Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Nakazawa, Osamu; Takiya, Hiroaki; Murakami, Masashi; Donomae, Yasushi; Meguro, Yoshihiro
JAEA-Review 2023-012, 6 Pages, 2023/08
JAEA-Review-2023-012.pdf:0.93MB
The selection of back-end technology development issues to be prioritized and their schedule of the Japan Atomic Energy Agency (JAEA) have been put together as the "Strategic Roadmap for Back-end Technology Development." The results of questionnaires on development technologies (seeds) and technical issues (needs) within JAEA conducted in FY2022 were reflected in the selection. The issues were extracted from among those that match the seeds and needs, from the perspective of early implementation in the work front and the perspective of common issues, and nine themes were selected. We will build a cross-organizational implementation framework within JAEA and aim to implement the development results in the work front as well as social implementation.
Hayafune, Hiroki; Maeda, Seiichiro; Ohshima, Hiroyuki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 61(11), p.798 - 803, 2019/11
In the "Strategic Roadmap" of Fast Reactor Development decided at the Inter-Ministerial Council for Nuclear Power in December 2018, the development works for the around next 10 years were identified, and the role of JAEA was presented. In response, JAEA has prepared a framework for R&D plans for about 5 years on the fast reactor technology and the fuel cycle technology (reprocessing, fuel manufacturing, fuel and material development). In the future, JAEA will promote independent R&D works based on these plans, and provide the obtained R&D results together with various testing functions of JAEA to the activities of the private sector, etc. Through these actions, JAEA will actively contribute to the future fast reactor development. This article outlines JAEA's policy and the R&D items (development of ARKADIA; Advanced Reactor Knowledge- and AI-Aided Design Integration Approach through the whole Plant Life Cycle, development of standards and standards system, development of safety improvement technology, research in the fuel cycle technology), the policy of international cooperation, the human resource development, and the future perspective were explained.
Washiya, Tadahiro; Suzuki, Shunichi*
Nihon Kikai Gakkai-Shi, 122(1211), p.13 - 15, 2019/10
International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) was held in last May. In this report, the out-line of Track-2: Debris Removal Strategy, Risk, Debris Characterization was described. In this Task-2, "1F decommissioning strategy plan 2019" was introduced by NDF, and challenging tasks are explained. As 1F's R&D issues, debris cutting technologies, dust generation behavior analysis and risk evaluation was reported. In this paper, the out-line of Track 2 will be reported.
Nishihata, Yasuo; Tanaka, Hirohisa*; Mitachi, Senshu*; Kasai, Hideaki*
Kogyo Zairyo, 62(5), p.41 - 44, 2014/05
no abstracts in English
Degueldre, C.*; Akie, Hiroshi; Boczar, P.*; Chauvin, N.*; Meyer, M.*; Troyanov, V.*
Proceedings of GLOBAL2003 Atoms for Prosperity; Updating Eisenhower's Global Vision for Nuclear Energy (CD-ROM), p.1967 - 1973, 2003/11
no abstracts in English
Konishi, Satoshi; Okano, Kunihiko*; Tokimatsu, Koji*; Ito, Keishiro*; Ogawa, Yuichi*
Fusion Engineering and Design, 69(1-4), p.523 - 529, 2003/09
Times Cited Count：4 Percentile：31.64(Nuclear Science & Technology)no abstracts in English
Morimoto, Kyoichi; Shibata, Atsuhiro; Shigetome, Yoshiaki
JNC TN8200 2001-006, 19 Pages, 2001/12
JNC-TN8200-2001-006.pdf:0.92MB
Global2001 (International Conference: "Back-End of the Fuel Cycle: From Research to Solutions ") was held for six days from September 9 to September 14 in Paris in France. In this year, there were about 420 participants from each country and about 70 people participated from Japan. This conference consisted of the reactor and fuel cycle field, the reprocessing field, the disposal field, and the non-proliferation field, etc. The main topics of this conference were the back end of the nuclear fuel cycle, the management of long-lived nuclide, the advanced concept of reactor and fuels. Advanced fuel recycle technology division reported about the feasibility study on commercialized FR cycle systems, the nuclear fuel and the reprocessing process in the oral session and poster session. Each report was audited and information was collected. It is possible to refer to information on Global2001 by the following homepages. http://www.cea.fr/conferences/global2001 /index.him*
Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji*
Genshiryoku eye, 47(7), p.60 - 64, 2001/07
no abstracts in English
Ehrlich, K.*; Bloom, E. E.*; Kondo, Tatsuo
Journal of Nuclear Materials, 283-287(1), p.79 - 88, 2000/12
Times Cited Count：86 Percentile：97.75(Materials Science, Multidisciplinary)no abstracts in English
; Inagaki, Tatsutoshi*
JNC TY1400 2000-004, 464 Pages, 2000/08
JNC-TY1400-2000-004.pdf:19.55MB
None
; Inagaki, Tatsutoshi*
JNC TY1400 2000-003, 92 Pages, 2000/08
Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power company (JAPCO, that is the representative of the electric utilities in Japan) have established a new organization to develop a commercialized fast breeder reactor (FBR) cycle system since July 1, 1999 and feasibility studies (F/S) have been undertaken in order to determine the promising concepts and to define the necessary R&D tasks. In the first two-year phase, a number of candidate concepts will be selected from various options, featuring innovative technologies. In the F/S, the options are evaluated and conceptual designs are examined considering the attainable perspectives for following: (1) ensuring safety, (2) economic competitiveness to future LWRs, (3) efficient utilization of resources, (4) reduction of environmental burden and (5) enhancement of nuclear non-proliferation. The F/S should also guide the necessary R&D to commercialize FBR cycle system.
