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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 721 records · Page 40
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

RNA-guided transcriptional activation via CRISPR/dCas9 mimics overexpression phenotypes in Arabidopsis

Clustered regularly interspaced short palindromic repeats (CRISPR) and the CRISPR associated protein 9 (Cas9) system allows effective gene modification through RNA-guided DNA targeting. The Cas9 has undergone a series of functional alterations from the original active endonuclease to partially or completely deactivated Cas9. The catalytically deactivated Cas9 (dCas9) offers a platform to regulate transcriptional expression with the addition of activator or repressor domains. We redesigned a CRISPR/Cas9 activation system by adding the p65 transactivating subunit of NF-kappa B and a heat-shock factor 1 (HSF) activation domain to dCas9 bound with the VP64 (tetramer of VP16) activation domain for application in plants. The redesigned CRISPR/Cas9 activation system was tested in Arabidopsis to increase endogenous transcriptional levels of production of anthocyanin pigment 1 (PAP1) and Arabidopsis thaliana vacuolar H+ -pyrophosphatase (AVP1). The expression of PAP1 was increased two- to three-fold and the activated plants exhibited purple leaves similar to that of PAP1 overexpressors. The AVP1 gene expression was increased two- to five-fold in transgenic plants. In comparison to the wild type, AVP1 activated plants had increased leaf numbers, larger single-leaf areas and improved tolerance to drought stress. The AVP1 activated plants showed similar phenotypes to AVP1 overexpressors. Therefore, the redesigned CRISPR/Cas9 activation system containing modified p65-HSF provides a simple approach for producing activated plants by upregulating endogenous transcriptional levels.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2019

Interface-driven mechanisms in cubic/noncubic nanolaminates at different scales

Superior structural properties of materials are generally desired in harsh environments, such as elevated temperatures, high rates of impact, and radiation. Composite nanolaminates, built with alternating stacks of crystalline layers, each with nanoscale individual thickness, are proving to exhibit many of these target properties. In principle, the nanolaminate concept can be applied to any twophase, bimetallic system; however, for a number of reasons, they have been limited to combinations of metals with a cubic crystal structure. There is growing demand to increase the number of advanced material systems containing noncubic metals, since these metals bear several desirable intrinsic properties. In this article, we cover recent modeling and experimental efforts to understand the complexity in structure, mechanisms, and behavior of noncubic/cubic nanolaminates. We hope this article will facilitate and encourage future studies in this promising area.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Structuring, stochastic behavior, and charge storage capacity of redox-active microemulsions formulated with mixtures of toluene and ionic liquid as oil phase

Oil/water microemulsions (µEs) are promising electrolytes for redox flow batteries (RFBs) as they simultaneously improve charge capacity and ionic conductivity. Here, we report the successful formulation of bicontinuous µEs where the oil phase is a solution of trihexyltetradecylphosphonium bis(trifluoromethylsulfonyl)amide ionic liquid in toluene with redox-active ferrocene. We examined the effect of supporting electrolyte anion (NO 3 – , Cl – , ClO 4 – ) on the structure, reactivity, transport, and electrolytic performance of redox-active µEs. Neutron scattering and nuclear magnetic resonance showed that the domain size increased as Cl – > NO 3 – > ClO 4 – while the ferrocene diffusion coefficient increased as NO 3 – > Cl – > ClO 4 – . Scanning electrochemical microscopy indicated anion-dependent current fluctuations during electrolysis, with ClO 4 – exhibiting the least high-frequency oscillations, which correlate to the highest charge and discharge capacity and reversibility. All ionic liquid containing systems improved the performance of toluene-based µEs, highlighting new design principles for these electrolytes.

diffusion↗
DOE OSTI2017

Classification of the Z-Pinch Waste Stream as Low-Level Waste for Disposal

The purpose of this document is to describe the waste stream from Z-Pinch Residual Waste Project that due to worker safety concerns and operational efficiency is a candidate for blending Transuranic and low level waste together and can be safely packaged as low-level waste consistent with DOE Order 435.1 requirements and NRC guidance 10 CFR 61.42. This waste stream consists of the Pu-ICE post-shot containment systems, including plutonium targets, generated from the Z Machine experiments requested by LANL and conducted by SNL/NM. In the past, this TRU waste was shipped back to LANL after Sandia sends the TRU data package to LANL to certify the characterization (by CCP), transport and disposition at WIPP (CBFO) per LANL MOU-0066. The Low Level Waste is managed, characterized, shipped and disposed of at NNSS by SNL/NM per Sandia MOU # 11-S-560.

