Severe accident (SA) codes and their core degradation models have to deal with strongly nonlinear and discontinuous phenomena. In the application of uncertainty quantification to SA simulations, the combination of such phenomena may lead to a strong increase in the uncertainty propagated through the simulation, as well as to the chaotic behavior of the output variables. In this framework, the application of the limit surface search method of the RAVEN tool is proposed for a case where cliff-edge effects of SA phenomena determine a bifurcation of an output figure of merit. The algorithm is based on a predictive method making use of a support vector machine model, and it is applied with the aim of separating those input values that lead to different phenomenologies among the uncertainty calculations. The case study is in regard to the uncertainty analysis of the ASTEC code simulation of the QUENCH6 experimental test conducted in the framework of the International Atomic Energy Agency Coordinated Research Project I31033.

An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170 GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, X-divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

In the framework of the activities coordinated by the EUROfusion consortium, the Water thermal-HYDRAulic (W-HYDRA) experimental platform is being built at the ENEA Brasimone Research Centre in order to support the development of the Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB). In particular, this infrastructure will make possible the installation and testing of prototypical mock-ups under relevant working conditions, such as the First Wall (FW), the manifold and the Steam Generator (SG). Moreover, it will represent an integral test facility for the investigation of phenomena characteristic of WCLL BB concept, such as the PbLi/water interaction. Finally, the collection of data coming from the different planned experimental campaigns will allow to qualify and validate numerical models and codes currently adopted for the design of components, as well as for the modelling of complex phenomena typical of the WCLL BB. In order to come to a definitive design of the different facilities constituting the experimental platform, several design analyses assessing the thermal, hydraulic and structural performances of the different facilities and components are necessary. The paper reports a highlight of the W-HYDRA platform with a general description of the facilities. Some of the most relevant design studies carried out so far are reported as well, highlighting their impact on the evolution of the design.

As the horizon of nuclear energy expands with the advent of small modular reactors, Generation IV reactors, and fusion reactors, there is a growing perspective that the licensing process could benefit from a more comprehensive approach. Moving beyond traditional deterministic and probabilistic risk assessment analyses might pave the way for a novel safety analysis paradigm propelled by the increasing computational power at our disposal. This paper explores different methodologies that can improve the outcomes of nuclear safety analysis. These range from uncertainty quantification techniques, aimed at enhancing the precision of safety margins, to deploying dynamic event trees by driving system code simulations, capturing the potential evolutions of severe accidents. These methodologies offer a better understanding of the management and consequences of nuclear accident scenarios, significantly improving the accuracy and efficiency of safety predictions compared to traditional methods. Specific case studies illustrate the practical application of these advanced techniques, demonstrating substantial improvements in predicting and managing the dynamics of severe accidents. These findings underscore the effectiveness of these methodologies in enhancing risk assessment capabilities and informing decision-making processes for nuclear safety management. The paper also emphasizes the importance of adaptability and continuous evolution, a call for action to address emerging nuclear safety concerns and highlights the utility of advanced tools like RAVEN.

One of the two most promising Breeding Blanket (BB) concepts to be chosen as driver design of the EU-DEMO fusion reactor is the Water-Cooled Lead Lithium (WCLL) BB. A crucial point of this key component is the cooling of the BB structural elements and of the Lithium-Lead alloy, used as neutronic multiplier and Tritium breeder. Indeed, the high neutronic flux needed for breeding implies high volumetric power affecting the Breeding Zone (BZ) materials. The current BB layout employs water-cooled Double Wall Tubes (DWTs) with a Cshaped configuration for the BZ cooling. Despite the WCLL BB have reached a mature design in the last years, some open issues remain to be solved to increase the reliability and performance of this technology. The present paper describes a promising upgrade of the BZ cooling layout adopting helical-shaped DWTs. This solution has the potential to increase the BB reliability, cooling performance, Tritium Breeding Ratio and cooling water flowpath simplicity. The new proposed layout is described in the paper, along with results of CFD analyses carried out to investigate the new cooling performances in BZ. Advantages and drawbacks with respect to the current BB layout have been also highlighted in the paper.

The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB's overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed.

