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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Technical Evaluation of Accelerated Basin De-Inventory Material Addition to Sludge Batch 11

Savannah River Nuclear Solutions has a need to discard spent nuclear fuel (SNF), currently stored in L Basin, to the Defense Waste Processing Facility (DWPF) for vitrification. The Department of Energy (DOE) has approved the Accelerated Basin De-inventory (ABD) Program for discarding SNF via transfers from H-Canyon to the Savannah River Site (SRS) Liquid Waste (LW) system. The first ABD discards will occur during the preparation of Sludge Batch (SB) 11. An initial impact evaluation of the LW flowsheet was performed by the Savannah River National Laboratory (SRNL) prior to the approval of the ABD Program. This evaluation addressed the LW downstream facilities based on the current H-Canyon flowsheet sequence for the average ABD discard. The flowsheet evaluation only included aluminum-clad SNF, specifically Materials Test Reactor (MTR) fuel and High Flux Isotope Reactor (HFIR) fuel similar to the planned SB11 discard. Following this evaluation, the flowsheet has been slightly altered to address (i) new nuclear criticality safety controls for DWPF that credit a higher amount of gadolinium as a neutron poison for all of the enriched uranium contained in a SB and (ii) potential additions of the H-Canyon neutralized fuel stream prior to the Low Temperature Aluminum Dissolution (LTAD) process in Tank 51. The early introduction of ABD material into the SB assembly process is being investigated to provide flexibility regarding transfer opportunities for H-Canyon to Tank 51 for SB11 and future sludge batches.

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

Film condensation with high heat fluxes and scaled experiments using pure steam for reactor containment cooling

Condensation tests were performed using a newly developed test facility for scaling the passive containment cooling system (PCCS) to a small modular reactor (SMR). The PCCS of the SMR plays a pivotal role in ensuring greater safety, reliability, and compactness than what is afforded by traditional reactors. Therefore, a well-designed PCCS is essential to SMRs. However, previous studies and test data were unsuitable for scaling, due to high variation in the test geometry and operating conditions. This study intends to close this research gap by using a novel designed scaled test facility consisting of vertical condensing test sections featuring 1-, 2-, and 4-inch-diameter condensing tubes with annular water cooling, and by applying superheated and saturated steam with different steam mass flow ranges of 5–25 g/s. Further, the primary test data, including axial temperatures, mass flow rates, and pressures, were used in conjunction with a standard data reduction method to estimate critical parameters such as heat fluxes, heat transfer coefficients, and condensation rates. These scaled test data would support improving empirical correlations and validating condensation models to identify scaling distortion for SMR PCCSs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

An experimental investigation of flame and autoignition behavior of propane

We report autoignition delay time data are one important means to develop, quantify, and validate fundamental understanding of combustion chemistry at low temperatures (T<1200 K). However, low-temperature chemistry often has higher uncertainties and scatter in the experimental data compared with high-temperature ignition data (T>1200 K). In this study, autoignition properties of propane and oxygen mixtures were investigated using the University of Michigan rapid compression facility in order to understand the effects of ignition regimes on low-temperature ignition data. For the first time for propane, autoignition delay times were determined from pressure histories, and autoignition characteristics were simultaneously recorded using highspeed imaging of the test section through a transparent end-wall. Propane mixtures with fuel-to-O 2 equivalence ratios of φ = 0.25 and φ = 0.5 and O 2 -to-inert gas molar ratios of 1:3.76 were studied over the pressure range of 8.9 to 11.3 atm and the temperature range of 930 – 1070 K. The results showed homogeneous or strong autoignition occurred for all φ = 0.25 experiments, and inhomogeneous or mixed autoignition occurred for all φ = 0.5 experiments. While a limited temperature range is covered in the study, importantly the data span predicted transitions in autoignition behavior, allowing validation of autoignition regime hypotheses. Specifically, the results agree well with strong-autoignition limits proposed based on the Sankaran Criterion. The autoignition delay time data at the strong-ignition conditions are in excellent agreement with predictions using a well-validated detailed reaction mechanism from the literature and a zerodimensional modeling assumption. However, the experimental data at the mixed autoignition conditions were systematically faster than the model predictions, particularly at lower temperatures (T< ~970 K). The results are an important addition to the growing body of data in the literature that show mixed autoignition phenomena are important sources of the higher scatter observed in the low-temperature autoignition data for propane and other fuels. The results are discussed in terms of different methods to capture the effects of pre-autoignition heat release associated with mixed autoignition conditions and thereby address some of the discrepancies between kinetic modeling and experimental measurements.

