Interview with Mr François GAUCHÉ, CEA

Interview on the ASTRID project

Mr GAUCHÉ is the Manager of the “Generation IV Reactors” Program at the Nuclear Energy Division of CEA, the “French Alternatives Energies and Atomic Energy Commission”. SFEN Young Generation has interviewed him on the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration).

Mr GAUCHÉ, could you please introduce the ASTRID project to us?

The ASTRID project consists in the R&D, design and development of a 4th Generation Sodium Fast Reactor Prototype, the level of safety of which will be at least equivalent to the third generation of LWR and will integrate into its design the lessons learnt from the Fukushima accident.

The project involves several companies of the nuclear industry:

  • CEA, the contracting owner, is in charge of core design,
  • AREVA is in charge of Nuclear Steam Supply System, Instrumentation and Control systems and Nuclear Auxiliaries,
  • EDF provides support and experience to the contracting owner and performs safety studies.
  • Alstom Power Systems brings its worldwide expertise to the design of the energy conversion part of the plant
  • Other companies are involved in various parts of the design (TOSHIBA, COMEX Nucléaire, ASTRIUM, Rolls-Royce, JACOBS France)

What is exactly a Sodium Fast Reactor?

The most widespread reactor technology uses water as a coolant, whereas Sodium Fast Reactors are cooled by liquid sodium. Water slows down neutrons, and liquid sodium does not. That is why nuclear reactions in SFRs are governed by neutrons of high speed, and it explains the name of the Fast Neutron Reactors. On the contrary, nuclear reactions in water-cooled reactors are governed by low speed neutrons (“thermal” neutrons).
In nuclear reactors, fissile nuclei can be split by neutron absorption into “fission products”, releasing at the same time energy (used for electricity generation) and neutrons. These neutrons can either produce new fissions (chain reaction), or be captured by nuclei, or leak out of the reactor.

There is a remarkable fact that, although the most part of the natural uranium – made of uranium-238 – is not fissile, it can be transformed into fissile plutonium-239 by capture of a neutron: this is called conversion.
In a thermal neutron reactor, the ratio between conversions and fissions is always less than 1, thus the quantity of fissile material decreases in the reactor over the cycle. In a fast neutron reactor, it is possible to obtain a balance between conversions and fissions (iso-breeding mode) or even more conversions than fissions (breeding mode).
Using that possibility, the spent fuel unloaded from a Fast Neutron Reactor can contain as much fissile material as when it started. Of course, uranium-238 is consumed in the process.

There are limitations due to the fission products (that hinder the chain reaction) and to the dose supported by materials like the fuel cladding. These limitations rule the duration of the cycle, at the end of which it is required to unload the fuel, recycle the uranium and plutonium while removing the waste products, and reload new fresh fuel.

ASTRID will operate in iso-breeding mode, so as to stabilize the quantity of plutonium in its fuel cycle.
Fast Neutron Reactors can also burn so-called minor actinides, i.e. isotopes that were produced in a reactor by neutron capture on plutonium (neptunium, americium and curium). ASTRID will provide demonstration capabilities for such a process that is also called transmutation.

fission pu

Basics of Uranium-Plutonium chain reaction within SFRs

Have there already been such reactors throughout the world?
Many SFRs have already been built and operated. The most powerful SFR ever built was the French “Superphenix” reactor. It was able to deliver up to a power of 1200 MW on the electrical grid and has been operated during 12 years. The accumulation of all SFRs years of operation has a total of approximately 400 years of experience in this technology. Today, several SFRs are under operation (Japan1, Russia, India and China).
What are the challenges associated with this reactor technology?

Liquid sodium is harder to handle than water. Indeed, hot liquid sodium inflames in contact with air, and it reacts chemically in contact with water. Not really attractive at first sight, right?

The choice of such a coolant for a Fast Neutron Reactor is based on several criteria. The first one is not to slow down neutrons. That is why water is excluded. But other key parameters (such as thermal properties, viscosity, compatibility with steel, etc.) are of utmost importance as well. There are other possible coolants that can let the fast-neutron reactions take place. However, following an analysis on advantages and drawbacks, taking into account safety and operability considerations, it is difficult to find a good replacement for sodium.
For example, liquid lead could be considered as coolant instead of sodium. But one of its major disadvantages is the narrow range of operation: indeed, above 480°C lead gets highly corrosive for the steel that the reactor vessel, circuits and fuel cladding are made of. On the other side, the reactor must not be cooled less than 400°C to ensure that lead does not freeze in the circuit.

