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JAEA Reports

Radiation monitoring using manned helicopter around the Nuclear Power Station in the fiscal year 2015 (Contract research)

Sanada, Yukihisa; Munakata, Masahiro; Mori, Airi; Ishizaki, Azusa; Shimada, Kazumasa; Hirouchi, Jun; Nishizawa, Yukiyasu; Urabe, Yoshimi; Nakanishi, Chika*; Yamada, Tsutomu*; et al.

JAEA-Research 2016-016, 131 Pages, 2016/10


By the nuclear disaster of Fukushima Daiichi Nuclear Power Station (FDNPS), Tokyo Electric Power Company (TEPCO), caused by the East Japan earthquake and the following tsunami occurred on March 11, 2011, a large amount of radioactive materials was released from the NPS. After the nuclear disaster, airborne radiation monitoring using manned helicopter was conducted around FDNPS. In addition, background dose rate monitoring was conducted around Sendai Nuclear Power Station. These results of the aerial radiation monitoring using the manned helicopter in the fiscal 2015 were summarized in the report.

Journal Articles

Improvement of slow purging and slow pumping system on the change exchange system in the J-PARC RCS

Tobita, Norimitsu; Yoshimoto, Masahiro; Takeda, Osamu; Saeki, Riuji; Yamazaki, Yoshio; Kinsho, Michikazu; Muto, Masayoshi*

Proceedings of 12th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1350 - 1354, 2015/09

no abstracts in English

Journal Articles

Retrievement of the charge stripping foil in J-PARC RCS

Tobita, Norimitsu; Yoshimoto, Masahiro; Yamazaki, Yoshio; Saeki, Riuji; Okabe, Kota; Kinsho, Michikazu; Takeda, Osamu*; Muto, Masayoshi*

Proceedings of 10th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.915 - 919, 2014/06

The charge conversion foil used with a J-PARC 3GeV synchrotron (RCS: Rapid Cycling Synchrotron) is a thin film made from carbon about 1 micrometer thick, and it radioactivates it by continuing being irradiated with a beam. Moreover, generally it is thought that degradation progresses and foil itself breaks easily. However, when dealing with the foil after irradiation, the measure against the danger of the contamination and the contamination in the living body by foil dispersing is one of the subjects. So, in RCS, the foil exchange booth for collecting the radioactivated foil safely and certainly was installed. Even when dispersing foil temporarily, the radioactivated foil can be shut up only in Booth and a worker's contamination and contamination of work area could be prevented. Moreover, when it sees from a viewpoint of the performance gain of foil, analysis and observation of the collected foil are one of the important issues. Then, in order to observe the radioactivated foil after beam irradiation, the transparent protective case which can be sealed with a foil frame simple substance was developed. In this announcement, the equipment developed in order to collect the charge conversion foil after beam irradiation, and the established technique are announced in detail.

Journal Articles

Preperation of the charge stripping foil in J-PARC RCS

Saeki, Riuji; Yoshimoto, Masahiro; Yamazaki, Yoshio; Tobita, Norimitsu; Okabe, Kota; Kinsho, Michikazu; Takeda, Osamu*; Muto, Masayoshi*

Proceedings of 10th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.523 - 526, 2014/06

RCS has loaded with the foil of 15 sheets including a reserve into equipment so that it can exchange in a short time, when foil is damaged also in a beam operating period. It is difficult for foil to be made of a thin film about 1 micrometer thick, and to treat as it is. Then, foil is fixed to the frame which stuck the SiC fiber, and foil is not touched, but only a frame is held, and it enabled it to operate it. The following preparations are needed as new foil exchange work. (1) Exfoliation and recovery of foil which have been vapor-deposited to glass substrate. (2) Dryness and logging of exfoliative foil. (3) Preparation of SiC wire, and attachment on frame. (4) Fix foil to a frame. (5) Although charge on a magazine-rack was performed manually altogether until now, there were many work man days, and the quality of the prepared foil had variation. Then, equipment required in order to secure reproducibility was developed. The technique for working efficiently simultaneously was established. In this announcement, the technique established until now and the developed jigs are announced in detail.

Journal Articles

Investigation of advanced divertor magnetic configuration for DEMO tokamak reactor

Asakura, Nobuyuki; Shinya, Kichiro*; Tobita, Kenji; Hoshino, Kazuo; Shimizu, Katsuhiro; Uto, Hiroyasu; Someya, Yoji; Nakamura, Makoto; Ono, Noriyasu*; Kobayashi, Masahiro*; et al.