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

Comparison Of A Neutron Kinetics Parameter For A Polyethylene Moderated Highly Enriched Uranium System

This paper examines the comparison of MCNP® code’s capability to calculate kinetics parameters effectively for a thermal system containing highly enriched uranium (HEU). The Rossi-α parameter was chosen for this examination because it is relatively easy to measure as well as easy to calculate using MCNP®’s kopts card. The Rossi-α also incorporates many other parameters of interest in nuclear kinetics most of which are more difficult to precisely measure. The comparison looks at two different nuclear data libraries for comparison to the experimental data. These libraries are ENDF/BVI (.66c) and ENDF/BVII (.80c).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Iodobutane Deep Bed Adsorption Test Report

The Department of Energy (DOE) Nuclear Technology Research and Development (NTRD) Program Material Recovery and Waste Form Development (MRWFD) Campaign Off-gas Sigma Team has supported research and development on iodine control and iodine waste forms for the past several years. As a part of this on-going research, two deep-bed iodobutane adsorption tests have been performed with higher NO x levels. High (>1,000) iodine decontamination factors (DFs, the ratios of the inlet iodine concentrations divided by the outlet iodine concentrations, when the gas flowrate is constant) were achieved with both the silver aerogel and silver zeolite sorbents. The iodine adsorption mass transfer zone (the depth of the sorbent bed in which adsorption is occurring at any specific time) for these tests was estimated at 4 inches for both sorbents under the test conditions. Silver utilization for silver aerogel sorbent reached 41% (up to 12 g adsorbed iodine per 100 g silver aerogel sorbent), while silver utilization reached 43% (up to 6.8 g adsorbed iodine per 100 g silver zeolite sorbent). These values are generally consistent with results for prior methyl iodide adsorption tests. The results also indicate that, during iodine adsorption process, the organic iodide molecule decomposes so that the iodine can be adsorbed. Even when sorbent breakthrough is reached, the organic iodide molecule continues to decompose; and the unadsorbed iodine that passes through bed is no longer in the form of the original organic iodide; but in the form of iodine compounds that are soluble in NaOH solutions, such as HI or I 2 . This research and development has included iodine adsorption tests using a laboratory-scale test system containing segmented fixed beds of iodine adsorbents in series. Non-radioactive synthetic gas mixtures were blended from air, nitrogen, NO, NO 2 , water, and the target iodine compounds to represent dissolver off-gas (DOG) streams from used nuclear fuel aqueous reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Crack Growth Rate and Large Plate Demonstration of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report describes the body of work involving testing and analyses performed in FY18 and FY19 at the Savannah River National Laboratory and at Korea University to determine chloride-induced stress corrosion crack (CISCC) growth rates in weldments of austenitic stainless steel plate material. The primary work involved: several separate crack growth rate test campaigns with new testing setups using laboratory-scale corrosion specimens; analyses to estimate the stress intensity factors for flaws postulated to be present in the residual stress fields of axial and circumferential weldments in a prototypic dry storage canister; and establishment of a “large plate” (approximately 51×46 cm with thickness 16 mm and weight about 30 kg) CISCC test that uses a plate with a circumferential weldment harvested from the Sandia National Laboratories’ full-size mockup dry storage canister. The tests for crack growth rate were conducted on ASTM E1861 bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse the brine to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The crack growth rate test results matched previous experimental data in the open literature. These literature data had been compiled by SNL and EPRI to develop a temperature-dependent (K-insensitive) crack growth rate model to support flaw disposition in the ASME Boiler and Pressure Vessel Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.” The knowledge of the salt, temperature, humidity conditions for CISCC including the threshold stress intensity factor, $K_{ISCC}$, were used to setup the large plate test in which a set of partthrough- wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a constant 75% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC. The visual results at the surface of the plate show corrosion product staining (brown rust appearance) but with no CISCC emanating from the machined notches after approximately 5 months exposure to date. The plate test will continue in FY20 and phased-array ultrasonic testing and other NDE techniques will be applied to identify and characterize CISCC that may emanate from the machined defects. The notches and any CISCC from them will be sectioned at the conclusion of the test to verify NDE characterization, and to develop effective through-wall crack growth rates from cracks that would likely be branched.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Evaluating Corrosion Effects on the Stainless Steel Components of the SAVY-4000/Hagan Nuclear Material Storage Containers: FY19 Update

This report provides an update of the FY19 corrosion observations, laboratory studies, and mitigation efforts for the nuclear material storage program at Los Alamos National Laboratory. Nuclear material packages outside of an approved engineered contamination barrier must meet packaging, surveillance, and testing requirements designed to protect workers from airborne contamination per Department of Energy (DOE) Manual 441.1-1, Nuclear Material Packaging.1 The SAVY-4000 containers were developed at Los Alamos National Laboratory (LANL) in conjunction with Nuclear Filter Technology as a manually-compliant container system for staging and storage of plutonium for LANL and the DOE complex.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Molten Salt Reactor Experiment Benchmark Evaluation

