Refine your search:     
Report No.
 - 
Search Results: Records 1-11 displayed on this page of 11
  • 1

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

Phase transition and domain formation in ferroaxial crystals

Hayashida, Takeshi*; Uemura, Yohei*; Kimura, Kenta*; Matsuoka, Satoshi*; Hagihara, Masato; Hirose, Sakyo*; Morioka, Hitoshi*; Hasegawa, Tatsuo*; Kimura, Tsuyoshi*

Physical Review Materials (Internet), 5(12), p.124409_1 - 124409_10, 2021/12

Journal Articles

Development and application of a $$^3$$He neutron spin filter at J-PARC

Okudaira, Takuya; Oku, Takayuki; Ino, Takashi*; Hayashida, Hirotoshi*; Kira, Hiroshi*; Sakai, Kenji; Hiroi, Kosuke; Takahashi, Shingo*; Aizawa, Kazuya; Endo, Hitoshi*; et al.

Nuclear Instruments and Methods in Physics Research A, 977, p.164301_1 - 164301_8, 2020/10

 Times Cited Count:4 Percentile:76.41(Instruments & Instrumentation)

JAEA Reports

An Investigation for long-term storage of a spent zeolite adsorption vessel; Estimation of washing out salt component in a spent zeolite adsorption vessel, 1

Sato, Hiroyuki; Terada, Atsuhiko; Hayashida, Hitoshi; Kamiji, Yu; Kobayashi, Jun; Yamagishi, Isao; Morita, Keisuke; Kato, Chiaki

JAEA-Research 2013-042, 25 Pages, 2014/03

JAEA-Research-2013-042.pdf:5.13MB

Spent zeolite adsorption vessels in the Fukushima No.1 nuclear power plant are kept for long-term with washing out with fresh water for prevention of corrosion remaining salt component in vessel. However, corrosion result is concerned by residual concentration of salt component, washing out experiment is carried out using actual and unspent adsorption vessel (KURION). KURION adsorption vessel is filled with 1,650 ppm of sodium chloride (1,000 ppm of chloride ion) and washed out with pure water for estimating washing effect in this experiment. Pure water is streamed with volume flow rate 4.5 m$$^{3}$$/h, chloride concentration in vessel is measured with drainage sample water. 1,000 ppm of chloride concentration is decreased till 0.5 ppm and below by washing out with about double pure water volume of adsorbing material filling volume in vessel, washing out is more effective in KURION adsorption vessel.

Journal Articles

Evaluation of water distribution in a small operating fuel cell using neutron color image intensifier

Yasuda, Ryo; Nitto, Koichi*; Konagai, Chikara*; Shiozawa, Masahiro*; Takenaka, Nobuyuki*; Asano, Hitoshi*; Murakawa, Hideki*; Sugimoto, Katsumi*; Nojima, Takehiro; Hayashida, Hirotoshi; et al.

Nuclear Instruments and Methods in Physics Research A, 651(1), p.268 - 272, 2011/09

 Times Cited Count:7 Percentile:53.57(Instruments & Instrumentation)

Neutron radiography is one of useful tools to visualize water behavior in fuel cells under operation. In order to observe the detailed information about the water distribution in MEA and GDL in fuel cells, a high spatial resolution and high sensitivity neutron imaging system are required. We developed an imaging system using the neutron color imaging intensifier and continuously observed water distribution in operating a fuel cell. By using the system, a small type fuel cell under operation was continuously observed at the TNRF in every 20 sec. In the results, the water area was appeared from GDL and MEA, and expanded to the channel of the cathode side. On the other hand, voltage was gradually reduced with the operation time, and steeply dropped. It is considered that voltage drop was caused by blockage of gas flow due to the piling up water in the channel of the cathode side.

