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At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Radiation Hardened Foam Cold Test Plan - Phase II: Foam Characterization Testing and Environmental Chamber Testing of FoamBag Fixative Foam

This document outlines the Phase II test objectives and implementation plan for a down-select foam fixative technology intended to facilitate activities in support of the Savannah River Site (SRS) F/H labs deactivation and decommissioning (D&D) efforts. It is a collaborative effort between Savannah River National Laboratory (SRNL), Florida International University (FIU), and the SRS F/H labs team intended to test and evaluate the potential of a polyurethane resin foam in mitigating the release of contamination during dismantling operations on radioactively contaminated piping in legacy facilities. The cold test plan addresses specific requirements highlighted by site and safety personnel and will be executed at FIU and SRNL test and lab facilities. Results from the cold tests will inform the hot test at F/H labs, which will use the foam fixative to confine and/or isolate residual contamination within a 3-dimensional void space of Hastelloy C-22 piping designated for removal from the area and transported to a designated disposal facility. Phase I testing was previously conducted using Hilti CP620 fixative foam. Results from Phase I testing indicate Hilti CP-620 fixative foam is incompatible with SRS site hot taps as an effective foam delivery method into Hastelloy C-22 piping. Phase II testing will be conducted using FoamBag TM fixative foam as an alternative fixative foam option. Phase II testing will address eight test objectives: (1) evaluation of the adhesion and bonding properties of FoamBag TM in Hastelloy C-22 piping, (2) evaluation of the adhesion of FoamBag TM in piping under various moisture conditions, (3) determination of the heat profile of FoamBag TM during curing, (4) determination of the internal pipe pressure after FoamBag TM deployment and curing, (5) conduct a leak test to determine if FoamBag TM is effective at creating a full seal within piping, (6) headspace testing of FoamBag TM to determine if there are any associated off-gas hazards during FoamBag TM curing, (7) conduct environmental chamber testing of FoamBag TM to understand how environmental parameters impacts FoamBag TM curing, and (8) conduct fire testing to evaluate FoamBag TM fire retardant characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Characterization of Aluminum Metal Powders for Plutonium-238 Program

As part of radioisotope thermoelectric generator (RTG) production for power sources in deep space exploration, Oak Ridge National Laboratory (ORNL) produces plutonium-238 ( 238 Pu). To produce 238 Pu, neptunium-237 ( 237 Np) targets are fabricated at ORNL and subsequently irradiated at the High Flux Isotope Reactor (HFIR) at ORNL and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Specifically, current flowsheets utilize neptunium dioxide ( 237 NpO 2 ) targets. Aluminum (Al) powder is blended with 237 NpO 2 target material to improve thermal conductivity during irradiation. The current composition of 237 NpO 2 /Al pellets (i.e., cermets, or ceramic–metallics) is 20% 237 NpO 2 , 70% Al, and 10% void space. Al powders utilized by the 238 Pu program are high-fired under vacuum before blending into targets. The primary objective of this work is to measure the physical, chemical, and thermal properties of Al before and after the high-firing process. Properties of interest include the morphology, surface texture, particle size, crystal phase(s), and thermal conductivity of the material. These properties are measured with existing materials characterization equipment at ORNL, including powder x-ray diffraction (pXRD), scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS), laser flash analysis (LFA), and thermogravimetric analysis (TGA). A secondary objective of this work is to compare the Al powders before and after the high-firing procedure to determine the effects of high-firing on chemical, physical, and thermal properties and to determine the efficacy of the high-firing process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Multidimensional Diagnostics of the Interface Evolutions in Solid-State Lithium Batteries

