Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 207

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Uto, Hiroyasu; Sakamoto, Yoshiteru; Gulden, W.*

Nuclear Fusion, 55(12), p.123008_1 - 123008_7, 2015/12

 Times Cited Count:4 Percentile:67.21(Physics, Fluids & Plasmas)

Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. The thermohydraulic analysis results suggests that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. As for the in-vessel LOCA, it was found that the pressure in the vacuum vessel reaches its design value due to the LOCA even though a pressure suppression system is in service. As for the ex-vessel LOCA, the pressure load to the tokamak hall due to the double-ended break of the primary cooling pipe was found to be so large that integrity of the hall was crucially challenged. Mitigations of the loads to the confinement barriers are also discussed.

Journal Articles

Behavior of high burnup advanced fuels for LWR during design-basis accidents

Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Sugiyama, Tomoyuki

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2015), Part.2 (Internet), p.10 - 18, 2015/09

Advanced fuels which consist of cladding materials with high corrosion resistance and pellets with lower fission gas release have been developed by utilities and fuel vendors to improve fuel performance even in the high burnup region and also raise the safety level of current nuclear power plants to a higher one. In order to evaluate the adequacy of present safety criteria and safety margins in terms of such advanced fuels and provide a database for future regulation on them, Japan Atomic Energy Agency (JAEA) has started a new extensive research program called ALPS-II program (Phase II of Advanced LWR Fuel Performance and Safety program). This program is primarily composed of tests simulating a reactivity-initiated accident (RIA) and a loss-of-coolant accident (LOCA) on high burnup advanced fuels shipped from European nuclear power plants. This paper describes an outline of this program and some experimental results with respect to RIA and LOCA which have been obtained in this program.

Journal Articles

New reactor cavity cooling system using novel shape for HTGRs and VHTRs

Takamatsu, Kuniyoshi; Hu, R.*

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12

continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 ($$^{circ}$$C). The RCCS uses novel shape so that the heat released from the RPV can be removed efficiently with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design greatly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting owing to overheating the fuels.

Journal Articles

Key aspects of the safety study of a water-cooled fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10

Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.

Journal Articles

Analysis of accident scenarios of a water-cooled tokamak DEMO

Nakamura, Makoto; Ibano, Kenzo*; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Ogawa, Yuichi*

Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10

Of late in Japan, a design study has been undertaken of a tokamak fusion DEMO with pressurized water coolant and solid pebble bed breeding blanket, but safety characteristics of this type of DEMO have not been well examined. In this paper, thermohydraulics analysis of in-vessel and ex-vessel loss-of-coolant accidents of a water-cooled tokamak DEMO is reported. Safety characteristics of water-cooled DEMO, particularly possible loads onto confinement barriers, are discussed based on the thermohydraulics analysis results. Measures to reduce such loads are also proposed.

Journal Articles

LOCA and RIA studies at JAERI

Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi

HPR-362, Vol.2, 12 Pages, 2004/05

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss of coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI.

JAEA Reports

Annual report on operation, utilization and technical development of Hot Laboratories; April 1 2002 to March 31 2003

Department of Hot Laboratories

JAERI-Review 2003-038, 106 Pages, 2003/12

JAERI-Review-2003-038.pdf:9.36MB

no abstracts in English

JAEA Reports

Literature survey of thermal-hydraulic studies on super-critical pressurized water

Kurihara, Ryoichi; Watanabe, Kenichi*; Konishi, Satoshi

JAERI-Review 2003-020, 37 Pages, 2003/07

JAERI-Review-2003-020.pdf:2.08MB

no abstracts in English

Journal Articles

Experimental results of functional performance of a vacuum vessel pressure suppression system in ITER

Shibata, Mitsuhiko; Takase, Kazuyuki; Watanabe, Hironori; Akimoto, Hajime

Fusion Engineering and Design, 63-64, p.217 - 222, 2002/12

 Times Cited Count:3 Percentile:81.08

no abstracts in English

Journal Articles

Consideration for the safety of a fusion power plant

Konishi, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 78(11), p.1157 - 1164, 2002/11

Based on the fundamental approach for safety of ITER, possible extension to assure safety of fusion power plant was considered. Although the entire analysis and licensing preparation are specific for ITER, its methodology that take full advantage of inherent feature of fusion is expected to be applied to fundamental logic of fusion power plant. Both energy and radioactive source terms that could be potential hazards are typically of order of days of operation rather than a year in the case of fission. Major difference from the test reactor ITER identified was power blanket, coolant loop and generator train that will hold high temperature and considerable amount of tritium. It is anticipated that tritium inventory and most of the tritium safety plant would essentially be same as ITER, tritium recovery and removal from blanket loop will dominate the fusion power plant tritium systems. Such a tritium system will actively remove tritium at the daily throughput comparable with the order of plant inventory. This feature suggest no dedicated off-normal systems are needed to assure safety of fusion plant from the aspect of environmental tritium release.

