As part of preparations for the public debate on the 2019-2021 French National Plan for the Management of Radioactive Materials and Waste (2019-2021 PNGMDR), the President of the National Public Debate Commi​ssion asked the French Institute for Radiological Protection and Nuclear Safety (IRSN) to carry out an assessment of the dry storage of spent nuclear fuels containing mixed uranium and plutonium oxide (MOX) or enriched reprocessed uranium oxide (ERU).

This report complements the IRSN n°2019-00181 report on concepts and safety issues related to the storage of spent nuclear fuel published in March 2019 (French issue in June 2018) in response to a request from the Parliamentary Committee of Inquiry into the Safety and Security of Nuclear Facilities in France.

IRSN examined, on the one hand the potential compatibility with dry storage of some of the MOX or ERU spent fuels currently stored underwater, and, on the other hand, the potential changes to transport and dry storage concepts in order to raise the reference residual heat values at present accepted, namely below 6 kW for transport and below 2 kW for dry storage.

In conclusion, IRSN’s assessment did not reveal any factors that would rule out storing in dry conditions some of the MOX and ERU fuels currently stored underwater. Nevertheless, the various possible options should be examined, incorporating the related safety and radiation protection requirements as well as all industrial constraints.

The French version of this report was pu​blished in April 2019, and the English version in January 2020.

Download IRSN Report 2019-00903: Assessment of dry storage possibilities for MOX or ERU spent fuels (PDF)

​An earthquake occurred on November 11, 2019 in the Ardèche region in Le Teil municipality at approximately 10 km to the west of the city Montelimar. The ground motion lasted several seconds in the vicinity of the epicenter and was felt by the population in the south-east of France, particularly up in Saint-Etienne, Grenoble, Lyon, Montpellier and Marseille.

Download the November 12th information report: Le Teil Earthquake of November 11, 2019

IRSN is upgrading its operating experience feedback (OPEX) system to improve its application in the field of expertise. Several experiments have already demonstrated the interest for implementing business intelligence, machine learning and cognitive search technologies. To extract knowledge and provide insights through data science, the Institute also granted access to “Sapide”, its database of significant events reported by the operators, for the magazine “Contexte”.

Like any risk activity, the nuclear industry relies on experience feedback to improve the safety of nuclear facilities. Any deficiency is analyzed to tackle the deficiency’s source and the past analyses are also used over time to learn about new topics. At IRSN, the Sapide’s database capitalizes on the experience feedback regarding French pressurized-water reactors (PWRs) since 1993. It contains 26,500 events, reported by EDF since the commissioning of the first two 900 MWe French PWRs in Fessenheim (Alsace) in 1979.

Sapide has evolved to track the changing technological and regulatory environments. At a very early stage, the Institute digitalized and categorized the oldest documents to facilitate the retrieval of information by experts and to contribute to the assessment of risk control in terms of safety, security, radiation protection and environment protection.

However, data processing, which relies on pre-established categories, restricts application of the database, in particular the identification of new subjects to investigate such as changes of both processes and organizations. "What is counted in the historical approaches of the treatment of experience feedback is not always interpretable in terms of risk control", summarizes Karine Herviou, director of Safety approaches, Systems and New Reactors at IRSN.

In the view of all these, IRSN launched the OPEX project, which would be rolled out gradually over the next 3 years. The query in Sapide had to be improved to help the experts exerts the most from the data in order to render more relevant opinions. The range of opinions could vary from the analysis of an event, a trend, a specific theme to an overall assessment of the risk management by various French operators.

According to Hervé Bodineau, the manager of the PWR safety department at IRSN, "raw data can be used to identify new subjects and refine knowledge on existing subject". This scheme has been tested to prepare the commissioning tests of the EPR reactor in Flamanville. As a part of that experiment, an assessment was carried out by exploiting the data from previous commissioning tests in the 1980s and 1990s.

This new scheme requires methods and tools that can make data more meaningful and adapt its restitution according to the uses. That should also allow a transversal analysis, which combines different disciplines (safety, human and environmental radiation protection), different types of facilities, technical and organizational aspects involved in the occurrence of events. "One of the expectations of the OPEX project is to be able to learn more from the significant events reported by EDF" illustrates Hervé Bodineau.

The OPEX project is exploring several technological opportunities, through three proofs of concept (POC):

• In terms of business intelligence, the visualization of data and analytical elements for decision support;
• Automatic language processing and machine learning in support of event analysis;
• The use of search engines implementing principles of cognitive search.