As one of the six advanced reactor concepts selected by the 2002 Generation IV roadmap as a technology meriting future research, Molten Salt Reactors (MSRs) attracted broad attention and multiple private and public entities are working towards its commercialization. Facing stringent regulatory requirements, validations of computational codes used to calculate and prove the safety characteristics of MSRs play a key role. This project developed the world-first, MSR-related reactor physics benchmark based on a series of zero-power experiments conducted at the Molten Salt Reactor Experiment (MSRE). The benchmark was successfully reviewed by the International Reactor Physics Experiment Evaluation Project (IRPhEP) committee and is available in the IRPhEP handbook starting form the 2019 edition. To benchmark experimental data from the MSRE, a three-dimensional high-fidelity model was developed using the Monte Carlo neutron transport code Serpent 2 and ad-hoc methods have been implemented to account for the unique feature of fuel salt motion. The calculated effective multiplication factor, k eff , for the first criticality experiment, when 235 U was progressively added to the fuel salt in order to achieve criticality with stationary salt and isothermal conditions, was 1.02132 (±3 pcm). The total uncertainty for experimental k eff was estimated to be 420 pcm. The calculated k eff is 2.154% larger than the experimental and benchmark model value, which is approximately five times the benchmark model uncertainty. It is to be noted that, for systems containing large volume of graphite (or other carbonaceous materials), Monte Carlo codes tend to overestimate the keff of the benchmark model by 1% to 2%. The bias is possibly due to uncertainties in the impurity content of the graphite blocks, uncertainty in the neutron capture cross section of carbon or uncertainty in the properties of nuclear-grade graphite. The calculated reactivity coefficient of 235 U concentration (0.2228 ± 0.0014, represented as the change of reactivity over the relative change of 235 U mass in loop) matches well with the experimental value (0.223 ± 0.006), strengthening the confidence on the accurate representation of the fuel salt composition in the MSRE benchmark. Additional reactivity effect calculations, including the control rod bank worth, reactivity effects of fuel circulation and isothermal and fuel temperature coefficients show good agreement with the experimental values (within 1σ) as well.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Biodefense Knowledge Management System

After the Amerithrax attacks, the Department of Homeland Security (DHS) recognized the need to link technical scientific information with law enforcement and national security information in the fight against bioterrorism. The Biodefense Knowledge Center (BKC) was formed to provide timely analysis of technical biodefense-related topics to support DHS and its customers in the interagency. The Biodefense Knowledge Management System (BKMS) was created as the backbone for information management within the BKC. It is a scalable, web-based knowledge management system for organizing disparate data sources. By combining information from 35+ data sources, the BKMS provides a searchable repository for technical information including scientific reports, journal articles, news articles, research funding databases, patent databases, and genomic datasets. The current system contains over 58 million documents and 450 million genomic records, and operates at several classification levels.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2020

An Evaluation of a Neutron Time Correlated Interrogation Method for Measurement of Fissile Content in Research Reactor Spent Fuel Assemblies

There are many research reactors worldwide that have been, or are being converted from the use of high enrichment uranium (HEU) to low enrichment uranium (LEU, < 20% enriched). The verification of fissile content and initial enrichment of the spent fuel is needed for the effective safeguards of the fuel. The advanced experimental fuel counter (AEFC) was developed for the measurement of spent fuel rods and assemblies from research reactors for safeguards verification. This measurement system contains components for active neutron interrogation, passive neutron totals counting, neutron coincidence counting, and gross gamma-ray counting. This report presents the first application of the time correlated interrogation technique for the measurement of the 235 U content in research reactor spent fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252 Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by coincidence counting. The doubles rates are enhanced by having the neutron trigger events from both the 252 Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission of the 252 Cf source is 3.76 and the induced fission neutrons for 235 U is 2.44, so the number of neutrons that are produced is higher than for random neutron interrogation. This high effective neutron number increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252 Cf are reduced by the water in the sample cavity and the polyethylene surrounding the 3 He detector tubes. This method of active neutron interrogation has been applied to the measurement of spent research reactor (IRT) fuel assemblies. The advanced experimental fuel counter (AEFC) was used to compare the TCIF method with the typically used AmLi neutron interrogation source that emits neutrons that are random in time. The statistical uncertainty for the use of the random neutron source (AmLi) and the time correlated source ( 252 Cf) for spent fuel interrogations was evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

