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Concerns within the nuclear data community led to substantial increases of Neutron Data Standards (NDS) uncertainties from its previous to the current version. For example, those associated with the NDS reference cross section 239Pu(n,f) increased from 0.6–1.6% to 1.3–1.7% from 0.1–20 MeV. These cross sections, among others, were adopted, e.g., by ENDF/B-VII.1 (previous NDS) and ENDF/B-VIII.0 (current NDS). There has been a strong desire to be able to validate these increases based on objective criteria given their impact on our understanding of various application uncertainties. Here, the “Physical Uncertainty Bounds” method (PUBs) by Vaughan et al. is applied to validate evaluated uncertainties obtained by a statistical analysis of experimental data. We investigate with PUBs whether ENDF/B-VII.1 or ENDF/B-VIII.0 239Pu(n,f) cross-section uncertainties are more realistic given the information content used for the actual evaluation. It is shown that the associated conservative (1.5–1.8%) and minimal realistic (1.1–1.3%) uncertainty bounds obtained by PUBs enclose ENDF/B-VIII.0 uncertainties and indicate that ENDF/B-VII.1 uncertainties are underestimated.

Effect of the [U(IV)/U(III)] ratio of fuel salt on selective chromium corrosion and tellurium intergranular cracking (IGC) of Hastelloy N alloy in the LiF-BeF2-UF4 salt mixture was investigated. The chromium corrosion of Hastelloy N alloy is caused by the oxidation of chromium on the alloy surface by reaction with UF4. The tellurium IGC of Hastelloy N alloy is caused by the diffusion of tellurium along the grain boundaries with the formation of unstable tellurides with based metals and alloying additives. Results indicate that the selective chromium corrosion and the tellurium IGC of the Hastelloy N alloy in fuel salt can be controlled by the [U(IV)]/[U(III)] ratio. The tellurium IGC of Hastelloy N alloy exposed in fuel LiF-BeF2-UF4 salt can be avoided. For temperatures up to 760 °C the selective chromium corrosion can be minimized to the acceptable level when the [U(IV)]/[U(III)] ratio of fuel salt is bellow 30–40.

Within the H2020 INSIDER project, the main objective of work package 3 (WP3) is to draft a sampling guide for initial nuclear site characterization in constraint environments, before decommissioning, based on a statistical approach. The second task of WP3 aims at developing a strategy for sampling in the field of initial nuclear site characterization in view of decommissioning, with the most important goal to guide the end user to appropriate statistical methods (including, but not limited to those identified during the first overview task) to use for data analysis and sampling design. To aid the end user in applying this strategy, a user-friendly application for guiding the end user through the contents of the strategy and the initial characterization process is also developed.

In the present work the new data on correlation and covariance matrices have been calculated for all recommended data on temporal parameters of delayed neutrons in 6- and 8-group models. The method for producing the correlation and covariance data for the temporal DN parameters is described. The table of changes is listed in the text. Some examples of the data generated on basis of experimental DN decay curves (for IPPE data) and data generated on the basis of decay curves restored using temporal parameters (for works of another authors) are listed too.

Aiming at economical optimization, the characterisation of the biological shield of the Belgian Reactor 3 is one of the three use cases intended to validate the integrated characterization methodology developed within the INSIDER project. Pre-existing data were used to define the sampling design strategy. The additional sampling and analysis program consisted of total gamma measurements at the inner surface of the biological shield (secondary data) and gamma spectrometry measurements on drill core samples (primary data). The newly acquired data is supplemented with the historical available data. The full data set currently consists of a total of 283 secondary and 379 primary data points. Preliminary calculations already provide a clear-cut representation of the three different end-stage classes: unconditional clearance, conditional clearance and radioactive waste. On the short term, the current model will be further refined and completed with proper risk evaluation. On the longer term, we envisage a global uncertainty calculation and sensitivity analysis of the entire process.

The Euratom Research and Training Programme 2014–2018 and its extension 2019–2020 (the Euratom Programme) is implemented through direct actions in fission − i.e. research performed by the Commission's Joint Research Centre (JRC), and through indirect actions in fission– i.e. via competitive calls for proposals, and in fusion − i.e. through a comprehensive named-beneficiary co-fund action managed by the Commission's Directorate-General for Research & Innovation (RTD). The general objective of the Programme is “to pursue nuclear research and training activities with an emphasis on the continuous improvement of nuclear safety, security and radiation protection, in particular to potentially contribute to the long-term decarbonisation of the energy system in a safe, efficient and secure way.” The Programme is an integral part of Horizon 2020, the EU Framework Programme for Research and Innovation. The direct actions implemented by the JRC constitute an important part of the Euratom Programme and pursue specific objectives covering: nuclear safety, radioactive waste management, decommissioning, emergency preparedness; nuclear security, safeguards and non-proliferation; standardisation; knowledge management; education and training; and support to the policy of the Union on these fields. The JRC multi-annual work programme for nuclear activities fully reflects the aforementioned objectives. It is structured in about 20 projects, and allocates 48% of its resources to nuclear safety, waste management, decommissioning and emergency preparedness, 33% to nuclear security, safeguards and non-proliferation, 12% to reference standards, nuclear science and non-energy applications and 7% to education, training and knowledge management. To ensure that direct actions are in line with and complement the research and training needs of Member States, JRC is continuously interacting with the main research and scientific institutions in the EU, and actively participating in several technological platforms and associations. JRC also participates as part of the consortia in indirect actions, which allows JRC scientist to engage in top level scientific research, and yields maintaining and further developing JRC's scientific excellence. At the same time, the members of the consortia can have access to unique research infrastructure. The participation of JRC in indirect actions can be improved by exploiting synergies inside the Euratom Programme, and also with the future Horizon Europe Framework Programme. In preparation of the next Euratom Programme 2021–2025, two pilot projects on knowledge management and on open access to JRC research infrastructure will explore and test this improved involvement of JRC in indirect actions. The paper highlights some of the achievements of recent JRC direct actions with a focus on the interaction with EU MS research organisations, as well as some of the most important elements of the Commission Proposal for the next (2021–2025) Euratom Programme, with a focus on the new positioning of the JRC as regards its participation in indirect actions.

