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; Sasaki, Makoto; *; *;
JNC TN9400 2001-113, 219 Pages, 2001/09
JNC-TN9400-2001-113.pdf:10.46MB
In the phase-I of the Feasibility Studies (F/S) on Commercialized Fast Reactor Cycle Systems from July 1999 to the end of FY2000, technology candidates of various commercialized fast reactor (FR) recycle concepts were investigated, in response to the JNC middle long term enterprise plan. This report describes about the core concept study of sodium-cooled oxide-fuel fast breeder reactors as a part of the investigation. The major results of JFY2000 are as follows: (1)One of the promising candidates of large and medium scale Na cooled oxide cores is radial heterogeneous core with axial blanket partial elimination driver fuel sub-assemblies, which achieves re-criticality free and high breeding capability. Core thermal hydraulic design is one of the technical matters of this concept to be investigated in detail. (2)There are concerns of core breeding capability of large scale Na cooled oxide core with the inner-duct sub-assembly concept. 0ne of the possible solutions of the concerns is selection radial heterogeneous core concept with inner-duct sub-assembly. (3)Results of core shielding analysis showed that feasible radial shielding thickness is 2 sub-assemblies layer for the core with radial blanket and 3 sub-assemblies layer for the core without radial blanket.
; Sasaki, Makoto; *; *;
JNC TN9400 2001-112, 174 Pages, 2001/09
JNC-TN9400-2001-112.pdf:8.09MB
As a part of phase-I of the Feasibility Studies of Commercialized Fast Reactor Cycle Systems (F/S), fast reactor core characteristics sensitivity study has been performed to understand the relationship between core performances of candidate concepts in the F/S and fuel specification variations, which correspond to the candidates of advanced fuel cycle technology concepts in the F/S, including fuel isotopic compositions. The major results of JFY2000 study are as follows: (1)It is indicated by neutronic calculation that change of core characteristic is not significant even the cases of variation of TRU composition and residual fission products in the recycled fuel which corresponds to advanced fuel cycle candidates. And such change is within a range in which significant modification of core design would not be required. (2)The core characteristic sensitivity study with oxide fuel concept options such as pellet, vi-pack, etc. indicated that the fuel smeared density variation has certain contribution to the core characteristic, especially to the breeding ration. The breeding ratio was calculated to be below l.2 even in the radial heterogeneous core if the fuel smeared density is as low as 80%TD. (3)Accessibility of irradiated radial blanket sub-assembly is evaluated in a viewpoint of proliferation resistance of core concept with radial blanket. The results showed that the heavy shielding and remote handling, similar to the general reprocessing plants, are indispensable to handle the irradiated radial blanket even after 5 years cooling.
; Hayashi, Hideyuki; ; ; ; Takaki, Naoyuki;
JNC TY9400 2001-011, 493 Pages, 2001/03
JNC-TY9400-2001-011.pdf:20.55MB
no abstracts in English
; Hayashi, Hideyuki; ; ; *; ;
JNC TY9400 2000-021, 452 Pages, 2000/03
JNC-TY9400-2000-021.pdf:16.64MB
no abstracts in English
; Hayashi, Hideyuki; ; ; ; ;
JNC TN9400 2000-070, 146 Pages, 2000/03
Feasibility studies(F/S) have been undertaken since July, 1999 in order to determine promising concepts of a commercialized fast reactor cycle system and to define the related necessary R&D tasks. ln the phase l(FYs of 1999-2000) of this F/S, a number of conceptual FBR candidates are evaluated. As for this study, a parameter survey on core characteristics of lead cooled fast breeder reactors (FBRs) has been performed. This report describes the intermediate results obtained in the first FY of the phase l. BREST-300 (Russia) is selected as one example for the parameter survey by using the JNC original analysis method, because it is easy to obtain enough information of the core design and its characteristics. The comparison of core characteristics has been performed under the same thermo-hydraulic conditions between lead cooled FBRs and sodium cooled ones. As a result, problems to be resolved have been listed up, and their core characteristics have been evaluated from the target review points of the F/S. The results have been obtained are as follows: (1)High breeding (internal conversion ratio 
1) of BREST-300 is mainly due to loading nitride fuel, though the effect of reflecting neutrons is high in the lead coolant. (2)lt may be difficult to reach 150GWd/t due to surface erosion and FCMl of fuel claddings. (3)The maximum fuel cladding temperature of lead cooled FBRs becomes about 40
C higher than sodium cooled ones under the same cooling condition. (4)Fuel pin pitch of lead cooled FBRs becomes larger than that of sodium cooled ones, under the coolant flow condition where 
T
and bundle pressure drop are the same for both cases. Therefore, breeding of the former is not always superior to the latter. From this study core characteristics of lead cooled FBRs, about which we had no experience of design studies so far, was almost clear.
; Hayashi, Hideyuki; ; ; *; ;
JNC TN9400 2000-068, 337 Pages, 2000/03
JNC-TN9400-2000-068.pdf:12.64MB
Feasibility studies(F/S) have been undertaken since July, 1999 in order to determine promising concepts of a commercialized fast reactor cycle system and to define the related necessary R&D tasks. ln the phase l(FYs of 1999-2000) of this F/S, a number of conceptual candidates are selected from the following 5 viewpoints: (a)ensuring safety, (b)economic competitiveness to future LWRs, (c)efficient utilization of resources, (d)reduction of environmental burden, (e)enhancement of nuclear non-proliferation. As for this study based on the above viewpoints, core characteristics of sodium-cooled fast breeder reactors have been surveyed and classified in the combinations of fuels (MOX, metal and nitride). and power output scales. As a result, R&D items to be performed have been proposed, a data base to select candidate reactor concepts has been prepared. The intermediate results obtained in the first FY of the phase l are as follows: (1)There is a limitation in expansion of operation duration for large scale FBRs with MOX fuel. ln case of the reactor with a short doubling time, it is possible to obtain doubling time less than 30 years. (2)The MA transmutation ratio per cycle is about 11% in case of MOX fuel with 5 weight% MA. The difference of this ratio among MOX, metal and nitride fuels is small. (3)A low decontamination fuel with 2 volume% FP may be possible to be used in FBR core designs. (4)The concept of re-criticality prevention may be possible by adoption of a fuel assembly with partly removed axial blanket fuel and a radial heterogeneous core. (5)There is no significant difference of core haracteristics between metal fuel and nitride one, which are suitable for the targets of the F/S.
; Ogawa, Shinta*; Hayafune, Hiroki; ; ; Tozawa, Katsuhiro; ;
PP.2739, (105), p.27 - 40, 1998/03
None
Takahashi, Kenji; Yamaguchi, Toshihiko; Onizawa, Takahiro; Kurosawa, Norifumi; Shiina, Akira; Tagawa, Akihiro; Ibaki, Shoji
no journal, ,
no abstracts in English
Shiina, Akira; Yamaguchi, Toshihiko; Onizawa, Takahiro; Kurosawa, Norifumi; Takahashi, Kenji; Ibaki, Shoji
no journal, ,
no abstracts in English