The POCs were used to establish the architecture of the OPEX project to help the experts to identify relevant lessons for risk control. "To go further, we must know how to complete and evolve our practices on experience feedback, says Karine Herviou. It is necessary to complete the descriptive approach with analytical elements, specific to experts work and focused on the uses".

Data Science: Another discipline to enhance the expertise

Does the age of a reactor have an influence on its safety? Is a surge in events reported by EDF a sign of premature ageing of a reactor? For Context magazine, which has just devoted an article to the question, the answer is rather negative. To get to this point, the magazine analyzed in statistical and mathematical form the Sapide database. This data journalism report was conducted in parallel with the OPEX project.

"The database is not public. Therefore, IRSN has decided to partially open it after informing the operators", says Hervé Bodineau. This collaboration has confirmed Sapide's robustness and the potential of data science to carry out expert missions.

Thus, it is possible to reconsider old data to answer new questions. Statistical processing may also allow, under certain conditions, the identification of weak signals to be investigated by the operators.

Along with their NEA peers, a number of IRSN experts will contribute actively to three projects to prepare for the decommissioning of the damaged Fukushima Daiichi reactors: one project to prepare the recovery and analysis of fuel debris (PreADES PreADES [http://www.oecd-nea.org/jointproj/preades.html]), another to examine the state of the damaged reactors (ARC-F) in more detail, and another to thermodynamically characterize the fuel debris and fission products (TCOFF).

IRSN intends to share its expertise internationally: by contributing to the synthesis of knowledge acquired from the Fukushima Daiichi accident, by contributing its expertise to the analysis of debris samples from the damaged reactors, by upgrading its thermodynamic databases (NUCLEA, MEPHISTA).

IRSN expects that its participation in these three projects will help to strengthen its expertise capacity on severe accidents: it will gain a better understanding of the limits of the various tools and methods it has developed through its studies and research, by applying them to the reality of an accident that has been thoroughly analyzed.

On 16 April 2018, IRSN and the CEA successfully completed a first test simulating an accident situation in the Cabri reactor at Cadarache as part of an experimental programme. This experiment was conducted on a new pressurised water loop and is the first step in the CIP international research programme to improve reactor safety.

Purpose of the test

The Cabri International Programme (CIP) seeks to improve knowledge of fuel behaviour in pressurised water reactors (PWR) during an accident involving a sudden increase of power in the reactor. The first test under the programme was conducted on 16 April 2018, on the renovated CABRI reactor, equipped with an experimental pressurised water loop [1]. This loop will be used to analyse fuel behaviour in a thermal-hydraulic configuration, representative of a PWR. The instrumentation installed by IRSN in the CABRI reactor can take vital measurements for interpreting tests. The Hodoscope is the only equipment of its kind in the world to provide real-time measurements of the power along the rod tested, its elongation and any potential fuel relocation.

The purpose of the first test is to ensure that the new loop works properly, study boiling following heat transfer from the fuel rod to the water, and the behaviour of fuel in the event of sudden and mass heat injection.

According to Jean-Christophe Niel, Director General of IRSN, “the next step in the programme hinges on testing the irradiated fuel rods in an environment as close as possible to that of a pressurised water reactor core. These conditions have never before been created in this kind of experiment. The aim is to obtain the knowledge required to assess the safety of PWR reactors.” The test will study heat exchange between the coolant and rod cladding, along with the thermal interaction between the ejected fuel fragments and water in the event of cladding rupture.

IRSN research on fuel behaviour in accident situations

IRSN research focuses on fuel behaviour under various accident situations that could affect pressurised water reactors:

• uncontrolled development of the nuclear reaction resulting from ejection of a control rod assembly formed of absorbent rods that help control the nuclear reaction;
• loss of coolant accident (LOCA) caused by rupture of the reactor coolant pressure boundary;
• dewatering of a spent fuel pool, in particular subsequent to feedback from the Fukushima-Daiichi accident.

These research projects are essential for developing the expertise of IRSN teams.

Jean-Christophe Niel underlines that “these research projects have objectives shared by the international community and need to promote partnerships in France and abroad with IRSN counterpart institutions, in particular the ETSON network [2], and with research bodies such as the CEA, with whom we are working hard to ensure the success of the Cabri International Programme (CIP).”

In addition to the experimental programmes conducted in the CABRI reactor, IRSN research draws on a large database of international experiments, comprising results from reactor tests (NSSR in Japan, BIGR in Russia, etc.) and analytical tests on cladding behaviour and heat transfers. IRSN is also participating in a number of international research programmes, in particular an OECD project in Halden, Norway, and another in Studsvik, Sweden. It is also carrying out projects funded by the French National Research Agency (ANR) in a post-Fukushima context, on fuel behaviour in a spent fuel tank in the event of dewatering, and on the coolant of a fuel assembly deformed during a LOCA. In addition to large-scale tests under representative conditions, earlier work is carried out with academics to develop and approve the models later integrated into simulation software used for safety studies.

