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Journal Articles

Giant multiple caloric effects in charge transition ferrimagnet

Kosugi, Yoshihisa*; Goto, Matato*; Tan, Z.*; Kan, Daisuke*; Isobe, Masahiko*; Yoshii, Kenji; Mizumaki, Masaichiro*; Fujita, Asaya*; Takagi, Hidenori*; Shimakawa, Yuichi*

Scientific Reports (Internet), 11(1), p.12682_1 - 12682_8, 2021/06

 Times Cited Count:0 Percentile:0(Multidisciplinary Sciences)

Caloric effects of solids provide more efficient and environment-friendly innovative refrigeration systems compared to the widely-used conventional vapor compressive cooling systems. Exploring novel caloric materials is challenging but critically important in developing future technologies. Here we discovered that the quadruple perovskite structure ferrimagnet BiCu$$_{3}$$Cr$$_{4}$$O$$_{12}$$ shows a large multicaloric effect at the first-order charge transition occurred around 190 K. Large latent heat and the corresponding isothermal entropy changes 28.2 J K$$^{-1}$$ kg$$^{-1}$$ can be fully utilized by applying both magnetic fields (magnetocaloric effect) and pressure (barocaloric effect). Adiabatic temperature changes reach 3.9 K for the 50 kOe magnetic field and 4.8 K for the 4.9 kbar pressure, and thus highly efficient thermal controls are achieved by multiple ways.

Journal Articles

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950$$^{circ}$$C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Journal Articles

A Model intercomparison of atmospheric $$^{137}$$Cs concentrations from the Fukushima Daiichi Nuclear Power Plant accident, phase III; Simulation with an identical source term and meteorological field at 1-km resolution

Sato, Yosuke*; Sekiyama, Tsuyoshi*; Fang, S.*; Kajino, Mizuo*; Qu$'e$rel, A.*; Qu$'e$lo, D.*; Kondo, Hiroaki*; Terada, Hiroaki; Kadowaki, Masanao; Takigawa, Masayuki*; et al.

Atmospheric Environment; X (Internet), 7, p.100086_1 - 100086_12, 2020/10

The third model intercomparison project for investigating the atmospheric behavior of $$^{137}$$Cs emitted during the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident (FDNPP-MIP) was conducted. A finer horizontal grid spacing (1 km) was used than in the previous FDNPP-MIP. Nine of the models used in the previous FDNPP-MIP were also used, and all models used identical source terms and meteorological fields. Our analyses indicated that most of the observed high atmospheric $$^{137}$$Cs concentrations were well simulated, and the good performance of some models improved the performance of the multi-model ensemble. The analyses also confirmed that the use of a finer grid resolution resulted in the meteorological field near FDNPP being better reproduced. The good representation of the wind field resulted in the reasonable simulation of the narrow distribution of high deposition amount to the northwest of FDNPP and the reduction of the overestimation over the area to the south of FDNPP. In contrast, the performance of the models in simulating plumes observed over the Nakadori area, the northern part of Gunma, and the Tokyo metropolitan area was slightly worse.

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 4; Investigation of fuel loading effects in BWR spent fuel rack

Tojo, Masayuki*; Kanazawa, Toru*; Nakashima, Kazuo*; Iwamoto, Tatsuya*; Kobayashi, Kensuke*; Goto, Daisuke*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05

In this study, fuel loading effects in BWR spent fuel rack accidents are widely investigated using three-dimensional analysis methods from both nuclear and thermal hydraulics viewpoints, including: (a) Decay heat of spent fuel after discharge, (b) The maximum temperature of spent fuel cladding in the spent fuel rack depending on heat transfer phenomena, and (c) Criticality of the spent fuel rack after collapsing of the fuel due to a severe accidents in the BWR spent fuel pool (SFP).

Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.

Journal Articles

Model intercomparison of atmospheric $$^{137}$$Cs from the Fukushima Daiichi Nuclear Power Plant accident; Simulations based on identical input data

Sato, Yosuke*; Takigawa, Masayuki*; Sekiyama, Tsuyoshi*; Kajino, Mizuo*; Terada, Hiroaki; Nagai, Haruyasu; Kondo, Hiroaki*; Uchida, Junya*; Goto, Daisuke*; Qu$'e$lo, D.*; et al.

