Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Fuel Development”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Coupling of Thermochemistry Solver THERMOCHIMICA with MOOSE/BISON

THERMOCHIMICA is a computational library for chemical thermodynamics. It determines a unique combination of phases and their compositions for a prescribed chemical composition, temperature and pressure. The solver and the underlying thermodynamic models can be used to estimate the chemical state and various constitutive and transport properties necessary for nuclear fuel performance modeling. Our NEAMS research project focuses on the development of nuclear fuel thermodynamics models and computational thermodynamics equilibrium solvers, and their implementation into Idaho National Laboratory (INL) nuclear fuel performance simulation system MOOSE/BISON. This report documents the software integration of THERMOCHIMICA with MOOSE/BISON using the current INL development framework GitLab. The THERMOCHIMICA implementation is demonstrated on modeling oxygen related phenomena in irradiated fuel, such as calculation of oxygen to metal ratio in the fluorite phase, oxygen partial pressure, oxygen chemical potential and oxygen transport. Experimental measurements from the open literature were used to validate the implemented models and illustrate functionality of the developed thermodynamics module. The calculations are based on chemical element inventory provided by neutronics, isotopic depletion, transmutation and decay calculations in the SCALE system. The burnup calculation in BISON can also be used for elemental composition although it provides only a limited set. Coupling of THERMOCHIMICA with other codes that can provide chemical composition of irradiated fuel for BISON simulations is currently being explored.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Mesoscale Modeling of the Effects of Accelerated Burnup on UO2 Microstructural Evolution

Accelerating the nuclear fuel qualification process will rely on some combination of advanced modeling and simulation techniques with accelerated irradiation testing and separate effects experiments to enable the development of new fuel concepts in a shorter time frame. One of the key challenges to successfully leveraging accelerated irradiation tests will be understanding the artifacts that may be introduced with accelerated accumulation of dose and/or burnup. This work presents phase field (MARMOT) simulations of the evolution of representative 2D UO2 microstructures up to 40 MWd/kgU. Simulations were performed under both commercial light water reactor fuel conditions as well as those that would be expected for highly accelerated (~10x) burnup conditions similar to those used in the MiniFuel irradiations in Oak Ridge National Laboratory’s High Flux Isotope Reactor. The phase field model was coupled with a discrete nucleation algorithm to model re- structuring at high burnup. The effect of the different fission rates in both microstructures was investigated at two temperatures: 650?C and 800?C. The lower temperature simulations both showed an onset of restructuring at nearly 60 MWd/kgU. More extensive restructuring was obtained in the MiniFuel microstructure compared with that of the PWR fuel. At 800?C, no restructuring was obtained as a result of the thermally activated diffusion of Xe atoms and U vacancies to fission gas bubbles, which reduces the nucleation driving force. These results highlight the importance of using modeling and simulation tools to inform the environmental conditions during targeted accelerated irradiation tests to extract the most useful fuel performance data.

accelerated fuel qualification, Phase Field, Restr↗
DOE OSTI2016

Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Status Report on the Development of a Thermodynamic Database for Use with Metallic Fuel Design

The scope of this FY18 milestone involves the development of a high fidelity thermodynamic database for the U-Mo-Ti-Zr system to support the accelerated design of new pseudo binary U-xM metallic fuel, where M is a ternary metallic additive composed of Mo, Ti and Zr and x is the additive amount. Based on the previous assessment of the Mo-Ti-Zr ternary system performed in FY17, a complete set of thermodynamic model parameters describing the U-Mo-Ti-Zr quaternary system over the entire composition and temperature range have been developed in this work using the CALPHAD (CALculation of PHAse Diagrams) approach with inputs from both experimental measurements and density functional theory (DFT) total energy calculations. The two ternary intermetallic compounds, the delta phase in the U-Ti-Zr system and the cubic C15 Laves phase in the Mo-Ti-Zr system, are modeled using physically-based two-sublattice compound energy formalism. The liquid, bcc and hcp phases are described using the random solution model. Using the present database, the melting temperatures and bcc onset temperatures of U-xM metallic fuels are predicted as a function of both additive composition and amount. An optimal metallic additive composition has been identified in this work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Benefits from developments in the field of Decommissioning for Fukushima Daiichi fuel debris retrieval: Remote-Controlled Laser Cutting Process - 17110

