NUGENIA, the European association for research on Generation II and III nuclear reactors, gathered 260 participants in April 2016 in Marseille. The meeting gave IRSN and its partners the opportunity to discuss the projects they plan to propose in response to the call for projects 2016-2017 launched by the European Commission, under the H2020 Framework Program.

The meeting, to which the European Commission and the AIEA were invited, provided an opportunity for IRSN and members of NUGENIA to discuss also ongoing projects selected in the first call for projects for 2014-2015.

The program for this second call for projects, like the first one, focuses largely on research relative to generation II reactors built up to the end of the 1990’s and generation III reactors such as EPR under construction in Normandy, France.

The subjects discussed by the key partners included extending the service life of current nuclear power plant, passive safety systems for generation II and III reactors, small modular reactors, and severe reactor accidents and their consequences.

Work on finalizing the safety-related research proposals is now underway for submission this fall. The European Commission will notify the results in March 2017.

More informations: Website of NUGENIA - NUclear GENeration II & III Association

The CABRI research reactor has once again achieved criticality after several years of renovation and modification work. Operational preparation for the CABRI International Program (CIP) for the study of reactivity accidents can now go ahead. Led by IRSN under the aegis of the NEA, CIP involves 18 partners from 12 countries.

The divergence of the CABRI research reactor in October 2015 is the culmination of several years of renovation and modification of the facility. This achievement was the last technical step prior to the operational implementation of the Cabri International Program (CIP), an international research program led by IRSN under the aegis of the OECD NEA and devoted to reactivity initiated accident experiments.

During a reactivity accident, some of the fuel assemblies making up the reactor core can be subject to a significant power surge for a few milliseconds, exposing the fuel elements to extreme loads, which may cause their destruction. It is therefore essential to verify experimentally the effects of such loads on fuel elements, in order to understand safety margins and adapt regulatory safety criteria where necessary. This is the goal of the fuel tests conducted in CABRI, a reactor with outstanding characteristics which make it unlike any other facility in the world.

CABRI is designed to produce rapid power transients. Its new water loop reproduces the pressure (155 bar) and temperature (280°C) conditions observed in a pressurized water reactor, so it can be used to study all the physical phenomena(7) involved in a reactivity accident, along with their interactions. The CIP program consists of ten tests on different types of fuel (UO2 or MOX) at different burnups, using different cladding materials. Among other things, the program will provide a fuller understanding of thermal and mechanical processes that are observed in the fuel rods and that can impair their integrity, in particular during the advanced phase of the accident.

Learn more: summary of the CABRI research program

In a new report, IRSN presents the main characteristics of passive safety systems and outlines the main difficulties associated with assessing the performance and reliability of passive safety systems for nuclear reactors, as well as priority research areas to be developed in order to overcome these difficulties.

Pressurized water reactors currently operating in France are equipped with active safety systems requiring a power source, such as an electrical power supply, and also include passive safety features (nuclear fission reaction control and shutdown rods, hydrogen recombiners, etc.).

Certain nuclear reactor designs currently under construction or development make more extensive use of passive safety systems in order to bring the reactor to a safe shutdown state and maintain this state for a long period of time without need for human intervention and with limited reliance on support functions.

IRSN considerations to date on passive safety systems have led to the identification of a number of intrinsic difficulties, particularly concerning the performance and reliability assessment of such systems.

Further research is required in order to properly assess the performance and reliability of passive safety systems to be implemented in new reactor designs. Initial considerations have already been identified as part of IRSN's scientific strategy, with emphasis on necessary understanding of physical phenomena influencing the operation of passive safety systems, necessary simulation capabilities for such phenomena, and necessary testing for validation of simulation software.

IRSN pursues this research within the framework of joint actions with foreign organisations so as to ensure fruitful exchanges and benefit from available knowledge.

Download IRSN report on passive safety systems for nuclear reactors (PDF)

IRSN has recently launched a national program of experiments called PROGRES to determine how agglomerations of debris in a reactor core could be cooled by injecting water into the vessel.

Retaining the corium resulting from a core meltdown accident within the reactor vessel or containment represents a major safety objective for the protection of people and the environment. 

For this reason, IRSN, as leader of the European IVMR project, has recently launched a national program of experiments called PROGRES to determine how agglomerations of debris in a reactor core could be cooled by injecting water into the vessel to enhance the retention strategy based on external vessel cooling.

Can the corium resulting from core degradation in a damaged reactor be cooled effectively enough to slow and prevent the progression of the accident towards total core meltdown and potential reactor vessel failure? Can it be stabilized and retained in the reactor vessel, even for higher power reactors (900 MWe and above)? 

Expected to run for five years under IRSN’s leadership, the European IVMR project (In-Vessel Melt Retention), which brings together 23 partners, raises key questions regarding corium behavior in the vessel in the event of a severe accident. 

In this context, one part of IRSN’s PROGRES program will study more specifically the cooling of the various types of debris agglomeration that may form in the vessel of a damaged reactor. Conducted at IRSN's PEARL facility, the program will subsequently extend beyond the vessel, assuming its failure. In fact, whether the debris bed is located inside the reactor core, at the bottom of the vessel or in the reactor pit, the phenomena surrounding corium cooling can be formulated in the same scientific terms.