We present a cross-cutting review of three on-going Horizon 2020 projects (ADVISE, NOMAD, TEAM CABLES) and one already finished FP7 project (HARMONICS), which address the reliability of safety-relevant components and systems in nuclear power plants, with a scope ranging from the pressure vessel and primary loop to safety-critical software systems and electrical cables. The paper discusses scientific challenges faced in the beginning and achievements made throughout the projects, including the industrial impact and lessons learned. Two particular aspects highlighted concern the way the projects sought contact with end users, and the balance between industrial and academic partners. The paper concludes with an outlook on follow-up issues related to the long term operation of nuclear power plants.

Clean energy production is a challenge, which was so far addressed mainly in the electric power sector. More energy is needed in the form of heat for both district heating and industry. Nuclear power is the only technology fulfilling all 3 sustainability dimensions, namely economy, security of supply and environment. In this context, the European Nuclear Cogeneration Industrial Initiative (NC2I) has launched the projects NC2I-R and GEMINI+ aiming to prepare the deployment of High Temperature Gas-cooled Reactors (HTGR) for this purpose.

Recent EURATOM research efforts on Emergency Preparedness and Response (EP&R) have been focussed on programs addressing some main knowledge gaps clearly identified in the outcomes of investigations carried out in Europe in response to the Fukushima accident. The PREPARE and FASTNET projects tried to solve similar problems adopting very complementary and synergic approaches. The main achievements of both projects are detailed in this paper. In particular, the problem of the fast estimation of time-dependent, long-lasting Source Terms is discussed. This problem is not only a technical one, but is also related to the experience and skill of the code users. As the EP&R is spanning a wide range in Europe, certainly far beyond the borders of individual states, it is mandatory creating a common and shared understanding of emergencies. Both PREPARE and FASTNET recognized the fundamental role of exercises to increase the experience of emergency responders in Europe. A general recommendation can then be formulated, in that more efforts should be dedicated in the future to the realization of such important exercises.

Countries operating nuclear power plants have to deal with the tasks connected to spent fuel and high-level radioactive waste management. There is international consensus that, at this time, deep geological disposal represents the safest and most sustainable option as the end point of the management of high-level waste and spent fuel considered as waste. There are countries with longer timescale for deep geological repository (DGR) implementation, meaning that the planned date of commissioning of their respective DGRs is around 2060. For these countries cooperation, knowledge transfer, participation in RD&D programmes (like EURAD) and adaptation of good international practice could help in implementing their own programmes. In the paper the challenges and needs of a country with longer implementation timescale for DGR will be introduced through the example of Hungary.

For more than 40 years, considerable scientific and technical knowledge has been acquired in Europe in the field of radioactive waste management, both for near-surface disposal and geological disposal. RD&D will continue to be necessary to develop, maintain and consolidate knowledge throughout the stepwise development, operation and closure of disposal facilities, which will be spread over many decades and make this knowledge available to all end users. Recently, the EC has promoted a step-change in pan-European research cooperation between EU Member States' national programmes by promoting the setting-up of inclusive research joint programmes in Europe gathering those organisations with scientific and technical responsibilities and a national mandate for research in radioactive waste management. Based on the positive achievement of the JOPRAD project (2015–2017), the EC confirmed in 2017 its willingness to co-fund such a Joint Programme in the field of RWM. The RWM community therefore pursued the efforts to establish the Founding Documents (Vision, Strategic Research Agenda, Roadmap, Deployment) and a Work Plan for a first implementation phase of 5-years (2019–2024). In June 2019 the Joint Programme − EURAD − was accepted by the European Commission.

In this work, two different thermodynamic softwares, ANGE using the TBASE database, and OPENCALPHAD using the TAF-ID (Thermodynamics of Advanced Fuels – International Database), have been integrated into the GERMINAL V2 fuel performance code (of the PLEIADES platform) in order to evaluate the chemical state of (U, Pu)O2 fuel and fission products in sodium cooled fast reactors. A model to calculate the composition and the thickness of the “Joint-Oxyde Gaine” (JOG) fission product layer in the fuel-clad gap has been developed. Five fuel pins with a final burnup ranging between 3.8 and 13.4% FIMA (Fissions per Initial Metal Atom) have been simulated, and the calculated width of the fission product layer have been compared with post irradiation examinations. The two different thermodynamic softwares have been compared in terms of computation time and predicted fuel-to-clad gap chemistry. The main elements and phases encountered in the fission productlayer have been identified, and the impact of the changing oxygen potential has been explored.