Among the numerous challenges towards the decommissioning of Fukushima Dai-ichi reactors, removing fuel debris is certainly one of the tougher ones, requiring several technical developments. One of them, related to laser cutting of fuel debris is being developed by a French consortium 'ONET Technologies, CEA and IRSN'. Remote controlled laser cutting was implemented for the first time in a real decommissioning workshop in 2015, in UP1 reprocessing plant. In addition to laser cutting development in air and underwater, key aspects of fuel debris cutting of representative of corium debris simulants was performed to test the quantities of aerosols, gas and dusts released during the debris laser cutting. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Advanced Fuel Cycle Technology: Special Session in Honor of Dr. Michael Lineberry

The US DOE recently initiated an effort to develop accident tolerant fuel designs for potential use in commercial power reactors. Evaluation of various fuel design concepts will require a broad array of testing that will include performance attributes at both steady state and transient irradiation conditions. The first stage of the transient testing program is intended to establish the relative performance limits of each proposed concept and to support development of first-draft fuel performance models. It is anticipated that this data can subsequently be used as the basis for larger scale qualification testing. This initial stage of the testing program is outlined in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development and demonstration of a methodology to evaluate high burnup fuel susceptibility to pulverization under a loss of coolant transient

For economic reasons, the US nuclear industry is renewing efforts to build a technical basis to extend rod average burnup limits above the current regulatory burnup limit of 62 GWd/MTU. The primary driver is to increase pressurized water reactor cycle lengths to 24 months, reducing the number of fresh fuel assemblies and core design constraints, thereby making core energy utilization more efficient. However, fuel pellet fragmentation and pulverization, termed high burnup fuel fragmentation (HBFF), has been observed in the high burnup (>90 GWd/MTU) Halden loss-of-coolant-accident (LOCA) integral test series. The issue gained attention when fuel fragmentation and pulverization were also observed closer to the current US regulatory limit during the US Nuclear Regulatory Commission (NRC) sponsored out-of-core integral test at Studsvik Nuclear in early 2011. This led to NRC concerns with potential changes to fuel and core designs relative to fuel pellet pulverization. In a letter to the NRC Commissioners, the staff specifically identified a need to “…define the boundary of safe operation for key fuel design and operating parameters,” stating that “the staff is challenged to evaluate the acceptability of future fuel design advancements and fuel utilization changes.” As such, it can be concluded that HBFF and potential dispersal into the reactor coolant system introduces additional complications in light-water reactor (LWR) fuel safety evaluations. However, it is not clear how much fuel will be susceptible to HBFF; nor has there been a methodology developed to evaluate fuel susceptibility to HBFF. To that end, this paper proposes an analysis methodology to assess fuel susceptibility to HBFF during LOCA scenarios. The work presented here uses the BISON fuel performance code to evaluate a representative pressurized water reactor fuel rod exposed to a rod average burnup of 75 GWd/MTU. Sensitivity studies investigated the impact of the peak cladding temperature, transient fission gas released, and pre-transient fission gas release on cladding ballooning and burst timing. Subsequently, a methodology to assess fuel susceptibility to HBFF will be developed based on experimental data published in the open literature. The methodology will then be demonstrated by calculating the mass of fuel susceptibility to HBFF. The BISON results conclude that increasing peak cladding temperature drastically decreased time to failure, and decreased balloon size both of which have been confirmed experimentally. Additionally, the effect of pre-transient and transient fission gas release affected cladding balloon size and burst timing. Finally, fuel susceptibility to HBFF significantly decreased as a function of peak cladding temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Thermodynamic modeling of hydrogen fueling process from high-pressure storage tank to vehicle tank

This study develops a hydrogen fueling station (HFS) thermodynamic model that simulates the actual fueling process in which hydrogen is supplied from a high-pressure (HP) storage tank into a fuel cell electric vehicle (FCEV) tank. To make the model as accurate as possible, we use the same components and specifications as in actual HFSs, such as a pressure control valve, a pre-cooling system, and an FCEV tank. After the components and their specifications are set, pressure and temperature profiles are set as the HP tank supply conditions. Based on the pressure and temperature profiles, the model solves for the temperature, pressure, and mass flow rate of hydrogen at each downstream position, including the inside of the vehicle tank. The values predicted by the model are compared with experimental data, and we show that the developed model makes it possible to accurately simulate those values at any position during the fueling process.