So the mandatory temperature range of the reactor at all times would be between 400 and 480°C, which is not that comfortable, e.g. for maintenance in “cold” state. In order to widen that 80°C range, Bismuth element can be combined with lead to lower its freezing temperature. Unfortunately, under radiation, bismuth is transmuted into Polonium-210, which is a highly radiotoxic isotope.

As another example, instead of liquid sodium, a gas might be used, like helium. But gases have a low thermal inertia, which means that these types of reactors are sensitive to depressurization. Safety commands in that case to take high margins and use materials that can withstand up to 1600°C: this technology is not available today and will need significant R&D before a proof of its feasibility.

These two examples show that there is no such thing as the “perfect” coolant. The weaknesses of sodium are well known and engineered barriers can be designed to control them, so that as a result of a multi-criteria analysis, sodium is worldwide considered as the reference choice for fast neutron reactors.

sfr sodium
Circuits of a Sodium Fast Reactor

How does the ASTRID project cope with the requirements for safety?

There are three main issues to deal with: air-sodium fires, sodium-water reactions and a more technical issue known as “positive void-coefficient”.

So as to ban sodium fires in case of a contact with air, on top of design provisions and quality control of the piping, the rooms where sodium circuits are located can be filled with nitrogen instead of air. Moreover, the vessel is made of three layers: the main steel vessel is contained in a safety steel vessel, which is in turn contained in a concrete vessel-shaped pit with steel-liner. The “pool-type” reactor design benefits from a fully integrated primary circuit, i.e. all the equipment is contained in the vessel, instead of drawing pipes out of the three-layer vessel. That makes the reactor mechanically stronger and provides the guarantee that the sodium cannot physically escape the primary circuit.

In classical designs where steam generators provide steam to a conventional turbine connected to an electrical generator, sodium-water reactions can occur and need to be addressed. To make sure the consequences of such sodium-water reaction do not affect the primary circuit, one possibility is to limit the size of the steam generators (modular steam generators). They are installed on a so-called “intermediate circuit” which is a second sodium circuit to provide for an additional barrier between the primary circuit and the environment, so that the water-sodium interface, located in the steam generator between the intermediate circuit and the water circuit, is far away from the nuclear material. Another more radical solution currently studied is to replace the water by another fluid, pure nitrogen for instance.

There is a last drawback at using sodium: contrary to most water-cooled reactors, SFRs’ void-coefficient is positive in classical designs, which means that in case of coolant boiling, the core reactivity increases and leads to a power excursion. To avoid that, the shape of the core for ASTRID has been designed to get a very low or negative void-coefficient and thus to avoid the power excursion in case of loss of cooling accident: this is a major safety improvement compared to former design of Sodium Fast Reactors.

Once the issues are tackled, what are the advantages of sodium?

“Pool-type” Sodium Fast Reactors have a thermal inertia combined with the boiling margin around 20 times greater than for water-cooled reactors. That is of great help in accident studies: in the case of a loss of reactor cooling, the low kinetics of the accident increases the “grace period” to take the necessary actions to bring back the plant in a safe state and avoid severe accidents of reactor core meltdowns.

Another good point is that the operating temperature of SFRs is around 550°C, which allows heat exchanges of great yield with air. Thus, the heat sink, usually a sea or a river, can be diversified: in case of loss of the main heat sink, dedicated safety systems transfer the heat directly to air via dedicated heat exchangers. Such systems can be designed as passive systems operating under natural circulation (heated sodium going up, and cooled sodium getting back down) is quite efficient in SFRs. This passive feature is of course very interesting in the frame of Post-Fukushima studies, as focus of studies of total loss of electrical supply gets more important.

Unlike PWRs, that are pressurized at 155 bar, SFR vessels are at almost the atmospheric pressure: this eliminates by design pressure-related loss of coolant events.

