Guidance Document for Integrated Safety Assessment Methodology (ISAM) - (GDI): EC JRC report prepared for GIF Risk and Safety Working Group

A key objective of the Generation IV (Gen IV) International Forum’s Risk and Safety Working Group (RSWG) is the development and the qualification of an integrated methodology that can be used to evaluate and document the safety of Gen IV nuclear systems. The presentation of the GIF’s Integrated Safety Assessment Methodology (ISAM) to different stakeholders, nuclear safety experts and the individual Gen IV System Steering Committees has produced a series of comments and suggestions expressing the explicit need for having a more detailed description/justification about the "integration" of the different ISAM tools, as well as the request for further practical guidelines for its application. This EC JRC report prepared as EURATOM contribution to the GIF Risk and Safety Working Group is a Guidance Document for ISAM (GDI) developed to answer to the comments and suggestions with a view to facilitate the use of the methodology and to provide the users with further help for the ISAM implementation. In first instance the GDI document addresses the demonstration of the consistency and the adequateness of ISAM for the safety related “design / assessment” process while defining the possible role and contribution of each ISAM tool versus the different plant design status (pre-conceptual, conceptual, final; i.e. the step-by-step application of ISAM). The consistency with the current requirements and recommendations applicable to future nuclear systems is also outlined. In addition, while taking into consideration the experience gained with application of ISAM to different innovative design solutions, the GDI also provides a set of examples with the intent to help the designers to develop their own applications.JRC.F.5-Nuclear Reactor Safety Assessmen

Adaptación y aplicación del código TRACE para el análisis de transitorios en diseños de reactores rápidos refrigerados por plomo

El Generation IV International Forum aglutina los diferentes esfuerzos internacionales en I+D para el desarrollo de una nueva generación de reactores nucleares. Entre las tecnologías identificadas destacan los reactores rápidos refrigerados por metales líquidos, tales como el sodio o el plomo, con gran potencial para cumplir con los ambiciosos objetivos marcados. La falta de experiencia operativa previa obliga al uso de herramientas capaces de simular el comportamiento de los sistemas basados en esta tecnología. El artículo expone las modificaciones implementadas en el código TRACE para incluir las tablas termodinámicas del plomo líquido extraídas de resultados experimentales. A continuación, explica el proceso seguido para el desarrollo de un modelo termohidráulico para el prototipo ALFRED y el análisis de una selección de transitorios representativos realizado en el marco de proyectos internacionales de investigación. El estudio demuestra la aplicabilidad del código TRACE para simular diseños de reactores rápidos refrigerados por plomo y expone los altos márgenes de seguridad con los que cuenta esta tecnología para acomodar los transitorios más severos identificados en su estudio de seguridad.Lázaro Chueca, A.; Ammirabile, L.; Martorell Alsina, SS. (2014). Adaptación y aplicación del código TRACE para el análisis de transitorios en diseños de reactores rápidos refrigerados por plomo. Sociedad Nuclear Española. http://hdl.handle.net/10251/71945

Radiomic Gradient in Peritumoural Tissue of Liver Metastases: A Biomarker for Clinical Practice? Analysing Density, Entropy, and Uniformity Variations with Distance from the Tumour

The radiomic analysis of the tissue surrounding colorectal liver metastases (CRLM) enhances the prediction accuracy of pathology data and survival. We explored the variation of the textural features in the peritumoural tissue as the distance from CRLM increases. We considered patients with hypodense CRLMs >10 mm and high-quality computed tomography (CT). In the portal phase, we segmented (1) the tumour, (2) a series of concentric rims at a progressively increasing distance from CRLM (from one to ten millimetres), and (3) a cylinder of normal parenchyma (Liver-VOI). Sixty-three CRLMs in 51 patients were analysed. Median peritumoural HU values were similar to Liver-VOI, except for the first millimetre around the CRLM. Entropy progressively decreased (from 3.11 of CRLM to 2.54 of Liver-VOI), while uniformity increased (from 0.135 to 0.199, p < 0.001). At 10 mm from CRLM, entropy was similar to the Liver-VOI in 62% of cases and uniformity in 46%. In small CRLMs (≤30 mm) and responders to chemotherapy, normalisation of entropy and uniformity values occurred in a higher proportion of cases and at a shorter distance. The radiomic analysis of the parenchyma surrounding CRLMs unveiled a wide halo of progressively decreasing entropy and increasing uniformity despite a normal radiological aspect. Underlying pathology data should be investigated

