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Source scaling comparison and validation in Central Italy: data intensive direct S waves versus the sparse data coda envelope methodology

SUMMARY Robustness of source parameter estimates is a fundamental issue in understanding the relationships between small and large events; however, it is difficult to assess how much of the variability of the source parameters can be attributed to the physical source characteristics or to the uncertainties of the methods and data used to estimate the values. In this study, we apply the coda method by Mayeda et al. using the coda calibration tool (CCT), a freely available Java-based code (https://github.com/LLNL/coda-calibration-tool) to obtain a regional calibration for Central Italy for estimating stable source parameters. We demonstrate the power of the coda technique in this region and show that it provides the same robustness in source parameter estimation as a data-driven methodology [generalized inversion technique (GIT)], but with much fewer calibration events and stations. The Central Italy region is ideal for both GIT and coda approaches as it is characterized by high-quality data, including recent well-recorded seismic sequences such as L'Aquila (2009) and Amatrice–Norcia–Visso (2016–2017). This allows us to apply data-driven methods such as GIT and coda-based methods that require few, but high-quality data. The data set for GIT analysis includes ∼5000 earthquakes and more than 600 stations, while for coda analysis we used a small subset of 39 events spanning 3.5 < Mw < 6.33 and 14 well-distributed broad-band stations. For the common calibration events, as well as an additional 247 events (∼1.7 < Mw < ∼5.0) not used in either calibration, we find excellent agreement between GIT-derived and CCT-derived source spectra. This confirms the ability of the coda approach to obtain stable source parameters even with few calibration events and stations. Even reducing the coda calibration data set by 75 per cent, we found no appreciable degradation in performance. This validation of the coda calibration approach over a broad range of event size demonstrates that this procedure, once extended to other regions, represents a powerful tool for future routine applications to homogeneously evaluate robust source parameters on a national scale. Furthermore, the coda calibration procedure can homogenize the Mw estimates for small and large events without the necessity of introducing any conversion scale between narrow-band measures such as local magnitude (ML) and Mw, which has been shown to introduce significant bias.

Morasca, Paola (ORCID:0000000265254867)↗

SU-E-I-62: Assessing Radiation Dose Reduction and CT Image Optimization Through the Measurement and Analysis of the Detector Quantum Efficiency (DQE) of CT Images Using Different Beam Hardening Filters

Purpose: Reducing patient dose while maintaining (or even improving) image quality is one of the foremost goals in CT imaging. To this end, we consider the feasibility of optimizing CT scan protocols in conjunction with the application of different beam-hardening filtrations and assess this augmentation through noise-power spectrum (NPS) and detector quantum efficiency (DQE) analysis. Methods: American College of Radiology (ACR) and Catphan phantoms (The Phantom Laboratory) were scanned with a 64 slice CT scanner when additional filtration of thickness and composition (e.g., copper, nickel, tantalum, titanium, and tungsten) had been applied. A MATLAB-based code was employed to calculate the image of noise NPS. The Catphan Image Owl software suite was then used to compute the modulated transfer function (MTF) responses of the scanner. The DQE for each additional filter, including the inherent filtration, was then computed from these values. Finally, CT dose index (CTDIvol) values were obtained for each applied filtration through the use of a 100 mm pencil ionization chamber and CT dose phantom. Results: NPS, MTF, and DQE values were computed for each applied filtration and compared to the reference case of inherent beam-hardening filtration only. Results showed that the NPS values were reduced between 5 and 12% compared to inherent filtration case. Additionally, CTDIvol values were reduced between 15 and 27% depending on the composition of filtration applied. However, no noticeable changes in image contrast-to-noise ratios were noted. Conclusion: The reduction in the quanta noise section of the NPS profile found in this phantom-based study is encouraging. The reduction in both noise and dose through the application of beam-hardening filters is reflected in our phantom image quality. However, further investigation is needed to ascertain the applicability of this approach to reducing patient dose while maintaining diagnostically acceptable image qualities in a clinical setting.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

AGR 3/4 Irradiation Test Final As Run Report

Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5×1025 n/m2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the temperature in the twelve capsules relatively flat in a range of temperatures suitable for the measurement of fission product diffusion in compact matrix and structural graphite materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR 3/4 Irradiation Test Final As Run Report, Revision 1

Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5 × 10 25 n/m 2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the temperature in the twelve capsules relatively flat in a range of temperatures suitable for the measurement of fission product diffusion in compact matrix and structural graphite materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Status of SAS4A/SASSYS-1 Software Development and Application (FY2024)

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over fifty years. It has been identified as a critical element of safety analysis capabilities for the U.S. Department of Energy and is utilized within industry to perform the transient safety analyses required to support the licensing of Liquid Metal-cooled Fast Reactors (LMFRs). This report summarizes the code development and update activities carried out during FY2024. In FY2024, programmatic activities focused on key improvements to software useability, such as enhanced user interfaces for reactivity feedback modeling, improvements in stability/useability of the Code Manual, and improvements to the acceptance testing infrastructure, including automation of acceptance testing and generation of the Acceptance Testing Report. To support end user applications, an open training was held, a semi-public forum was maintained, and a practical benchmarking and validation matrix was developed which allowed limitations of existing testing capabilities to be assessed. The existing fuel models were also enhanced with improved modeling capabilities and testing for the oxide and annular fuel models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Post-DTL Beam Delivery

Ensuring that the beam delivered from the upgraded Front-End (FE) meets the Key Performance Parameters (KPPs) at each user facility is critical to the success of the LANSCE Accelerator Modernization Project (LAMP). For a high-intensity, multi-user facility like LANSCE, compliance with beam loss and radiation thresholds is as important as the charge delivered to each target. While early LAMPF/LANSCE operations relied on iterative tuning to minimize losses from beam halo and tail particles, the new FE may introduce different beam distributions and loss modes—making predictive modeling essential. To manage this, the F2E (Front-End to End) effort is developing detailed particle-tracking models that reflect realistic beamline conditions, including halo formation and expected diagnostic readings. These "snapshot" simulations aim to benchmark live machine performance at a given moment. This will help quantify how beam quality from the new FE will propagate downstream through the facility. Only by validating these models can we confidently assess and mitigate the potential impacts of the LAMP FE on beam delivery. Post-DTL, the beam splits to serve five major user facilities. Historically, low-energy beam transport has been modeled using TRACE, and higher-energy sections with TRANSPORT. These have now been unified into MAD-X format and validated with codes such as Elegant, pyOrbit, XSuite, Impact-Z, and HPSim. The primary focus now is on accurate modeling of full particle distributions (including beam halo) as they traverse the accelerator and beamlines to each experimental station. All models are at various stages of validation with empirical data.

43 PARTICLE ACCELERATORS↗

A New Analytic-Adaptive Model for EGS Assessment, Development and Management Support

To increase understanding of the energy extraction capacity of Enhanced Geothermal System(s) (EGS), a numerical model development and application project is completed. The general objective of the project is to develop and apply a new, data-coupled Thermal-Hydrological-Mechanical-Chemical (T-H-M-C) model in which the four internal components can be freely selected from existing simulation software without merging and cross-combining a diverse set of computational codes. Eight tasks are completed during the project period. The results are reported in five publications, an MS thesis, twelve quarterly, and two annual reports to DOE. Two US patents have also been issued during the project period, with one patent application originated prior to the start of the project. The “Multiphase Physical Transport Modeling Method and Modeling System” (U.S. Patent 8,396,693 B2, 2013), a key element in the GHE sub-model solution, is successfully used for EGS studies. The “Geothermal Energy Extraction System and Method" invention (U.S. Patent 8,430,166 B2, 2013) originates from the time of project performance, describing a new fluid flow control solution. The new, coupled T-H-M-C numerical model will help analyzing and designing new, efficient EGS systems.

15 GEOTHERMAL ENERGY↗

A New Analytic-Adaptive Model for EGS Assessment, Development and Management Support

To increase understanding of the energy extraction capacity of Enhanced Geothermal System(s) (EGS), a numerical model development and application project is completed. The general objective of the project is to develop and apply a new, data-coupled Thermal-Hydrological-Mechanical-Chemical (T-H-M-C) model in which the four internal components can be freely selected from existing simulation software without merging and cross-combining a diverse set of computational codes. Eight tasks are completed during the project period. The results are reported in five publications, an MS thesis, twelve quarterly, and two annual reports to DOE. Two US patents have also been issued during the project period, with one patent application originated prior to the start of the project. The “Multiphase Physical Transport Modeling Method and Modeling System” (U.S. Patent 8,396,693 B2, 2013), a key element in the GHE sub-model solution, is successfully used for EGS studies. The “Geothermal Energy Extraction System and Method" invention (U.S. Patent 8,430,166 B2, 2013) originates from the time of project performance, describing a new fluid flow control solution. The new, coupled T-H-M-C numerical model will help analyzing and designing new, efficient EGS systems.