First principles calculations in support of Pu aging: calculating the effects of lattice imperfections on thermodynamics

Plutonium (Pu) has, in theory, well defined crystal structures: its atoms are arranged in regular spatial patterns. But Pu is radioactive, and as its nuclei decay those regular spatial patterns are interrupted. The interruptions are lattice imperfections, which are known to affect how materials respond to their environment. To adequately model Pu, we need to know which lattice imperfections are present, how they interact with each other, and how they affect the material’s response to its environment. The work presented here aims to use density functional theory (DFT) calculations to begin to answer the latter, in particular, how individual lattice imperfections affect measurable effects including thermal expansion (the change in volume in response to a change in temperature), heat capacity (the amount of thermal energy needed to change a material’s temperature), and elastic moduli (a material’s resistance to applied stresses). Pu poses many computational challenges. The $\textit{f}$ electrons require special attention. Of all the elements, Pu has the largest number of electrons that must be included in the calculations. Thermal effects require calculating the phonons (the lattice vibrations), which for systems containing lattice imperfections demand large, complex computational cells - but computational resources limit the size and complexity. With careful restructuring of how the calculations are performed, all these challenges have been met to enable calculations that provide insight into how lattice imperfections affect Pu’s response to its environment. Reported here are the computational challenges and the advances developed to meet them, along with the first results showing the strong effect that one prototype lattice imperfection (an interstitial Pu atom in a delta-phase Pu lattice) has on Pu’s response to its environment.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Study of Ionic Mass Transport in Non-Conventional Electrolytes for Energy Storage and Carbon Capture Applications

In the field of energy storage, electrolytes are the materials or liquids used to transport ions or redox active chemicals to facilitate an electrochemical process. Conventional electrolytes are the most common form of electrolyte used and consist of salts dissolved in a liquid solvent medium. While conventional electrolytes are generally inexpensive and work well with lithium-ion batteries, they have many shortfalls however, such as corrosion of the cell, flammability, toxicity, price, and incompatibility with other metals. This project investigates the fundamental properties in nonconventional electrolyte systems with an initial focus on eutectics being made. Using electrochemical methods ( Cyclic Voltammetry (CV), Impedance, Conductance, etc.), we are studying the study of electrolyte systems containing ions used for thermal, chemical, or electrochemical energy storage such as Li + , Mg 2+ , Na + , K + , Al 3+ , CO 3 2- , Cl - , and O 2- . This research looks at how ion mobility effects the electrochemical characteristics and properties in green energy applications.

25 ENERGY STORAGE↗
DOE OSTI2021

Measurement of radiolytic hydrogen generation and impact of drying treatments on reactor exposed and surrogate aluminum materials

Technical challenges associated with dry storage of Aluminum-clad Spent Nuclear Fuel (ASNF) include a need to better understand the potential extent of gaseous molecular hydrogen (H 2 ) production through radiolytic degradation of the aluminum (oxy)hydroxide films present on ASNF. The characterization of radiolytic gas generation from ASNF (oxy)hydroxide layers has been identified as a key knowledge gap which poses a technical challenge to the long-term storage of ASNF. Task 2 of the action plan for the extended (>50 years) dry storage of ASNF addresses this gap. Previous radiolysis studies performed under Task 2 established baseline estimates of H 2 generation rates from the attendant hydrated oxides. The results confirmed that net radiolytic H 2 production has a dependency on absorbed gamma dose, as well as relative humidity and cover gas composition (air, nitrogen, and argon). Further experiments revealed that the physisorbed water on the samples may significantly impact the radiolytic H 2 yield. This phenomenon complicates the determination of H 2 generation rates for hydrated oxides, which may lead to inaccurate modeling predictions of the long-term H 2 yields in sealed storage systems containing ASNF, particularly when compared to spent fuel casks which have undergone some drying process intended to remove physically- and chemically-bound water. This report describes the testing methods utilized and the hydrogen generation results obtained in an investigation of the effects of gamma irradiation on aluminum materials for a variety of (oxy)hydroxide surface compositions and drying conditions. The testing methods included small-area aluminum material testing in ampules, and large-area aluminum material testing in steel vessels. Test material preparation, irradiation, and radiolytic H 2 measurement methods are summarized. The measured H 2 concentrations, which reflect the variable initial hydrated inventory and drying treatments, are compared to one another as well as to previously published data, to identify the primary factors affecting radiolytic H 2 generation rates and potential equilibrium H 2 concentrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Radiation Effects on Line Heat Detection Cable AP Hot Cells Fire Detection at Brookhaven National Laboratory

