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Ohgama, Kazuya; Kawashima, Katsuyuki*; Oki, Shigeo
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code 
-FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750 MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.
Kawashima, Katsuyuki; Sugino, Kazuteru; Oki, Shigeo; Okubo, Tsutomu
Nuclear Technology, 185(3), p.270 - 280, 2014/03
Times Cited Count：1 Percentile：8.88(Nuclear Science & Technology)Although the sodium void reactivity is limited up to 6 dollars in the current JSFR design, it should be significant to perform design studies of the low sodium void reactivity core besides the reference design, to increase the design margin considering any influence of the TRU fuel compositions. In this study, the BUMPY core is proposed as the low sodium void core concept, in which the partial-length fuels with upper sodium plenum are interspersed within the core, causing the steps in fuel length in the neighboring fuel assemblies. The void reactivity is considerably reduced due to the upward and lateral neutron leakage from the fuel region to the sodium plenum upon voiding. The BUMPY core is applied to the JSFR design. The calculated void reactivity of the BUMPY core is 2.5 dollars, which is considerably reduced from 5.3 dollars for the reference core. Moreover, the Doppler coefficient is almost the same as the reference core.
Meiliza, Y.; Oki, Shigeo; Kawashima, Katsuyuki; Okubo, Tsutomu
Progress in Nuclear Energy, 70, p.270 - 278, 2014/01
Times Cited Count：2 Percentile：16.44(Nuclear Science & Technology)Meiliza, Y.; Oki, Shigeo; Kawashima, Katsuyuki; Okubo, Tsutomu
Journal of Nuclear Science and Technology, 50(6), p.615 - 628, 2013/05
Times Cited Count：1 Percentile：10.69(Nuclear Science & Technology)Uematsu, Mari Mariannu; Sugino, Kazuteru; Kawashima, Katsuyuki; Okano, Yasushi; Yamaji, Akifumi; Naganuma, Masayuki; Oki, Shigeo; Okubo, Tsutomu; Ota, Hirokazu*; Ogata, Takanari*; et al.
JAEA-Research 2012-041, 126 Pages, 2013/02
JAEA-Research-2012-041.pdf:16.49MB
The characteristics of sodium-cooled metal fuel core compared to MOX fuel core are given by its higher heavy metal density and superior neutron economy. By taking advantage of these characteristics and allowing flexibility in metal fuel specification and core design conditions as sodium void reactivity and bundle pressure drop, core design with high burnup, high breeding ratio and low fuel inventory features will be achievable. On ground of the major achievements in metal fuels utilization as driver fuels in sodium fast reactors in U.S., the metal fuel core concept is selected as a possible alternative of MOX fuel core concept in FaCT project. This report describes the following items as a result of the joint study on "Reactor core and fuel design of metal fuel core of sodium-cooled fast reactor" conducted by JAEA and CRIEPI during 4 years from fiscal year 2009 to 2012.
Maruyama, Shuhei; Oki, Shigeo; Okubo, Tsutomu; Kawashima, Katsuyuki; Mizuno, Tomoyasu
Journal of Nuclear Science and Technology, 49(6), p.640 - 654, 2012/06
Times Cited Count：3 Percentile：24.98(Nuclear Science & Technology)This study shows the good correlations in FBR core characteristics, and find out the mechanism of their correlations with the aid of sensitivity analyses. It has been clarified that Doppler coefficient turns to have the correlations with the other core characteristics by considering the constraint of the criticality requirement for fuel composition variations. The finding of the correlations makes easy to specify the ranges of core reactivity control and core safety properties which are important for core design in determining core specification and performance. It gives significant information for FBR core design in the transition stage. Moreover, as an application of the above-mentioned correlations, a simplified burnup reactivity index is developed for rapid and rational estimation of the core characteristic variations. By using this index and the correlations, the core characteristic variations can be estimated for various fuel compositions without repeating core calculations.
Kawashima, Katsuyuki; Ogawa, Takashi; Oki, Shigeo; Okubo, Tsutomu; Mizuno, Tomoyasu
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
Sodium-cooled fast reactor core design considerations are made to improve the proliferation resistance by focusing on the plutonium generated in the UO
blanket in the frame of the Fast Reactor Cycle Technology Development (FaCT) project. The appropriate design and treatments of the UO blanket help to reduce the intrinsic proliferation potentials. Based on the 1500 MWe FaCT reference core, the three different cores (radial blanket-free core, the core with the low-enriched MOX fuel, and the core with MA-doped UO fuel) are configured to meet the provisional proliferation resistance criteria as well as the core performance targets.
Kawashima, Katsuyuki; Okubo, Tsutomu; Mizuno, Tomoyasu
Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12
As part of the FaCT Project, the study is focusing on to evaluate the effects of power distribution skewing on the fuel temperature at the TOP state in the large commercial-base Fast Reactor (JSFR), in order to help streamline the safety evaluation methods of the JSFR. It is shown that the considerations of the power distribution skewing increase the average fuel temperature rise, leading to more negative reactivity feedback which decreases the rate of hot pin temperature rise during TOP, and the peak linear power is significantly mitigated by local feedback reactivities due to temperature rise in the fuel assemblies adjacent to the withdrawn control rod. A conventional calculation model which neither considers the power distribution skewing effect nor the local feedback reactivities effect gives the conservative results of the peak linear power in the TOP event.