Integral Severe Accident (SA) codes are aimed at providing an exhaustive coverage of all the main phenomena taking place in a core melt accident. Today, these deterministic codes have reached a high level of maturity for the simulation of operating reactors and the nuclear technical community is starting to extend their applicability to advanced reactor designs, as Small Modular Reactor (SMR). In the framework of the NUGENIA TA-2 ASCOM (ASTEC COMmunity) collaborative project, a generic input-deck based on the IRIS design has been developed for the ASTEC code. The generic SMR ASTEC model has been already proved able to simulate the main thermal–hydraulic phenomena driving the passive mitigation of a SBLOCA in Design Basis Accident (DBA) conditions. The same initiator event, regarding the guillotine break of a Direct Vessel Injection (DVI) line, will be assumed for the simulation of beyond-design scenarios by considering the unavailability of selected passive safety systems. The results of the ASTEC simulations of four Beyond Design Basis Accidents (BDBAs) (study carried out with ASTEC V2.2, IRSN all rights reserved, [2021]) will be analyzed and discussed against the reference DBA sequence in the present paper. This study is aimed at proving the first insights about the capability of the ASTEC model of a generic IRIS reactor to be used in BDBA and in SA analyses, if significant core degradation takes place. In addition, it characterizes the role played by each safety system in SMR passive mitigation strategy and give the possibility to characterize the phenomenologies specific of SMR designs.

A large research effort is currently ongoing within the framework of the EUROfusion consortium for the study and design of a water-cooled lithium–lead (WCLL) breeding blanket (BB). This concept will be tested in ITER through the installation of a test blanket module (TBM) and it is one of the two candidates adopted as driver BBs in DEMO. In this framework, at the ENEA research centre of Brasimone, the realization of the experimental platform, W-HYDRA, is envisaged. The platform is dedicated to the support of the development of WCLL BB and ITER TBM and the investigation of the DEMO balance of plants. One of the most important experimental infrastructures is the water-loop facility, the aim of which is to provide water at a high pressure and temperature (PWR conditions), with a sufficient mass-flow rate and power for the experimental testing of BB and TBM components. The facility will be equipped with a vacuum chamber and an electron beam gun for the reproduction of high surface heat flux on plasma-facing components. In the present work, the design of a prototypical mock-up (MU) of the WCLL BB first wall is described. The MU is used to investigate the thermal, hydraulic and structural behavior of the current first-wall design under relevant heat loads at the expected operational conditions. The delineation of the main experimental test’s features and the instrumentation needed is assessed in the paper. A preliminary CFD calculation on the prototypical MU and the computational results are also presented.

In the last decades, taking into account the operational experience of fission nuclear reactors, the nuclear international technical community started the development of advanced reactor designs in order to satisfy the demand of the people to improve the safety of NPPs and to take into consideration the industry needs to improve the economic efficiency and reduce the capital costs of nuclear power technology. In this framework Small Modular Reactors (SMR) adopting light water as coolant, taking advantage of the experience developed in current larger scale LWR, can bring advantages in terms of increasing “inherent safety” due to the integral type configuration, lower nominal power, and the adoption of passive mitigation strategy. Starting from some of the research activities on code applications developed by the authors along the last decade, and available in the public scientific literature, the target of this paper is to give some insights and recommendations for the future development of new research activities, in national and international frameworks, in relation to thermal hydraulics of SMRs.

The Management and Uncertainties of Severe Accidents (MUSA) project, funded in HORIZON 2020 and coor-dinated by CIEMAT (Spain), aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) focusing on Source Term (ST). In this framework, the ob-jectives of the Innovative Management of Spent Fuel Pool Accidents (IMSFP - WP6), led by IRSN (France), are to quantify and rank the uncertainties affecting accident analyses in a Spent Fuel Pool (SFP), to review existing and contemplated SA management measures and systems and to assess their possible benefits in terms of reduction of radiological consequences.To quantify the propagation of the uncertainties of the input parameters to the output uncertainties of severe accident codes (ASTEC, MELCOR, RELAP/SCDAP), a diverse set of uncertainty quantification (UQ) tools (DAKOTA, RAVEN, SUNSET, SUSA) are used. The statistical framework used by the different UQ-tools is similar e.g. pure random (Monte Carlo) and Latin hypercube sampling (LHS).Fourteen partners from three different world regions are involved in the WP6 activities. The target of this paper is to describe the achievements during the first three years of the project. In a first part, a description is given of the SFP accidental scenario, of the key target variables and radionuclides chosen as ST Figures of Merit (FoM) and of the identified uncertainty sources in models and input parameters. A key element when defining the SFP scenario has been the consideration (or not) of the reactor building, as it is expected to significantly affect analyses. In a second part, the first insights coming out from the calculation phase of the project are presented. The review of existing SA management measures is also exposed, as well as systems whose benefits will be assessed in the second phase of the project. Finally, challenges that arise from such an exercise are discussed, as well as major difficulties found when applying UQ methodologies to SFP scenarios and solutions adopted.

