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Journal Articles

DNA damage induction during localized chronic exposure to an insoluble radioactive microparticle

Matsuya, Yusuke; Satou, Yukihiko; Hamada, Nobuyuki*; Date, Hiroyuki*; Ishikawa, Masayori*; Sato, Tatsuhiko

Scientific Reports (Internet), 9(1), p.10365_1 - 10365_9, 2019/07

 Times Cited Count:2 Percentile:40(Multidisciplinary Sciences)

Insoluble radioactive microparticles (so called Cs-bearing particles) have been assumed to adhere in the long term to trachea after aspirated into respiratory system, leading to heterogeneous dose distribution within healthy tissue around the particles. The biological effects posed by such a particle remain unclear. Here, we show cumulative DNA damage in cultured cells proximal and distal to the particle under localized chronic exposure in comparison with uniform exposure. We placed the particle-contained microcapillary onto a glass-base dish containing normal human lung cells in vitro, and observed a significant change in nuclear $$gamma$$-H2AX foci after 24 h or 48 h exposure to the particle. The dose calculation by a Monte Carlo simulation and the comparison with nuclear foci under uniform exposure suggested that the localized exposure to a Cs-bearing particle leads to not only signal-induced DNA damage to distal cells but also the reduction of DNA damage induction yield to proximal cells (protective effects). Considering the small organ dose, the conventional radiation risk assessment is adequate. This study is the first to quantify the spatial distribution of cumulative DNA lesions under heterogeneous exposure by insoluble Cs-bearing particles.

Journal Articles

Dry cleaning process test for fuel assembly of fast reactor plant system, 1; Pilot scale test for fuel pin bundle

Kudo, Hideyuki*; Otani, Yuichi*; Hara, Masahide*; Kato, Atsushi; Ishikawa, Nobuyuki; Otaka, Masahiko; Nagai, Keiichi; Saito, Junichi; Ara, Kuniaki; Ide, Akihiro*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 10 Pages, 2019/05

A next generation SFR in Japan has adopted an advanced dry cleaning system which consists of the argon gas blowing process to reduce the amount of metallic residual sodium remaining on spent fuel assemblies. This paper describes experimental and analytical work focusing on the amount of residual sodium remaining on a fuel pin bundle before and after the argon gas blowing process. The experiments were conducted using a sodium test loop and a short specimen consisting of a 7 pin bundle. The effects of the blowing gas velocity and the blowing time were quantitatively analyzed in the experiments. On the basis of these experimental results, evaluation models predicting the amount of the residual sodium were constructed.

Journal Articles

Dry cleaning process test for fuel assembly of fast reactor plant system, 2; Laboratory scale test for fuel assembly and evaluation of the amount of residual sodium

Ide, Akihiro*; Kudo, Hideyuki*; Inuzuka, Taisuke*; Hara, Masahide*; Kato, Atsushi; Ishikawa, Nobuyuki; Otaka, Masahiko; Nagai, Keiichi; Saito, Junichi; Ara, Kuniaki

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 10 Pages, 2019/05

A next generation SFR in Japan has adopted an advanced dry cleaning system which consists of the following process of argon gas blowing to reduce the amount of metallic sodium, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP without using storage containers. This three-step process increases economic competitiveness and reduces waste products. In this Research and Development work, the amount of residual sodium and performance of the dry cleaning process were investigated. This paper describes experimental and analytical work for all parts of a fuel assembly except for a fuel pin bundle.

Journal Articles

CIELO collaboration summary results; International evaluations of neutron reactions on uranium, plutonium, iron, oxygen and hydrogen

Chadwick, M. B.*; Capote, R.*; Trkov, A.*; Herman, M. W.*; Brown, D. A.*; Hale, G. M.*; Kahler, A. C.*; Talou, P.*; Plompen, A. J.*; Schillebeeckx, P.*; et al.

Nuclear Data Sheets, 148, p.189 - 213, 2018/02

 Times Cited Count:20 Percentile:3.34(Physics, Nuclear)

The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear facilities - $$^{235}$$U, $$^{238}$$U, $$^{239}$$Pu, $$^{56}$$Fe, $$^{16}$$O and $$^{1}$$H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality. This report summarizes our results and outlines plans for the next phase of this collaboration.

