Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 5854

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

Anisotropic thermal lattice expansion and crystallographic structure of strontium aluminide within Al-10Sr alloy as measured by in-situ neutron diffraction

Liss, K.-D.*; Harjo, S.; Kawasaki, Takuro; Aizawa, Kazuya; Xu, P. G.

Journal of Alloys and Compounds, 869, p.159232_1 - 159232_9, 2021/07

Journal Articles

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2; Results of severe accident analyses for unit 3

Lind, T.*; Pellegrini, M.*; Herranz, L. E.*; Sonnenkalb, M.*; Nishi, Yoshihisa*; Tamaki, Hitoshi; Cousin, F.*; Fernandez Moguel, L.*; Andrews, N.*; Sevon, T.*

Nuclear Engineering and Design, 376, p.111138_1 - 111138_12, 2021/05

This is the third part of the three part paper describing the accidents at the FDNPS as analyzed in the Phase 2 of the OECD/NEA project "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant" (BSAF). In this paper, we describe the accident progression in unit 3. In the BSAF project, eight organizations from five countries analyzed severe accident scenarios for Unit 3 at the Fukushima Daiichi site using different severe accident codes. The present paper for Unit 3 describes the findings of the comparison of the participants' results against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on the status of the reactor pressure vessel, melt release and fission product release and transport. Unit 3 specific aspects, e.g., the complicated accident progression following repeated containment venting actuations and attempts at coolant injection at the time of the major core degradation, are highlighted and points of consensus as well as remaining uncertainties and data needs will be summarized. FP transport is analyzed, and the calculation results are compared with dose rate measurements in the containment. The release of I-131 and Cs-137 to the environment is compared with analysis conducted by using WSPEEDI code.

Journal Articles

Evaluation of high-energy delayed gamma-ray spectra dependence on interrogation timing patterns

Rodriguez, D.; Bogucarska, T.*; Koizumi, Mitsuo; Lee, H.-J.; Pedersen, B.*; Rossi, F.; Takahashi, Toon; Varasano, G.*

Nuclear Instruments and Methods in Physics Research A, 997, p.165146_1 - 165146_13, 2021/05

 Times Cited Count:0 Percentile:100(Instruments & Instrumentation)

JAEA Reports

Data report of ROSA/LSTF experiment SB-PV-09; 1.9% pressure vessel top small break LOCA with SG depressurization and gas inflow

Takeda, Takeshi

JAEA-Data/Code 2021-006, 61 Pages, 2021/04

JAEA-Data-Code-2021-006.pdf:2.78MB

An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.

Journal Articles

Microstructural stability of ODS steel after very long-term creep test

Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki

Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700$$^{circ}$$C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.

Journal Articles

Application of probabilistic fracture mechanics to reactor pressure vessel using PASCAL4 code

Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*

Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04

 Times Cited Count:0

Journal Articles

Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors

Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03

Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by $$^{6}$$Li(n$$_{t}$$,$$alpha$$)$$^{3}$$H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.

Journal Articles

Iodine-129 in the Tokai Reprocessing Plant and the environment

Nakano, Masanao

Hoken Butsuri (Internet), 56(1), p.17 - 25, 2021/03

The Tokai Reprocessing Plant is the first reprocessing plant in Japan which started hot test in 1977, and had reprocessed 1140 tons of spent nuclear fuel by May 2007. The gaseous and liquid radioactive wastes has been discharged to the environment. Since iodine-129 ($$^{129}$$I) is one of the most important nuclides for environmental impact assessment. Therefore, $$^{129}$$I in the exhaust and effluent has been controlled, and a precise analysis method of $$^{129}$$I in the environmental samples was developed, and the concentration of 129I in the environment was investigated. This report presents an overview of these activities. Not limited to $$^{129}$$I on reprocessing facilities, it is essential for nuclear operators to reduce the amount released to the environment in the spirit of ALARA, and to continuously develop the further upgrading environmental monitoring methods and evaluation methods in order to foster a sense of safety and security among residents living in the vicinity of the facilities.

