Sodium Experiments of Buoyancy-Driven Penetration Flow into Low-Power Subassemblies in a Sodium-Cooled Fast Reactor During Natural Circulation Decay Heat Removal

Kamide, Hideki; Kobayashi, Jun; Hayashi, Kenji

doi:10.13182/nt11-a12511

Cited by 8 publications

(3 citation statements)

References 7 publications

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“…Because sodium's heat conductivity is large, the temperature fluctuation around the interface of the thermal stratification in the upper plenum affects the RV's components and wall. Other phenomena are core-plenum interactions (Kamide, et al, 1994;Nishimura, et al, 2000) comprising the penetration flow into the fuel subassembly, the interwrapper flow (IWF), and the radial heat transfer between neighboring fuel subassemblies. The penetration flow (Kamide, et al, 1997(Kamide, et al, , 2009 means cold sodium provided from the D-DHX covers the top of the core and penetrates the fuel subassemblies from their outlets.…”

Section: Introductionmentioning

confidence: 99%

“…Other phenomena are core-plenum interactions (Kamide, et al, 1994;Nishimura, et al, 2000) comprising the penetration flow into the fuel subassembly, the interwrapper flow (IWF), and the radial heat transfer between neighboring fuel subassemblies. The penetration flow (Kamide, et al, 1997(Kamide, et al, , 2009 means cold sodium provided from the D-DHX covers the top of the core and penetrates the fuel subassemblies from their outlets. At the lower power fuel subassemblies under low flowrate conditions in NC, the cold sodium flow penetrates the upper neutron shielding region due to the strong negative buoyancy force.…”

Section: Introductionmentioning

confidence: 99%

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Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Hamase

Miyake

Imai

et al. 2022

Mechanical Engineering Journal

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A direct reactor auxiliary cooling system (DRACS) under natural circulation (NC) conditions with a dipped-type direct heat exchanger (D-DHX) in the upper plenum of a reactor vessel (RV) has been investigated for enhancing the safety of sodium-cooled fast reactors. Studies of the past have revealed that core-plenum interactions, which consists of penetration of the coolant from D-DHXs into the subassemblies and the narrow gap between them (IWF: inter-wrapper flow), and the heat transfer through a wrapper tube among subassemblies (radial heat transfer), occurred and increased core cooling performance during the DRACS operation. Therefore, a multidimensional thermal-hydraulics analysis model in the RV using a computational fluid dynamics (CFD) code (RV-CFD model) was developed to evaluate the core cooling performance. For the design study, the RV-CFD model must simulate reasonable calculation costs while maintaining accuracy. In this study, the subchannel analysis method using the CFD code for fuel subassemblies (subchannel CFD model) was applied to the RV-CFD model. In the subchannel CFD model, the porous media approach was used to consider local geometry in the fuel subassembly, and the effective heat conductivity coefficients in a diffusion term of the energy equation were set to fit the actual radial thermal diffusion between subchannels. Two numerical simulations were compared to the experimental data obtained from the sodium experimental apparatus PLANDTL-1. In the first case, the focus was only the radial heat transfer without the D-DHX operation. In another case with the D-DHX operation, the IWF noticeably occurred, and the focus was on the core-plenum thermal interaction. The calculated sodium temperature in the core correlated well with the experimental results. The RV-CFD with subchannel CFD model was validated for core-plenum interactions during the DRACS with the D-DHX operation under NC conditions.

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Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Hamase

Miyake

Imai

et al. 2022

Mechanical Engineering Journal

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“…It is needed to predict the thermal-hydraulic in the plant not only under the normal operating conditions but also under the severe accident (Kamide, 2017a;Kamide, 2017b). Japan Atomic Energy Agency (JAEA) is implementing the experimental and numerical study on thermal-hydraulics for the sodium-cooled fast reactor plant to establish the evaluation method based on the numerical simulation to confirm the coolability of the decay heat removal system (DHRS) which is driven by the natural circulation under both the normal operating condition and the severe accident conditions Kamide, 2001;Kamide, 2011;Ono, 2016;Watanabe, 2015). The sodium test loop, the PLANDTL-2, was built to investigate the thermal-hydraulics in the sodium-cooled fast reactor, which consists of the simulated core, two types of decay heat removal system, the secondary loop, the air coolers.…”

Section: Introductionmentioning

confidence: 99%

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono

Tanaka

Miyake

et al. 2020

Mechanical Engineering Journal

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Fully natural circulation decay heat removal systems (DHRSs) are adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is needed to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. The transient analysis simulating the reactor scram reveals that the 3-dimensional large scale flow structure is developed through the gaps in the whole of the core area during the reactor scram. The steady-state analysis coinciding Richardson number between the PLANDTL-2 and the reactor operation condition reveals that the hot spot and cold spot appear depending on the location of the DHX, which is caused by the complex thermalhydraulic phenomena driven by the natural circulation. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

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Key technologies for future sodium-cooled fast reactors

Aizawa

Ando

Chikazawa

et al. 2022

Sodium-Cooled Fast Reactors

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Sodium Experiments of Buoyancy-Driven Penetration Flow into Low-Power Subassemblies in a Sodium-Cooled Fast Reactor During Natural Circulation Decay Heat Removal

Cited by 8 publications

References 7 publications

Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Core thermal-hydraulics analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Key technologies for future sodium-cooled fast reactors

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