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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.

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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A large scale system approach for stability analysis of multi-unit SMR plants - 248

In this paper we present a framework based on large scale system theory for analyzing the stability of coupled advanced high temperature molten salt small modular reactors. It has been suggested that small modular reactors be coupled through a common molten salt energy repository, called a salt vault. Using vector Lyapunov functions the connective stability of multiple reactors coupled through a salt vault can be guaranteed. In this theory, scalar Lyapunov functions are assigned to portions of the composite system state space. Functions are constructed for each small modular reactor (SMR), each Brayton cycle, and the salt vault. These functions are then aggregated, becoming components of a vector Lyapunov function. If this vector function possesses special properties, the composite system is stable. Bounds on the connective stability of a multi-reactor plant can be found in terms of a free parameter, in this case the mass of molten salt in the salt vault. Examining bounds on the connective stability allows one to use control theory as a tool for making performance and economic design decisions related to the size of the salt vault. This technique can produce a constraint that is used in a design optimization problem for finding the smallest-possible salt vault that ensures stability. It can also be used as an indicator of interconnection and its effect on system behavior. This, in turn, could be used to guide multi-unit control system design to ensure that highly coupled systems behave favorably. The theory for multi-unit control and operation is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Verification of severe accident management guideline (SAMG) entry condition for OPR1000 - 367

The current severe accident prevention and mitigation strategies for OPR 1000 have emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs), respectively. The main objective of SAMG is to prevent a release of radioactive materials from releasing into environment and mitigate severe accident phenomena during severe accident. The current SAMG entry condition is met when the core exit temperature (CET) reaches 650 deg. C in OPR 1000. Peak cladding temperature (PCT), which represents core integrity more precisely than CET does, cannot be used since it is not measurable in the main control room (MCR). Current SAMG entry condition for every kind of accident scenarios is not reasonable, since entry condition is determined without considering operator action time. Therefore, SAMG entry condition need to be verified with consider operator action time. Thus, an analysis methodology for SAMG entry condition was developed to verify the SAMG entry condition. A severe accident DB were also developed in order to analyze the SAMG entry condition. The analysis results of DB show that available operator action time from the SAMG entrance to RV fail is shorter than the required time in most cases. As a result, the proper SAMG entry condition was suggested through the results of DB analysis. When the SAMG entry condition is changed from CET to suggested condition, the available action time is increased about 350 to 4300 sec. Also, the suggested entry condition was verified through the simulation which considered operation action time and reflecting the suggested entry condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of Performance Validation Tool for the NPCS for APR1400

Availability of nuclear power plant is highly dependent on the quality of major control systems including NSSS Control Systems. It has been experienced that a small flaw hidden among the control functions resulted in large operational transient or even reactor trip. The proactive quality control of these systems is not easy because of the dynamical characteristics of control algorithms. In a stand-alone or open loop state, it is not possible to perform integral system tests because they need control loads to form an appropriate closed control loop. In the construction plant, Power Ascension Test (PAT) is the first chance to witness the overall performance of the control systems. Because the PAT is the last activity of plant construction schedule, the cost of corrective actions is very high if any functional defect is found in this stage. To minimize risk caused by late detection of control system defects, we developed a validation tool as a proactive test device for the NSSS Process Control System (NPCS) for the SKN Unit 3 which is the first APR1400 plant in Korea. In this paper, the features of the dynamic performance test tool, test methodology, and the test results are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Temperature-programmed deoxygenation of acetic acid on molybdenum carbide catalysts

