Development and Validation of a Computational Tool for Fusion Reactors\u27 System Analysis

On the roadmap to fusion energy the development and the operation of a demonstration power plant (DEMO) is the next step after ITER, a key facility currently devoted to the exploration of the physics aspects for self-sustained fusion plasmas with sizes and fusion power comparable to those attended in fusion power plants (FPP). Fusion systems codes are essential computational tools aimed to simulate the physics and the engineering features of a FPP. The main objective of a system code is to find one (or more) reactor configurations which simultaneously comply with physics operational limits, engineering constraints and net electric output requirements. As such simulation tools need to scope many design solutions over a large parameter phase space, they rely on rather basic physics and engineering models (mostly at zero or one-dimensional level) and on a relatively large number of input specifications. Within the conceptual design of a FPP, systems codes are interfaced to the detailed transport codes and engineering platforms, which operate in much larger time scales. To fill the gap between systems and the detailed transport and engineering codes the high-fidelity system/design tool MIRA (Modular Integrated Reactor Analysis) has been developed. MIRA relies on a modular structure and provides a refined FPP system analysis, with the primary goal of generating a more robust plant baseline. It incorporates into a unique computing environment a mathematical algorithm for the utmost tokamak fusion problems, including two-dimensional plasma magnetic equilibrium and core physics, transport of neutron and photon radiations emitted from the plasma and electromagnetic and engineering characterization of the toroidal field (TF) and poloidal field (PF) field coil systems. Most of the implemented modules rely on higher spatial resolution compared to presently available system codes, such as PROCESS. The multiphysics MIRA approach has been applied to the DEMO 2015 baseline, generated by means of the PROCESS system code. The analysis has been carried out by taking an identical set of input assumptions and requirements (e.g. same fusion power, major radius and aspect ratio) and observing the response on certain figures of merit. This verification study has featured the violation of some constraining conditions imposed on plasma safety factor, TF ripple and plasma burn time. The DEMO 2015 baseline has been found not in line with all the imposed requirements and constraints, hence necessitates a set of active measures on some of the input parameters. Such measures have been reported in form of parameter scans, where three variables have been identified, such as plasma internal inductance, blanket breeding zone inboard thickness and vacuum vessel/TF coil gap radial outboard width. The addressed sensitivity analyses have shown non-trivial inter-parametric dependencies, never explored in fusion system analyses. For instance, large influences of the plasma internal inductance on safety factor, plasma shape, density and temperature features, peak divertor flux and plasma burn time have been observed. Moreover, an optimal overall breeding blanket + TF coil inboard width has been observed with respect to the maximization of the plasma burn time, representing a meeting point between neutronic tritium breeding and technological limits in central solenoid and TF coils superconducting cables. These outcomes have inspired important changes in the way of designing a tokamak reactor like DEMO, where more extended analyses of the key physics and engineering aspects of the reactor can speed up and improve the design process of a FPP

Tritium transport analysis in HCPB DEMO blanket with the FUS-TPC Code (KIT Scientific Reports ; 7642)

In thermonuclear fusion reactors, the fuel is an high temperature deuterium-tritium plasma, in which tritium is bred by lithium isotopes present inside solid ceramic breeder (e.g. Li-Orthosilicate) or inside liquid eutectic alloys (e.g. Pb-16Li alloy). In the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the Helium Cooled Pebble Bed (HCPB) blanket concept or transported in solution by the owing alloy in the Helium Cooled Lead Lithium (HCLL) blanket concept. Tritium produced in the breeding blanket by neutrons interacting with lithium nuclei can enter the metal structures, and can be lost by permeation to the environment. Tritium in metallic components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents and the need for maintenance. In this study the problem of tritium transport in HCPB DEMO blanket from the generation inside the solid breeder to the release into the environment has been studied and analyzed by means of the computational code FUS-TPC (Fusion Devoted-Tritium Permeation Code). The code has been originally developed to study the tritium transport in HCLL blanket and it is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. The code has the main goal to estimate the total tritium losses into the environment and the tritium inventories inside the breeder, inside the multiplier, inside the purge gas and the main coolant loops and inside the structural materials. Different simulations of the code were performed by adopting the configuration of the European HCPB blanket for DEMO. Total tritium losses from a generic fusion power plant, is often considered a key parameter to evaluate the tritium containment capabilities (added to tritium inventories) of a certain nuclear plant. Without any tritium control techniques, permeation can be quite significant, thus some tritium transport mitigation devices are required. The code is able to model and compute different tritium fluxes exchanged in the overall tritium system. A sensitivity study for the tritium losses and inventories is performed in this work

Reducing the environmental impact of surgery on a global scale: systematic review and co-prioritization with healthcare workers in 132 countries

Abstract Background Healthcare cannot achieve net-zero carbon without addressing operating theatres. The aim of this study was to prioritize feasible interventions to reduce the environmental impact of operating theatres. Methods This study adopted a four-phase Delphi consensus co-prioritization methodology. In phase 1, a systematic review of published interventions and global consultation of perioperative healthcare professionals were used to longlist interventions. In phase 2, iterative thematic analysis consolidated comparable interventions into a shortlist. In phase 3, the shortlist was co-prioritized based on patient and clinician views on acceptability, feasibility, and safety. In phase 4, ranked lists of interventions were presented by their relevance to high-income countries and low–middle-income countries. Results In phase 1, 43 interventions were identified, which had low uptake in practice according to 3042 professionals globally. In phase 2, a shortlist of 15 intervention domains was generated. In phase 3, interventions were deemed acceptable for more than 90 per cent of patients except for reducing general anaesthesia (84 per cent) and re-sterilization of ‘single-use’ consumables (86 per cent). In phase 4, the top three shortlisted interventions for high-income countries were: introducing recycling; reducing use of anaesthetic gases; and appropriate clinical waste processing. In phase 4, the top three shortlisted interventions for low–middle-income countries were: introducing reusable surgical devices; reducing use of consumables; and reducing the use of general anaesthesia. Conclusion This is a step toward environmentally sustainable operating environments with actionable interventions applicable to both high– and low–middle–income countries

Tritium control in fusion reactor materials: A model for Tritium Extracting System

In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT)

Asthma in patients admitted to emergency department for COVID-19: prevalence and risk of hospitalization

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