JAEA Reports

Study on sodium viewing technique applying sodium gamma-rays emission tomography; Verification of basic principle by analytical investigation

Hirabayashi, Masaru; Otaka, Masahiko; Hayashida, Hitoshi; Ara, Kuniaki

JNC TN9400 2003-016, 35 Pages, 2003/04

JNC-TN9400-2003-016.pdf:1.45MB

To confirm structural integrity of a primary cooling system and in-vessel components in a sodium-cooled fast breeder reactor, monitoring and inspection technique applying Gamma-rays emitted from sodium are proposed. The basic principle is as follows. As radioisotope $$^{24}$$Na decays, photons are emitted and a fraction of these photons penetrate materials. If the number of these photons is counted by radiation detectors, an image of gamma-rays source is reconstructed by a computed tomography technique. In this report, Applicability and problems concerned with the technique are investigated. Main results are as follows: (1)To verify applicability, the technique was analytically investigated based on gamma-rays emitted from sodium coolant in a typical pipe of a primary cooling system. As a result, it was confirmed that the image of gamma-rays source could be reconstructed. (2)A required time to measure in a spatial resolution of about 1mm was investigated in the detection efficiency of 20%. The time was about 4 minutes per section by a thousand detectors in the typical pipe of a primary cooling system. And in a typical steam generator, the time was about 2 days per section by ten thousand detectors. (3)To realize a fluoroscopic inspection system, it is necessary that the principle should be verified by experimental researches. Main equipments of the system are a collimator, radiation detector, scanner, signal processing device and image processing device. As a spatial resolution is decide by the collimator, the shape must be evaluated by experimental researches and analytical investigation.

JAEA Reports

The Report of inspection and repair technology of sodium cooled reactors

Kisohara, Naoyuki; Uchita, Masato; Konomura, Mamoru; Kasai, Shigeo; Soman, Yoshindo; Shimakawa, Yoshio; Hori, Toru; Chikazawa, Yoshitaka; Miyahara, Shinya; Hamada, Hirotsugu; et al.

JNC TN9400 2003-002, 109 Pages, 2002/12

JNC-TN9400-2003-002.pdf:8.12MB

Sodium is the most promising candidate of an FBR coolant because of its excellent properties such as high thermal conductivity. Whereas, sodium reacts with water/air and its opaqueness makes it difficult to inspect sodium components. These weaknesses of sodium affect not only plant safety but also plant availability (economy). To overcome these sodium weak points, the appropriate countermeasure must be adopted to commercialized FBR plants. This report describes the working group activities for sodium/water reaction of steam generators (SG), in-service inspection for sodium components and sodium leak due to sodium components boundary failure. The prospect of each countermeasure is discussed in the viewpoint of the commercialized FBR plants. (1)Sodium/water reaction. The principle of the countermeasure for sodium/water reaction accidents was organized in the viewpoint of economy (the investment of SG and the plant availability). The countermeasures to restrain failure propagation were investigated for a large-sized SG. Preliminary analysis revealed the possibility of minimizing tubes failure propagation by improving the leak detection system and the blow down system. Detailed failure propagation analysis will be required and the early water leak detection system and rapid blow down system must be evaluated to realize its performance. (2)In-service inspection (ISI&R). The viewpoint of the commercialized plant's ISI&R was organized by comparing with the prototype reactor's ISI&R method. We also investigated short-term ISI&R methods without sodium draining to prevent the degrading of the plant availability, however, it is difficult to realize them with the present technology. Hereafter, the ISI&R of the commercialized plants must be defined by considering its characteristics. (3)Sodium leak from the components. This report organized the basic countermeasure policy for primary and secondary sodium leak accidents. Double-wall structure of sodium piping was ...

JAEA Reports

The ultrasonic wave thermometer sodium test, 1; A summary of test results of the externally mounted ultrasonic transducer for pipe-flow

Hayashida, Hitoshi; Kokaki, Nobuhisa; Ueda, Masashi; Isozaki, Tadashi; Ara, Kuniaki

JNC TN9400 98-001, 54 Pages, 1998/10

JNC-TN9400-98-001.pdf:1.39MB

Based on the temperature dependence of the velocity of sound in sodium, an ultrasonic thermometer that measures the temperature of sodium non-intrusively is being developed. The principle of the device is based on the propagation time of an acoustic pulse wave, and the back calculation of the sodium temperature. As the part of the development a test was actually carried out in sodium pipe-flow in order to evaluate various aspects of realizing the ultrasonic wave thermometer. The results and conclusions to date are as follows: (1)Within the present test range, the ultrasonic wave thermometer appears relatively insensitive to flow velocity of sodium, pressure of the cover gas and the impurity concentration in sodium. The calculated error of the measured thermometry was in the experiment about 1 $$^{circ}$$C, a smaller value than the expected 2.5$$^{circ}$$C of the system. (2)The ultrasonic thermometer has only been used wherein the thermal expansion coefficient was known and with 200 $$^{circ}$$C as the reference temperatures. For the entire temperature range tested the difference between this approach and a two-point calibration over a temperature range is only expected to be about 1 $$^{circ}$$C. (3)By using the mean value of multiple ultrasonic wave transmit and receive measurements, a value whereby the ultrasonic propagation time was stabilized is obtained. (4)As acoustic coupling material between the ultrasonic transducer and piping, a copper plate was found to be more suitable than a specialized acoustic bonding material. A weight equivalent, area distributed force of 2.0kg/mm$$^{2}$$ was used to press the test copper plate to the pipe. A slightly smaller force appears more than sufficient as well. (5)We found that mounting the ultrasonic transducer to the exterior surface of the pipe by a clamping method is sufficient such that no welding is needed. (6)The in-sodium test period was about 2 months. No noticeable change in measurement characteristics of the ...