Hot cells are used to safely shield workers from high radiation environments while allowing a process to take place within the cell's shielded enclosure, often with the assistance of remote manipulators. The interior of a cell may not be accessible by personnel for years at a time, either due to the lingering radiation environment or due to the constant use of these specialized pieces of equipment. High radiation levels are known to cause undesirable effects and changes to different materials in varying degrees. Selecting equipment for a radiation environment requires an evaluation of materials used (preferably completed assembles), to ensure all interactions of materials are known and ensure that the equipment will have a high degree of reliability and functionality over the intended life of the facility. Hot cells contain systems to support the primary purpose of the enclosure. Fire detection is one such support system. Safety of hot cell operations often establish limits on the amounts of combustibles with the cells, but the presence of some combustible materials is unavoidable. Fire detection is provided to quickly detect a fire and summon manual intervention before damage is too great. Detection may also be used to initiate fire suppression systems. As a safety system, fire detection needs to remain functional to protect the facility throughout the facility's life. Equally important is that the false alarms from the fire detection are minimized to retain confidence in the alarm system and avoid unnecessary interruptions in the hot cell operations. There are many types of fire detection available with various pros and cons. Not all types of fire detection are suited for hot cell application.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Development of High-Temperature Bonding Techniques to Enable High-Temperature Static or Dynamic Strain Measurements

Current light-water nuclear reactors rely on a variety of different sensors and sensor applications to meet their structural health monitoring needs throughout the entirety of the reactor primary, secondary, and containment systems. Optical fiber–based sensor technologies could provide solutions to reduce the sensor system footprint while enhancing the measurement fidelity and spatial resolution by leveraging distributed monitoring techniques. Moreover, advanced reactors may require optical fiber–based sensors for structural health monitoring because their operating temperatures will exceed the limits of conventional transducers used to acquire dynamic strain or acoustic data in nuclear power plants. Therefore, this report describes experiments targeting the development of high-temperature bonding techniques that would allow for potentially long lengths of fibers to be bonded to metallic reactor components in advanced reactor systems. The high temperatures experienced within target application, next-generation nuclear reactors, necessitate a high-temperature resistant bond to limit the amount of tension on the fiber at the target application temperature. The primary bonding method investigated in this work is brazing; hot-rolling has also been investigated to a lesser extent. Both techniques are well-suited to bonding optical fibers to large reactor components such as primary coolant piping, pressure vessels, or heat exchangers. Optical frequency domain reflectometry was used to monitor the strain in metal-coated optical fibers before, during, and after the high-temperature bonding process. On select optical fibers that were successfully bonded, additional thermal cycling was performed to assess the extent to which the fiber remained bonded based on the expected thermal expansion of the test specimen material. The results of the various experiments yielded the following general conclusions: (1) brazing is a viable technique for bonding and allows significant compressive strain to be applied to the fiber at room temperature; (2) hot-rolling is a viable technique as well, which has been more optimized than the brazing technique for bonding, but less residual compressive strain has been observed with this technique; and (3) both techniques will need further development and optimization to demonstrate bonding of a long length of fiber that can provably operate at relevant temperatures for an advanced nuclear reactor application.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Organic Iodine Capture from Vessel Off-gas

The Department of Energy (DOE) Nuclear Fuel Cycle and Supply Chain Program Material Recovery and Waste Form Development (MRWFD) Campaign Off-gas Sigma Team has performed research and development on iodine control and iodine waste forms for the past several years. This research and development has included iodine adsorption tests using a laboratory-scale test system containing multiple-segmented fixed beds of iodine adsorbents. Iodine capture performance has been tested using non-radioactive synthetic gas mixtures blended from air, nitrogen, NO, NO2, water, and the target iodine compounds diatomic iodine (I2), representing likely inorganic iodides, and methyl iodide (iodomethane, CH3I) and 1-iodobutane (butyl iodide) a surrogates for potential organic iodides, that could exist in dissolver off-gas (DOG) and vessel off-gas (VOG) streams from used nuclear fuel aqueous reprocessing.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2026

Design and Prototyping of the 162.5 MHz PIP-II RF Reference-Line Station

The Proton Improvement Plan II (PIP-II) Reference Line distributes phase-stable radio-frequency signals so accelerating cavities transfer energy to the proton beam at the correct point in each RF cycle. The full system contains 162.5, 325, and 650 MHz sections; this project focuses on the 162.5 MHz station. Commercial filters, amplifiers, splitters, mixers, and hardware were assembled into a compact, serviceable heat-plate layout and tested to verify performance metrics at 162.5 MHz.

Subedi, Harsheet [Unlisted, US]↗