In recent years, the application of Best Estimate Plus Uncertainty methodologies in the frame of Severe Accidents (SAs) has gained significant momentum. Both researchers and regulatory bodies in the field recognize the importance of quantifying the uncertainties associated with SA codes results as well as sensitivity analysis’ relevance in understanding the variables driving the calculated uncertainty. In this framework, the current work aims to deliver a thorough overview of the historical evolution of the application of sensitivity analysis techniques within the SA domain over the past five decades, detailing their primary focus, geographical context, main techniques and key documents. Highlighting how sensitivity analysis evolved over the years, the paper underscores its critical role within nuclear safety assessments. This review offers both a detailed historical perspective and insights into future directions for research, emphasizing the need for a balance between computational efficiency and model accuracy, and suggesting the integration of machine learning techniques to enhance future analyses.

Along the Management and Uncertainty of Severe Accidents (MUSA) project, attention was paid to the unfolding of uncertainty analysis when dealing with severe accident (SA) scenarios. While the quantification of the uncertainty linked to SA simulations' results was the main focus of the project, some efforts were also addressed to the identification of the variables being the root of it. To this end, a complementary sensitivity analysis was deemed to be of high importance. Following this path, the present paper reports the advancements made in the attempt to enhance and optimize the sensitivity analysis process. More commonly used sensitivity analysis techniques, such as correlation coefficients or simple regression, are complemented by more advanced techniques through the integration of feature selection algorithms. As a further step, a testing phase is foreseen; in particular, the selected sensitivity methods are applied against a SA scenario, namely, an unmitigated station blackout in a pressurized water reactor. Outcomes according to the different techniques are reported and compared, with a certain level of agreement being shown. The analysis also highlighted the need to support the application of sensitivity methods with expert judgment to corroborate the physical consistency of the obtained results.

Multi-scale methodologies have been developed and applied successfully in the frame of nuclear fuel performance analyses, but the complexity of the tools involved hinders their extensive application. Gaps in modelling capabilities of specific input/outputs in particular limits code-to-code communication. In this work, we propose a multi-fidelity methodology to tackle this issue. The application presented here concerns the inclusion of a meso-scale module describing fission gas behaviour (SCIANTIX) in a fuel performance code (FRAPCON). A critical input parameter of the meso-scale module, the local hydro-static stress in the fuel, is not predicted by such fuel performance code, hence limiting this coupling. This gap is filled by using a second fuel performance code (TRANSURANUS) to construct a virtual dataset of local hydro-static stress values, on which an artificial neural network is trained and included in the FRAPCON/SCIANTIX coupled suite. This multi-fidelity methodology is demonstrated by simulating the Ris & oslash; AN3 irradiation experiment.

The recently completed Horizon-2020 project "Management and Uncertainties of Severe Accidents (MUSA)" has reviewed uncertainty sources and Uncertainty Quantification methodology for assessing Severe Accidents (SA), and has made a substantial effort at stimulating uncertainty applications in predicting the radiological Source Term of reactor and Spent Fuel Pool accident scenarios. The key motivation of the project has been to bring the advantages of the Best Estimate Plus Uncertainty approach to the field of Severe Accident modelling. With respect to deterministic analyses, expected gains are avoiding adopting conservative assumptions, identifying uncertainty bands of estimates, and gaining insights into dominating uncertain parameters. Also, the benefits for understanding and improving Accident Management were to be explored. The reactor applications brought together a large group of participants that set out to apply uncertainty analysis (UA) within their field of SA modelling expertise - in particular reactor types, but also SA code used (ASTEC, MELCOR, MAAP, RELAP/SCDAPSIM), uncertainty quantification tools used (DAKOTA, SUSA, URANIE, self-developed tools based on Python code), detailed accident scenarios, and in some cases SAM actions. The setting up of the analyses, challenges faced during that phase, and solutions explored, are described in Brumm et al. ANE 191 (2023). This paper synthesizes the reactor-application work at the end of the project. Analyses of 23 partners are presented in different categories, depending on whether their main goal is/are (i) uncertainty bands of simulation results; (ii) the understanding of dominating uncertainties in specific sub-models of the SA code; (iii) improving the understanding of specific accident scenarios, with or without the application of SAM actions; or, (iv) a demonstration of the tools used and developed, and of the capability to carry out an uncertainty analysis in the presence of the challenges faced. A cross-section of the partners' results is presented and briefly discussed, to provide an overview of the work done, and to encourage accessing and studying the project deliverables that are open to the public. Furthermore, the partners' experiences made during the project have been evaluated and are presented as good practice recommendations. The paper ends with conclusions on the level of readiness of UA in SA modelling, on the determination of governing uncertainties, and on the analysis of SAM actions.

