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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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Pelletization with Spark Plasma Sintering and Characterization of Metal Iodides: An Assessment of Long-Term Radioiodine Immobilization Options

Four promising iodine “getter” materials (Ag, Cu, Bi, and Sn) for radioiodine capture were assessed in their pure metal-iodide (MI x ) pelletized forms to compare relative chemical durabilities. To study chemical durability, commercial MI x compounds of AgI, BiI 3 , BiOI, CuI, and SnI 4 were converted to dense monolithic pellets using spark plasma sintering. Semidynamic leach testing in the form of modified ASTM C1308 tests was then performed on the pellets in two different forms including unmounted (as-pressed) specimens (i.e., “U”) and epoxy-mounted specimens (i.e., “M”) with polished surfaces. The chemical durability results and sample characterizations showed that three of the five MI x compounds tested (i.e., AgI, CuI, and BiOI) displayed moderate to high leach resistances. Further, the remaining two MI x compounds (i.e., BiI 3 and SnI 4 ), which are both desirable iodine waste forms due to their high iodine loading capacities, readily decomposed during leach testing, indicated by crystallographic changes in the specimens as well as large amounts of iodine detected in the leachate solutions. The instabilities of BiI 3 and SnI 4 raise uncertainties for using the base metals/cations (i.e., Bi 0 /Bi 3+ and Sn 0 /Sn 4+ , respectively) as viable getters for radioiodine capture due to likely poor waste form chemical durabilities after capture and consolidation into waste forms.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Low-Cost, Efficient and Durable High Temperature Wireless Sensors by Direct Write Additive Manufacturing for Application in Fossil Energy Systems

In-situ monitoring of physical parameters in fossil energy (FE) systems is challenging because of their high operating temperatures and oxidizing environments. This research aims utilize Aerosol Jet (AJ) nanoparticle printing, an advanced additive manufacturing method, to fabricate high temperature sensors for FE applications. Sensor films were fabricated by the AJ printing method with an excellent control over film microstructure and properties, in addition to creating less waste. Printing of six different materials, namely, silver (Ag), nickel (Ni), nickel-chromium (NiCr) alloy, copper (Cu), copper-nickel (CuNi), and carbon nanotubes (CNTs) was carried out for their evaluation of as sensor films. The films were then characterized using impedance spectroscopy, scanning electron microscopy (SEM), transmission electron microscopy (TEM), x-ray diffraction (XRD), and x-ray photoelectron spectroscopy (XPS). Film microstructure was correlated with the sensor electrical and mechanical behavior at different temperatures (24-500 °C) and frequencies (0.02 - 300 kHz). This knowledge was then used to fabricate, test, and characterize strain and temperature sensors for high temperature applications. The strain sensors were shown to have a very low thermal strain up to 500 °C and a 57% higher gauge factor compared to the commercially available counterparts. This enhance performance was attributed to the inherent porosity in the sensor films that allowed the films to exhibit a metamaterial-like behavior when deformed. The AJ printed temperature sensors were shown to have the highest Seebeck coefficient amongst all the film-based temperature sensors. In addition, fundamental aspects of crystal structure and phase transformations in Ba(Fe0.7Ta0.3)O3-δ (BFTO), a high temperature sensor material, was investigated along with 3D printed antennas that can transmit sensor signals at high temperatures. Two PhD students, one from first generation college-going underrepresented minority community (URM), two postdoctoral fellows, and several masters and undergraduate students were supported by this award. This work opens up the possibility of realizing low-cost high-performance sensor devices using advanced manufacturing techniques for high temperature FE applications.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2021

SNF Canister Coatings for Corrosion Prevention and Mitigation (FY21 Status Report)

This report summarizes the current actives in FY21 related to the effort by Sandia National Laboratories to identify and test coating materials for the prevention, mitigation, and repair of spent nuclear fuel dry storage canisters against potential chloride-induced stress corrosion cracking. This work follows up on the details provided in Sandia National Laboratories FY20 report on the same topic, which provided a detailed description of the specific coating properties desired for application and implementation on spent nuclear fuel canisters, as well as provided detail into several different coatings and their applicability to coat spent nuclear fuel canisters. In FY21, Sandia National Laboratories has engaged with private industry to create a Memorandum of Understanding and established a collaborative R&D program building off the analytical and laboratory capabilities at Sandia National Laboratories and the material design and synthesis capabilities of private industry. The resulting Memorandum of Understanding included four companies to date (Oxford Performance Materials, White Horse R&D, Luna Innovations, and Flora Coating) proposing six different coating technologies (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin, silane-based polyurethane hybrid with and without a Znrich primer, and a quasi-ceramic sol-gel polyurethane hybrid) to be tested, evaluated, and optimized for their potential use for this application. This report provides a detailed description of each of the coating systems proposed by the participating industry partners. It also provides a description of the planned experimental actives to be performed by Sandia National Laboratories including physical tests, electrochemical tests, and characterization methods. These analyses will be used to identify specific ways to further improve coating technologies toward their application and implementation on spent nuclear fuel canisters. In FY21, Sandia National Laboratories began baseline testing of the base metal material in according with activities of the Memorandum of Understanding. In FY22, Sandia National Laboratories will receive coated coupons from each of the participating industry partners and begin characterization, physical, and electrochemical testing following the test plan described herein.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Decontamination of urban surfaces contaminated with radioactive materials and consequent onsite recycling of the waste water