Comparisons of calculated and experimental isotopic compositions of used nuclear fuels can provide valuable information on the quality of nuclear data involved in neutronic calculations. The experimental database used in the present study − containing more than a thousand isotopic ratio measurements for UOX and MOX fuels with burnup ranging from 10 GWd/t up to 85 GWd/t − allowed to investigate 45 isotopic ratios covering a large number of actinides (U, Np, Pu, Am and Cm) and fission products (Nd, Cs, Sm, Eu, Gd, Ru, Ce, Tc, Mo, Ag and Rh). The Integral Data Assimilation procedure implemented in the CONRAD code was used to provide nuclear data trends with realistic uncertainties for Pressurized Water Reactors (PWRs) applications. Results confirm the quality of the 235U, 239Pu and 241Pu neutron capture cross sections available in the JEFF-3.1.1 library; slight increases of +1.2 ± 2.4%, +0.5 ± 2.2% and +1.2 ± 4.2% are respectively suggested, these all being within the limits of the quoted uncertainties. Additional trends on the capture cross sections were also obtained for other actinides (236U, 238Pu, 240Pu, 242Pu, 241Am, 243Am, 245Cm) and fission products (103Rh, 153Eu, 154Eu) as well as for the 238U(n,2n) and 237Np(n,2n) reactions. Meaningful trends for the cumulative fission yields of 144Ce, 133Cs, 137Cs and 106Ru for the 235U(nth,f) and 239Pu(nth,f) reactions are also reported.

In this paper, a proper orthogonal decomposition based reduced-order model is presented for parametrized multiphysics computations. Our application physics is Doppler feedback in a simplified model of the molten salt fast reactor concept. The reduced model is created using the method of snapshots where the offline training set is obtained by exercising a full-order model created with the OpenFOAM based multiphysics solver, GeN-Foam. The steady state models solve the multi-group diffusion k-eigenvalue equations with moving precursors together with the energy equation. A fixed velocity field is assumed throughout the computations, hence the momentum and continuity equations are not solved. The discrete empirical interpolation method is used for the efficient coupling of the ROM solvers, while the input parameter space is surveyed using the improved distributed latin hypercube sampling algorithm.

In reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident.

In a liquid fuel molten salt reactor (MSR) a key factor to consider upon its design is the strong coupling between different physics present such as neutronics, thermo-mechanics and thermal-hydraulics. Focusing in the thermal-hydraulics aspect, it is required that the heat transfer is well characterized. For this purpose, turbulent models used for FLiNaK flow must be valid, and its thermophysical properties must be accurately described. In the literature, there are several expressions for each material property, with differences that can be significant. The goal of this study is to demonstrate and quantify the impact that the uncertainty in thermophysical properties has on key metrics of thermal hydraulic importance for MSRs, in particular on the heat transfer coefficient. In order to achieve this, computational fluid dynamics (CFD) simulations using the RANS k-ω SST model were compared to published experiment data on molten salt. Various correlations for FLiNaK’s material properties were used. It was observed that the uncertainty in FLiNaK’s thermophysical properties lead to a significant variance in the heat coefficient. Motivated by this, additional CFD simulations were done to obtain sensitivity coefficients for each thermophysical property. With this information, the effect of the variation of each one of the material properties on the heat transfer coefficient was quantified performing a one factor at a time approach (OAT). The results of this sensitivity analysis showed that the most critical thermophysical properties of FLiNaK towards the determination of the heat transfer coefficient are the viscosity and the thermal conductivity. More specifically the dimensionless sensitivity coefficient, which is defined as the percent variation of the heat transfer with respect to the percent variation of the respective property, was −0.51 and 0.67 respectively. According to the different correlations, the maximum percent variations for these properties is 18% and 26% respectively, which yields a variation in the predicted heat transfer coefficient as high as 9% and 17% for the viscosity and thermal conductivity, respectively. It was also demonstrated that the Nusselt number trends found from the simulations were captured much better using the Sieder Tate correlation than the Dittus Boelter correlation. Future work accommodating additional turbulence models and higher fidelity physics will help to determine whether the Sieder Tate expression truly captures the physics of interest or whether the agreement seen in the current work is simply reflective of the single turbulence model employed.

To demonstrate the robustness of extraction systems against radiolytic degradation is nowadays one of the limiting points to ensure a safe and stable operation for advanced nuclear fuel strategies. In this paper, is addressed the stability of one of most promising extractants (TODGA, N,N,N′,N′-tetraoctyldiglycolamide) but also the importance of designing realistic model to simulate and study the degradation of the systems. For that, new irradiations experiments were carried out where mixture between phases and the oxygen content have been taken into account. Extraction behaviour and composition of the organic phases after γ-irradiation have been measured and compared. Although TODGA studies are applicable to many processes currently under development, this work is focus on Grouped Actinides Extraction (GANEX) process development.

Coarse mesh nodal methods are widely used in the analysis of nuclear reactors. However, these methods provide only average values of the neutron fluxes. From a safety point of view, it is important to have an accurate analysis of the pin to pin flux distribution that nodal methods are not able to provide. Many articles have been published that make use of mathematical techniques to determine flux distributions. Most of these techniques use expansion functions to estimate these distributions. The expansion coefficients of these works are determined by conditions that take into account the average values of certain fluxes supplied by the nodal methods. There are also methods that employ analytical solutions of the neutron diffusion equation. This article presents a different approach for calculating the pin to pin neutron flux distribution for a PWR reactor. The developed method uses support vector regression (SVR) technique to determine this pin to pin neutron flux. The SVR technique uses average data computed with the Nodal Expansion Method (NEM) for learning purposes. A total of 70% of the computed data were used for training and 30% for validation, using multifold-cross-validation. Two fuel elements were removed from the training and validation sets, to test the method. Less than 2% errors were found when compared to the values ​​obtained by the nodal expansion method (NEM), using a fine-mesh spatial discretization. We concluded that use of SVR to reconstruct pin to pin fluxes is another option, which will be of great value in fuel reload calculations, since the same parameters will be applied to all cycles, thus expediting calculations when compared to standard procedure calculations.