15 GEOTHERMAL ENERGY↗

Development of an Efficient Approach to Perform Neutronics Simulations for Plutonium-238 Production

Conversion of 238Pu decay heat into usable electricity is imperative to power National Aeronautics and Space Administration (NASA) deep space exploration missions; however, the current stockpile of 238Pu is diminishing and the quality is less than ideal. In response, the US Department of Energy and NASA have undertaken a program to reestablish a domestic 238Pu production program and a technology demonstration sub-project has been initiated. Neutronics simulations for 238Pu production play a vital role in this project because the results guide reactor safety-basis, target design and optimization, and post-irradiation examination activities. A new, efficient neutronics simulation tool written in Python was developed to evaluate, with the highest fidelity possible with approved tools, the time-dependent nuclide evolution and heat deposition rates in 238Pu production targets irradiated in the High Flux Isotope Reactor (HFIR). The Python Activation and Heat Deposition Script (PAHDS) was developed specifically for experiment analysis in HFIR and couples the MCNP5 and SCALE 6.1.3 software quality assured tools to take advantage of an existing high-fidelity MCNP HFIR model, the most up-to-date ORIGEN code, and the most up-to-date nuclear data. Three cycle simulations were performed with PAHDS implementing ENDF/B-VII.0, ENDF/B-VII.1, and the Hybrid Library GPD-Rev0 cross-section libraries. The 238Pu production results were benchmarked against VESTA-obtained results and the impact of various cross-section libraries on the calculated metrics were assessed.

depletion↗

Transient performance of accident tolerant fuel in design basis accidents

A set of transient analysis has been performed to assess the performance of candidate accident tolerant fuel (ATF) against reference UO{sub 2} fuel with Zircaloy cladding under design basis accident (DBA) conditions. This study examined two ATF options, U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding. The accident analyses were conducted using the system code TRACE for a three-loop PWR plant. PARCS standalone full-core calculations were done for each fuel/cladding combination to provide inputs to the TRACE point-kinetics model. The design basis accidents analyzed were a reactivity insertion accident (RIA) and a large break loss-of-coolant accident (LBLOCA). The RIA was simulated by the insertion of a reactivity ramp of $1 (one dollar) in 0.1 s at hot full power (HFP) condition. The LBLOCA was a simulation of a double-ended pipe break in one of the cold legs of the primary loops. Differences in thermal responses of the fuel and the cladding are attributed to their thermal properties. The U{sub 3}Si{sub 2} fuel with its lower volumetric heat capacity and higher thermal conductivity (i.e. higher thermal diffusivity) generally results in a lower fuel temperature than the UO{sub 2} fuel and a correspondingly faster thermal response. Additional calculations for the U{sub 3}Si{sub 2} fuel shows that in a RIA the cladding material, SiC versus Zr, has little impact on the maximum fuel centerline temperature and the PCT (peak cladding temperature). An increase in the hot rod power factor is shown to raise the centerline fuel temperature in a RIA but has little impact on the PCT. The analysis results show that for a relatively fast decay heat driven transient, such as the LBLOCA, the thermal-hydraulics and fuel/cladding thermal properties dictate the overall response more so than the reactivity feedback. The results show that the PCTs in a LBLOCA are similar for all fuel/cladding combinations considered in this study. The LBLOCA analysis also notes the impact of the asymmetry in the system response. The fuel rods in the sector of the core connected to the hot leg of the broken loop (Sector 1 in the PWR model) is shown to have a much lower PCT than fuel rods in the other five sectors of the core. Results of this study indicate that for the two DBAs considered (RIA and LBLOCA) the transient behavior and the resulting peak cladding temperature of the two ATF options (U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding) are similar to the standard light-water reactor fuel (UO{sub 2} fuel with Zircaloy cladding). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Post-Remediation Radiological Dose Assessment, Painesville Site, Painesville, Ohio