All-solid-state lithium metal batteries hold tremendous potential as next-generation energy storage solutions. However, their performance is hindered by complex chemo-mechanical instabilities at the interface, which encompass electrolyte decomposition, loss of contact due to volume changes, and the growth of lithium filaments. Given the complexity of these failure mechanisms, it is imperative to employ multidimensional diagnostics that leverage a combination of tools capable of quantifying formed voids and filaments, identifying the chemical and mechanical properties of lithium dendrites and electrolyte decomposition products, as well as monitoring the evolution of these processes in situ. The primary objective of this project is to develop a comprehensive characterization platform for solid-state battery interfaces, integrating FIB-SEM tomography, ToF-SIMS, in-SEM nanoindentation, and in-SIMS AFM-based stiffness mapping to enable structural, chemical, and mechanical analyses. The project team has accomplished significant milestones, including: 1) The successful development of a characterization platform that combines FIB-SEM, ToF-SIMS, nanoindentation, and AFM. 2) The creation of solid-state cells at various scales. 3) Real-time visualization of the plating and stripping of lithium. 4) Operando analysis of the chemical composition and evolution of electrolyte decomposition products, encompassing intermediate and metastable species, at both the cathode and anode interfaces. 5) In situ analysis of mechanical properties coupled with chemical mapping. 6) The implementation of the multiscale characterization technique has yielded valuable insights into cell failure mechanisms and has led to the development of a solid-state lithium cell that exhibits exceptional durability, surpassing 2500 cycles. This project has resulted in 10 publications, 1 invention disclosure, 8 manuscripts under review or in preparation.

25 ENERGY STORAGE↗
DOE OSTI2023

In-situ and ex-situ characterization of ion-irradiated AM materials

Additive manufacturing (AM) has attracted increasing attention in recent years as a new way of making high-quality components for nuclear reactors. While AM materials are compositionally similar to their conventionally produced counterparts, they do possess different microstructures, such as dislocation cells and chemical inhomogeneity, that can lead to different mechanical properties and performance behavior. In this study, the irradiation response of AM materials was investigated. In-situ and ex-situ ion irradiations were performed on AM316L and AM316H stainless steels (SS) at 300 and 600°C. The influence of the dislocation cell structure on the evolution of irradiation-induced dislocation loops was evident at 600ºC, but was much weaker at 300ºC. No voids were observed with the in-situ ion irradiation up to 10 dpa at both temperatures. Post-irradiation energy dispersive spectroscopy showed radiation-induced segregation (RIS) near grain boundaries and the formation of Cr-rich oxides throughout the matrix. The extent of segregation at dislocation cell walls varies with dose. Nanoindentation tests performed on the AM316L SS irradiated at 600ºC showed a complex dose dependence with softening at low doses and hardening at high doses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Cyberguardians and STEM Warriors (Final Technical Report (FTR))

In the past decade, solar power, with and without energy storage has become the fastest growing source of energy generation in the world. In the U.S., solar employment more than doubled from 105,145 jobs in 2011 to 255,037 jobs in 2021, four times faster than the U.S. job growth rate overall. These factors, combined with technology advancements, creates a skills gap that puts tremendous stress on society to deliver the workers to fill the open job requisitions. The Cyberguardians and STEM Warriors project (Cyberguardians) was designed to address the trained-worker shortage in the energy industry in three ways: 1) by developing educational curriculum that addresses DER technology changes; 2) by delivering curriculum to prospective workers, including military veterans and their families, via universities, community colleges, and vocational training outlets; and 3) introducing individuals who have completed training to employers that can hire them. Cyberguardians exceeded its curriculum goals by producing 27 academic units of university-accredited material (12 total courses) covering energy fundamentals, smart inverters, Distributed Energy Resource (DER) data communication, cybersecurity, standardization, certification, data analytics, and IEEE 1547 standard topics. The North American Board of Certified Energy Practitioners (NABCEP) also accredited the material for use in their credential program. Seven instructors were recruited and trained, and six academic institutions (University of California San Diego, State University of New York, North Carolina State University, Harper Community College, Green Village Academy, and the SunSpec Alliance) were enlisted, meeting program goals. All course material was published under the Creative Commons license and made available royalty free, thus providing a long-lasting public benefit. The program’s outreach program vastly exceeded program goals and incorporated the efforts of 13 outreach partners (11 of which are veteran focused), an advisory board representing 15 companies, webinars and 10’s of thousands of email messages sent to prospective students and hiring managers. Despite these efforts, the global pandemic depressed anticipated program participation by about a third. Still, a total of 396 students enrolled and 289 completed the courses and were accredited. The job applicant task achieved similar results (111 realized vs a 174 goal) but reported job placement was weaker at (9 realized vs. a 51 goal). The Cyberguardians program fills a critical void for cost-effective, royalty-free curriculum and training pertaining to DER technologies and cybersecurity that prospective energy workers must possess to be effective in the 21 st century. On this basis alone, the investment of taxpayer funds will pay dividends for years to come.