Notes:

1. Two initial tests were conducted in a “sodium” loop before renovation of the reactor and the development of the pressurised water loop. It will be possible to compare these results with results from the tests conducted in the pressurised water loop.
2. The main role of technical safety organisations (TSO) is to scientifically assess the safety of nuclear facilities and the radiation risks. Within the international network, Etson, they now also work on research, discussions concerning future choices, training and even communication with society on nuclear and radiation risks.

CEA CABRI research reactor at Cadarach

The CEA operates the CABRI research reactor on behalf of IRSN. It is its experimental facility to study the behaviour of nuclear fuel in the event of an accidental power increase, located at the CEA Cadarache research centre in the South of France (Bouches-du-Rhône).

The reactor was built in 1962 and has been adapted since its construction for the purposes of safety studies on French nuclear facilities, from fast neutron reactors to recent pressurised water reactors.

It has therefore undergone a major renovation programme (civil engineering, fire protection, etc.) in order to comply with current safety standards and install an experimental pressurised water loop. As acknowledged by François Gauché, Director of Nuclear Energy at the CEA, “this renovation programme required significant resources from the CEA and IRSN over a long period of time. This first pressurised water loop test is a token of our collective success.”

One specific property of the CABRI reactor is the rod system in its core, which uses an ultra-rapid depressurisation phenomenon to change the reactor power by a factor of around 200,000, from 0.1 MW to 20,000 MW in just a few milliseconds, for a very short duration of 10 to 100 ms.

About the CEA
The CEA is a public research body that works in four fields: defence and security, nuclear and renewable energies, technological research for industry and fundamental research. Drawing on its widely acknowledged expertise, the CEA actively participates in collaborative projects with a large number of academic and industrial partners. With its 16,000 researchers and staff, it is a major player in the European Research Area and its international presence is constantly growing. www.cea.fr

About IRSN
IRSN is a public body with industrial and commercial activities (EPIC). Its missions are now defined by French Act no. 2015-992 of 17 August 2015 on the energy transition for green growth (TECV). It is the public expert in France on nuclear and radiation risks. IRSN contributes to public policy on nuclear safety and human and environmental protection against ionising radiation. It is a research and assessment body, working with all players affected by these policies, while retaining its independent position. IRSN operates under the joint authority of the Ministries of the Environment, Research, Energy, Health and Defence.

The implementation of the defence in depth principle leads applicants to define provisions to prevent accidents, including severe accidents, and to limit their consequences should they occur. Some severe accident situations potentially lead to large early radiological releases, for which it appears impossible to define realistic and demonstrable provisions to limit their consequences according to current knowledge and the techniques available: applicants shall then use the “practical elimination” approach.

IRSN publishes an orientation document which defines “practical elimination” approach and its place in safety demonstration for nuclear power plants involving pressurized water reactors.

Download the document: The practical elimination approach of accident situations for water-cooled nuclear power reactors (document updated in March 2021)

From their joint analysis of the files submitted by Areva NP, IRSN (the French technical support organisation), and the Nuclear Pressure Equipment Department (DEP) of ASN (the French nuclear safety authority) conclude that if the serviceability of the closure head and lower head of the Flamanville 3 reactor pressure vessel is demonstrated, in-service inspections must be implemented. At this stage, the feasibility of these inspections appears to have been acquired for the lower head. The same is not true for the closure head.

At the end of 2014, Areva NP discovered a manufacturing anomaly in the closure head and lower head of the Flamanville 3 reactor pressure vessel. This anomaly undermines certain mechanical characteristics of the steel of these components, in particular its toughness, i.e. its ability to withstand crack propagation in the event of a pre-existing flaw.

In 2015, Areva NP proposed an approach to demonstrate the sufficiency of the steel's toughnessfor the closure head and lower head of the pressure vessel. This approach was examined jointly by IRSN and the Nuclear Pressure Equipment Department (DEP) of ASN, and reviewed by the Advisory Committee for nuclear pressure equipment (GP ESPN) on 30 September 2015. ASN then took position and considered acceptable, in principle, the demonstration approach proposed by Areva NP.

At the end of 2016, Areva NP transmitted its analysis of the consequences of the anomaly of the closure head and lower head of the Flamanville EPR reactor pressure vessel. Areva NP concludes that the two components are serviceable.

