Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “technical issues”

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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize plant safety and operational margins. Advanced MOOSE tools that are needed to complete this integrated evaluation model are: RAVEN, RELAP-7, BISON, and MAMMOTH.

10CFR50.46C↗
DOE OSTI2015

PORFLOW Simulations Supporting Saltstone Disposal Unit Design Optimization

SRNL was requested by SRR to perform PORFLOW simulations to support potential cost-saving design modifications to future Saltstone Disposal Units in Z-Area (SRR-CWDA-2015-00120). The design sensitivity cases are defined in a modeling input specification document SRR-CWDA-2015-00133 Rev. 1. A high-level description of PORFLOW modeling and interpretation of results are provided in SRR-CWDA-2015-00169. The present report focuses on underlying technical issues and details of PORFLOW modeling not addressed by the input specification and results interpretation documents. Design checking of PORFLOW modeling is documented in SRNL-L3200-2015-00146.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

The Status and Promise of Advanced M&V: An Overview of “M&V 2.0” Methods, Tools, and Applications

Advanced measurement and verification (M&V) of energy efficiency savings, often referred to as M&V 2.0 or advanced M&V, is currently an object of much industry attention. Thus far, however, there has been a lack of clarity about what techniques M&V 2.0 includes, how those techniques differ from traditional approaches, what the key considerations are for their use, and what value propositions M&V 2.0 presents to different stakeholders. The objective of this paper is to provide background information and frame key discussion points related to advanced M&V. The paper identifies the benefits, methods, and requirements of advanced M&V and outlines key technical issues for applying these methods. It presents an overview of the distinguishing elements of M&V 2.0 tools and of how the industry is addressing needs for tool testing, consistency, and standardization, and it identifies opportunities for collaboration. In this paper, we consider two key features of M&V 2.0: (1) automated analytics that can provide ongoing, near-real-time savings estimates, and (2) increased data granularity in terms of frequency, volume, or end-use detail. Greater data granularity for large numbers of customers, such as that derived from comprehensive implementation of advanced metering infrastructure (AMI) systems, leads to very large data volumes. This drives interest in automated processing systems. It is worth noting, however, that automated processing can provide value even when applied to less granular data, such as monthly consumption data series. Likewise, more granular data, such as interval or end-use data, delivers value with or without automated processing, provided the processing is manageable. But it is the combination of greater data detail with automated processing that offers the greatest opportunity for value. Using M&V methods that capture load shapes together with automated processing1 can determine savings in near-real time to provide stakeholders with more timely and detailed information. This information can be used to inform ongoing building operations, provide early input on energy efficiency program design, or assess the impact of efficiency by location and time of day. Stakeholders who can make use of such information include regulators, energy efficiency program administrators, program evaluators, contractors and aggregators, building owners, the investment community, and grid planners. Although each stakeholder has its own priorities and challenges related to savings measurement and verification, the potential exists for all to draw from a single set of efficiency valuation data. Such an integrated approach could provide a base consistency across stakeholder uses.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2016

Argillite And Crystalline Disposal Research: Accomplishments And Path-Forward.

The intention of this document is to provide a path-forward for research and development (R&D) for two host rock media-specific (argillite and crystalline) disposal research work packages within the Used Fuel Disposition Campaign (UFDC). The two work packages, Argillite Disposal R&D and Crystalline Disposal R&D, support the achievement of the overarching mission and objectives of the Department of Energy Office of Nuclear Energy Fuel Cycle Technologies Program. These two work packages cover many of the fundamental technical issues that will have multiple implications to other disposal research work packages by bridging knowledge gaps to support the development of the safety case. The path-forward begins with the assumption of target dates that are set out in the January 2013 DOE Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (http://energy.gov/downloads/strategy-management-and-disposal-used-nuclear-fuel-and-high-levelradioactive- waste). The path-forward will be maintained as a living document and will be updated as needed in response to available funding and the progress of multiple R&D tasks in the Used Fuel Disposition Campaign and the Fuel Cycle Technologies Program. This path forward is developed based on the report of “Used Fuel Disposition Campaign Disposal Research and Development Roadmap (FCR&D-USED- 2011-000065 REV0)” (DOE, 2011). This document delineates the goals and objectives of the UFDC R&D program, needs for generic disposal concept design, and summarizes the prioritization of R&D issues.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Migration of Older to New Digital Control Systems in Nuclear Power Plant Main Control Rooms

