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Journal Articles

IAEA low enriched uranium bank; Towards the realization of international management initiatives

Tamai, Hiroshi; Tazaki, Makiko; Suda, Kazunori

Nippon Genshiryoku Gakkai-Shi, 60(1), p.25 - 29, 2018/01

IAEA Low Enriched Uranium Bank, which is one of international management initiatives of nuclear materials operated by IAEA, will be realized soon. During increasing concern on proliferation risk of sensitive nuclear technologies as well as in this century the potential acquisition by terrorists, the IAEA bank will offer the fuel assurance aiming at decreasing incentive for acquiring those sensitive technologies. Throughout the argument on the criteria for the fuel supply, the bank site and its requirement have been established and will be in operation next year. The background, significance, and development of this initiative are described.

Journal Articles

International trends on nuclear security and Japan's contribution to nuclear security; Brief summary of the 2016 Nuclear Security Summit and challenges for strengthening global nuclear security

Tazaki, Makiko; Suda, Kazunori

Nippon Genshiryoku Gakkai-Shi, 58(10), p.594 - 598, 2016/10

The 4th (and the last) Nuclear Security Summit was held in Washington D.C. in 31 March and 1 April 2016. Brief reviews of pas 3 nuclear security summits, including the 4th one, Japan contribution to the summits, future challenges of post nuclear security summits and Japan role are described in the paper.

JAEA Reports

Study on the application of CANDLE burnup strategy to several nuclear reactors, JAERI's nuclear research promotion program, H13-002 (Contract research)

Sekimoto, Hiroshi*

JAERI-Tech 2005-008, 111 Pages, 2005/03


The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. When this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the bunrup of discharged fuel is about 40%. It means that the nuclear energy can be utilized for many hundreds years without new mining, enrichment and reprocessing, and the amount of spent fuel can be reduced considerably. Compared to fast reactors, application of CANDLE burnup to prismatic fuel high-temperature gas cooled reactors is very easy. In this report, the applications of CANDLE burnup to both these types of reactors are studied.

Journal Articles

Criticality safety benchmark experiment on 10% enriched uranyl nitrate solution using a 28-cm-thickness slab core

Yamamoto, Toshihiro; Miyoshi, Yoshinori; Kikuchi, Tsukasa*; Watanabe, Shoichi

Journal of Nuclear Science and Technology, 39(7), p.789 - 799, 2002/07

 Times Cited Count:4 Percentile:68.69(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

The 3rd technological meeting of Tokai reprocessing plant

Maki, Akira; ; Taguchi, Katsuya; ; Shimizu, Ryo; Shoji, Kenji;

JNC-TN8410 2001-012, 185 Pages, 2001/04


"The third technological meeting of Tokai Reprocessing plant (TRP)" was held in JNFL Rokkasyo site on March 14$$^{th}$$, 2001. The technical meetings have been held in the past two times. The first one was about the present status and future plan of the TRP and second one was about safety evaluation work on the TRP. At this time, the meeting focussed on the corrosion experrience, in-service inspection technology and future maintenance plan. The report contains the proceedings, transparancies and questionnaires of the meeting are contained.

JAEA Reports

Estimation of biases for inserted reactivity estimation of JCO criticality accident

Yamamoto, Toshihiro; Nakamura, Takemi*; Miyoshi, Yoshinori

JAERI-Data/Code 2001-001, 30 Pages, 2001/02


no abstracts in English

JAEA Reports

Study on dissolution of UO$$_{2}$$ to obtain the high U solution

; *; Sakurai, Koji*; *; *; *

JNC-TN8400 2000-032, 98 Pages, 2000/12


Concerning the preparation of high U solution for the crystallization process and the application of UO$$_{2}$$ powder dissolution to that, the effects of final U concentration, dissolution temperature, nitric acid concentration and powder size on the dissolution of UO$$_{2}$$ powder in the nitric acid where the final U concentration was $$sim$$800g/L were investigated. The experimental results showed that the solubility of UO$$_{2}$$ decreased with the increase of final UO$$_{2}$$ concentration and powder size, and with the decrease of dissolution temperature and nitric acid concentration. It was also confirmed that in the condition where the final U concentration was sufficiently lower than the solubility of U,,UO$$_{2}$$ dissolution behavior in the high U solution could be estimated with the equation based on the fragmentation model which we had already reported. Based on these experimental results, the dissolution behavior of irradiated MOX fuel in high U solution was estimated and the possibility of supplying high U solution to the crystallization process was discussed. In the preparation of high U solution for the crystallization process, it was estimated that the present dissolution process (dissolution for fuel pieces of about 3cm long) needed a lot of time to obtain a high dissolution yield, but it was shorted drastically by the pulverization of fuel pieces. The burst of off-gas at the early in the dissolution of fuel powder seems to be avoidable with setting the appropriate dissolution condition, and it is important to optimize the dissolution condition with considering the capacity of off-gas treatment process.

