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Journal Articles

Comprehensive analysis and evaluation of Fukushima Daiichi Nuclear Power Station Unit 3

Yamashita, Takuya; Honda, Takeshi*; Mizokami, Masato*; Nozaki, Kenichiro*; Suzuki, Hiroyuki*; Pellegrini, M.*; Sakai, Takeshi*; Sato, Ikken; Mizokami, Shinya*

Nuclear Technology, 26 Pages, 2023/00

Journal Articles

Design and actual performance of J-PARC 3 GeV rapid cycling synchrotron for high-intensity operation

Yamamoto, Kazami; Kinsho, Michikazu; Hayashi, Naoki; Saha, P. K.; Tamura, Fumihiko; Yamamoto, Masanobu; Tani, Norio; Takayanagi, Tomohiro; Kamiya, Junichiro; Shobuda, Yoshihiro; et al.

Journal of Nuclear Science and Technology, 59(9), p.1174 - 1205, 2022/09

 Times Cited Count:1 Percentile:71.47(Nuclear Science & Technology)

In the Japan Proton Accelerator Research Complex, the purpose of the 3 GeV rapid cycling synchrotron (RCS) is to accelerate a 1 MW, high-intensity proton beam. To achieve beam operation at a repetition rate of 25 Hz at high intensities, the RCS was elaborately designed. After starting the RCS operation, we carefully verified the validity of its design and made certain improvements to establish a reliable operation at higher power as possible. Consequently, we demonstrated beam operation at a high power, namely, 1 MW. We then summarized the design, actual performance, and improvements of the RCS to achieve a 1 MW beam.

Journal Articles

Quasifree neutron knockout reaction reveals a small $$s$$-Orbital component in the Borromean nucleus $$^{17}$$B

Yang, Z. H.*; Kubota, Yuki*; Corsi, A.*; Yoshida, Kazuki; Sun, X.-X.*; Li, J. G.*; Kimura, Masaaki*; Michel, N.*; Ogata, Kazuyuki*; Yuan, C. X.*; et al.

Physical Review Letters, 126(8), p.082501_1 - 082501_8, 2021/02


 Times Cited Count:25 Percentile:96.89(Physics, Multidisciplinary)

A quasifree ($$p$$,$$pn$$) experiment was performed to study the structure of the Borromean nucleus $$^{17}$$B, which had long been considered to have a neutron halo. By analyzing the momentum distributions and exclusive cross sections, we obtained the spectroscopic factors for $$1s_{1/2}$$ and $$0d_{5/2}$$ orbitals, and a surprisingly small percentage of 9(2)% was determined for $$1s_{1/2}$$. Our finding of such a small $$1s_{1/2}$$ component and the halo features reported in prior experiments can be explained by the deformed relativistic Hartree-Bogoliubov theory in continuum, revealing a definite but not dominant neutron halo in $$^{17}$$B. The present work gives the smallest $$s$$- or $$p$$-orbital component among known nuclei exhibiting halo features and implies that the dominant occupation of $$s$$ or $$p$$ orbitals is not a prerequisite for the occurrence of a neutron halo.

Journal Articles

Comprehensive analysis and evaluation of Fukushima Daiichi Nuclear Power Station Unit 2

Yamashita, Takuya; Sato, Ikken; Honda, Takeshi*; Nozaki, Kenichiro*; Suzuki, Hiroyuki*; Pellegrini, M.*; Sakai, Takeshi*; Mizokami, Shinya*

Nuclear Technology, 206(10), p.1517 - 1537, 2020/10

 Times Cited Count:10 Percentile:86.11(Nuclear Science & Technology)

JAEA Reports

Waste liquid treatment for uranium liquid waste containing impurities

Sato, Yoshiyuki; Aono, Ryuji; Haraga, Tomoko; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Testing 2019-003, 20 Pages, 2019/12


In the Radioactive Waste Management Technology Section, the radioactive liquid waste generated in the test using natural uranium in the past has been stored based on the contents of permission. Although we decided to perform solidification treatment in order to reduce the risk in storage, no rational treatment method has been established so far. Therefore, we examined adsorption treatment of natural uranium using uranium adsorbent (Tannix), and finally stabilized treatment by cement solidification. The treatment methods and findings obtained for a series of operations in waste liquid treatment are summarized in this report for reference when treating similar liquid waste.

