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Journal Articles

Outline of the OECD/NEA/ARC-F Project

Nakatsuka, Toru; Maeda, Toshikatsu; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1650 - 1656, 2019/08

The OECD/NEA is launching a new project named "Analysis of Information from Reactor Buildings and Containment Vessels of Fukushima Daiichi Nuclear Power Station (ARC-F)" Project. This project will serve as the successor to the precedent NEA project, "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Phase II" which investigated the accident scenarios, associated fission products behavior in the damaged units and source term to the environment. The ARC-F project comprises three tasks: Task 1: Refinement of analysis for accident scenarios and associated fission product transportation and dispersion; Task 2: Compilation and management of data and information; and Task 3: Discussion for future long-term project. Japan Atomic Energy Agency is the operating agent, responsible to lead all the tasks. Duration of the project is from January 2019 to December 2021 and the final report is planned to be published in 2022.

Journal Articles

Shell evolution beyond $$Z$$=28 and $$N$$=50; Spectroscopy of $$^{81,82,83,84}$$Zn

Shand, C. M.*; Podoly$'a$k, Zs.*; G$'o$rska, M.*; Doornenbal, P.*; Obertelli, A.*; Nowacki, F.*; Otsuka, T.*; Sieja, K.*; Tostevin, J. A.*; Tsunoda, T.*; et al.

Physics Letters B, 773, p.492 - 497, 2017/10

 Times Cited Count:8 Percentile:14.67(Astronomy & Astrophysics)

Journal Articles

Shape evolution in neutron-rich krypton isotopes beyond N=60; First spectroscopy of $$^{98,100}$$Kr

Flavigny, F.*; Doornenbal, P.*; Obertelli, A.*; Delaroche, J.-P.*; Girod, M.*; Libert, J.*; Rodriguez, T. R.*; Authelet, G.*; Baba, H.*; Calvet, D.*; et al.

Physical Review Letters, 118(24), p.242501_1 - 242501_6, 2017/06

 Times Cited Count:14 Percentile:9.07(Physics, Multidisciplinary)

Journal Articles

$$gamma$$ decay of unbound neutron-hole states in $$^{133}$$Sn

Vaquero, V.*; Jungclaus, A.*; Doornenbal, P.*; Wimmer, K.*; Gargano, A.*; Tostevin, J. A.*; Chen, S.*; N$'a$cher, E.*; Sahin, E.*; Shiga, Yoshiaki*; et al.

Physical Review Letters, 118(20), p.202502_1 - 202502_5, 2017/05

 Times Cited Count:8 Percentile:18.97(Physics, Multidisciplinary)

Journal Articles

Low-lying structure and shape evolution in neutron-rich Se isotopes

Chen, S.*; Doornenbal, P.*; Obertelli, A.*; Rodriguez, T. R.*; Authelet, G.*; Baba, H.*; Calvet, D.*; Ch$^a$teau, F.*; Corsi, A.*; Delbart, A.*; et al.

Physical Review C, 95(4), p.041302_1 - 041302_6, 2017/04

 Times Cited Count:11 Percentile:8.11(Physics, Nuclear)

Journal Articles

Are there signatures of harmonic oscillator shells far from stability?; First spectroscopy of $$^{110}$$Zr

Paul, N.*; Corsi, A.*; Obertelli, A.*; Doornenbal, P.*; Authelet, G.*; Baba, H.*; Bally, B.*; Bender, M.*; Calvet, D.*; Ch$^a$teau, F.*; et al.

Physical Review Letters, 118(3), p.032501_1 - 032501_7, 2017/01

 Times Cited Count:14 Percentile:9.07(Physics, Multidisciplinary)

JAEA Reports

Internship using nuclear facilities in Oarai Research and Development Center

Takemoto, Noriyuki; Itagaki, Wataru; Kimura, Nobuaki; Ishitsuka, Etsuo; Nakatsuka, Toru; Hori, Naohiko; Ooka, Makoto; Ito, Haruhiko

JAEA-Review 2013-063, 34 Pages, 2014/03

JAEA-Review-2013-063.pdf:8.46MB

Nuclear energy is important from a viewpoint of economy and energy security in Japan. However, the lack of nuclear engineers and scientists in future is concerned after the sever accident of TEPCO's Fukushima Daiichi Nuclear Power Station has occurred. Institute of National Colleges of Technology planned to carry out training programs for human resource development of nuclear energy field including on-site training in nuclear facilities. Oarai Research and Development Center in Japan Atomic Energy Agency cooperatively carried out an internship for nuclear disaster prevention and safety utilizing the nuclear facilities such as the JMTR. Thirty two students joined in total in the internship from FY 2011 to FY2013. In this paper, contents and results of the internship are reported.