A post-remediation radiological dose assessment was conducted for the Formerly Utilized Sites Remedial Action Program (FUSRAP) Painesville Site by using the measured residual concentrations of the radionuclides of concern following the completion of remedial action. The site’s FUSRAP-related constituents of concern (COCs) are radionuclides associated with past activities conducted by the Manhattan Engineer District (MED) and Atomic Energy Commission (AEC) in support of the Nation’s early atomic energy and weapons program, and include radium-226 (Ra-226), thorium-230 (Th-230), thorium-232 (Th-232), and total uranium (Total U). Remedial actions were conducted at the site in accordance with the Final Record of Decision for the Painesville Site, Painesville, Ohio (ROD) (USACE 2006). In the ROD, the U.S. Army Corps of Engineers (USACE) determined that the U.S. Nuclear Regulatory Commission (NRC) standards for decommissioning of licensed facilities found in Title 10, Part 20.1402 of the Code of Federal Regulations (10 CFR 20.1402) and in the Ohio Administrative Code (OAC) 3701:1-38-22(B) are applicable or relevant and appropriate requirements (ARARs) for the site. On the basis of the ARAR analysis, a total effective dose equivalent (TEDE) goal of 25 millirem per year (mrem/yr) was assumed for the site with a construction worker considered as the average member of the critical group. Derived concentration guideline levels (DCGLs) for residual radioactivity of the COCs in soil were developed for the Painesville Site using the RESRAD (Version 6.22) computer code (Yu et al. 2001), which was designed to estimate radiation doses and risks from RESidual RADioactive materials. The DCGLs represent activity levels above site average background activity corresponding to the 25-mrem/yr criterion for a construction worker scenario. The DCGLs are above the background activity concentrations and represent average concentration guidelines. The DCGLs and average background activity concentrations are provided in Table 1. The Final Feasibility Study Addendum (USACE 2005) provides a complete description of the derivation of the DCGLs for the Painesville Site. A post-remediation radiological dose assessment was performed to determine the potential radiation doses to a construction worker, consistent with the approach used to develop the DCGLs for the Painesville Site given in the Final Feasibility Study Addendum (USACE 2005) and ROD (USACE 2006). A second more conservative dose assessment was performed to determine the potential radiation doses to an urban resident who is assumed to live and garden on the site. To further evaluate the long-term protectiveness of the remedial action, a third dose assessment was conducted to determine the potential radiation doses to a hypothetical subsistence farmer. The primary difference between exposure to an urban resident versus a conservative subsistence farmer would be an increase in the amount of foodstuffs (homegrown produce, milk, and beef) produced/cultivated and consumed on-site.

54 ENVIRONMENTAL SCIENCES↗

Assessment of Screen-Covered Grooved Sodium Heat Pipes for Microreactor Applications

New nuclear reactor designs that incorporate heat pipes are being investigated for possible near-term deployment in terrestrial applications. Here, this study explores the use of screen-covered axially grooved sodium heat pipes and their applicability for providing heat removal for microreactors. A sodium working fluid is appropriate for microreactors operating in the 5 to 20 MW(thermal) range at approximately 650°C. HTPIPE, a legacy software code, was validated for the case of screen-covered grooves and used to perform steady-state analyses to determine the performance limits of a proposed heat pipe design. The performance limits of a sodium heat pipe with a screen-covered square grooved wick structure is compared to that of an equivalent heat pipe with an annular wick. In a horizontal orientation at an operating temperature of 650°C,the performance limits for the heat pipe with an annular wick configuration are 15% higher than for the screen-covered grooved wick. At operating temperatures below 777°C, the annular wick outperforms the screen-covered grooved wick, and at temperatures above 777°C, the screen-covered grooved wick outperforms the annular wick. However, the marginal performance gain at higher temperatures may not justify the use of heat pipes with a screen-covered grooved wick structure due to increased manufacturing costs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Developments in SRW Code and Sirepo Framework Supporting Simulation of Time-Dependent Coherent X-ray Scattering Experiments