A new approach to PWR simplification is presented, in which a compact Reactor Coolant System (RCS) configuration is introduced, particularly suited for a power level in the range of 600 MWe. Customary PWR primary system components are eliminated to achieve this RCS simplification. For example, RCS pressure control through a “self-pressurization” mode, with core exit at saturation temperature with less than 1% steam, allows elimination of a pressurizer. Also, mechanical control rods are replaced by reactivity control using negative moderator void and temperature coefficient together with variable speed primary pumps, and with an upgrade in the safety boration function. Decay heat removal in shutdown conditions is realized through the secondary side rather than through primary side equipment. The compact RCS can be installed in a small volume, high-pressure containment. The containment is divided into two leak-tight zones separated by a partition plate. Safety equipment installed in one of the two zones will be protected against adverse ambient conditions from leaks or breaks in the other zone. The partition facilitates management of coolant inventory within the RCS and the containment following RCS leaks or breaks. In particular, the safety injection system as commonly known, consisting of accumulators and multiple stages of injection pumps can be discarded and replaced by gravity-driven flooding tanks. Space available around major RCS components is adequate to avoid compromising accessibility during maintenance or in-service inspection operations. In addition, the two-zone, high-pressure containment provides extra margins in severe accident mitigation. Finally, the proposed containment has a much smaller size than customary large dry containments in PWR practice and it can be anticipated that Nuclear Island building size will similarly be reduced.

The PETALE experimental programme in the CROCUS reactor intends to provide integral measurements to constrain stainless steel nuclear data. This article presents the tools and the methodology developed to design and optimize the experiments, and its operating principle. Two acceleration techniques have been implemented in the Serpent2 code to perform a Total Monte Carlo uncertainty propagation using variance reduction and correlated sampling technique. Their application to the estimation of the expected reaction rates in dosimeters is also discussed, together with the estimation of the impact of the nuisance parameters of aluminium used in the experiment structures.

A correlated sampling technique has been implemented to estimate the impact of cross section modifications on the neutron transport and in Monte Carlo simulations in one single calculation. This implementation has been coupled to a Total Monte Carlo approach which consists in propagating nuclear data uncertainties with random cross section files. The TMC-CS (Total Monte Carlo with Correlated Sampling) approach offers an interesting speed-up of the associated computation time. This methodology is detailed in this paper, together with two application cases to validate and illustrate the gain provided by this technique: the highly enriched uranium/iron metal core reflected by a stainless-steel reflector HMI-001 benchmark, and the PETALE experimental programme in the CROCUS zero-power light water reactor.

The paper highlights the main features of some Euratom projects, which have been running recently in support to education, training and mobility in the nuclear fields. The described projects address various critical aspects of nuclear knowledge management, aiming at maintaining the wealth of nuclear expertise in Europe in an environment characterised by decreased attractiveness of nuclear careers. In an effort to broaden the cooperation and to further extend the opportunities for mobility, some projects ran in parallel with similar initiatives undertaken beyond the European borders. The lesson learnt in terms of successes achieved and critical aspects revealed by the different actions are finally discussed also considering recent recommendations and assessed scenarios by the European Commission for the decarbonisation of the energy sector.

We consider a model of neutron-nuclear wave burning. The traveling wave of nuclear burning of the medium is initiated by an external neutron source and is the basis for the new generation reactors the so-called “traveling-wave reactors”. We develop a model of nuclear traveling wave burning, for which it is possible to draw a Newton’s second law analogy with a mechanical dissipative system. On the basis of this analogy, we find that the wave velocity has a continuous spectrum bounded below. Within the framework of the new model, we show the autowave to be possible for certain neutron energies only. Also we find that two burning modes are possible depending on the control parameters: a traveling autowave and a wave driven by an external neutron source.

The Fukushima-Daiichi accidents in 2011 underlined the importance of severe accident management (SAM), including external events, in nuclear power plants (NPP) and the need of implementing efficient mitigation strategies. To this end, the Euratom work programmes for 2012 and 2013 was focused on nuclear safety, in particular on the management of a possible severe accident at the European level. Relying upon the outcomes of the successful Euratom SARNET and SARNET2 projects, new projects were launched addressing the highest priority issues, aimed at reducing the uncertainties still affecting the main phenomena. Among them, PASSAM and IVMR project led by IRSN, ALISA and SAFEST projects led by KIT, CESAM led by GRS and sCO2-HeRO lead by the University of Duisburg-Essen. The aim of the present paper is to give an overview on the main outcomes of these projects.

This manuscript presents important material challenges regarding innovative Gen-II/III nuclear systems and research reactors. The challenges are discussed alongside the key achievements so far realised within the framework of 4 EU-funded projects: H2020 IL TROVATORE, FP7 MULTIMETAL, FP7 MATTER and FP7 SCWR-FQT. All the four Projects deal with innovative researches on materials to enhance the safety of nuclear reactors. IL TROVATORE proposes new materials for fuel cladding of PWR reactors and tests in order to really find out an “Accident Tolerant Fuel” (ATF). MULTIMETAL focused on optimization of dissimilar welds fabrication having considered the field performances and dedicated experiments. MATTER carried on methodological and experimental studies on the use of grade 91 steel in the harsh environment of liquid metal cooled EU fast reactors. SCWR-FQT focused on fuel qualification of Supercritical Water Reactor including the selection of the better material to resist the associated high thermal flux.

