Conference Agenda

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Session Overview
Session
3.03: Fast Neutron Reactors – Lead Fast Reactor
Time:
Monday, 16/Mar/2020:
1:30pm - 3:00pm

Session Chair: Christian Latge, CEA, France
Location: L-1011

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Presentations

Lead-Bismuth and Lead as Coolants for Fast Reactors

Georgii Ilyich Toshinskii1,2, Aleksander Vladislavovich Dedul1, Oleg Gennadyevich Komlev1, Aleksey Viktorovich Kondaurov1, Vladimir Viktorovich Petrochenko1

1JSC "AKME-egineering"; 2JSC State Scientific Center of Russian Federation - Institute for Physics and Power Engineering

Nuclear power (NP) of the future must be large-scale to ensure sustainable development. This is possible when using fast reactors (FRs) with a breeding ratio equal to or more than one operating in a closed nuclear fuel cycle.

In addition, in order to overcome the radiophobia of the population, the safety of future nuclear power plants (NPPs) should mainly be based on a high level of inherent self-protection of reactors, deterministically excluding the causes of severe accidents requiring evacuation of the population.

The use of FRs with a high level of inherent self-protection (absence of potential compression energy and chemical energy in the primary coolant), eliminates the conflict between the safety requirements and the requirements of the economy, inherent in conventional reactors. Such coolants are heavy liquid metal coolants (HLMC): a lead-bismuth eutectic alloy, mastered under the operating conditions of Russian nuclear submarine reactors, and lead.

The mentioned conflict is manifested in the fact that with increasing safety requirements, economic indicators worsen. This is due to the fact that to reduce the likelihood of a severe accident, an increase in the number of safety systems and protective barriers is required.

Thus, the use of FRs with a high level of inherent self-protection creates the prerequisites for constructing NPPs not only with a higher level of safety, but also with a higher level of competitiveness.

The report presents a comparative analysis of the advantages and disadvantages of these coolants, identifies areas of preferred use of FRs with lead-bismuth and lead. It is concluded that the use of FRs with HLMC is promising after gaining experience in operating the first samples of such reactors.

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* Corresponding author E-mail: toshinsky@ippe.ru



MYRRHA, the Belgian ADS programme: new developments in the primary systems

Didier De Bruyn, Graham Kennedy, Rafaël Fernandez

SCK•CEN

The purpose of the MYRRHA programme is to demonstrate the Accelerator Driven System (ADS) concept at pre-industrial scale, to demonstrate transmutation and to serve as a flexible and multipurpose irradiation facility.

The design of the MYRRHA facility (accelerator and subcritical reactor) has progressed through various framework programmes of the European Commission in the context of Partitioning and Transmutation, among others EUROTRANS (2005-2010), CDT (2009-2012) and more recently MAXSIMA and MYRTE.

Since the beginning of MYRRHA in 1998, SCK•CEN has launched a strong R&D support programme to address the main design challenges, in particular those related to the use of liquid Lead-Bismuth Eutectic (LBE) as reactor coolant and spallation target. The various experimental R&D LBE facilities have been presented in our ICAPP 2019 paper, together with the present structure of the MYRRHA programme after approval by the Belgian government of a significant budget (558 M€). The aim of the present structure is to obtain a first infrastructure operational (a 100 MeV accelerator together with dedicated proton target facilities) in 2026.

The different components of the primary systems were presented in our ICAPP 2018 paper, but as isolated components. The integration of the system components is in progress and will be described with more detail in the oral presentation. The primary systems design has to continue, as 2026 is also the milestone for decision to build the remaining part of the MYRRHA facility, namely the accelerator up to 600 MeV connected to the subcritical reactor.

In this paper, we describe some thermomechanical results obtained with the LBE facilities and the corresponding numerical modelling. We also present the new core configuration (for which a detailed presentation is in preparation for another conference end of this year).



Analysis of the SGTR accident for safety justification of two- circuits lead cooled reactor

I.S. Khomyakov, V.I. Rachkov, I.E. Shvetsov, I.R. Suslov

Innovation and Technology Center of the Project "PRORYV"

Lead cooled reactor BREST-OD-300 is developing as a part of Russian federal project "PRORYV". Two- circuits scheme is used in the reactor for heat removal. The special feature of two- circuits reactor is the potential danger of water steam ingression in the core in the case of large leakage in steam generator initiated, for example, SGTR (steam generator tube rupture). First of all the concern is caused by the possibility of positive reactivity insertion in the case of water steam ingression in the central part of the core because that can effect on nuclear criticality safety.

The analysis of physical phenomena that are important for correct prediction of SGTR accident consequences was performed and it was concluded that for correct modeling of the transient in the core in the case of water steam injection it is required:

a) Complex modeling of neutronic and thermal-hydraulic characteristics due to their interrelation;

b) 3D modeling of the accident development due to strong non-symmetry of space distributions of vapor concentration in the core during the accident.

3D multi-physics (neutronics + thermal-hydraulics) UNICO-2F code was developed for study of SGTR accident. Code calculates unsteady 3D space distributions of coolant velocity, pressure and temperature, space distributions of vapor concentration and heat release density in the core.

The analysis of BREST-OD-300 rector parameters under SGTR accident conditions was performed and it was shown that even for most conservative scenario of the accident the maximum (during the transient) fuel pin cladding temperature is kept below acceptable limits. Therefore, the self-protection of BREST-OD-300 against SGTR accident is confirmed.



 
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