Physical optics simulations for beamlines and experiments are essential for the effective use of synchrotron light source facilities such as NSLS-II at BNL. The SRW software package supports such source-to-detector simulations for coherent X-ray scattering and imaging experiments through its Python interface and Sirepo browser-based graphical user interface. This allows one to define custom sample models, assess the feasibility of an experiment, and estimate most appropriate beamline settings before using valuable beamtime. We discuss the recent use of general-purpose GPU resources and coherent mode decomposition algorithms in SRW to accelerate physical optics simulations with partially coherent X-rays. To illustrate these new capabilities, we describe simulations of typical time series of partially coherent scattering images used in X-ray Photon Correlation Spectroscopy (XPCS) experiments; aiming to characterize the nanoscale dynamics of a disordered sample, representing a solution of nanoparticles undergoing Brownian diffusion.

36 MATERIALS SCIENCE↗

High-Burnup LOCA Burst Susceptibility BISON Analysis in PWRs and BWRs

Accurately assessing high-burnup fuel behavior during loss-of-coolant accidents (LOCAs) is essential for understanding fuel fragmentation, relocation, and dispersal (FFRD) risks across the US light-water reactor fleet. This work updates previous Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program multiphysics LOCA analyses for a pressurized water reactor (PWR) and a boiling water reactor (BWR) by incorporating recent model and material property advancements in the BISON fuel performance code, including a high-burnup structure (HBS) model, revised cladding burst criteria, and updated thermal–mechanical correlations. This update was needed to support ongoing industry initiatives and upcoming regulatory changes. Full-core, rod-resolved operating histories generated using Virtual Environment for Reactor Analysis (VERA) and system-level LOCA conditions obtained from TRACE were applied to statistically representative rod samples in BISON to evaluate burst behavior and FFRD susceptibility. These calculations used two cladding burst correlations and three fuel pulverization models so that the predictions of these models could be compared. The updated PWR simulations show markedly improved numerical stability as the number of crashed simulations decreased by 95% compared to the previous study, and hence higher confidence in results. The updated PWR simulations predicted cladding bursts exclusively among once-burned, high-power rods, with two different cladding burst models identifying the same burst-susceptible population. Resulting FFRD susceptibility estimates are significantly reduced compared with earlier studies, driven by cooler predicted fuel and plenum temperatures, lower hoop strains, and reduced fission gas release in the updated models. In contrast, none of the BWR rods were predicted to burst under either burst criterion, reaffirming minimal BWR FFRD susceptibility even with updated HBS and material models. Comparisons between the PWR and BWR end-of-cycle predictions are made. Comparison with prior work highlights significant shifts in PWR fuel performance metrics and confirmation of earlier BWR conclusions. Overall, the updated results underscore the importance of having high-resolution detailed modeling capability and continuously integrating evolving material models and physics into high-resolution multiphysics simulations. The unified assessment presented here strengthens confidence in predicting high-burnup LOCA behavior by improving agreement between different cladding burst correlations. These results also provide an improved foundation for future BISON model development, FFRD susceptibility calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Enhancing Operational Performance and Productivity Benefits in Breweries through Smart Manufacturing Technologies

This article uses case studies to demonstrate the potential of smart manufacturing (SM) and Internet of Things (IoT) technologies to enhance operational performance and productivity in industry. The analysis highlights benefits such as cost reduction, production flexibility, shorter product time-to-market, energy/water efficiency and environmental impact reduction, and increased productivity. To illustrate the effectiveness of SM and IoT approaches, the authors sought out manufacturers currently implementing or seeking to implement SM and IoT technologies. The authors identified beer brewing (NAICS Code 312120) as a rapidly expanding industry whose members appear eager to implement SM technologies to optimize production lines by revealing bottlenecks and identifying performance-reducing nodes. This article presents two case studies based on SM and IoT technologies in breweries. It briefly describes a systematic framework introduced elsewhere by the authors and uses it to assess the energy productivity and competitiveness of SM applications in breweries. This article addresses questions concerning the information and communications technology infrastructure needed to build smart breweries and how corporations simplify the installation and deployment of SM and IoT components. This article was originally published in the Proceedings of the 2019 ACEEE Summer Study on Energy Efficiency in Industry.