JAEA Reports

2013 training using JMTR and related facilities as advanced research infrastructures

Takemoto, Noriyuki; Kimura, Nobuaki; Hanakawa, Hiroki; Shibata, Akira; Matsui, Yoshinori; Nakamura, Jinichi; Ishitsuka, Etsuo; Nakatsuka, Toru; Ito, Haruhiko

JAEA-Review 2013-058, 42 Pages, 2014/02

JAEA-Review-2013-058.pdf:4.95MB

Practical training courses using the JMTR and related facilities as an advanced research infrastructures have been carried out in Japan Atomic Energy Agency since FY2010 from a viewpoint of the nuclear human resource development and the securing. In FY2013, "Training course for foreign young researchers and engineers" was carried out from July 8th to July 26th, and "Training course using JMTR and related facilities as advanced research infrastructures" for domestic young researchers and engineers was carried out from July 29th to August 9th. 18 young researchers and engineers were joined in each training course, and 36 trainees in total studied about basic nuclear research and technology through the lecture and training about the reactor operation management, safety management, irradiation test, etc. in the JMTR. The results of these courses are reported in this paper.

JAEA Reports

2012 training using JMTR and related facilities as advanced research infrastructures

Kimura, Nobuaki; Takemoto, Noriyuki; Ooka, Makoto; Ishitsuka, Etsuo; Nakatsuka, Toru; Ito, Haruhiko; Ishihara, Masahiro

JAEA-Review 2012-055, 40 Pages, 2013/03

JAEA-Review-2012-055.pdf:93.64MB

Training courses using JMTR and related facilities as advanced research infrastructures have been newly organized for domestic students, young researchers and engineers since FY2010 from a viewpoint of nuclear human resource development in order to support global expansion of nuclear power industry. In FY 2012, two courses were carried for foreign as well as Japanese young researchers and engineers in order to carry out effective practical training. For the foreigner course, 16 young researchers and engineers were joined from July 23rd to August 10th. For the Japanese course, total 35 young researchers and engineers were joined two courses from August 20th to August 31st and from September 3rd to September 14th. Lectures of these training courses were consisted from basics of nuclear energy to its application, especially for irradiation tests in Motrin this paper, results of these foreigners and Japanese training courses are reported.

Journal Articles

Thermal-hydraulic calculation for simplified fuel assembly of super fast reactor using two-fluid model analysis code ACE-3D

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 5 Pages, 2012/00

In the present paper, thermal-hydraulic behavior in a simplified fuel assembly of the supercritical water cooled fast reactor (Super Fast Reactor) was analyzed with the three-dimensional two-fluid model analysis code ACE-3D. The analytical geometry simulates a 19-rod assembly, which is one of the most simplified geometry of the SCWR fuel assembly and includes three kinds of different subchannel types; (1) adjoining to the channel box, (2): next to type (1), and (3): located inside types (1) and (2). It was confirmed that the MCST satisfies a thermal design criteria to ensure fuel and cladding integrity.

Journal Articles

Numerical analysis on thermal-hydraulics of supercritical water flowing in a tight-lattice fuel bundle

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Progress in Nuclear Science and Technology (Internet), 2, p.143 - 146, 2011/10

To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of supercritical water reactor (Super Fast Reactor), a simplified 19-rod fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D. In this calculation, a one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power distribution are to be the same as the steady state condition of the reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Effect of grid spacers is taken into account in the analysis. Maximum cladding surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 628$$^{circ}$$C that is almost the same as results without grid spacers. The predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650$$^{circ}$$C.

