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DOE OSTI2018

Spectral Decomposition and Other Seismic Attributes for Gas Hydrate Prospecting

Studying the sediments at the base of gas hydrate stability is ideal for determining the seismic response to gas hydrate saturation. First, assuming gas migration to the shallow section, this area is more likely to have concentrated gas hydrate because it encompasses the zone in which upward moving buoyant gas transitions to form immobile gas hydrate deposits. Second, this zone is interesting because these areas have the potential to show a hydrate filled zone and a gas filled zone within the same sediments. Third, the fundamental measurement within seismic data is impedance contrasts between velocity*density layers. High saturation gas hydrates and free gas inhabit opposite ends of these measurements making the study of this zone ideal for investigating the seismic characteristics of gas hydrate and, hence, the investigation of other seismic attributes that may indicate gas hydrate fill.

gas hydrate↗

DOE OSTI2019

Analysis of Cygnus Electrical Signals

The United States initiated the Subcritical Experiment (SCE) program to support a stockpile stewardship mission shortly after the 1992 moratorium on underground nuclear testing was established. Many SCE’s are conducted at the Nevada National Security Site (NNSS) in Nevada. Cygnus is a high-energy radiation generating device (RGD) located and operated at the NNSS and is a primary diagnostic for the SCE program. The Cygnus Dual Beam Radiographic Facility consists of two identical radiographic sources, Cygnus 1 and Cygnus 2. From creation of the high power V-I drive to energy transport and X-ray conversion at the rod-pinch diode, the Cygnus machines utilize the following components: oil-filled Marx generator, water-filled pulse-forming line (PFL), water-filled coaxial transmission line (CTL), three-cell vacuum induction voltage adder (IVA), and rod-pinch diode. The diode pulse has the following electrical specifications: 2.25 MV, 60 kA, 60 ns. Each source has the following X-ray specifications: 1 mm diameter, 4 rad at 1 m, and 50 ns radiation pulse. SCEs are both single-event and high-value; therefore, a high level of performance in reliability and reproducibility are key issues of Cygnus. Prior to executing such a SCE, there are a formidable number of shots (e.g., each of the two Cygnus RGDs charging and discharging properly into the rod-pinch diode load) that must be executed to determine reliability and reproducibility of the Cygnus RGDs. For every shot on Cygnus, voltages and currents along the machine are recorded and analyzed. In this paper we summarize attributes of the voltage and current waveforms at different locations using distribution plots. These distribution plots are used to quantify the reliability and reproducibility for Cygnus. This analysis methodology will provide a standard to trouble shoot results of individual Cygnus shots, and also to monitor long term Cygnus performance. Accordingly, this type of extensive analysis will contribute to optimized Cygnus performance on SCEs.

42 ENGINEERING↗

DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW PPT

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Nuclear thermal propulsion fuel element packing device

A packing device for filling fuel elements with a powder through a fill aperture in an outer shell of the fuel element, including a stationary base, a clamp assembly including a body defining a bore therein, the bore being configured to slidably receive a fuel element therein, wherein the clamp assembly is movable along a vertical axis with respect to the stationary base, a cam assembly including a cam and a drive motor configured to rotate the cam, wherein rotation of the cam alternatingly raises the clamp assembly up along the vertical axis and subsequently drops the clamping assembly, and a powder reservoir assembly including a powder reservoir and a fill needle in fluid communication with the powder reservoir.

Lake, Russell L.↗

DOE OSTI2022

Systems and methods for electrical energy storage

The present disclosure relates to an electrical energy storage apparatus which forms an interpenetrating, three dimensional structure. The structure may have a first non-planar channel filled with an anode material to form an anode, and a second non-planar channel adjacent the first non-planar channel filled with a cathode material to form a cathode. A third non-planar channel may be formed adjacent the first and second non-planar channels and filled with an electrolyte. The first, second and third channels are formed so as to be interpenetrating and form a spatially dense, three dimensional structure. A first current collector is in communication with the first non-planar channel and forms a first electrode, while a second current collector is in communication with the second non-planar channel and forms a second electrode. A separator layers separates the current collectors.