Fusion Science and Technology, 63(1T), p.70 - 75, 2013/05

no abstracts in English

Journal Articles

Investigation of advanced divertor magnetic configuration for Demo tokamak reactor

Asakura, Nobuyuki; Shinya, Kichiro*; Tobita, Kenji; Hoshino, Kazuo; Shimizu, Katsuhiro; Uto, Hiroyasu; Someya, Yoji; Nakamura, Makoto; Ono, Noriyasu*; Kobayashi, Masahiro*; et al.

Fusion Science and Technology, 63(1T), p.70 - 75, 2013/05

 Times Cited Count:12 Percentile:71.51(Nuclear Science & Technology)

Design study of poloidal field coil (PFC) locations and current distribution for the advanced divertor in the Demo tokamak reactor was presented. Concept of the super-X divertor (SXD) for Demo reactor has an outer divertor leg longer than the conventional divertor, and it extends outboard to increase both the target wetted area and connection length to the outer target ($$L_{//}$$). Equilibrium calculation code, TOSCA, was developed by introducing two parameters, i.e. super-X null radius ($$R_{SX}$$) and a ratio of the poloidal flux at the super-X null to that at the separatrix ($$f_{SX}$$). Some SXD magnetic configurations with minimal number of PFCs located outside toroidal field coil (TFC) were Demonstrated. Locations of the divertor target were also investigated. It was found that the flux expansion can be increased up to 4-10 depending on the target location and $$f_{SX}$$, and that SXD has an advantage to increase $$L_{//}$$ with $$f_{SX}$$. Thus, the divertor plasma temperature is expected to decrease at the same upstream plasma density. On the other hand, large currents for the divertor PFCs were necessary. Other arrangements of PFCs such as (1) larger $$R_{SX}$$ and (2) inside TFC, can reduce the PFC currents.

JAEA Reports

Consideration on the earthquake-resistant margin of equipment base-bolts receiving the shearing load during seismic movements

Sakaguchi, Shinobu; Tachibana, Ikuya; Koshino, Katsuhiko; Shirozu, Hidetomo; Shirai, Nobutoshi; Imamoto, Nobuo; Tomita, Tsuneo; Tobita, Hiroo; Yamanaka, Atsushi; Kobayashi, Daisuke; et al.

JAEA-Technology 2011-006, 24 Pages, 2011/03


In the Niigataken Chuetsu-oki Earthquake on 2007, observed earthquake motion exceeded design base at the Kashiwazaki Kariwa nuclear plant. However, there was no earthquake damage in safety important equipments to stop reactor, to cool reactor, and to contain radioactive materials. One of this reason is said that many safety margin are included in the design and the permissible value. To know more accurate safety margin, shearing force examinations for the base bolts were conducted. In examinations, delegate test-bolts were made; the test bolts were selected from heavier equipment in Tokai Reprocessing Plant. In this report, the shearing strength obtained from the examinations shows more accurate safety margin.

JAEA Reports

Numerical analysis on thermal-hydraulic behavior in natural convection capsules

Inaba, Yoshitomo; Ogawa, Mitsuhiro; Yamaura, Takayuki; Tobita, Masahiro

JAEA-Technology 2009-032, 51 Pages, 2009/07


The fuel transient tests for light water reactors are to be carried out in the Japan Materials Testing Reactor (JMTR), and the capsule-type test facilities (fuel transient test capsules) are to be used in the tests. In order to investigate the thermal-hydraulic behavior in the capsules, the multi-dimensional two-fluid model code ACE-3D is used. At first, the functions of ACE-3D were expanded for the pre-process and the post-process. Then, the BWR power calibration test capsule, which had been tested in JMTR, was modeled, and the BWR power calibration tests were simulated numerically for the verification of ACE-3D. The numerical results agreed well with the test data. As a result, it was found that ACE-3D is applicable to the numerical simulation of the fuel transient tests. In addition, the fuel transient tests with a natural convection capsule were simulated numerically with ACE-3D, and the thermal-hydraulic behavior in the capsule was investigated.

