Search NASA⌕ Search

SEARCH · Search NASA

Results for “Tritium Processing System”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

220 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Study of an Unrefined Humate Solution as a Possible Attenuation-based Remedy for Uranium Contamination in Acidic Groundwater - 20351

The Savannah River Site (SRS), joined the environmental cleanup program in 1981 after the Cold War, and was designated as a hazardous waste management facility. About 1.8 billion gallons of acid wastes were disposed into the F/H Area Seepage Basin that included many radionuclides and dissolved metals, resulting in highly contaminated groundwater plumes with pH of 3 - 5.5. The acidity of the plumes contributes to the mobility of several constituents of concern (COC) such as H-3, U-238, I-129, and Sr-90 for the F-Area plume and H-3, Sr-90 and mercury for the H-Area plume. An active treatment (pump-and-treat and re-injection) system was constructed and put in place in 1997 to address the removal of metals and radionuclides from the contaminated groundwater. As remediation projects advanced, active treatments were transitioned to more passive or enhanced-passive approaches such as the subsurface barrier with base injection system, which has been controlling the mobility of contaminants. However, new and more efficient attenuation-based remedies are always sought. Several studies have proposed that humic substances can be used to remediate sites contaminated with heavy metals by creating a permeable reactive barrier. Humic substances (HS) are major components of soil organic matter. HS are polyfunctional organic macromolecules formed by the chemo-microbiological decomposition of biomass or dead organic matter. These substances are usually divided into three main fractions: humin (insoluble at all pH values), humic acid (soluble at pH greater than 3.5), and fulvic acid (soluble at all pH values). Humic substances are major components of the soil at SRS and, in addition, are also helpful in the removal process of uranium. This research focuses on uranium (VI), which is a key contaminant of concern in the F-Area groundwater plume. The interaction of uranium with sediment in the presence or absence of humic substances involves complex mechanisms that are not yet well understood. The interaction of U(VI) with humic substances can affect the adsorption of U on sediment, altering its mobility in the subsurface. The objective of this study is to study the mobility of uranium in the presence of humic acid and to determine if sediments amended by humic acid can enhance the sorption of heavy metals onto sediments. In this research, experiments were conducted using background (clean) F-Area aquifer sediments to understand and predict uranium mobility. A blended material, humate, containing both humic and fulvic acids and a chemically modified humate (KW15 modified humic or mod-HA) was studied as a possible amendment for uranium remediation in SRS groundwater. Batch experiments were conducted in triplicates using 20 ml of aqueous suspension in DI water containing SRS sediment and mod-HA and studied the effect of pH, kinetics of humate sorption onto SRS sediment and also studied the sorption of uranium onto mod-HA amended SRS sediment. Results indicate that mod-HA amended sediment increases uranium removal significantly. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Enhanced Performance Fast Reactors with Engineered Passive Safety Systems