14 SOLAR ENERGY↗
DOE OSTI2023

Generalized kinetic model for defect evolution in irradiated materials

A fundamental and open problem in materials science is to determine how the structural properties of irradiated and self-irradiated materials evolve in time. Structural defects generated by irradiation lead to changes in a material’s macroscopic properties, and these macroscopic changes, such as volumetric swelling, can in turn give rise significant alterations of the functionality of a material. Here, we develop a unified analytical model to understand the complex interplay between aggregation, fragmentation, and recombination processes involving mobile clusters of defects such as bubbles, voids, and interstitial atoms in materials subject to radiative sources. Specifically, we employ a mean field approach to derive a system of coupled kinetic equations that describes both the time evolution of the density of each type of defect and how clusters of each defect form and grow as the material ages.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Probing Supercritical Phase Transition using Ultrafast X-ray Diagnostics (Final Technical Report)

Over recent years, several hypotheses have been put forward to elucidate supercritical transition states. Despite significant progress on the fundamental understanding of fluids at these supercritical conditions, important questions concerning the microstructure and dynamical processes remain. A particular research issue is hereby the fundamental understanding of the morphology of the molecular microstructure and its effect on the macroscopic behaviour and thermodynamic response functions. It is widely believed that the supercritical state is homogeneous without structural and dynamic observables to distinguish between a liquid and a vapor. However, recent investigations have identified regions of distinct liquid-like or vapor-like properties even under supercritical conditions. Specifically, it was shown that the transition between liquid-like and vapor-like states occurs across an extension to the coexistence line, marked by almost discontinuous changes in fluid properties. This transition was first identified experimentally by Nishikawa and Tanaka and stands in contrast to the classical presentation of the supercritical state space as a featureless, homogeneous domain. Structurally, the most important properties of supercritical fluids are the dynamic heterogeneities and local density fluctuations that are present within the fluid state at the microscopic level. These inhomogeneities are associated with the formation of molecular clusters of various sizes with liquid-like properties separated by voids of unbound gas-phase molecules which continuously restructure itself at picosecond timescales. By utilizing ultrafast X-ray Photon Correlation Spectroscopy (XPCS) at the Linac Coherent Light Source (LCLS) and X-ray synchrotron measurements as an experimental probe of these structural changes at the molecular level we seek to elucidate the higher-order phase transition from a liquid-like to a vapor-like state in the region extending the critical point. Open questions we seek to address particularly with XPCS are: (i) the underlying mechanisms responsible for the density fluctuations at supercritical conditions and how these dynamical processes at the molecular level affect the thermodynamic response functions and (ii) on what timescales do dynamical processes evolve that are associated with thermalization, molecular diffusion, and intermolecular cluster transfer.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Dual-Purpose Canister Filling Demonstration Project Progress Report at ORNL, 2023

The US DOE Office of Nuclear Energy is investigating the feasibility of direct disposal of dual-purpose canisters (DPCs) in a hypothetical geological repository to offset the potential requirement to repackage spent nuclear fuel (SNF) from existing DPCs into smaller, disposal-ready canisters. Oak Ridge National Laboratory (ORNL) is currently evaluating the feasibility of filling void space in loaded DPCs with an engineered material to prevent a criticality event caused by groundwater/moderator intrusion. Metal alloys are being investigated as a filler material because of their relatively low viscosities when molten, which may facilitate their injection via an existing drainpipe that runs almost the full length of the DPC. ORNL’s strategy for evaluating filler viability includes simulations and physical demonstrations of filling and casting behavior, as well as evaluations of materials for compatibility in the repository environment. Filling of DPCs in this manner is expected to mitigate the risk associated with a post-closure criticality event during the repository performance assessment time frame (10,000 years or greater). Efforts in this fiscal year focused on (1) destructive analysis of experimental filler castings made in FY 2022, (2) a report outlining a conceptual design of a DPC filling facility (Fortner et al., 2023, M3SF 23OR010305044/ ORNL/SPR-2023/2921, May 31, 2023), (3) a report on affected features, events, and processes (FEPs) due to DPC filling (Price et al., 2023, M3SF-23SN010305093, issuance pending), (4) developing and testing a more practical alloy filler based upon a Sn-Al eutectic, and (5) modeling the heating and cooling dynamics of a DPC subjected to molten metal filling. The preliminary results from each of these tasks support the feasibility of filling DPCs with metal as a strategy against the possibility of criticality in the repository. The FY 2022 casting was found to penetrate even very small orifices in the mold and internal structures. Preliminary testing of the Sn-Al eutectic indicate little interaction of the melt with Zircaloy cladding. Thermal modelling shows that a filled DPC will cool to manageable temperatures within 2-3 days, which is likely manageable in a facility.

