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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Seismotectonic evidence for subduction beneath the Eastern Greater Caucasus

SUMMARY We generated high-resolution 3-D seismic velocity models as well as a relocated earthquake catalogue across the eastern Greater Caucasus and Kura basins. This work was done using data from the recently upgraded Republic Seismological Survey Center's (RSSC) seismic network. We generated our tomographic images of crustal velocity structure in Azerbaijan using double-difference inversions (i.e. tomoDD and hypoDD). Earthquake catalogues from the RSSC between 2011 and 2016 were used; these catalogues include absolute arrival times of 103 288 P- and 120 952 S-wave traveltime picks for 7574 events recorded at 35 stations in Azerbaijan. Beginning with a layered, 1-D velocity model that was estimated using VELEST, we inverted simultaneously for relative location, Vp and Vs on a 3-D grid with dimensions 670 × 445 × 45 km, with a uniform grid spacing of 55 × 55 × 5 km for all of Azerbaijan. We observe that the relocated hypocentres cluster into two depth ranges, at the surface and at depth, that appear to correspond to major fault zones and the top of a subducting plate. Additionally, we note intermediate depth seismicity (∼50–60 km) beneath the Kura Basin, and a northward deepening of earthquake depths. Seismic velocities vary significantly throughout the study region; we observe very slow velocities throughout the Kura Basin between 5 and 15 km, and elevated velocities at 20–35 km. The wholesale velocity structure and seismic structure of Kura Basin strongly mirrors that of the Caspian Sea, which suggests that the geodynamics of the Caspian continue westwards into Azerbaijan. The key results of this study suggest that the northward subduction observed in the Caspian Sea continues beneath the Eastern Greater Caucasus, as well as provides evidence for active faulting along the southern margin of the mountain range.

Gunnels, Michael↗
DOE OSTI2021

Solving the sample size problem for resource selection functions

Abstract Sample size sufficiency is a critical consideration for estimating resource selection functions (RSFs) from GPS‐based animal telemetry. Cited thresholds for sufficiency include a number of captured animals and as many relocations per animal N as possible. These thresholds render many RSF‐based studies misleading if large sample sizes were truly insufficient, or unpublishable if small sample sizes were sufficient but failed to meet reviewer expectations. We provide the first comprehensive solution for RSF sample size by deriving closed‐form mathematical expressions for the number of animals M and the number of relocations per animal N required for model outputs to a given degree of precision. The sample sizes needed depend on just 3 biologically meaningful quantities: habitat selection strength, variation in individual selection and a novel measure of landscape complexity, which we define rigorously. The mathematical expressions are calculable for any environmental dataset at any spatial scale and are applicable to any study involving resource selection (including sessile organisms). We validate our analytical solutions using globally relevant empirical data including 5,678,623 GPS locations from 511 animals from 10 species (omnivores, carnivores and herbivores living in boreal, temperate and tropical forests, montane woodlands, swamps and Arctic tundra). Our analytic expressions show that the required M and N must decline with increasing selection strength and increasing landscape complexity, and this decline is insensitive to the definition of availability used in the analysis. Our results demonstrate that the most biologically relevant effects on the utilization distribution (i.e. those landscape conditions with the greatest absolute magnitude of resource selection) can often be estimated with much fewer than animals. We identify several critical steps in implementing these equations, including (a) a priori selection of expected model coefficients and (b) regular sampling of background (pseudoabsence) data within a given definition of availability. We discuss possible methods to identify a priori expectations for habitat selection coefficients, effects of scale on RSF estimation and caveats for rare species applications. We argue that these equations should be a mandatory component for all future RSF studies.

Street, Garrett M.↗
DOE OSTI2024

Wasatch Fault Structure from Machine Learning Arrival Times and High-Precision Earthquake Locations