42 ENGINEERING↗
DOE OSTI2020

Microreactor Agile Nonnuclear Experimental Testbed Test Plan

Microreactors are an attractive technology option for kick-starting nuclear innovation if they can be operated at high temperature, yielding high power conversion thermal efficiencies comparable or better than in commercial light water reactors. Microreactors are currently the smallest variation of Small Modular Reactors (SMRs). SMRs are “newer generation reactors designed to generate electric power up to 300 MWe and whose components and systems can be shop-fabricated and then transported as modules to the sites for installation as demand arises.” (IAEA, 2016). Vendors are developing microreactor designs to provide an affordable, potentially mobile source of electricity - see Fig. 1 for an example of a microreactor on a semi-truck. Various microreactor designs are possible including heat pipe- and gascooled options, which are the focus of the nonnuclear testing described in this document. In heat pipe microreactors, high-temperature heat pipes using liquid sodium or potassium working fluid transport fission heat from the core to a heat removal section which in turn transfers heat to the power conversion system working fluid. In a gas-cooled design, He or other gas will flow through a solid monolith of material and transfer heat as the temperature of the gas increases through a heat exchanger to a power conversion unit. The logistics of all these processes will be examined and tested through a series of articles at the nonnuclear test bed at Idaho National Laboratory (INL), the Microreactor Agile Nonnuclear Experiment Testbed (MAGNET) facility. Microreactors designed to produce power of 0.1-20 MWt offer the potential for more affordable nuclear energy for a range of applications. In a heat pipe microreactor, heat pipes, fuel rods, and/or moderator are intermixed in the reactor core assembly. Heat pipes extend from the core region into the heat removal section where the power conversion unit working fluid flows through holes or channels, transferring heat from the heat pipes to the working fluid. In a gas-cooled microreactor, gas flows through the solid monolith region and up into the heat exchanger region, transferring heat to the working fluid. For initial testing, the heat removal working fluid can be a low pressure gas for testing that addresses thermal stresses. In the final application, heat addition to the power conversion working fluid typically occurs at high pressure, supporting operation of an air-Brayton, supercritical CO 2 (SCO2), or He-recuperated Brayton cycle. Various stages of the steps above will be demonstrated through the tests described in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Experiments with a prototype titanium hot cavity surface ionization source intended for electromagnetic separation of radioactive samarium and other lanthanide elements

This paper reports experimental results of a prototype titanium surface ionization source. For the first time, a lanthanide ion beam has been produced with a surface ionizer composed completely of titanium metal. Titanium does not readily activate with neutron irradiation. This offers the potential for inserting an ion source made of titanium directly into a reactor with a pre-loaded non-radioactive lanthanide target. This seamlessly integrates target irradiation with isotope separation, eliminating post irradiation sample manipulation. Samarium ion beam currents up to 960 nA have been produced in an off-line test bench equipped with rudimentary beam optics. This is a crucial step toward the development of an ionization source adopted for the electromagnetic isotope separator (EMIS) facility, which has been designed for high throughput separations of radioactive 153 Sm and other lanthanides of interest in the field of nuclear medicine. The ion current and important factors affecting the performance of the ion source, such as the ionizer temperature and thermal gradient, are discussed. Lastly, the experimental results are presented together with a discussion of future modifications to optimize the overall surface ionization source performance.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Mechanical response of HFIR-irradiated M5FRAMATOME cladding under simple and complex loading conditions