This report summarizes the work and results of NEUP Project #15-8251, "Enhanced Performance Fast Reactors with Engineered Passive Safety System," performed from 2015-2018 at UC Berkeley and Argonne National Laboratory. This project concerns the incorporation of a novel engineered passive safety system, called the Autonomous Reactivity Control (ARC) system, into Sodium-cooled Fast Reactors (SFRs) to improve their transient and economic performance. The first part of this project consisted of fleshing out the basic principles and the manufacturing process for an SFR fuel assembly with the ARC system installed, with the details provided in high-quality, step-by-step CAD drawings. Once the detailed design is laid out, high-fidelity simulations were performed to optimize the dimensions of important ARC system components and to quantify the standalone time-response of ARC systems in a typical SFR environment. These high-fidelity simulations were used to create reduced-order models of ARC system response which could be incorporated into the SAS4A/SASSYS-1 whole-core transient models. In order to accurately account for the time-response of ARC systems in whole-core transient simulations, modifications were made to the SAS4A/SASSYS-1 transient code. These new capabilities were tested and incorporated into the production version of the code, and have been used extensively in this analysis. Before performing whole-core transient calculations, the two reference cores selected for this study to cover a wide design space were simulated and characterized. The two reference cores, (1) a medium-sized oxide-fueled Advanced Burner Reactor and (2) a large metal-fueled Breed-and-Burn core, challenge conventional means of achieving passive safety. Because the two cores vary substantially, simulating the ARC system in both allows for conclusions to be drawn about the applicability of ARC systems in different SFRs. Once both cores were fully characterized, transient calculations were performed to understand the complicated interplay between the ARC system and other inherent reactivity feedback mechanisms. By performing parametric studies of the transient behavior, ARC system designs that best improve transient performance in each core were identified. Once transient improvements were established, the transient calculations were extended to improve reactor economics through design modifications. Upon establishing the potential benefits of ARC system inclusion, the study examines the negative neutronic implications of incorporating ARC systems into the reference cores. The study concludes by discussing a number of knowledge gaps that exist which may have important implications for ARC system implementation. Ultimately, the results of this project indicate that ARC systems may provide an effective means to improve transient performance in certain types of systems. The transient simulations in the oxide Advanced Burner Reactor showed that significant margin to coolant boiling and fuel melting could be gained through the use of a properly designed ARC system, thus improving inherent safety. While oscillatory behavior was induced in the transient response for certain combinations of ARC system parameters, the oscillation-free design space that can be effectively used to improve performance was found to be large, and the additional margin was demonstrated to allow economical improvements to the core design. Conversely, difficulties were encountered in the Breed-and-Burn core due to fundamental aspects of the core design. The combination of a particularly strong positive coolant void feedback and the use of metallic fuel allowed for diverging oscillations to manifest in response to certain transient initiators, and in the end it was found that the current ARC system design could not simultaneously enable passive safety in all transients examined. In order for complete passive safety to be achieved, a mechanism for slowing the reverse actuation of ARC systems is necessary. Regardless, the standard ARC system was able to improve the performance of two out of the three examined transients by increasing the margin to boiling in Loss of Heat Sink transients and avoiding boiling in Transient Overpower accidents. Additionally, no significant barriers to implementation were found, although the issue of tritium production should be closely monitored in any core incorporating ARC systems. Still, a number of technology gaps exist, and these are discussed at the conclusion of the report to recommend directions for future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Progress toward hydro-equivalent ignition in OMEGA direct-drive DT-layered implosions

Considerable progress has been made in deuterium-tritium-layered implosion experiments on the OMEGA Laser System, bringing the prospects for thermonuclear ignition in direct-drive configurations with megajoule-class lasers closer to reality. Doing so has required navigating the balance between improved 1D performance and multidimensional stability. Using statistical modeling based on over 350 cryogenic implosions to identify various degradation mechanisms, and combined with multidimensional simulations and experimental techniques such as target offsets to combat residual flows, core conditions have repeatably been achieved that extrapolate to the burning-plasma state when scaled to 2.15 MJ of symmetric laser illumination. Using high implosion velocities (⁠> 450 km/s) and moderately high adiabats (⁠~5⁠), these experiments produced record-high scaled Lawson parameters in direct drive equal to 89 ± 2% of that required for ignition with expected yields of up to 1.5 ± 0.2 MJ. To improve these results still further, focused physics studies are performed to improve physics understanding and identify routes to even greater performance. Recent studies include investigations into the impact of mounting features, laser imprint, reduced fuel temperatures, and greater on-target intensities through subscale experiments. This manuscript gives a summary of the cryogenic direct-drive program on the OMEGA laser, including routes taken to achieve the current best performance, the status of recent focused physics investigations, and future designs—such as target solutions to laser imprint and reducing vapor density to increase convergence—that are expected lead to the demonstration of hydro-equivalent ignition on OMEGA.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

U.S. Fusion D-Li Neutron Irradiation Facility: Fusion Prototypic Neutron Source (FPNS) Technology Study