Abstract On 18 March 2020, a magnitude 5.7 earthquake hit the Salt Lake valley in the state of Utah, United States. Using a dense geophone deployment and machine learning (ML), an additional several thousand events were detected and located. Currently, both the mainshock and the majority of the aftershocks are suspected to have occurred on or near a deeper portion of the Salt Lake segment of the Wasatch fault—part of a large range-bounding fault system thought to be capable of generating an Mw 7.2 earthquake. However, a small subset of aftershocks may have occurred on a portion of the more steeply, eastward dipping, and poorly understood West Valley fault. Unfortunately, the catalog locations and lack of focal mechanisms for this subset of aftershocks provide only a crude constraint on the true fault structure. To better illuminate fault structure, we relocate the ML-generated catalog with a range of magnitudes from −2 to 4.6, using: (1) NonLinLoc, a nonlinear location algorithm, (2) source-specific station terms, and (3) waveform coherence. We further compute first-motion focal mechanisms for 68 events. Results of the relocation suggest a simpler, minimally listric Wasatch fault geometry, contrary to what has been previously proposed. We also find that analysis of the focal mechanisms and waveform similarity indicates minimal event similarity throughout the Magna sequence, suggesting a highly complex and heterogeneous rupture zone, as opposed to rupture on a single plane. These findings suggest an increased seismic hazard due to the overall shallowness of the earthquake sequence and highly varied rupture mechanisms.

Geochemistry & Geophysics↗
DOE OSTI2023

Generating Mixed Patterns of Residential Segregation: An Evolutionary Approach

The Schelling model of residential segregation has demonstrated that even the slightest preference for neighbors of the same race can be amplified into community-wide segregation. However, these models are unable to simulate mixed, coexisting patterns of segregation and integration, which have been seen to exist in cities. Using evolutionary model discovery we demonstrate how including social factors beyond racial bias when modeling relocation behavior enables the emergence of strongly mixed patterns. Our results indicate that the emergence of mixed patterns is better explained by multiple factors influencing the decision to relocate; the most important being the interaction of nonlinear, rapidly diminishing racial bias with a recent, historical tendency to move. Additionally, preference for less isolated neighborhoods or preference for neighborhoods with longer residing neighbors may produce weaker mixed patterns. Finally, this work highlights the importance of exploring the influence of multiple hypothesized factors of decision making, and their interactions, within agent rules, when studying emergent outcomes generated by agent-based models of complex social systems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI1994

Nuclear Safety [Vol. 35, No. 2, July-December 1994]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 179 Consideration of Postaccident Consequences in the Determination of Safety Objectives for Future Nuclear Power Plants in France, D. Queniart, A. Sugier, and J. Lochard; ACCIDENT ANALYSIS: 187 Nuclear Safety Research: The Phebus FP Severe Accident Experimental Program, P. von der Hardt, A. V. Jones, C. Lecomte, and A. Tattegrain; 205 Containment Performance Analysis of the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory, S. H. Kim, R. P. Taleyarkhan, and V. Georgevich; 213 Assessment of Fission Product Deposits in the Reactor Coolant System: The DEVAP Program, G. Le Marois and M. Megnin; 222 Erratum to “A Review of the Available Information on the Triggering Stage of a Steam Explosion," Vol. 35, No. 1, CONTROL AND INSTRUMENTATION: 223 Effects of Normal Aging on Calibration and Response Time of Nuclear Plant Resistance Temperature Detectors and Pressure Sensors, H. M. Hashemian; DESIGN FEATURES: 235 Defense in Depth Against the Hydrogen Risk—A European Research Program, F. Fineschi; ENVIRONMENTAL EFFECTS: 246 Technical Note: A Preliminary Analysis of the Risks to Hong Kong Resulting from Potential Accidents of Daya Bay Nuclear Power Plant, Z. Shi and X. Wei; OPERATING EXPERIENCES: 253 Reactor Shutdown Experience, Compiled by J. W. Cletcher; SPECIAL SECTION ON TMI-2 VESSEL INVESTIGATION PROJECT: 256 Three Mile Island—New Findings 15 Years After the Accident A. M. Rubin and E. Beckjord; 269 Relocation of Molten Material to the TMI-2 Lower Head, J. R. Wolf, D. W. Akers, and L. A. Neimark; 280 Insight Into the TMI-2 Core Material Relocation Through Examination of Instrument Tube Nozzles, L. A. Neimark; 288 Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head, D. W. Akers and B. K. Schuetz; 301 Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head, D. R. Diercks and G. E. Korth; 313 Margin-to-Failure Calculations for the TMI-2 Vessel, J. Rempe, L. Stickler, S. Chavez, G. Thinnes, R. Witt, and M. Corradini; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 328 1993 Accident Sequence Precursor (ASP) Program Results, L. N. Vanden Heuvel, J. W. Cletcher, D. A. Copinger, J. W. Minarick, B. W. Dolan, and P. D. O’Reilley; RECENT DEVELOPMENTS: 339 Reports, Standards, and Safety Guides, D. S. Queener; 345 Proposed Rule Changes as of June 30, 1994; ANNOUNCEMENTS: 234 Thirty-First Annual Meeting of the National Council on Radiation Protection and Measurements; 356 1995 International Incineration Conference; 358 Fifth International Controls and Instrumentation Conference; 359 ANS International Topical Meeting on Safety of Operating Reactors; 359 Fifth International Conference on Nuclear Criticality Safety; 360 International Conference on Probabilistic Safety Assessment Methodology and Applications; 351 The Authors; 357 Reviewers of Nuclear Safety, Vol. 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Brentwood Lessons Learned Project Report