Temperature programmed reaction (TPRxn) is a simple yet powerful tool for screening solid catalyst performance at a variety of conditions. A TPRxn system includes a reactor, furnace, gas and vapor sources, flow control, instrumentation to quantify reaction products (e.g., gas chromatograph), and instrumentation to monitor the reaction in real time (e.g., mass spectrometer). Here, we apply the TPRxn methodology to study molybdenum carbide catalysts for the deoxygenation of acetic acid, an important reaction among many in the upgrading/stabilization of biomass pyrolysis vapors. TPRxn is used to evaluate catalyst activity and selectivity and to test hypothetical reaction pathways (e.g., decarbonylation, ketonization, and hydrogenation). Furthermore, the results of the TPRxn study of acetic acid deoxygenation show that molybdenum carbide is an active catalyst for this reaction at temperatures above ca. 300 °C and that the reaction favors deoxygenation (i.e., C-O bond-breaking) products at temperatures below ca. 400 °C and decarbonylation (i.e., C-C bond-breaking) products at temperatures above ca. 400 °C.

09 BIOMASS FUELS↗
DOE OSTI2020

Sensor Drift Estimation for Reactor Systems by Fusing Multiple Sensor Measurements

A nuclear power plant is instrumented with a variety of sensors that monitor its variables to estimate the state and initiate safety actions, if needed. We address the problem of estimating drifts or errors in sensor measurements due to factors such as calibration changes. We propose an information fusion method that uses measurements from other sensors to generate an estimate of a sensor measurement, and its difference from an actual measurement provides an error estimate. We present two fusers based on the ensemble of trees and support vector machine that are trained using sensor measurements collected at an emulated test loop of a pressurized water reactor under no-drift conditions. We present error estimates for a differential pressure sensor of the heat exchanger of the primary coolant system, under twenty controlled scenarios using the test loop. Both positive and negative errors are captured by both methods in scenarios involving calibration drifts, blocking, air gap and electromagnetic interference. The root mean square errors of estimated drifts are typically within 2% percent of the maximum.

Rao, Nageswara↗
DOE OSTI2017

A SUMMARY COMPARISON OF DESIGN EVALUATION TECHNIQUES

The United States’ fleet of Nuclear Reactors is considering modernizing their control rooms and instrumentation as an effort to avoid component obsolescence, keep up with competing industries, and cater to a new work force among others. Multiple technologies have potential to improve the control room. In attempts to evaluate the available options researchers employ a variety of measures to ensure the best candidate is used in a modernizing effort. The NRC is in charge of ensuring any new design has been thoroughly vetted before approving a license. Laid out in NUREG-0711 “Human Factors Engineering Program Review Model” is the review criteria for design validation. As there are a variety of measures currently used to evaluate candidate technologies, this paper seeks to identify weaknesses in the common measures used to design and validate technology in control room modernization efforts. Identifying where measurements are not meeting the criteria will help the control room modernization research platform identify where measurement development needs to occur. This will help ensure that current performance measurements are producing as reliable results as possible to select the right technology to integrate into nuclear operating control rooms.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

The study on overall I and C system scheme for floating nuclear power plant ACP100S and ACP25S - 189

Floating nuclear power plant (FNPP) is built on the offshore platform and it is defined as a kind of Small and Modular Reactor (SMR) due to its power level. The FNPP can provide power supply, fresh water and high temperature steam and other products. So it can be used as special energy sources, such as regional power supply, district heating and offshore oil exploitation. At present, two types of FNPP, named ACP100S and ACP25S, are developed by China National Nuclear Corporation (CNNC). In this paper, the overall technical parameters and overall technical scheme of ACP100S and ACP25S are introduced firstly. On the basis of these schemes, the different arrangements of nuclear power plant which are used for meeting the different energy demands are discussed. Due to the different configuration of nuclear power plant, the requirements for the Instrument and Control (I and C) system are different, so the paper discusses some overall configurations of I and C system under the several typical arrangements models, i.e., the single reactor and the single turbine, the double reactors and double turbines, and the double reactors and the single turbine. Furthermore, the characteristics of these overall I and C configurations are analyzed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Study of the open loop and closed loop oscillator techniques