Journal Articles

The CIELO collaboration; Progress in international evaluations of neutron reactions on Oxygen, Iron, Uranium and Plutonium

Chadwick, M. B.*; Capote, R.*; Trkov, A.*; Kahler, A. C.*; Herman, M. W.*; Brown, D. A.*; Hale, G. M.*; Pigni, M.*; Dunn, M.*; Leal, L.*; et al.

EPJ Web of Conferences, 146, p.02001_1 - 02001_9, 2017/09

 Times Cited Count:5 Percentile:1.34

The CIELO collaboration has studied neutron cross sections on nuclides ($$^{16}$$O, $$^{56}$$Fe, $$^{235,238}$$U and $$^{239}$$Pu) that significantly impact criticality in nuclear technologies with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality.

Journal Articles

The Verification tests of the melting conditions for homogenization of metallic LLW at the JAEA

Nakashio, Nobuyuki; Osugi, Takeshi; Iseda, Hirokatsu; Tohei, Toshio; Sudo, Tomoyuki; Ishikawa, Joji; Mitsuda, Motoyuki; Yokobori, Tomohiko; Kozawa, Kazushige; Momma, Toshiyuki; et al.

Journal of Nuclear Science and Technology, 53(1), p.139 - 145, 2016/01

 Times Cited Count:1 Percentile:82.51(Nuclear Science & Technology)

no abstracts in English

Journal Articles

The CIELO Collaboration; Neutron reactions on $$^1$$H, $$^{16}$$O, $$^{56}$$Fe, $$^{235,238}$$U, and $$^{239}$$Pu

Chadwick, M. B.*; Dupont, E.*; Bauge, E.*; Blokhin, A.*; Bouland, O.*; Brown, D. A.*; Capote, R.*; Carlson, A. D.*; Danon, Y.*; De Saint Jean, C.*; et al.

Nuclear Data Sheets, 118, p.1 - 25, 2014/04

 Times Cited Count:88 Percentile:1.05(Physics, Nuclear)

CIELO (Collaborative International Evaluated Library Organization) provides a new working paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts from across the international nuclear reaction data community to identify and document discrepancies among existing evaluated data libraries, measured data, and model calculation interpretations, and aims to make progress in reconciling these discrepancies to create more accurate ENDF-formatted files. The focus will initially be on a small number of the highest-priority isotopes, namely $$^{1}$$H, $$^{16}$$O, $$^{56}$$Fe, $$^{235,238}$$U, and $$^{239}$$Pu. This paper identifies discrepancies between various evaluations of the highest priority isotopes. The evaluated data for these materials in the existing nuclear data libraries are reviewed, and some integral properties are given. The paper summarizes a program of nuclear science and computational work needed to create the new CIELO nuclear data evaluations.

Journal Articles

Radiation distribution measurement using plastic scintillating optical fibers for survey of radioactive contamination in wide area

Ito, Chikara; Ito, Keisuke; Ishikawa, Takashi; Yoshida, Akihiro; Sanada, Yukihisa; Torii, Tatsuo; Notomi, Akihiro*; Wakabayashi, Genichiro*; Miyazaki, Nobuyuki*

Hoshasen, 39(1), p.7 - 11, 2013/09

no abstracts in English

Journal Articles

Nonlinear simulation of energetic particle modes in JT-60U

Bierwage, A.; Aiba, Nobuyuki; Shinohara, Koji; Todo, Yasushi*; Deng, W.*; Ishikawa, Masao; Matsunaga, Go; Yagi, Masatoshi

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2012/10

Journal Articles

Nonlinear simulation of energetic particle modes in high-beta tokamak plasma

Bierwage, A.; Aiba, Nobuyuki; Todo, Yasushi*; Deng, W.*; Ishikawa, Masao; Matsunaga, Go; Shinohara, Koji; Yagi, Masatoshi

