LFR safety approach and main ELFR safety analysis results

This paper summarizes the approach to safety for the LFR systems, developed on the basis of the recommendations of the Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) and taking into account the fundamental safety objectives and the Defence-in-Depth approach, as described by IAEA Safety Guides, as well as the Safety quantitative objectives reported in the European Utilities Requirements (EUR). LEADER project activities are focused on the resolution of the key issues as they emerged from the 6th FP ELSY project attempting to reach a new industrial size European Lead-cooled Fast Reactor (ELFR) configuration. Apart from the safety approach, the main results of the ELFR safety transient analysis, where the most important design basis condition (DBC) and design extension condition (DEC) transient initiators were re-analyzed using the system codes RELAP5 (ENEA), TRACE-FRED (PSI), SIM-LFR (KIT) and SIMMER (CIRTEN), are summarized

Thermal-hydraulics of internally heated molten salts and application to the Molten Salt Fast Reactor

The Molten Salt Reactors (MSR) are an innovative kind of nuclear reactors and are presently considered in the framework of the Generation IV International Forum (GIF-IV) for their promising performances in terms of low resource utilization, waste minimization and enhanced safety. A unique feature of MSRs is that molten fluoride salts play the distinctive role of both fuel (heat source) and coolant. The presence of an internal heat generation perturbs the temperature field and consequences are to be expected on the heat transfer characteristics of the molten salts. In this paper, the problem of heat transfer for internally heated fluids in a straight circular channel is first faced on a theoretical ground. The effect of internal heat generation is demonstrated to be described by a corrective factor applied to traditional correlations for the Nusselt number. It is shown that the corrective factor can be fully characterized by making explicit the dependency on Reynolds and Prandtl numbers. On this basis, a preliminary correlation is proposed for the case of molten fluoride salts by interpolating the results provided by an analytic approach previously developed at the Politecnico di Milano. The experimental facility and the related measuring procedure for testing the proposed correlation are then presented. Finally, the developed correlation is used to carry out a parametric investigation on the effect of internal heat generation on the main out-of-core components of the Molten Salt Fast Reactor (MSFR), the reference circulating-fuel MSR design in the GIF-IV. The volumetric power determines higher temperatures at the channel wall, but the effect is significant only in case of large diameters and/or low velocities

Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like Sodium Fast Reactor

The EU 7th Framework ESNII+ project was launched in 2013 with the strategic orientation of preparing ESNII for Horizon 2020. ESNII stands for the European Industrial Initiative on Nuclear Energy, created by the European Commission in 2010 to promote the development of a new generation of nuclear systems in order to provide a sustainable solution to cope with Europe’s growing energy needs while meeting the greenhouse gas emissions reduction target. The designs selected by the ESNII+ project are technological demonstrators of Generation-IV systems. The prototype for the sodium cooled fast reactor technology is ASTRID (standing for Advanced Sodium Technological Reactor for Industrial Demonstration), which building phase is foreseen to be initiated in 2019. The ASTRID core has a peculiar design which was created in order to tackle the main neutronic challenge of sodium cooled fast reactors: the inherent overall positive reactivity feedback in case of sodium boiling occurring in the core. Indeed, the core is claimed by its designers to have an overall negative reactivity feedback in this scenario. This feature was demonstrated for an ASTRID-like core within the ESNII+ framework studies performed by nine European institutions. In order to shift the paradigm towards best-estimate plus uncertainties, the nuclear data sensitivity analysis and uncertainty propagation on reactivity coefficients has to be carried out. The goal of this work is to assess the impact of nuclear data uncertainties on sodium boiling reactivity feedback coefficients in order to get a more complete picture of the actual safety margins of the ASTRID low void-core design. The nuclear data sensitivity analysis is performed in parallel using SCALE-TSUNAMI 3D and the newly developed GPT SERPENT 2 module. A comparison is carried out between both methodologies. Uncertainty on the sodium boiling reactivity feedbacks is then calculated using TSAR module of SCALE and the necessary safety margins conclusions are drawn

Lumbopelvic transpedicular fixation of vertically unstable pelvic ring injuries

Introduction Identification of a proper fixation of the posterior pelvic ring is of paramount importance in treatment of patients with vertically unstable pelvic injuries. Material and methods Outcomes of 29 patients with polytrauma and vertically unstable pelvic injuries treated at Level I Trauma Center between 2013 and 2017 were analyzed. The mean age of the patients was 34.8 ± 99 years. The severity of the injuries and patients’ condition were evaluated using Injury Severity Score (ISS), VPKh-P (MT), VPKh-SP, and Yu. N. Tsibin scales (1975) to determine the sequence of treatment and diagnostic procedures. Classification offered by Pape H. C. (2005) was used to evaluate physiological condition. The ISS score was 27.1 ± 9.9. All patients underwent computed tomography (CT) scan of pelvic for preoperative planning. Lumbopelvic transpedicular fixation (LPTF) was employed as a definitive treatment of vertically unstable pelvic ring fractures in all clinical observations. Posterior half-ring morphology, a need for decompression of the nerve roots of the sacral plexus, timing of surgery were considered to decide on LPTF configuration. Results Three-month-to-six-year follow-ups of 22 patients showed good and excellent results achieved in 72.7 % of the cases that are in line with findings reported in the literature. Discussion Biomechanically adequate method of internal fixation is the method of choice in the definitive treatment of vertical unstable pelvic injuries with the possibility of decompression of compromised neural structures. Lumbopelvic fixation with the possibility of simultaneous access for decompression of neural structures is the most optimal technique for these complicated injuries

MINIMALLY INVASIVE LUMBAR-PELVIC STABILIZATION FOR UNSTABLE PELVIC RING INJURIES

Reconstructive operations for unstable  pelvic ring injuries  in most cases are performed  at later  date  after  trauma (period  of complete  stabilization of the vital functions). The paper presents  treatment outcomes  of three  patients with vertically unstable  pelvic ring injuries where minimally invasive lumbar-pelvic fixation with pedicle screws was applied. The morphology  of sacrum  injury  determined a configuration of the  lumbar-pelvic transpedicular system. In all cases the final surgery was performed  in the early period of traumatic disease, which made it possible to restore  the anatomy of the pelvic ring and obtain good functional outcomes

The assessment of direct results of drainage operations in the biliary tract in patients with obstructive jaundice

The article presents the results of a clinical examination of 88 patients aged 22 to 91 years who underwent drainage operations for obstructive jaundice in the Gomel Clinical Oncology Center between January 2017 and October 2019. The analysis of drainage interventions on the biliary tract, the duration of jaundice, hospitalization, complications.В статье представлены результаты клинического обследования 88 пациентов в возрасте от 22 до 91 года, перенесших дренирующие операции по поводу механической желтухи в Гомельском клиническом онкологическом диспансере в период с января 2017 по октябрь 2019 года. Проводился анализ дренирующих вмешательств на желчевыводящих путях, длительность желтухи, госпитализации, осложнений

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