The United States (U.S.) Department of Energy (DOE) Office of Nuclear Energy (NE) has the primary mission to advance nuclear power by resolving socio-technical issues through research and development (R&D). One DOE-NE activity supporting this mission is the Light Water Reactor Sustainability (LWRS) program. LWRS has the overall objective to sustain the operation of existing commercial nuclear power plants (NPPs) through conducting R&D across multiple “pathways,” or R&D focus areas. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies pathway conducts targeted R&D to address aging and reliability concerns with the legacy instrumentation and control (I&C) and related information systems in operating U.S. NPPs. This work involves (1) ensuring that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) implementing digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Under the LWRS Advanced II&C pathway, Human Factors experts at Idaho National Laboratory (INL) have been conducting R&D in support of NPP main control room (MCR) modernization activities. Work in prior years has focused on migrating analog I&C systems to new digital I&C systems (). In fiscal year 2016 (FY16), one new focus area for this research is migrating older digital I&C systems to new and advanced digital I&C systems. This report summarizes a plan for conducting a digital-to-digital migration of a legacy digital I&C system to a new digital I&C system in support of control room modernization activities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Completion Report for Well ER-4-1 Corrective Action Unit 97: Yucca Flat/Climax Mine, Revision 0

Well ER-4-1 was drilled for the U.S. Department of Energy, Nevada National Security Administration Nevada Field Office in support of the Underground Test Area (UGTA) Activity. The well was drilled and completed from March 23 to April 13, 2016, as part of the Corrective Action Investigation Plan (CAIP) for Yucca Flat/Climax Mine Corrective Action Unit (CAU) 97. The primary purpose of the well was to collect hydrogeologic data to assist in validating concepts of the groundwater flow system within the Yucca Flat/Climax Mine CAU, and to test for potential radionuclides in groundwater from the STRAIT (U4a) underground test. The completed well includes one piezometer (p1), to a depth of 663.16 meters (m) (2,175.71 feet [ft]) below ground surface (bgs) and open from the Alluvial aquifer (AA3) to the Oak Spring Butte confining unit (OSBCU) hydrostratigraphic units; and a main completion (m1), which includes 6.625-inch (in.) casing with slotted interval (m1) installed to 906.80 m (2,975.05 ft) bgs in the Lower carbonate aquifer (LCA). A 13.375-in. diameter surface casing was installed from the surface to a depth of 809.00 m (2,654.21 ft) bgs. Well ER-4-1 experienced a number of technical issues during drilling, including borehole instability and sloughing conditions. An intermediate, 10.75-in./9.625-in. casing string was installed to 856.94 m (2,811.48 ft) bgs to control these issues. Borehole stability and erosion problems appear to be associated with the Tunnel Formation (Tn) and the Older tunnel beds (Ton). Overall efforts to stabilize the borehole were successful. Data collected during borehole construction include composite drill cutting samples collected every 3.0 m (10 ft), a partial suite of geophysical logs to a maximum depth of 766.57 m (2,515 ft) bgs, water-quality measurements (including tritium), water-level measurements, and two depth-discrete bailer samples collected at 538.89 m and 646.18 m (1,768 ft and 2,120 ft) bgs respectively. The well penetrated 187.45 m (615 ft) of Quaternary/Tertiary alluvium (QTa), 671.47 m (2,203 ft) of Tertiary Volcanic rocks (Tv), and 66.20 m (217.19 ft) of Paleozoic rocks (|). The stratigraphy and lithology were generally as expected with some exceptions. The top of Paleozoic rocks (|) was predicted to occur at 822.35 m (2,698 ft) bgs and was intercepted at 858.93 m (2,818 ft), a difference of 36.58 m (120 ft). As expected, the Paleozoic rocks (|) are the principal water producing formation in Well ER-4-1. Depth to water was measured after drilling as follows: In the piezometers: p1 at 320.39 m (1,051.16 ft) bgs, (measured January 4, 2017); and in the main production casing interval: m1 at 539.17 m (1,768.92 ft) bgs, (measured December 12, 2016) Geophysical logs and depth-discrete bailer sample analytical results suggest likely zones of prompt injection (underground-test-related) fission products from 472.44 to 481.48 m (1,550 to 1,580 ft) bgs and at approximately 539.50 m (1,770 ft) bgs. Subsequent work at Well ER-4-1 will be included in future reports. Field measurements for tritium were mostly below the Safe Drinking Water Act limit (20,000 picocuries per liter) with the exception of two zones showing elevated tritium concentrations. The first zone is located at approximately 365.76 to 390.14 m (1,200 to 1,280 ft) bgs and a second zone at approximately 542.54 to 566.93 m (1,780 to 1,860 ft) bgs. All Fluid Management Plan requirements were met.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Supplemental Information for New York State Standardized Interconnection Requirements