This paper presents some aspects of a five year study undertaken at Idaho State University of the measurement of very small worth reactivity samples comparing open and closed loop oscillator techniques. The study conclusively demonstrated the equivalency of the two techniques with regard to uncertainties in reactivity values, i.e., limited by reactor noise. As those results are thoroughly documented in recent publications, in this paper we will concentrate on the support work that was necessary. For example, we describe in some detail the construction and calibration of a pilot rod for the closed loop system. We discuss the campaign to measure the required reactor parameters necessary for inverse-kinetics. Finally, we briefly discuss the transfer of the open loop technique to other reactor systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Experimental and numerical validation of an ultrasonic visualization system for nuclear reactor application

This paper deals with the design of ultrasonic imaging systems for the next generation of nuclear reactors cooled by liquid metal. Indeed, a generation IV research nuclear reactor is being developed by the Belgian Nuclear Research Center (SCK-CEN) in the frame of the MYRRHA project (for Multipurpose hYbrid Research Reactor for High-tech Applications). This sub-critical/critical reactor is cooled by Lead-Bismuth Eutectic (LBE). The opacity of this liquid metal requires the development of an ultrasonic visualization system for internal inspection and object detection. But due to the peculiar conditions met in the core of the reactor, velocity as well as temperature gradients are expected and are likely to affect directly the ultrasonic propagation. The objective of this work is to validate the ultrasonic imaging strategy by tests performed in a dedicated test rig and by numerical simulations using a ray-tracing method. The experimental investigations have been performed on a specific water facility reproducing conditions similar to those encountered in the core of the MYRRHA reactor. These conditions include the propagation over large distance, and the presence of temperature and velocity gradients. In the MYRRHA reactor application, the distance to be travelled by the acoustic waves of the visualization system is about 5 m, including a reflection. The acoustic absorption, the scattering losses, the beam divergence and the transmitted energy during reflection have been determined as a function of the travelled distance. The experimental values are compared with the literature for validation. The presence of temperature and velocity gradients in the core of the reactor is due to the coolant circulation. These gradients are about 5 K over 0.1 m and 1 m/s over 0.2 m, respectively, and are reproduced in the facility for investigating their influence on the propagation of ultrasounds. The experimental data are used for improving and validating a ray-tracing algorithm. This algorithm is then applied to the full scale application for validating the ultrasonic visualization system for nuclear reactor application. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

In service inspection and repair of sodium cooled ASTRID prototype

In the frame of the large R and D work which is performed for the future ASTRID sodium cooled prototype, In Service Inspection and Repair (ISI and R) has been identified as a major issue to be taken into account in order to enlarge the plant safety, to consolidate its availability and to protect the associated investment. After the first part of pre-conceptual design phase (2008-2012), the running second part of pre-conceptual phase (2013-2015) allows to increase the ISI and R tool ability for immersed sodium structures of ASTRID, at about 200 deg. C, on the basis of consolidated specifications and thanks to their qualification through more and more realistic laboratory tests and simulation with CIVA code. ISI and R items are being developed and qualified during a pluri-annual program which mainly deals with the reactor block structures, the primary components and circuit, and the Power Conversion System. It ensures a strong connection between the reactor designers and inspection specialists, as the optimization of inspectability and repairability is looked at: this already induced specific rules for design, in order to shorten and ease the ISI and R operations, which have been merged into RCC-MRx rules. In the frame of increasing technology readiness level with corresponding performance demonstration, this paper presents R and D dealing with the ISI and R items: it highlights the sensor development (both ultrasonic and electromagnetic concepts, compatible with sodium at 200 deg. C), then their applications for ASTRID structure control (under sodium telemetry, imaging and NDE). Activity for repair is also presented (a single laser tool for sodium sweeping, machining and welding), and finally the effort for associated robotic (generic program for ASTRID applications, specific technological tools for sodium medium, tight immersed bell). The main results of testing and simulation are given for telemetry, vision, NDE applications, laser process repair and under sodium sealing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of a diversity and defense-in-depth strategy for the CNNC Fuqing, Fangjiashan and Hainan nuclear plants - 379