This work forms part of a strategic IRSN goal – underpinned by the lessons learned from the Chernobyl and Fukushima accidents – to fill the remaining gaps in knowledge in order to help develop effective solutions to prevent, limit and control large radioactive releases in the event of a severe accident.

More information about the IVMR projet

In an article published on 8th June 2015, the French online journal Mediapart mentions “a confidential report” from IRSN about the safety valves on the European Pressurized Reactor (EPR) under construction at the Flamanville nuclear power plant, and another report by IRSN on the manufacturing defects in the lower and upper head of the reactor vessel.

IRSN examines the design of the EPR reactor for many years. The technical instruction conducted by IRSN is assessed on the basis of documents submitted by EDF: these documents lead to technical exchanges between IRSN, EDF, the designer AREVA and the French nuclear safety authority (ASN). These discussions are not public. At the end of its examination, IRSN transmits the conclusions of its expertise to ASN in the form of a “review”. After the decision of the French safety authority, IRSN’s review may be published on its website. As part of the instruction of the EPR reactor under construction in Flamanville, IRSN has already submitted about 150 reviews on many subjects (equipment, accident studies ...).

In 2014, IRSN began the review of the pressurizer safety valves design of the EPR: the pressurizer's role is to maintain a given pressure in the reactor coolant system, and is equipped with three safety valves in order to ensure protection against overpressure. The examination performed by IRSN aimed to ensure that these safety valves meet the requirements according to the high reliability expected from this equipment for the reactor coolant protection (opening pressure, closing pressure…).In this context, IRSN had technical meetings with EDF, and witnessed the qualification tests performed in France for the safety valves: these tests in particular revealed an unexpected behavior, including a failure to open at the expected pressure.

At a meeting in February 2015, IRSN presented to ASN the status of its examination, including the analysis of risks of failure of the safety valves and the results of ongoing qualification tests. The report mentioned by Mediapart is actually a “working document” (slides), prepared for the meeting, which reports IRSN questions at that time. The technical instruction of the pressurizer safety valves design of the EPR is still ongoing and still lead to technical exchanges with EDF. IRSN expects to transmit the conclusions of its examination to ASN during the summer of 2015. These conclusions may include recommendations for additional or corrective actions.

The article also mentions a report by IRSN on the manufacturing defects in the lower and upper head of the Flamanville EPR reactor vessel. This information comes from a first IRSN review transmitted to ASN in early April 2015. The instruction of this subject will continue on the basis of additional elements submitted by AREVA, elements which now require thorough analysis by IRSN, with the assistance of external experts if necessary.

Download this information note in PDF

As part of the third ten-year inspection of 1,300 MWe reactors, IRSN has published a summary (in French) of its assessment of actions taken by EDF regarding severe accidents, on the basis of which ASN has issued a position in letters addressed to EDF on November 25, 2013 and January 20, 2014.

Rather logically, in connection with the third ten-year inspection of these reactors and with respect to severe accidents, EDF is proposing to focus on improving radioactive materials containment and on reducing the probability of scenarios likely to lead to early or significant radioactive releases.

While finding the facility modifications proposed by EDF to be pertinent in principle, IRSN's assessment identified several potential avenues for improvement, in particular concerning EDF’s planned measures to eliminate any possibility of a steam explosion in the event of a reactor vessel breach following a core meltdown.

In the field of level 2 probabilistic safety assessments, used to assess the probability of occurrence of an undesirable event and its potential consequences in terms of radioactive releases to the environment, IRSN noted that EDF's initiative was generally consistent with the practices of other countries. However, the Institute also found that certain risks – such as hydrogen combustion in the space between the two superposed containment buildings (a specificity of these reactors) – were not taken into account and considered that additional studies were needed to give a more realistic grasp of certain environmental release risks. For the operator, the challenge is to demonstrate that its proposals help achieve the lowest level of risk possible, given the intrinsic technological characteristics of these reactors.

More information in French: IRSN's summary of assessment of actions taken by EDF regarding severe accidents

A nuclear emergency response drill involving a shipment of radioactive materials between the FBFC plant in Romans-sur-Isère in southeastern France and the port of Antwerp in northwestern Belgium was carried out on April 2, 2014.

The scenario was based on a fictional collision, followed by a fire, between a truck carrying containers of enriched uranium hexafluoride and a tanker truck on the border of France and Belgium. The drill was designed primarily to test relations between the authorities in the two countries and to coordinate local resources. It will serve as the basis for a larger-scale field exercise to be organized in the near future.

Carried out as part of the project called “Innovative integrative tools and platforms to be prepared for radiological emergencies and post-accident response in Europe” (PREPARE), a project designed to develop emergency response management tools funded by the European Commission and performed with IRSN's involvement, the drill was unique in that it involved testing the relations of two bordering countries, Belgium and France, in a nuclear emergency situation. Based at the IRSN Technical Center for Emergency Response near Paris, the drill brought together some thirty people from the FANC (Belgian Federal Agency for Nuclear Control) and ASN (the competent authorities of Belgium and France), their respective technical safety organizations (Bel V for Belgium and IRSN for France), and the French carrier TN International.