Journal Articles

Study on sudden loss of cryogenic coolant accident happened in the hydrogen isotope separation system for fusion reactor

Iwai, Yasunori; Nakamura, Hirofumi; Konishi, Satoshi; Nishi, Masataka; Willms, R. S.*

Fusion Science and Technology, 41(3), p.668 - 672, 2002/05

no abstracts in English

Journal Articles

Numerical visualization of water-vapor flow configurations in fusion reactors during ingress of coolant events

Ose, Yasuo*; Takase, Kazuyuki; Yoshida, Hiroyuki; Akimoto, Hajime

Proceedings of 10th International Conference on Nuclear Engineering (ICONE 10) (CD-ROM), 8 Pages, 2002/00

no abstracts in English

Journal Articles

Experimental verification of effectiveness of integrated pressure suppression systems in fusion reactors during in-vessel loss of coolant events

Takase, Kazuyuki; Akimoto, Hajime

Nuclear Fusion, 41(12), p.1873 - 1883, 2001/12

 Times Cited Count:3 Percentile:87.17(Physics, Fluids & Plasmas)

no abstracts in English

JAEA Reports

Transient analyses of mercury loop for spallation target

Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro

JAERI-Tech 2001-052, 41 Pages, 2001/08

JAERI-Tech-2001-052.pdf:1.99MB

no abstracts in English

Journal Articles

Numerical analysis on ingress of coolant event in vacuum vessel using modified TRAC-BF1

Kurihara, Ryoichi; Ajima, Toshio*; Ueda, Shuzo; Seki, Yasushi

Journal of Nuclear Science and Technology, 38(7), p.571 - 576, 2001/07

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Analysis of pressure rise in an ITER-Like Fusion reactor during in-vessel loca by a modified TRAC-PF1 code

Takase, Kazuyuki; Ose, Yasuo*; Akimoto, Hajime

Fusion Technology, 39(2-Part.2), p.1050 - 1055, 2001/03

no abstracts in English

Journal Articles

Experimental verification of integrated pressure suppression systems in fusion reactors at in-vessel loss-of-coolant events

Takase, Kazuyuki; Akimoto, Hajime

Proceedings of IAEA 18th Fusion Energy Conference (CD-ROM), 5 Pages, 2001/00

no abstracts in English

JAEA Reports

CompalisonoFnlermohydraulicCharacteristicsintheuseofvariousCoolants

; ; *; Yamaguchi, Akira

JNC-TN9400 2000-109, 96 Pages, 2000/11

JNC-TN9400-2000-109.pdf:9.56MB

Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiaied based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1)Ther is no remarkable difference betweeen liquid sodium and luquid Pb-Bi in characteristics of internal flows and free surface charatristics based on Fr number. (2)the AQUA-VOF code has a potentiall enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [thermal Stratification Phenomena] (1)On-set position of thermal entainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. 0n the other hand, the position in the case of C0$$_{2}$$ gas was shifted to upstream side with decreasing of Ri number. (2)Destruction speed of the thermal stratyification interface was dependent on thermal diffusivity as fluid properties. therefor it was concluded that an elimination method is necessary for the interface generated in C0$$_{2}$$ gas. [thermal Striping Phenomena] (1)Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO$$_{2}$$ gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2)To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it isnecessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phenomena] (1)Fundamental behavior of the natural convection in various coolant follows buoyant jet ....

Journal Articles

Experimental and analytical studies on water-vapor two-phase flow behavior under low pressure

Takase, Kazuyuki; Ose, Yasuo*; Akimoto, Hajime

Nippon Kikai Gakkai Ryutai Kogaku Bumon Koenkai 2000 Koen Rombunshu (CD-ROM), 4 Pages, 2000/09

no abstracts in English

JAEA Reports

None

; Inagaki, Tatsutoshi*

JNC-TY1400 2000-004, 464 Pages, 2000/08

JNC-TY1400-2000-004.pdf:19.55MB

None

207 (Records 1-20 displayed on this page)