This study investigates the mechanical behavior of High Flux Isotope Reactor (HFIR) irradiated M5FRAMATOME cladding under simple and complex loading conditions through axial tensile and reversible cyclic bending. Axial tension specimens were pre-machined prior to HFIR irradiation while cyclic bend specimens were inserted as intact tubes. Tests articles were neutron-irradiated to 4 and 16 dpa, and specimens were tested at ORNL's hotcell facilities. The axial tension tests were conducted under constant displacement control, and the reversible cyclic bend tests were performed using ORNL's Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) apparatus. Results showed that mechanical response of Cr-coated and uncoated M5FRAMATOME cladding were similar and independent of irradiation dose for axial tension tests, while Cr-coated specimens’ reversible cyclic bend behavior differed from uncoated counterparts. For all tests, irradiation temperature showed a significant impact on the mechanical behavior. Below 280°C, all axial tensile specimens whether coated or not behaved similarly. Above 280°C, YS and UTS showed decrease with increasing irradiation temperature. A similar behavior was also observed in cyclic bend tests as well. The mechanical damage during cyclic bend tests was linked to damage accumulation in unirradiated Cr-coated zircaloy-4 specimens, and the effect of irradiation temperature was related to changing characteristics of defect mobility during high temperature irradiation.

Cinbiz, Nedim [ORNL] (ORCID:0000000346268515)↗
DOE OSTI2024

Performance Evaluation of Drain Water Heat Recovery Exchangers for Heat Pump Water Heaters

Water heating constitutes approximately 18% of energy consumption and is the second largest energy expenses in United States homes. Heat pump water heaters (HPWHs) are energy efficient technologies with lower carbon footprints as compared to conventional water heating technologies, such as gas and electrical resistance heaters. The performance of water-source HPWHs can be improved by recovering heat from blackwater using drain heat recovery heat exchangers. Depending on water draw patterns of single family and multifamily residences, the operation and heat transfer performance of these heat exchangers are highly transient. This paper examines the thermo-hydraulic performance of a drain heat recovery heat exchanger during its transient and steady-state operations. The heat recovery performance of the exchanger was evaluated at different inlet temperatures and flow rates ranging from 9°C to 36°C and 0.03 kg/s to 0.28 kg/s, respectively. The obtained effectiveness varies from 45%–85% depending on the operating conditions. The test facility, steady-state and transient experimental procedures are reported in detail. Further, the effect of operating conditions on the effectiveness is discussed. The experimental approach and results of the study will provide insights into the transient operation of drain recovery heat exchangers as well as, guide sizing for various steady state conditions.

Krishnan, Easwaran↗
DOE OSTI2023

Nb$_{3}$Sn SRF Photogun High Power Test at Cryogenic Temperatures

Superconducting RF (SRF) photoguns are emerging as promising candidates to produce highly stable electrons for Ultra fast Electron Diffraction/Microscopy (EUD/UEM) applications due to the ultrahigh shot-to-shot stability compared to room temperature RF photoguns. SRF technology was prohibitively expensive for industrial use until two recent advancements: Nb 3 Sn and conduction cooling. SRF gun can provide a CW operation capability while consuming only several Watts of RF power which eliminates the need of an expensive high power RF system and saves a facility footprint. Euclid is developing a continuous wave (CW), 1.5-cell, MeV-scale SRF conduction cooled photogun operating at 1.3 GHz. Here, in this paper, we present first high power results of the gun covered with Nb 3 Sn.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Fluorine Limits and Impacts in High-Level Waste Glass Compositions