Duoss, Eric B.↗

DOE OSTI2022

Start-up and Emergency Shutdown Modeling for a Coal-fired 10MWe sCO₂ Power Plant

As part of a Phase II Front End Engineering Design (FEED) study for the DOE Fossil Fuel Large Scale Pilots program, investigations of novel start-up and emergency shutdown methods for a stoker-fed, coal-fired, 10MWe sCO₂ power plant proposal were performed. These investigations were performed using 1D transient system models of the plant created within the GT-SUITE system modeling platform. The system model components were created using vendor quotation data (heat exchangers), performance maps (turbomachines), or FMU models (fired heater). Previously, Echogen Power Systems (Echogen) has used a sCO₂ power cycle start-up method that fills the system with liquid CO₂ to facilitate easy use of an electrically-driven start pump to transition the system from initial CO₂ fill to turbo-compressor initialization. The novel start-up method presented attempts this transition with a minimal filling of the system with liquid CO₂, to reduce the total CO₂ inventory required. This method leaves the start pump vulnerable to a two phase inlet condition as system pressure is below the CO₂ saturation pressure for the condenser cooling water temperature (Tcw). System model cases at Tcw of 18°C, 26°C, and 32°C, corresponding to cold, design, and hot ambient days, were analyzed to determine if the SP inlet condition could be kept as a subcooled liquid and estimate the CO₂ inventory reduction amount./p> Stoker-fed, coal-fired heaters continue to emit heat for some time even after emergency shutdown from events such as a power failure. This heat emission would lead to a failure of the fired heater, as the metal overheats, if the CO₂ flow is shut off. The emergency shutdown method presented utilizes CO₂ vented by a controllable vent valve (CRV) to provide CO₂ cooling flow to the fired heater. To determine the effectiveness of this method at keeping the fired heater peak tubing metal temperature below the ASME material temperature limit, for pressures below 5 MPa, of 816 °C, system model cases with CRV diameters ranging from 4” to 10”, and an alternative CO₂ vent routing with CRV size of 6”, were investigated./p>

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2023

Lessons Learned From Ventilation and Glovebox Flooding Via Overfilling of the Wet Vacuum System in a Plutonium Facility

In the summer of 2021, operations personnel requested maintenance personnel to fill a wet vacuum (Wetvac) seal water tank in a plutonium facility at Los Alamos National Laboratory (LANL). After the filling was finished, the fill valve of the Wetvac seal water tank was left open and a ‘dead-man’ valve on an isolation line failed to close properly. After a series of other failures, the water exited the glovebox and contaminated three rooms of the first floor of the facility and an underlying area in the basement. A retrospective analysis of this event determined that there was no risk of a criticality due to the inherent design features of the ventilation system. This paper will explain the driving factors behind how this event was able to happen, as well as some of the major lessons learned that can be universally applied to other nuclear facilities. One of the main causes of the event was the delegation of work to maintenance personnel that were not qualified for the work being asked of them. Additionally, there was a lack of coordination between criticality safety engineers and facility operations equipment owners regarding equipment and connections present on the seal water tanks for the Wetvac, specifically the presence of overflow drain lines. The respective criticality safety evaluation credited an overflow drain line being present on the seal water tanks that was later found to have not been present. Lastly, there are lessons to be learned regarding equipment configuration and designing systems with safety in mind.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Decommissioning of a facility in a historical reactor at SRS: achieving both historical significance preservation and effective in-situ decommissioning - 15275

