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; ; Hayashi, Kenji; Momoi, K.
Nuclear Engineering and Design, 200, p.157 - 175, 2000/00
Times Cited Count：28 Percentile：84.32(Nuclear Science & Technology)None
Kimura, Nobuyuki; Nishimura, Motohiko; Momoi, K.; Hayashi, Kenji; Kamide, Hideki
PNC TN9410 97-046, 69 Pages, 1997/04
To enhance reliability and safety of FBR, taking advantage of its inherent properties is of importance. From this point of view, natural circulation decay heat removal (NC/DHR) has been studied in which no active components such as pumps are used. DRACS (Direct Reactor Auxiliary Cooling System) is an option of NC/DHR systems. which causes cold coolant flow from DHX (Direct Heat Exchanger) penetrating into inter wrapper gaps: IWF (Inter-Wrapper Flow). Another option for NC/DHR is to use PRACS (Primary Reactor Auxiliary Cooling System) in which no remarkable IWF occurs. Thermal-hydraulic behavior in the core depends on interactions among auxiliary reactor cooling system, IHX (Intermediate Heat Exchanger), and the secondary loop during NC/DHR. Such interactions have been studied with the test rig called PLANDTL-DHX equipped with DRACS and PRACS. In this study, influence of operating condition of the auxiliary cooling systems and the secondary loop of IHX were examined on the core thermal-hydraulic bchaviors. In the present paper, one-dimensional network analyses using LEDHER code are reported. The analyses were performed on steady tests using two models: a model taking account of an inter-subassembly heat transfer only, and a model simulated both IWF and the inter-subassembly heat transfer. The calculation method was validated through comparisons with the experimental results. In the cases cooled by PRACS or IHX, two calculated models showcd good agreements with the experiments regarding the natural circulation flow rate, the temperature distribution in the core and the temperature at the inlet/outlet of the heat exchangers. However, in the case cooled by DRACS operated, the model without flow pass of IWF could not simulate the experiments with respect to the natural circulation flow rate (12% larger than experiment) and temperature profiles in the inter wrapper gaps. On the other hand, the model taking account of IWF simulated the experiments in good ...
Momoi, K.; Hayashi, Kenji; Kamide, Hideki; Nishimura, Motohiko; Kokaki, Nobuhisa
PNC TN9410 97-047, 93 Pages, 1997/03
The evaluation of core thermohydraulics under natural circulation conditions is of significance in order to utilize passive safety features of fast reactors. When the heat exchangers of the decay heat removal system are operated in the upper plenum of a reactor vessel, cold sodium provided by the heat exchangers can penetrate into the gap regions between fuel subassemblies; thiS natural convection phenomenon is called inter-wrapper flow (IWF). During natural circulation decay heat removal, IWF will significantly modify the flow and temperature distributions in the subassemblies. IWF can decrease the temperature in the subassemblies. On the other hand, the natural circulation head will be reduced by temperature reduction in the upper non-heated section of subassemblies due to the IWF cooling. These positive and negative effects of IWF are our main concerns in this report. Sodium experiments were carried out to investigate these phenomena. In a test section, seven subassemblies are bundled and connected to an upper plenum with a heat exchanger. The experiments were carried out under steady state conditions. Experimental parameters were power in the core and flow resistance in the primary loop. Decrease of natural circulation flows in the subassemblies were recognized. Inter-subassembly flow redistribution was also seen due to larger cooling in outer 6 subassemblies and smaller cooling in the center subassembly. In the extremely low flow conditions (large flow resistance in the primary loop), reverse flow was registered in 2 or 3 outer subassemblies. Cooling effect of IWF was also observed. It consisted of direct cooling through the wrapper tube, flow redistribution among subassemblies (higher flow rate for hotter subchannel), and cold reverse flow from the upper plenum. When the flow resistance was small in the primary loop, i.e., flow rate was larger than 1% of reactor rated conditions (based on subassembly averaged flow velocity), the cooling effects and ...
