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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 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Data Validation Package, July 2016 Groundwater Sampling at the Shirley Basin South, Wyoming, Disposal Site November 2016

Sampling Period: July 14-15, 2016 The 2004 Long-Term Surveillance Plan for the Shirley Basin South (UMTRCA Title II) Disposal Site, Carbon County, Wyoming, requires annual monitoring to verify continued compliance with the pertinent alternate concentration limits (ACLs) and Wyoming Class III (livestock use) groundwater protection standards. Planned monitoring locations are shown in Attachment 1, Sampling and Analysis Work Order. Point-of-compliance (POC) wells 19-DC, 5-DC, and 5-SC, and monitoring wells 10-DC, 110-DC, 112-DC, 113-DC, 40-SC, 54-SC, 100-SC, 102-SC, and K.G.S.#3 were sampled. POC well 51-SC and downgradient well 101-SC were dry at the time of sampling. The water level was measured at each sampled well. See Attachment 2, Trip Report for additional details. Sampling and analyses were conducted in accordance with the Sampling and Analysis Plan for the U S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and­ analysis-plan-us-department-energy-office-legacy-management-sites). ACLs are approved for cadmium, chromium, lead, nickel, radium-226, radium-228, selenium, thorium-230, and uranium in site groundwater. Time-concentration graphs of the contaminants of concern in POC wells are included in Attachment 3, Data Presentation. The only ACL exceedance in a POC well was radium-228 in well 5-DC where the concentration was 30.7 picocuries per liter (pCi/L), exceeding the ACL of 25.7 pCi/L. Concentrations of sulfate and total dissolved solids continue to exceed their respective Wyoming Class III groundwater protection standards for livestock use in wells 5-DC, 5-SC, and 54-SC as they have done throughout the sampling history; however, there is no livestock use of the water from these aquifers at the site, and no constituent concentrations exceed groundwater protection standards at the wells near the site boundary.

Carbon County, Wyoming↗
DOE OSTI2016

High power ring methods and accelerator driven subcritical reactor application

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g., PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the transverse beam dynamics. The results obtained allow to develop a correction scheme to minimize the tune variations of the FFAG. This is the cornerstone of a new fixed tune non-scaling FFAG that represents a potential candidate for high power applications. As part of the developments towards high power at the KURRI FFAG, beam dynamics studies have to account for space charge effects. In that framework, models have been installed in the tracking code ZGOUBI to account for the self-interaction of the particles in the accelerator. Application to the FFAG studies is shown. Finally, one focused on the ADSR concept as a candidate to solve the problem of nuclear waste. In order to establish the accelerator requirements, one compared the performance of ADSR with other conventional critical reactors by means of the levelized cost of energy. A general comparison between the different accelerator technologies that can satisfy these requirements is finally presented. In summary, the main drawback of the ADSR technology is the high Levelized Cost Of Energy compared to other advanced reactor concepts that do not employ an accelerator. Nowadays, this is a show-stopper for any industrial application aiming at producing energy (without dealing with the waste problem). Besides, the reactor is not intrinsically safer than critical reactor concepts, given the complexity of managing the target interface between the accelerator and the reactor core.

43 PARTICLE ACCELERATORS↗
DOE OSTI2017

Grain Growth and Precipitation Behavior of Iridium Alloy DOP-26 During Long Term Aging

The influence of long term aging on grain growth and precipitate sizes and spatial distribution in iridium alloy DOP-26 was studied. Samples of DOP-26 were fabricated using the new process, recrystallized for 1 hour (h) at 1375 C, then aged at either 1300, 1400, or 1500 C for times ranging from 50 to 10,000 h. Grain size measurements (vertical and horizontal mean linear intercept and horizontal and vertical projection) and analyses of iridium-thorium precipitates (size and spacing) were made on the longitudinal, transverse, and rolling surfaces of the as-recrystallized and aged specimens from which the two-dimensional spatial distribution and mean sizes of the precipitates were obtained. The results obtained from this study are intended to provide input to grain growth models.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Project Overview: LA13-FY13-123-PD08

