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Journal Articles

Direct ${it in-situ}$ temperature measurement for lamp-based heating device

Sumita, Takehiro; Sudo, Ayako; Takano, Masahide; Ikeda, Atsushi

Science and Technology of Advanced Materials; Methods (Internet), 2(1), p.50 - 54, 2022/02

Journal Articles

The Effects of steam addition on the unstable behavior of hydrogen-air lean premixed flames under the adiabatic and non-adiabatic conditions

Furuyama, Taisei*; Thwe, T. A.; Katsumi, Toshiyuki; Kobayashi, Hideaki*; Kadowaki, Satoshi

Nihon Kikai Gakkai Rombunshu (Internet), 87(898), p.21-00107_1 - 21-00107_12, 2021/06

The effects of steam addition on the unstable behavior of hydrogen-air lean premixed flames under adiabatic and non-adiabatic conditions were investigated by numerical calculations. Adopting a detailed chemical reaction mechanism of hydrogen-oxyfuel combustion modeled by 17 reversible reactions of 8 active species and diluents, a two-dimensional unsteady reaction flow was treated based on the compressible Navier-Stokes equation. As the steam addition and heat loss increased, the burning velocity of a planar flame decreased and the normalized burning velocity increased. The addition of water vapor promotes the unstable behavior of the hydrogen-air lean premixed flame. This is because the thermal diffusivity of the gas decreases and the diffusion-thermal instability increases. The effect of adding water vapor on the instability of hydrogen premixed flames is a new finding, and it is expected to connect with hydrogen explosion-prevention measures as in NPP.

Journal Articles

Effects of pressure and heat loss on the unstable motion of cellular-flame fronts caused by intrinsic instability in hydrogen-air lean premixed flames

Kadowaki, Satoshi; Thwe, T. A.; Furuyama, Taisei*; Kawata, Kazumasa*; Katsumi, Toshiyuki; Kobayashi, Hideaki*

Journal of Thermal Science and Technology (Internet), 16(2), p.20-00491_1 - 20-00491_12, 2021/00

Effects of pressure and heat loss on the unstable motion of cellular-flame fronts in hydrogen-air lean premixed flames were numerically investigated. The reaction mechanism for hydrogen-oxygen combustion was modeled with seventeen reversible reactions of eight reactive species and a diluent. Two-dimensional unsteady reactive flow was treated, and the compressibility, viscosity, heat conduction, molecular diffusion and heat loss were taken into account. As the pressure became higher, the maximum growth rate increased and the unstable range widened. These were due mainly to the decrease of flame thickness. The burning velocity of a cellular flame normalized by that of a planar flame increased as the pressure became higher and the heat loss became larger. This indicated that the pressure and heat loss affected strongly the unstable motion of cellular-flame fronts. In addition, the fractal dimension became larger, which denoted that the flame shape became more complicated.

Journal Articles

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021113_1 - 021113_9, 2020/04

Journal Articles

Three-dimensional cellular premixed flames generated by hydrodynamic and diffusive-thermal instabilities (Effects of unburned-gas temperature and heat loss)

Kadowaki, Satoshi; Nogami, Masato*; Thwe, T. A.; Katsumi, Toshiyuki*; Yamazaki, Wataru*; Kobayashi, Hideaki*

Nihon Kikai Gakkai Rombunshu (Internet), 85(879), p.19-00274_1 - 19-00274_13, 2019/11

We dealt with three-dimensional cellular premixed flames generated by hydrodynamic and diffusive-thermal instabilities to elucidate the effects of unburned-gas temperature and heat loss by adopting the three-dimensional compressible Navier-Stokes equation. As the unburned-gas temperature became lower and the heat loss became larger, the growth rate decreased and the unstable range narrowed. With a decrease of unburned-gas temperature, the normalized growth rate increased and the normalized unstable range widened, which was because the temperature ratio of burned and unburned gases became larger. The obtained hexagonal cellular fronts were qualitatively consistent with the experimental results. As the heat loss became larger, the burning velocity of a cellular flame normalized by that of a planar flame increased. This was because diffusive-thermal effects became stronger owing to the increase of apparent Zeldovich number caused by the decrease of flame temperature.

