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JAEA Reports

Development of ESRAD2 program for estimation of spatial radioactivity distribution based on kriging; User's manual

Ishigami, Tsutomu; Shimada, Taro; Seki, Masaya; Mukai, Masayuki

JAEA-Data/Code 2015-019, 122 Pages, 2015/12

JAEA-Data-Code-2015-019.pdf:3.65MB

In ensuring compliance with the criterion of site release as the final stage of termination of decommissioning of nuclear facilities, it is supposed to confirm the radioactivity concentration obtained by measurement in the site is less than or equal to the concentration corresponding to the criterion. It is needed to estimate the distribution and mean of radioactivity concentration in the evaluation unit using a number of measured data. It is further needed to compare the estimated result with the concentration corresponding to the criterion of site release and to decide if the evaluation unit should comply with the criterion. The estimated result exhibits uncertainty depending on the number of measurement points, which results in a certain probability of the occurrence of decision error according to the uncertainty. It is important to decide the number of measurement points required by revealing a relationship of the error probability to the number of measurement points for site security. We have developed the ESRAD2 (Estimation of Spatial RadioActivity Distribution program version 2), which is an extended version of the existing ESRAD, for estimating the mean of radioactivity concentration and calculating the number of measurement points required according to the error probability. This report describes a method for ensuring compliance with the criterion of site release, structure and functions, input file format, output examples, execution method of ESRAD2, and sample run with ESRAD2.

Journal Articles

Progress in long-pulse production of powerful negative ion beams for JT-60SA and ITER

Kojima, Atsushi; Umeda, Naotaka; Hanada, Masaya; Yoshida, Masafumi; Kashiwagi, Mieko; Tobari, Hiroyuki; Watanabe, Kazuhiro; Akino, Noboru; Komata, Masao; Mogaki, Kazuhiko; et al.

Nuclear Fusion, 55(6), p.063006_1 - 063006_9, 2015/06

 Times Cited Count:20 Percentile:9.28(Physics, Fluids & Plasmas)

Significant progresses in the extension of pulse durations of powerful negative ion beams have been made to realize the neutral beam injectors for JT-60SA and ITER. In order to overcome common issues of the long pulse production/acceleration of negative ion beams in JT-60SA and ITER, the new technologies have been developed in the JT-60SA ion source and the MeV accelerator in Japan Atomic Energy Agency. As for the long pulse production of high-current negative ions for JT-60SA ion source, the pulse durations have been successfully increased from 30 s at 13 A on JT-60U to 100 s at 15 A by modifying the JT-60SA ion source, which satisfies the required pulse duration of 100 s and 70% of the rated beam current for JT-60SA. This progress was based on the R&D efforts for the temperature control of the plasma grid and uniform negative ion productions with the modified tent-shaped filter field configuration. Moreover, the each parameter of the required beam energy, current and pulse has been achieved individually by these R&D efforts. The developed techniques are useful to design the ITER ion source because the sustainment of the cesium coverage in large extraction area is one of the common issues between JT-60SA and ITER. As for the long pulse acceleration of high power density beams in the MeV accelerator for ITER, the pulse duration of MeV-class negative ion beams has been extended by more than 2 orders of magnitude by modifying the extraction grid with a high cooling capability and a high-transmission of negative ions. A long pulse acceleration of 60 s has been achieved at 70 MW/m$$^{2}$$ (683 keV, 100 A/m$$^{2}$$) which has reached to the power density of JT-60SA level of 65 MW/m$$^{2}$$.

Journal Articles

Compact neutron sources for energy and security

Uesaka, Mitsuru*; Kobayashi, Hitoshi*; Kureta, Masatoshi; Nakatsuka, Shigehiro*; Nishimura, Kazuya*; Igashira, Masayuki*; Hori, Junichi*; Kiyanagi, Yoshiaki*; Tagi, Kazuhiro*; Seki, Toshichika*; et al.