JAEA Reports

Dose evaluation for fuel transient test

Iimura, Koichi; Ogawa, Mitsuhiro; Tomita, Kenji; Tobita, Masahiro

JAEA-Technology 2009-021, 71 Pages, 2009/05


The preparation of a fuel transient test using the JMTR is advanced to conduct its irradiation test from 2011 F.Y. after re-operation of the JMTR. The fuel behavior for high burn-up BWR's under power ramping condition will be evaluated in simulating the BWR environmental condition using the shroud irradiation facility (Oarai Shroud Facility No.1) and $$^{3}$$He power-control type BOCA (Boiling Water Capsule) irradiation facility, which is composed of the capsule control device, $$^{3}$$He power-control device and boiling water capsule. In order to change the fuel irradiation conditions so as to treat high burn-up fuels (from 50 GWD/t-UO$$_{2}$$ to 110 GWD/t-U), it is necessary to revaluate the dose for the safety evaluation at the test fuel failure. In this report, evaluations for equivalent dose rate of each device and exposure dose of handling operators when all fission products released in the coolant of the capsule control device and the BOCA at fuel failure in the fuel transient test are summarized.

JAEA Reports

Dose evaluation of external exposure by direct and skyshine gamma rays of nuclear fuel handling facilities at JMTR

Ogawa, Mitsuhiro; Iimura, Koichi; Tomita, Kenji; Tobita, Masahiro

JAEA-Technology 2009-017, 254 Pages, 2009/05


In JMTR, upgrade of irradiation facilities is advanced to re-operate from 2011 F.Y. In order to irradiate test fuels of high-burnup, external exposure reassessment by direct and skyshine gamma rays of the nuclear fuel handling facility at JMTR was performed. In evaluation method, radiation source of maximum use of the nuclear fuel was calculated by using ORIGEN2 code. Dose equivalent rate for supervised area boundary was calculated by modeling reactor building at using shielding calculation codes QAD-CGGP2 and G33-GP2. As a result of evaluation, it was confirmed that the effective dose equivalent during year was low enough at supervised area boundary of the JMTR site.

Journal Articles

Research and development of nuclear fusion

Ushigusa, Kenkichi; Seki, Masahiro; Ninomiya, Hiromasa; Norimatsu, Takayoshi*; Kamada, Yutaka; Mori, Masahiro; Okuno, Kiyoshi; Shibanuma, Kiyoshi; Inoue, Takashi; Sakamoto, Keishi; et al.

Genshiryoku Handobukku, p.906 - 1029, 2007/11

no abstracts in English

Journal Articles

Intermittent $$beta$$ collapse after NBCD turn-off in JT-60U fully non-inductive reversed shear discharges

Takei, Nahoko; Nakamura, Yukiharu; Ushigome, Masahiro*; Suzuki, Takahiro; Aiba, Nobuyuki; Takechi, Manabu; Tobita, Kenji; Takase, Yuichi*; Fukuyama, Atsushi*; Jardin, S. C.*

Plasma Physics and Controlled Fusion, 49(3), p.335 - 345, 2007/03

 Times Cited Count:7 Percentile:27.14(Physics, Fluids & Plasmas)

Non-disruptive $$beta$$-collapses with a regular intermittency have been observed after a forced turn-off of neutral beam current drive (NBCD) in JT-60U fully non-inductive, reversed shear (RS) discharges. Self-consistent transport simulations with improved core confinement and linear MHD stability analysis have first clarified that redistribution of return current induced after the NBCD turn-off lowers the safety factor of magnetic shear reversal, leading to the n =1 kink-ballooning instability with localized modes around internal transport barrier (ITB). It was also pointed out that an increase of the bootstrap current under continuous NB heating can lead to ITB reconstruction and thus causes subsequent beta-collapses.

Journal Articles

Development of in-pile capsule for IASCC study at JMTR

Matsui, Yoshinori; Hanawa, Satoshi; Ide, Hiroshi; Tobita, Masahiro*; Hosokawa, Jinsaku; Onuma, Yuichi; Kawamata, Kazuo; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Saito, Junichi; et al.

JAEA-Conf 2006-003, p.105 - 114, 2006/05

Irradiation assisted stress corrosion cracking (IASCC) caused by the simultaneous effects of radiation, stress and high temperature water environment is considered to be one of the critical concerns of in-core structural materials not only for light water reactors (LWRs) but also for water-cooled fusion reactors. In the research field of IASCC, post-irradiation examinations (PIEs) for irradiated materials have been mainly carried out, because there are many difficulties on SCC tests under neutron irradiation environment. Hence we have embarked on a development of the test techniques for performing the in-pile SCC tests. In this paper, we describe the developed several in-pile test techniques and the current status of in-pile SCC tests at Japan Materials Testing Reactor (JMTR).