Journal Articles

Numerical investigation on unstable behaviors of cellular premixed flames at low Lewis numbers based on the diffusive-thermal model and compressible Navier-Stokes equations

Thwe, T. A.; Kadowaki, Satoshi; Hino, Ryutaro

Journal of Thermal Science and Technology (Internet), 13(2), p.18-00457_1 - 18-00457_12, 2018/12

 Times Cited Count:0 Percentile:0.01(Thermodynamics)

Two dimensional unsteady calculations of reactive flows were performed in large domain to investigate the unstable behaviors of cellular premixed flames at low Lewis numbers based on the diffusive-thermal (D-T) model and compressible Navier-Stokes (N-S) equations. The growth rates obtained by the compressible N-S equations were large and the unstable ranges were wide compared with those obtained by the D-T model equations. When the length of computational domain increased, the number of small cells separated from large cells of the cellular flame increased drastically. The stronger unstable behaviors and the larger average burning velocities were observed especially in the numerical results based on the compressible N-S equations. In addition, the fractal dimension obtained by the compressible N-S equations was larger than that by the D-T model equations. Moreover, we confirmed that the radiative heat loss promoted the instability of premixed flames at low Lewis numbers.

Journal Articles

Free convective heat transfer experiment to validate air-cooling performance analysis of fuel debris

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Journal Articles

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

Abe, Yuta; Yamashita, Takuya; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Journal Articles

Evaluation and demonstration of cutting the fuel assembly heating examination by AWJ

Maruyama, Shinichiro*; Watatani, Satoshi*

Mitsui Sumitomo Kensetsu Gijutsu Kenkyu Kaihatsu Hokoku, (15), p.107 - 112, 2017/10

It is essential to estimate characteristics and forms of fuel debris for safe and reliable removing at the decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F). For the estimation, melting behavior of fuel assembly in the accident is being researched. To proceed the research, the fuel debris were need to cut, and the abrasive water jet (AWJ) which had enough results for cutting ceramic material or mixed material of zirconium alloy and stainless. The test results demonstrated that AWJ could cut the fuel assembly and accumulated the cutting data which will be subservient when removing the fuel debris in future.

Journal Articles

Heat treatment of phosphate-modified cementitious matrices for safe storage of secondary radioactive aqueous wastes in Fukushima Daiichi Nuclear Power Plant

Irisawa, Keita; Taniguchi, Takumi; Namiki, Masahiro; Garc$'i$a-Lodeiro, I.*; Osugi, Takeshi; Sakakibara, Tetsuro; Nakazawa, Osamu; Meguro, Yoshihiro; Kinoshita, Hajime*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

A solidification technique with minimized water content is being developed using phosphate cements for the safe storage of secondary radioactive wastes in the Fukushima Daiichi Nuclear Power Plant. Conventional cement systems become solidified via hydration reactions, and need a certain water content. Phosphate cement systems, however, become solidified via an acid-base reaction, and so they only require water mainly for reasons of workability. A reduced water content of phosphate cement systems is beneficial for the immobilization of the radioactive wastes from mitigating the potential to generate hydrogen gas by the radiolysis of water by radioactive wastes. The current study investigated the water content and mineralogy of calcium aluminate cement (CAC) and phosphate-modified CAC (CAP) cured in open systems at 60, 90 and 120 $$^{circ}$$C and in a closed system at 20 $$^{circ}$$C as a reference case. Water contents in both the CAC and the CAP were seen to decrease as curing progressed. For $$geq$$ 90 $$^{circ}$$C, the CAP contained less water than CAC. Free water in CAC converted to structural water by heat treatment, but this was not the case for CAP. An orthophosphate hydrate salt, a precursor phase of hydroxyapatite, was found in CAP when cured at 20 and 60 $$^{circ}$$C, and a mixture of the orthophosphate hydrate salt and hydroxyapatite, Ca$$_{10}$$(PO$$_{4}$$)$$_{6}$$(OH)$$_{2}$$, were formed in the CAP when cured at 90 $$^{circ}$$C. Phosphate products in CAP cured at 120 $$^{circ}$$C appears to consist of a different phosphate phase compared with the CAP cured at 20, 60 and 90 $$^{circ}$$C.

Journal Articles

Development of non-transfer type plasma heating technology to address CMR behavior during severe accident with BWR design conditions

Abe, Yuta; Sato, Ikken; Nakagiri, Toshio; Ishimi, Akihiro; Nagae, Yuji

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Journal Articles

Investigation of countermeasure against local temperature rise in vessel cooling system in loss of core cooling test without nuclear heating

Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro

Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

JAEA Reports

Study of HTGR contribution to Japan's CO$$_{2}$$ emission reduction goal in 2050

Kamiji, Yu; Suzuki, Koichi*; Yan, X.