The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions, LLC (SRNS). In 2004, US DOE recognized SRS as a structure within the Cold War Historic District of national, state and local significance because it is composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. US DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. The Disassembly Basin (DB) in C Reactor was evaluated for action under CERCLA. Data indicated there were over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 million (M) liters (L) of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. The need for an action was identified to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. In-situ decommissioning was selected as the remedy. The remedial action proceeded in stages: -Above-grade modifications to the interior of DB building -Relocation of irradiated scrap and reactor components to the deepest segments of the DB - Construction of a metal building to house evaporators - Forced evaporation of the DB's radiologically contaminated water - Grouting with special hybrid grouts - Removal of the evaporators and temporary building. Since the building itself was not demolished and will remain unchanged in appearance with the exception of filling the DB with grout, this ISD removal action/methodology did not adversely impact the historical significance of the 105-C Reactor Building, in accordance with the National Historic Preservation Act, nor did it adversely affect the ability to determine the full extent and details of preservation at some later date. Both special hybrid grout materials met all requirements. Overall average 7-day compressive strength for the dry area mix was 1.54 MPa (223 psi) and for the underwater mix 2.94 MPa (427 psi).Overall 28-day compressive strength for the dry area mix was 2.88 MPa (417 psi) and for the underwater mix 4.99 MPa (724 psi). A limited amount of testing was also performed at 91 days, where the overall compressive strength for the dry area mix was found to be 9.40 MPa (1363 psi) and for the underwater mix 13.98 MPa (2028 psi). All compressive strength values far exceeded the 0.34 MPa (50 psi) minimum requirements. The special, flowable-fill grouts designed and used for P- and R-Reactor ISD and used again for this 105-C ISD removal action resulted in considerable labor, cost, and schedule savings versus conventional materials. As an added bonus/benefit, the CO{sub 2} footprint of the SRS C-Reactor DB ISD was minimized by using a small amount of cement and byproduct material to produce the structural fill materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Engineering scale demonstration of a prospective cast Stone process - 15584

A cementitious waste form is one of the alternatives being evaluated for Supplemental Immobilization of Hanford Low Activity Waste (LAW). Washington River Protection Solutions (WRPS) is collecting data on cementitious or pozzolanic waste forms such as Cast Stone. The goal of this project for WRPS is to obtain data on the performance of the Cast Stone waste form for immobilizing LAW. As part of the data package being assembled, an engineering-scale demonstration with non-radioactive simulants was performed at the Savannah River National Laboratory (SRNL) using the Scaled Continuous Processing Facility (SCPF) to fill a container with simulated Cast Stone grout. Legacy salt solution from previous Hanford salt waste testing was trimmed to resemble the average composition generated from the Hanford Tank Waste Operation Simulator (HTWOS) used in the screening tests. The dry blend materials, ordinary portland cement (OPC), Class F fly ash, and ground granulated blast furnace slag (GGBFS or BFS) were obtained from Lafarge North America in Pasco, WA. Over three days, the SCPF was used to fill a 6,056 liter (1600 gallon) container with simulated Cast Stone grout. The tank was instrumented with x-, y-, and z-axis thermocouples to monitor curing temperature and two formed core sampling vials. The target production rate was 5.7 L/min (1.5 gpm). This required a salt solution flow rate of approximately 3.78 L/min (1 gpm) and a premix feed rate of approximately 263 kh/h (580 lb/h). The final surface slope at a fill height of 100 cm was 2.5-3.8 cm across the 260 cm (8.5 ft) diameter tank. During processing, grout was collected from both the mixer discharge and the discharge into the tank. These samples were stored in a humid environment either in a closed box proximal to the tank or inside the laboratory. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Conceptual Model and Risk Assessment of Waste Package Evolution - 15643

A generic disposal facility model was created to mirror disposal of low-level radioactive waste in a humid environment. The standard waste package used is a painted and primed carbon steel box stacked four high and placed next to one another within the disposal facility, preventing outward deformations of individual boxes. Waste is placed in the facility in sections, and as each section is filled it is covered with a layer of natural soil. It was assumed that boxes are exposed to natural precipitation rates for approximately 12 months, reduced to natural background infiltration to simulate the soil cover, and dropping to 1% of background infiltration when the facility is filled and a water-shedding interim cover is installed. It was also assumed that each box retains its structural integrity at the time of disposal, with the exception of the top box, which has its lid collapse inward from the overlying soil cover. Waste packages experience different rates of corrosion, differential amounts of infiltration, and fill with liquid at different rates based on when and where they are placed in the facility. Using historical literature and DOE site reports, calculations were made to estimate a range of waste package failure outcomes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Environmental and Radiological Readiness of Fixatives, Foams, and Intumescent Coatings for D and D Applications - 18553