This document is intended to aid in the understanding and application of the New York State Standardized Interconnection Requirements (SIR) and Application Process for New Distributed Generators 5 MW or Less Connected in Parallel with Utility Distribution Systems, and it aims to provide supplemental information and discussion on selected topics relevant to the SIR. This guide focuses on technical issues that have to date resulted in the majority of utility findings within the context of interconnecting photovoltaic (PV) inverters. This guide provides background on the overall issue and related mitigation measures for selected topics, including substation backfeeding, anti-islanding and considerations for monitoring and controlling distributed energy resources (DER).

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2018

Proceedings of the 8th US/German Workshop on Salt Repository Research Design and Operation.

This document records the Proceedings of the 2017 gathering of salt repository nations. In a spirit of mutual support, technical issues are dissected, led capably by subject matter experts. As before, it is not possible to explore all contemporary issues regarding nuclear waste disposal in salt formations. Instead, the group focused on a few selected issues to be pursued in depth, while at the same time acknowledging and recording ancillary issues.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Gas Migration Project: Risk Assessment Tool and Computational Analyses to Investigate Wellbore/Mine Interactions, Secretary's Potash Area, Southeastern New Mexico

The Bureau of Land Management (BLM), US Department of the Interior has asked Sandia National Laboratories (SNL) to perform scientific studies relevant to technical issues that arise in the development of co-located resources of potash and petroleum in southeastern New Mexico in the Secretary’s Potash Area. The BLM manages resource development, issues permits and interacts with the State of New Mexico in the process of developing regulations, in an environment where many issues are disputed by industry stakeholders. The present report is a deliverable of the study of the potential for gas migration from a wellbore to a mine opening in the event of wellbore leakage, a risk scenario about which there is disagreement among stakeholders and little previous site specific analysis. One goal of this study was to develop a framework that required collaboratively developed inputs and analytical approaches in order to encourage stakeholder participation and to employ ranges of data values and scenarios. SNL presents here a description of a basic risk assessment (RA) framework that will fulfill the initial steps of meeting that goal. SNL used the gas migration problem to set up example conceptual models, parameter sets and computer models and as a foundation for future development of RA to support BLM resource development.

02 PETROLEUM↗
DOE OSTI2018

Structural Health Monitoring of Nuclear Spent Fuel Storage Facilities

Interim storage of spent nuclear fuel from reactor sites has gained additional importance and urgency for resolving waste-management-related technical issues. To ensure that nuclear power remains clean energy, monitoring has been identified by DOE as a high priority cross-cutting need, necessary to determine and predict the degradation state of the systems, structures, and components (SSCs) important to safety (ITS). Therefore, nondestructive structural condition monitoring becomes a need to be installed on existing or to be integrated into future storage system to quantify the state of health or to guarantee the safe operation of nuclear power plants (NPPs) during their extended life span. In this project, the lead university and the collaborating national laboratory teamed to develop a nuclear structural health monitoring (n-SHM) system based on in-situ piezoelectric sensing technologies that can monitor structural degradation and aging for nuclear spent fuel DCSS and similar structures. We also aimed to identify and quantify possible influences of nuclear spent fuel environment (temperature and radiation) to the piezoelectric sensor system and come up with adequate solutions and guidelines therefore. We have therefore developed analytical model for piezoelectric based n-SHM methods, with considerations of temperature and irradiation influence on the model of sensing and algorithms in acoustic emission (AE), guided ultrasonic waves (GUW), and electromechanical impedance spectroscopy (EMIS). On the other side, experimentally the temperature and irradiation influence on the piezoelectric sensors and sensing capabilities were investigated. Both short-term and long-term irradiation investigation with our collaborating national laboratory were performed. Moreover, we developed multi-modal sensing, validated in laboratory setup, and conducted the testing on the We performed multi-modal sensing development, verification and validation tests on very complex structures including a medium-scale vacuum drying chamber and a small-scale mockup canister available for the desired testing. Our work developed the potential candidate for long term structural health monitoring of spent fuel canister through piezoelectric wafer sensors and provided the sensing methodologies based on AE and GUW methodologies. It overall provides an innovative system and methodology for enhancing the safe operation of nuclear power plant. All major accomplishments planned in the original proposal were successfully achieved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Progress Summary: 12-Month Compression Set and Load Retention of Thermally Accelerated Aged Siloxane Specimens (Additively Manufactured by the Direct-Ink-Write (DIW) 3D Printing Technique) Was Measured