Enhancing rapid remediation strategies is paramount for recovery after a large-scale nuclear contamination event in an urban environment. Some current strategies recommend use of readily available equipment, materials, and facilities to expedite recovery. For example, applying pressurized water to contaminated surfaces may effectively remove radioactive contamination. In this study, a commercial power washer removes soluble forms of 152 Eu 3 + , 85 Sr 2 + , and 137 Cs + contamination from common porous building materials, and computer simulations characterize the recycling of the resultant contaminated wash water. Pressure washing the porous building materials under spray conditions typical with do-it-yourself units improved decontamination factors (DFs) for 152 Eu compared to low-pressure application of tap water (majority of two-tailed t-test p-values < 0.1), but pressure did not improve DFs for 137 Cs or 85 Sr. For both pressurized and low-pressure applications, adding potassium ions (K + ) to promote ion exchange reactions produced significantly higher DFs for tested radionuclides on asphalt, brick, and concrete. The resultant contaminated wash water can be processed through self-prepared chemical filtration beds of clay and sand. Modeled in a prior study, the beds yielded linear trends (R 2 > 0.98) in sensitivity analyses between most bed configuration variables and bed performance variables, permitting flexible ad-hoc bed design. The experimental and simulation results led to estimates of the remediation rate and waste generated after cleaning 250 m 2 of cesium-contaminated concrete from the combined deployment of a power washer and two different mobile treatment beds. Furthermore, the first treatment bed was designed to reduce treatment time and processed 1900 L of wash solution in 70 min using 880 kg of clay/sand infill material. Designed to reduce the solid waste generated, the second bed processed the same solution volume in 1040 min (17 h) using 170 kg of clay/sand infill material. The results of this analysis warrant further investigation of power washing with recycled salt solution as an effective rapid decontamination method with manageable waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Integration of Large-Scale Electrical Imaging into Geological Framework Development and Refinement

Geologic framework models (GFMs) are critical to the construction of reliable simulation models of groundwater flow and contaminant transport. To support GFM development, direct information (e.g., core samples, fluid samples, hydraulic testing) tends to be sparse and separated by large distances relative to the spatial scales of aquifer heterogeneity. There are additional challenges associated with highly contaminated legacy waste sites, where drilling is particularly costly, and invasive sampling requires specialized handling and disposal of hazardous materials. At these sites in particular, non-invasive geophysical imaging can play an important role in filling spatial gaps between boreholes and reducing characterization costs by optimizing and minimizing the number of necessary boreholes. Here this paper presents a case study demonstrating the use of large-scale (> 30 km 2 ) electrical mapping to identify hydrostratigraphy and potential paleochannels at the Hanford Site, located in Washington State, USA. In two field campaigns, over 36 line-kilometers of electrical resistivity tomography (ERT) data were collected along 14 transects. ERT surveys were sited and performed to image critical aspects (e.g., paleochannels, stratigraphic contacts) of the subsurface, demonstrating a general workflow for integrating ERT with GFM development. Inconsistencies between the GFM and ERT were catalogued to provide a basis for future site characterization using complementary geophysical methods and (or) direct sampling.

58 GEOSCIENCES↗
DOE OSTI2021

Metagenomic and metatranscriptomic analysis reveals enrichment for xenobiotic‐degrading bacterial specialists and xenobiotic‐degrading genes in a Canadian Prairie two‐cell biobed system

Summary Biobeds are agriculture‐based bioremediation tools used to safely contain and microbially degrade on‐farm pesticide waste and rinsate, thereby reducing the negative environmental impacts associated with pesticide use. While these engineered ecosystems demonstrate efficient pesticide removal, the microbiomes in these environments remain largely understudied both taxonomically and functionally. This study used metagenomic and metatranscriptomic techniques to characterize the microbial community in a two‐cell Canadian biobed system before and after a field season of pesticide application. These culture‐independent approaches identified an enrichment of xenobiotic‐degrading bacteria, such as Afipia , Sphingopyxis and Pseudomonas , and enrichment and transcription of xenobiotic‐degrading genes, such as peroxidases, oxygenases, and hydroxylases, among others; we were able to directly link the transcription of these genes to Pseudomonas , Oligotropha , Mesorhizobium , Rhodopseudomonas , and Stenotrophomonas taxa.