During D and D operations of legacy nuclear facilities, the ability to effectively stabilize contamination in large cavities is a significant challenge. For items that must be removed, such as gloveboxes and entry hoods, a method of simply filling the volume to stabilize any residual contamination after contamination efforts have concluded would be of great benefit, and further the As Low As Reasonably Achievable (ALARA) culture of the DOE complex. The development, testing, and characterization of commercial fixatives and foams for applicability towards fixating contamination provides a potential structural reinforcement and fire resiliency in legacy nuclear facility equipment. Much of the equipment, such as gloveboxes, shielded cells, and radiological hoods left over from nuclear processing leave large cavities. Filling these voids after decontamination efforts provides a means of shielding workers from excess radiation exposure while immobilizing residual contamination. As such, we look to continue the development of fixatives/intumescent coatings from the previous year and test commercially available polymer-based foam products for use in filling these cavities, providing a secondary level of protection and shielding during the D and D process This research is funded as part of the In-Situ Decommissioning and Technology Development and sponsored by DOE- Office of Environmental Management Technology Development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Removal of Historic Low-Level Radioactive Waste from the Highland Drive South and Alexander Street Ravines - 19155

In Port Hope, the Highland Drive South Ravine (HDSR) is the discharge area for groundwater originating from the closed Highland Drive Landfill site. The presence of low-level radioactive waste (LLRW) in the HDSR is the result of a contaminated groundwater plume flowing from the landfill site. The contaminant plume contains elevated concentrations of uranium and arsenic and extends from the landfill south to discharge groundwater to shallow soils in a wet discharge area within the ravine. The groundwater discharges to Hunt's Ponds and Highland Drive South Creek, which are immediately downstream of the wet discharge area and will be included in the remediation design. Remediation and environmental management plans for HDSR have been developed within the framework of the Port Hope Project. Recent subsurface investigations have identified LLRW at greater depths than previously anticipated. Other investigations of surrounding properties have identified LLRW that is primarily from filling activities in backyards. The steep topography of the site and the presence of the creek and seepage face present remediation challenges. Also in Port Hope, the LLRW in the Alexander Street Ravine Area is the result of end dumping, filling of private backyards and stack deposition from the former refinery. The original site extent was focused on the deeper contamination due to end dumping and included a portion of Alexander Creek. However, recent investigations have confirmed other fill areas and the extent of stack deposition that are generally contiguous with the end dumping area. Restricted access to the site will result in all remediation in this area taking place via a single access point along the narrow existing Hayward Street. While both of these sites are ravines in Port Hope containing LLRW, the origins of contamination and remediation approaches are very different. This paper will compare these two challenging remediation sites. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Co-Disposal Waste Package Loading Options for DOE SNF and HLW - 20331

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

High Performance Heat Pipe Power Transient Testing at SPHERE Facility

Microreactors are being researched, designed, and built at Idaho National Laboratory (INL). Microreactors are small reactors defined at less than 20MW of power. These reactor concepts are also being looked at throughout industry for various applications. An important aspect of these reactor designs is economic feasibility i.e. lower overnight capital cost. The driving factors for implementing microreactors are quick setup and takedown, minimal operators, and the ability to manufacture them readily and to fit in mid-sized containers for transport. A specific area of research to aid in successful integration of these factors within the designs is passive heat removal of the core’s thermal power. Interest in heat pipes to achieve this passive heat removal has been shown across multiple industry partners. Because of this interest, INL has developed a test facility to facilitate experimental tests for sodium filled heat pipes. INL has developed the Single Primary Heat Extraction and Removal Emulator (SPHERE) facility to run experiments on high performance, sodium filled heat pipes. As mentioned above, heat pipes are passive heat transfer devices. Radially, heat pipes are broken up into an outer wall, a small annular gap, a wick structure, and a centerline gap. They function by utilizing latent heat transfer. Heat pipes are traditionally separated into three regions, an evaporator (heat input), an adiabatic region, and finally a condenser region (heat removal). As heat is being applied to the evaporator, the working fluid undergoes a phase change to a vapor. This phase change causes a differential pressure across the axial length of the pipe driving flow down the center gap of the heat pipe. The vapor flows down past the adiabatic region to the condenser where the heat is removed. This heat removal forces the working fluid to phase change back to a liquid. The wick structure is then utilized to drive the flow back towards the evaporator by capillary forces. This backflow is aided by the annular gap. Because this heat transfer mechanism functions with latent heat transfer, the heat pipe is close to isothermal down the axial length. Heat pipes can operate under a wide range of working fluids. Considerations for these working fluids are primarily driven by operating temperatures amongst other important factors based around overall performance. Sodium filled heat pipes operate from 450°C up to 900°C. This temperature range works well for the current microreactor designs. In conjunction with this experimental capability, INL has developed a modeling software to simulate heat pipe physics within reactor cores. This modeling software is called Sockeye and functions under the established INL Multiphysics Object Oriented Simulation Environment (MOOSE). SPHERE also supports Sockeye development by providing the modeling team with experimental data on an array of setups and operating parameters to support validation efforts. A power transient experiment was performed utilizing the SPHERE facility to continue to aid with Sockeye development. The testing followed a proposed test plan to ramp up and down the temperature of the heat pipe. Sockeye models steady state heat pipe operation with high accuracy, the data provided by the power transient testing aims to assist with the validation efforts and further enhance transient modeling capability of the tool [2].