Journal Articles

Super fast reactor R&D projects in Japan, 4; Numerical estimation of thermal-hydraulic characteristics of supercritical fluids in tight-lattice bundles by three-dimensional two-fluid model analysis code ACE-3D

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 5 Pages, 2011/05

To analyze thermal hydraulics in the core of supercritical-water-cooled reactors, JAEA has been improved a three-dimensional two-fluid model analysis code ACE-3D, which has been developed originally for two-phase flow of LWRs. Heat transfer experiments of supercritical fluids flowing in a tube, a vertical annular channel around a heater pin and 7-rod bundles were analyzed with the improved ACE-3D to assess the prediction performance of the code at supercritical region. As a result, it was confirmed that the calculated wall surface temperatures agreed with the measured results. To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of Super Fast Reactor, a simplified 19-rod fuel assembly was analyzed. Maximum clad surface temperature was observed at the position facing to the narrowest gap on the center rod near the outlet and the value was 901K. The predicted MCST satisfies thermal design criteria to ensure fuel and cladding integrity.

Journal Articles

Outline of research and development of thermal-hydraulics and safety of Japanese Supercritical Water Cooled Reactor (JSCWR) project

Nakatsuka, Toru; Mori, Hideo*; Akiba, Miyuki*; Ezato, Koichiro; Yasuoka, Makoto*

Proceedings of 5th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-5) (CD-ROM), 12 Pages, 2011/03

In the thermal-hydraulic area of Japanese Supercritical Water Cooled Reactor (JSCWR) project, the main objective is to provide high-precision heat transfer and hydraulics resistance correlations of supercritical water which are necessary for the conceptual design of the core and fuel. For this purpose, a database was constructed from literature survey and previous research results. The most suitable correlation applied for circular tubes was selected based on the database and the range of application and predictive accuracy were defined. A thermal-hydraulics analysis code has been developed based on large eddy simulation, which is selected for simulation of the heat transfer deterioration, to give detailed information of thermal-hydraulics phenomena in a fuel bundle.

Journal Articles

Assessment of applicability of two-fluid model code ACE-3D to heat transfer test of supercritical water flowing in an annular channel

Nakatsuka, Toru; Ezato, Koichiro; Misawa, Takeharu; Seki, Yohji; Yoshida, Hiroyuki; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Takase, Kazuyuki

Journal of Nuclear Science and Technology, 47(12), p.1118 - 1123, 2010/12

 Times Cited Count:1 Percentile:87.11(Nuclear Science & Technology)

In order to perform efficiently the thermal design of the supercritical water reactor (SCWR), it is important to assess the thermal hydraulics in rod bundles of the core. Japan Atomic Energy Agency (JAEA) has been improved the three-dimensional two-fluid model analysis code ACE-3D, which has been developed originally for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water at supercritical region. In the present paper, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which was performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code may be applicable to prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of SCWR core.

Journal Articles

Numerical analysis on thermal-hydraulics of supercritical water flowing in a tight-lattice fuel bundle

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 4 Pages, 2010/10

To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of supercritical water reactor (Super Fast Reactor), a simplified 19-rod fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D. In this calculation, a one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power distribution are to be the same as the steady state condition of the reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Effect of grid spacers is taken into account in the analysis. Maximum clad surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 628$$^{circ}$$C that is almost the same as results without grid spacers. The predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650$$^{circ}$$C.

Journal Articles

Development of prediction method of turbulent heat transfer for thermal-hydraulic design of supercritical water reactor

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Nippon Konsoryu Gakkai Nenkai Koenkai 2010 Koen Rombunshu, p.348 - 349, 2010/07

In the thermal hydraulic design of supercritical water-cooled reactor, it is required to establish a thermal hydraulic design method which can precisely predict heat transfer deterioration of supercritical water as the core coolant. Assessments of applicability of turbulence models used in design methods have not been sufficiently performed, since the mechanism of heat transfer deterioration has not been clearly figured out yet. Japan Atomic Energy Agency has started developing prediction method of heat transfer deterioration with large eddy simulation to improve the thermal hydraulic design accuracy. In the present study, simulation results of heat transfer test with Freon are reported.