Journal of Geophysical Research; Atmospheres, 123(20), p.11748 - 11765, 2018/10

 Times Cited Count:18 Percentile:75.7(Meteorology & Atmospheric Sciences)

A model intercomparison of the atmospheric dispersion of $$^{137}$$Cs emitted following the Fukushima Daiichi Nuclear Power Plant accident was conducted by 12 models to understand the behavior of $$^{137}$$Cs in the atmosphere. The same meteorological data, horizontal grid resolution, and an emission inventory were applied to all the models to focus on the model variability originating from the processes included in each model. The multi-model ensemble captured 40% of the observed $$^{137}$$Cs events, and the figure-of-merit in space for the total deposition of $$^{137}$$Cs exceeded 80. Our analyses indicated that the meteorological data were most critical for reproducing the $$^{137}$$Cs events. The results also revealed that the differences among the models were originated from the deposition and diffusion processes when the meteorological field was simulated well. However, the models with strong diffusion tended to overestimate the $$^{137}$$Cs concentrations.

Journal Articles

Analytical study of the applicability of FeCrAl-ODS cladding for BWR

Takano, Sho*; Kusagaya, Kazuyuki*; Goto, Daisuke*; Sakamoto, Kan*; Yamashita, Shinichiro

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

We focused on one of accident tolerant fuel (ATF) materials, Oxide Dispersion Strengthened Fe-Cr-Al Steel (FeCrAl-ODS). There is a reasonable prospect that FeCrAl-ODS is applied to BWRs, but relatively high neutron absorption should be compensated. To decrease adverse neutron economic impact, thin FeCrAl-ODS cladding was designed, and we evaluated characteristics of a core into which 9$$times$$9 Advanced BWR (ABWR) bundles with thin FeCrAl-ODS claddings were loaded. Thin FeCrAl-ODS water rods and channel boxes were also applied. We confirmed that FeCrAl-ODS core reactivity was sufficient by increasing enrichment of UO$$_{2}$$ fuel under the limit of 5 wt%. Moreover, some representative FeCrAl-ODS core characteristics were comparable to zircaloy core. We also confirmed that fuel thermal-mechanical behaviors of thin FeCrAl-ODS cladding at normal operation and transient conditions were acceptable. These results led to a conclusion that FeCrAl-ODS was applicable to BWR in the analysis range of this study.

Journal Articles

Study on oxidation behavior of cladding for accident conditions in spent fuel pool

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo*; Tojo, Masayuki*; Goto, Daisuke*

Fushoku Boshoku Kyokai Dai-62-Kai Zairyo To Kankyo Toronkai Koenshu (CD-ROM), p.23 - 24, 2015/11

In order to clarify the air oxidation behavior of the cladding at high temperatures for study on improvement of safety for accident conditions in spent fuel pool, the oxidation tests for both small specimens under constant temperature conditions and long specimens under loss of coolant simulated temperature conditions were carried out, and the knowledge for influence of both temperature gradient and preoxide film on oxidation behavior of the cladding were obtained in this study.

Journal Articles

Near term test plan using HTTR (High Temperature engineering Test Reactor)

Takada, Shoji; Iigaki, Kazuhiko; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Ono, Masato; Yanagi, Shunki; Nishihara, Tetsuo; Fukaya, Yuji; Goto, Minoru; et al.

Nuclear Engineering and Design, 271, p.472 - 478, 2014/05

 Times Cited Count:5 Percentile:44.02(Nuclear Science & Technology)

JAEA has carried out research and development to establish the technical basis of HTGRs using HTTR. To connect hydrogen production system to HTTR, it is necessary to ensure the reactor dynamics when thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the reactor dynamics stability and to validate plant dynamics codes. It will be confirmed that the reactor become stable state during losing a part of removed heat at heat-sink. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted to confirm the validity of calculated temperature coefficient of reactivity. A LOFC test using HTTR has been carried out to verify the inherent safety under the condition of LOFC while the reactor shut-down system disabled.