International audience; From 2010 to 2019, the French Alternative Energies and Atomic Commission (CEA) associated with industrial partners realized the Basic Design of a prototype Sodium Fast Reactor. This project was called ASTRID (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration). ASTRID design studies were financed through governmental funds until the end of the basic design. These funds covered also the design studies for the core manufacturing workshop, the refurbishment or construction of large test loops. One year before the term of this Basic Design phase (in 2018), industrial partners, CEA and the French State conducted a review of fast neutrons reactors and fuel cycle strategy. The review which is now translated into the Multiannual Energy Program concluded that the perspective of industrial deployment of Fast Reactors is more distant. Yet it has been concluded to keep this option open, requiring to maintain competences, and to progress on technological barriers and further develop know-how. The strategy for complete closure of nuclear fuel cycle is maintained as a long-term sustainability objective (in the second half of the 21st century). Therefore, as a direct consequence of this decision, the ASTRID project stopped at the end of 2019 at its Basic Design phase. Quickly the question raised on the Knowledge Management (KM) and Know-How capitalization of the huge amount of studies and results realized during ten years (around 23 000 technical documents). Moreover the challenge was to realize this KM process in less than one year, before the ASTRID project team definitive split. The paper is presenting an innovative KM methodology which has been created and specifically performed on the ASTRID project. It is based on a series of interviews and video recordings, all transformed into some New KM tools called “MOOK” (MOOK for Management of Organized Online Knowledge). All these MOOKs considered as “data rich contents” are then inter-connected and linked by the ASTRID Product Breakdown Structure to some fundamental documents, for a comprehensive and quick mapping of the project. They finally form an efficient KM tool recorded in a PLM Software (PLM for Product Lifecycle Management). Thus the ASTRID project team has realized a high level and easy-to-use “GPS” (Global Positioning System) tool to keep the ASTRID history, context, knowledge and know-how for years. This KM methodology can be easily adapted to other nuclear projects and needs.

The ASGARD project (2012–2016) was designed to tackle the challenge the multi-dimensional questions dealing with the recyclability of novel nuclear fuels. These dimensions are: the scientific achievements, investigating how to increase the industrial applicability of the fabrication of these novel fuels, the bridging of the often separate physics and chemical communities in connection with nuclear fuel cycles and finally to create an ambitious education and training platform. This will be offered to younger scientists and will include a broadening of their experience by international exchange with relevant facilities. At the end of the project 27 papers in peer reviewed journals were published and it is expected that the real number will be the double. The training and integration success was evidenced by the fruitful implementation of the Travel Fund as well as the unique schools, e.g. practical and theoretical handling of plutonium.

International audience; Safe management of radioactive waste is challenging to waste producers and waste management organisations. Deployment of thermal treatment technologies can provide significant improvements: volume reduction, waste passivation, organics destruction, safety demonstration facilitation, etc. The EC-funded THERAMIN project enables an EU-wide strategic review and assessment of the value of thermal treatment technologies applicable to Low and Intermediate Level waste streams (ion exchange media, soft operational waste, sludges, organic waste, and liquids). THERAMIN compiles an EU-wide database of wastes, which could be treated by thermal technologies and documents available thermal technologies. Applicability and benefits of technologies to the identified waste streams will be evaluated through full-scale demonstration tests by project partners. Safety case implications will also be assessed through the study of the disposability of thermally treated waste products. This paper will communicate the strategic aims of the ongoing project and highlight some key findings and results achieved to date.

International audience; The H2020 Euratom European project INSIDER(Improved Nuclear Site Characterization for waste minimization in decommissioning under constrained environment) was launched in June 2017 for a duration of 4 years; itcurrently includes 17 partners from 10 European countries.The project is focused onto radiological characterization applied to waste-driven integrated approaches,including the sampling overall strategy and design. Itsobjectives are to improve the management of waste withmedium (MA) and high radioactivity (HA) levels comingfrom nuclear sites or facilities under D&D (Decommissioning and dismantling) and/or other constrained environments. The optimization criteria refer to operationaldecommissioning efficiency, safety and costs. The outcomeof the project will increase knowledge on the amounts andcharacteristics of radioactive waste resulting from D&Dand increase confidence on the sound definition ofsubsequent storage and disposal end points.

The deterministic simple least square (LS) approach is employed in the covariance analysis of the total neutron cross section (n,tot) calculated by a microscopic optical potential, CTOM, which is based on a fundamental theory − Dirac Brueckner Hartree Fock. The sensitivity to the CTOM parameters is firstly systematically calculated for 77 stable nuclei in the range 12C–208Pb within neutron energy 5–200 MeV. Then, an equivalent covariance of experimental data (EVexp) is constructed to describe the experimental data uncertainties and the systematic difference between experimental data and CTOM calculation. The variance and covariance of EVexp matrix are both evaluated via the Gaussian analysis to the ratios of measured (n,tot) cross sections and the CTOM calculations. In addition, a technique named “selection of effective points (SEP)” is suggested additionally to reduce the influence of the Peelle's Pertinent Puzzle problem in this work.