This paper provides insights into the response of the advanced technology fuel (ATF) cladding FeCrAl during postulated design basis accident (DBA) and design extension condition without significant fuel degradation (DEC-A) scenarios. Such insights are gained through the development and application of in-house extensions of the FRAPCON/FRAPTRAN fuel performance codes, coupled with the statistical tool DAKOTA, within the framework of a loss-of-coolant accident (LOCA) safety evaluation methodology. While most of the specific FeCrAl models and correlations embedded in these extensions have been documented in the existing literature, the derivation of an instantaneous plasticity model describing the strain-hardening behaviour of FeCrAl alloy C26M is presented for the first time in this paper. The application of the methodology to the DEC-A/LOCA scenario suggests an improved performance of the advanced cladding material, as it maintains its integrity, in contrast to Zircaloy. However, in the DBA/LOCA scenario, no significant differences between these cladding materials were observed.

This study focuses on the release of the most relevant volatile fission products from the hot sodium pools. From an exhaustive review of the open literature, a critical review of a model based on the film theory and its estimates in comparison to data from the NALA experimental program has been conducted. Due to certain assumptions embedded in the approach, the fitting-to-data transport coefficients derived and some inconsistencies found between its formulation and the reported estimates, an alternate approach including other phenomena anticipated in the scenario is here proposed. Based on diffusive and convective mechanisms in the gas phase, the assumption of the analogy between heat and mass transport (HMT) and the ideal-dilute solutions laws to set the fission products concentration at the Na pool interface, a good agreement has been found with experimental data, which mean a substantial enhancement of qualitative and quantitative predictability while maintaining a conservative nature.

Severe Accident (SA) research is currently facing new challenges coming from changes in the energy and computer science sectors. Therefore, it is imperative to reassess the current status in the area to optimize where the research resources should go to reach even higher safety standards both in the running Nuclear Power Plants (NPP) and the upcoming new designs, particularly Water-Cooled Small Modular Reactors (WC-SMR). Three Horizon Euratom projects stand out in such context. SEAKNOT (SEvere Accident research and KNOwledge managemenT) is progressing in setting a SA research roadmap by ranking the major phenomena involved in terms of knowledge and safety significance in LWRs (large water-cooled reactors) and SMRs, at the same time that it is strengthening paths for Education & Training (E&T) on SA for forthcoming generations of researchers and engineers. SASPAM-SA (Safety Analysis of SMR with PAssive Mitigation strategies-Severe Accident) is supplying valuable information on phenomena, boundary and accident conditions that might prevail in WC-SMRs, specifically integral PWR (iPWR). The project allows the assessment of the applicability of the current state-of-the-art simulation codes and the relevance of large reactor experiments to iPWRs. Different SA mitigation strategies, like In-Vessel Melt Retention (IVMR), are being explored. Finally, ASSAS (Artificial intelligence for Simulation of Severe AccidentS) is working to prove the possibility to develop fast-running SA simulators thanks to Artificial Intelligence, to support training, engineering and emergency response. This paper discusses the major progress made in the three projects and their complementarity contributes to a safer nuclear energy production.

Chromium coatings are being developed for advanced technology fuel (ATF) claddings, offering negligible corrosion during normal operation, improved resistance to high-temperature steam oxidation, and superior high-temperature strength, the latter two being of utmost relevance during design basis accidents (DBAs). Demonstrating the improved response of Cr-coated Zircaloy requires the development or extension of fuel performance codes to coating simulations.In this work, material models and correlations for Cr-coated Zircaloy cladding have been derived or obtained from the literature and implemented into TRANSURANUS and the FRAPTRAN-TUmech suite. These extended tools have been used to simulate two complementary LOCA tests: QUENCH-L1 rod 4 (out-of-pile bundle test on fresh Zircaloy cladding) and IFA-650.10 (in-pile single rod test on high-burnup Zircaloy-UO2 fuel), enabling a gradual cross-verification of results between codes and a comparative performance analysis between coated and uncoated cladding.The results indicate negligible impact of coating properties other than creep on the burst time. While the superior high-temperature creep resistance of coated cladding slightly delays the burst time, additional burst data would be necessary to draw sound conclusions on the balloon size. Regarding the modelling approach, treating the coated cladding as a composite material through the definition of effective properties might result in worse performance relative to uncoated cladding, contradicting experimental observations. Therefore, the separate modelling of the coating and the cladding is recommended.

