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

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

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

Imperial Users onl

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

EThOS - Electronic Theses Online ServiceGBUnited Kingdo

Studies on Super-critical Water Reactors Fuel Assemblies using the Sub-channel Code COBRA-EN

In the Generation IV International Forum (GIF) programme, the Super-Critical Water Reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however, exist fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under super-critical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the super-critical region of water have also been incorporated in the code. As part of the efforts to assess the new code capabilities, both a hexagonal and a square fuel assembly have been analysed. In the latter case the use of the code in combination with the neutronic code MCNP has allowed investigating on the use of hydride fuel in super-critical fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (wall, wire wrap models).JRC.DDG.F.4-Safety of future nuclear reactor

Analysis of the MT-3 Clad Ballooning Reflood Test using the Multi-rod Coupled MATARE Code

If PWR fuel rods balloon in a co-located fashion following a large break LOCA, it is possible that an un-coolable region could be formed. The MT-3 test was one of the few large-scale experiments performed to investigate this. In this paper we report the analysis of this experiment using a novel combined thermal-hydraulic and structural-mechanics model, able to model multiple distinct rods in a coupled fashion. The sensitivity of the behaviour to (in particular) the channel droplet distribution makes a full ¿first principles¿ validation impractical, but good agreement is achieved between observed and predicted blockage fractions, and the observed incoherence of ballooning is reproduced well.JRC.DG.F.4-Safety of future nuclear reactor