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

Goodbye Hartmann trial: a prospective, international, multicenter, observational study on the current use of a surgical procedure developed a century ago

Background: Literature suggests colonic resection and primary anastomosis (RPA) instead of Hartmann's procedure (HP) for the treatment of left-sided colonic emergencies. We aim to evaluate the surgical options globally used to treat patients with acute left-sided colonic emergencies and the factors that leading to the choice of treatment, comparing HP and RPA. Methods: This is a prospective, international, multicenter, observational study registered on ClinicalTrials.gov. A total 1215 patients with left-sided colonic emergencies who required surgery were included from 204 centers during the period of March 1, 2020, to May 31, 2020. with a 1-year follow-up. Results: 564 patients (43.1%) were females. The mean age was 65.9 ± 15.6 years. HP was performed in 697 (57.3%) patients and RPA in 384 (31.6%) cases. Complicated acute diverticulitis was the most common cause of left-sided colonic emergencies (40.2%), followed by colorectal malignancy (36.6%). Severe complications (Clavien-Dindo ≥ 3b) were higher in the HP group (P < 0.001). 30-day mortality was higher in HP patients (13.7%), especially in case of bowel perforation and diffused peritonitis. 1-year follow-up showed no differences on ostomy reversal rate between HP and RPA. (P = 0.127). A backward likelihood logistic regression model showed that RPA was preferred in younger patients, having low ASA score (≤ 3), in case of large bowel obstruction, absence of colonic ischemia, longer time from admission to surgery, operating early at the day working hours, by a surgeon who performed more than 50 colorectal resections. Conclusions: After 100 years since the first Hartmann's procedure, HP remains the most common treatment for left-sided colorectal emergencies. Treatment's choice depends on patient characteristics, the time of surgery and the experience of the surgeon. RPA should be considered as the gold standard for surgery, with HP being an exception

EQL3D: ERANOS based equilibrium fuel cycle procedure for fast reactors

The advanced fast reactors of the fourth generation should be capable to breed their own fuel from poorly fissile 238U and to recycle the actinides from their own spent fuel. However, this recycling or actually the closure of fuel cycle has negative impact on the safety parameters. The goal of this work is to develop a numerical tool, which can simulate and confirm the capability of these reactors to operate with closed fuel cycle, and which can evaluate their safety parameters. The tool is named equilibrium fuel cycle procedure for fast reactors (EQL3D) and is based on the ERANOS 2.1 code platform. Equilibrium cycle or virtually equilibrium method for considering the homogeneous recycling of actinides is a known approach; however, in EQL3D the equilibrium method is newly applied for hexagonal-z 3D and r-z 2D core geometries and typically 33 energy-group neutron-flux calculations. The utilization of hexagonal-z 3D geometry enables to characterize the equilibrium cycle for complex reloading patterns within a multi-batch scheme. Furthermore, EQL3D enables comparison of the advanced fast reactors on a common basis of their equilibrium cycle reactivity swing, fuel composition, breeding gain and safety-related parameters. The Gas-cooled Fast Reactor (GFR) was selected for verification and optimization of the EQL3D procedure. The GFR geometry was based on an international neutronics benchmark with a simple setup and potential for latter upgrade. It was used to show the impact of several EQL3D options e.g. different isotope evolution models, geometry selection, or cross-section recalculation frequency, on the equilibrium parameters. The results demonstrate the capability of the procedure to calculate the equilibrium fuel cycle for advanced fast reactors. Among others, also the ability of GFR benchmark core to be operated with closed fuel cycle is shown

Calculations of Reactivity-Initiated Transients in Gas-Cooled Fast Reactors Using the Code System FAST

The FAST code system is a general tool for analyzing advanced reactors from the viewpoint of the static and dynamic behavior of the whole reactor system. It includes an integrated three-dimensional representation of the core neutronics, appropriate modeling of the core thermal-hydraulics and fuel pin behavior, coupled to models of the reactor primary and secondary systems. Use is made largely of well-established individual neutronic, thermal-hydraulic and fuel behavior modules. Clearly, it is important to verify the individual parts of the code, including the links between them. The paper is focused on this detailed verification procedure. Steady-state conditions, as well as the transient behavior of hypothetical reactivity-initiated accidents, are investigated for two specific gas-cooled fast reactors. While the first system, a CO2-cooled CAPRA-CADRA core, is loaded with Superphénix-like MOX fuel, the second system being analyzed, a He-cooled Generation IV-like core, uses ceramic (U,Pu)C fuel dispersed in a silicon-carbide matrix. In the current study, the TRAC/PARCS elements of FAST are compared with the 3D-kinetics stand-alone ERANOS/KIN-3D code, which is considered state-of-the-art, using as far as possible equivalent options. A new methodology is proposed to improve a diffusion-theory, coarse-group PARCS-solution by scaling the original cross-section derivatives and input kinetic parameters

Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies

FAST code system: Review of recent developments and near-future plans

The FAST code system is currently being developed and used at the Paul Scherrer Institut for static and transient analysis of the main Generation 4 fast-spectrum reactor concepts: sodium-, helium-, and gas-cooled fast reactors. The code system includes the ERANOS code system for static neutronics calculations, as well as coupled TRACE/PARCS/FRED for neutron kinetics, thermal hydraulic, and fuel transient analysis. The paper presents the status of the recent developments in neutronics (new 3D procedure for equilibrium cycle simulation and new transient cross section generation procedure), in thermal hydraulics and chemistry (equations-of-state for new coolants, two-phase flow models for sodium, and new model for oxide layer buildup in heavy-metal flow), and in fuel behavior (new model for the dispersed gas-cooled fast reactor fuel). Near-future plans for the further development of FAST are outlined. 2010 by ASME

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 detailed design 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 voiding 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 voiding 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 the two methodologies. Uncertainty on the sodium reactivity feedbacks is then calculated using TSAR module of SCALE and the necessary safety margins conclusions are drawn