Russell, Jennifer N.↗
DOE OSTI2020

Novel complex ceramic oxides, Ln 2 TiO 5 (Ln = La, Sm, Gd, Tb, Dy, Ho, Er, and Yb), for polyphase nuclear waste‐forms

Abstract As part of a broader study of ceramic nuclear waste‐forms, four different lanthanide titanates were fabricated; La 0.1 Sm 0.1 Gd 0.1 Tb 0.1 Dy 0.3 Ho 0.1 Er 0.2 YbTiO 5 , Sm 0.3 Gd 0.3 Dy 0.3 Yb 1.1 TiO 5 , Sm 0.1 Gd 0.4 Dy 0.4 Yb 1.1 TiO 5 , and Sm 0.2 Gd 0.2 Dy 0.2 Yb 1.4 TiO 5 . The aim was to produce single‐phase novel materials with cubic symmetry, capable of incorporating a wide variety of cations and with acceptable radiation tolerance. The chemistry flexibility and radiation tolerance are some of the major desirable properties for nuclear waste‐form materials. By using multiple lanthanides the average lanthanide radius can be controlled and consequently the structure, along with properties such as radiation tolerance. The radiation tolerance was assessed using in situ 1 MeV krypton irradiation and transmission electron microscopy characterization. Those materials for which cubic symmetry was achieved displayed better radiation tolerance; a greater critical fluence of ions ( F c ) was required for the crystalline to amorphous transition, and a lower temperature was required to maintain crystallinity ( T c ) during irradiation.

Aughterson, Robert D.↗
DOE OSTI2024

Method Development and Implementation for Characterizing Uncertainties in Tank Leak Volumes for Composite Analysis.

The purpose of this environmental calculation file (ECF) is to provide a summary of methodology, inputs, assumptions, and results of a Special Analysis (SA) titled “Method Development and Implementation for Characterizing Uncertainties in Tank Leak Volumes for the Updated Composite Analysis.” This SA is one of the three SAs conducted in Fiscal Year (FY) 2024 for the Composite Analysis (CA) for the Hanford Site (DOE/RL-2019-52, Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau (FY 2020)), and it is performed following the Unreviewed Composite Analysis Question Evaluation procedure, UCAQ-24-01-E (UCAQ-24-01-E, Rev. 0).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Atom Probe Tomography Investigation of Clustering in Model P 2 O 5 -Doped Borosilicate Glasses for Nuclear Waste Vitrification

Atom probe tomography (APT) has been utilized to investigate the microstructure of two model borosilicate glasses designed to understand the solubility limits of phosphorous pentoxide (P 2 O 5 ). This component is found in certain high-level radioactive defence wastes destined for vitrification, where phase separation can potentially lead to a number of issues relating to the processing of the glass and its long-term chemical and structural stability. The development of suitable focused ion beam (FIB)-preparation routes and APT analysis conditions were initially determined for the model glasses, before examining their detailed microstructures. In a 3.0 mol% P 2 O 5 -doped glass, both visual inspection and sensitive statistical analysis of the APT data show homogeneous microstructures, while raising the content to 4.0 mol% initiates the formation of phosphorus-enriched nanoscale precipitates. This study confirms the expected inhomogeneities and phase separation of these glasses and offers routes to characterizing these at near-atomic scale resolution using APT.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Qualification of Processing Batches for Tank Closure Cesium Removal Through In-Tank Batch Contact Testing - 20470

The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2023

Analysis of the Solid Products from the OSCAR and the AOWG Trash Processing Systems