JAEA-Technology 2016-010, 24 Pages, 2016/07

JAEA-Technology-2016-010.pdf:1.05MB

Japanese government has set the goal of reducing CO$$_{2}$$ emission by 26% in 2030 below the 2013 level, in longer term, by 80% below the 1990 level. To achieve the goals, various measures should be taken. The GTHTR300, a commercial High Temperature Gas-cooled Reactor (HTGR) design being developed by JAEA offers spectrum of heat applications by using its high temperature heat up to 950$$^{circ}$$C. The potential contribution of CO$$_{2}$$ emission reduction by HTGR is estimated considering domestic and overseas deployment of the GTHTR300. The best estimate for domestic CO$$_{2}$$ reduction is 2.07$$times$$10$$^{8}$$ ton- CO$$_{2}$$/yr and that from oversea is 2.25$$times$$10$$^{8}$$ ton- CO$$_{2}$$/yr. The sum of these is about 47% of 9.13$$times$$10$$^{8}$$ ton- CO$$_{2}$$/yr which is CO$$_{2}$$ reduction target in 2050, for which deployment of 52 plants in Japan and 113 plants abroad, with each plant containing four 600 MWt reactor units, is required.

Journal Articles

Preparation for a new experimental program addressing core-material-relocation behavior during severe accident with BWR design conditions; Conduction of preparatory tests applying non-transfer-type plasma heating technology

Abe, Yuta; Sato, Ikken; Ishimi, Akihiro; Nakagiri, Toshio; Nagae, Yuji

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm $$times$$ 107 mm $$times$$ 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using ca. 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.

Journal Articles

Pool nucleate boiling for seawater containing minerals

Uesawa, Shinichiro; Koizumi, Yasuo; Yoshida, Hiroyuki

Proceedings of 9th International Conference on Multiphase Flow (ICMF 2016) (CD-ROM), 6 Pages, 2016/05

Journal Articles

Computational and experimental examination of simulated core damage and relocation dynamics of a BWR fuel assembly

Hanus, G.*; Sato, Ikken; Iwama, Tatsuya*

Proceedings of International Waste Management Symposia 2016 (WM 2016) (Internet), 12 Pages, 2016/03

JAEA plans a large-scale test to evaluate damage and relocation behavior of BWR core materials consisting of fuel rods, channel boxes, control blade and lower support structures. Its purpose is to contribute to understanding of core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing, JAEA desires preliminary investigations to examine melting test pieces. The purpose of such tests is to verify the materials and test piece will be heated by plasma to the target temperature (ca.2900K) and to collect data about the material relocation behavior. Results from preliminary computational simulations are presented illustrating the effectiveness of a 150 kW non-transferred plasma jet. An experimental test program using the computational analyses as a basis and a plasma torch is described.

Journal Articles

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 2; Heat transfer and flow visualization experiment by using internally heated annulus

Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Yoshida, Hiroyuki

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11

Journal Articles

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 1; Outline of the research project

Yoshida, Hiroyuki; Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 9 Pages, 2015/11

Journal Articles

Nuclear heat supply fluctuation test by non-nuclear heating using HTTR

Takada, Shoji; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Inaba, Yoshitomo; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

To investigate the safety design criteria of heat utilization system for the HTGRs, it is necessary to evaluate the effect of fluctuation of thermal load on the reactor. The nuclear heat supply fluctuation test by non-nuclear heating was carried out to simulate the nuclear heat supply test which is carried out in the nuclear powered operation. The test data is used to verify the numerical code to calculate the temperature of core bottom structure to carry out the safety evaluation of abnormal events in the heat utilization system. In the test, the helium gas temperature was heated up to 120$$^{circ}$$C. A sufficiently high temperature disturbance was imposed on the reactor inlet temperature. It was found that the response of temperatures of metallic components such as side shielding blocks was faster than those of graphite blocks in the core bottom structure, which was significantly affected by the heat capacities of components, the level of imposed disturbance and heat transfer performance.

Journal Articles

Investigation of characteristics of natural circulation of water in vessel cooling system in loss of core cooling test without nuclear heating

Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05

In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1$$^{circ}$$C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.

127 (Records 1-20 displayed on this page)