The impact of elevated fluorine (F) content on Hanford high-level waste (HLW) glasses has not previously been studied in detail. This effort represents the first systematic study to determine what F concentration limits should be used for the design of alkali-borosilicate-based Hanford Waste Treatment and Immobilization Plant (WTP) HLW glasses, and to document the technical basis for that limit. If alkali borosilicate glass made from Hanford HLW can accommodate a large amount of F, the large capital costs for complex sludge washing facilities may be avoided, as would much of the operational costs and negative schedule impacts associated with handling the large volumes of water required to dissolve these salts. In order to determine a limit for F in likely HLW glass compositions, an evaluation was conducted on glasses with F ≤ 0.90 mass% from previous nuclear waste glass studies. The collected dataset contains 239 glasses (232 HLW glasses and 7 LAW glasses) including 109 glasses with 0.9 ≤ F mass% ≤ 2.5, 116 with 2.5 < F mass% ≤ 8.0, and 14 with F mass% ≥ 8 (max. F mass% = 17.42). The collected composition and property data were analyzed to determine the basis for the F tolerance, i.e. the maximum F concentration that can be processed without potential issues. Fluorine volatility, product consistency test (PCT) response, liquidus temperature (T L ), glass melt viscosity, and crystallinity have been investigated. No limits for F concentration can be made based on F volatility, T L , or glass melt viscosity, because the data show that high F in glasses do not indicate, with high probability, being restricted by those property constrains. However, crystallinity and PCT response were used to estimate the F tolerance. The results show that glasses with high F (≥ 0.90 mass%) are more likely to form large fractions of F-containing crystal phases which may increase PCT responses, i.e. decrease the glass durability. Based on the results of crystallinity and PCT data, the F tolerance of 4.5 mass% is estimated. There is no evidence of other glass components, such as calcium oxides and alkali metal oxides have combined impacts with F on the glass properties. Overall, the available high-F glass data is limited, especially in the designed HLW glass composition regions. Future work on formulation and testing of HLW glasses with F ≥ 0.9 mass% will close the data gaps and expand operational flexibility with respect to the fluoride tolerances. Volatility of F from melters and corrosion of materials in contact with glass melts are important for processing of high-F wastes; yet no test data are currently available. It is recommended tests be conducted to address these two potential issues.

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

Experimental Comparison of Hydrogen Refueling with Directly Pressurized vs. Cascade Method

This paper presents a comparative analysis of two hydrogen station configurations during the refueling process: the conventional “directly pressurized refueling process” and the innovative “cascade refueling process.” The objective of the cascade process is to refuel vehicles without the need for booster compressors. The experiments were conducted at the Hydrogen Research and Fueling Facility located at California State University, Los Angeles. In the cascade refueling process, the facility buffer tanks were utilized as high-pressure storage, enabling the refueling operation. Three different scenarios were tested: one involving the cascade refueling process and two involving compressor-driven refueling processes. On average, each refueling event delivered 1.6 kg of hydrogen. Although the cascade refueling process using the high-pressure buffer tanks did not achieve the pressure target, it resulted in a notable improvement in the nozzle outlet temperature trend, reducing it by approximately 8 °C. Moreover, the overall hydrogen chiller load for the two directly pressurized refuelings was 66 Wh/kg and 62 Wh/kg, respectively, whereas the cascading process only required 55 Wh/kg. This represents a 20% and 12% reduction in energy consumption compared to the scenarios involving booster compressors during fueling. The observed refueling range of 150–350 bar showed that the cascade process consistently required 12–20% less energy for hydrogen chilling. Additionally, the nozzle outlet temperature demonstrated an approximate 8 °C improvement within this pressure range. These findings indicate that further improvements can be expected in the high-pressure region, specifically above 350 bar. This research suggests the potential for significant improvements in the high-pressure range, emphasizing the viability of the cascade refueling process as a promising alternative to the direct compression approach.

08 HYDROGEN↗
DOE OSTI2021

Characterization of Embedded Sensors in Stainless Steel Test Articles and Design/Planning for MAGNET Testing