Sensitivity and uncertainty analysis in error propagation studies are carried out based on nuclear data uncertainty information available in the basic nuclear data libraries such as ENDF, JEFF, JENDL and others. The uncertainty files (covariance matrices) are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). The intent of this paper is to discuss the use of the RPCM in contrast to the use of a cross section covariance matrix (CSCM) representation. For so, the resonance evaluation for 48Ti in the resonance range from 10−5 eV to 400 keV is used. The RPCM is translated into two distinct CSCM by using a coarser and a finer group structures. The results are presented and discussions of the feasibility of using these covariance representations are depicted.

International audience; Nowadays, reliance on nuclear models to interpolate or extrapolate between experimental data points is very common, for nuclear data evaluation. It is also well known that the knowledge of nuclear reaction mechanisms is at best approximate, and that their modeling relies on many parameters which do not have a precise physical meaning outside of their specific implementations in nuclear model codes: they carry both specific physical information, and effective information that is related to the deficiencies of the model itself. Therefore, to improve the uncertainties associated with evaluated nuclear data, the models themselves must be refined so that their parameters can be rigorously derived from theory. Examples of such a process will be given for a wide sample of models like: detailed theory of compound nucleus decay through multiple nucleon or gamma emission, or refinements to the width fluctuation factor of the Hauser-Feshbach model. All these examples will illustrate the reduction in the effective components of nuclear model parameters, through the reduced dynamics of parameter adjustment needed to account for experimental data. The significant progress, recently achieved for the non-fission channels, also highlights the difficult path ahead to improve our quantitative understanding of fission in a similar way: by relying on microscopic theory.

In the last decade or so, estimating uncertainties associated with nuclear data has become an almost mandatory step in any new nuclear data evaluation. The mathematics needed to infer such estimates look deceptively simple, masking the hidden complexities due to imprecise and contradictory experimental data and natural limitations of simplified physics models. Through examples of evaluated covariance matrices for the soon-to-be-released U.S. ENDF/B-VIII.0 library, e.g., cross sections, spectrum, multiplicity, this paper discusses some uncertainty quantification methodologies in use today, their strengths, their pitfalls, and alternative approaches that have proved to be highly successful in other fields. The important issue of how to interpret and use the covariance matrices coming out of the evaluated nuclear data libraries is discussed.

Theoretical α-decay half-lives of some nuclei from ground state to ground state are calculated using different nuclear potential model including Coulomb proximity potential (CPPM), Royer proximity potential and Broglia and Winther 1991. The calculated values comparing with experimental data, it is observed that the CPPM model is in good agreement with the experimental data.

Both the availability and the quality of covariance data improved over the last years and many recent cross-section evaluations, such as JENDL-4.0, ENDF/B-VII.1, JEFF-3.3, etc. include new covariance data compilations. However, several gaps and inconsistencies still persist. Although most modern nuclear data evaluations are based on similar (or even same) sets of experimental data, and the agreement in the results obtained using different cross-sections is reasonably good, larger discrepancies were observed among the corresponding covariance data. This suggests that the differences in the covariance matrix evaluations reflect more the differences in the (mathematical) approaches used and possibly in the interpretations of the experimental data, rather than the different nuclear experimental data used. Furthermore, “tuning” and adjustments are often used in the process of nuclear data evaluations. In principle, if adjustments or “tunings” are used in the evaluation of cross-section then the covariance matrices should reflect the cross-correlations introduced in this process. However, the presently available cross-section covariance matrices include practically no cross-material correlation terms, although some evidence indicate that tuning is present. Experience in using covariance matrices of different origin (such as JEFF, JENDL, ENDF, TENDL, SCALE, etc.) in sensitivity and uncertainty analysis of vast list of cases ranging from fission to fusion and from criticality, kinetics and shielding to adjustment applications are presented. The status of the available covariance and future needs in the areas including secondary angular and energy distributions is addressed.

International audience; Delayed-neutrons are of great importance for reactor operations. Current abundances derive from either a measurement performed in 1957 by Keepin or by a summation calculation performed by Brady and England in 1989. In this work, a code has been written to compute a new set of delayed-neutron abundances as well as to estimate uncertainties and correlations through a Monte Carlo method and a Bayesian inference. An experiment will take place in the future to verify the validity of the calculated quantities.

MYRRHA is a multi-purpose research reactor able to operate in sub-critical and critical modes and currently in the design phase at SCK•CEN. The choice of LBE was driven by its chemical stability, low melting temperature, high boiling point, low chemical reactivity with water and air and a good neutronic performance. As a drawback, the neutron capture in 209Bi results in the production of 210Po, a highly radiotoxic alpha emitter with relatively short half-life (≈138 days). The 210Po production represents a major safety concern that has to be addressed for the reactor licensing. In this work we used the ALEPH-2 burnup code to accurately calculate the 210Po production in a MYRRHA operating cycle. The impact of using different nuclear data libraries was evaluated and the reliability of the results was determined by quantifying the uncertainty of the 210Po concentration. The uncertainty quantification was carried out sampling the currently available nuclear data covariance matrices with the SANDY code. Also, estimates of the sensitivity profiles were obtained with a linear regression approach. The activation yield of the 209Bi neutron capture reaction was assessed as the largest nuclear data source of uncertainty, however the lack of covariances for such data represent a capital drawback for the 210Po content prediction.