Journal Articles

Development of evaluation method of thermal-hydraulic stability of once-through steam generator by enhanced TRAC-BF1

Nakatsuka, Toru; Liu, W.; Yoshida, Hiroyuki; Takase, Kazuyuki

Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.279 - 280, 2010/06

To assess the stability of once-through steam generators in FBR, Japan Atomic Energy Agency has been developing a prediction method for thermal-hydraulic instability based on system analysis code TRAC-BF1. In the present paper, to simulate the primary coolant in steam generators, thermal property of sodium was incorporated to the code and the VESSEL component was improved to handle two different fluids of primary sodium and secondary water. These added functions were assessed with a simplified steam generator model calculation by altering primary coolant fluid as water and sodium. It was confirmed that heat transfer at steam generators was properly evaluated for the case that primary coolant is sodium as well as water.

Journal Articles

Study on effect of local power distribution of fuel assembly on critical power of Reduced-Moderation Water Reactor (RMWR)

Nakatsuka, Toru; Nakano, Yoshihiro; Okubo, Tsutomu

Nippon Genshiryoku Gakkai Wabun Rombunshi, 9(2), p.139 - 149, 2010/06

The viability of fuel assembly designs of Reduced-Moderation Water Reactor (RMWR) with fewer kinds of plutonium enrichment of MOX fuel which may result in high local peaking factor in peripheral rods were assessed in the present report. Critical powers of 217-rod bundles with peripheral peaks for upper and lower MOX regions of double-flat core of the RMWR were calculated by a subchannel analysis code NASCA. Peripheral peaking with the corresponding local peaking factor for the uniform plutonium enrichment design yields almost the same critical power as for the flat power distribution. Reduction in fuel fabrication burden may be possible by decreasing the number of the kind of plutonium fuel enrichment while maintaining the same thermal-hydraulic margin as the fuel assembly design with five enrichment types of MOX fuels.

Journal Articles

Preliminary assessment of flow stability of once-through steam generator by TRAC-BF1

Nakatsuka, Toru; Liu, W.; Yoshida, Hiroyuki; Takase, Kazuyuki

Nippon Kikai Gakkai Netsu Kogaku Konfarensu 2009 Koen Rombunshu, p.269 - 270, 2009/11

To assess the stability of components which comprises parallel channels like steam generators in fast breeder reactors, Japan Atomic Energy Agency has been developing a prediction method for thermal-hydraulic instability based on system analysis code TRAC-BF1. In the present paper, TRAC-BF1 code was modified to simulate both primary and secondary coolant flow with VESSEL component. The calculation model was established and tested with preliminary simulations in which the primary sodium was replaced by high pressure water.

Journal Articles

Numerical simulations on turbulent heat transfer characteristics of supercritical pressure fluids

Nakatsuka, Toru; Takase, Kazuyuki; Yoshida, Hiroyuki; Misawa, Takeharu

Proceedings of 2009 ASME International Mechanical Engineering Congress & Exposition (IMECE 2009) (CD-ROM), 8 Pages, 2009/11

As one of next generation nuclear reactors, development of a supercritical pressure water reactor (SCWR) has been performed. In order to design the SCWR, it is necessary to investigate thermal-hydraulic characteristics in the SCWR core precisely. As for those characteristics, many experimental studies have been conducted from the former in each country using circular tubes, annular channels and the simulated fuel bundles. An objective of this study is to clarify the prediction accuracy of the turbulent heat transfer characteristics in the supercritical pressure fluids for the SCWR design. From the experimental results of the supercritical pressure fluids flowing upward in a vertical circular tube, it was confirmed that the turbulent heat transfer coefficient suddenly decreases under the high heat flux condition. Although many numerical studies have been done in order to confirm the deterioration of turbulent heat transfer in supercritical pressure fluids, it is important to choose a suitable turbulence model to obtain high prediction accuracy. Then, the prediction accuracy on the deteriorated turbulent heat transfer was investigated numerically using four kinds of turbulence models, SKE, MKE, RSM and LES. The predicted result of each turbulence model was compared with the experimental results of the concentric smooth annulus and vertical circular tube. From the results of the present study, it was found that LES is the most effective to simulate the deterioration of the turbulent heat transfer at the supercritical pressure fluids.

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