The nuclear industry is pursuing microreactors that can be factory assembled and deployed to remote regions for reliable power generation. One class of microreactors uses a monolithic metal core block coupled to heat pipes for heat rejection, which results in significant thermal stresses in the monolithic structures. This work describes the initial characterization and test plan for evaluating stainless steel test articles fabricated with embedded sensors for measuring heat pipe performance limits, as well as spatially distributed temperatures and strains during electrically heated thermal testing. The electrically heated testing will be performed in the non-nuclear Microreactor Agile Non-Nuclear Testbed and Single Primary Heat Extraction and Removal Emulator facilities located at Idaho National Laboratory. The goals of these tests are to (1) accurately monitor temperature and strain distributions that result from differential thermal expansion in the test articles and (2) quantify heat rejection limits of heat pipes as a function of operating temperature and working fluid during steady-state and transient operations. More generally, the ability to monitor component and system health during microreactor operation is attractive for providing a high sensor density to inform a limited number of microreactor operators to ultimately reduce operation and maintenance costs and move toward semi-autonomous operation. This report discusses the characterization of embedded thermocouples and fiber optic sensors in relevant test articles, including cylindrical pipes and hexagonal monolithic test articles for heat pipe-based reactors. The sensors were embedded by placing them in machined channels and then building up additional material by using ultrasonic additive manufacturing (UAM). UAM is a solid-state welding process that uses downward pressure and a lateral scrubbing motion to bond thin metal foils to a base material layer by layer. The ultrasonic welding process relies on the plastic deformation of the metal—as opposed to typical melting and solidification—to break oxide scales and bond the metal layers. The characterization of these embedded sensors included evaluating fiber optic signal attenuation, observing residual strain in the fibers, investigating microstructural and mechanical aspects, and demonstrating the sensors under various thermal loads and acoustic vibrations. Post-embedding characterization showed a fine grain structure (<1 μm) near the interfaces of the bonded foils as a result of severe deformation from the welding process. A large increase in hardness was observed at the foil interfaces and the fiber/matrix interface compared with the bulk matrix. Even when compared with the SS304 interfaces, the higher hardness observed around the embedded fiber suggests a higher degree of deformation due to the soft metal coating around the silica fiber core. The distributed fiber-optic temperature sensors and embedded thermocouples reliably measured temperature distributions during steady-state and transient thermal testing. The embedded fiber-optic sensors reliably measured strain during both transient and steady-state testing and properly identified resonant frequencies during acoustic testing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Non-contact Thermophysical Characterization of Solids and Fluids for Gen3 Concentrating Solar Power

In this project, we aim to develop Modulation Photothermal Radiometry (MPR), which is an attractive non-contact tool to evaluate thermophysical properties of high-temperature solids and HTFs that are otherwise time-consuming and difficult to measure and for in-situ testing of these properties. Therefore, the main objective of the project is the development and application of MPR to characterize high temperature thermophysical properties of heat transfer fluids (HTFs) and the solids (tubing and solar absorbing coating) in various components and sub-systems used in CSP plants. The measurements will be first demonstrated on a wide range of materials at UCSD and then the tool will be transferred to other facilities within the Gen3 CSP portfolio.

14 SOLAR ENERGY↗
DOE OSTI2020

ORNL Post-irradiation Examination (LAMDA) Final Report on Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

Irradiation effects on ferritic-martensitic steel Grade 92 were investigated in this project. Samples from two heats of Grade 92, i.e., G92-2b and NF616, were selected, which were irradiated in the High Flux Isotope Reactor (HFIR) and the Advanced Test Reactor (ATR), respectively. Samples from two heats of ferriticmartensitic steel T91, irradiated in the ATR, were also selected as references for the Grade 92 samples. The selected G92-2b samples were irradiated to 0.5–14.7 displacements per atom (dpa) at two temperature regimes: 400–497 °C and 683–720°C. The selection of the high irradiation temperatures up to ~720°C is to test the performance at such upper bound temperatures, which would provide insights for the upper bound applicable temperature for Grade 92. The selected NF616 and T91 samples were irradiated to 3–8.2 dpa at 292–448°C. Except for T91 heat2 samples in 3-mm diameter discs, all the other selected samples are of type SS-J2 miniature tensile specimens. Two Nuclear Science User Facilities (NSUF), i.e., the Irradiated Materials Examination and Testing (IMET) hot cell facility and the Low Activation Materials Design and Analysis (LAMDA) facility, were used in the Post-Irradiation Examination (PIE) activities for the selected samples, which include: Tensile tests at room temperature (IMET); Fractography of the tensile fracture surfaces (LAMDA: scanning electron microscopy and energy; dispersive spectroscopy); Vickers hardness measurements (LAMDA); and Microstructural characterization (LAMDA: electron backscatter diffraction, focused-ion beam, and transmission electron microscopy).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Hybrid Solar Heat Generation Modelling and Cases: Preprint