Studies on fission yields have a major impact on the characterization and the understanding of the fission process and are mandatory for reactor applications. Fission yield evaluation represents the synthesis of experimental and theoretical knowledge to perform the best estimation of mass, isotopic and isomeric yields. Today, the output of fission yield evaluation is available as a function of isotopic yields. Without the explicitness of evaluation covariance data, mass yield uncertainties are greater than those of isotopic yields. This is in contradiction with experimental knowledge where the abundance of mass yield measurements is dominant. These last years, different covariance matrices have been suggested but the experimental part of those are neglected. The collaboration between the LPSC Grenoble and the CEA Cadarache starts a new program in the field of the evaluation of fission products in addition to the current experimental program at Institut Laue-Langevin. The goal is to define a new methodology of evaluation based on statistical tests to define the different experimental sets in agreement, giving different solutions for different analysis choices. This study deals with the thermal neutron induced fission of 235U. The mix of data is non-unique and this topic will be discussed using the Shannon entropy criterion in the framework of the statistical methodology proposed.

This paper presents a Bayesian approach based on integral experiments to create correlations between different isotopes which do not appear with differential data. A simple Bayesian set of equations is presented with random nuclear data, similarly to the usual methods applied with differential data. As a consequence, updated nuclear data (cross sections, [see formula in PDF], fission neutron spectra and covariance matrices) are obtained, leading to better integral results. An example for 235U and 238U is proposed taking into account the Bigten criticality benchmark.

OPAL is a 20 MWth multi-purpose open-pool type Research Reactor located at Lucas Heights, Australia. It was designed, built and commissioned by INVAP between 2000 and 2006 and it has been operated by the Australia Nuclear Science and Technology Organization (ANSTO) showing a very good overall performance. On November 2016, OPAL reached 10 years of continuous operation, becoming one of the most reliable and available in its kind worldwide, with an unbeaten record of being fully operational 307 days a year. One of the enhanced safety features present in this state-of-art reactor is the availability of an independent, diverse and redundant Second Shutdown System (SSS), which consists in the drainage of the heavy water reflector contained in the Reflector Vessel. As far as high quality experimental data is available from reactor commissioning and operation stages and even from early component design validation stages, several models both regarding neutronic and thermo-hydraulic approaches have been developed during recent years using advanced calculations tools and the novel capabilities to couple them. These advanced models were developed in order to assess the capability of such codes to simulate and predict complex behaviours and develop highly detail analysis. In this framework, INVAP developed a three-dimensional CFD model that represents the detailed hydraulic behaviour of the Second Shutdown System for an actuation scenario, where the heavy water drainage 3D temporal profiles inside the Reflector Vessel can be obtained. This model was validated, comparing the computational results with experimental measurements performed in a real-size physical model built by INVAP during early OPAL design engineering stages. Furthermore, detailed 3D Serpent Monte Carlo models are also available, which have been already validated with experimental data from reactor commissioning and operating cycles. In the present work the neutronic and thermohydraulic models, available for OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

International audience; In the current paper, we investigate the application of the Equivalent Generalized Perturbation Theory (EGPT) to derive trends and associated covariances on the neutron capture cross section of one major fission product for both light water reactors and sodium-cooled fast reactors which is Rhodium-103. To do so, we have considered the ERMINE-V/ZONA1 & ZONA3 fast spectrum experiment and the MAESTRO thermal-spectrum experiment, where samples of these materials were oscillated in the MINERVE facility. In the paper, the theoretical formulation of EPGT is described and its derivation in the special case of the close loop oscillation technique where the reactivity worth is determined thanks to a power control system. A numerical benchmark is presented to assess the relevance of sensitivity coefficients provided by EGPT against direct perturbations where the microscopic cross sections are manually changed before calculating the adjoint and forward flux. The breakdown between direct and indirect contributions in the sensitivity analysis of the sample reactivity worth is presented and discussed, with the impact of using a calibration reference sample to normalize the measured reactivity worth. Finally, the assimilation of integral trends is done with the CONRAD code, using C/E comparisons between TRIPOLI4/JEFF3.2 calculations and experimental results and the sensitivity coefficients provided by the EGPT. Preliminary results of this study are showing that the JEFF3.2 evaluation of $^{103}$Rh gives satisfactory agreements in both thermal and fast spectrum experiments and that the combination of them can lead to a significant uncertainty reduction on the capture cross section, from $\pm$5% to $\pm$3% in the resolved resonance range (1 eV-10 keV) and from $\pm$8% to $\pm$5% in the unresolved resonance range (10 keV-1 MeV).

Nuclear data users’ requirements for uncertainty data started already in the seventies, when several fast reactor projects did use extensively “statistical data adjustments” to meet data improvement for core and shielding design. However, it was only ∼20–30 years later that a major effort started to produce scientifically based covariance data and in particular since ∼2005. Most work has been done since then with spectacular achievements and enhanced understanding both of the uncertainty evaluation process and of the data utilization in V&V. This paper summarizes some key developments and still open challenges.

We present the beam dynamics and electromagnetic studies of a 3 MeV, 325 MHz H− radio frequency quadrupole (RFQ) accelerator for the proposed Indian Spallation Neutron Source project. We have followed a design approach, where the emittance growth and the losses are minimized by keeping the tune depression ratio larger than 0.5. The transverse cross-section of RFQ is designed at a frequency lower than the operating frequency, so that the tuners have their nominal position inside the RFQ cavity. This has resulted in an improvement of the tuning range, and the efficiency of tuners to correct the field errors in the RFQ. The vane-tip modulations have been modelled in CST-MWS code, and its effect on the field flatness and the resonant frequency has been studied. The deterioration in the field flatness due to vane-tip modulations is reduced to an acceptable level with the help of tuners. Details of the error study and the higher order mode study along with mode stabilization technique are also described in the paper.