The principal purpose of the exercise was to test relations between the authorities and their coordination of local resources, and secondly to identify necessary improvements, both in terms of the formal relations between the authorities and technical support organizations and in terms of coordinating local resources. IRSN, ASN, FANC, Bel V and TNI are planning to hold another field exercise on a larger scale in the near future.

The ERMSAR (European Review Meeting on Severe Accident Research) conference organized by IRSN in Avignon (France) in October 2013 gave a broad view of four years of R&D relating to major nuclear reactor accidents. This event was the opportunity to promote the decision taken by the SARNET network of excellence to continue its activities in the framework of the NUGENIA association (Nuclear Generation II & III Association).

The SARNET network of excellence reached a special milestone with the 6th ERMSAR conference, attended by 137 participants from 61 organizations and 25 countries, confirming its status as a leading international event on major nuclear reactor accidents. Among other things, the ERMSAR conference defined the highest priority areas for future research, considering the first feedback of the Fukushima Daiichi accident. The future research topics will notably focus on developing technology systems aiming to limit the consequences of such accidents.

It was officially announced that, after operating for eight years under the European Commission’s supervision, SARNET will continue to run as one Technical Area of the NUGENIA association, of which IRSN is a founding member, and which is recognized as the key European vector for research on the safety of generation II and III reactors. IRSN will continue to coordinate the network, confirming its leadership position with regard to this crucial safety research issues.

More information:

Presentation of SARNET

NUGENIA website

The publication by Reuters on August 5, 2013 of a news report about the situation at the Fukushima Daiichi nuclear power plant (read) has revived questions concerning the management of the contaminated water on the site of Fukushima-Daiichi.

For IRSN, there was no sudden aggravation of the situation in recent days, but statements by the authority present at the site reminding the operator TEPCO of the need to put in place as quickly as possible corrective actions regarding the diffuse contamination of the Pacific Ocean.

Volumes of contaminated water at the site are estimated at several hundreds of thousands of cubic meter. The natural phenomena that led to the accident that affected TEPCO’s Fukushima Daiichi nuclear power plant on March 11, 2011 also led to flooding of the site leading to an accumulation of water in the basements of the power plant buildings. Furthermore, since the accident, the water used to cool the damaged cores of the reactors has been flowing into the basements of the buildings from where it is pumped in order to be re-used, after treatment, to cool the reactors.

However, the galleries below the plant are not completely sealed; there is a suspicion of contamination of groundwater. Tepco is trying to strengthen the leak tightness of the ground by injection of sealing products and by creating a first barrier between the facilities and the ocean (expected to be completed by mid-2014).

For more information on the situation, download the information notes by IRSN:

Fukushima Daiichi nuclear accident: Management of contaminated water from the damaged reactors (PDF)

Fukushima Daiichi nuclear accident: Contamination of the ground between the damaged reactors and the Pacific Ocean  (PDF)

IRSN and GRS are jointly developing the ASTEC software system dedicated to nuclear meltdowns in the different types of reactors currently in operation. A January 2013 meeting of the ASTEC Users Club in Aix-en-Provence in southeastern France bolstered its position as an international reference and consolidated its scope of applications, which has progressively been extended to include most types of reactors, existing and future.

Over the past decade, ASTEC has earned its status as an international reference in the simulation of core meltdown accidents in pressurized water reactors – including the Russian-designed VVER – with the help of the SARNET network [1] in particular.

Today, no fewer than 36 organizations use ASTEC, and three major events have just strengthened its position as the leading reference in this type of software:

• the selection of ASTEC by JRC/IET [2] for use in its core meltdown accident research laboratory,
• China Nuclear Power Engineering’s decision to acquire a five-year commercial user license for ASTEC,
• the beginning of the Code for European Severe Accident Management [3] project of the 7th FRDP.

The modular nature of ASTEC allows its scope of application to be extended to a wide range of reactors such as boiling water reactors, to enable a detailed analysis of the Fukushima Daiichi accidents and of the IPHWR [4] for collaboration with India’s BARC [5]. ASTEC is already being applied to naval propulsion reactors, but will benefit from certain improvements following IRSN-led development work as part of an agreement with the CEA/DAM [6].

For accidents in other types of reactors, ASTEC is being used to model accidents involving the ingress of air or water into the vacuum vessel of fusion facilities such as the ITER and is being adapted for sodium-cooled fast reactors as part of the 7th FRDP project entitled "Joint Advanced Severe Accidents Modelling and Integration for Na-Cooled Fast Neutron Reactors" coordinated by IRSN.

Notes:

1. Excellence network specialized in research on severe nuclear reactor accidents.
2. Institute for Energy and Transport of the Joint Research Centre of the European Union, located in Petten (the Netherlands).
3. Coordinated by GRS working closely with IRSN, this project brings 17 partners together to improve the modeling and functionality  of the ASTEC software for severe accident management.
4. Indian Pressurized Heavy Water Reactors (developed by India).
5. Bhabha Atomic Research Centre.
6. Military applications division of the French Alternative Energies and Atomic Energy Commission.