The purpose of this paper is to provide an overview of the development of the Diversity and Defense-in- Depth (D3) strategy for the China National Nuclear Corporation Fuqing, Fangjiashan, and Hainan Nuclear Plants. The Nuclear Power Plant (NPP) Digital Control System (DCS) and advanced control room was completed by Invensys and China Nuclear Power Engineering (CNPE)/Nuclear Power Institute of China (NPIC) for these eight new Chinese built Pressurized Water Reactor plants. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

HFIR Core Modeling and Analysis for the PROSPECT Reactor Antineutrino Experiment

Nuclear reactors have been employed as a copious source of antineutrinos for over six decades. Research initially directed at the study of antineutrino physics has led to applications of these highly penetrating particles to reactor safeguards because the antineutrino flux and spectrum can be related to the state of the reactor core, including burnup evolution. One of the main challenges of antineutrino-based safeguards is operation of the detectors near surface (within the reactor containment or in the close proximity). Lack of overburden results in increased contribution to the detector noise from cosmic radiation. Similarly to safeguards, experiments to definitively confirm or exclude the existence of sterile neutrino oscillations require detector technologies capable of operating at short baselines (tens of meters) at the surface. The Precision Reactor Oscillation and Spectrum Experiment (PROSPECT) at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is an experiment with a main objective to make a precision measurement of the flux and energy spectrum of antineutrinos emitted from nuclear reactors in order to search for the oscillation signature of sterile neutrinos. Fissions of highly-enriched uranium of HFIR will serve as a source of antineutrinos. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Safety-Related Instrumentation & Control Pilot Upgrade Initiation Phase Implementation Report

This research report (1) describes the process followed and products developed during the SR I&C Pilot Project Initial Scoping Phase, and (2) captures lessons learned. Exelon Generation and LWRS collaborated to develop a Digital Transformation Strategy as part of a larger Advanced Concept of Operations. The proposed SR I&C Pilot Upgrade provides a foundation stone for this Digital Transformation that will improve plant safety, reliability, and operational performance while lowering plant Total Cost of Ownership (TCO). Initial Scoping Phase activities for this Pilot Project have been performed in accordance with industry processes that have been adapted to better support digital upgrades. These processes include IP-ENG-001, Standard Design Process (SDP) [Reference 2], NISP-EN-04, Standard Digital Engineering Process (SDEP) [Reference 3], and Electric Power Research Institute (EPRI) Report 3002011816, Digital Engineering Guide (DEG). Completing Initial Scoping Phase Engineering and Operations, Licensing, and Project Management Activities was necessary to sufficiently bound the scope, schedule, and estimated cost of the Project to enable utility management to authorize moving into the Conceptual Design Phase. A significant finding of the Business Case Analysis (BCA) methodology developed and applied as part of this effort was that the growth rate of material costs for sustaining the operation of obsolete SR I&C equipment is accelerating. This directly contributed to the Project Economic Analysis created to justify continuing the Project. Project Initial Scoping Phase lessons learned have also been captured to assist the larger industry in understanding the Digital Transformation Strategy and SR I&C Pilot Project Initial Scoping Phase efforts. This is in keeping with the public/private partnership that has been established between the Department of Energy (DOE) and Exelon for this effort with engagement from the NRC. By addressing first-of-a-kind (FOAK) risks and capturing lessons learned, the SR I&C Pilot Upgrade Project addresses technical, regulatory, and business risks to enable subsequent implementers of similar upgrades.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Safety-Related Instrumentation and Control Pilot Upgrade (Initial Scoping Phase Implementation Report)