08 HYDROGEN↗
DOE OSTI2017

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and characteristics. While INL has developed Designs A and B, LANL is evaluating a Hot Isostatic Pressed reactor configuration to overcome the drawbacks of the monolithic core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Capabilities Development for Transient Testing of Advanced Nuclear Fuels at TREAT

The TREAT facility is a unique capability at the Idaho National Laboratory currently being prepared for resumption of nuclear transient testing. Accordingly, designs for several transient irradiation tests are being pursued to enable development of advanced nuclear fuels and materials. In addition to the reactor itself, the foundation for TREAT’s capabilities also include a suite of irradiation vehicles and supporting infrastructure to provide the desired specimen boundary conditions while supporting a variety of instrumentation needs. The challenge of creating these vehicles, especially since many of the modern data needs were not historically addressed in TREAT experiment vehicles, has necessitated a sizeable engineering effort. This effort is currently underway and maturing rapidly. This paper summarizes the status, future plans, and rationale for TREAT experiment vehicle capabilities. Much of the current progress is focused around understanding and demonstrating the behavior of fuel design with enhanced accident tolerance in water-cooled reactors. Additionally, several related efforts are underway to facilitate transient testing in liquid sodium, inert gas, and steam environments. This paper discusses why such a variety of capabilities are needed, outlines plans to accomplish them, and describes the relationship between transient data needs and the irradiation hardware that will support the gathering of this information.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Heat deposition analysis for the High Flux Isotope Reactor’s HEU and LEU core models

The High Flux Isotope Reactor at Oak Ridge National Laboratory is an 85 MW th pressurized light-water-cooled and -moderated flux-trap type research reactor. The reactor is used to conduct numerous experiments, advancing various scientific and engineering disciplines. As part of an ongoing program sponsored by the US Department of Energy National Nuclear Security Administration Office of Material Management and Minimization, studies are being performed to assess the feasibility of converting the reactor’s highly enriched uranium fuel to low-enriched uranium fuel. To support this conversion project, reference models with representative experiment target loading and explicit fuel plate representation were developed and benchmarked for both fuels to (1) allow for consistent comparison between designs for both fuel types and (2) assess the potential impact of low-enriched uranium conversion. These high-fidelity models were used to conduct heat deposition analyses at the beginning and end of the reactor cycle and are presented herein. This article (1) discusses the High Flux Isotope Reactor models developed to facilitate detailed heat deposition analyses of the reactor’s highly enriched and low-enriched uranium cores, (2) examines the computational approach for performing heat deposition analysis, which includes a discussion on the methodology for calculating the amount of energy released per fission, heating rates, power and volumetric heating rates, and (3) provides results calculated throughout various regions of the highly enriched and low-enriched uranium core at the beginning and end of the reactor cycle. These are the first detailed high-fidelity heat deposition analyses for the High Flux Isotope Reactor’s highly enriched and low-enriched core models with explicit fuel plate representation. Lastly, these analyses are used to compare heat distributions obtained for both fuel designs at the beginning and end of the reactor cycle, and they are essential for enabling comprehensive thermal hydraulics and safety analyses that require detailed estimates of the heat source within all of the reactor’s fuel element regions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Poly(aryl piperidinium) membranes and ionomers for hydroxide exchange membrane fuel cells