Plasma and Fusion Research (Internet), 7, p.2403081_1 - 2403081_4, 2012/07

Journal Articles

Design study and comparative evaluation of JSFR failed fuel detection system

Aizawa, Kosuke; Chikazawa, Yoshitaka; Ishikawa, Nobuyuki; Kubo, Shigenobu; Okazaki, Hitoshi*; Mito, Makoto*; Tozawa, Katsuhiro*; Hayashi, Masateru*

Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.465 - 474, 2012/06

A conceptual design study of an advanced sodium-cooled fast reactor JSFR has progressed in the "Fast Reactor Cycle Technology Development (FaCT)" project in Japan. JSFR has two failed fuel detection systems in the core. One is a failed fuel detection (FFD) system which continuously monitors a fission product from failed fuel subassembly. The other is a failed fuel detection and location (FFDL) system which locates when it receives signals from FFD. In this study, requirements to the FFD-DN and the FFD-DN design to meet the requirements were investigated for the commercial and demonstration JSFR. For the FFDL systems, experiences in the previous fast reactors and the research and development of FFDL system for JSFR were investigated. Operation experiences of the Selector-valve FFDL system were accumulated in PFR and Phenix. Tagging-gas system experiences were accumulated in EBR-II and FFTF.

Journal Articles

Design study on safety protection system of JSFR

Ishikawa, Nobuyuki; Chikazawa, Yoshitaka; Fujita, Kaoru; Yamada, Yumi*; Okazaki, Hitoshi*; Suzuki, Shinichi*

Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.483 - 489, 2012/06

The development of safety protection system for JSFR is progressed in terms of logic circuits, selection of trip signals and its setting values for reactor trip. In addition, it is necessitated to evaluate the satisfaction for requirements of the safety protection system by safety analyses considering comprehensive parameter ranges. For this purpose, we will report the current status of the development focusing on the evaluation results for satisfaction of safety protection system based on safety standard.

Journal Articles

Consideration on effective Pu utilization in high conversion type LWR for better transition to FBR cycle

Ishikawa, Nobuyuki; Okubo, Tsutomu

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 4 Pages, 2011/12

It is necessary to consider the transition from LWR to FBR cycle in terms of reactor introduction point of view since the FBR necessitate LWR originated plutonium in its introduction stage. The plutonium necessary for introduction of FBR is supplied from reprocessing of spent fuel of LWR, not only in the form of UOX fuel but also MOX one. In this sense, the MOX spent fuel from LWR plays a role of plutonium supply source for FBR so that it may require effective use of plutonium in LWR in advance for introduction of FBR. The high conversion type LWR (HC-LWR) may be a preferable option for plutonium utilization in LWR especially taking into account the transition to FBR cycle, since it can preserve much amount of plutonium with good isotopic composition for FBR. Considering these points, in this study, the characteristics of HC-LWR are evaluated in terms of effective plutonium utilization for transition to FBR cycle.

Journal Articles

Nonlinear hybrid simulations of energetic particle modes in realistic tokamak flux surface geometry

Bierwage, A.; Todo, Yasushi*; Aiba, Nobuyuki; Shinohara, Koji; Ishikawa, Masao; Yagi, Masatoshi

Plasma and Fusion Research (Internet), 6, p.2403109_1 - 2403109_5, 2011/08

Journal Articles

Critical condition for hydrogen induced cold cracking of 980 MPa class weld metal

Ishikawa, Nobuyuki*; Sueyoshi, Hitoshi*; Suzuki, Hiroshi; Akita, Koichi

Yosetsu Gakkai Rombunshu (Internet), 29(3), p.218 - 224, 2011/08

Cold cracking test was conducted using Y-grooved constraint weld joint by intentionally introducing hydrogen gas, and critical hydrogen content in the weld metal to cold cracking was determined. Residual stress distribution in the weld meal was measured by neutron diffraction technique. The root portion showed highest tensile stress of over 1110 MPa in the transverse direction, which is the same as the crack opening direction.

JAEA Reports

Verification of improvement of the casting process in metal melting system

Tohei, Toshio; Nakashio, Nobuyuki; Osugi, Takeshi; Ishikawa, Joji; Mizoguchi, Takafumi; Hanawa, Ritsu; Someya, Keita*; Takahashi, Kenji*; Iseda, Hirokatsu; Kozawa, Kazushige; et al.