JAEA Reports

SIMMER-III: A Computer Program for LMFR Core Disruptive Accident Analysis; Version 3.A Model Summary and Program Description

Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kamiyama, Kenji; Kondo, Satoru; Morita, Koji*; Fischer, E. A.; Brear, D. J.; Shirakawa, Noriyuki*; Cao, X.; et al.

JNC TN9400 2003-071, 340 Pages, 2003/08


An advanced safety analysis computer code, SIMMER-III, has been developed to investigate postulated core disruptive accidents in liquid-metal fast reactors (LMFRs). SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-dependent neutron kinetics model. By completing and integrating all the physical models originally intended at the beginning of this code development project, SIMMER-III is now applicable to integral reactor calculations and other complex multiphase flow problems. A systematic code assessment program, conducted in collaboration with European research organizations, has shown that the advanced features of the code have resolved many of the limitations and problem areas in the previous SIMMER-II code. In this report, the models, numerical algorithms and code features of SIMMER-III Version 3.A are described along with detailed program description. Areas which require future model refinement are also discussed. SIMMER-III Version 3.A, a coupled fluid-dynamics and neutronics code system, is expected to significantly improve the flexibility and reliability of LMFR safety analyses.

JAEA Reports

SIMMER-IV: A Three-Dimensional Computer Program for LMFR Core Disruptive Accident Analysis; Version 2.A Model Summary and Program Description

Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kondo, Satoru; Morita, Koji*; Sugaya, Masaaki*; Mizuno, Masahiro*; Hosono, Seigo*; Kondo, Teppei*

JNC TN9400 2003-070, 333 Pages, 2003/08


An advanced safety analysis computer code, SIMMER-III, has been developed at Japan Nuclear Cycle Development Institute (JNC) to more realistically investigate postulated core disruptive accidents in liquid-metal fast reactors. The two-dimensional framework of SIMMER-III fluid dynamics has been extended to three dimensions to a new code, SIMMER-IV, which is currently (in Version 2) coupled with the three-dimensional neutronics model. With the completion of the SIMMER-IV version, the applicability of the code is further enhanced and the many of the known limitations in SIMMER-III are eliminated. The sample calculations demonstrated the general validity of SIMMER-IV.This report describes SIMMER-IV Version 2.A, by documenting the models, numerical algorithms and code features, along with the program description and input and output information to aid the users.

JAEA Reports

Development of capsule design support subprograms for 3-dimensional temperature calculation using FEM code NISA

Tobita, Masahiro*; Matsui, Yoshinori

JAERI-Tech 2003-042, 132 Pages, 2003/03


Prediction of irradiation temperature is one of the important issues in the design of the capsule for irradiation test. Many kinds of capsules with complex structure have been designed for recent irradiation requests, and three-dimensional (3D) temperature calculation becomes inevitable for the evaluation of irradiation temperature. For such 3D calculation, however, many works are usually needed for input data preparation, and a lot of time and resources are necessary for parametric studies in the design. To improve such situation, JAERI introduced 3D-FEM (finite element method) code NISA (Numerically Integrated elements for System Analysis) and developed several subprograms, which enabled to support input preparation works in the capsule design. The 3D temperature calculation of the capsule are able to carried out in much easier way by the help of the subprograms, and specific features in the irradiation tests such as non-uniform gamma heating in the capsule, becomes to be considered.

Journal Articles

Energetic particle experiments in JT-60U and their implications for a fusion reactor

Tobita, Kenji; Kusama, Yoshinori; Shinohara, Koji; Nishitani, Takeo; Kimura, Haruyuki; Kramer, G. J.*; Nemoto, Masahiro*; Kondoh, Takashi; Oikawa, Toshihiro; Morioka, Atsuhiko; et al.

