostatic electrolysis with a limited amount of Zr dissolution. The ICP-AES analysis of the anode residues confirmed that a high dissolution ratio of actinides (U; $$>$$ 99.6%, Pu; 99.9%) was successfully demonstrated at both electrolyses.

Journal Articles

Recent progress of JAEA-CRIEPI joint study for metal pyroreprocessing at CPF

Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Koizumi, Tsutomu; Kurata, Masaki*; Yahagi, Noboru*

Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1269 - 1273, 2009/09

JAEA is developing the pyroreprocessing by collaboration with CRIEPI. The test using U began in 2002, and the test using PuO$$_{2}$$ and unirradiated MOX were ended in 2008. The reduction of UO$$_{2}$$ pellets by using Li-reduction method, the electrowinning using reduced pellets, the separation of adhered salt with deposit by distillation, and the ingot formation of deposit were performed. As a result, 99% of the loaded U is recovered as metal ingot. The tests similar to U tests were performed by using PuO$$_{2}$$. As a result, Pu was successfully recovered with U metal. In the MOX test, the mass balance of Pu was maintained at $$sim$$100% with respect to the initial amount. We try to form the U-Pu-Zr alloy by using reduced MOX. After 2009, the process development that uses the alloy will be continued.

Journal Articles

Ingot formation using uranium dendrites recovered by electrolysis in LiCl-KCl-PuCl$$_{3}$$-UCl$$_{3}$$ melt

Fukushima, Mineo; Nakayoshi, Akira; Kitawaki, Shinichi; Kurata, Masaki*; Yahagi, Noboru*

Proceedings of 3rd International ATALANTE Conference (ATALANTE 2008) (CD-ROM), 4 Pages, 2008/05

Products on solid cathodes recovered from metal pyrochemical processing were processed to obtain uranium ingot. Studies on process conditions for uranium forming, assay recovered uranium products and by-products and evaluation of mass balance were carried out. In this tests, it is confirmed that uranium ingots can be obtained with heating the products up to more than melting temperature of metal uranium under normal pressure because adhered salt cover the uranium not to oxidize it during uranium cohering. Volatilization of americium is very small under the condition of high temperature.

Journal Articles

Basic knowledge on treating various wastes generated from practical operation of metal pyro-reprocessing

Nakayoshi, Akira; Kitawaki, Shinichi; Fukushima, Mineo; Kurata, Masaki*; Yahagi, Noboru*

Proceedings of International Symposium on EcoTopia Science 2007 (ISETS '07) (CD-ROM), p.1062 - 1066, 2007/11

Pyro-reprocessing is one of the promising reprocessing methods for recycling spent nuclear fuels generated from fast reactors. Comparing to the conventional aqueous-processes, following benefits are expected when introducing the pyro-reprocessing, such as reduction of environmental burden, enhancement of proliferation-resistant, enhancement of economical potential, efficient utilization of nuclear resources. The pyro-reprocessing will therefore become more attractive not only in developed countries regarding nuclear energy, but also in developing countries. As for reducing environmental burden, the most important subject is establishment of the nuclear fuel cycle, in which actinide elements are closed. Various kinds of intermediate waste which contains actinide elements are formed in the practical operation not only from the main steps of the pyro-reprocessing but also from related sub-streams.

Oral presentation

Development of pyrometallurgical reprocessing; Mass balance of electrorefining process

Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Yahagi, Noboru*; Kurata, Masaki*

no journal, , 

no abstracts in English

Oral presentation

Development of metal pyro-processing, 2; Sequential process test for electro-chemical reduction of MOX pellet

Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo; Kurata, Masaki*; Yahagi, Noboru*

no journal, , 

no abstracts in English

Oral presentation

Sequential process tests for metal pyro-reprocessing

Kurata, Masaki*; Yahagi, Noboru*; Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo

no journal, , 

To demonstrate the real process of metal pyro-reprocessing of nuclear spent fuels, sequential tests of major step were performed, such as electro-chemical reduction of MOX pellets and electro-chemical recovery of U metal and U-Pu alloy, in 1/1000 scale w.r.t. the real process. Variation in various operation parameters, product composition, and actinide distribution were investigated in the sequential tests.

Oral presentation

Development of pyro-metallurgical reprocessing, 3; Sequential test for electro-refining using U-Pu alloy

Kurata, Masaki*; Yahagi, Noboru*; Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo

no journal, , 

Sequential test for electro-refining using U-Pu alloy that was obtained by reduction of MOX pellets was performed. The test was consisted of four electrolysis using three solid cathodes and a liquid cadmium cathode. Test results shows that (1) material balance of nuclear materials were satisfactorily closed in whole process, (2) concentration of Pu and Am in uranium products were less than 1 ppm respectively, (3) U metal reacted with oxygen impurities in media as an oxygen getter.

Oral presentation

Sequential process test for metal pyro-processing using U, Pu, and Am

Kurata, Masaki*; Yahagi, Noboru*; Kitawaki, Shinichi; Nakayoshi, Akira; Fukushima, Mineo

no journal, , 

CRIEPI and JAEA are continuing a collaboration study for metal pyro-processing, in which sequential process tests have been performed under practical conditions using 1kg of molten salt baths. Recent results on two kinds of sequential process test are reported. The former is the electro-chemical reduction test of MOX pellet in order to form U-Pu alloy ingots. The latter is the electrolysis test using a combination of U-Pu alloy anode and solid or liquid cadmium cathode to recover U-product or U-Pu-Am alloy product, respectively. Variation in electrode potentials, current efficiency, molten salt composition, Pu- or Am-impurity in the U-product, U/Pu or Am/Pu separation factor in the U-Pu alloy product, and etc. were measured. These tests reproduced practical operating conditions at a scale of 1/1000 that of the actual process.

59 (Records 1-20 displayed on this page)