The purpose of this report is to document lessons learned in the installation of the hydrogen fueling station at the National Park Service Brentwood site in Washington, D.C., to help further the deployment of hydrogen infrastructure required to support hydrogen and other fuel cell technologies. Hydrogen fueling is the most difficult infrastructure component to build and permit. Hydrogen fueling can include augmenting hydrogen fueling capability to existing conventional fuel fueling stations as well as building brand new hydrogen fueling stations. This report was produced as part of the Brentwood Lessons Learned project. The project consisted of transplanting an existing modular hydrogen fueling station from Connecticut to the National Park Service Brentwood site. This relocation required design and construction at the Brentwood site to accommodate the existing station design as well as installation and validation of the updated station. One of the most important lessons learned was that simply moving an existing modular station to an operating site was not necessarily straight-forward - performing the relocation required significant effort and cost. The station has to function at the selected operating site and this functionality requires a power supply, building supports connecting to an existing alarm system, electrical grounding and lighting, providing nitrogen for purging, and providing deionized water if an electrolyzer is part of the station package. Most importantly, the station has to fit into the existing site both spatially and operationally and not disrupt existing operations at the site. All of this coordination and integration requires logistical planning and project management. The idea that a hydrogen fueling station can be simply dropped onto a site and made immediately operational is generally not realistic. Other important lessons learned include that delineating the boundaries of the multiple jurisdictions that have authority over a project for all parties involved in the project are key to an efficient approval process; and site investigation is necessary when integrating a new station design onto an existing site, particularly an older existing site that may have limited documentation on the site history and operations. The lessons learned for permitting and subcontracting construction work can be applied to other similar sites and to commercial sites.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2017

Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Development of a LOCA Experimental Benchmark for BISON - Milestone L2:FMC.P15.08

Since 2013 CASL has been planning and working towards a PWR LOCA Challenge Problem. Since BISON is the CASL fuel rod simulation platform, it plays a central role in that activity. The goal of the FY-17 CASL milestone reported here was to develop an Experimental Benchmark for LOCA analysis using BISON. Activities encompassed both code development and validation. This report provides a current snapshot of BISON’s capability for LOCA behavior including: 1) a summary of code extensions to facilitate accident analysis, 2) a series of separate effects tests, and 3) initial code validation to complex integral rod LOCA behavior. Conclusions from each of these main activities are summarized below. The key material and behavior models required to address transient high-temperature phenomena occurring during LOCAs in a standard PWR have now been implemented in BISON. These apply specifically to UO2 fuel, Zircaloy cladding and water coolant. During FY-17 important new capability to address axial UO2 fuel relocation (for 1.5D geometry) and account for oxidation energy deposition in cladding, were included. Planned future development efforts include extending the axial fuel relocation model from 1.5D to 2D/3D and improving BISON’s fission gas release model to include transient gas release associated with the high burnup structure (HBS) in high burnup fuel. Additionally, investigation of potential anisotropic creep behavior in Zircaloy cladding is planned, assuming sufficient experimental data are available to support this effort. A substantial number of separate effects validation cases (42 tests from 3 experimental series) have been completed to compare BISON predictions to measured ballooning and burst behavior for Zircaloy cladding. Such experiments include a wide variety of pressures, temperatures and loading rates. In general, BISON predictions of burst temperature, pressure and burst time are very reasonable. For one experimental series, however, involving both very high temperatures and strain rates, BISON systematically over-predicts the cladding hoop strain. Investigation of this discrepancy is an important and planned activity. With another experiment an effort was made to investigate 3D cladding response due to an azimuthal temperature variation. Results indicated 3D effects are potentially important in fuel rod analysis during LOCAs and will be further investigated in the future. BISON validation to a series of integral fuel rod experiments has also been completed. These experiments involve all fuel and cladding phenomena relevant to LOCA conditions, and can include complexities associated with irradiated fuel relative to fresh fuel. Such experiments also generally include complex thermal-hydraulic boundary conditions. Four experiments (6 rods) have been considered to date including simulated fuel (ZrO2) and both fresh and high-burnup UO2. Test rods ranged from rodlets to full length commercial PWR fuel rods. As with the separate effects experiments, BISON predictions of burst temperature, pressure and burst time are generally very reasonable. Comparisons to cladding peak strain and rod outer diameter axial profiles are less satisfactory, and identify material models and possibly modeling approximations (e.g., 2D-RZ vs 3D geometries) requiring additional investigation. Validation of BISON for integral rod LOCA behavior is by no means complete, with additional cases planned.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