Reviews of Accelerator Science and Technology, 8, p.181 - 207, 2015/00

We choose nuclear data and nuclear material inspection for energy application and nondestructive testing of explosive and hidden nuclear materials for security application. 90 keV electrostatic accelerators of deuterium are commercially available for nondestructive testing. For nuclear data measurement, electrostatic ion accelerators and L-band and S-band electron linear accelerators (linac) are used for the neutron source. Compact or mobile X-band electron linac neutron sources are under development. Compact proton linac neutron source is used for nondestructive testing especially water in solids. Several efforts for more neutron intensity using proton and deuteron accelerators are also introduced.

Journal Articles

Development of PRW welding technology for 9Cr-ODS cladding tube

Seki, Masayuki; Kihara, Yoshiyuki; Kaito, Takeji; Tsukada, Tatsuya*; Motoki, Kazuhiko*; Hirako, Kazuhito*

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Oxide Dispersion Strengthened (ODS) steel has been developed as an advanced fuel cladding tube for sodium cooled fast reactors in Japan Atomic Energy Agency (JAEA) to attain the target burn up of 150 GWd/t in the bundle average because of its excellent swelling resistance and high mechanical strength in high temperature. If conventional TIG welding is applied to the ODS welding, it is difficult to obtain necessary mechanical strength at the weld zone because of the formation of porosity. It is formed by the argon bubbles which initially dissolve in the matrix and grow up under the high temperature during welding. Therefore JAEA has been conducted the development of pressurized resistance welding (PRW) technology for ODS cladding tube, which is one of the solid state welding methods. This paper describes in the development of PRW technology, an ultrasonic test method for detecting weld defects, the result of machine strength measurement examination in weld part and the result of fuel pin irradiation examination using nuclear reactor.

Journal Articles

Thermo-hydraulic testing and integrity of ITER test blanket module (TBM) first wall mock-up in JAEA

Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio

Fusion Engineering and Design, 85(7-9), p.1255 - 1260, 2010/12

 Times Cited Count:11 Percentile:29.27(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Assessment of applicability of two-fluid model code ACE-3D to heat transfer test of supercritical water flowing in an annular channel

Nakatsuka, Toru; Ezato, Koichiro; Misawa, Takeharu; Seki, Yohji; Yoshida, Hiroyuki; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Takase, Kazuyuki

Journal of Nuclear Science and Technology, 47(12), p.1118 - 1123, 2010/12

 Times Cited Count:1 Percentile:87.11(Nuclear Science & Technology)

In order to perform efficiently the thermal design of the supercritical water reactor (SCWR), it is important to assess the thermal hydraulics in rod bundles of the core. Japan Atomic Energy Agency (JAEA) has been improved the three-dimensional two-fluid model analysis code ACE-3D, which has been developed originally for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water at supercritical region. In the present paper, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which was performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code may be applicable to prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of SCWR core.

Journal Articles

Procurement of Nb$$_3$$Sn superconducting conductors in ITER

Nabara, Yoshihiro; Isono, Takaaki; Nunoya, Yoshihiko; Koizumi, Norikiyo; Hamada, Kazuya; Matsui, Kunihiro; Hemmi, Tsutomu; Kawano, Katsumi; Uno, Yasuhiro*; Seki, Shuichi*; et al.

Journal of Plasma and Fusion Research SERIES, Vol.9, p.270 - 275, 2010/08

Journal Articles

Thermal hydraulics and mechanics research on fusion blanket system

Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 12 Pages, 2009/09

In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding/neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding/n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R&Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics.

Journal Articles

Progress of design and R&D of water cooled solid breeder test blanket module

Tsuru, Daigo; Enoeda, Mikio; Hirose, Takanori; Tanigawa, Hisashi; Ezato, Koichiro; Yokoyama, Kenji; Dairaku, Masayuki; Seki, Yohji; Suzuki, Satoshi; Mori, Kensuke*; et al.

Fusion Science and Technology, 56(2), p.875 - 882, 2009/08

 Times Cited Count:7 Percentile:45.86(Nuclear Science & Technology)

As the primary candidate of ITER Test Blanket Module (TBM) for the first day of ITER operation, development of Water Cooled Solid Breeder (WCSB) TBM has been performed toward the TBM milestones, which are necessary for acceptance of the TBM in ITER for testing from the first day of plasma operation. Milestones of ITER TBMs prior to the installation consist of milestones on safety assessment, module qualification and design integration in ITER. This paper overviews the recent achievements for preparation of the WCSB TBM for ITER day-1 operation, toward the TBM milestones.

