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Sun, Haomin; Kunugi, Tomoaki*; Yokomine, Takehiko*; Shen, X.*; Hibiki, Takashi*
Experimental Thermal and Fluid Science, 154, p.111171_1 - 111171_24, 2024/05
Oizumi, Akito; Akie, Hiroshi
JAEA-Technology 2023-017, 93 Pages, 2023/12
JAEA-Technology-2023-017.pdf:8.45MB
After the decision of decommissioning JMTR (Japan Materials Testing Reactor), Japan Atomic Energy Agency investigated the possibility to construct a new irradiation test reactor to succeed JMTR (post-JMTR), and the final report of the investigated result was submitted to the Ministry of Education, Culture, Sports, Science and Technology on March 30th 2021. This investigation was carried out in 4 steps of (1) selection of reactor type, (2) reactor core plans studies, (3) neutronic studies, (4) thermal studies, and was finally (5) considered and evaluated. This JAEA-Technology report summarizes the process and the results of (3) neutronic studies. Neutron fluxes were calculated at irradiation sample positions in the investigated cores, the standard core and the compact core, and the calculated fluxes satisfied the required irradiation capability. It was also evaluated the two investigated cores' continuous reactor operation time in days in one refueling cycle, and the results guaranteed an operation days equality with that of existing JMTR. In addition, neutronic characteristics of the cores were estimated, such as power distribution in the core, control rod reactivity worth, reactivity coefficients, distribution of fuel burnup rate of each fuel element, and kinetics parameters. The evaluated neutronic characteristics were used in the post-JMTR final investigation report to confirm the neutronic feasibility by comparing with the neutronic limiting values of existing JMTR, and to estimate the cooling capability to make the core thermally feasible.
Omer, M.; Shizuma, Toshiyuki*; Hajima, Ryoichi*; Koizumi, Mitsuo
Dai-43-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 3 Pages, 2022/11
Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
The new prediction method of critical heat flux (CHF) of the fuel assemblies based on the mechanism is proposed in this study. The prediction method of CHF based on the mechanism has been needed for a long time to enhance the safety analysis and reduce the design cost. From several experimental findings of the liquid-vapor behavior near the heating surface from the nucleate boiling to the CHF, the authors consider that the macrolayer dryout model will be appropriate to predict the CHF under the reactor condition. The prediction method of the macrolayer thickness and the passage period of vapor mass in the fuel assemblies are needed to predict CHF from the macrolayer dryout model. In this study, the CHF under the forced convection is evaluated by combining the prediction methods for the macrolayer thickness and passage period of vapor mass, which are proposed by authors. The prediction of the CHF under the forced convection is examined and compared with the experimental data.
Ho, H. Q.; Ishitsuka, Etsuo; Iigaki, Kazuhiko
Recent Contributions to Physics, 82(3), p.16 - 20, 2022/09
Shimo, Michito*; Niwa, Masakazu; Miyakawa, Kazuya; Yasue, Kenichi*; Tonokura, Kenichi*; Tokunaga, Tomochika*
Fukada Chishitsu Kenkyujo Nempo, (23), p.21 - 34, 2022/00
no abstracts in English
Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(12), p.820 - 824, 2021/12
The evaluation method of the critical heat flux based on the mechanism is needed for the efficient design and development of fuel in reactors and the appropriate safety evaluation. In this paper, the current researches relating to the mechanism of the critical heat flux are reviewed, and the issue to be considered in the future are discussed.
Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 178, p.121637_1 - 121637_24, 2021/10
Times Cited Count：7 Percentile：58.99(Thermodynamics)Nakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi
Journal of Nuclear Science and Technology, 58(10), p.1061 - 1070, 2021/10
Times Cited Count：5 Percentile：56.94(Nuclear Science & Technology)In a well-thermalized neutron field, it is principally possible to drive a thermal-neutron capture cross-section without considering an epithermal neutron component. This was demonstrated by a neutron activation method using the graphite thermal column (TC-Pn) of the Kyoto University Research Reactor. First, in order to confirm that the graphite thermal column was a well-thermalized neutron field, neutron irradiation was performed with neutron flux monitors: 
Au, 
Co, 
Sc, 
Cu, and 
Mo. The TC-Pn was confirmed to be extremely thermalized on the basis of Westcott's convention, because the thermal-neutron flux component took a constant value regardless of the sensitivity of each flux monitor to epithermal neutrons. Next, as a demonstration, the thermal-neutron capture cross section of 
Ta(n,
)
Ta reaction was measured using the graphite thermal column, and then derived to be 20.5
0.4 barn, which supported the evaluated value of 20.40.3 barn. The Ta nuclide could be useful as a flux monitor that complements the sensitivity between Au and Mo monitors.
Ono, Ayako; Sakashita, Hiroto*; Yoshida, Hiroyuki
Heat Transfer Engineering, 42(21), p.1775 - 1788, 2021/00
Times Cited Count：3 Percentile：23.27(Thermodynamics)In this study, the macrolayer formation model is proposed to predict the critical heat flux in the saturated and subcooled pool boiling based on the macrolayer dryout model. This model concept is based on the results of the previous experiments. In the model, the nucleation site is assumed to distribute based on the Poisson distribution. Combining the proposed macrolayer formation model and macrolayer dryout model, the CHFs up to subcooling 40K were predicted and they are successfully good agreement with the experimental data. Moreover, the concept of the model was confirmed by the numerical simulation using the TPFIT.