JAEA Reports

Analyse on the BFS critical experiments; An analysis on the BFS-62-1 assembly

; Iwai, Takehiko*;

JNC-TN9400 2000-098, 182 Pages, 2000/07


In order to support the Russian excess weapons plutonium disposition, the international collaboration has been started between Japan Nuclear Cycle Development Institute (JNC) and Russian Institute of Physics and Power Engineering (IPPE). In the frame of the collaboration, JNC has carried out analyses on the BFS-62 assemblies that are constructed in the fast reactor critical experimental facility BFS-2 of IPPE. This report summarizes an experimental analysis on the BFS-62-1 assembly, which is the first core of the BFS-62 series. The core contains the enriched U0$$_{2}$$ fuel surrounded by the U0$$_{2}$$ blanket. The standard analytical method for fast reactors has been applied, which was used for the JUPITER and other experimental analyses. Due to the lack of the analytical data the 2D RZ core calculation was mainly used. The 3D XYZ core calculation was applied only for the preliminary evaluation. Further in terms of the utilization of the BFS experimental analysis data for the standard data base for FBR core design, consistency evaluation with JUPITER experimental analysis data has been performed using the cross-section adjustment method. As the result of analyses, good agreement was obtained between calculations and experiments for the criticality and the reaction rate ratio. However, it was found that accurate evaluation of the reaction rate distribution was impossible without exact consideration of the arrangement of the two types of sodium (with and without hydrogen impurity), which can be accommodated by the 3D core analysis, thus it was essentia1. In addition, it was clarifie that there was a room for an improvement of the result on the reaction rate distribution in the blanket and shielding regions. The application of the 3D core calculation improved the result on the control rod worth because 3D core model can more exactly consider the shape of the control rod. Furthermore it was judged that the result of the analysis on the sodium void reactivity .....

JAEA Reports

Experimental analyses results on the BFS 58-1-I1 critical assemblies

; Sato, Wakaei*; Iwai, Takehiko*

JNC-TN9400 2000-096, 113 Pages, 2000/06


This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0$$_{2}$$ fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0$$_{2}$$ zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...

JAEA Reports

Criticality safety evaluation in Tokai reprocessing plant

Shirai, Nobutoshi; ; ; Shirozu, Hidetomo; ; Hayashi, Shinichiro;

JNC-TN8410 2000-006, 116 Pages, 2000/04


Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 "Criticality safety of single unit" in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units.

Journal Articles

Current status of criticality safety experiment in NUCEF and its enhancement of facility function toward Pu experiment

Takeshita, Isao; Oono, Akio; Izawa, Naoki*; Miyoshi, Yoshinori; Maeda, Atsushi; Sugikawa, Susumu; Miyauchi, Masakatsu

Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), p.1512 - 1576, 1999/09

no abstracts in English

JAEA Reports

Safety analysis of JMTR core with 6-MEU fuel elements and 16-LEU fuel elements

Tabata, Toshio; Komukai, Bunsaku; Nagao, Yoshiharu; Shimakawa, Satoshi; Koike, Sumio; Takeda, Takashi; Fujiki, Kazuo

JAERI-Tech 99-021, 68 Pages, 1999/03


no abstracts in English

JAEA Reports

Annual report of STACY operation in F.Y. 1997; 280mm thickness slab core・10% enriched uranyl nitrate solution (Contract research)

; ; Hirose, Hideyuki; *; *; Murakami, Kiyonobu; Takahashi, Tsukasa; Sakuraba, Koichi; Miyauchi, Masakatsu; ; et al.

JAERI-Tech 98-023, 66 Pages, 1998/06


no abstracts in English

JAEA Reports

Journal Articles

Application of a Phoswich detector for simultaneous counting of $$alpha$$- and $$beta$$($$gamma$$)-rays in a rotating drum-cell type monitor

Usuda, Shigekazu; Yasuda, Kenichiro; Sakurai, Satoshi; Takahashi, Toshiyuki; Gunji, Hideho*; P.Howarth*

INMM 39th Annual Meeting Proceedings (CD-ROM), 27, 6 Pages, 1998/00

no abstracts in English

Journal Articles

Experience in the implementation of physical protection measures of nuclear material at the JAERI Tokai Establishment

; ; Tsuruta, Harumichi

Physical Protection of Nuclear Materials:Experience in Regulation,Implementation and Operations, p.195 - 200, 1998/00

no abstracts in English

Journal Articles

Improvement of a rotating drum-cell type alpha monitor and its performance test

Usuda, Shigekazu; Yasuda, Kenichiro; Sakurai, Satoshi; Takahashi, Toshiyuki; Gunji, Hideho*

Dai-18-Kai Kaku Busshitsu Kanri Gakkai (INMM) Nippon Shibu Nenji Taikai Hobunshu, p.142 - 148, 1997/11

no abstracts in English

JAEA Reports


PNC-TN1440 97-005, 76 Pages, 1997/09


no abstracts in English

JAEA Reports

Annual report of STACY in 1995; 600mm diameter cylindrical core and 10% enriched uranyl nitrate solution

; ; Hirose, Hideyuki; *; *; Oono, Akio; Sakuraba, Koichi; Izawa, Naoki; Tonoike, Kotaro; *; et al.

JAERI-Tech 97-005, 107 Pages, 1997/03


no abstracts in English

JAEA Reports

Development of a standard data base for FBR core nuclear design (VI): JUPITER-II experimental data book


PNC-TN9450 96-052, 694 Pages, 1996/10


The present report compiles the experimental data of JUPITER Phase-II, which was a joint research program between U.S. DOE and PNC of Japan, using the ZPPR facility, which stands for Zero Power Physics Reactor at ANL-Idaho in l982 to l984. The JUPITER-II experiment was a series of critical experiments for conventional radial heterogeneous cores of 650 MWe class LMFBR, including six experimental cores. The nuclear characteristics recorded here include criticality, control rod reactivity, reaction rate distribution, sodium void reactivity, sample reactivity, Doppler reactivity and gamma heating. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The detail of experimental data is thoroughly recorded here so as to re-analyze these experiments in future. In addition, these experimental data are installed in the computer system at OEC for convenience of analytical code input.

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