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

LightMAT: Continous Fiber Malleable Thermoset Composites with sub-1-minute dwell times; validation of impact performance and evaluation of the efficacy of the compression forming process

Recent major advances in polymer chemistry are enabling a paradigm shift in the area of processing and re-processing/recycling of advanced composite materials. Inexpensive compression molding with short cycle times, traditionally available for thermoplastic composites only, are now possible for the thermoset resins based on polyimine chemistry. Participant will collaborate with the LightMAT consortium Labs ORNL, SNL, and PNNL to non-destructively characterize microscopic structural defects (such as voids and resin-rich areas), and perform high speed impact characterization of its malleable thermoset CFRP's. This work is critical to the validation and adoption of these materials for the automotive market and towards DOE-VTO lightweighting mission objectives.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Scalable Manufacturing of Efficient Perovskite/Silicon Tandem Modules

Crystalline-Si (c-Si) technology produces excellent solar cells with conversion efficiencies of up to 26.7%—nearly their practical limit of 27%—and has been commercialized extensively to produce panels below $\$0.40$/W. Techno-economic analysis shows that module efficiency will continue to be a primary cost driver because of high balance-ofsystems costs, and there is no path to higher efficiencies with single-junction silicon cells. In this project, three research teams from University of North Carolina, Arizona State University and National Renewable National Laboratory worked together to develop perovskite-silicon tandem cells with a high throughput process which can handle 5000 wafers per hour, enabled by very low added CAPEX of ~$20k for a doctor-blade coater. This can potentially increase the module efficiency to 30% (with grid) with small increase of cost, which will drive down the cost of silicon modules to be at least 16% cheaper than present silicon PERC modules. We also explored the alternative narrow bandgap perovskites as silicon replacement for all perovskite tandem cells. This project has substantially advanced the progress of solution-process perovskitessilicon tandem solar cells and perovskite-perovskite tandem solar cells with many inventions and discoveries, evidenced by 18 publications and 6 invention disclosures. Several notable examples include 1) we developed the new idea of small pyramid in combination with solution grown perovskites and demonstrated record-efficiency of 28.5% for 1 cm2 on textured silicon bottom cells and 25.2% for 24 cm2 on chemically etched silicon bottom cells for perovskite-silicon tandem cells. 2) We developed perovskite ink that can coat perovskites onto texture silicon without voids;, and new device structure to enhance the yield of fabrication by reducing shunting 3) We identified the origin of open circuit voltage on mixed halide wide bandgap perovskites, and came with a solution for this problem, reducing the voltage loss to a record small value, which can potentially push the perovskite-silicon tandem cell efficiency to over 31%; 4) We have developed efficient alternative low bandgap semiconductors, i.e. gradient-doped Sn based perovskites by Ba ions, new oxide hole transport layers which increase the efficiency of perovskite/perovskite tandem cells reached record efficiency of 26.3%; 5) We developed bifacial perovskite/perovskite tandem cells by overcoming contacts issues and regaining light absorption, further boosting the equivalent efficiency to 29.3% under 1 sun illumination with 30% albedo light; 6) We developed the first all perovskite tandem module fabricated in air by discovery a combination of oxidization and reduction couples, yield an aperture efficiency of 21.6% for all perovskite mininodules, exceeding that of single junction perovskite minimodules. These discoveries not only accelerate the commercialization of perovskite-silicon tandem solar cells, but also provide guidance in designing other type of perovskite solar cell technologies.