Fusion Science and Technology (JT-60 Special Issue), 42(2-3), p.315 - 326, 2002/09

 Times Cited Count:8 Percentile:49.16(Nuclear Science & Technology)

Energetic particle experiments in JT-60U are summarized, mainly covering ripple loss and Alfven eigenmodes (AE modes). Significant loss was observed for 85 keV NBI ions and fusion-produced tritons increased when toroidal field ripple at the plasma surface, especially in reversed shear plasma. Measurement of hot spots on the first wall due to ripple loss confirmed agreement with code predictions, validating the modeling incorporated in an orbit-following Monte Carlo code. A variety of AE modes were destabilized in ICRF minority heating and negative-ion-based NBI (N-NBI) heating. Most of the observed modes are gap modes identified to be TAE, EAE and NAE. Interesting finding is pulsating modes accompanying frequency sweep, which were destabilized by N-NBI and sometimes induced a beam ion loss of up to 25%. Also presented are energetic particle issues in auxiliary heating with ICRF and N-NBI.

Journal Articles

Diagnostics system of JT-60U

Sugie, Tatsuo; Hatae, Takaki; Koide, Yoshihiko; Fujita, Takaaki; Kusama, Yoshinori; Nishitani, Takeo; Isayama, Akihiko; Sato, Masayasu; Shinohara, Koji; Asakura, Nobuyuki; et al.

Fusion Science and Technology (JT-60 Special Issue), 42(2-3), p.482 - 511, 2002/09

 Times Cited Count:6 Percentile:3.03(Nuclear Science & Technology)

The diagnostic system of JT-60U (JT-60upgrade) is composed of about 50 individual diagnostic devices. Recently, the detailed radial profile measurements of plasma parameters have been improved, so that the internal structure of plasmas has been explored. The understanding of plasma confinement has been enhanced by density and temperature fluctuation measurements using a mm-wave reflectometer and electron cyclotron emission measurements respectively. In addition, the real-time control experiments of electron density, neutron yield, radiated power and electron temperature gradient have been carried out successfully by corresponding diagnostic devices. These measurements and the real time control contribute to improving plasma performance. Diagnostic devices for next generation fusion devices such as a CO2 laser interferometer/polarimeter and a CO2 laser collective Thomson scattering system have been developed.

Journal Articles

Fusion plasma performance and confinement studies on JT-60 and JT-60U

Kamada, Yutaka; Fujita, Takaaki; Ishida, Shinichi; Kikuchi, Mitsuru; Ide, Shunsuke; Takizuka, Tomonori; Shirai, Hiroshi; Koide, Yoshihiko; Fukuda, Takeshi; Hosogane, Nobuyuki; et al.

Fusion Science and Technology (JT-60 Special Issue), 42(2-3), p.185 - 254, 2002/09

 Times Cited Count:30 Percentile:48.48(Nuclear Science & Technology)

With the main aim of providing physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has been developing and optimizing the operational concepts, and extending the discharge regimes toward sustainment of high integrated performance in the reactor relevant parameter regime. In addition to achievement of the equivalent break-even condition (QDTeq up to 1.25) and a high fusion triple product = 1.5E21 m-3skeV, JT-60U has demonstrated the integrated performance of high confinement, high beta-N, full non-inductive current drive with a large fraction of bootstrap current in the reversed magnetic shear and in the high-beta-p ELMy H mode plasmas characterized by both internal and edge transport barriers. The key factors in optimizing these plasmas are profile and shape controls. As represented by discovery of various Internal Transport Barriers, JT-60/JT-60U has been emphasizing freedom and restriction of profiles in various confinement modes. JT-60U has demonstrated applicability of these high confinement modes to ITER and also clarified remaining issues.

JAEA Reports

Evaluation of dose equivalent rate for IASCC water control unit

Tobita, Masahiro*; Itabashi, Yukio

JAERI-Tech 2002-042, 40 Pages, 2002/03


In relation to aging of light water reactors (LWRs), Irradiation Assisted Stress Corrosion Cracking (IASCC) has been regarded as a significant and urgent issue for reliability of in-core components of LWRs. It is essential for IASCC studies to irradiate test materials under well-controlled of Boiling Water Reactor (BWR) conditions simulating the in-core environment. Therefore, the study for the design of the new water control unit to supply high temperature water into saturated temperature capsules in the Japan Materials Testing Reactor (JMTR) has been carried out. This report summarizes the results of estimation using ORIGEN-2 and QAD-CGGP2 codes of dose equivalent rate on outer surface of the concrete wall of installation room and dose equivalent rate around the ion-exchangers where the highest dose equivalent rate is expected in the unit after the reactor shutdown.

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