Oki, Shigeo; Meiliza, Y.; Kawashima, Katsuyuki; Okubo, Tsutomu
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.561 - 570, 2011/05
Yamaji, Akifumi; Kawashima, Katsuyuki; Oki, Shigeo; Mizuno, Tomoyasu; Okubo, Tsutomu
Nuclear Technology, 171(2), p.153 - 160, 2010/08
Times Cited Count：5 Percentile：35.74(Nuclear Science & Technology)The homogeneous MA loading core with 3wt% MAs is used as a reference design to evaluate the impact of the americium target in-core loading (20wt% MAs) on reactivity characteristics and ULOF response of sodium-cooled MOX-FBR. The Am target loading method of this study can flatten core radial reactivity worth distributions and effectively reduce reactivity insertion into the core during ULOF. As the result, the core power increase speed during ULOF is reduced. The maximum fuel temperature of the target region does not become particularly high compared with that of the inner core and it is much lower than the melting point. It is promising from the viewpoints of the reactivity characteristics and ULOF response.
Maruyama, Shuhei; Kawashima, Katsuyuki; Oki, Shigeo; Mizuno, Tomoyasu; Okubo, Tsutomu
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.1548 - 1556, 2009/09
Kawashima, Katsuyuki; Maruyama, Shuhei; Oki, Shigeo; Mizuno, Tomoyasu
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9288_1 - 9288_7, 2009/05
750 MWe MOX fuel fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters.
Oki, Shigeo; Ogawa, Takashi; Kobayashi, Noboru; Naganuma, Masayuki; Kawashima, Katsuyuki; Maruyama, Shuhei; Mizuno, Tomoyasu; Tanaka, Toshihiko*
Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 10 Pages, 2008/09
Conceptual design studies of sodium-cooled fast reactor core are performed in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan. The representative MOX fuel core and the metal fuel core exert excellent performances on safety and reliability, sustainability, economic competitiveness, and nuclear non-proliferation. This paper reviews their feature in terms of reactor physics, and describes recent progress in design studies. In the recent design studies, much interest has been taken in the fuel composition change in the transition stage from light water reactors to fast breeder reactors. The core flexibility is also shown to fulfil the refined objectives such as high breeding and an enhancement of non-proliferation property.
Kawashima, Katsuyuki; Mizuno, Tomoyasu
JAEA-Research 2008-047, 12 Pages, 2008/04
JAEA-Research-2008-047.pdf:2.5MB
As a part of the FaCT Project, the local feedback reactivities effect on power distributions during control rod withdrawal was studied for a large fast reactor. In the UTOP transient analysis based on the point reactor kinetics model, power distribution is usually given by the steady-state core neutronics calculation performed prior to the transient analysis. However, the power distribution during transient could be different from the one in the steady-state core condition because of local feedback reactivities effect induced by the temperature rise in the fuel assemblies adjacent to the withdrawn control rod. The core calculation model is configured to take into account the transient temperature rise. It is shown that the core radial power distribution skewing is significantly mitigated by considering the local feedback reactivities due to temperature rise. The calculated maximum power density ratio is 1.30, compared with 1.39 calculated under the steady-state core condition.
; 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 2000-068, 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.
Kawashima, Katsuyuki; Ikegami, Tetsuo
PNC TN9410 98-038, 42 Pages, 1998/03
The fast neutron fluence plays the major role in limiting the higher fuel burnup in large fast reactor cores with austenitic stainless steel cladding and duct. Therefore, the core design considerations are performed to reduce the peak fast fluence to average burnup ratio (PFB ratio). The important findings include as follows: (1) Key design parameters are defined which are related to the PFB ratio, and the formula for evaluating the PFB ratio of the fuel is established as: PFB ratio 
power peaking factor/fissile enrichment 
irradiation time / core averaged irradiation time. (2) Available design methods to reduce the PFB ratio include the flux shaping by the operating control rods and the decrease of irradiation time due to the multiple batch refueling scheme. (3) These design methods are applied to the 1000 MWe homogeneous core design, resulting in the 15-18% reduction in the PFB ratio without significant penalties on the other core characteristics such as peak power density and discharge burnup.
Miki, Kazuyoshi*; Kawashima, Katsuyuki*; Inoue, Kotaro*
Journal of Nuclear Science and Technology, 18(1), p.71 - 73, 1981/01
Kawashima, Katsuyuki; Mizuno, Tomoyasu
no journal, ,
A local feedback reactivities effect on power distributions during control rod withdrawal is analyzed in the sodium-cooled large fast reactor core. In the steady-state core model simulating the transient overpower conditions, core radial power distribution skewing is mitigated by considering the reactivities due to the temperature rise in the fuel assemblies neighboring the withdrawn control rod.
