scholarly journals Simplified modeling of liquid sodium medium with temperature and velocity gradient using real thermal-hydraulic data. Application to ultrasonic thermometry in sodium fast reactor

2013 ◽  
Author(s):  
N. Massacret ◽  
J. Moysan ◽  
M. A. Ploix ◽  
J. P. Jeannot ◽  
G. Corneloup
Keyword(s):  
Fast Reactor ◽  
Velocity Gradient ◽  
Liquid Sodium ◽  
Sodium Fast Reactor ◽  
Data Application ◽  
Simplified Modeling

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.

Design and first operation of the MACARON irradiation experiment of Sodium Fast reactor absorber pins in the BOR-60 reactor

2021 ◽  
Vol 135 ◽  
pp. 103676
Author(s):  
T. Lambert ◽  
J.M. Escleine ◽  
B. Fontaine ◽  
S. Eremin ◽  
E. Muraleva ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Irradiation Experiment ◽  
Sodium Fast Reactor

Validation of TRACE capability to simulate unprotected transients in Sodium Fast Reactor using FFTF LOFWST Test #13

2021 ◽  
Vol 164 ◽  
pp. 108600
Author(s):  
Shibao Wang ◽  
Konstantin Mikityuk ◽  
Petrovic Dorde ◽  
Dalin Zhang ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Sodium Fast Reactor

Decay Heat Characterization for the European Sodium Fast Reactor

2021 ◽  
Author(s):  
Antonio Jiménez-Carrascosa ◽  
Nuria Garcia Herranz ◽  
Jiri Krepel ◽  
Marat Margulis ◽  
Una Baker ◽  
...  
Keyword(s):  
Fast Reactor ◽  
State Of The Art ◽  
Nuclear Data ◽  
Summation Method ◽  
Heat Power ◽  
Coupled Transport ◽  
Decay Heat ◽  
Sodium Fast Reactor ◽  
Detailed Assessment ◽  
Commercial Size

Abstract In this work a detailed assessment of the decay heat power for the commercial-size European Sodium-cooled Fast Reactor (ESFR) at the end of its equilibrium cycle has been performed. The summation method has been used to compute very accurate spatial- and time-dependent decay heat by employing state-of-the-art coupled transport-depletion computational codes and nuclear data. This detailed map provides basic information for subsequent transient calculations of the ESFR. A comprehensive analysis of the decay heat has been carried out and interdependencies among decay heat and different parameters characterizing the core state prior to shutdown, such as discharge burnup or type of fuel material, have been identified. That analysis has served as a basis to develop analytic functions to reconstruct the spatial-dependent decay heat power for the ESFR for cooling times within the first day after shutdown.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

2021 ◽  
Author(s):  
Janos Bodi ◽  
Alexander Ponomarev ◽  
Evaldas Bubelis ◽  
Konstantin Mikityuk
Keyword(s):  
Power Plant ◽  
Fast Reactor ◽  
Heat Removal ◽  
Current Paper ◽  
Decay Heat ◽  
Sodium Fast Reactor ◽  
Station Blackout ◽  
Safety Assessments ◽  
The Individual

Abstract As part of the ESFR-SMART project, safety assessments are being conducted on the updated European Sodium Fast Reactor (ESFR) design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In the current paper, the analyzed accidental scenario is the so called Protected Station Blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and SIM-SFR system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRS). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project have been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

New Reactor Safety Measures for the European Sodium Fast Reactor - Part I: Conceptual Design

2021 ◽  
Author(s):  
Joel Guidez ◽  
Janos Bodi ◽  
Konstantin Mikityuk ◽  
Enrico Girardi ◽  
Bernard Carluec
Keyword(s):  
Fast Reactor ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Lessons Learned ◽  
Reactor Safety ◽  
Fukushima Accident ◽  
Safety Level ◽  
Safety Measures ◽  
Sodium Fast Reactor ◽  
High Thermal Inertia

Abstract Following up the previous CP-ESFR project, the ESFR-SMART project considers the safety objectives envisaged for Generation-IV reactors, taking into account the lessons learned from the Fukushima accident, in order to increase the safety level of the European Sodium Fast Reactor (ESFR). In accordance with these objectives, guidelines have been defined to drive the ESFR-SMART developments, mainly simplifying the design and using all the positive features of Sodium Fast Reactors (SFR), such as low coolant pressure, efficiency of natural convection, possibility of decay heat removal (DHR) by atmospheric air, high thermal inertia and long grace period before a human intervention is needed. In this paper, a set of new ambitious safety measures is introduced for further evaluation within the project. The proposed set aims at consistency with the main lines of safety evolutions since the Fukushima accident, but it does not yet constitute the final comprehensive safety analysis. The paper gives a first review of the new propositions to enhance the ESFR safety, leading to a simplified reactor, forgiving and including a lot of passivity. This first version is supported by the various project tasks in order to assess the relevance of the whole design in comparison to the final safety objectives.

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