Interim activity status report of "the group for investigation of reasonable safety assurance based on graded approach" (from September, 2019 to September, 2020)

Yonomoto, Taisuke; Nakashima, Hiroshi*; Sono, Hiroki; Kishimoto, Katsumi; Izawa, Kazuhiko; Kinase, Masami; Osa, Akihiko; Ogawa, Kazuhiko; Horiguchi, Hironori; Inoi, Hiroyuki; et al.

JAEA-Review 2020-056, 51 Pages, 2021/03

JAEA-Review-2020-056.pdf:3.26MB

A group named as "The group for investigation of reasonable safety assurance based on graded approach", which consists of about 10 staffs from Sector of Nuclear Science Research, Safety and Nuclear Security Administration Department, departments for management of nuclear facility, Sector of Nuclear Safety Research and Emergency Preparedness, aims to realize effective graded approach (GA) about management of facilities and regulatory compliance of JAEA. The group started its activities in September, 2019 and has had discussions through 10 meetings and email communications. In the meetings, basic ideas of GA, status of compliance with new regulatory standards at each facility, new inspection system, etc were discussed, while individual investigation at each facility were shared among the members. This report is compiled with expectation that it will help promote rational and effective safety management based on GA by sharing contents of the activity widely inside and outside JAEA.

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.

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.

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

http://www-pub.iaea.org/mtcd/meetings/PDFplus/2009/cn176/cn176_BoeS.pdf

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.

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

https://doi.org/10.15669/pnst.2.143

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 628C 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 650C.

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

https://doi.org/10.3327/jnst.47.1118

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.

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

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

Nihon 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.

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

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

https://doi.org/10.3327/taesj.J09.011

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.

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.

Numerical simulation on thermal-hydraulic characteristics in fuel assemblies of supercritical water cooled reactors using two-fluid model analysis code ACE-3D

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

Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1690 - 1693, 2009/09

In order to assess the thermal hydraulics in fuel assemblies of the supercritical water cooled reactor (SCWR) core, JAEA has been enhancing the three-dimensional two-fluid model analysis code ACE-3D to predict thermal-hydraulic behavior of the SCWR. As a part of these assessments, the present paper describes numerical analysis results on thermal-hydraulic characteristics in the fuel assembly based on the design of the supercritical water cooled fast reactor (Super Fast Reactor) concept. 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), adjacent to the channel box; (2), next to type (1); and (3), located inside types (1) and (2). The results show the influence of existence of the channel box and variation of the rod surface temperature profiles in the circumferential direction.