14 SOLAR ENERGY↗
DOE OSTI2020

FAST-1.0: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients: Developed under NQA-1-2017

Fuel Analysis under Steady-state and Transients (FAST) is the US Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation, including creep, fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure and void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heattransfer correlations, for normal operation through postulated accidents for today’s US-based light water reactor (LWR) fuel designs. FAST-1.0 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.0, which is the first official version of this code. This document is one of a series of documents on FAST; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

High Temperature Tensile Testing of Molybdenum Before and After Exposure in Flowing He at 600°C to 1000°C

ORNL supported NorthStar Medical Radioisotope’s accelerator-based concept to produce the medical isotope molybdenum-99 (Mo-99). The Mo targets are expected to experience in the accelerator temperature ranging from 200°C to 1000°C, and previous studies have demonstrated the embrittlement of Mo targets at room temperature when exposed to impure He at 800 to 1000°C. The goal of this project was to evaluate potential embrittlement at temperature ranging from 600°C to 1000°C for three Mo materials exposed at the same temperatures for 100h in high purity He (O 2 < 5ppm (µl/l)) with a flow rate of ~1m/s. Enriched Mo-100 (aMo) and conventional pressed and sintered Mo (UHP) disks were provided by NorthStar while low carbon arc cast rolled Mo, fabricated at ORNL, was used for comparison. The as fabricated rolled Mo exhibited superior tensile strength due to the alloy very fine grain size with limited ductility except at 1000°C with the elongation at rupture reaching 30%. The UHP Mo disks showed the lowest strength but highest ductility at 20-1000°C with significant strain hardening during tensile testing. The ductility of the aMo material was quite low at all temperatures due to a high density of voids at grain boundaries. Significant embrittlement of the aMo material was observed after exposure at 600°C to 1000°C, most likely because of oxygen diffusion and segregation at grain boundaries. No embrittlement was observed for the UHP Mo material, with excellent elongation at rupture for some of the specimens oxidized in He and tested at 600-1000°C. Significant variation in ductility from one UHP Mo specimen to another was, however, measured, both in the as fabricated and He-oxidized conditions. The rolled Mo was not affected by oxidation in He at 800°C, but a decrease in strength and increase in ductility was observed at 1000°C, likely due to the material recrystallization.

36 MATERIALS SCIENCE↗
DOE OSTI2022

FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients (Developed under NQA-1-2017)

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: * heat conduction through the fuel and other materials * heat transfer from the cladding-to-coolant * cladding elastic and plastic deformation, including creep * fuel-cladding mechanical interaction * fission gas release from the fuel * rod internal pressure and void volume * cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.1 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.1 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Accuracy Enhancement of Nuclear Power Plant Simulators Utilizing High Accuracy Simulation Predictions