JAEA-Technology 2010-008, 28 Pages, 2010/06

JAEA-Technology-2010-008.pdf:5.0MB

The Waste Volume Reduction Facility (WVRF) was constructed for volume reduction and the chemical stabilization of the low level radioactive waste in the Nuclear Science Research Institute of JAEA. The metal melting system in the WVRF treats radioactive metal waste. From the experience of trial operations, the improvement has conducted on the casting process in the metal melting system. The performance of the improved casting process was verified through the trial operations from Oct. 2008. In this report, we describe the reduction of the processing time, of the utilities consumption, of the load of maintenance on the improved casting process.

JAEA Reports

Research on high conversion type FLWR (HC-FLWR) core

Nakano, Yoshihiro; Fukaya, Yuji; Akie, Hiroshi; Ishikawa, Nobuyuki; Okubo, Tsutomu; Uchikawa, Sadao

JAEA-Research 2009-061, 92 Pages, 2010/03

JAEA-Research-2009-061.pdf:9.5MB

A series of research on a high conversion type innovative water reactor for flexible fuel cycle (FLWR) has been conducted. This FLWR is a boiling water reactor (BWR) with a tight triangular fuel rod lattice and the uranium plutonium mixed oxide (MOX) fuel. FLWR is designed for two types of cores to be developed in succession. The preceding core is a high conversion type FLWR (HC-FLWR) and the other core is Reduced Moderation Water Reactor (RMWR) of which the conversion ratio is more than 1.0. Three design studies and a senario study on HC-FLWR are presented in this report. The first design study is for a representative core. The second one is for a transition core from HC-FLWR to RMWR. In the transition core, both assemblies for HC-FLWR and RMWR exist. The third one is for a core to recycle minor actinides (MAs). Regarding to the scenario study, based on design results of the representative core, effective plutonium utilization in future LWR was considered within general framework.

Journal Articles

Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

Ishikawa, Nobuyuki; Okubo, Tsutomu

Annals of Nuclear Energy, 36(5), p.650 - 658, 2009/05

 Times Cited Count:7 Percentile:49.23(Nuclear Science & Technology)

The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the flow response characteristics of single fuel assembly concerning channel flow response characteristics were evaluated. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss on account of high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity feedback coefficient.

Journal Articles

General consideration on effective plutonium utilization in future LWRs

Ishikawa, Nobuyuki; Okubo, Tsutomu

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9071_1 - 9071_8, 2009/05

In this study, the potential of mixed oxide fueled light water reactors (MOX-LWRs), especially focusing on the high conversion type LWRs (HC-LWRs) such as FLWR are evaluated in terms of both economic aspect and effective use of plutonium. For economics consideration, relative economics positions of MOX-LWRs are clarified comparing the cost of electricity for uranium fueled LWRs (U-LWRs), MOX-LWRs and fast breeder reactors (FBRs) assuming future natural uranium price raise and variation of parameters such as construction cost and capacity factor. Also the economic superiority of MOX utilization against the uranium use is mentioned from the view point of plutonium credit concerning to the front-end fuel cycle cost. In terms of effective use of plutonium, comparative evaluations on plutonium mass balance in the cases of HC-LWR and high moderation type LWRs (HM-LWRs) taking into account plutonium quality (ratio of fissile to total plutonium) constraint in multiple recycling are performed as representative MOX utilization cases. Through this evaluation, the advantageous features of plutonium multiple recycling by HC-LWR are clarified.

Journal Articles

Design and scenario studies on FLWR for effective use of Pu

Iwamura, Takamichi; Ishikawa, Nobuyuki; Okubo, Tsutomu

Proceedings of 4th Asian Specialist Meeting on Future Small-Sized LWR Development, p.11_1 - 11_9, 2007/11

An advanced LWR concept of Innovative Water Reactor for Flexible fuel cycle (FLWR) has been established based on the well-experienced LWR technologies. The feature of this concept is that the high conversion type core (HC-FLWR) with small technical gap from current LWR technologies can be proceed to the breeding type FLWR core, named Reduced-Moderation Water Reactor (RMWR) under the same core configuration and reactor systems. This paper describes the investigations on designs and introduction scenario of FLWR.

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