One promising approach to reduce the cost of fuel cell systems is to develop hydroxide exchange membrane fuel cells (HEMFCs), which open up the possibility of platinum-group-metal-free catalysts and low-cost bipolar plates. However, scalable alkaline polyelectrolytes (hydroxide exchange membranes and hydroxide exchange ionomers), a key component of HEMFCs, with desired properties are currently unavailable, which presents a major barrier to the development of HEMFCs. Here we show hydroxide exchange membranes and hydroxide exchange ionomers based on poly(aryl piperidinium) (PAP) that simultaneously possess adequate ionic conductivity, chemical stability, mechanical robustness, gas separation and selective solubility. Here, these properties originate from the combination of the piperidinium cation and the rigid ether-bond-free aryl backbone. A low-Pt membrane electrode assembly with a Ag-based cathode using PAP materials showed an excellent peak power density of 920 mW cm -2 and operated stably at a constant current density of 500 mA cm -2 for 300 h with H 2 /CO 2 -free air at 95 °C.

25 ENERGY STORAGE↗
DOE OSTI2025

SEM Milestone Report for MARMOT Validation Project

MARMOT is a meso-scale fuel code under development by NEAMS designed to predict the microstructure evolution of fuel materials to determine updated thermal and mechanical property models. These updated models can be used by engineering scale codes such as BISON. To support its development and validation the MARMOT Validation Project (MVP) was initiated and consists of extensive microstructure and thermal characterization of U-10Zr fuel material before and at discreet levels of irradiation. This report summarizes the SEM and EDS characterization of the lowest burnup samples (1.5% FIMA) and how they compare to pre-irradiation SEM and EDS.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Stationary Direct Methanol Fuel Cells Using Pure Methanol

This project developed stationary direct methanol fuel cells (DMFCs) using pure methanol as the fuel, integrating complementary institutional expertise on platinum group metals free (PGM-free) cathode catalyst (UB), supported anode catalysts (KSU), electrode fabrication, characterization, and optimization (CMU), and multi-phase mass transfer (KU). The proposed cost-effective prototype will deliver reliable and durable performance using a methanol-tolerant PGM-free cathode and a high-performance anode with low PtRu loading. At the end of the project, the research team is expected to deliver a single cell (50 cm 2 ) with ≤3 mg PGM /cm 2 that achieves 300 mW/cm 2 using pure methanol.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2019

Spent Nuclear Fuel and High-Level Radioactive Waste Inventory Report

This report provides information on the inventory of commercial spent fuel (SNF) and high-level radioactive waste (HLW) in the United States, as well as non-commercial SNF and HLW in the U.S. Department of Energy (DOE) complex. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future commercial nuclear power generation scenarios. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves). This report was produced for the U.S. Department of Energy (DOE) to support various analyses on options for storage and transport of SNF and HLW and sponsored by DOE’s Office of Spent Fuel and Waste Disposition (SFWD). The report draws from and complements a previously issued report, Fuel Cycle Potential Waste Inventory for Disposition, developed for DOE’s Used Nuclear Fuel Research & Development Campaign. The current report is not intended as a revision to either Fuel Cycle Potential Waste Inventory for Disposition or Evaluation of Options for Permanent Geologic Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste in Support of a Comprehensive National Nuclear Fuel Cycle Strategy from which some underlying data is used as referenced in the current report. Rather, this report is intended as a stand-alone report providing estimates of current and projected SNF and HLW inventory and has been generated for SFWD planning purposes. This is a technical report that does not take into account contractual limitations or obligations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). For example, under the provisions of the Standard Contract, spent nuclear fuel in multi-assembly canisters is not an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this report conflict with the provisions of the Standard Contract, the Standard Contract governs the obligations of the parties, and this report in no manner supersedes, overrides, or amends the Standard Contract. This report reflects technical work which could support future decision making by DOE. No inferences should be drawn from this report regarding future actions by DOE, which are limited both by the terms of the Standard Contract and a lack of Congressional appropriations for the Department to fulfill its obligations under the Nuclear Waste Policy Act including licensing and construction of a spent nuclear fuel repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Contribution of IAEA coordinated research projects to light water reactors advanced technology fuel testing and simulation