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

A Complexity Agnostic Clustering Engine for Time Projection Chambers and its Implementation in FPGA

A clustering functional block implemented in field-programable-gate-array (FPGA) for time projection chambers (TPC) operating with predictable time regardless the complexity of the event is described in this paper. The clustering functional block reorganizes input data and the hits data belonging to the same clusters are output together for further process in the later stages. The clustering operation consists of two phases, data filling phase and data outputting phase, and the later uses the same number of clock cycles as the data filling phase. The clustering block can accommodate events with arbitrary number of clusters and number of hits per cluster as long as the total number of hits is within a predesigned limit. The operation time is exactly twice of the data filling time with no residual O(n2) term. The clustering block has been implemented with operating frequency of 200 MHz in a low-cost FPGA evaluation module and test results confirm the expected performance.

Wu, Jinyuan [Fermilab] (ORCID:0000000344329521)↗

DOE OSTI2019

Fuel Cell Electric Vehicle Durability and Fuel Cell Performance

The objectives of this project are to validate hydrogen fuel cell electric vehicles (FCEVs) in real-world settings and to identify the current status and evolution of the technology. The analysis objectively assesses progress toward targets and market needs defined by the U.S. Department of Energy and stakeholders, provides feedback to hydrogen research and development, and publishes results for key stakeholder use and investment decisions. Fiscal year 2018 objectives focus on analysis and reporting of FCEV driving range, fuel economy, drive and fill behaviors, durability, fill performance, and fuel cell performance. This report specifically addresses the topics of fuel cell stack durability and fuel cell performance.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2017

Grain Boundary Engineering for Achieving High Thermoelectric Performance in n-Type Skutterudites

Grain or phase boundaries play a critical role in the carrier and phonon transport in bulk thermoelectric materials. Previous investigations about controlling boundaries primarily focused on the reducing grain size or forming nanoinclusions. Herein, liquid phase compaction method is first used to fabricate the Yb–filled CoSb 3 with excess Sb content, which shows the typical feature of low–angle grain boundaries with dense dislocation arrays. Seebeck coefficients show a dramatic increase via energy filtering effect through dislocation arrays with little deterioration on the carrier mobility, which significantly enhances the power factor over a broad temperature range with a high room–temperature value around 47 μW cm –2 K –1 . Simultaneously, the lattice thermal conductivity could be further suppressed via scattering phonons via dense dislocation scattering. As a result, the highest average figure of merit ZT of ≈1.08 from 300 to 850 K could be realized, comparable to the best reported result of single or triple–filled Skutterudites. Furthermore, this work clearly points out that low–angle grain boundaries fabricated by liquid phase compaction method could concurrently optimize the electrical and thermal transport properties leading to an obvious enhancement of both power factor and ZT .

36 MATERIALS SCIENCE↗

DOE OSTI2019

A method for the characterization of intra-pixel response of infrared sensor

Due to the response variations within a pixel, response of sensor varies with the positions of the point target. The accurate measurement of intra-pixel response of the sensor is the key to high-precision centroid estimation and radiometric calibration. In this paper, we establish an intra-pixel response model based on the pixel structure and use Gauss function to describe the beam spot profile. We further calculate the full width at half maximum (FWHM) of the Gauss function and the fill factor of the infrared sensor using the grid search algorithm. Experimental results show that the Gauss function FWHM differences among different pixels using the same radius could be less than 6%. The fill factors are not constant in different pixels because of errors in the process of manufacture and assembly, and the error between model and measured values is less than 3.6%. This work could be useful for stellar calibration and centroid estimation.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