Renewable thermal energy systems (RTES) harness renewable energy sources to provide services for space heating and cooling, district heating, domestic hot water, and industrial process heat (IPH). The use of low-pressure steam generated by the combustion of fossil fuels is common today to provide process heat for industrial facilities. Solar IPH (SIPH) technologies could economically replace the steam or heat needs at many industrial sites by providing high-temperature pressurized hot water, a heat transfer fluid (HTF) such as synthetic-oil, or direct steam (Kurup and Turchi, 2015). RTES could be hybridized with technology options or combined with the existing heat supply (e.g. fuels), to give options for targeted IPH application and the reduction of fuel consumption. This work has tested hybrid system modelling approaches. Initial results show when a natural gas (NG) burner that feeds an IPH application of 300°C, has both air and NG streams pre-heated with a solar field/RTES exit temperature of 180°C (via an HTF), a 13% NG offset is possible. NG offsets reach up to 26%, when the RTES exit temperatures are at 300°C for a given annual capacity factor of 24%. This can be even higher with addition of thermal energy storage (TES).

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Accelerating Nuclear Fuels and Materials Qualification by Multi-Level Irradiation Experiment Campaign

The advanced reactor technologies feature fuels, coolants, and materials that promise safer operating conditions under normal and accident scenarios. However, the nuclear fuels and materials qualifications require several decades (for example, new reactor fuel qualification from conceptualization requires about 20 years). Therefore, accelerating nuclear fuels and materials qualification is essential, and it can be achieved by combining high through-put materials irradiation and testing, advanced post-irradiation examinations, and Multiphysics modeling. This paper addressed the associated challenges in accelerating nuclear fuels and material qualification for new and advanced reactor designs, which differ based on fuel, coolant, operating conditions, and structural materials. These challenges vary for radiation level, operating conditions (e.g., temperature and pressure), and coolant type (e.g., corrosion environment). In addition, the challenges and limitations in modeling tools, experimental facilities, and licensing guidelines are also discussed, and a general solution path forward is recommended.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Presentation: Accelerating Nuclear Fuels and Materials Qualification by Multi-Level Irradiation Experiment Campaign

The advanced reactor technologies feature fuels, coolants, and materials that promise safer operating conditions under normal and accident scenarios. However, the nuclear fuels and materials qualifications require several decades (for example, new reactor fuel qualification from conceptualization requires about 20 years). Therefore, accelerating nuclear fuels and materials qualification is essential, and it can be achieved by combining high through-put materials irradiation and testing, advanced post-irradiation examinations, and Multiphysics modeling. This paper addressed the associated challenges in accelerating nuclear fuels and material qualification for new and advanced reactor designs, which differ based on fuel, coolant, operating conditions, and structural materials. These challenges vary for radiation level, operating conditions (e.g., temperature and pressure), and coolant type (e.g., corrosion environment). In addition, the challenges and limitations in modeling tools, experimental facilities, and licensing guidelines are also discussed, and a general solution path forward is recommended.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

LLNL Macroscopic Anisotropic Explosives Research at INL National Security Test Range - Test Results

A select team of 23 engineers, scientists, and explosives specialists from LLNL, LANL, INL, and Marine Raiders from Marine Special Operations Command (MARSOC) and U.S. Special Operations Command (SOCOM) assembled during the second week of November at the INL National Security Test Range near Idaho Falls to investigate and demonstrate fundamental principles of explosives anisotropy. Today's explosives are isotropic in their detonation performance. That is, no matter what direction a detonation runs through bulk explosive, the performance is the same; whereas, anisotropic explosives exhibit different performance, depending on which direction the detonation wave moves through the explosive. The ANISO Team worked in subfreezing temperatures on the Snake River Plain, carrying out 55 experimental explosives shots in four days that lead to a clear understanding of the performance and behavior of an assembly of small, linerless, C4 shaped charges. These shots clearly demonstrated, for the first time, on a macroscopic scale, the principle of anisotropy in measured progression of the detonation through the explosive assembly. The outputs of nine piezo timing pins in the explosive assembly clearly showed detonation progressing through the assembly faster than nominal detonation velocity and moving slower than nominal detonation velocity in the opposite direction. Basic data from these experiments will be used to design and construct explosives assemblies that will be shot in the LLNL High Explosives Applications Facility's (HEAF). These experimental tests will provide refined basic data that will then be used by modelers to develop high explosives models. Computer simulations using these models will then be run to predict performance and design inhomogeneous, anisotropic bulk explosive charges that will be tested at LLNL.

33 ADVANCED PROPULSION SYSTEMS↗