Journal Articles

Oxidation of silicon carbide in steam studied by laser heating

Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Furumoto, Kenichiro*; Sato, Hisaki*; Ishibashi, Ryo*; Yamashita, Shinichiro

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.670 - 674, 2019/09

Journal Articles

Safe and rapid development of capillary electrophoresis for ultratrace uranyl ions in radioactive samples by way of fluorescent probe selection for actinide ions from a chemical library

Haraga, Tomoko; Ouchi, Kazuki; Sato, Yoshiyuki; Hoshino, Hitoshi*; Tanana, Rei*; Fujihara, Takashi*; Kurokawa, Hideki*; Shibukawa, Masami*; Ishimori, Kenichiro; Kameo, Yutaka; et al.

Analytica Chimica Acta, 1032, p.188 - 196, 2018/11

 Times Cited Count:11 Percentile:49.7(Chemistry, Analytical)

The development of safe, rapid and highly sensitive analytical methods for radioactive samples, especially actinide (An) ions, represents an important challenge. Here we propose a methodology for selecting appropriate emissive probes for An ions with very low consumption and emission of radioactivity by capillary electrophoresis-laser-induced fluorescence detection (CE-LIF), using a small chemical library of probes with eight different chelating moieties. It was found that the emissive probe, which possesses the tetradentate chelating moiety, was suitable for detecting uranyl ions. The detection limit for the uranyl-probe complex using CE-LIF combined with dynamic ternary complexation and on-capillary concentration techniques was determined to be 0.7 ppt. This method was successfully applied to real radioactive liquid samples collected from nuclear facilities.

Journal Articles

Network system operation for J-PARC accelerators

Kamikubota, Norihiko*; Yamada, Shuei*; Sato, Kenichiro*; Kikuzawa, Nobuhiro; Yamamoto, Noboru*; Yoshida, Susumu*; Nemoto, Hiroyuki*

Proceedings of 16th International Conference on Accelerator and Large Experimental Physics Control Systems (ICALEPCS 2017) (Internet), p.1470 - 1473, 2018/01

no abstracts in English

JAEA Reports

Development of $$^{93}$$Zr, $$^{93}$$Mo, $$^{107}$$Pd and $$^{126}$$Sn analytical methods for radioactive waste from Fukushima Daiichi Nuclear Power Station

Aono, Ryuji; Sato, Yoshiyuki; Shimada, Asako; Tanaka, Kiwamu; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka

JAEA-Technology 2017-025, 32 Pages, 2017/11


We have developed analytical methods for $$^{93}$$Zr, $$^{93}$$Mo, $$^{107}$$Pd and $$^{126}$$Sn, which are considered important in terms of the safety assessment of radioactive waste disposal. The methods are specialized for the wastes left after Fukushima accident. As the main analytical sample, we assumed accumulated water / treated water collected at Fukushima Daiichi Nuclear Power Station. As for $$^{93}$$Zr, $$^{93}$$Mo, $$^{107}$$Pd and $$^{126}$$Sn contained in this accumulated water / treated water, we have worked on the development of separation and purification method of target nuclide and improvement of recovery, and summarized these results in this report.

Journal Articles

Study on the seismic buckling evaluation method for Mod. 9Cr-1Mo steel cylindrical vessel

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Kubo, Koji*; Sato, Kenichiro*; Wakai, Takashi; Shimomura, Kenta

Nihon Kikai Gakkai M&M 2017 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.591 - 595, 2017/10

no abstracts in English

Journal Articles

Radiochemical analysis of rubble collected from around and inside reactor buildings at Units 1 to 4 in Fukushima Daiichi Nuclear Power Station

Sato, Yoshiyuki; Aono, Ryuji; Konda, Miki; Tanaka, Kiwamu; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 13 Pages, 2017/00

no abstracts in English

Journal Articles

Radiochemical analysis of rubble collected from Fukushima Daiichi Nuclear Power Station

Sato, Yoshiyuki; Tanaka, Kiwamu; Ueno, Takashi; Ishimori, Kenichiro; Kameo, Yutaka