Journal Articles

Visualization of the boron distribution in core material melting and relocation specimen by neutron energy resolving method

Abe, Yuta; Tsuchikawa, Yusuke; Kai, Tetsuya; Matsumoto, Yoshihiro*; Parker, J. D.*; Shinohara, Takenao; Oishi, Yuji*; Kamiyama, Takashi*; Nagae, Yuji; Sato, Ikken

JPS Conference Proceedings (Internet), 33, p.011075_1 - 011075_6, 2021/03

Journal Articles

Study of neutron-nuclear spin correlation term with a polarized Xe target

Sakai, Kenji; Oku, Takayuki; Okudaira, Takuya; Kai, Tetsuya; Harada, Masahide; Hiroi, Kosuke; Hayashida, Hirotoshi*; Kakurai, Kazuhisa*; Shimizu, Hirohiko*; Hirota, Katsuya*; et al.

JPS Conference Proceedings (Internet), 33, p.011116_1 - 011116_6, 2021/03

In neutron fundamental physics, study of correlation term $${bf s}cdot{bf I}$$ of a neutron spin $${bf s}$$ and a target nuclear spin $${bf I}$$ is important because $${bf s}cdot{bf I}$$ term interferes to parity non-conserving (PNC) and time reversal non-conserving terms. For this study, a xenon (Xe) is an interesting nucleus because it has been observed an enhancement of PNC effect around neutron resonance peaks, and polarizes up to $$ sim 10^{-1}$$ by using a spin exchange optical pumping (SEOP) method. We would plan to develop a polarized Xe gas target with a compact in-situ SEOP system, and to study $${bf s}cdot{bf I}$$ term by utilizing epithermal neutron beams supplied from a high intense pulsed spallation neutron source. As the first step, we attempted to measure neutron polarizing ability caused by $${bf s}cdot{bf I}$$ term at a 9.6 eV s-wave resonance peak of $$^{129}$$Xe at BL10 in MLF, by detecting change $$Delta R$$ of ratio between neutron transmissions with the polarized and unpolarized Xe target. After demonstrating that our apparatus could detect small change ($$Delta R_{rm DB} , {approx},10^{-2}$$) of neutron transmissions caused by Doppler broadening effect, a signified value of $$Delta R$$ has been obtained as preliminary results. For analyzing the obtained $$Delta R$$ in detail, we are improving our nuclear magnetic resonance and electron paramagnetic resonance systems for evaluating Xe polarization independently of neutron beams.

Journal Articles

Thermal-neutron capture cross sections and resonance integrals of the $$^{243}$$Am(n,$$gamma$$)$$^{rm 244g}$$Am and $$^{243}$$Am(n,$$gamma$$)$$^{rm 244m+g}$$Am reactions

Nakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Kimura, Atsushi

Journal of Nuclear Science and Technology, 58(3), p.259 - 277, 2021/03

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

Research and development were made for accuracy improvement of neutron capture cross section data on $$^{243}$$Am among minor actinides. First, the emission probabilities of decay $$gamma$$ rays were obtained with high accuracy, and the amount of the ground state of $$^{244}$$Am produced by reactor neutron irradiation of $$^{243}$$Am was examined by $$gamma$$-ray measurement. Next, the total amount of isomer and ground states was examined by $$alpha$$-ray measurement. Thermal-neutron capture cross sections and resonance integrals were derived both for the $$^{243}$$Am(n,$$gamma$$)$$^{rm 244g}$$Am and for $$^{243}$$Am(n,$$gamma$$)$$^{rm 244m+g}$$Am reactions.

Journal Articles

Effect of gas microbubble injection and narrow channel structure on cavitation damage in mercury target vessel

Naoe, Takashi; Kinoshita, Hidetaka; Kogawa, Hiroyuki; Wakui, Takashi; Wakai, Eiichi; Haga, Katsuhiro; Takada, Hiroshi

Materials Science Forum, 1024, p.111 - 120, 2021/03

The mercury target vessel for the at the J-PARC neutron source is severely damaged by the cavitation caused by proton beam-induced pressure waves in mercury. To mitigate the cavitation damage, we adopted a double-walled structure with a narrow channel for the mercury at the beam window of the vessel. In addition, gas microbubbles were injected into the mercury to suppress the pressure waves. The front end of the vessel was cut out to inspect the effect of the damage mitigation technologies on the interior surface. The results showed that the double-walled target facing the mercury with gas microbubbles operating at 1812 MWh for an average power of 434 kW had equivalent damage to the single-walled target without microbubbles operating 1048 MWh for average power of 181 kW. The erosion depth due to cavitation in the narrow channel was clearly smaller than it was on the wall facing the bubbling mercury