JAEA Reports

The development of liquid surface visualization and void detection system for sodium; Imaging characteristic examination of sound propagation simulation and underwater basic test

Hayashida, Hitoshi; Hirabayashi, Masaru

PNC TN9410 97-053, 29 Pages, 1997/05

PNC-TN9410-97-053.pdf:1.72MB

On the FBR plant, it is necessary that cover gas would not involve in the primary coolant, and especially the measures is very important in order to compact the reactor on the design of DFBR. Thus the behavior of free liquid surface must be evaluated in sodium pool tests at the design phase. And also from the viewpoint of safety analyses, the void detection system which measure the void in the reactor vessel inlet piping is also developed as a measurement system for confirming that the void has not been entrapped in the coolant. Then on the liquid surface visualization system which image the behavior of liquid sodium surface and the void detection system which image the detected void using the ultrasonic wave, the sound propagation simulations and the underwater basic tests were carried out to obtain a knowledge on the basic imaging characteristics. As a first step, the sound propagation simulations were carried out to evaluate the imaging characteristics of the vortex model. As the next step, the measurement system was produced tentatively in order to image the fluctuating liquid surface and the rising void in the underwater basic tests. The signals from some ultrasonic transducers to image them were processed by the cross correlation method and the aperture synthesis method. However the imaging processed in off-line because of the limitation of thc signal processing ability. As the results, it was confirmed that the vortex model could be imaged by the sound propagation simulations. And the fluctuating liquid surface and the rising void could be also imaged by the measurement system in the underwater basic tests. However since the slope of vortex can't be imaged well, the most suitable arrangement of the transducers will be examined to image the slope. And it will be necessary for the measurement system to confirm that the size and shape of void can be imaged accurately.

JAEA Reports

Large-scaled thermohydraulic tests plan for cooling systems in fast reactors; Experimental models of reactor vessel and the primary cooling system

Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.

PNC TN9410 96-279, 51 Pages, 1996/08

PNC-TN9410-96-279.pdf:2.92MB

Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.

Oral presentation

Evaluation of water distribution in a fuel cell under operation by neutron radiography

Yasuda, Ryo; Shiozawa, Masahiro*; Takenaka, Nobuyuki*; Asano, Hitoshi*; Hayashida, Hirotoshi; Sakai, Takuro; Honda, Mitsunori; Iikura, Hiroshi; Nojima, Takehiro; Matsubayashi, Masahito

no journal, , 

Neutron radiography is an effective water diagnostic tool for fuel cells. We continuously observed water behavior in a small type fuel cell under operation by high performance neutron radiography system using Neutron image intensifier. Water generated in the cell was initially accumulated in region of MEA and GDL, extending to a channel gradually. We will report and discuss about relationship between voltage drop and the water behavior in presentation.

Patent

INTERMEDIATE HEAT EXCHANGER-INCORPORATED TYPE STEAM GENERATOR

林田 均; 荒 邦章

not registered

JP, 10/138306  Patent licensing information  Patent publication (In Japanese)

An intermediate heat exchanger-incorporated type steam generator having an intermediate heat exchanger tube20and a steam generating heat exchanger tube22disposed separately in a vessel10storing a secondary coolant14therein. A pump mechanism is made of an electromagnetic pump mechanism formed by an electromagnetic driving coil24provided on an outer circumference of the vessel and a magnetic core26attached to an inner cylinder12disposed in the vessel. At least one porous plate or slotted plate18is preferably disposed between the intermediate heat exchanger tube and steam generating heat exchanger tube.

11 (Records 1-11 displayed on this page)
  • 1