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 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.

Optimization of the European Sodium Fast Reactor Secondary Sodium Loop as Part of the ESFR-SMART Project

2021 ◽  
Author(s):  
Joel Guidez ◽  
Janos Bodi ◽  
Konstantin Mikityuk ◽  
Enrico Girardi
Keyword(s):  
Fast Reactor ◽  
Reactor Safety ◽  
Secondary Circuit ◽  
Safety Measures ◽  
Design Option ◽  
Sodium Fast Reactor ◽  
The Cost ◽  
Research Tools

Abstract Based on feedback from existing reactors and current projects, the European Sodium Fast Reactor Safety Measures Assessment and Research Tools (ESFR SMART) project proposes an optimization of the secondary circuit with the main aim of improving safety. Besides, the optimization also leads to a simplification of the circuits and therefore to a reduction of the cost of the reactor. For the implementation of the proposed new design option, some points require further R&D to validate their feasibility.

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.

HORIZON-2020 ESFR-SMART PROJECT ON SODIUM FAST REACTOR SAFETY: STATUS AFTER FIRST 15 MONTHS

2019 ◽  
Vol 2019.27 (0) ◽  
pp. 2181 ◽  
Author(s):  
K. Mikityuk ◽  
E. Girardi ◽  
J. Krepel ◽  
E. Bubelis ◽  
E. Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Safety ◽  
Horizon 2020 ◽  
Sodium Fast Reactor

scholarly journals Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

2011 ◽  
Author(s):  
Juan Carbajo ◽  
Hae-Yong Jeong ◽  
Roald Wigeland ◽  
Michael Corradini ◽  
Rodney Cannon Schmidt ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Accident Analysis ◽  
Reactor Safety ◽  
Sodium Fast Reactor ◽  
Computer Codes

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 Sodium fast reactor safety and licensing research plan. Volume I.

2012 ◽  
Author(s):  
Tanju Sofu ◽  
Jeffrey L. LaChance ◽  
R. Bari ◽  
Roald Wigeland ◽  
Matthew R. Denman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Safety ◽  
Sodium Fast Reactor ◽  
Research Plan

Evaluation of the Passive Reactor Shutdown System Performance in European Sodium Fast Reactor

2021 ◽  
Author(s):  
Evaldas Bubelis ◽  
Michael Schikorr ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
System Performance ◽  
Negative Consequences ◽  
Work Package ◽  
Safety Measures ◽  
Sodium Fast Reactor ◽  
Core Meltdown ◽  
Reactor Shutdown ◽  
Design Options ◽  
The Impact

Abstract The ESFR-SMART European project (Contract number: 754501) focuses on the development of innovative safety design options for European Sodium-cooled Fast Reactor (ESFR). The task of Work Package 1.3 is to assess the impact of the new safety measures on the reactor behaviour in the transients protected by either active or passive reactor shutdown systems. The aim of Task 4 in this Work Package is to evaluate the passive reactor shutdown system performance in the ESFR core. This paper deals with the results of this evaluation, which is based on the analysis of four transients passively protected by 12 Diversified Shutdown Device (DSD) rods. Simulations have been done with the SIM-SFR system code and demonstrated that DSD rods are capable to shutdown ESFR in a timely manner, in order to avoid the negative consequences of the analyzed transients. Although a total loss of heat sink transient is a practically eliminated event, it was included in the analysis to estimate the grace time before the core meltdown.

Research on Emergency System of Injection Estimation of Reactor Core Recovery and Reactor Core Decay Heat Removal

2017 ◽  
Author(s):  
Liao Feiye ◽  
Jiang Pingting ◽  
Liu Wang ◽  
He Dongyu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Reactor Core ◽  
Heat Removal ◽  
Lessons Learned ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Core Recovery ◽  
Emergency System

One of the lessons learned from Fukushima accident is that the existing procedures used in Nuclear Power Plants (NPPs) are not executed effectively and quickly enough after such an extended accident, for the accident is complex and people are too nervous in such a situation. Thus, emergency system that helps to raise diagnosis efficiency is necessary. In the paper, a quick diagnosis system on injection estimation of reactor core recovery and decay heat removal injection estimation is developed to meet the urgent needs and strengthen requirements for the training and application among utilities and nuclear regulators. The system will assist regulators to quickly know whether the currently flow will probably recover the reactor core, or whether the current injection capacity is sufficient to quench and recover the reactor core, directly after input present parameters into the system. In the system, Matlab method is used, and intuitive insights are considered, which is propitious to give immediate graphical interface and reduce possibility of human error.

scholarly journals Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

2008 ◽  
Vol 2008 ◽  
pp. 1-8
Author(s):  
A. Kaliatka ◽  
E. Uspuras ◽  
M. Vaisnoras
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Severe Accident ◽  
Cooling Circuit ◽  
High Thermal Inertia ◽  
The Moment ◽  
Control Rods

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

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