Journal Articles

New design of high power mercury target vessel of J-PARC

Wakui, Takashi; Wakai, Eiichi; Kogawa, Hiroyuki; Naoe, Takashi; Hanano, Kohei*; Haga, Katsuhiro; Shimada, Tsubasa*; Kanomata, Kenichi*

Materials Science Forum, 1024, p.145 - 150, 2021/03

To realize a high beam power operation at the J-PARC, a mercury target vessel covered with water shroud was developed. In the first step, to realize an operation at 500 kW, the basic structure of the initial design was followed and the connection method between the mercury vessel and the water shroud was changed. Additionally, the operation at a beam power of 500 kW was realized in approximately eight months. In the second step, to realize the operation at 1 MW, the new structure in which only rear ends of vessels were connected was investigated. Cooling of the mercury vessel is used to reduce thermal stress and thick vessels of the water shroud are used to increase stiffness for the internal pressure; therefore, it was adopted. The stress in each vessel was lower than the allowable stress based on the pressure vessel code criteria prescribed in the Japan Industrial Standard, and confirmation was obtained that the operation with a beam power of 1 MW could be conducted.

Journal Articles

Pressure-dependent structure of methanol-water mixtures up to 1.2 GPa; Neutron diffraction experiments and molecular dynamics simulations

Temleitner, L.*; Hattori, Takanori; Abe, Jun*; Nakajima, Yoichi*; Pusztai, L.*

Molecules (Internet), 26(5), p.1218_1 - 1218_12, 2021/03

 Times Cited Count:0

Total structure factors of per-deuterated methanol and heavy water, CD$$_{3}$$OD and D$$_{2}$$O, have been determined across the entire composition range at pressures of up to 1.2 GPa, by neutron diffraction. Largest variations due to increasing pressure were observed below $$Q=$$ 5 $AA$^{-1}$$, mostly as shifts of the first and second maxima. Molecular dynamics computer simulations been conducted at the experimental pressures to interpret neutron diffraction results. The peak shifts mentioned above could be qualitatively reproduced by simulations. In order to reveal the influence of changing pressure on the local intermolecular structure, simulated structures have been analyzed in terms of hydrogen bond related partial radial distribution functions and size distributions of hydrogen bonded cyclic entities. Distinct differences between pressure dependent structures of water-rich and methanol-rich composition regions have been revealed.

Journal Articles

Determination of parameters for an equation to obtain natural background radiation using KURAMA-II loaded with C12137-01 type CsI(Tl) detector

Ando, Masaki; Matsuda, Norihiro; Saito, Kimiaki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 20(1), p.34 - 39, 2021/03

We measured count rates and air dose rates at 11 measurement points where the influence of the Fukushima Dai-ichi nuclear power plant accident could be ignored to obtain parameters for a background equation applying to KURAMA-II loaded with the high sensitivity CsI(Tl) detector, C12137-01. It was found that the sensitivity of KURAMA-II (C12137-01) was about 10 times or more for background measurement, compared with KURAMA-II loaded with the standard type CsI(Tl) detector, C12137. A background equation for the energy range of 1400-2000 keV was determined as, y ($$mu$$Sv/h)=0.062 x (cps). We evaluated background air dose rates using KURAMA-II (C12137-01) for 71 municipalities and compared them with the previous study using KURAMA-II (C12137). Evaluated background air dose rates in this study were almost equal to those in the previous study. We confirmed that the background equation evaluated in this study was applicable for the KURAMA-II (C12137-01).