Hoken Butsuri, 51(4), p.209 - 217, 2016/12

A large amount of contaminated rubbles were generated by the accident at the Fukushima Daiichi Nuclear Power Station (F1NPS). For safe decommissioning of F1NPS, it is important to evaluate the composition and concentration of radionuclides in the rubbles. In this paper, to characterize the rubbles collected at F1NPS in Unit-1, Unit-2 and Unit-3, radiochemical analysis was operated. As a result of radiochemical analysis, $$gamma$$-ray-emitting nuclides $$^{60}$$Co, $$^{137}$$Cs and $$^{152,154}$$Eu, $$beta$$-ray-emitting nuclides $$^{3}$$H, $$^{14}$$C, $$^{90}$$Sr and $$^{99}$$Tc, and $$alpha$$-particle-emitting nuclides $$^{238,239+240}$$Pu, $$^{241}$$Am and $$^{244}$$Cm were detected. In contrast, $$^{94}$$Nb and $$^{152}$$Eu concentrations were below the detection limit. Measured radioactive concentrations implied that $$^{3}$$H, $$^{14}$$C, $$^{60}$$Co and $$^{90}$$Sr concentrations depended on $$^{137}$$Cs concentration respectively. This analysis was characterized the radioactivity concentrations of the rubbles.

Journal Articles

The Welded joint strength reduction factors of modified 9Cr-1Mo Steel for the advanced loop-type sodium cooled fast reactor

Yamashita, Takuya; Wakai, Takashi; Onizawa, Takashi; Sato, Kenichiro*; Yamamoto, Kenji*

Journal of Pressure Vessel Technology, 138(6), p.061407_1 - 061407_6, 2016/12

 Times Cited Count:0 Percentile:0(Engineering, Mechanical)

Journal Articles

Strength of 316FR joints welded by Type 316FR/16-8-2 filler metals

Yamashita, Takuya; Nagae, Yuji; Sato, Kenichiro*; Yamamoto, Kenji*

Journal of Pressure Vessel Technology, 138(2), p.024501_1 - 024501_7, 2016/04

 Times Cited Count:0 Percentile:0(Engineering, Mechanical)

Journal Articles

A Study on evaluation method of penetrate crack length for LBB assessment of fast reactor pipes

Wakai, Takashi; Machida, Hideo*; Sato, Kenichiro*

Nihon Kikai Gakkai M&M 2015 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), 3 Pages, 2015/11

This paper describes a through-wall crack length evaluation procedure applicable to Leak Before Break (LBB) assessment of Japan Sodium cooled Fast Reactor (JSFR) pipes made of Mod.9Cr-1Mo steel. In LBB assessment of JSFR pipes, it is required to calculate virtual through-wall crack length, though the crack growth is quite small under design condition. Generally, it is known that the through-wall crack length depends on loading condition, namely the load ratio between tensile and bending and that the length under pure bending load condition is largest. This study proposes a simplified method to evaluate the through-wall crack length both for axial and circumferential cracks as a function of load ratio and fatigue crack growth characteristics. Using the method, through-wall crack length can be predicted as far as we know the loading condition and material properties.

Journal Articles

J-integral evaluation method for a through wall crack in thin-walled large diameter pipes made of Mod.9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Sato, Kenichiro*

Nihon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09

This paper describes a J-integral evaluation procedure applicable to unstable failure analysis for a circumferential through wall crack in a pipe. Japan Sodium cooled Fast Reactor (JSFR) pipes are made of Mod.9Cr-1Mo steel. The fracture toughness of the material is inferior to that of conventional austenitic stainless steels. In addition, JSFR pipe has small thickness and large diameter and displacement controlled load is predominant. Therefore, the load balance in such piping system changes by crack extension and 2 parameter method using J-integral is applicable to unstable failure analysis for the pipes under such conditions. As a J-integral evaluation method for circumferential through wall crack in a cylinder, EPRI has proposed a fully plastic solution method. However, the geometry of JSFR pipe and material characteristics of Mod.9Cr-1Mo steel exceed the applicable range of EPRI's method. Therefore, a series of elastic, elastoplastic and plastic finite element analyses (FEA) were performed for a pipe with a circumferential through-wall crack to establish a J-integral evaluation method applicable to such conditions. J-integrals obtained from the FEA were resolved into elastic, local plastic and fully plastic components. Each component was expressed as a function of analytical parameter, such as pipe geometries, crack size, material characteristics and so on. As a result, a simplified J-integral evaluation method was proposed.

