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.

scholarly journals Assessment of Radiation Doses for Lungs due to Some Radionuclides Released from a Hypothetical Nuclear Power Reactor Accident

2015 ◽  
Vol 17 (1) ◽  
pp. 30-37
Author(s):  
Md Moniruzzaman Khan ◽  
AHM Ruhul Quddus ◽  
Mir Md Akramuzzaman ◽  
Abdus Sattar Mollah
Keyword(s):  
Human Body ◽  
Nuclear Power ◽  
Reactor Core ◽  
Reactor Power ◽  
Dose Effect ◽  
Severe Accident ◽  
Reactor Accident ◽  
Nuclear Power Reactor ◽  
North East ◽  
Sensitive Organ

Different radionuclides are emitted from the reactor core after the nuclear accident. These radionuclides are entered into human body through different pathways, which damage the cells. The dose consequence to the sensitive organ like lungs of human body is considered in the present study to show the dose effect for various radionuclides from a hypothetical nuclear reactor accident. The calculations were made with the in-house developed computer program “RaDARRA”. Cardinal directions like E, ENE, ESE, N, NE, NNE, NNW, NW, S, SE, SSE, SSW, SW, W, WNW and WSW are considered to observe the dose effect along the directions. For the calculations, lungs dose arising from 8 radionuclides e.g., 89Sr, 91Y, 95Zr, 95Nb, 131I, 133I, 140Ba and 144Ce have been considered. Of all these radionuclides the maximum and minimum dose contribution mainly come from 144Ce (30%) and 95Nb (4.43%). It is marked that dose is maximum along North East (NE) direction for all the distances and for all types of the radioisotopes. Methodology used in the present study can also be utilized for any type of severe accident and any type of reactor power. DOI: http://dx.doi.org/10.3329/bjnm.v17i1.22489 Bangladesh J. Nuclear Med. 17(1): 30-37, January 2014

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.

Applying the Serpent-DYN3D code sequence for the decay heat analysis of metallic fuel sodium fast reactor

2019 ◽  
Vol 125 ◽  
pp. 291-306 ◽  
Author(s):  
G. Pereira ◽  
A.E. Johnson ◽  
Y. Bilodid ◽  
E. Fridman ◽  
D. Kotlyar
Keyword(s):  
Fast Reactor ◽  
Code Sequence ◽  
Metallic Fuel ◽  
Decay Heat ◽  
Sodium Fast Reactor

scholarly journals Advanced sodium fast reactor accident source terms :

2010 ◽  
Author(s):  
Dana Auburn Powers ◽  
Bernard Clement ◽  
Richard Denning ◽  
Shuji Ohno ◽  
Roland Zeyen
Keyword(s):  
Fast Reactor ◽  
Source Terms ◽  
Reactor Accident ◽  
Sodium Fast Reactor
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