Safety Strategy of JSFR Eliminating Severe Recriticality Events and Establishing In-Vessel Retention in the Core Disruptive Accident

Sato, Ikken; Tobita, Yoshiharu; Konishi, Kensuke; Kamiyama, Kenji; Toyooka, Jun-ichi; Nakai, Ryodai; Kubo, S.; Kotake, Shoji; Koyama, Kazuya; Vassiliev, Yuri S.; Vurim, Alexander; Zuev, V. A.; Kolodeshnikov, Alexander A.

doi:10.1080/18811248.2011.9711733

Cited by 48 publications

(10 citation statements)

References 5 publications

(2 reference statements)

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“…The heat fluxes from the melt to the wall were very high in the EAGLE in-pile experiments (Sato, et al, 2011). Although this heat flux level was similar to the SCARABEE BE+3 experiment, the difference between the ID1/ID2 and BE+3 experiments were the formation of molten fuel pool by the transient and the steady-state nuclear heating, respectively.…”

Section: Experimental Database 31 Duct-wall Failure Behaviormentioning

confidence: 55%

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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Section: Experimental Database 31 Duct-wall Failure Behaviormentioning

confidence: 55%

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

Yamano

Tobita

2014

Mechanical Engineering Journal

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“…In the early stage of CDA such as 10s, there would not be issue since near half of all the fuel is coolable on the core catcher even in the most pessimistic values of the porosity and particle diameter of the debris bed. To more suppress the fuel debris accumulation on the core catcher, installing a modified FAIDUS (Fuel Assembly with Inner DUct Structure) (Suzuki et al, 2014;Sato et al, 2011) and/or a CFV (low sodium void worth) core (Serre et al, 2017), etc. could be useful.…”

Section: Demonstration Of Parametric Calculations Using Annmentioning

confidence: 99%

“…One of the future generation Sodium-cooled Fast Reactor (SFR) design strategies is to achieve In-Vessel Retention (IVR) for Core Disruptive Accident (CDA) (Suzuki et al, 2014;Sato et al, 2011). In CDA, molten fuel is discharged from the core region and becomes solidified particle debris with a fuel-coolant interaction.…”

Section: Introductionmentioning

confidence: 99%

Study on application of artificial neural network to debris bed coolability calculations for sodium-cooled fast reactors

Matsuo¹,

Sasa²,

Saito³

et al. 2020

Mechanical Engineering Journal

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Understanding the effect of uncertainties of Core Disruptive Accident (CDA) scenarios on debris bed coolability on a core catcher is required for decision making on design options to mitigate a CDA consequence. For the understanding, a huge number of calculations are required but are extremely difficult to perform because a huge number of calculations require much calculation time to solve non-steady equations in the coolability calculation model. Thus, we applied Artificial Neural Network (ANN), which is one of models for machine learning, to debris bed coolability calculations. The application of ANN is expected to exponentially improve the calculation speed of debris bed coolability because ANN provides results from experimental rules learned through training without solving non-steady equations. The application is in three steps. Firstly, we created many data for training ANN and validating the trained ANN through coolability calculations parameterizing main dominant inputs (particle diameter of debris bed, porosity of debris bed, etc.) by using Latin hypercube sampling. Secondly, ANN was trained and validated with the created data. The accuracy rate of the results by the ANN to the validation data exceeded 99%. In addition, the calculation time using ANN was micro seconds order. Finally, through demonstration calculations, it was confirmed that we can easily understand the effect of uncertainties of CDA scenarios on debris bed coolability owing to results visualization based on a huge number of parametric calculations using ANN. Thus, the application of ANN to debris bed coolability calculations should contribute to the decision making on design options to mitigate a CDA consequence.

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“…Elimination of the severe power burst events in the Core Disruptive Accident (CDA) is intended [7]. The design strategy for it is to control the potential of excessive void reactivity insertion in the initiating phase of CDA by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation by introducing an inner duct in a fuel subassembly as shown in Figure 1 to discharge molten fuel from the core on the other hand.…”

Section: Examples Of Safety Approach In Countries Developing Fast Reamentioning

confidence: 99%

Review of safety improvement on sodium-cooled fast reactors after Fukushima accident

Takeda¹,

Shimazu²,

Foad³

et al. 2012

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Several countries are developing and deploying SFRs even after the accident at Tokyo Electric Power Company’s Fukushima Dai-Ichi Nuclear Power Station. However, the Fukushima accident prompted all countries to redefine the fast reactor programs. The drastic safety enhancement is the most important issue to be established. In light of this situation, key essence of the safety improvement is reviewed in this paper by referring the achievements of the recent International Workshop on Prevention and Mitigation of Severe Accidents in SFRs which was held by Japan Atomic Energy Agency (JAEA) in cooperation with the International Atomic Energy Agency (IAEA) in June, 2012 and the findings published in the past journals including those of the International Conference on Fast Reactor and Related Fuel Cycles (FR09) held by IAEA in December, 2009

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Safety Strategy of JSFR Eliminating Severe Recriticality Events and Establishing In-Vessel Retention in the Core Disruptive Accident

Cited by 48 publications

References 5 publications

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

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

Study on application of artificial neural network to debris bed coolability calculations for sodium-cooled fast reactors

Review of safety improvement on sodium-cooled fast reactors after Fukushima accident

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