The result of a wall failure in-pile experiment under the EAGLE project

Konishi, Kensuke; Toyooka, Jun-ichi; Kamiyama, Kenji; Sato, Ikken; Kubo, S.; Kotake, Shoji; Koyama, Kazuya; Vurim, Alexander; Gaidaichuk, V. A.; Pakhnits, Alexander V.; Vassiliev, Yuri S.

doi:10.1016/j.nucengdes.2007.03.012

Cited by 38 publications

(10 citation statements)

References 4 publications

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“…The dispersed fuel is relocated mainly on the intermediate isolation plate (top level of the core). The above key phenomena have been confirmed by experimental study [23]. Figure 9 shows the fuel discharge capability for the modified FAIDUS design during the early discharge phase, which was evaluated by the SIMMER-III code.…”

Section: Cda Evaluationmentioning

confidence: 54%

Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

Yamano

Kubo²,

Shimakawa³

et al. 2012

Science and Technology of Nuclear Installations

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As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR) adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS) in design extension conditions (DECs). A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs) and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

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Section: Cda Evaluationmentioning

confidence: 54%

Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

Yamano

Kubo²,

Shimakawa³

et al. 2012

Science and Technology of Nuclear Installations

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“…By contrast, the discharge of molten fuel through CRGTs would provide negative reactivity feedback. Through the EAGLE [6,7] experimental programs, an effective database has been established for the fuel discharge behaviors in the transition phase. Based on this database, the following aspects have become well understood: (i) thermal loading from disrupted core materials to CRGT structure, (ii) failure of CRGT structure and formation of discharge path, (iii) reasonable treatment of FCI pressure, and (iv) blockage possibility inside CRGTs.…”

Section: Transition Phasementioning

confidence: 99%

“…Although the conservative evaluations are conducted so as to assess the effect of phenomenological or analytical uncertainties, conservative conditions due to such uncertainties are not superposed because of this lower Core Damage Frequency. In the present study, the event progression derived by ULOF is evaluated based on the concept described above, with the knowledge newly obtained through CABRI experiments [4,5] and EAGLE projects [6,7], which are efficiently reflected in the evaluation methodologies and computational analytical tools.…”

Section: Introductionmentioning

confidence: 99%

A preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Suzuki

Tobita

Kawada

et al. 2015

Nuclear Engineering and Technology

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Core disruptive accident (CDA) In-vessel retention (IVR) Sodium-cooled fast reactor (SFR) Unprotected loss of flow (ULOF) SAS4A SIMMER-III/IV a b s t r a c t In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of invessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

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“…The rapid duct wall failure behavior by quickly formed melt that was was observed mainly by thermocouples in the EAGLE in-pile experiments (Konishi, et al, 2006). The WF experiment observed two different wall failure behaviors of gas-gap and sodium-gap walls using about 2 kg of fuel pins (Konishi, et al, 2007). The FD experiment used about 8 kg of fuel pins to generate the molten pool that gave thermal load onto the duct wall without sodium cooling.…”

Section: Experimental Database 31 Duct-wall Failure Behaviormentioning

confidence: 99%

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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The result of a wall failure in-pile experiment under the EAGLE project

Cited by 38 publications

References 4 publications

Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

A preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

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

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