Knowledge from recent investigations on sloshing motion in a liquid pool with solid particles for severe accident analyses of sodium-cooled fast reactor

The helium-cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected for further development in the frame of the Generation IV International Forum (GIF). Since no gas cooled fast reactor has ever been built, a small demonstration reactor is necessary on the road towards the full-scale GFR reactor. A concept of this demonstrator is called ALLEGRO. The French Commissariat à l’énergie atomique et aux énergies alternatives (CEA) developed between 2001–2009 a pre-conceptual design of both the full-scale GFR called GFR2400 and the small demonstration unit called ALLEGRO (75 MWt). Since 2013 ALLEGRO has been under development by several partners from Czech Republic, France, Hungary, Poland and Slovakia. No severe accident study of ALLEGRO using a dedicated computer code has been published so far. This paper is the first attempt to perform computer simulations of the ALLEGRO CEA 2009 concept, using MELCOR version 2.1. A model of the ALLEGRO CEA 2009 concept has been developed with the aim to perform safety analyses; to confirm that MELCOR can be used for such a study, to investigate what scenarios lead to a severe accident and to study in detail the progression of the severe accident during the in-vessel phase. Several pressurized and depressurized protected scenarios were investigated; four of them are presented in this paper. It was observed that even long-lasting station blackout (SBO) without further failures of the passive safety systems does not lead to a severe accident as long as there is enough water in the decay heat removal (DHR) system. Loss of coolant (LOCA) transients with DHR system in the forced-convection mode can lead to peak cladding temperatures causing limited core damage in the early phase of the accidents, but without further development into core meltdown. On the other hand, LOCA combined with SBO leads to excessive core melting in orders of minutes, which represents a weak point of ALLEGRO 2009 concept. Recommendations were formulated for the further development of the ALLEGRO concept.

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Numerical Investigation of Corium Coolability in Core Catcher: Sensitivity to Modeling Parameters

Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation ◽

10.1115/icone26-81841 ◽

2018 ◽

Cited By ~ 1

Author(s):

Liancheng Guo ◽

Andrei Rineiski

Keyword(s):

Fast Reactor ◽

Large Scale ◽

Mass Transfer Process ◽

Severe Accident ◽

Liquid Sodium ◽

Numerical Experience ◽

The Core ◽

Sodium Fast Reactor ◽

Core Catcher ◽

Fuel Pin

To avoid settling of molten materials directly on the vessel wall in severe accident sequences, the implementation of a ‘core catcher’ device in the lower plenum of sodium fast reactor designs is considered. The device is to collect, retain and cool the debris, created when the corium falls down and accumulates in the core catcher, while interacting with surrounding coolant. This Fuel-Coolant Interaction (FCI) leads to a potentially energetic heat and mass transfer process which may threaten the vessel integrity. For simulations of severe accidents, including FCI, the SIMMER code family is employed at KIT. SIMMER-III and SIMMER-IV are advanced tools for the core disruptive accidents (CDA) analysis of liquid-metal fast reactors (LMFRs) and other GEN-IV systems. They are 2D/3D multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics codes coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. However, the experience of SIMMER application to simulation of corium relocation and related FCI is limited. It should be mentioned that the SIMMER code was not firstly developed for the FCI simulation. However, the related models show its basic capability in such complicate multiphase phenomena. The objective of the study was to preliminarily apply this code in a large-scale simulation. An in-vessel model based on European Sodium Fast Reactor (ESFR) was established and calculated by the SIMMER code. In addition, a sensitivity analysis on some modeling parameters is also conducted to examine their impacts. The characteristics of the debris in the core catcher region, such as debris mass and composition are compared. Besides that, the pressure history in this region, the mass of generated sodium vapor and average temperature of liquid sodium, which can be considered as FCI quantitative parameters, are also discussed. It is expected that the present study can provide some numerical experience of the SIMMER code in plant-scale corium relocation and related FCI simulation.

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ESFR-SMART Core Safety Measures and Their Preliminary Assessment

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4052588 ◽

2021 ◽

Author(s):

Andrei Rineiski ◽

Clément Mériot ◽

Marco Marchetti ◽

Jiri Krepel ◽

Christine Coquelet ◽

...

Keyword(s):

Fast Reactor ◽

Severe Accident ◽

Minor Actinides ◽

Preliminary Assessment ◽

Safety Measures ◽

Horizon 2020 ◽

The Core ◽

Severe Accidents ◽

Sodium Fast Reactor ◽

Void Effect

Abstract A large 3600 MW-thermal European Sodium Fast Reactor (ESFR) concept has been studied in Horizon-2020 ESFR-SMART (ESFR Safety Measures Assessment and Research Tools) project since September 2017, following an earlier EURATOM project, CP-ESFR. In the paper, we describe new ESFR core safety measures focused on prevention and mitigation of severe accidents. In particular, we propose a new core configuration for reducing the sodium void effect, introduce passive shutdown systems, and implement special paths in the core for facilitation of molten fuel discharge in order to avoid re-criticalities after a hypothetical severe accident. We describe and assess the control and shutdown system, and consider options for burning minor actinides.

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Conceptual Study on Recriticality Prevention Core Having Duplex Pellets with Neutron Absorber in Outer Core in a Fast Reactor

Science and Technology of Nuclear Installations ◽

10.1155/2019/2753789 ◽

2019 ◽

Vol 2019 ◽

pp. 1-6

Author(s):

Toshio Wakabayashi ◽

Makoto Takahashi ◽

Naoyuki Takaki ◽

Yoshiaki Tachi ◽

Mari Yano

Keyword(s):

Fast Reactor ◽

Reactor Core ◽

Thermal Characteristics ◽

Outer Core ◽

Severe Accident ◽

Monte Carlo Code ◽

Core Concept ◽

Neutron Absorber ◽

The Core ◽

Nuclear Data Library

In a fast reactor, we evaluated a new core concept that prevents severe recriticality after whole-scale molten formation in a severe accident. A core concept in which Duplex pellets including neutron absorber are loaded in the outer core has been proposed. Analysis by the continuous energy model Monte Carlo code MVP using the JENDL-4.0 nuclear data library revealed that this fast reactor core has large negative reactivity due to fuel melting at the time of a severe accident, so that the core prevents recriticality. Regarding the core nuclear and thermal characteristics, the loading of Duplex pellets including neutron absorber in the outer core caused no significant differences from the normal core without Duplex pellets.

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