Journal Articles

Development and achievements on the high power ECRF system in JT-60U

Moriyama, Shinichi; Kobayashi, Takayuki; Isayama, Akihiko; Terakado, Masayuki; Sawahata, Masayuki; Suzuki, Sadaaki; Yokokura, Kenji; Shimono, Mitsugu; Hasegawa, Koichi; Hiranai, Shinichi; et al.

Nuclear Fusion, 49(8), p.085001_1 - 085001_7, 2009/07

 Times Cited Count:14 Percentile:42(Physics, Fluids & Plasmas)

In the gyrotron development in JT-60U ECRF system, output power of 1.5 MW for 1 s has been achieved at 110 GHz. It is the world highest power oscillation $$>$$ 1 s. In addition to the carefully designed cavity and collector in view of thermal stress, an RF shield for the adjustment bellows, and a low-dielectric-loss DC break enabled this achievement. Power modulation technique by anode voltage control was improved to obtain high modulation frequency and 5 kHz has been achieved for NTM stabilizing experiments. Long pulse demonstration of 0.4 MW, 30 s injection to the plasma has been achieved with real time control of anode/cathode-heater. It has been confirmed that the temperature of cooled components were saturated and no evidence of damage were found. An innovative antenna having wide range of beam steering capability with linearly-moving-mirror concept has been designed for long pulse. Beam profile and mechanical strength analyses shows the feasibility of the antenna.

Journal Articles

Research and development on Water-Cooled Solid Breeder Test Blanket Module in JAEA

Enoeda, Mikio; Tanigawa, Hisashi; Tsuru, Daigo; Hirose, Takanori; Ezato, Koichiro; Yokoyama, Kenji; Dairaku, Masayuki; Seki, Yohji; Suzuki, Satoshi; Mori, Kensuke*; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Journal Articles

Research and development on plasma facing components for fusion reactors in JAEA

Suzuki, Satoshi; Ezato, Koichiro; Yokoyama, Kenji; Dairaku, Masayuki; Enoeda, Mikio; Tanigawa, Hisashi; Tsuru, Daigo; Seki, Yohji; Mori, Kensuke*; Nishi, Hiroshi; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

This paper presents the present status of R&D activities on plasma facing components (PFCs) for fusion reactors, such as ITER and fusion demonstration reactor (DEMO). The PFCs as typified by divertor and first wall components are subjected to high heat flux and particle flux from fusion plasma. It is essential to have sufficient heat removal capability and robust structure against those loadings. JAEA has been carried out to develop the ITER-PFCs which consist of copper alloys and armor materials with high thermal conductivity, such as carbon fiber composites, tungsten and beryllium. The demonstration of the thermomechanical performance has successfully been made under close mutual cooperation between the participant countries of ITER. Currently, the activity on the development of the ITER-PFCs is in a qualification phase prior to the bulk production for construction. In JAEA, the R&Ds on the DEMO-PFCs is being made in parallel with the development activity of the ITER-PFCs.

Journal Articles

Long pulse/high power ECRF system development in JT-60U

Moriyama, Shinichi; Kobayashi, Takayuki; Isayama, Akihiko; Terakado, Masayuki; Sawahata, Masayuki; Suzuki, Sadaaki; Yokokura, Kenji; Shimono, Mitsugu; Hasegawa, Koichi; Hiranai, Shinichi; et al.