3D Crustal Tomography Model of Utah

The ability to accurately locate seismic events is necessary for treaty monitoring. When using techniques that rely on the comparison of observed and predicted travel times to obtain these locations, it is important that the estimated travel times and their estimated uncertainties are also accurate. The methodology of Ballard et al. (2016a) has been used in the past to generate an accurate 3D tomographic global model of compressional wave slowness (the SAndia LoS Alamos 3D tomography model, i.e. SALSA3D). To re-establish functionality and to broaden the capabilities of the method to local distances, we have applied the methodology of Ballard et al. (2016a) to local data in Utah. This report details the results of the initial model generated, including relocations performed using analyst picked mining events at West Ridge Mine and three ground-truth events at Bingham Mine. We were successfully able to generate a feasible tomography model that resulted in reasonable relocations of the mining events.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Compare predictions of transient fission gas release by empirical and mechanistic models to experiments in high burnup UO 2 fuel

Understanding and predicting fuel performance at high burnup require improving our understanding of transient fission gas release. High-burnup operations enable new mechanisms of fission gas release, which affect fuel performance. The Nuclear Regulatory Commission has recently published its interpretation of existing fuel fragmentation, relocation, and dispersal data in a research information letter. There, transient fission gas release was identified as one of the main factors that contributes to fuel fragmentation, relocation, and dispersal, and therefore limits fuel extension to high burnup. However, transient fission gas release is a complex phenomenon that cannot be fully described by simple empirical descriptions. This report summarizes the development of a mechanistic model for high-burnup transient fission gas release in the fuel performance code BISON. This research was supported by the Nuclear Energy Advanced Modeling and Simulation program during fiscal year 2023 to improve our understanding of high-burnup transient fission gas release and ability to predict it as a function of operation history. To support the development of a mechanistic transient fission gas release model, the existing Simple Integrated Fission Gas Release and Swelling (Sifgrs) model in BISON has been completely refactored to make it more modular and extensible. This effort supports the model's application to high-burnup conditions, its extension to other fuel forms, and the continuous improvement of its current features. Once refactoring was completed, models for high-burnup structure formation, fission gas transfer from non-restructured fuel to high-burnup structure, high-burnup structure intragranular and intergranular fission gas behavior, high-burnup structure bubble evolution, fuel pulverization, and the resulting transient fission gas release were tested and implemented in the Simple Integrated Fission Gas Release and Swelling (Sifgrs) model or tightly coupled to it. The new mechanistic model was then compared to an empirical model developed in parallel by a Nuclear Energy University Program project using a Studsvik high-burnup loss-of-coolant-accident assessment case. Finally, the report details the preliminary BISON results for a benchmark activity organized by the Nuclear Energy Agency to evaluate fuel performance codes' predictive capabilities for burst fission gas release. This work represents an important step toward a mechanistic understanding of fission gas release in high-burnup conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Modelling the Liquid Waste Operation at the Savannah River Site