Journal Articles

Single-crystal growth and magnetic phase diagram of the enantiopure crystal of NdPt$$_2$$B

Sato, Yoshiki*; Honda, Fuminori*; Maurya, A.*; Shimizu, Yusei*; Nakamura, Ai*; Homma, Yoshiya*; Li, D.*; Haga, Yoshinori; Aoki, Dai*

Physical Review Materials (Internet), 5(3), p.034411_1 - 034411_9, 2021/03

JAEA Reports

Background radiation monitoring using manned helicopter for application of technique of nuclear emergency response in the fiscal year 2019 (Contract research)

Futemma, Akira; Sanada, Yukihisa; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Sato, Kazuhiko*; Haginoya, Masashi*; Matsunaga, Yuki*; Kikuchi, Hikaru*; Ishizaki, Azusa; et al.

JAEA-Technology 2020-019, 128 Pages, 2021/02

JAEA-Technology-2020-019.pdf:15.75MB

A large amount of radioactive material was released by the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company, caused by the Great East Japan Earthquake and the following tsunami on March 11, 2011. After the nuclear disaster, airborne radiation monitoring using manned helicopter has been utilized to grasp rapidly and widely the distribution of the radioactive materials around FDNPS. We prepare the data of background radiation dose, geomorphic characteristics and the controlled airspace around nuclear facilities of the whole country in order to make effective use of the monitoring technique as a way of emergency radiation monitoring and supply the results during accidents of the facilities. Furthermore, the airborne radiation monitoring has been conducted in Integrated Nuclear Emergency Response Drill to increase effectiveness of the monitoring. This report is summarized that the knowledge as noted above achieved by the aerial radiation monitoring around Higashidori nuclear power station, the nuclear fuel reprocessing plant in Rokkasho village and Shika nuclear power station, the full details of the aerial radiation monitoring in Integrated Nuclear Emergency Response Drill in the fiscal 2019. In addition, examination's progress aimed at introduction of airborne radiation monitoring using unmanned helicopter during nuclear disaster and the technical issues are summarized in this report.

JAEA Reports

Outline of Regional Workshops held in 2006 - 2017 by the International Atomic Energy Agency in the proposal of Nuclear Emergency Preparedness Group of the Asian Nuclear Safety Network

Okuno, Hiroshi; Yamamoto, Kazuya

JAEA-Review 2020-066, 32 Pages, 2021/02

JAEA-Review-2020-066.pdf:3.01MB

The International Atomic Energy Agency (abbreviated as IAEA) has been implementing the Asian Nuclear Safety Network (abbreviated as ANSN) activities since 2002. As part of this effort, Topical Group on Emergency Preparedness and Response (abbreviated as EPRTG) for nuclear or radiation disasters was established in 2006 under the umbrella of the ANSN. Based on the EPRTG proposal, the IAEA conducted 23 Asian regional workshops in the 12 years from 2006 to 2017. Typical topical fields of the regional workshops were nuclear emergency drills, emergency medical care, long-term response after nuclear/radiological emergency, international cooperation, national nuclear disaster prevention system. The Japan Atomic Energy Agency has produced coordinators for EPRTG since its establishment and has led its activities since then. This report summarizes the Asian regional workshops conducted by the IAEA based on the recommendations of the EPRTG.

JAEA Reports

Research and development of radiation-resistant sensor for fuel debris by integrating advanced measurement technologies (Contract research); FY2019 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; High Energy Accelerator Research Organization*

JAEA-Review 2020-058, 101 Pages, 2021/02

JAEA-Review-2020-058.pdf:5.58MB

JAEA/CLADS had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project in FY2019. Among the adopted proposals in FY2018, this report summarizes the research results of the "Research and Development of Radiation-resistant Sensor for Fuel Debris by Integrating Advanced Measurement Technologies" conducted in FY2019.

JAEA Reports

Summaries of research and development activities by using supercomputer system of JAEA in FY2019 (April 1, 2019 - March 31, 2020)

HPC Technology Promotion Office

JAEA-Review 2020-021, 215 Pages, 2021/02

JAEA-Review-2020-021.pdf:13.11MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. As shown in the fact that about 20 percent of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2019, the system was used for R&D aiming to restore Fukushima (environmental recovery and nuclear installation decommissioning) as a priority issue, as well as for JAEA's major projects such as research and development of fast reactor cycle technology, research for safety improvement in the field of nuclear energy, and basic nuclear science and engineering research. This report presents a great number of R&D results accomplished by using the system in FY2019, as well as user support, operational records and overviews of the system, and so on.

5854 (Records 1-20 displayed on this page)