The Fukushima-Daiichi accident in 2011 highlighted the pressing need for enhanced design and safety analysis of nuclear fuels, especially under accident conditions in nuclear power plants (NPPs). To address this need, the international nuclear fuel community has developed several concepts of Accident Tolerant and Advanced Technology Fuels (ATFs) for light water reactors. The International Atomic Energy Agency (IAEA), at the request of its Member States, has initiated a series of three Coordinated Research Projects (CRPs) to aid in the development and testing of these ATFs. Further, these projects, known as FUMAC, ACTOF, and ATF-TS, focus on the experimentation and simulation of ATFs under various accident conditions, including design basis accidents and design extension conditions. This paper provides a comprehensive overview of the main activities, as well as the achieved or anticipated results of these three CRPs. The insights and advancements gained from these projects will enable IAEA Member States to further refine nuclear fission technology, thereby contributing to efforts aimed at mitigating climate change.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

The Challenge Ahead: A Critical Perspective on Meeting U.S. Growth Targets for Sustainable Aviation Fuel

With demand for jet fuel expected to more than double by 2050 and triple by 2070, continued and accelerated efforts to decarbonize the global aviation sector are needed to curtail rising emissions and avert the worst outcomes of climate change. In response to this call-to-action, the United States has unveiled the Sustainable Aviation Fuel (SAF) Grand Challenge, which seeks to expedite the development of alternative fuel pathways that offer a minimum of a 50% reduction in life cycle greenhouse gas emissions compared to conventional jet fuel. However, there is growing concern on whether sufficient biomass will be readily available to satisfy the projected sharp rise in demand for SAF. Carbon dioxide point-source emissions combined with rising supply from direct air capture efforts have the potential to complement and offset any supply gaps for biomass-derived SAF and other use cases. Independent of the chosen feedstock and conversion technology, numerous challenges persist across SAF production in seeking to drive down fuel cost and carbon intensity. To hit the 2030 target of 3 billion gal/yr, this means about a 100% compound annual growth rate in volumetric SAF production (about 130 times scale-up) will need to be realized. To achieve the stretch goal of 35 billion gallons of SAF per year by 2050, volumetric production must increase by a factor of 12 from the 3-billion-gal/year 2030 target of SAF Grand Challenge. The comparatively longer runway for 2050 combined with significant production volume growth suggests a more inclusive "all-hands-on-deck"-style approach will likely play a role. A key component in success lies in the coordinated research, development, demonstration, and deployment efforts by multiple federal agencies and industry partnerships.

09 BIOMASS FUELS↗
DOE OSTI2025

Comparison of a Full-Scale and a 1:10 Scale Low-Speed Two-Stroke Marine Engine Using Computational Fluid Dynamics

International marine shipping is a growing component of international trade; a vast majority of all the world’s goods are being transported on large ocean-going vessels. The International Maritime Organization (IMO) introduced the Energy Efficiency Design Index in 2013, a regulatory framework of associated metrics for reducing emissions of CO 2 per tonne-mile from shipping by approximately 10% each decade. Therefore, decarbonizing the maritime sector requires the development of new fuel sources. Because of the extremely large physical size of the internal combustion engines present in shipping vessels, experimental iterative development of the engine and fuel system is cost-prohibitive. Thus, the ability to perform combustion system development in a scaled platform that can be more easily operated and modeled computationally is of interest. To that end, scaling relationships are needed to translate the results from a smaller engine to a larger counterpart. Scaling studies to date have been restricted to low scaling ratios, four-stroke light-duty engines, and under-resolved computational fluid dynamic simulations that likely do not accurately capture the physics of scaling. In this work, computational models of a 1:10 scale and a full-scale two-stroke crosshead low-speed marine engine were created and validated against experiments obtained in a real 1:10 scale engine installed at Oak Ridge National Laboratory. Further, due to the large size of the full-scale engine, the model required large high-performance computing resources to be evaluated. The availability of high-performance computing resources at the Department of Energy’s Leadership Computing Facilities is an enabler of the current work. The results of the small- and large-scale engine simulations were compared to analyze the effectiveness of the appropriate scaling laws under these extreme scaling ratio conditions.

33 ADVANCED PROPULSION SYSTEMS↗