A primary engineering challenge for the realization of fusion energy involves the development of high-performance structural and plasma-facing materials with dimensional stability and resistance to neutron degradation of mechanical properties. The harsh fusion environment, including high levels of helium production and other transmutation products along with high operating temperatures, poses an extreme challenge for materials. Identifying materials that can withstand these conditions is a high priority for fusion energy research. Critical to resolving this gap in scientific understanding is developing a neutron source facility with the ability to mimic the extreme conditions in a fusion reactor. Working with the Department of Energy (DOE) Office of Fusion Energy Sciences (FES), Oak Ridge National Laboratory (ORNL) and the Virtual Laboratory for Technology (VLT) organized a workshop to discuss the need for a Fusion Prototypic Neutron Source (FPNS). [1] While there are a variety of uses for FPNS neutrons, the main driver of US interest is expanding the scientific understanding of materials degradation in the deuterium-tritium (D-T) fusion environment. FPNS will help build the scientific foundation required to enable design of a next step fusion device. Using specifications from previous projects as a guide, key parameters were agreed upon for a near term, moderate cost US facility that would be complementary to the more extensive plans of the International Fusion Materials Irradiation Facility (IFMIF). Types of facilities that could potentially meet those criteria were identified. A scoping study for three technologies was initiated to answer three critical questions: 1) Can the concept meet the key parameters, 2) What is the estimated cost, and 3) What is the estimated schedule. One approach selected for the scoping study is a D-Li stripping source similar to the design of the IFMIF but at a lower beam power and atomic displacement rate for cost and schedule reasons. The purpose of this report is to document the results of the D-Li stripping source option study, providing answers to the three questions. The approach to generating a fusion prototypic neutron source using a deuteron beam incident on a flowing lithium target has been studied for more than 40 years. As a result of the research and development recently executed primarily in Europe and Japan, the ability of this approach to meet key parameters (damage parameters, irradiation volume) outlined in the US workshop for a near-term, moderate cost fusion prototypic neutron source facility is not in doubt. The main question that must be answered is the cost of such a facility. As a result, a strong focus was placed on understanding alternatives for developing the facility and their relative costs. The project was broken into four components for the cost estimate: the ion source, accelerator, lithium target system, and conventional facilities. The cost of the accelerator is an important cost driver for the D-Li stripping concept. To fully understand the range of the accelerator cost, multiple alternatives were considered including normal conducting and superconducting linacs and the frequency of the radiofrequency quadrupole. Multiple previous design studies dating from 1989 through 2017 were examined (costs were escalated to 2019 dollars) and compared. After analysis, two options were selected for costing, one superconducting and one room temperature. We found similar project costs for the two accelerator options, on the order of $100M including design and project management but excluding contingency. Including everything except the conventional facilities, the estimate for the neutron source facility is estimated to be $220M without contingency. Additional significant costs for the facility are the conventional costs arising from the need to construct buildings for the accelerator and target facilities, bringing electrical power to the site including a new substation, and removing the heat from the facility using heat exchangers. While ORNL would oversee this work, much of the effort would be executed by contractors. Estimates for this effort was based on similar projects that are currently being executed at ORNL or were executed at ORNL in the recent past. The estimate for conventional facilities for the neutron facility is $173M without contingency, bringing the total project cost to $393M without contingency. With a 50% contingency, the estimated cost is $589M. The estimate range was established using guidance from the Association for the Advancement of Cost Engineering (AACE). Specifically, guidance from recommended practice 17R-97 provides insight on the expected accuracy range of estimates based on the degree of project definition and the position of the project in the stage-gate process or maturity timeline. An AACE Class 5 type estimate is appropriate for concept screening. The expected accuracy range includes a low end of -50% to -20% and a high end of +30% to +100% around the estimated cost. For the D-Li stripping neutron source, a -20% to +100% cost range was applied for a recommended cost range of $471M to $1,179M. Less emphasis was placed on determining the project schedule due to the fact that the project costs were large and may eliminate this option from consideration. The development of the accelerator facility may require on the order of five years without contingency. A more realistic schedule with contingency may be seven years. However, this estimate should be scrutinized in a preconceptual design activity if a more realistic estimate is needed. Finally, one lesson learned in this process is that the requirements for a near-term, moderate cost facility drives the alternative selection toward existing facilities and infrastructure. The D-Li stripping neutron source option requires a green field site due to the lack of existing buildings and infrastructure that can be reused for this effort. As a result, the conventional facilities costs are a significant fraction of the total facility cost. Other neutron source options where buildings, electrical power, and cooling infrastructure already exist may be able to meet the requirements for a near term, moderate cost facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