The assessment of cladding degradation mechanisms under anticipated DBA LOCA conditions requires analytical tools capable of properly modelling the processes involved in clad deformation and burst. The present work proves that the deviations found in FRAPTRAN simulations of burst and LOCA tests, mostly concerning time to failure, come to a good extent from the high temperature creep law, whereas failure limits barely play any role in the deviations. According to an analysis conducted with an alternative model, one might conclude that a Norton-type formulation, once extended to irradiated conditions, might substantially enhance FRAPTRAN accuracy. Besides, this work shows how the formulation adopted to estimate plastic strains up to the instability strain also plays an important role in the time-to-failure prediction.

The Management and Uncertainties of Severe Accidents (MUSA) project was funded in HORIZON 2020 and is coordinated by CIEMAT (Spain). The project aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) analysis, focusing on source term figures of merit. The Application of Uncertainty Quantification (UQ) Methods against Integral Experiments (AUQMIE - Work Package 4 (WP4)), led by ENEA (Italy), was devoted to apply and test UQ methodologies adopting the internationally recognized PHEBUS FPT1 test. FPT1 was chosen to test UQ methodologies because, even though it is a simplified SA scenario, it was representative of the in-vessel phase of a severe accident initiated by a break in the cold leg of a PWR primary circuit. WP4 served as a platform to identify and discuss the issues encountered in the application of UQ methodol-ogies to SA analyses (e.g. discuss the UQ methodology, perform the coupling between the SA codes and the UQ tools, define the results post-processing methods, etc.). The purpose of this paper is to describe the MUSA PHEBUS FPT1 uncertainty application exercise with the related specifications and the methodologies used by the partners to perform the UQ exercise. The main outcomes and lessons learned of the analysis are: scripting was in general needed for the SA code and uncertainty tool coupling and to have more flexibility; particular attention should be devoted to the proper choice of the input uncertain parameters; outlier values of figures of merit should be carefully analyzed; the computational time is a key element to perform UQ in SA; the large number of uncertain input parameters may complicate the interpretation of correlation or sensitivity analysis; there is the need for a statistically solid handling of failed calculations.

The safe storage of Spent Nuclear Fuel (SNF) within a dry cask system must be guaranteed for any operating condition. In particular, the Peak Cladding Temperature (PCT) is currently employed as a measure of the fuel integrity in normal conditions. The same fuel integrity has to be ensured also in off-normal and accidental conditions by keeping the PCT below the regulatory limits. In this regard, the work proposed in this paper aims at evaluating the performance of a concrete-based dry cask in off-normal and accidental conditions. More specifically, the influence of a partial to complete blockage of the air inlet openings on the PCT is assessed by means of a MELCOR model of the HI-STORM 100S cask. Attention is also paid to the temperature of the external concrete cask in order to check the integrity of the concrete itself.

Since 2012, many NEA member countries have implemented deterministic safety analyses for operating nuclear power plants under design extension conditions without significant fuel degradation or core melt (DEC-A). However, variations persist among these countries in defining DEC-A scenarios and acceptance criteria, validation and application of computer codes, development and application of deterministic safety analysis methods. Furthermore, there is a dearth of shared international experience and methodologies among various stakeholders, including regulatory authorities, technical safety or support organizations, utilities, engineering and consulting companies. To address these gaps, the OECD/NEA initiated a project in 2021, titled "Good Practices for Analyses of Design Extension Conditions without Significant Fuel Degradation for Operating Nuclear Power Plants" (or "DEC-A"), under the auspices of the Working Group on Accident Management and Analysis (WGAMA) and the Working Group on Fuel Safety (WGFS). The DEC-A project aims to review and summarize the current requirements, knowledge status, and best practices in NEA member countries. This paper outlines the objectives and scope of the OECD/NEA DEC-A project, and presents the findings from the review and discussions for each task.