Lastly, one of the best assets of sodium is that we have 400 years of cumulated operating experience with it, so that we know its strengths, that can be used for instance to design efficient safety systems, and its weaknesses for which we can design dedicated engineered barriers. Let us not forget that safety is improved by learning lessons from the experience.

Actually, what is the point in working on a new technology of reactors?
This comes from the need to better use and recycle nuclear matters that are uranium, plutonium and minor actinides.

Natural uranium is composed of two isotopes: Uranium-238 is present at 99,3%, Uranium-235 for 0,7% of natural uranium.

Water-cooled reactors – like the 58 reactors currently in operation in France – mostly “burn” Uranium-235 out of fuels made of enriched uranium. Even if another type of fuel can be partially used (MOX fuel, i.e. plutonium-uranium oxide), this means that a maximum of 1% of the energetic content of natural uranium is used, leaving the larger part in form of depleted uranium or reprocessed uranium.

In the enrichment process, the percentage of uranium-235 is increased in the fuel, leaving aside depleted uranium. For example, in a typical open cycle for a 63GWe fleet, the enrichment of 9600 tons of natural uranium leaves 8400 tons of depleted uranium aside containing almost only uranium-238.

On earth, there are 189 billions of tons of oil, 187 Tm3 of Natural Gas, 860 billions of tons of coal (2) and 4 millions of tons of Natural Uranium (3). If we consider the uranium is used only in thermal neutrons reactors, converting these stocks into energy makes the following chart:

total energetic

Total energetic contents of various sources of energy according to confirmed stocks
Purple: Coal / Pink: Oil / Orange: Natural Gas
Green: Uranium in Thermal Neutrons Reactors

This uranium-238 cannot be used in water-cooled reactors but could be used in fast neutron reactors, multiplying the energy content of uranium by a factor of more than 100. Thus the chart becomes the following:

energetic total

Total energetic contents of various sources of energy according to confirmed stocks
Purple: Coal / Pink: Oil / Orange: Natural Gas
Green: Uranium in Thermal Neutrons Reactors

Global reserves of this uranium-235 (the 1%-part) could be exhausted in less than a century if the rate of use follows the current trend. On the contrary, global reserves of coal are high enough to let coal power plants be operated for centuries, which could lead to an environmental disaster.

However, the nuclear reactions that take place in Sodium Fast Reactors use the uranium-238, the exhaustion of which is forecasted after several millennia of electricity consuming! On top of that, due to uranium enrichment activities, several countries already own great amounts of depleted uranium, France included. This depleted uranium can be used to fuel Sodium Fast Reactors, enabling to secure the uranium supply.

In a view to keep alternatives to CO2 emissions, SFRs stand as a millennium-sustainable, economically viable and carbon-free source of energy.

France has a leading position in the nuclear energy industry, for its experience and its high safety standards. Thus it has the duty to keep an eye on the long-term strategy to adopt. If France does not get involved in the future of the nuclear industry, other countries will take the lead and their safety standards might become global standards, for better or for worse.

General interest commands to do whatever is possible to stop emitting greenhouse gases. That is why the future of nuclear energy must be developed now, by countries that have experience and have credibility as regards improving more and more nuclear safety.

What are the milestones of the ASTRID project?

We have completed at the end of 2012 the first phase of the conceptual design. The main safety orientations of the reactor have been presented to the French Nuclear Safety Authority in a document sent in June 2012. Within 3 years, a major document, called the Safety Options File, will be written. It will gather all the strategy and rules that will be applied to go on in the reactor design. Thanks to this document, the basic design of the reactor will be carried out, leading to the writing of a Preliminary Safety Analysis Report, in 2019. First criticality could be achieved in 2025, so that the operation of ASTRID provides sufficient feedback of experience for commercial deployment from 2040.


What is your motivation for a program with such a late end?

This program is not meant to optimise short-term profit. It is meant to prepare the future of energy sources. I strongly believe that centralized, intensive energy sources are needed, and that nuclear energy can continue to give our country a competitive advantage. In a few decades, thanks to the ASTRID project, there will still be some economic, sustainable and carbon-free energy sources to prevent the release of greenhouse gases. This is the reason of my involvement in the ASTRID project.





(2) Source : BP statistical Review of World Energy, June 2011

(3) Source : Red Book, 2009 edition (RRA)