Two trash conversion systems that process space logistical trash for conversion into gases and liquids have minor solid product formation consisting of primarily high carbon ash, inorganic salts, and aluminum metal flakes. These solid products necessarily comprise part of the system output because the thermal degradation processes intentionally operate at temperatures below aluminum vaporization and do not convert all solid inorganics or carbon-based compounds into gaseous compounds (i.e., CO 2 or CH 4 ). The two systems included the Orbital Syngas/Commodity Augmentation Reactor (OSCAR) combustion system funded by the NASA Space Technology Mission Directorate (STMD) located at NASA’s Kennedy Space Center, and the Advanced Organic Waste Gasifier (AOWG) steam reforming system developed by Pioneer Astronautics in Lakewood, Colorado, and funded by the NASA Small Business Innovation Research (SBIR) Exploration Systems Development Mission Directorate (ESDMD). OSCAR residual solids were collected from suborbital flight and ground-based laboratory testing, while all AOWG samples were collected from ground testing. This report discusses the solid product characterization, including elemental analyses performed with scanning electron microscopy (SEM) with energy dispersive X-ray spectroscopy (EDS), and a surface chemistry performed with x-ray photoelectron spectroscopy (XPS). A particle size analysis was performed with sieves, followed by SEM analyses. These analyses were performed to elucidate what high value products can be extracted or utilized from the solid products for follow on infusion into chemical feedstocks, manufacturing additives, or in-situ construction-based material needs.

OSCAR↗
DOE OSTI2023

Reactive Extrusion of Waste Plastics with Compatibilizer and Lightly Pyrolyzed Crumb Rubber for Asphalt Modification

Modifying asphalt is a potentially high-value application for reusing waste plastics because of the high-volume usage of asphalt in highway construction. However, simply blending hot plastics and asphalt encounters difficulties related to the poor solubility of polymers, which limits the formation of a swollen network with asphalt molecules. The polymer phases also tend to coalesce and separate from asphalt during high-temperature storage in static conditions. The present study developed an innovative process to stabilize waste plastics in asphalt and improve binder storage stability by using lightly pyrolyzed crumb rubber together with a chemical compatibilizer. Both polymers were extruded to produce a thermoplastic elastomer (TPE) for asphalt modification. The mechanical performance and chemical reactions of TPEs were characterized via tension test and Fourier transform infrared spectroscopy. The storage stability and rheological properties of modified binder blends were evaluated through laboratory asphalt stability test and dynamic shear rheometer test. Polymer phases and network structures were characterized through optical microscopy. It was found that the pyrolyzed and reactive extrusion process improved the rubber solubility and polymer interaction, and therefore the storage stability of modified binder blends. The co-existence of rigid plastic and soft rubbery regimes in an entangled network provided a promising pathway to improve the mechanical performance of asphalt binders in both high- and low-temperature domains.

Ma, Yuetan↗
DOE OSTI2022

MELCOR Input Model for Spent Fuel Transportation and Storage Canister

This report documents the progress and current results of the MELCOR spent fuel cask input model. The MELCOR model is being developed to investigate aerosol transport and deposition given the aerosol physical phenomena models within MELCOR. To perform the analyses, a general portrayal of the MAGNASTOR® cask system has been employed; however, this system was selected based on available information to provide a reasonable representation of a spent fuel cask. The analytical results are not intended to characterize the performance of the MAGNASTOR® cask. Instead, the provided results are intended to enhance our general understanding of the aerosol behavior within casks and the validity of current models. The current model efforts are being performed to investigate hypothetical UO 2 release from failed fuel pins within a spent fuel cask. The existing MELCOR model of the MAGNASTOR® cask system has been adapted to permit future comparative analyses with the GOTHIC representation of the MAGNASTOR® cask. To support this comparison, the PNNL model characteristics that are unrelated to the aerosol modeling were applied to the MELCOR model. These characteristics included improved comparability of the axial fidelity, total spent fuel power, fuel pin axial power profile, and heat losses from cannister. The thermal-hydraulic solutions are improved within the capability of the MELCOR code and will permit better overall agreement with the GOTHIC results. Detailed results are presented on the thermal-hydraulic analysis of the MELCOR cask as well as characterization of UO 2 aerosol dispersion and deposition within the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1980

Assessment of the use of space technology in the monitoring of oil spills and ocean pollution: Technical volume. Executive summary

The potential of space systems and technology for detecting and monitoring ocean oil spills and waste pollution was assessed as well as the impact of this application on communication and data handling systems. Agencies charged with responsibilities in this area were identified and their measurement requirements were ascertained in order to determine the spatial resolution needed to characterize operational and accidental discharges. Microwave and optical sensors and sensing techniques were evaluated as candidate system elements. Capabilities are described for the following: synthetic aperture radar, microwave scatterometer, passive microwave radiometer, microwave altimeter, electro-optical sensors currently used in airborne detection, existing space-based optical sensors, the thematic mapper, and the pointable optical linear array.

