COMPLEX OF STUDIES TO SUPPORT SAFETY OF FRENCH FAST REACTOR PROJECT

30 Projects to the 30th Anniversary of NNC RK!

Project No.25. 2019–2022 COMPLEX OF STUDIES TO SUPPORT SAFETY OF FRENCH FAST REACTOR PROJECT

Safety aspects of accidents involving fast reactor core melting have been and continue to be a major topic of discussion in the design and operation of nuclear reactors, and reactor safety requirements have become increasingly stringent over the years, and one of the factors that must be studied in detail to develop measures to mitigate the consequences of a severe accident is the recurrence of criticality.

When the core melts and the fuel and solid absorbers are redistributed in the corium, there is a risk that critical conditions for neutron multiplication, i.e. the phenomenon of recurring criticality, may arise in the reactor, which in turn may lead to additional overheating of the damaged reactor and a massive release of radioactivity beyond the shielding barriers. The occurrence of recurring criticality is a more likely event in the event of a severe fast reactor accident due to the relatively higher enrichment of the fuel used compared to other types of power reactors.

The most significant way to obtain experimental data on the behavior of reactor fuel in transient and accident modes of operation are reactor experiments, in which the maximum approximation to real operating conditions can be achieved, and, therefore, the fuel behavior can correspond as closely as possible to the real.

The need for such research is no exception for France, which is one of the world leaders in nuclear power and ranks second in the world in terms of the amount of energy generated by nuclear power plants, and first in terms of the share of nuclear power in the country's energy sector. The French Atomic Energy and Alternative Energies Commissariat is the French nuclear industry's leading research organization. The Commissariat's main mission is fundamental and applied research into the use of nuclear energy. The organization emphasizes the importance to research programs dedicated to the fourth generation reactors, including fast-neutron reactors, effectively operating in a closed fuel cycle.

The program for the development of advanced nuclear power plants in France was based on sodium-cooled fast reactors. The first demonstration reactor ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is to be a visible result of this program.

The ASTRID reactor has a number of characteristic features that require a large amount of additional research aimed at confirming the safety of the chosen design, including studying the behavior of reactor core elements during emergency situations progression.

One of the most suitable model fuel assemblies for advanced reactor testing is the pulsed graphite reactor IGR, the technical characteristics of which provide the ability to simulate severe accidents in a wide range of specified energy release, and the combination of the reactor dynamic characteristics and its loop parameters provide broad experimental capabilities and test conditions for nuclear facilities under various design and beyond design basis accidents.

In December 2011, the first visit of the French Commissariat staff to Kurchatov took place regarding the preparation and conduct of experimental studies at the IGR reactor in support of the ASTRID reactor project. During the working meeting, the Kazakhstan side presented the main results of previously conducted studies on the behavior of reactor fuel under severe accident conditions, and the French specialists expressed interest in considering the possibility of conducting this type of tests with respect to elements of the ASTRID reactor core.

By the next working meeting, held in October 2012 in Kurchatov, the team from France had already prepared specific proposals on technical and economic studies to justify the possibility of implementing an experimental program aimed at studying, under reactor experiment conditions, the development processes of a severe accident in a reactor with a model fuel assembly of ASTRID reactor. This proposal was accepted by the Kazakhstan side, and in 2013, studies were initiated to justify the possibility of conducting experiments under the program provisionally named SAIGA (Severe Accident In-pile Tests for Generation IV reactors and ASTRID project), which aims to study the option of reducing the probability of recurrence of criticality.

The results of technical and economic studies were presented by Kazakhstani specialists at a working meeting held in May 2014 in Cadarache Research Center. The results of the validation of the possibility of performing in-reactor experiments at the IGR reactor were evaluated by the French side as positive, and the meeting resulted in the signing of a protocol of intentions of the parties to discuss the preparation and conduct of experiments on the IGR reactor to justify the safety.

The objectives of the main and additional studies were defined by the French side by mid-2015, and by the end of the same year, the work to justify the possibility of implementing SAIGA program was completed.  In late February – early March 2016, a team of Kazakhstani specialists visited Сadaraсhe to discuss the results of the completed additional part of feasibility studies that resulted to a decision to prepare the tender documents for participation in SAIGA contract tender. During 2017 and 2018, negotiations were actively conducted to discuss the order, scope and conditions of work under the contract, as well as to prepare a package of documents for participation in the tender. As a result, the contract was signed in June 2019, and in early July a team of French specialists arrived at NNC to participate in the first (within the contract) technical meeting at which the initial data and specifications for the development of the experimental device, sodium circuit and justifications of experimental conditions were presented.

The SAIGA contract is intended for seven years and consists of two parts. The first part was performed during 2019-2020 and was aimed at solving several tasks - the development of technical projects of the experimental device and sodium circuit; the calculation-theoretical analysis of neutronic and thermal-hydraulic characteristics of the experimental device and sodium circuit, the justification of test modes on a research reactor. Since 2021, the National Nuclear Centre of the Republic of Kazakhstan has been carrying out the research part of the SAIGA program.