Hayashi, Kenji; Momoi, K.; Nishimura, Motohiko; Kamide, Hideki
PNC TN9410 97-045, 68 Pages, 1997/03
Steady state sodium experiments were performed to investigate interactions between the heat transport systems, i.c., the primary system, the secondary system, and the decay heat removal system, during the natural circulation decay heat removal in FBRs. The PLANDTL-DHX test rig was used for the experiments. The core model has seven subassemblies; the center assembly simulates pin bundle geometry of a core fuel subassembly in a large scale FBR and consists of 37 pins, six outer subassemblies consists of 7 pins. As the decay heat removal system, Dirtct Reactor Auxiliary Cooling System (DRACS) and Primary Reactor Auxiliary Cooling System (PRACS) can be selected. Experiments were carried out under natural circulation conditions in the primary loop and force convection conditions in the decay heat removal system. In cases using DRACS, natural circulation flow rate in the primary loop was smaller by 20% than that in cases using PRACS due to the low temperature in the upper plenum and also in the upper non-heated section of the core. When natural circulation was allowed in the secondary heat transport system, the natural circulation flow rate in the primary system increased in spite of the operation of DRACS. In cases using DRACS, inter-subassembly flow redistribution occurred; the center subassembly had larger flow rate than those in outer subassemblies due to the low natural circulation head in the outer subassemblies which were cooled by the inter-wrapper flow (IWF). The highest temperature in the core was reduced by IWF via not only the direct cooling effect but also the inter-subassembly flow redistribution. Temperature fluctuations around the PRACS cooling coil installed in the IHX were registered under the natural circulation conditions in the primary system. The amplitude of fluctuation was less than 20
C and small on the points of structural integrity.
; Kamide, Hideki; Hayashi, Kenji; Momoi, K.
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.2, p.903 - 913, 1997/00
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Momoi, K.; Hayashi, Kenji; Nishimura, Motohiko; Kamide, Hideki
PNC TN9410 96-280, 146 Pages, 1996/10
The complicated thermal hydraulics which was not observed in forced circulation occurs in the core during natural circulation decay heat removal in Fast Breeder Reactors (FBRs). Especially, in a FBR which has the auxiliary cooling system of DRACS (Direct Reactor Auxiliary Cooling System) type with direct heat exchangers (DHXs) immersed in the reactor plenum, cold sodium provided by the DHXs penetrates into inter-wrapper gaps in the core. This phenomena called Inter-wrapper Flow (IWF) will influence the thermal hydraulics in the core. Further, thermal-interaction between the cooling systems makes the natural circulation flow rate in the primary loop change and will influence the thermal hydraulics in the core. The purpose of this study is to grasp the interaction between the cooling systems and its influence on the thermal hydraulics in the core during natural circulation decay heat removal in FBRs. Transient Sodium experiments which simulated transitions from forced to natural circulation in reactors were performed under several kinds of operating conditions of the IHX secondary system and of the auxiliary cooling system using PLANDTL-DHX test facility with DRACS and PRACS (Primary Reactor Auxiliary Cooling System). In the DRACS, the heat removed by natural circulation in the IHX secondary system influenced the natural circulation flow rate in the primary loop more than the operation conditions of DRACS. When natural circulation in the IHX secondary loop was stopped, the IHX outlet temperature rapidly increased and the natural circulation flow rate in the primary loop decreased. In addition, reverse flows were detected in the outer subassemblies. The temperature rise in the heated length of the center subassembly was reduced by the cooling effect of the inter-subassembly heat transfer to the reverse flow subassemblies and IWF. When the natural circulation in the IHX secondary loop was continued, the natural circulation flow rate of 1% level of rated ...
Kamide, Hideki; Nishimura, Motohiko; Hayashi, Kenji; Momoi, K.; Miyake, Yasuhiro*
PNC TN9410 96-268, 79 Pages, 1996/09
It is considerably effective to utilize the natural circulation on advance of reliability of the decay heat removal systems of Fast Breeder Reactors. The natural circulation dose not depend on components which needs external power sources like pumps. This increases the reliability of the decay heat removal systems. However, thermohydraulics in the core have complex characteristics under the natural circulation. Under low flow conditions, buoyancy effects and heat transfer from high temperature subassemblies to low temperature subassemblies, i.e. inter-subassembly heat transfer, will significantly modify the flow and temperature distributions in the subassemblies. Thus, development of an evaluation method for the core thermohydraulic is significant to utilize the natural circulation. A multi-subassembly analysis method using the three-dimensional thermohydraulic analysis code, AQUA, was developed to predict thermohydraulics in the subassemblies with influence of the inter-subassembly heat transfer. In this method, each subassembly is modeled in individual mesh region of multi-region model of AQUA and the staggered half-pin mesh arrangement was applied in each subassembly. The heat transfer between the subassemblies was simulated by a thermal structure model. The analysis method was applied to two sodium experiments where three or seven subassemblies were modeled in simulated cores. The experimental analyses showed that the multi-subassembly analysis method could evaluate the thermohydraulics in the subassemblies.
Fujiwara, Shizuo*; Arata, Yoji*; Hayakawa, Naohiro; Momoi, Hironao*
Bulletin of the Chemical Society of Japan, 35(10), 1658 Pages, 1962/00
Times Cited Count：14no abstracts in English
Kamide, Hideki; Hayashi, Kenji; Momoi, K.
International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Vol. 2, ,
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Kamide, Hideki; Hayashi, Kenji; ; Momoi, K.
Proceedings of 2nd International Topical Meeting on Advanced Reactors Safety (ARS '97), Vol.2, ,
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