Motivation: Accelerated aging of polymeric materials is needed to predict their long-term mechanical functionality. Two key metrics used to evaluate their performance are compression set and load retention. A baseline(time=0)mechanical assessment was done prior to initiating the accelerated aging process to acquire initial thickness and load values for future calculation of compression set and load retention.Specimens are beingaged under compressive strain (13%for the current studies) at four different temperatures (25, 40, 55, and 70°C) for at least 1 year; compression set and load retention will be measured at various timepoints. Time-temperature superposition (TTS) will be applied to the isothermal data to obtain long-term predictions. Significance: Compression set is increasing with time and temperature, as expected. The increasing compression set is accompanied by decreasing load retention.The relativelylow compression set indicates the material is performing well.In a previous aging study, the legacy S5370 RTV foam exhibited 40% relative compression set after 12 months at 70°C, which is more than twice that of the DIW material. Progress:Data has been acquired out to the 12-month timepoint. Relative compression set and load retention versus time are shown in Fig. 1.A technical issue with the method of measuring load retention was discovered and previously reported load retention values have been adjusted; the current load retention values are estimates. TTS predictions are shown in Fig. 2.

36 MATERIALS SCIENCE↗
DOE OSTI2018

MELCOR Modeling of Non-LWR Systems Draft Report for the Nuclear Regulatory Commission

This report provides an overview of technical issues and design features relevant to advanced reactors and reviews MELCOR's current readiness for modeling accidents in such reactor types. This report describes advanced reactor physics models currently available or under development, and gauges the level of effort required to develop new models and capabilities applicable to assessing advanced reactor safety issues. Finally, this report reviews the available database that can be used in verification and validation of new models. Four general advanced reactor types are considered in this report: 1) High Temperature Gas-Cooled Reactor (HTGR); 2) Sodium Fast Reactor (SFR); 3) Molten Salt Reactor (MSR); 4) Fluoride Salt-Cooled High Temperature Reactor (FHR)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Efficient Solution Methods for Large-scale Optimization Problems Constrained by Time-dependent Partial Differential Equations (Final Report)

The objectives of this project are to devise and analyze efficient solvers for several classes of PDE- constrained optimization problems. The project was initially focused on, but not restricted to time- dependent PDE-constraints such as parabolic equations, time-dependent fluid flows (Navier-Stokes equations), hyperbolic equations (Burgers equation, shallow water equations). Other constraints considered are PDEs with stochastic coefficients. Additional elements of interest are the handling of problems with control and/or state inequality constraints or non-differentiable regularization terms, and boundary controls. A direction of work not initially listed among the goals, which was triggered by conversations with DOE scientists from the Sandia National Labs, is to develop efficient solvers for optimization- based domain-decomposition PDE-solvers. Also targeted in the project are a number of technical issues necessary for facilitating the integration of the methods resulted from this project with existing high-performance packages developed at DOE national labs; this integration is expected to further enable the scientific community to use the novel methods. An example of such a task is the need to develop algebraic multigrid (AMG) versions of the algorithms. A list of accomplishments in several areas is detailed as well as a list of publications.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

High Voltage Modeling and Testing of Transformer, Line Interface Devices, and Bulk System Components Under Electromagnetic Pulse, Geomagnetic Disturbance, and other Abnormal Transients