This analysis was the subject of a joint ASN DEP - IRSN assessment, the conclusions of which were presented on 26 and 27 June 2017 to the GP ESPN. From their instruction, ASN DEP and IRSN conclude that if Areva NP has demonstrated the serviceability of the closure head and lower head of the pressure vessel of the EPR reactor in Flamanville, in-service inspections must be implemented to periodically check this equipment during the operation of the installation (expected operating life of 60 years).

At this stage, the feasibility of these inspections appears to have been acquired for the lower head. The same is not true for the closure head: failing to implement these inspections, ASN DEP and IRSN consider that the replacement of this closure head should be carried out a few years away.

More information on the assessment of the consequences of the anomaly

IRSN’s information report of 28 June 2017: Note on the assessment of the closure head and lower head of the Flamanville 3 EPR reactor pressure vessel

ASN DEP-IRSN joint report of 15 June 2017: Analysis of the consequences of the anomaly in the Flamanville EPR reactor pressure vessel head domes on their serviceability

More information on the analysis of the demonstration approach

ASN DEP-IRSN joint report of 15 September 2015: Analysis of the procedure proposed by AREVA to prove adequate toughness of the domes of the Flamanville 3 EPR reactor pressure vessel lower head and closure head

Opinion of the Advisory Committee for nuclear pressure equipment  (GP ESPN) of 30 September 2015: Opinion concerning the procedure proposed by AREVA to demonstrate the adequate toughness of the domes of the Flamanville 3 EPR reactor pressure vessel (RPV) bottom head and closure head

Position statement of ASN: Letter of 14 December 2015 from ASN to Areva

Position statement of ASN: Interim review of the approach to demonstrate the adequate toughness of the reactor pressure vessel upper and lower domes - 26 September 2016

On 30 November 2016, IRSN submitted to ASN its assessment of the risks of fracture in steam generators (SG’s) with abnormally high levels of carbon in the steel that makes up their channel head.

The local carbon levels measured in these channel heads are in the region of 0.4 % compared to the maximum expected value of 0.22%. This anomaly meant that a re-examination of the risks of fracture of those steam generators was required since the mechanical properties of the steel have been altered.

The assessment relates to the SG’s in 900 MWe reactors whose channel heads were manufactured by the company Japan Casting and Forging Corporation (JCFC).

The evaluation of the risks of failure in the SG’s consists of examining the risks of fracture initiation resulting from a known or postulated crack in the steel. Thus, IRSN evaluated in particular the extent of the cracks postulated by EDF, the loads that give rise to stresses likely to open the postulated cracks, i.e. the thermal shocks that can affect the SG’s under all reactor operating conditions, as well as the mechanical properties of the steel, taking account of the preliminary data provided by EDF for the steel showing 0.4 % carbon levels.

The evaluation conducted by IRSN has led it to conclude that there is no risk of fracture for the SG’s manufactured by JCFC and fitted in the 900 MWe reactors covered by the assessment (Bugey 4, Dampierre 3, Fessenheim 1, Gravelines 2 and 4, Saint-Laurent-des-Eaux B1, Tricastin 1, 2, 3 and 4), with the exception of three reactors (Bugey 4, Fessenheim 1 and Tricastin 4) subject to the recommendations of IRSN relating to limiting the extent of the potential thermal shocks and the results of the inspections requested by ASN. For the three aforementioned reactors, the study still needs to be finalised and submitted by EDF.

In order to reaffirm its assessment of the mechanical properties of the steels showing high carbon levels, IRSN has called on the assistance of its Belgian expert assessor counterpart, BEL-V, which is a member of the European ETSON network. ASN and IRSN also went to Japan to visit the manufacturer JCFC in order to better understand the cause of the excessive carbon levels observed.

IRSN assessment has been published on its website and presented to the press during a joint conference with ASN on 5 December 2016. It has been be presented to the High Committee for Transparency and Information on Nuclear Safety (HCTISN) at its meeting on 6 December 2016.

Moreover, in conjunction with the National Association of Local Information Commissions and Committees (ANCCLI) and ASN, a meeting has been organised for February 2017 to hold an exchange of views with civil society stakeholders as an extension to the technical dialogue launched with ANCCLI, the Flamanville Local Information Commission and ASN on the topic of similar anomalies affecting the reactor vessel at the EPR in Flamanville.

Download IRSN information notice of 5 December 2016: IRSN assessment of the safety of reactors equipped with steam generators whose channel heads contain an abnormally high level of carbon