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.

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 New Reactor Safety Measures for the European Sodium Fast Reactor. Part II: Preliminary Assessment

2021 ◽  
Author(s):  
Joel Guidez ◽  
Janos Bodi ◽  
Konstantin Mikityuk ◽  
Enrico Girardi ◽  
Jeremy Bittan ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Heat Removal ◽  
Reactor Safety ◽  
Preliminary Assessment ◽  
Safety Measures ◽  
Decay Heat ◽  
European Project ◽  
Sodium Fast Reactor ◽  
Core Catcher ◽  
Gen Iv

Abstract The European project ESFR SMART offers innovative options of a sodium fast reactor to improve its safety. This paper explains the results of preliminary calculations made of the main options to verify the big lines of their feasibility. Design propositions and calculations are here provided of following innovative options: removal of the safety vessel, innovative decay heat removal systems, core catcher, thermal pumps and secondary loops. In conclusion, all these options seem able to fulfil the big lines of new safety rules for GEN-IV reactors. A status of the R&D necessary to validate these new options is also proposed.

Neutronic Analysis of the European Sodium Fast Reactor: Part I - Fresh Core Results

2020 ◽  
Author(s):  
Emil Fridman ◽  
Francisco Álvarez Velarde ◽  
Pablo Romojaro Otero ◽  
Haileyesus Tsige-Tamirat ◽  
Antonio Jiménez-Carrascosa ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Core State ◽  
Computational Tools ◽  
Horizon 2020 ◽  
The Core ◽  
Continuous Energy ◽  
Sodium Fast Reactor ◽  
Computer Codes ◽  
Calibration And Verification

Abstract In the framework of the Horizon 2020 project ESFR-SMART (2017-2021), the European Sodium Fast Reactor (ESFR) core was updated through a safety-related modification and optimization of the core design from the earlier FP7 CP-ESFR project (2009-2013). This study is dedicated to neutronic analyses of the improved ESFR core design. The conducted work is reported in two parts. Part I deals with the evaluation of the safety-related neutronic parameters of the fresh Beginning-of-Life (BOL) core carried out by 8 organizations using both continuous energy Monte Carlo and deterministic computer codes. In addition to the neutronics characterization of the core, a special emphasis was put on the calibration and verification of the computational tools involved in the analyses. Part II is devoted to once-through and realistic batch-wise burnup calculations aiming at the establishing of the equilibrium core state, which will later serve as a basis for detailed safety analyses.

Neutronic Analysis of the European Sodium Fast Reactor: Part II - Burnup Results

2020 ◽  
Author(s):  
Emil Fridman ◽  
Francisco Álvarez Velarde ◽  
Pablo Romojaro Otero ◽  
Haileyesus Tsige-Tamirat ◽  
Antonio Jiménez-Carrascosa ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Core State ◽  
Computational Tools ◽  
Horizon 2020 ◽  
The Core ◽  
Continuous Energy ◽  
Sodium Fast Reactor ◽  
Computer Codes ◽  
Calibration And Verification

Abstract In the framework of the Horizon 2020 project ESFR-SMART (2017-2021), the European Sodium Fast Reactor (ESFR) core was updated through a safety-related modification and optimization of the core design from the earlier FP7 CP-ESFR project (2009-2013). This study is dedicated to neutronic analyses of the improved ESFR core design. The conducted work is reported in two parts. Part I deals with the evaluation of the safety-related neutronic parameters of the fresh Beginning-of-Life (BOL) core carried out by 8 organizations using both continuous energy Monte Carlo and deterministic computer codes. In addition to the neutronics characterization of the core, a special emphasis was put on the calibration and verification of the computational tools involved in the analyses. Part II is devoted to once-through and realistic batch-wise burnup calculations aiming at the establishing of the equilibrium core state, which will later serve as a basis for detailed safety analyses.

Word from the Horizon-2020 EU ESFR-SMART Project Coordinator

2021 ◽  
Author(s):  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Safety Measures ◽  
Horizon 2020 ◽  
Project Coordinator ◽  
Support Organizations ◽  
Sodium Fast Reactor ◽  
Medium Enterprise ◽  
Radiation Science ◽  
Asme Journal ◽  
Research Centres

Abstract This Special Issue of the ASME Journal of Nuclear Engineering and Radiation Science includes 2 editorials and 25 technical papers presenting main achievements of the Horizon-2020 European Union ESFR-SMART project (European Sodium Fast Reactor Safety Measures Assessment and Research Tools) supported by the EURATOM grant and launched in September 2017. ESFR-SMART gathers a consortium of 19 organizations (see Figure 1): Research centres, industries, universities, Technical and Scientific Support Organizations as well as Small to Medium Enterprise aiming at enhancing further the safety of Generation-IV Sodium Fast Reactors (SFRs) and, in particular, of the commercial-size European Sodium Fast Reactor (ESFR) in accordance with the European Sustainable Nuclear Industrial Initiative (ESNII) roadmap.

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.

scholarly journals Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Hao Yu ◽  
Minjun Peng
Keyword(s):  
High Pressure ◽  
Low Pressure ◽  
Lower Pressure ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Pressurized Water ◽  
The Core ◽  
Severe Accidents ◽  
Accident Management ◽  
Degradation Studies

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

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