Modeling Successes Dounreay – Site Closure Optimization and Risk Reduction Re-structuring and streamlining a complex project plan to achieve site decommissioning in reduced time and at reduced cost, through automated task prioritization. Idaho AMWTP – Multi-stream Optimization Speeding up a complex facility with many process lines, by utilizing all of the process lines more efficiently and reducing overall plant downtime, by coordinating repairs and shutdowns. AWE –Waste Repackaging and Relocation Relocating schedule for waste material from an old storage facility to a new purpose-built one, safely, and in a timely fashion that removed the need to reassure the safety of the old facility. Sellafield – Effluent, Sludge and HAW Treatment Improving a wide range of chemical engineered processes including aqueous effluent treatment, chemically-reactive sludge management and waste vitrification processing. Strategic Petroleum Reserve – Cavern Management Linking chemistry models of petroleum prediction, through engineering cavern management to financial modeling of gasoline sales to benefit US government

Jung, Andrew W.↗
DOE OSTI2023

Los Alamos National Laboratory Floodplain Assessment for the TA-72 Guard Post 10 Expedited Vehicle Lane Project

The National Nuclear Security Administration (NNSA), a semi-autonomous agency within the U.S. Department of Energy (DOE), is proposing new construction in lower Sandia Canyon at Technical Area (TA) 72 at the Guard Post 10 vehicle inspection station at Los Alamos National Laboratory (LANL). The proposed project is intended to provide expedited inspection times for vehicles transporting time sensitive loads, such as concrete and asphalt. The project activities within the 100-year floodplain include installation of 1) installation of an expedited vehicle inspection lane, 2) relocation of a 4-strand wire fence, 3) possible relocation of at least one power pole, and 4) a laydown area (Figures 1 and 2).

42 ENGINEERING↗
DOE OSTI2023

Demonstration of new fracture criteria based on micro-structure and compare to empirical model and measurements

The US nuclear industry is currently exploring extending the peak rod average burnup limits above the current regulatory limit of 62 GWd/tU. A potential concern for fuel exceeding the burnup limit is the fragmentation during a temperature transient such as a loss-of-coolant accident (LOCA). In the event of cladding failure, the fragmented fuel could relocate and disperse within the reactor. The potential impact of fuel fragmentation, relocation and dispersal (FFRD) on licensing assumptions for burnup extension has focused a great amount of efforts and research on this topic.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Recent High Burnup LOCA Testing at Oak Ridge National Laboratory

To improve fuel cycle economics, increasing the fuel burnup limit in light-water reactors requires a solid technical foundation. Observations from experiments at Halden and Studsvik highlighted severe fuel fragmentation during loss-of-coolant-accident (LOCA) conditions, indicating a need for further technical considerations. These experiments suggest that the fragmentation threshold for high-burnup fuel might be influenced by pre-transient power levels. Consequently, additional LOCA test data are essential to complement existing findings and to deepen our understanding. Oak Ridge National Laboratory’s Severe Accident Test Station has been instrumental in advancing knowledge about high-burnup fuel fragmentation, relocation, and dispersal. This milestone report details two high-burnup LOCA tests designed to evaluate the effects of terminal temperature and grid spacers (or cladding restraints) on fuel fragmentation and relocation susceptibility. In addition, BISON fuel performance modeling and out-of-cell benchmark testing were conducted to better interpret the in-cell test results. These tests were developed in collaboration with fuel vendors to ensure that the results provide valuable data for topical reports and support the Nuclear Regulatory Commission’s review.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

The 'Site of Things': Remediation in the Information Age - 18088

More than just a payload, machines now haul data as well. Wood Environment and Infrastructure's maiden voyage into civil engineering automation at the Port Granby Project Long-Term Waste Management Facility demonstrated enormous potential for the remediation industry. The Port Granby Project will relocate approximately 450,000 cubic meters of historic low-level radioactive waste and marginally contaminated soils, located at an existing waste management facility on the shoreline of Lake Ontario, to a new, engineered aboveground mound to be built about a kilometer north of the current site. Sounds simple, right? Just move a bunch of dirt up the road? Hardly, this is marginally contaminated soil situated near Lake Ontario in a beautiful little village an hour east of Toronto. We must relocate safely, meaning limiting the number of staff in the field and efficiently to deliver according to our agreed upon budget and schedule. As if that wasn't enough we were asked to keep track of where each truckload of dirt was excavated from (with a high degree of accuracy), where it was placed in the new engineered disposal facility, how much it weighed and the amount of radioactivity in each load... a virtual chain of custody for each load. Presented is an overview of the Information Technology that supports the Port Granby Project and how the Data Management System (DMS) is being leveraged to make the project safer, more effective and efficient. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Feasibility study for detection of fuel assemblies state inside sealed dry storage casks using external gamma flux measurements - 25445