This project aims at quantitatively analyzing possible knowledge gaps related to the risk of Electromagnetic Pulse (EMP) impact on high-voltage, power transformers (transmission-class), that represent the most critical power grid assets. In relation to the electric infrastructure, some past studies indicate that the grid is overall fairly robust and resilient against most EMP threats. This assessment stems, in part, from estimates of the peak voltage that can be induced on a transmission power line for a nominal EMP fast-transient, compared with the typically very high basic insulation level that is featured by the high-voltage infrastructure. The project provides additional insight toward a more comprehensive risk evaluation of the grid EMP and Geomagnetic Disturbances (GMD) vulnerabilities. This work is aligned with the U.S. Dept. of Energy current research thrust in support of the development of a highly secure and resilient power infrastructure. Project milestones have been devised to address the fundamental technical issues in support of protection of large power transformer against EMP threats. More specifically, this project included parallel research tasks focused on the following topics: analysis of EMP-induced, fast-transient, current and voltages on a variety of power line configurations and EMP impact scenarios, detailed characterization of EMP damage potential due to the voltage surge, from the impact on the power line, to the transformer windings, GMD and EMP slow transients and impact from harmonics generation and reactive power flow on a complex grid layout, and identification of risk mitigation paths, with emphasis on GIC damage prevention for large power transformers, and on impact protection options from a grid stability perspective.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2019

Aluminum Clad Spent Nuclear Fuel Task 5: Oxide Layer Response to Drying Experiment Test Plan

This document presents a test plan developed to assess the fundamental effectiveness of two widely accepted processes for drying spent nuclear fuel: vacuum drying and helium forced-gas drying. The experimental activities planned for Task 5 follow from the “Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan.” The objectives are to evaluate the response of the oxide corrosion layer on aluminum clad spent nuclear fuel (ASNF), processed at temperatures above 100°C, to understand and compare the behavior of oxide corrosion layer compositions resulting from a range of credible procedural variations during the transition of ASNF from wet storage to dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Aluminum Clad Spent Nuclear Fuel Task 5: Oxide Layer Response to Drying Experiment Test Plan

This document presents a test plan developed to assess the fundamental effectiveness of two widely accepted processes for drying spent nuclear fuel: vacuum drying and helium forced-gas drying. The experimental activities planned for Task 5 follow from the “Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Technical Issues Action Plan.”[Connolly, INL/EXT-17-43908] The objectives are to evaluate the response of the oxide corrosion layer on aluminum clad spent nuclear fuel (ASNF), processed at temperatures above 100°C, to understand and compare the behavior of oxide corrosion layer compositions resulting from a range of credible procedural variations during the transition of ASNF from wet storage to dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Report on Hydrogen Isotope Permeability/Hermeticity of SiC-based Cladding

The continuous SiC fiber-reinforced SiC matrix ceramic composites have been perceived as one of the leading candidate materials for accident tolerant fuel cladding in light water reactors. Potential loss of the fission products containment due to penetrating cracking has been identified as one of the most critical technical issues for SiC-based nuclear fuel cladding. As such, evaluation of the hermeticity of SiC/SiC composite cladding subject to ligh water reactor (LWR) relevant environment is critically important. In last two years, we have established the capability to measure the gas permeation flux through tubular samples and reported preliminary results of the hermeticity of unirradiated SiC/SiC composite tubes and one neutron irradiated Chemical Vapor Deposition (CVD) SiC tube. In this study, a permeation testing station was employed to measure the deuterium permeation of as-fabricated and coated SiC/SiC composite tubes following high heat flux neutron irradiation. The results indicate that CVD SiC tubes survived both the high and low heat flux neutron irradiations. SiC/SiC composite tubes with and without coating following high heat flux neutron irradiation lost gas tightness. Coating on SiC/SiC composite tubes helped to reduce the deuterium permeation flux compared to the uncoated samples. Better coating and/or improved architectural design and manufacturing quality is needed to ensure the hermeticity of SiC/SiC composite tubes. A more robust understanding of the correlation between applied gas pressure and measured leak rate is needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Converting the MCNP Runtape to HDF5

The current runtape in MCNP is a Fortran unformatted IO dump of approximately 800 variables used for restarting an MCNP simulation. This format has a number of technical issues for those who interact with it. First, Fortran unformatted IO is not guaranteed to be portable. As a result, a runtape generated on one machine is not necessarily readable on another. Second, for users who wish to extract data from the runtape, there is no simple approach. It is impossible to extract one variable without knowing the size and position of every other one. Finally, the runtape itself is not backwards compatible such that a future version of MCNP could open it. Users are expected to keep around versions of MCNP that are compatible with saved runtapes.

97 MATHEMATICS AND COMPUTING↗