Experimental Analyses by SIMMER-III on Debris-Bed Coolability and Metallic Fuel Freezing Behavior

Yamano, Hidemasa; Tobita, Yoshiharu

doi:10.1299/jpes.5.2

Cited by 3 publications

(2 citation statements)

References 22 publications

(17 reference statements)

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“…Fundamental studies concerning numerical methods are performed to support the code development of COMPASS. The past results of the present project can be found in the references (Koshizuka et al, 2006(Koshizuka et al, , 2007(Koshizuka et al, , 2008a(Koshizuka et al, , 2008b(Koshizuka et al, , 2009a(Koshizuka et al, , 2009b(Koshizuka et al, , 2009cYamano and Tobita, 2009, 2010a, 2010bZhang et al, 2009Zhang et al, , 2010Yamamoto et al, 2008Yamamoto et al, , 2009Shirakawa et al, 2009;Uehara et al, 2009;Himi et al, 2008Himi et al, , 2009.…”

Section: Mps (Moving Particle Semi-implicit) Methods Was Developed To mentioning

confidence: 87%

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Morita

Zhang

Koshizuka

et al. 2011

Nuclear Engineering and Design

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A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include 1) fuel pin failure and disruption, 2) molten pool boiling, 3) melt freezing and blockage formation, 4) duct wall failure, 5) low-energy disruptive core motion, 6) debris-bed coolability, and 7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. Simulations of GEYSER and THEFIS experiments were performed for dispersion 2 and freezing behaviors of molten materials in narrow flow channels. In particular, the latter experiment using melt-solid mixture is also related to fundamental behavior of low energy disruptive core. CABRI-TPA2 experiment was simulated for boiling behavior of molten core pool. Expected mechanism of heat transfer between molten fuel and steel mixture was reproduced by the simulation. Analyses of structural dynamics using elastoplastic mechanics and fracture criteria were performed for SCARABEE BE+3 and CABRI E7 experiments. These two analyses are especially focused on thermal and mechanical failure of steel duct wall and fuel pin, respectively. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.

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Section: Mps (Moving Particle Semi-implicit) Methods Was Developed To mentioning

confidence: 87%

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Morita

Zhang

Koshizuka

et al. 2011

Nuclear Engineering and Design

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“…．Lipinski は，沸騰領域におけるドライアウト熱流束を評価するための１点モデル及び 1 次元モデルを提案した (13) ．Nakamura と Lipinski は，Na 冷却材中で生成する小径粒子のデブリベッドにおいては，Lipinski の 1 次元モデルのうち，乱流項と重力が無視できると考え，基礎式からそれらを削除することにより，沸騰を伴うデブリベッドの冷却挙動を２次元的に評価可能な DEBRIS-MD コードを提案した (14) ．DEBRIS-MD と同様の仮定に基づく 1 次元モデルをプラントフローネットワークモデルと組合せた解析コード (15) も報告されている．Stubos らは，毛管力及び層流に加え，重力を考慮し，さらに沸騰領域における温度と飽和度の関係を，沸騰開始時とドライアウト時のデブリベッド温度と飽和度を適切に平準化した関数により結ぶことで表現し，沸騰領域の拡大とディスターバンスの開始を予測可能なモデルを報告した (16) ．その他，SIMMER-Ⅲコードによる ACRR-D10 試験解析 (17) なども報告されている．…”

Section: 緒言unclassified

Study on Debris Bed Cooling Model for Sodium-Cooled Fast Reactor

Matsuo¹

2013

Trans.JSME, Ser.B

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This paper describes a debris-bed cooling model for a Sodium-Cooled Fast Reactor (SFR). The model aims at addressing not only the debris-bed consisted of the particles of the diameter with the hundreds of micrometers typically assumed with SFR but also the debris-bed with larger diameter particles. Then, it can cope with wider range of the scenario in the severe accident of SFR. Pressure losses of a laminar flow and a turbulent flow, gravity force and capillary force are considered in the basic equations. The present study model is validated through the analysis of the Sandia ACRR-D10 in-pile experiments utilizing real materials to simulate the SFR environment. Sensitivity analyses are performed by using the present study model to evaluate the influence of gravity force, turbulent flow pressure loss and the diameter of particles. It is shown from the sensitivity analysis, that gravity force influences the temperature and the liquid saturation in the debris-bed with larger diameter particles, while the turbulent flow pressure loss has small influence with the range of these sensitivity analyses.

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Experimental analyses by SIMMER-III on duct-wall failure and fuel discharge/relocation behavior

Yamano

Tobita

2014

Mechanical Engineering Journal

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This paper describes experimental analyses using SIMMER-III and-IV, which are respectively two-and three-dimensional multi-component multi-phase Eulerian fluid-dynamics codes, for the purpose of integral code validation. Two topics of key phenomena in core disruptive accidents of sodium-cooled fast reactors are presented in this paper: duct-wall failure and fuel discharge/relocation behavior. To analyze the duct-wall failure behavior, the SCARABEE BE+3 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall event progression; which was characterized by coolant boiling, clad melting, fuel failure, molten pool formation, duct-wall failure, etc.; observed in the experiment. The CAMEL C6 experiment investigated the fuel discharge and relocation behavior through a simulated control rod guide tube, which is important in evaluating the neutronic reactivity. SIMMER-IV well simulated fuel-coolant interaction, sodium voiding, fuel relocation behavior observed in the experiment. These experimental analyses indicated the validity of the SIMMER-III and-IV computer code for the duct wall failure and fuel discharge/relocation behavior. : Sodium-cooled fast reactor, Core disruptive accident, Multi-phase flow, Simulation code, SIMMER-III, Wall failure, Fuel discharge, Fuel relocation 1. Introduction The SIMMER-III computer code has been developed and used to analyze the event progression of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs) in Japan Atomic Energy Agency (JAEA) (Tobita, et al., 1999). SIMMER-III is a two-dimensional multi-component multi-phase Eulerian fluid-dynamics code, coupled with fuel pin model and space-dependent neutronics model (Yamano, et al., 2003). This computer code has to simulate complex phenomena in CDAs, such as sodium voiding, fuel and clad melting, duct wall failure, molten core material discharge and relocation, and so on. A comprehensive and systematic assessment (verification and validation) program of the SIMMER-III code has been conducted for fundamental assessment of individual code models (Kondo, et al., 1997) and for integral code assessment (Kondo, et al., 1999). Areas of interests in the latter assessment problems were boiling pool dynamics, fuel relocation and freezing (Kamiyama, et al., 2006), fuel-coolant interaction (FCI) (Morita, et al., 1999), core expansion dynamics and disrupted core neutronics. Especially through sodium experimental calculations, it was demonstrated that SIMMER-III was well validated for simulating transient multi-phase flow phenomena occurring during CDAs (Tobita, et al., 2006). In parallel with the code assessment efforts, a three-dimensional SIMMER-IV code was developed with retaining exactly the same framework in physical models as SIMMER-III (Yamano, et al., 2008). Fuel relocation is very important in the CDA event progression because fuel discharge from the core can significantly mitigate the neutronic reactivity. In particular, effective is fuel discharge through a large-diameter duct, suc...

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Experimental Analyses by SIMMER-III on Debris-Bed Coolability and Metallic Fuel Freezing Behavior

Cited by 3 publications

References 22 publications

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Study on Debris Bed Cooling Model for Sodium-Cooled Fast Reactor

Experimental analyses by SIMMER-III on duct-wall failure and fuel discharge/relocation behavior

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