Severe Accident (SA) integral codes, such as the Accident Source Term Evaluation Code (ASTEC) developed by IRSN, are used to simulate the phenomena occurring during accident progression in Nuclear Power Plants (NPPs) up to the source term evaluation. Code validation against experimental data is fundamental to carry out deterministic safety analysis and apply these codes to NPPs. In addition, in the Best Estimate Plus Uncertainty (BEPU) framework, the quantification of the results uncertainty is needed. In the framework of the IAEA CRP I31033 "Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors", the QUENCH test-6 experiment, conducted at KIT, has been selected to develop an uncertainty analysis using the ASTEC v2.2b code. The accuracy of the best-estimate ASTEC simulation was evaluated with the Fast Fourier Transform Based Method (FFTBM) against the experimental data. Then, the uncertainty of the code results was quantified by using the probabilistic propagation of input uncertainties method, through the coupling of ASTEC with RAVEN (Risk Analysis and Virtual Environment). Beyond identifying the main sources of uncertainty affecting the simulated test, the outcomes of the work also include some general discussion on the uncertainty propagation in a SA sequence.

The analyses of postulated severe accident sequences play a key role for the international nuclear technical scientific community for the study of the effect of possible actions to prevent significant core degradation and mitigate source term release. To simulate the complexity of phenomena involved in a severe accident, computational tools, known as severe accident codes, have been developed in the last decades. In the framework of NUGENIA TA-2 ASCOM project, the analysis of an unmitigated 2-inch cold leg LOCA transient, occurring in a generic western three-loops PWR-900 MWe, has been carried out with the aim to give some insights on the modelling capabilities of these tools and to characterize the differences in the calculations results. The ASTEC V2.2b code (study carried out with ASTEC V2, IRSN all rights reserved, [2021]), and MELCOR 2.2 code have been used in this code-to-code benchmark exercise. In the postulated transient, the unavailability of all active injection coolant systems has been considered and only the injection of accumulators has been assumed as accident mitigation strategy.

This paper deals with a general methodology to evaluate the Source Term (ST) and the Radiological Consequences (RC) of a hypothetical Severe Accident (SA) at a Fukushima-like Spent Fuel Pool (SFP) by coupling ASTEC 2.1 and RASCAL 4.3 SA and consequence projections (CP) codes, respectively. The methodology consists of the following sequential steps: the ST provided by a prior simulation performed by ASTEC V2.1 code was used as input to RASCAL 4.3 code to make a RC analysis. This approach was developed as a preparatory study for the Management and Uncertainties in Severe Accident (MUSA) H2020 European Project, coordinated by CIEMAT, where the ENEA's Nuclear Installations safety laboratory is committed to performing an analysis on a Fukushima-like SFP with the aim to apply innovative management of SFP accidents (WP6) to mitigate the RC of the accident itself. To perform the RC studies that could have an impact on Italy, a Fukushima-like SFP was assumed located in one of the Italian cross-border NPP sites. The weather data adopted are both standard and real hourly meteorological data taken from more than one geographical location. The results of the RC for 96 h of ST release in a range of 160 km from the emission point are reported in terms of Total Effective Dose Equivalent (TEDE), Thyroid dose, and Cs-137 total ground deposition. The mitigating effect on ST and on RC of the cooling spray system (CSS) actuated with several pH values (i.e., 4,7,10) was also investigated.

After the Fukushima accident, the interest of the public to nuclear safety has growth and the international technical nuclear community has increased his attention in the investigation and the characterization of Severe Accident (SA) scenarios. In order to simulate the different, complex and multi-physical phenomena involved in a SA, computational tools, known as SA codes, have been developed in the last decades. In order to give some insights on the modelling capabilities of these tools and the differences in the calculation results, also related to the user-effect, an analysis of an unmitigated Station Black Out (SBO) occurring in a generic Western three-loops PWR 900 MWe has been carried out by the authors in the framework of the NUGENIA TA-2 ASCOM project. The simulation results of ASTEC code (study carried out with ASTEC V2, IRSN all rights reserved, [2019]), developed by IRSN, and MELCOR 2.2 code, developed by SANDIA for USNRC, have been compared and analyzed. The SBO scenario considered takes into account the intervention of the accumulators as only accident mitigation strategy. Several figures of merits related to the thermal-hydraulic (e.g. primary pressure, cladding temperature, etc.) and to the core degradation (e.g. hydrogen production, etc.) have been considered to describe the accident evolution until the vessel failure, for the two codes comparison.