Proceedings of 22nd IAEA Fusion Energy Conference (FEC 2008) (CD-ROM), 8 Pages, 2008/10

In the gyrotron development in JT-60U ECRF system, output power of 1.5 MW for 1 s has been achieved at 110 GHz. It is the world highest power oscillation $$>$$ 1 s. In addition to the carefully designed cavity and collector in view of thermal stress, an RF shield for the adjustment bellows, and a low-dielectric-loss DC break enabled this achievement. Power modulation technique by anode voltage control was improved to obtain high modulation frequency and 5 kHz has been achieved for NTM stabilizing experiments. Long pulse demonstration of 0.4 MW, 30 s injection to the plasma has been achieved with real time control of anode/cathode-heater. It has been confirmed that the temperature of cooled components were saturated and no evidence of damage were found. An innovative antenna having wide range of beam steering capability with linearly-moving-mirror concept has been designed for long pulse. Beam profile and mechanical strength analyses shows the feasibility of the antenna.

Journal Articles

Achievement of 1.5 MW, 1 s oscillation by the JT-60U gyrotron

Kobayashi, Takayuki; Moriyama, Shinichi; Seki, Masami; Sawahata, Masayuki; Terakado, Masayuki; Fujii, Tsuneyuki

Plasma and Fusion Research (Internet), 3, p.014_1 - 014_3, 2008/03

Gyrotrons are used for electron cyclotron heating (ECH) / current drive (ECCD) as high power millimeter wave sources in high performance plasma experiments. Pulse length from 0.1 to several seconds with high power is required in present tokamak experiments, such as JT-60U. However, 0.1 s oscillation had only been achieved in the power level of 1.5 MW. In JAEA, high power and long pulse oscillation experiments by using the latest JT-60U gyrotron have been tried to achieve power level of 1.5 MW and pulse length over 1 s. As a result, 1.5 MW/1 s oscillation has been successfully achieved by the fine optimization of operation parameters. In this paper, the first results of the oscillation experiment of 1.5 MW for 1 s and future plans of gyrotron improvements are described.

Journal Articles

Extended steady-state and high-beta regimes of net-current free heliotron plasmas in the Large Helical Device

Motojima, Osamu*; Yamada, Hiroshi*; Komori, Akio*; Oyabu, Nobuyoshi*; Muto, Takashi*; Kaneko, Osamu*; Kawahata, Kazuo*; Mito, Toshiyuki*; Ida, Katsumi*; Imagawa, Shinsaku*; et al.

Nuclear Fusion, 47(10), p.S668 - S676, 2007/10

 Times Cited Count:31 Percentile:20.71(Physics, Fluids & Plasmas)

The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as $$5$$$times$10^{20} m^{-3}$$, which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high-beta and long pulse.

JAEA Reports

Development of the power modulation technique in JT-60U ECH system

Terakado, Masayuki; Shimono, Mitsugu; Sawahata, Masayuki; Shinozaki, Shinichi; Igarashi, Koichi; Sato, Fumiaki; Wada, Kenji; Seki, Masami; Moriyama, Shinichi

JAEA-Technology 2007-053, 28 Pages, 2007/09

JAEA-Technology-2007-053.pdf:4.3MB

The electron cyclotron heating (ECH) system at 110 GHz are injected to JT-60U plasmas with pulse modulation at dozens to hundreds of Hz in order to measure heat conductivity of the plasma to investigate plasma confinement. The JT-60U ECH system has a unique feature to realize the pulse modulation by controlling the anode voltage of the triode gyrotron without chopping the main acceleration voltage. The typical depth of the modulation is 80 % at the modulation frequency range of 12.2 Hz to 500 Hz. However in the JT-60SA, higher modulation frequency of some kHz will be required to stabilize neoclassical tearing mode (NTM). The modulation techniques have been investigated and the modulation frequency of 3.5 kHz with the modulation depth of 84 % has been achieved. The modulation frequency up to 3 kHz is available in the pulse widths of the practical operation. As a next step, replacement of the parts in the anode voltage divider circuit is planned to achieve higher modulation frequency.

Journal Articles

Neutronics experimental study on tritium production in solid breeder blanket mockup with neutron reflector

Sato, Satoshi; Verzilov, Y.*; Ochiai, Kentaro; Wada, Masayuki*; Kutsukake, Chuzo; Tanaka, Shigeru; Abe, Yuichi; Seki, Masakazu; Oginuma, Yoshikazu*; Kawabe, Masaru*; et al.