Mechanistic multi-scale modelling holds the potential to inform fuel performance codes by incorporating high-fidelity models, algorithms, parameters, and material properties. In this context, meso-scale codes emerge as valuable tools for developing detailed models and performing separate verification and validation steps. This work focuses on SCIANTIX, an open-source 0D meso-scale code designed to describe the behaviour of gaseous and volatile fission products in nuclear oxide fuel. The code predominantly employs engineering physics-based behavioural models featuring computational times that align with typical fuel performance code requirements. Given the numerical foundation of the code, it is applicable to both stationary and transient conditions. Following a recent work outlining the standalone SCIANTIX (version 2.0) performance and its separate-effect validation database, we present its performance when coupled with fuel performance codes to simulate light water reactor fuel rods. The experiments selected for the comparative analysis constitute an initial integral validation database. The comparison focuses on conventional engineering quantities of interest, such as integral fission gas release, demonstrating the satisfactory performance of the code. Additionally, it highlights the potential advantages of multi-scale modelling over conventional semi-empirical approaches.

The analysis of the cladding mechanical performance is essential for predicting rod failure during the heat-up phase of a loss-of-coolant accident (LOCA). Previous modelling and simulation exercises have shown that there is room for improvement in the accuracy of fuel performance simulations under such conditions. In this work, TUmech, a simplified standalone version of the mechanical model in TRANSURANUS, has been developed and coupled with FRAPTRAN-2.0. After highlighting the most significant differences between both approaches, particularly in their mechanical treatment and material property correlations, simulations have been performed for IFA-650.2 (fresh fuel) and IFA-650.10 (high burnup fuel) of the Halden LOCA test series. The results obtained from both FRAPTRAN versions have been compared with experimental data and benchmarked against TRANSURANUS estimates, focusing on rod inner gas pressure, post-test cladding outer diameter, and burst time. In both scenarios, FRAPTRAN-TUmech has proven to enhance the accuracy in predicting these three variables compared to the FRAPTRAN default version. However, further validation is required before drawing definitive conclusions based on the encouraging results.

In the framework of the European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA), application of Best Estimate advanced codes for evaluating the radiological release under Design Basis Accident and Design Extended Conditions was considered. The aim is to assess the current modelling capabilities and to propose new calculation methodologies in order to produce more realistic evaluations of radioactive releases resulting from such accidents.In this paper, the studies are conducted for a reference Design Basis Accident Steam Generator Tube Rupture scenario in a 3-loop PWR-1000 reactor in terms of key thermal-hydraulic and radiological variables that condition and characterize the fission product release to the environment. The assessment includes a comparative study of the results performed by three organisations using three different computational tools and approaches. The outcomes highlighted issues considered for the design basis evaluation, modelling differences as well as some challenges and limitations in carrying out such analysis.

As a contribution to the transmission of knowledge on severe accidents, a methodology for developing a Phenomena Identification and Ranking Table (PIRT) for severe accident phenomena in light water nuclear reactors is proposed. The methodology takes into account the current state of knowledge as well as the importance for nuclear safety. The developed PIRT should identify the priorities for further research directions.

The European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA) was launched in September 2019, with a very broad participation: 11 countries, 17 participating organisations, including international organisations, utilities, regulators, technical support organisations, researchers and developers and was coordinated by IRSN. The main goal of the project was to assess the conservatisms in the radiological releases calculations in nuclear power plant (NPP) studies. The work here presented is focused on developing new calculation methodologies and updating computer code models to carry out more detailed assessments of source terms resulting from a steam generator tube rupture (SGTR) accident in the Design Extension conditions A (DEC-A) domain. For this purpose, participants in Work Package 2 (WP2) of R2CA developed and implemented simplified models in the Severe Accident (SA) and Thermal-Hydraulic (TH) codes that take into account iodine spiking and iodine transport phenomena within primary and secondary circuit. These methodologies will not only provide a better estimation of the radiological consequences, but should also serve to improve accident management procedures, innovative instrumentation development and early detection tools. The new approaches are tested in a SGTR + Steam Line Break Outside Containment scenario without significant fuel degradation in a three-loop Western 1000 MWe PWR. During the transient, operators are also assumed to implement emergency operating procedures (EOPs) in line with Westinghouse EOPs to limit the release of radioactivity and control the evolution of plant parameters. The calculations are performed by Tractebel and CIEMAT with the American SA tool MELCOR and Bel-V with the French TH code CATHARE. The results highlighted some limitations of implemented models in predicting iodine behaviour as well as small discrepancies in the TH evolution of the transient, both of which were analysed and discussed in detail.