Deterministic uncertainty propagation methods are certainly powerful and time-sparing but their access to uncertainties related to the power map remains difficult due to a lack of numerical convergence. On the contrary, stochastic methods do not face such an issue and they enable a more rigorous access to uncertainty related to the PFNS. Our method combines an innovative transport calculation chain and a stochastic way of propagating uncertainties on nuclear data: first, our calculation scheme consists in the calculation of assembly self-shielded cross sections and a pin-by-pin flux calculation on the whole core. Validation was done and the required CPU time is suitable to allow numerous calculations. Then, we sample nuclear cross sections with consistent probability distribution functions with a correlated optimized Latin Hypercube Sampling. Finally, we deduce the power map uncertainties from the study of the output response functions. We performed our study on the system described in the framework of the OECD/NEA Expert Group in Uncertainty Analysis in Modelling. Results show the 238U inelastic scattering cross section, the 235U PFNS, the elastic scattering cross section of 1H and the 56Fe cross sections as major contributors to the total uncertainty on the power map: the power tilt between central and peripheral assemblies using COMAC-V2 covariance library amounts to 5.4% (1σ) (respectively 7.4% (1σ) using COMAC-V0).

In the present paper, firstly, we review our previous works on uncertainty quantification (UQ) of reactor physics parameters. This consists of (1) development of numerical tools based on the depletion perturbation theory (DPT), (2) linearity of reactor physics parameters to nuclear data, (3) UQ of decay heat and its reduction, and (4) correlation between decay heat and β-delayed neutrons emission. Secondly, we show results of extensive calculations about UQ on decay heat with several different numerical conditions by the DPT-based capability of a reactor physics code system CBZ.

Fuel assembly (FA) structure without fuel rods is called FA skeleton which is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the Chashma Nuclear power plant-1 FA skeleton at room temperature. The finite element (FE) analysis has been performed using ANSYS, in order to determine the elongation of the FA skeleton as well as the location of max. stress and stresses developed in axial direction under tensile load of 9800 N or 2 g being the FA handling or lifting load [Y. Zhang et al., Fuel Assembly Design Report, SNERDI, China, 1994]. The FE model of grids, guide thimbles with dash-pots and flow holes has been developed using Shell 181. It has been observed that FA skeleton elongation values obtained through FE analysis and experiment are comparable and show linear behaviors. Moreover, the values of stresses obtained at different locations of the guide thimbles are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Therefore, validation of the FE methodology is confirmed. The values of stresses are less than the design limit of the materials used for the grid and the guide thimble. Therefore, the structural integrity criterion of CHASNUPP-1 FA skeleton is fulfilled safely.

Growing the oxide scale on the zirconium cladding of fuel elements in pressured-water reactors (PWR) is caused by the current of oxygen anions off the waterside to the metal through the layer of zirconia and by the strictly equal inverse electronic current. This process periodically speeds up the corrosion of the zirconium cladding in the aqueous coolant due to the breakaway of the dense part of oxide scale when its thickness reaches 2 mkm. It is shown that the electronic resistivity of zirconia is not limiting the zirconium oxidation at working temperatures. For gaining this limitation, a metal of lesser valence than zirconium has to be added to this oxide scale up to 15%. Then, oxygen vacancies arise in the complex zirconia, increase its band-gap, and thus, sharply decrease the electronic conductivity and form the solid oxide electrolyte whose growth is inhibited in contact with water at working temperatures of PWR.

Recently, increasing demands on the reduction of fuel cycle costs have led to higher burnup fuel designs. According to the erbia-credit super high burnup fuel concept, developed by mixing low content of erbia to UO2 powder directly after reconversion process so that all fuel pins in a given fuel assembly are homogeneously doped, the present study aims to characterize, from a neutronic point of view, a 17 × 17 pressurized water reactor assembly enriched to 10.27 wt.% in 235 U with an erbia content of 1 at.% (i.e. 0.7 wt.%) by means of the deterministic neutronic code APOLLO2. For this purpose, a simplified thermal-hydraulic analysis was performed in order to evaluate the effects on fuel thermal conductivity of adding erbia to uranium oxide. The results obtained allow to conclude that an Er-doped assembly enriched to >5 wt.% in 235 U represents an advantageous solution for very long fuel cycles, and it is so suited for very high burnups.

The deep geological repository (DGR) is considered as the definitive management solution for high-level waste (HLW). Countries defined different DGR implementation schedules, depending on their national context and political choices. We raise the question of the economic grounds of such political decisions by providing an economic analysis of different DGR schedules. We investigate the optimal timing for DGR commissioning based on available Nuclear Energy Agency (NEA) data (2013). Two scenarios are considered: (1) rescheduling the deployment of a DGR with the same initial operational period, and (2) rescheduling the deployment of a DGR with a shorter operational period, i.e. initial closure date. Given the long timescales of such projects, we also take into account the discounting effect. The first finding is that it appears more economically favorable to extend the interim storage than to dispose of the HLW immediately. Countries which chose “immediate” disposal are willing to accept higher costs to quickly solve the problem. Another interesting result is that there is an optimal solution with respect to the length of DGR operational period and the waste flow for disposal. Based on data provided by the Organisation for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA), we find an optimal operating period of about 15 years with a flow of 2000 tHM/year.