*
JNC TN1440 2000-007, 115 Pages, 2000/08
JNC-TN1440-2000-007.pdf:4.45MB
no abstracts in English
*
JNC TN1440 2000-005, 214 Pages, 2000/08
JNC-TN1440-2000-005.pdf:13.81MB
no abstracts in English
; ; ; ;
JNC TN9400 2000-066, 52 Pages, 2000/06
JNC-TN9400-2000-066.pdf:1.82MB
Phase I of feasibility studies on commercialized fast reactor system is being peformed for two years from Japanese Fiscal Year 1999. In this report, results of the study on fluid fuel reactors (especialiy a molten salt fast breeder reactor concept) are described from the viewpoint of technical and economical concerns of the plant system design. ln JFY1999, we have started to investigate the fluid fuel reactors as alternative concepts of sodium cooled FBR systems with MOX fuel, and selected the unique concept of a molten chloride fast, breeder reactor, whose U-Pu fuel cycle can be related to both light water reactors and fast breeder reactors on the basis of present technical data and design experiences. We selected a preliminary composition of molten fuel and conceptual plant design through evaluation of technical and economical issues essential for the molten salt reactors and then compared them with reference design concepts of sodium cooled FBR systems under limited information on the molten chloride fast breeder reactors. The following results were obtained. (1)The molten chloride fast breeder reactors have inherent safety features in the core and plant performances, ad the fluid fuel is quite promising for cost reduction of the fuel fabrication and reprocessing. (2)On the other hand, the inventory of the molten chloride fuel becomes high and thermal conductivity of the coolant is inferior compared to those of sodium cooled FBR systems, then, the size of main components such as lHX's becomes larger and the amount of construction materials is seems to be increased. (3)Furthermore economical vessel and piping materials which contact with the molten chloride salts are required to be developed. From the results, it is concluded that further steps to investigate the molten chloride fast breeder reactor concepts are too early to be conducted.
Ohtaki, Akira; ; ; *; *;
JNC TN9410 2000-006, 74 Pages, 2000/04
JNC-TN9410-2000-006.pdf:3.01MB
To evaluate materials balance in nuclear fuel cycle quickly and quantitatively the fuel cycle requirement code "FAMILY" was improved. And an accumulated TRU&LLFP quantity analysis code was developed. The contents are as follows: (1)A calculation ability of minor actinide production and expenditure was added to the "FAMILY" code. (2)An output program for the "FAMILY" calculation results was developed. (3)A simple version of "FAMILY" code was developed. (4)An analysis code for accumulated TRU&LLFP quantity in nuclear fuel cycle was developed.
Fujiwara, Masayuki; Mizuta, Shunji;
JNC TN9400 2000-050, 19 Pages, 2000/04
JNC-TN9400-2000-050.pdf:0.82MB
For evaluating the fast reactor system technology, it is important to evaluate the practical feasibility of ODS ferritic cdaddings, which is the most promising matelials to attain the goal of high coolant temperature and more than 150 GWd/t. Based on the results of their technology development, mass production process with highly economically benefit as well as manufacturing cost estimation of ODS ferritic claddings were preliminarily conducted. From the view point of future utility scale, the cost for manufacturig mother tubes has a dominant factor in the total manufacturing cost. The method to reduce the cost of mother tube manufacturing was also preliminarily investigated.
JNC TN9400 2000-038, 98 Pages, 2000/04
JNC-TN9400-2000-038.pdf:7.49MB
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
Nuclear Non-Proliferation Study Group*; *; Iwata, Shuichiro*; *; *; *;
JNC TN1400 2000-008, 81 Pages, 2000/04
JNC-TN1400-2000-008.pdf:3.44MB
no abstracts in English
; ;
JNC TN9400 2000-041, 29 Pages, 2000/03
JNC-TN9400-2000-041.pdf:1.18MB
Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.
gas cooled reactor; ; Mizuta, Shunji
JNC TN9400 2000-040, 41 Pages, 2000/03
JNC-TN9400-2000-040.pdf:0.85MB
The corrosion behavior of ferritic stainless steels applied to core components under C0 gas environment was investigated in order to be helpful to fuel design in C0 gas cooled reactor as the feasibility study for fast breeder reactor. The dependence of the corrosion behavior, before a breakaway occurs, on C0 gas temperature, Si and Cr contents of ferritic steels was determined quantitatively. The following correlations to calculate the metal loss thickness was established. X = 4.4w w = √(k
t) k = 
exp( - 5.45[Si]) exp( - 1.09[Cr]) exp( - 11253/T) = 1.65 10
4.40 10
X : metal loss thickness[
ml, w : corrosion weight gain [mg/cm
] k : parabola constant [(mg/cm)/hr], t : time [hr], : constant [Si] : Si content[wt.%], [Cr] : Cr content [wt.%], T : temperature [K]