Journal of Nuclear Science and Technology, 44(4), p.657 - 663, 2007/04

 Times Cited Count:5 Percentile:56.65(Nuclear Science & Technology)

Neutronics experiments have been performed for the solid breeder blanket using a DT neutron source at the FNS facility in JAEA. We have applied the blanket mockup composed of two enriched Li$$_{2}$$TiO$$_{3}$$ and three beryllium layers, and measured the detailed spatial distribution of the tritium production rate (TPR) using enriched Li$$_{2}$$CO$$_{3}$$ pellets. TPRs in the pellets have been measured by a liquid scintillation counter. Experiments have been done under a condition with a neutron reflector surrounding the DT neutron source. Numerical simulations have been performed using the MCNP-4C with the FENDL-2.0 and JENDL-3.3. The ranges of ratios of calculation results to experimental ones (C/Es) are 0.97-1.17 concerning with local TPR, and 1.04-1.09 for the integrated tritium production. It is found that the total integrated tritium production, which corresponds to tritium breeding ratio, can be predicted within uncertainty of 10% using the Monte Carlo calculation code and latest nuclear data libraries.

Journal Articles

Extended steady-state and high-beta regimes of net-current free heliotron plasmas in the large helical device

Motojima, Osamu*; Yamada, Hiroshi*; Komori, Akio*; Oyabu, Nobuyoshi*; Kaneko, Osamu*; Kawahata, Kazuo*; Mito, Toshiyuki*; Muto, Takashi*; Ida, Katsumi*; Imagawa, Shinsaku*; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 12 Pages, 2007/03

The performance of net-current free Heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fueling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an Internal Diffusion Barrier (IDB) by combination of efficient pumping of the local island divertor function and core fueling by pellet injection has realized a super dense core as high as 5$$times$$10$$^{20}$$m$$^{-3}$$, which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5 % and a discharge duration of 54-min. with a total input energy of 1.6 GJ (490 kW in average) are also highlighted. The progress of LHD experiments in these two years is overviewed with highlighting IDB, high $$beta$$ and long pulse.

Journal Articles

Acceleration of MeV-class energy, high-current-density H$$^{-}$$-ion beams for ITER neutral beam system

Taniguchi, Masaki; Inoue, Takashi; Kashiwagi, Mieko; Watanabe, Kazuhiro; Hanada, Masaya; Seki, Takayoshi*; Dairaku, Masayuki; Sakamoto, Keishi

Review of Scientific Instruments, 77(3), p.03A514_1 - 03A514_4, 2006/03

 Times Cited Count:9 Percentile:49.43(Instruments & Instrumentation)

In the ITER NB systems, conventional gas insulation technology cannot be utilized because of the conductivity of the insulation gas caused by the radiation from the tokamak plasma. To overcome this problem, a vacuum insulated beam source (VIBS), where the whole beam source is immersed in vacuum, has been developed in JAERI. Recently, voltage holding capability of the VIBS was drastically improved by installing the large stress ring and these progress enables us to perform the high power operation of the VIBS accelerator. For high current density H$$^{-}$$- beam acceleration, modifications were made on KAMABOKO source. At present, H$$^{-}$$- beam current density is 146 A/m$$^{2}$$ at 836 keV (input arc power; 40 kW, operation pressure; 0.3 Pa).The acceleration of 900 keV, 0.1 A level beam was accomplished for 175 shots during the test campaign. The beam acceleration was quite stable and the degradation of the voltage holding due to the beam acceleration and/or Cs seeding was not observed.

Journal Articles

Improvement of beam uniformity by magnetic filter optimization in a Cs-seeded large negative-ion source

Hanada, Masaya; Seki, Takayoshi*; Takado, Naoyuki*; Inoue, Takashi; Tobari, Hiroyuki; Mizuno, Takatoshi*; Hatayama, Akiyoshi*; Dairaku, Masayuki; Kashiwagi, Mieko; Sakamoto, Keishi; et al.

Review of Scientific Instruments, 77(3), p.03A515_1 - 03A515_3, 2006/03

 Times Cited Count:21 Percentile:25.31(Instruments & Instrumentation)

no abstracts in English

120 (Records 1-20 displayed on this page)