Misono, Toshiharu; Tsuruta, Tadahiko; Nakanishi, Takahiro; Sanada, Yukihisa; Shiribiki, Takehiko; Miyamoto, Kenji*; Urabe, Yoshimi*
JAEA-Research 2020-008, 166 Pages, 2020/10
JAEA-Research-2020-008.pdf:13.11MBJAEA-Research-2020-008(errata).pdf:0.92MB
After the accident at TEPCO Fukushima Daiichi Nuclear Power Station (1F), marine monitoring survey on radioactive substances have been conducted with financially supported by the Nuclear Regulatory Agency from FY2019. Results obtain in the project in FY2019 are presented in this report. Based on scientific grounds, the concept necessary for "progress of sea area monitoring" was arranged for the future medium- to long-term investigation of radiocesium concentrations. As basic information of survey frequency revise, a seabed topography and sediment distribution survey was conducted, and an attempt was made to understand the relationship between the seabed topography and the grain size distribution of bottom sediment. A columnar core sample was collected in the coastal area and analyzed for radioactive cesium concentration. In order to understand the dynamics of radioactive cesium contained in suspended matter flowing in from a river, suspended solids was collected using a sediment trap and the concentration of radioactive cesium was measured. We re-analyzed the towed monitoring data that had been implemented since 2013, and tried to improve the accuracy of the radioactive cesium distribution estimation map in the coastal area.
-ray beamOmer, M.; Shizuma, Toshiyuki*; Hajima, Ryoichi*
Nuclear Instruments and Methods in Physics Research A, 951, p.162998_1 - 162998_6, 2020/01
Times Cited Count：1 Percentile：13.56(Instruments & Instrumentation)Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08
An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. In our study, the numerical simulation with surface-tracking will be applied for the two-phase flow in fuel assemblies in order to obtain the detail data relating to the size and velocity of bubbles in the subchannel, which is needed to predict the CHF based on the mechanism. In this study, the numerical simulation of two-phase flow in 4
4 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. The simulation results are validated by the curve of flow regime for air-water under the adiabatic condition. The bubble and velocity of bubbles obtained by simulation results are analyzed.
Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10
The mechanism of Critical Heat Flux (CHF) remains to be clarified, even though it is important to evaluate the CHF for super high heat flux components such as light water reactors (LWRs). Some theoretical models to predict the CHF is proposed so far. A macrolayer formation model which is proposed in order to predict the CHF based on the macrolayer dryout model. In this model, it is assumed that the liquid is captured inside vapor mass at coalescence. In this study, the verification of the assumption of a macrolayer formation model by the numerical simulation of CMFD code, TPFIT, from the view point of hydrodynamics.
Takeda, Takeshi; Otsu, Iwao
Mechanical Engineering Journal (Internet), 5(4), p.18-00077_1 - 18-00077_14, 2018/08
Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
The mechanism of critical heat flux (CHF) for higher system pressure remains to be clarified, even though it is important to evaluate the CHF for the light water reactor (LWR) which is operated under the high pressure condition. In this study, the process of bubble coalescence was simulated by using a computational multi-fluid dynamics (CMFD) simulation code TPFIT under various system pressure in order to investigate the behavior of bubbles as a basic study. The growth of bubbles was simulated by blowing of vapor from a tiny orifice simulating bubble bottom. One or four orifices were located on the bottom surface in this simulation study. The numerical simulations were conducted by varying the pressure and temperature.
Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 14 Pages, 2017/09
An experimental study on the upward bubbly air-water flows in a vertical large-diameter square duct have been performed by using four-sensor probes. The four-sensor probe were applied in the local measurements at 3 axial positions along the flow direction to obtain interfacial area concentration, 3-D bubble velocity vector and bubble diameter. The obtained void fraction, interfacial area concentration, 3-D bubble velocity vector and bubble diameter provided valuable insight into the flow structure and will serve as a valuable database to develop the mechanistic models for interfacial area transport equation sources and sinks.
Takeda, Takeshi; Otsu, Iwao
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 11 Pages, 2017/07
Maeda, Makoto; Komeda, Masao; Tobita, Hiroshi; Ozu, Akira; Kureta, Masatoshi; Bogucarska, T.*; Crochemore, J. M.*; Varasano, G.*; Pedersen, B.*
Dai-37-Kai Kaku Busshitsu Kanri Gakkai Nihon Shibu Nenji Taikai Rombunshu (CD-ROM), 7 Pages, 2017/02
JAEA and EC/JRC are carrying out collaborative research to develop NDA techniques that can be utilized for quantification of high radioactive special nuclear materials such as spent fuel and next generation minor actinide fuels. In the research, reliability of neutron transport codes is important because it is utilized for design and development of a demonstration system of next-generation Differential Die-away (DDA) technique in JAEA. In order to evaluate the reliability, actual neutron flux distribution in a sample cavity was examined in PUNITA device using JRC type DDA technique and JAWAS-T device using JAEA type DDA technique, and then the measurement results were compared with the simulation results obtained by the neutron transport codes. The neutron flux distribution in the target matrix was also examined in the PUNITA and compared with the simulation results. We report on the measurement and simulation results of the neutron flux distribution and evaluation results of the reliability of the neutron transport codes.
Maeda, Makoto; Komeda, Masao; Ozu, Akira; Kureta, Masatoshi; Toh, Yosuke; Bogucarska, T.*; Crochemore, J. M.*; Varasano, G.*; Pedersen, B.*
EUR-28795-EN (Internet), p.694 - 701, 2017/00