Advanced small LWRs, are considered as one of the key design options for the development of a safer nuclear technology considering their inherent safety due to the adoption of passive mitigation strategies and lower nominal power. In the framework of deterministic safety analyses for Small Modular Reactors (SMRs), a model of a generic IRIS SMR was developed by using the severe accident code ASTEC (Study carried out with ASTEC V2, IRSN all rights reserved, [2019]). The ASTEC code thermal-hydraulics modules have been used for the reactor modelling, and the nodalization approach has been described in the present work. The objective of the paper is to analyze the code capability and the consequent applicability to model an integral-type reactor and to simulate the complex thermal-hydraulic phenomena occurring in a passive mitigation strategy. The analysis is based on 2-inches guillotine Direct Vessel Injection (DVI) line break transients. (C) 2021 Elsevier Ltd. All rights reserved.

The integrated ICE (Ingress-of-Coolant Event) facility, scaled 1/1600 with respect to the ITER-FEAT design, was built at JAERI with the aim of reproducing the phenomenology occurring in an ICE accident. An ICE occurs when a rupture in the coolant pipes causes the pressurized coolant to enter into the Plasma Chamber, which is held under high vacuum condition. A suppression system is used to mitigate the overpressurization and to prevent mechanical damages to the structures. The CPA module of the ASTEC severe accident code (Study carried out with ASTEC V2, IRSN all rights reserved, [2020]), has been adopted for the modelling and the simulation of a test conducted in the ICE facility. The experimental results of the main thermal-hydraulic parameters have been compared to the code results to characterize the ASTEC capability to predict the phenomenology of a low-pressure two-phase flow transient occurring in a fusion reactor. By coupling the ASTEC code with the uncertainty tool RAVEN, developed by Idaho National Laboratory, an uncertainty analysis has been conducted on the transient. The aim of the present activity is to investigate the dispersion and the sensitivity of the code response to the variation of selected uncertain input parameters, which could influence the simulation of an ICE. The activity also provides a first application of uncertainty analysis through the RAVEN-ASTEC coupling.

Among the Postulated Initiating Events in nuclear fusion plants, the Ingress of Coolant Event (ICE) in the Plasma Chamber is one of the main safety issues. In the present paper, the best estimate thermal-hydraulic system code TRACE, developed by USNRC, has been adopted to study the ICE, and it has been qualified based on experimental results obtained in the Integrated ICE facility at JAERI. A nodalization has been developed in the SNAP environment/architecture, using also the TRACE 3D Vessel component where multidimensional phenomena could occur. The accuracy of the code calculation has been assessed both from a qualitative and quantitative point of view. In addition, an Uncertainty Analysis (UA), with the probabilistic method to propagate the input uncertainties, has been performed to characterize the dispersion of the results. The analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in the SNAP environment/architecture. Results show the adequacy of the 3D nodalization and the capability of the code to follow the transient evolution also at a very low pressure. Response correlations have been computed to characterize the correlation between the selected uncertain input parameters and the Plasma Chamber pressure.

This paper deals with a general methodology to evaluate the Source Term (ST) and the Radiological Consequences (RC) of a Severe Accident (SA) at a Fukushima-like Spent Fuel Pool (SFP) by coupling ASTEC 2.1 and RASCAL 4.3 codes. Essentially, the ST provided by ASTEC is used as input to RASCAL to perform a RC analysis. This methodology was developed as a preparatory study for the Management and Uncertainties in Severe Accident (MUSA) H2020 European Project, coordinated by CIEMAT. Within MUSA project, the laboratory for the Safety of Nuclear Installation of ENEA is involved in the Innovative Management of SFP Accidents Work Package (WP6), coordinates by IRSN. Within WP6, ENEA is committed to perform an analysis on a Fukushima-like SFP with the aim to apply innovative measures on the SFP Severe Accident Management to mitigate the RC of the accident itself. In order to perform the RC studies, the Fukushima-like SFP has been assumed located in one of the Italian cross-border NPP sites. The weather data connected with the radionuclides transport in atmosphere phase are both standard and real hourly meteorological data. The results of the RC for 96 hours of ST release from the SFP in a range of 160 km from the emission point are reported in terms of Total Effective Dose Equivalent (TEDE), I-131 thyroid dose and Cs-137 total ground deposition. The mitigating effect on ST and on RC of the cooling spray system (CSS) actuated with several pH values (i.e., 4,7,10) was also investigated.