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Oral presentation

A Tritium removal for the decommissioning of FUGEN, 1; Applicability of the tritium removal technique for the actual equipment and pipe

Ando, Koji; Kadowaki, Haruhiko; Matsuo, Hidehiko; Yamane, Naoki; Matsushima, Akira

no journal, , 

no abstracts in English

Oral presentation

Estimation of inventory in the Fukushima-Daiichi Nuclear Power Plant

Nishihara, Kenji; Iwamoto, Hiroki; Suyama, Kenya

no journal, , 

no abstracts in English

Oral presentation

Development of nuclear materials assay system using alternative He-3 neutron detectors

Kureta, Masatoshi; Soyama, Kazuhiko; Nakamura, Hironobu; Seya, Michio; Ozu, Akira; Nakamura, Tatsuya; Haruyama, Mitsuo

no journal, , 

Alternative techniques to neutron detection by He-3 for safeguards and nuclear security systems are necessary to be developed since He-3 gas shortage is serious. With support of Japanese government (MEXT), we have conducted an R&D project of ZnS ceramic scintillator neutron detectors for non-destructive assay (NDA) of Pu in Fresh MOX fuel and other types of nuclear materials for a use of safeguards verification. Here we present the fundamental tests and the current status of the development of the Pu-NDA system for the demonstration.

Oral presentation

Development of pulsed neutron two and three-dimentional imaging system using high-speed video camera at J-PARC

Segawa, Mariko; Oi, Motoki; Kai, Tetsuya; Shinohara, Takenao; Kureta, Masatoshi; Sato, Hirotaka*

no journal, , 

no abstracts in English

Oral presentation

Development of numerical simulation method for relocation behavior of melting fuel in nuclear reactors, 1; Preliminary analysis of relocation of melting fuel to lower plenum

Yamashita, Susumu; Yoshida, Hiroyuki; Takase, Kazuyuki

no journal, , 

In the Fukushima accidents, fuel assemblies which were installed in the reactors were reached a high temperature by stop of the core cooling system with a power station black-out. As a result, it is considered that the core degradation has been introduced because the melting of the fuel rods occurred and the melting behavior was expanded. In order to elucidate a progress of the melting phenomena in the reactor core, a numerical simulation code which can be precisely evaluated the melting phenomena is required. Then, the melting behavior of the fuel assemblies in the reactor core was analyzed numerically using the three-dimensional multi-phase thermal-hydraulic simulation code. In the presentation, the numerical results on the melting behavior of the simulated fuel assemblies and reactor structures are reported for several different initial conditions.

Oral presentation

Alternative He-3 neutron detectors using solid scintillator for nuclear safeguards, 2; Optical guide property for scintillator modular detector

Ozu, Akira; Nakamura, Tatsuya; Takase, Misao; Kurata, Noritaka; Haruyama, Mitsuo; Soyama, Kazuhiko; Kureta, Masatoshi; Seya, Michio

no journal, , 

no abstracts in English

Oral presentation

Oral presentation

Development of numerical simulation for jet breakup behavior in complicated structure of BWR lower plenum, 2; Observation of jet breakup behavior with visualized experimental apparatus

Saito, Ryusuke*; Abe, Yutaka*; Kaneko, Akiko*; Suzuki, Takayuki; Yoshida, Hiroyuki; Nagase, Fumihisa

no journal, , 

no abstracts in English

Oral presentation

Tritium removal of FUGEN using a compact portable dehumidifier system of hollow fiber membrane separation, 2; Verification test of applicability

Kadowaki, Haruhiko; Matsuo, Hidehiko; Yamane, Naoki; Asakura, Yamato*; Matsushima, Akira

no journal, , 

no abstracts in English

Oral presentation

Development of prediction technology of two-phase flow dynamics under earthquake acceleration, 15; Numerical prediction of bubble behavior in simulated sub-channel under accelerating condition

Yoshida, Hiroyuki; Nagatake, Taku; Takase, Kazuyuki; Kaneko, Akiko*; Monji, Hideaki*; Abe, Yutaka*

no journal, , 

no abstracts in English

Oral presentation

Evaluation of covariance for JENDL-4.0, 1; Pb isotopes

Iwamoto, Osamu

no journal, , 

Covariance data have been evaluated for nuclear data of Pb isotopes in JENDL-4.0. The evaluation has been done for the cross sections such as total, inelastic-scattering, and (n,2n) reactions for the Pb isotopes of $$^{204}$$Pb, $$^{206}$$Pb, $$^{207}$$Pb and $$^{208}$$Pb in continuum energy region using CCONE-KALMAN code system. Model parameter covariance were deduced from uncertainties of cross sections estimated from experimental data using sensitivities of the parameters for optical model, level density and $$gamma$$-ray strength functions, etc. Cross correlations between reactions were also deduced from the unique parameter covariance for each isotope. Covariance of Legendre coefficients for elastic scattering angular distribution was evaluated. The evaluation methods and results are presented.