More recently, reactor core simulators for core designs associated with commercial nuclear power plants that utilize what is believed to be higher fidelity models have been developed. Features such as neutronics models that utilize transport equation solvers with fine spatial meshes and many energy-groups, thermal-hydraulic models that utilize sub-channel solvers with fine spatial mesh and capable of treating a wide range of fluid conditions, and fuel-coolant chemistry interaction models capable of treating CRUD deposition are to be found in these higher fidelity core simulators. These reactor core simulators require access to higher performance computers, characterized by many processors, cores and large memory. So associated with utilization of these simulators is access to high performance computers and ability to accommodate in one’s workflow longer execution times. By contrast, currently used core simulators by the nuclear industry can execute on engineering workstations and have execution times of seconds to minutes. The desirability for having short execution times is not only desired for support of time critical tasks but supports the mental process of decision making by engineers. The goal of the work reported upon here has the objective of retaining the fidelity of higher fidelity models while retaining the ability to utilize engineering workstations. Beyond the core simulator goal, additional goals of this work include incorporating the just described core simulator capability into a Nuclear Steam Supply System (NSSS) simulator, and to incorporate the resulting capability into an environment supportive of design and operational decision making associated with nuclear power stations. The model selected for the core neutronics model is the NESTLE code, for the core thermal-hydraulic model is the CTF code utilizing coarse mesh, and for the NSSS model is the RELAP5-3D code. WSC’s proprietary 3KEYMASTERTM platform is being used to provide software coupling, user interface, visualization, and reporting. The NESTLE core neutronics simulator was first integrated with the CTF core thermal-hydraulic simulator using CTF developed communication commands which are also used for CTF to communicate with RELAP5-3D under WSC’s proprietary 3KEYMASTERTM platform. To assure NESTLE prediction consistency with higher fidelity core neutronic simulators, buffer codes have been created to automatically generate from output files written by the VERA core simulator the NESTLE nodal neutronic parameter’ library, geometry, and pin-power reconstruction input files, thereby avoiding a number of challenges associated with utilizing lattice physics codes and providing consistency with VERA predictions. To treat absorber rod effects a multi-set library is utilized, where a set refers to a specific absorber rod fully inserted pattern. A coarse spatial mesh CTF model was developed with features added that support using CTF as envisioned in the engineering quality simulator. A hybrid meshing approach was implemented to allow for automated construction of models with mixed levels of refinement. Specifically, a core model could resolve some assemblies at a nodal level (4 subchannels per assembly) and others at a pin-resolution (one subchannel per coolant subchannel in the assembly). The intention is that this will allow for better resolution of limiting conditions such as DNBR and PCT, which are based on local rod and subchannel conditions. Further development was done of features that enhance the capabilities for the envisioned engineering quality simulator that has been developed, but now for RELAP-3D. The RELAP5-3D code development includes ability to model more than 999 components and the addition of the cross-channels turbulence mixing model and the void drift model that are implemented in CTF, aiming to achieve closer prediction agreement of the two codes for transient simulations, specifically, more accurate matches of the overall mass, momentum, and energy exchanges of both the liquid and gas phases between the neighboring core assemblies. Graphics were also developed for the Instructor Station for this project under WSC’s proprietary 3KEYMASTERTM platform to facilitate design and operational decision making.

42 ENGINEERING↗
DOE OSTI2024

Improved Creep Testing Approach for Bentonite EBS

In the United States, the nuclear reactor contributes 18.7 percent of the nation’s total electricity generation and produces a significant amount of spent nuclear fuel (SNF) and high-level waste. Because SNF is being stored for longer periods than initially envisioned, the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Science and Technology is assessing the technical performance of the SNF storage systems after extended durations. The concept of a sustainable geologic repository for SNF disposal relies heavily on the safety provided by multiple barriers, such as repository host rock, overlying rock formation, and human-made engineered barrier systems (EBS). One of the major components of EBS is bentonite-based buffer materials that isolate the nuclear-waste canisters from the host rock and fill the void left in the horizontally drilled boreholes. The resaturated bentonite has a swelling characteristic, which is useful for self-sealing the microcracks in the host rock and supporting the canister’s heavy weight. However, factors such as nonuniformity in the saturation, swelling pressure of bentonite, and the high temperature at the canister-bentonite interface may lead to instability of the waste canister inside the borehole. Therefore, evaluating the long-term deformation of bentonite-based material is necessary. RESPEC Company, LLC recently completed a preliminary study for the U.S. Department of Energy (DE SC0022804) that is attempting to understand the time-dependent deformation in the consolidated sand/bentonite (SB) specimens through triaxial creep experiments and improve the understanding of EBS performance under anticipated repository conditions. The project included developing a standard testing procedure for preparing the consolidated core specimen from a mixture of sand and bentonite in a 50:50 ratio by weight, fabricating a new pressure vessel in-house, modifying existing creep equipment to perform triaxial creep experiments on consolidated SB specimens, and performing six long-term triaxial creep experiments under low-deviatoric stress at two different temperature conditions, which is representative of hypothetical conditions that might be encountered in a geologic repository within a reasonable rate of success in completing creep experiments. The consolidated SB specimen was prepared using the novel specimen preparation procedure and had physical characteristics (e.g., moisture content and bulk density) reasonably comparable with the properties of bentonite-based buffer, which is proposed to be used in the EBS in the Swedish KBS-3H design and the Swiss design of SNF disposal in the horizontal drifts. The newly fabricated pressure vessel could withstand the confining pressure of up to 15 megapascals (MPa) while maintaining the test temperature of up to 90 degrees Celsius (°C) for long-term creep experiments. The preliminary results from the triaxial creep experiments suggested that the consolidated SB specimen may experience time-dependent deformation at a low-deviatoric stress state and room temperature condition. Based on the deformation recorded in the SB specimen from axial and radial linear variable differential transformers (LVDTs) and physical characteristic changes, it is apparent that the time-dependent deformation is a combination of consolidation and creep, which is influenced by the level of deviatoric stresses and temperature. The systematic creep testing of bentonite-based specimens helped develop a standard testing procedure applicable for analyzing the similar characteristics of shale or clay-like materials in different geologic conditions outside repository science, including in civil engineering and infrastructure projects, underground and open pit mining, and deep drilling.