Review of Euratom projects on design, safety assessment, R&D and licensing for ESNII/Gen-IV fast neutron systems

Nine Euratom projects started since late 2011 in support of the infrastructure and R&D of the seven fast reactor systems are briefly presented in the paper in terms of key objectives, results and recommendations

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

ELSMOR – towards European Licensing of Small Modular Reactors: Methodology recommendations for light-water small modular reactors safety assessment

Decarbonization of energy production is key in today’s societies and nuclear energy holds an essential place in this prospect. Besides heavy-duty electricity production, other industrial and communal needs could be served by integrating novel nuclear energy production systems, among which are low-power nuclear devices, like small modular reactors (SMRs). The ELSMOR (towards European Licensing of Small Modular Reactors) European project addresses this topic as an answer to the Horizon 2020 Euratom NFRP-2018-3 call. The consortium includes 15 partners from eight European countries, involving research institutes, major European nuclear companies and technical support organizations. The 3.5-year project, launched in September 2019, investigates selected safety features of light-water (LW) SMRs with focus on licensing aspects. Providing a comprehensive compliance framework that regulators can adopt and operate, the licensing process of such SMRs could be optimized, helping their deployment. In this prospect, as a result of ELSMOR’s work, this article gives an overview of the specific issues that LW-SMRs may bring about in the different domains of nuclear safety, in terms of: Methodological standpoints: safety goals, safety requirements, safety principles (defence-in-depth implementation); Main safety functions of reactivity control, decay heat removal and confinement management; Severe accident management; Other safety issues particular to SMRs: use of shared systems; performing of multi-unit probabilistic safety assessment (PSA); refuelling, spent fuel management, transport and disposal management. In this article, adequate methodologies are developed to deal with these issues and to help assess the safety of LW-SMRs. This work gives a precious synthesis of the safety assessment issues of LW-SMRs and of the associated methodologies developed in the context of the ELSMOR European project. The removal of fossil fuels in energy production is very important in today’s societies and nuclear energy plays an essential role in this. Besides large-scale electricity production, other industrial and communal needs could be solved by using new nuclear energy production systems, among which are low-power nuclear devices, like small modular reactors (SMRs). The ELSMOR (towards European Licensing of Small Modular Reactors) European project looks at this topic as an answer to the Horizon 2020 Euratom NFRP-2018-3 initiative. This project includes 15 partners from eight European countries, involving research institutes, major European nuclear companies and technical support organizations. The 3.5-year project, started in September 2019, investigates selected safety features of light-water (LW) SMRs with a focus on the licensing aspects. Providing a comprehensive compliance framework that regulators can use and operate, the licensing process of such SMRs could be optimized, helping their deployment. With this prospect, this article gives an overview of the specific subjects that LW-SMRs may bring in the different areas of nuclear safety (in particular: safety goals, safety requirements, nuclear safety functions: reactivity control, decay heat removal and confinement management, etc..). In this article, methods are developed to deal with these new subjects and to help assess the safety of LW-SMRs. This work gives an overview of the safety assessment issues of LW-SMRs and of the associated methods developed in the context of the ELSMOR European project

Coupled mechanical-thermohydraulic multi-pin deformation analysis of a PWR loss of coolant accident

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Coupled mechanical-thermohydraulic multi-pin deformation analysis of a PWR loss of coolant accident

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