Oral presentation

Experimental evaluations of seawater effects on thermal-hydraulic behavior at severe accident, 1; Outline of research plan

Liu, W.; Nagatake, Taku; Takase, Kazuyuki; Yoshida, Hiroyuki; Nagase, Fumihisa

no journal, , 

Severe accidents occurred in the reactors and spent fuel pools of the Fukushima No.1 nuclear power plant after the East Japan Earthquake and the large tsunami occurred on Mar.11, 2011. To cool down the reactors, seawater was injected into the cores. So far, core cooling with seawater has never been assumed and the effect of seawater on heat transfer in reactor core is not clear. To contribute to the understanding of the situations of the reactor cores, Japan Atomic Energy Agency has started the experimental research on the effect of seawater on thermal-hydraulic behaviors including heat transfer. Multiple experiments are to be performed in order to understand the phenomena of the thermal-hydraulic behaviors including the effect of salt precipitation, identify controlling factors, etc. This paper reports a summary of research plan and experimental apparatus to be manufactured.

Oral presentation

Oral presentation

Numerical analysis on heat transfer characteristics of supercritical pressure water in a heated tube based on three dimensional two-fluid model

Ose, Yasuo*; Suzuki, Takayuki; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

no journal, , 

no abstracts in English

Oral presentation

Development of analytical method for behavior of fuel melting by particle method, 1; Outline of the research plan

Nagatake, Taku; Takase, Kazuyuki; Furuya, Masahiro*; Yoshida, Hiroyuki; Nagase, Fumihisa

no journal, , 

no abstracts in English

Oral presentation

Numerical analysis of hydraulic behavior in venturi scrubber by TPFIT

Horiguchi, Naoki*; Abe, Yutaka*; Yoshida, Hiroyuki; Kaneko, Akiko*; Uesawa, Shinichiro*

no journal, , 

no abstracts in English

Oral presentation

Development of the neutron resonance densitometry using a pulsed neutron source, 2; Design study of a D-T pulsed neutron source for nuclear material quantification in melted fuel

Takamine, Jun; Kureta, Masatoshi; Harada, Hideo; Kitatani, Fumito; Koizumi, Mitsuo; Tsuchiya, Harufumi; Iimura, Hideki

no journal, , 

In the process of dismantlement and reclamation of Fukushima Daiichi Nuclear Power Plant, accounting for and control of nuclear material (NM) in the melted fuel (MF) may be demanded. In Japan Atomic Energy Agency (JAEA), The Neutron Resonance Densitometry (NRD) is proposed. In this method, continuous-energy neutrons are transmitted through NM in MF. And quantity of NM in MF is fixed using neutron resonance absorption peaks derived from measurement results. The prototype system using a compacted D-T fusion pulsed neutron source will be produced in next fiscal year. For getting enough Fixed-quantity precision using this system, it is important to decide arrangement, materials and shapes as high intensity and time resolution as we can. About several kinds of moderator system assumed to be suited for the system, Neutron spectrums and time dependent characteristics were evaluated using MCNP5. Knowledge about the configuration parameters derived from these results will be showed.

Oral presentation

Mixing factor for sub-channel analysis of wire-wrapped large diameter fuel pin bundle of sodium-cooled fast reactors; Effects of wire-wrap lead length

Okano, Yasushi; Doda, Norihiro; Ohshima, Hiroyuki; Okubo, Tsutomu

no journal, , 

no abstracts in English

Oral presentation

Alternative He-3 neutron detectors using solid scintillator for nuclear safeguard, 3; Multi-channel ADC / DSP board for scintillator modular detector

Ebine, Masumi; Birumachi, Atsushi; Nakamura, Tatsuya; Ozu, Akira; Takase, Misao; To, Kentaro; Sakasai, Kaoru; Soyama, Kazuhiko; Kureta, Masatoshi; Seya, Michio

no journal, , 

no abstracts in English

256 (Records 1-20 displayed on this page)