58 GEOSCIENCES↗
DOE OSTI2023

Continued Development and Advanced Testing of DPC Filler Cements (on FY22 R&D and Demonstration Activities) (Progress Report)

Commercial generation of energy by nuclear power plants in the United States (U.S.) has produced thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE). Utilities typically utilize the practice of storing this SNF in dual-purpose canisters (DPCs). DPCs were designed, licensed, and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a criticality event during SNF storage and transport, but were not designed or loaded to preclude the possibility of a criticality event during the regulated post-closure period following disposal, which could be up to 1,000,000 years (Price, 2019). There are several options being investigated that could facilitate the disposal of SNF stored in DPCs in a geologic repository (Hardin et al., 2015; SNL 2020b; SNL 2021b). These include: (1) repackage the SNF into canisters that are designed to prevent criticality during the regulated post-closure period following disposal, but with an increased disposal cost estimated at approximately $\$$20B in United States dollars (USD) (Freeze et al., 2019); (2) analysis of the probability and consequences of criticality from the direct disposal of DPCs during a 1,000,000-year post-closure period in several geologic disposal media (Price, 2019); and (3) filling the void space of a DPC with a material before its disposal that significantly limits the potential for criticality over the post-closure regulatory period. This report further investigates the third option, filling DPC already containing SNF with a material to limit the potential for criticality over the post-closure regulatory period.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Boiling-Water Reactor Testing Capability in the Advanced Test Reactor

I-Loop is an irradiation facility that is currently being installed at the Advanced Test Reactor. It is a two-loop test facility capable of performing Light Water Reactor (LWR) irradiations in prototypic coolant conditions. The two loops are being installed to be capable of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) pressure, temperature, and chemistry environments. Each loop is nominally dedicated as a BWR or PWR for simplicity of operations. In-reactor water loop testing that an I-Loop provides is key to the deployment of new accident tolerant fuel technologies and other advanced LWR fuel concepts. Currently, pressurized water loops are the only testing facilities available to test BWR fuel concepts. Their test environments are non-prototypic at higher pressure/temperature and at single-phase fluid flow conditions. This void in the LWR test bed capabilities is one that the I-Loop is uniquely situated to provide. This report discusses the mechanical design, thermal hydraulic calculations, and neutronic calculations of a proposed standard experiment of accident tolerant BWR fuel concepts. Mechanical design examines the geometry and features of the main components. Thermal hydraulic calculations examine the modeling and results of the two-phase flow options available. Lastly, neutronic calculations examine the Monte Carlo analysis of enrichment, heat rates, and flux spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Effect of PBX Prill Size on the Compressive Properties of PBX 9502

Plastic-Bonded Explosive (PBX) prills are agglomerates a few millimeters in diameter composed of High Explosive crystals non-uniformly distributed in and surrounded by a polymer binder. When pressed at elevated temperatures at or above the binder melt temperature the polymer flows to uniformly coat the HE crystals to create a well-consolidated compaction. In reality it does so imperfectly, such that x-ray tomographic scans often look like a collection of prills mashed together. It is speculated that the larger the prill, the more non-uniform the binder distribution, but this has never been proven. Even if that were not so, the larger the prill the farther binder must flow toward the middle to homogenize during pressing. Thus, the degree of homogenization depends in part on prill size. The degree to which binder flows during pressing in turn affects the void distribution within pressed charges, which one suspects will affect shock sensitivity and material strength. In this paper we explore, for ~30 PBX 9502 formulation batches using the same TATB powder lot, how prill size effects the compressive mechanical properties.

36 MATERIALS SCIENCE↗