This research report (1) describes the process followed and products developed during the SR I&C Pilot Project Initial Scoping Phase, and (2) captures lessons learned. Exelon Generation and LWRS collaborated to develop a Digital Transformation Strategy as part of a larger Advanced Concept of Operations. The proposed SR I&C Pilot Upgrade provides a foundation stone for this Digital Transformation that will improve plant safety, reliability, and operational performance while lowering plant Total Cost of Ownership (TCO). Initial Scoping Phase activities for this Pilot Project have been performed in accordance with industry processes that have been adapted to better support digital upgrades. These processes include IP-ENG-001, Standard Design Process (SDP), NISP-EN-04, Standard Digital Engineering Process (SDEP), and Electric Power Research Institute (EPRI) Report 3002011816, Digital Engineering Guide (DEG). Completing Initial Scoping Phase Engineering and Operations, Licensing, and Project Management Activities was necessary to sufficiently bound the scope, schedule, and estimated cost of the Project to enable utility management to authorize moving into the Conceptual Design Phase. A significant finding of the Business Case Analysis (BCA) methodology developed and applied as part of this effort was that the growth rate of material costs for sustaining the operation of obsolete SR I&C equipment is accelerating. This directly contributed to the Project Economic Analysis created to justify continuing the Project. Project Initial Scoping Phase lessons learned have also been captured to assist the larger industry in understanding the Digital Transformation Strategy and SR I&C Pilot Project Initial Scoping Phase efforts. This is in keeping with the public/private partnership that has been established between the Department of Energy (DOE) and Exelon for this effort with engagement from the NRC. By addressing first-of-a-kind (FOAK) risks and capturing lessons learned, the SR I&C Pilot Upgrade Project addresses technical, regulatory, and business risks to enable subsequent implementers of similar upgrade.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Formation of Aerosol Nanoparticles by Gas-Phase Hydrolysis Reaction of Uranium Hexafluoride

The aerosol physics of uranyl particle formation has been addressed in this research using advanced aerosol instrumentation and an aerosol dynamics model. Based on the research works, we conclude that the formation and growth of aerosol particles by gas-phase UF6 hydrolysis strongly depends on the availability of water molecules in our reactor conditions. The total number concentration of the UO 2 F 2 particulate material that could be produced in the hydrolysis reaction is also regulated primarily by the availability of water molecule concentration. The higher the water molecule concentration, the higher the number and the larger the size of UO 2 F 2 aerosol particles that could be produced in a reactor custom-built at ORNL. Although the aerosol reactor was enabling the study of particle formation kinetics, the instrumentation was still insufficient in characterizing the chemical composition of the produced particles as well as the time-dependent evolution of the particulate species. The temporal evolution could impact the eventual fate of the particles upon release to the environment (i.e., the physio-chemical transformation, transport, and removal). On uranyl particle formation kinetics, we found that the growth rates of aerosol particles appeared to approach a single number in the range of 0.05 ± 0.03 - 0.08 ± 0.04 nm/s, statistically, as the ω value becomes smaller than 1. The size of primary particles from the UF6 hydrolysis at water-deprived condition was estimated to be 3.6 ± 0.4 nm; the higher the availability of water molecules, the larger the primary particles. The ability to precisely control the availability of water molecules in the reaction could lead to the production of nearly monodispersed aerosol particles. In other words, the result suggests that one can precisely manipulate the size of UO 2 F 2 aerosol particles by controlling the water vapor availability and interaction of water molecules with U F6 in the reaction. This finding has significant implications in the engineering manufacturing of fuel powder materials and possibly to future development and deployment of an environmental sampling apparatus.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2023

Intern Poster Session

This poster has been developed for display during the 2023 Intern Poster Session. The poster is apart of the Nuclear Research and Experiments category. The focus of this poster is on the modifications made to the design of an integral effect test (IET) facility currently being designed for a new Small Modular Reactor (SMR) design. My contributions made to the IET facility are laid out along with an explanation on some of the aspects of designing a new nuclear reactor. This poster covers the ideas and importance of scaling and IET facilities with the context of the standards set by the NRC. There are also graphics of some other examples of IET facilities developed for other reactor designs. My contributions include designing the piping and instrumentation layout of the IET facility using CAD software. This is a key step in the process of designing this IET facility and needs to be designed to avoid excessive head loss.