42 ENGINEERING↗

In-situ thermomechanical measurements during burst testing of Zr claddings using digital image correlation and infrared thermography techniques

Burst testing has long been a relied-on method for assessing fuel cladding materials in light-water reactoraccident scenarios. However, burst testing historically provided minimal in-situ observations beyond a few thermocouple readings, pressure from transducers, and post-test deformation measurements. Acquisition of full cladding thermomechanical data throughout a burst test would be invaluable to fuel system qualification, material model development and fuel performance code validation, such as the BISON fuel performance code. Recently, methodologies for simultaneously applying 2-dimensional digital image correlation and infrared thermography techniques to quantify the relationship between temperature, internal cladding pressure, and deformation during burst testing were developed. This work details results from the application of these techniques to burst testing of Zircaloy-4 cladding segments subjected to loss-of-coolant accident transients in an air environment. Two types of tests were performed with: (a) cladding segments connected to a fixed pressure reservoir as well as (b) closed systems where internal pressure increased during transients. In-situ thermomechanical measurements including strain and thermal gradient data are presented, demonstrating the effectiveness of applying combined surface measurement techniques to burst testing

Bell, Sam↗

Final Technical Report for Year 5 Early Career Research Project "Viscosity and equation of state of hot and dense QCD matter"

The Section below summarizes research activities and achievements during the fifth (last) year of the PI’s Early Career Research Project (ECRP). Unlike the first four years of the project, the last year was not funded under the American Recovery and Reinvestment Act (ARRA). The ECRP advanced two main areas: i) radiative 3 ↔ 2 radiative transport, via development of a new computer code MPC/Grid that solves the Boltzmann transport equation in full 6+1D (3X+3V+time); and ii) application of relativistic hydrodynamics, via development of a self-consistent framework to convert viscous fluids to particles. In Year 5 we finalized thermalization studies with radiative gg ↔ ggg transport (Sec. 1.1.1) and used nonlinear covariant transport to assess the accuracy of fluid-to-particle conversion models (Sec. 1.1.2), calculated observables with self-consistent fluid-to-particle conversion from realistic viscous hydrodynamic evolution (Secs. 1.2.1 and 1.2.2), extended the covariant energy loss formulation to heavy quarks (Sec. 1.4.1) and studied energy loss in small systems (Sec. 1.4.2), and also investigated how much of the elliptic flow could have non-hydrodynamic origin (Sec 1.3). Years 1-4 of the ECRP were ARRA-funded and, therefore, they have their own report document ’Final Technical Report for Years 1-4 of the Early Career Research Project “Viscosity and equation of state of hot and dense QCD matter”’ (same award number DE-SC0004035). The PI’s group was also part of the DOE JET Topical Collaboration, a multi-institution project that overlapped in time significantly with the ECRP. Purdue achievements as part of the JET Top- ical Collaboration are in a separate report “Final Technical Report summarizing Purdue research activities as part of the DOE JET Topical Collaboration” (award DE-SC0004077).

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Risk-Informed Mechanistic Source Term Calculation (Final CRADA Report)

As part of an FY18 U.S. Department of Energy (DOE) Gateway for Accelerated Innovation in Nuclear (GAIN) award, a CRADA agreement was established between Argonne National Laboratory (Argonne) and Oklo. The CRADA focused on the following three tasks to assist Oklo with the regulatory licensing of their innovative nuclear reactor design: • Task 1 – Coupling and Demonstration Analysis using The Nek5000, Diablo, and OpenMC Codes: Establishing a coupling between the neutronic, structural, and thermohydraulic analysis codes and utilizing the code package for a high-fidelity demonstration analysis of radial core expansion due to temperature changes and its impact on core reactivity. • Task 2 – Mechanistic Source Term Analysis: Performing a realistic evaluation of the potential release of radionuclides to the environment during postulated transient event sequences. • Task 3 – Data in Support of Metal Fuel Qualification: An assessment of past postirradiation examination data of metal fuel utilized in the Argonne EBR-II reactor to assist in the potential qualification of the fuel typer for regulatory licensing. The details regarding each analysis and the associated conclusions were provided to Oklo in three separate project reports. Due to the utilization of proprietary and export controlled information, the current summary report was created, which provides an uncontrolled overview of each of the three tasks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