Alvarado, U. R.↗
DOE OSTI2024

Out-of-Pile Testing and Instrumentation Transient Water Irradiation System

Current initiatives to increase the burnup of conventional nuclear fuels past the approximate 62 GWd/t limit have been spurred on by direct savings to refueling and waste storage. The technical justification for a new license limit requires extensive qualification through experimental testing. Unlike beginning-of-life fuels, high-burnup fuels are more susceptible to fuel fragmentation, relocation, and dispersal (FFRD), therefore more data is needed to characterize fuels under key accident scenarios. The Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory has developed a testing apparatus architecture to test fuels and claddings at prototypic conditions. The Transient Water Irradiation System (TWIST) is the latest iteration of a testing device capable of conducting loss of coolant accidents (LOCAs) in TREAT. The Out-of-Pile Testing and Instrumentation TWIST (OPTI-TWIST) is an electrically heated device that is analogous to TWIST. OPTI-TWIST allows for detailed instrumentation and thermal-hydraulic characterization. TWIST ultimately aims to conduct the most advanced in-situ diagnostics to evaluate FFRD in a prototypic LOCA. Moreover, it will explore the phenomenological bifurcation of a decay-energy heat up driven LOCA and a stored-energy heat up driven LOCA. The instrumentation suite includes conventional thermocouples and pressure transducers in addition to an electro impedance sensor, an acoustic emission sensor, an optical pressure sensor, and an optical pyrometer. Characterizing these instruments in OPTI-TWIST eliminates complications of irradiation effects while preserving extreme thermal-hydraulic conditions. Finally, benchmarking both devices to a thermal-hydraulic code like the Reactor Excursion and Leak Analysis Program (RELAP)5-3D provides a unique opportunity for iteration. Pre-test predictions and post-test interpretations inform the physical designs, operational procedures, test conditions, and instrumentation types and positions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Characterization of Infrequent Samples from the Concentration, Storage, and Transfer Facility: Leak Detection Box (LDB) Drain Cell Sample: August 04, 2022 Sample

Savannah River Mission Completion Engineering (SRMC-E) requested that the Savannah River National Laboratory (SRNL) analyze the Concentration, Storage, and Transfer Facility (CSTF) samples from the following Tank Farm areas: the sump encasement, catch tank, drain cell, and waste tank annulus. In general, these CSTF samples will be analyzed on an infrequent basis and analyses will include detection for total beta/gamma activities, total alpha activity, free hydroxide, and pH measurements.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Demonstration of the On-the-Fly Shielding Analysis Method: Spent Fuel and Waste Disposition

This report documents work performed supporting the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management activities under work breakdown structure element 1.08.02.04.01, “Data and Tools Development, Validation, and Maintenance.” In particular, this report fulfills milestone M3SF-21OR020401016, “Implement on-the-fly dose analysis methodology in UNF-ST&DARDS” within work package SF-21OR02040101, “Commercial SNF Characterization - ORNL.” The Used Nuclear Fuel - Storage, Transportation & Disposal Analysis Resource and Data System (UNFST& DARDS) enables automated dose rate calculations for spent nuclear fuel (SNF) transportation packages and storage casks using a Monte Carlo radiation transport code. The explicit method uses a detailed model of the SNF system and its contents. Therefore, a dose rate calculation is required for each as-loaded transportation package or storage cask because the SNF assemblies within a canister typically have unique irradiation characteristics. An alternate method, referred to as the “on-the-fly” shielding analysis method, has been proposed that requires only a set of Monte Carlo dose rate calculations for each transportation packaging/storage cask design. The results of the Monte Carlo dose rate calculations are independent of the SNF assembly irradiation and decay characteristics. The dose rate values may then be combined with the radiation source strength of the SNF assemblies associated with a particular transportation packaging/storage cask design to determine actual dose rates. This report presents on-the-fly dose rate calculations for a representative SNF storage cask and verification of the on-the-fly dose rate calculation results by comparison with reference dose rate calculations using the explicit Monte Carlo dose rate calculation. The on-the-fly shielding analysis method was implemented in UNF-ST&DARDS. A Python program was developed to process the MAVRIC dose rate results obtained by source particle type, energy group, and fuel geometry region. A Python processor created binary files, which were saved as a special UNF-ST&DARDS library for on-the- fly shielding analyses. UNF-ST&DARDS uses the precalculated on-the-fly binary libraries generated by the Python data processor and directly executes the Python code for on-the-fly dose analysis. This Python code unzips the pre-generated binary files mentioned above, reads the data, and combines them with user-specified sources for dose and uncertainty calculations. The Python programs were verified using Excel calculations and by comparison with the values obtained with the MAVRIC post-processing utilities applied to the 3dmap files. This method can currently be used to determine dose rates for as-loaded HI-STORM FW storage casks. The UNF-ST&DARDS analysis wizard for on-the-fly shielding analysis is described in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