42 ENGINEERING↗
DOE OSTI2017

Equipment for subcriticality monitoring by pulse method - 240

At the present time methods and means of determining, evaluating, and monitoring the effective neutron multiplication factor or the subcriticality are becoming very important for validating and monitoring nuclear safety in connection with the accumulation of a large number of spent fuel assemblies and the transition to a more tight mode of their storage arrangement in the spent-fuel pools of NPP. Pulse method is an effective subcriticality monitoring tool for subcritical reactors and spent nuclear fuel storages. This method is based on the use neutron generator, fission chambers and instrumentation for data acquisition and processing. The measurement and control equipment was designed. It allows: neutron pulses generation with specified parameters; neutron flux response monitoring by measuring the count rate at predetermined set of sequential time intervals; data transmitting to the personal computer; data processing and subcriticality calculation; self diagnostics and health monitoring; data recording, archiving and storage. The special device is proposed to check the electronic equipment. This device simulates a signal of a neutron detector counting rate after neutron generator pulses appear. The paper presents hardware structure and composition, specification and software of the control unit, which provides neutron generator operation, data acquisition and processing. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Summary of Technical Peer Review on the Risk Assessment Framework proposed in Report INL/RPT-22-68656 for Digital Instrumentation and Control Systems

This report summarizes the peer review activities initiated by Idaho National Laboratory (INL) during fiscal year (FY) 2023 for the evaluation and improvement of the methodology developed under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a technical basis to support effective and secure DI&C technologies for digital upgrades/designs. A framework was proposed for this strategy, which aims to (1) provide a best-estimate, risk-informed capability to quantitatively and accurately estimate the risk impact of plant modernization, considering the introduction of high safety-significant safety-related (HSSSR) DI&C systems, (2) support and supplement existing risk-informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems, (4) assure the long-term safety and reliability of HSSSR DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. The R&D efforts of this project from FY 2019 through FY 2022 were focused on methodology improvement and demonstration of the proposed framework for the risk assessment and design optimization of safety-critical DI&C systems. Collaborations with the nuclear industry have been initiated to support the reliability and risk assessment of their DI&C systems by using the proposed framework. In FY 2023, the framework has reached to a point for a technical peer review and obtain stakeholder feedback. This peer review activity includes coordination of the reviews performed by a group of industry stakeholders, documentation of the peer review suggestions, providing resolutions and responses to the peer review comments. The objective of this technical peer review is to obtain representative feedback on the proposed framework to improve the technical qualities of its methodology and readiness for deployment to the industry. Feedback may identify potential areas for improvement and further development. The Subject Matter experts were invited to review the latest project report documenting the methodology developed in the project and provide evaluations of the technical qualities of the proposed framework and relevant methods. The reviewed project report is “An Integrated Framework for Risk Assessment of High Safety-significant Safety-related Digital Instrumentation and Control Systems in Nuclear Power Plants: Methodology and Demonstration” INL/RPT-22-68656 (short as “INL/RPT-22-68656” in this report). This peer review report documents the technical questions provided for technical peer review and introduces the technical peer reviewers from the stakeholders including nuclear utilities, regulators, and universities. Comments from technical peer reviewers and the resolutions and responses to these comments are outlined. Insights and lessons learned from the technical peer review are summarized in conclusions and future work. The primary audience of this report are DI&C designers, engineers, and probabilistic risk assessment (PRA) practitioners. This includes stakeholders, such as the nuclear utilities and regulators who consider the deployment and upgrade of DI&C systems, DI&C software developers and reviewers, and cybersecurity specialists.

99 GENERAL AND MISCELLANEOUS↗