Journal Articles

Requirements for fracture toughness to satisfy LBB behavior of a pipe made of high chromium steel

Machida, Hideo*; Wakai, Takashi; Sato, Kenichiro*

Nihon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09

The volumetric test for piping in a sodium cooled fast reactor (SFR) is difficult from the poor accessibility. Detection of a crack, therefore, is difficult before its penetration of a pipe wall, an SFR has a strategy to detect sodium leakage from a through wall crack before fracture of a pipe. Plant safety is ensured by shutting down a plant as soon as possible to detect small quantity of sodium leakage even if a crack penetrates a pipe wall. Consequently, it is important to ensure establishment of leakage-before-break (LBB) in this strategy. Effects of fracture resistance curve on fracture strength of a cracked pipe made of high chromium steel (Mod. 9Cr-1Mo steel), which is one of the candidates for fast reactor piping material, are evaluated in this study; and requirements for fracture resistance curve to achieve the LBB were proposed.

Journal Articles

Creep-fatigue tests of double-end notched bar made of Mod.9Cr-1Mo steel

Shimomura, Kenta; Kato, Shoichi; Wakai, Takashi; Ando, Masanori; Hirose, Yuichi*; Sato, Kenichiro*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

This paper describes experimental and analytical works to confirm that the design standard for SFR components sufficiently covers possible failure mechanisms. Creep-fatigue damage evaluation method in JSME design standard for SFR components has been constructed based on experiments and/or numerical analyses of conventional austenitic stainless steels, such as 304SS. Since the material characteristics of Mod.9Cr-1Mo steel are substantially different from those of austenitic stainless steels, it is required to verify the applicability of the design standards to the SFR components made of Mod.9Cr-1Mo steel. A series of uni-axial creep-fatigue tests were conducted using double-ended notch bar specimens made of Mod.9Cr-1Mo steel under displacement controlled condition with 30 minute holding. The curvature radii of the specimens were 1.6mm, 11.2mm and 40.0mm. The specimen having 1.6mm notch and 11.2mm notch failed from outer surface but the specimen having 40.0mm notch showed obvious internal crack nucleation. In addition, though total duration time of the creep-fatigue test was only 2,000 hours, a lot of creep voids and inter granular crack growth were observed. To clarify the cause of such peculiar failure, some additional experiments were performed, as well as some numerical analyses. We could point out that such a peculiar failure aspect might result from corresponding stress distribution in the cross section. As a result of a series of investigations, possible causes of such peculiar failure could be narrowed down. A future investigation plan was proposed to clarify the most significant cause.

JAEA Reports

The Periodic safety review report of Tokai Reprocessing Plant

Fukuda, Kazuhito; Tomioka, Kenichiro*; Omori, Satoru; Nakano, Takafumi; Nagasato, Yoshihiko

JAEA-Technology 2014-032, 566 Pages, 2014/11


The Periodic Safety Review of TRP is assessment of the validity of safety activities in order to get assurance for continuous operation by adding effective items to extract and to execute for TRP safety. We performed 4 items; for (1) evaluation of safety activity at TRP, as we confirmed organization was ordered and managed. For (2) evaluation of status of safety activities reflecting the latest knowledge, we confirmed improvement of safety was continued adequately reflecting from the experience for safety. For (3) technical review on aging for the safety related structures, systems and components, we evaluated the guaranty of safety under assumption of continuous maintenance till the next Periodic Safety Review. For (4) establishment of long term maintenance program, we found no additional activities into maintenance programs, however, for several installations we established a plan and utilized them for reliability.

Journal Articles

Application of capillary electrophoresis with laser-induced fluorescence detection for the determination of trace neodymium in spent nuclear fuel using complexation with an emissive macrocyclic polyaminocarboxylate probe

Haraga, Tomoko; Saito, Shingo*; Sato, Yoshiyuki; Asai, Shiho; Hanzawa, Yukiko; Hoshino, Hitoshi*; Shibukawa, Masami*; Ishimori, Kenichiro; Takahashi, Kuniaki

Analytical Sciences, 30(7), p.773 - 776, 2014/07

 Times Cited Count:6 Percentile:23.41(Chemistry, Analytical)

A simple and rapid method with low radiation exposure risk was developed for the determination of neodymium in spent nuclear fuel by CE with LIF detection using a fluorescent ligand having a macrocyclic hexadentate polyaminocarboxylate structure. The concentration of Nd(III) in a spent nuclear fuel sample was determined with no interference from various matrix elements, including lanthanides and uranium (at a 200-fold excess), with 92 $$pm$$ 3% recovery. This is due to method's high resolution based on establishing a ternary complex equilibrium during migration in which the hydroxyl ion plays an auxiliary role.

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