On the Multiple-Pin Modeling of the Fuel Bundle for the Simulation of the Initiating Phase of a Severe Accident in a Sodium Fast Reactor

2014 ◽  
Vol 178 (2) ◽  
pp. 202-224 ◽  
Author(s):  
M. Guyot ◽  
P. Gubernatis ◽  
C. Suteau
Keyword(s):  
Fast Reactor ◽  
Severe Accident ◽  
Sodium Fast Reactor ◽  
Fuel Bundle

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

2018 ◽  
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.

scholarly journals ESFR-SMART Core Safety Measures and Their Preliminary Assessment

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.

An Assessment of Corewide Coherency Effects in the Multichannel Modeling of the Initiating Phase of a Severe Accident in a Sodium Fast Reactor

2014 ◽  
Vol 185 (1) ◽  
pp. 21-38 ◽  
Author(s):  
M. Guyot ◽  
P. Gubernatis ◽  
C. Suteau ◽  
R. Le Tellier ◽  
J. Lecerf
Keyword(s):  
Fast Reactor ◽  
Severe Accident ◽  
Sodium Fast Reactor

Turbulent Transverse Plane PIV Measurements on a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle

2018 ◽  
Author(s):  
Nolan E. Goth ◽  
Philip Jones ◽  
Thien D. Nguyen ◽  
Rodolfo Vaghetto ◽  
Yassin Hassan
Keyword(s):  
Fluid Dynamics ◽  
Experimental Data ◽  
Computational Fluid Dynamics ◽  
Fast Reactor ◽  
Measurement Techniques ◽  
Transverse Plane ◽  
Index Of Refraction ◽  
Plane Flow ◽  
Sodium Fast Reactor ◽  
Fuel Bundle

This study recaps the experimental effort to characterize a transverse plane flow field in a tightly-packed, 61-pin, wire-wrapped, hexagonal fuel bundle prototypical for a sodium fast reactor. The motive was to produce high spatiotemporal experimental data for computational fluid dynamics turbulence model validation. The matched-index-of-refraction and laser-based optical measurement techniques were utilized on an isothermal experimental flow facility. Measurements were performed on a transverse plane perpendicular to the axial flow. Fluid flow in the three types of subchannels (corner, exterior, and interior) were quantified. All measurements have been performed at a bundle-averaged Reynolds number of approximately 10,400. Results include flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, and Reynolds stress. Of interest was the flow behavior around the restriction caused by the wire spacer and hexagonal duct wall, where recirculation regions formed. Regions of maximum and minimum momentum transfer coincided with regions of maximum and minimum fluctuations. These regions highlight locations of maximum and minimum cooling of the fuel pins. The experimental data will be used to benchmark computational fluid dynamics simulations of the sodium fast reactor fuel bundle using Reynolds-averaged Navier Stokes and large-eddy simulation turbulence modeling methods.

Preliminary Analysis of In-Vessel Corium Confinement and Cooling in a Large Sodium Fast Reactor

2012 ◽  
Author(s):  
Franco Polidoro ◽  
Flavio Parozzi
Keyword(s):  
Fast Reactor ◽  
Preliminary Analysis ◽  
Structural Integrity ◽  
Mechanical Energy ◽  
Severe Accident ◽  
Sodium Fast Reactor ◽  
Safety Margins ◽  
Core Catcher ◽  
Core Meltdown ◽  
Accident Conditions

Considering a reasonable range of core meltdown accidents that can be postulated for GenIV sodium fast reactors, good safety margins exist for corium confinement and cooling inside the reactor vessel. Coolable conditions can be reached with the adoption of an ad-hoc device in the lower plenum, i.e. core catcher, capable to intercept the downward motion of the molten material and assure its cooling. Such device has to be designed to withstand to extreme thermal-mechanical conditions that rise as consequence of the large mechanical energy release and high temperature of molten corium. As this study has been carried out in the frame of the Collaborative Project on European Sodium Fast Reactor (CP ESFR) of the 7th Framework Programme Euratom, on the basis of the postulated accident conditions assumed for a reference 1500 MWe pool-type sodium fast reactor, the present work provides a preliminary analysis of the thermal response of a possible core catcher placed within the vessel. The dynamic thermal behaviour of the corium-structure-coolant system is analyzed with the computer code CORIUM-2D, an original simulation tool developed by RSE - Ricerca Sul Sistema Energetico, with the aim to assess the thermal interaction among corium, structures and coolant under severe accident conditions in both Light Water Reactors (LWRs) and Liquid Metal Fast Breeder Reactors (LMFBRs). The results of the numerical simulations show that the steady-state coolable configuration of core debris and the structural integrity of main containment structures can be reached in a number of partial core meltdown situations.

The Study About the Transient Characteristics of Metallic Fuel Sodium Cooled Fast Reactor in the Unprotected Loss of Flow Accident Based on the SAS4A Code

2017 ◽  
Author(s):  
Pengrui Qiao ◽  
Wenjun Hu
Keyword(s):  
Fast Reactor ◽  
Reactor Power ◽  
Severe Accident ◽  
Inherent Safety ◽  
Metallic Fuel ◽  
Reactor Accident ◽  
Transient Characteristics ◽  
Sodium Fast Reactor ◽  
The Us ◽  
Development Direction

Unprotected loss of flow accident (ULOF) is the most typical severe accident in sodium cooled fast reactor, which is focused by scholars civil and abroad. Metallic fuel has different safety characteristics with the oxide fuel as the important development direction of future sodium fast reactor, accident analysis of which is also a research focus at home and abroad. This paper bases on one Cooperation Research Project proposed by ANL and organized by IAEA, analyses the Shut-down Removal Test-45R of the metallic fuel sodium cooled fast reactor EBR-II in the US with SAS4A code, to research the transient characteristics of it in ULOF accident. Studies have shown that, metallic fuel sodium cooled fast reactor has very good inherent safety performance, which can reduce the reactor power in ULOF accident through the negative feedback itself.

COMPUTATIONAL FLUID DYNAMIC INVESTIGATIONS ON EFFECTS OF CIRCULAR ROD INSERTS IN FAST REACTOR FUEL BUNDLE

2018 ◽  
Author(s):  
G Padmakumar ◽  
K. Velusamy ◽  
Bhamidi V. S. S. S. Prasad ◽  
P Lijukrishnan ◽  
P. Selvaraj
Keyword(s):  
Fast Reactor ◽  
Computational Fluid Dynamic ◽  
Fluid Dynamic ◽  
Reactor Fuel ◽  
Fuel Bundle ◽  
Fast Reactor Fuel
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