The license for the storage of spent nuclear fuel assemblies in dry casks is limited to 40 years. In Germany, a site for a final repository of spent fuel is not yet available. Therefore, this license has to be renewed for prolonged storage period. Currently, there is no experience with dry cask storage worldwide that goes beyond a period of 40 years. Beside regulatory and security aspects, there are questions concerning the long-term integrity of the spent fuel assemblies as this is of relevance for final transportation and reloading to final waste repository casks. Once the cask is filled and sealed, the knowledge about the state of the fuel assemblies is limited. This study investigates the feasibility of using external gamma flux measurements to detect the state of the spent fuel assemblies inside the cask. Monte-Carlo simulations were performed to evaluate the gamma flux distribution outside the sidewall of a cask with all intact fuel assemblies and a cask with one damaged fuel assembly, mimicking a fuel assembly with expanded rods and fuel relocation. The evaluation was performed for different location of the damaged fuel assembly and different storage times. The results of the investigation showed that the intensity and the shape of the gamma flux outside the sidewall of the cask are sufficient to identify the damaged fuel assembly, in case it is located near the inner wall of the cask. The detection of the inner damaged fuel assemblies, however, is less feasible. Due to the self-shielding of the fuel assemblies, the gamma photons which come from the inner fuel assemblies stopped within the fuel basket and assemblies materials. The gamma photons which do survive the fuel assemblies self-shielding reach the inner cask wall with not enough energy to escape the shielding of the thick cast iron wall. In general, it can be concluded that the more the damaged fuel assembly is close to the cask inner walls and that the damage is in a large scale (i.e., significant fuel relocation), the more it can be detected by external gamma flux measurements. (authors)

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

Updates to the Regional Seismic Travel Time (RSTT) Model: 1. Tomography

Abstract A function of global monitoring of nuclear explosions is the development of Earth models for predicting seismic travel times for more accurate calculation of event locations. Most monitoring agencies rely on fast, distance-dependent one-dimensional (1D) Earth models to calculate seismic event locations quickly and in near real-time. RSTT (Regional Seismic Travel Time) is a seismic velocity model and computer software package that captures the major effects of three-dimensional crust and upper mantle structure on regional seismic travel times, while still allowing for fast prediction speed (milliseconds). We describe updates to the RSTT model using a refined data set of regional phases (i.e., Pn, Pg, Sn, Lg) using the Bayesloc relative relocation algorithm. The tomographic inversion shown here acts to refine the previous RSTT public model ( rstt201404um ) and displays significant features related to areas of global tectonic complexity as well as further reduction in arrival residual values. Validation of the updated RSTT model demonstrates significant reduction in median epicenter mislocation (15.3 km) using all regional phases compared to the iasp91 1D model (22.1 km) as well as to the current station correction approach used at the Comprehensive Nuclear-Test-Ban Treaty Organization International Data Centre (18.9 km).

58 GEOSCIENCES↗
DOE OSTI2025

Loss-of-Coolant Accident Analysis of a High-Burnup Pressurized Water Reactor Core Design Using Gadolinia-Doped UO 2

High-burnup and extended enrichment fuels are of interest for extending the cycle lengths of pressurized water reactors from 18 to 24 months. Changes to the fuel design and core loading scheme have potentially significant safety implications due to power distribution effects, reduced thermal conductivity, and/or increased plenum pressures due to additional burnable poison loadings. Additionally, higher burnups result in increased material degradation and risk of fuel fragmentation, relocation, and dispersal (FFRD). A representative 24-month core design using gadolinia-doped UO 2 was analyzed for performance under large-break (LB) loss-of-coolant accident (LOCA) conditions using PARCS, RELAP5-3D, and BISON. Furthermore, all considered acceptance criteria were met, with no cases exceeding the 1477 K maximum cladding temperature or the post-quench ductility oxidation limit of 17% equivalent cladding reacted. Full-core FFRD susceptibility was estimated to be approximately 455 kg, though high uncertainties exist with current approaches for computing susceptibility. Undoped fuel rods are more likely to be limiting due to higher linear heat rates. Relatively high burnup and linear heat rate rods located in second batch assemblies are of greatest safety significance during LB LOCA for this high-burnup core design.

High burnup↗