JAEA Reports

Temperature coefficient measurement test of HTTR; Burn-up characteristic of temperature coefficients at reactor power 30 kW and 9 MW

Ono, Masato; Goto, Minoru; Shinohara, Masanori; Nojiri, Naoki; Tochio, Daisuke; Shimazaki, Yosuke; Yanagi, Shunki

JAEA-Technology 2013-001, 35 Pages, 2013/03

JAEA-Technology-2013-001.pdf:6.04MB

The temperature coefficient measurements of the HTTR have been carried out. In the beginning of the operation, temperature coefficients at the reactor power of 30 kW and 9 MW were obtained through 1999 to 2000. The operation days of the HTTR fuel reached 375 Effective Full Power Days (EFPD), which is over a half of design operation days (660 EFPD). The temperature coefficient measurements were conducted at the same power levels of 30 kW and 9 MW to evaluate burnup effect. Also, to measure temperature coefficient in high accuracy, technique of core temperature control and technique of core temperature homogenization were established.

Journal Articles

Long-term high-temperature operation of the HTTR

Goto, Minoru; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Hamamoto, Shimpei; Tachibana, Yukio

Nuclear Engineering and Design, 251, p.181 - 190, 2012/10

 Times Cited Count:18 Percentile:82.92(Nuclear Science & Technology)

30-days operation in rated operation mode and 50-days operation in high-temperature operation mode were performed to obtain various characteristic data of HTGR. The main test results are as follows:(1) Coated fuel particle (CFP) of the HTTR has excellent confinement ability of fission product which is the highest performance in the world, (2) The measurement temperature of the core internals is good agreement with the design value so that their structural integrity is maintained, and (3) The intermediate heat exchanger keeps excellent heat transfer performance from beginning of operation. Additionally, the following two issues were validated using the HTTR burnup data. (1) The effectiveness of rod-type burnable poisons on reactivity control in the HTTR, and (2) The whole core burnup calculation method for nuclear characteristics of the HTTR.

Journal Articles

Test plan using HTTR (High Temperature engineering Test Reactor)

Takada, Shoji; Iigaki, Kazuhiko; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Ono, Masato; Nishihara, Tetsuo; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; et al.

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 8 Pages, 2012/10

JAEA has carried out research and development to establish the technical basis of HTGRs using HTTR. LOFC test to verify the inherent safety of HTGR under the condition of loss of forced cooling while the reactor shut-down system disabled was initiated. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations for safety analysis, and changes with burnup because of variance of fuel compositions, which has been measured to confirm the validity of the calculated ones. In order to connect hydrogen production system to HTTR, it is necessary to ensure the reactor safety when thermal-load of the hydrogen production system is lost. Thermal load fluctuation test is planned to demonstrate the reactor safety and gain the test data for validation of the plant dynamics code. It will be confirmed that the reactor become stable state during a part of removed heat at HTTR heat-sink is lost.

Journal Articles

Azimuthal correlations of electrons from heavy-flavor decay with hadrons in $$p+p$$ and Au+Au collisions at $$sqrt{s_{NN}}$$ = 200 GeV

Adare, A.*; Afanasiev, S.*; Aidala, C.*; Ajitanand, N. N.*; Akiba, Yasuyuki*; Al-Bataineh, H.*; Alexander, J.*; Aoki, Kazuya*; Aphecetche, L.*; Aramaki, Y.*; et al.

Physical Review C, 83(4), p.044912_1 - 044912_16, 2011/04

 Times Cited Count:8 Percentile:52.9(Physics, Nuclear)

Measurements of electrons from the decay of open-heavy-flavor mesons have shown that the yields are suppressed in Au+Au collisions compared to expectations from binary-scaled $$p+p$$ collisions. Here we extend these studies to two particle correlations where one particle is an electron from the decay of a heavy flavor meson and the other is a charged hadron from either the decay of the heavy meson or from jet fragmentation. These measurements provide more detailed information about the interaction between heavy quarks and the quark-gluon matter. We find the away-side-jet shape and yield to be modified in Au+Au collisions compared to $$p+p$$ collisions.

JAEA Reports

High temperature continuous operation in the HTTR (HP-11); Summary of the test results in the high temperature operation mode

Takamatsu, Kuniyoshi; Ueta, Shohei; Sumita, Junya; Goto, Minoru; Hamamoto, Shimpei; Tochio, Daisuke; Nakagawa, Shigeaki

JAEA-Research 2010-038, 59 Pages, 2010/11

JAEA-Research-2010-038.pdf:4.6MB

Research and development and hydrogen production technologies by HTGRs will establish future hydrogen energy system. Additionally, the R&D will contribute to innovative hydrogen production technologies by nuclear powers as one of nuclear heat utilizations. We are making an effort to propose a prototype of reactor hydrogen production system until about 2020. Therefore, JAEA is promoting the R&D for confirming the technical basis of HTGRs with the HTTR in the first midterm plan. In 2007, we accomplished the rated power operation for continuous 30 days of the HTTR. In the high temperature operation for continuous 50 days, JAEA evaluated the experimental data such as core burn-up, helium purity control, performance of high temperature equipments, structural integrity in the core, etc. and demonstrated the nuclear thermal availability of heat source for thermo-chemical hydrogen production technology.

Journal Articles

Contribution to improvement of HTGR technology by using HTTR operation data

Nakagawa, Shigeaki; Tochio, Daisuke; Shinohara, Masanori; Nojiri, Naoki; Nishihara, Tetsuo; Goto, Minoru; Takamatsu, Kuniyoshi

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9476_1 - 9476_6, 2009/05

Journal Articles

Hexafluoro complex of rutherfordium in mixed HF/HNO$$_{3}$$ solutions

Toyoshima, Atsushi; Haba, Hiromitsu*; Tsukada, Kazuaki; Asai, Masato; Akiyama, Kazuhiko*; Goto, Shinichi*; Ishii, Yasuo; Nishinaka, Ichiro; Sato, Tetsuya; Nagame, Yuichiro; et al.

Radiochimica Acta, 96(3), p.125 - 134, 2008/03

 Times Cited Count:27 Percentile:86.79(Chemistry, Inorganic & Nuclear)

Formation of an anionic fluoride-complex of element 104, rutherfordium (Rf) produced in the $$^{248}$$Cm($$^{18}$$O,5n)$$^{261}$$Rf reaction was studied by an anion-exchange method based on an atom-at-a-time scale. It was found that the hexafluoro complex of Rf, [RfF$$_{6}$$]$$^{2-}$$, was formed in the studied fluoride ion concentrations of 0.0005 - 0.013 M. Formation of [RfF$$_{6}$$]$$^{2-}$$ was significantly different from that of the homologues Zr and Hf, [ZrF$$_{6}$$]$$^{2-}$$ and [HfF$$_{6}$$]$$^{2-}$$; the evaluated formation constant of [RfF$$_{6}$$]$$^{2-}$$ is at least one-order of magnitude smaller than those of [ZrF$$_{6}$$]$$^{2-}$$ and [HfF$$_{6}$$]$$^{2-}$$.

Journal Articles

Improvement of analysis technology for high temperature gas-cooled reactor by using data obtained in high temperature engineering test reactor

Nakagawa, Shigeaki; Tochio, Daisuke; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki

Journal of Power and Energy Systems (Internet), 2(1), p.83 - 91, 2008/00

The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30 MW at a reactor outlet coolant temperature of about 950 $$^{circ}$$C in April, 2004 during the "rise-to-power tests" confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with the high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.

JAEA Reports

Investigation of the loss of forced cooling test by using the High Temperature Engineering Test Reactor (HTTR) (Contract research)

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Tochio, Daisuke; Inaba, Yoshitomo; Goto, Minoru

JAEA-Technology 2007-056, 51 Pages, 2007/09

JAEA-Technology-2007-056.pdf:5.49MB

The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the HTTR are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the core residual heat follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur.

Journal Articles

R&D on safeguards environmental sample analysis at JAERI

Sakurai, Satoshi; Magara, Masaaki; Usuda, Shigekazu; Watanabe, Kazuo; Esaka, Fumitaka; Hirayama, Fumio; Lee, C. G.; Yasuda, Kenichiro; Kono, Nobuaki; Inagawa, Jun; et al.

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

no abstracts in English

Journal Articles

Development of analytical techniques for safeguards environmental samples; Bulk analysis

Hirayama, Fumio; Kurosawa, Setsumi; Magara, Masaaki; Ichimura, Seiji; Kono, Nobuaki; Suzuki, Daisuke; Inagawa, Jun; Goto, Mototsugu; Sakurai, Satoshi; Watanabe, Kazuo; et al.

KEK Proceedings 2005-4, p.184 - 192, 2005/08

no abstracts in English

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