This project was a two year, LANL-led effort focused on extending the application of ion emitter sources in thermal ionization mass spectrometry based analysis of trace actinide systems. Particular emphasis was geared towards thorium isotope quantitation, both as a TIMS “worst case” scenario in terms of ionization efficiency and a desirable target for future chronometric measurements. Additional points of focus included source performance characteristics and identification of next-generation emitter strategies addressing remaining development challenges.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

Large-Scale Power Production Potential on U.S. Department of Energy Lands

This report summarizes the potential for independent power producers to generate large-scale power on U.S. Department of Energy (DOE) lands and export that power into a larger power market, rather than serving on-site DOE loads. The report focuses primarily on the analysis of renewable energy (RE) technologies that are commercially viable at utility scale, including photovoltaics (PV), concentrating solar power (CSP), wind, biomass, landfill gas (LFG), waste to energy (WTE), and geothermal technologies. The report also summarizes the availability of fossil fuel, uranium, or thorium resources at 55 DOE sites.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2018

Formation, Stability, and Chemistry of Tetravalent Actinide Nanocrystals

First database with almost 4000 stable structures for thorium dioxide (ThO 2 ) n (n≤8) calculated with density functional theory. The figure shows the lowest energy minima for different units of ThO 2 and the energy landscape generated by the found stable structures for Th 8 O 16 . This highlights the structural richness of the ThO interactions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Advanced Fuel Cycle Cost Basis – 2017 Edition

This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

DE-NE0008277_PROTEUS final technical report 2018

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce such a comprehensive suite of experimental data to support any future GCFR endeavors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Demonstration of a Rapid HPLC-ICPMS Direct Coupling Technique Using IDMS- Project Report: Part II

Researchers working on chemical separations and inductively coupled plasma mass spectrometry have developed a rapid separation–direct analysis scheme to determine both concentration and isotopics of a suite of elements down to the low picogram level, with the ultimate goal of performing rapid analysis of postdetonation nuclear materials. The second phase in method development has shown that both the sensitivity and stability to achieve the precise, low-level analyses required can be maintained even when present in complex silicon- and uranium-based matrices. The methods analyte list has been expanded to incorporate a number of new analytes, notably the actinides plutonium, americium, curium, neptunium, and thorium and the key fission product zirconium. Key nonnatural fission product isotopic ratios in irradiated highly enriched uranium targets have been successfully measured at the low picogram level. The accuracy of these ratios was confirmed using an isotopic depletion and decay modeling software to within 1–2%.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

Development of a Two-Fluid Molten Salt Reactor Safeguards Model

This work outlines the development of a two-fluid molten salt reactor process and safeguards model. The model is split into two parts consisting of a process model and a safeguards model. The process model is based on a design by Flibe Energy, the Liquid-Fluoride Thorium Reactor, which is a two-fluid molten salt reactor that performs full salt processing on-site to remove fission products and re-fuel the reactor. The model simulates feed and consumption rates of the reactor fuel and blanket salts. The process model includes the reactor core and salt processing loops. The reactor core model has a robust architecture that allows for integration with other tools and data sets as they become available. A majority of the effort to date has been focused on the process model, and the safeguards model will be developed in detail in future work. A preliminary safeguards analysis was performed based on actinide inventories, and a preliminary materials accountancy approach was initialized. The results of this analysis are presented along with a description of the model development.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Off-Gas Condensate and Cast Stone Analysis Results

A sample of off-gas condensate produced during a demonstration test of the direct-feed low-activity waste (DFLAW) Radioactive Waste Test Platform using waste retrieved from Hanford storage tank AP-105 waste was analyzed for major cations, anions, total cyanide, Resource Conservation and Recovery Act (RCRA) metals, selected radionuclides, and selected organic compounds. The remaining off-gas condensate was concentrated through evaporation and then used to produce Cast Stone waste forms. After the solid waste forms cured for 28 days, U.S. Environmental Protection Agency (EPA) Method 1311 (EPA 1992) Toxicity Characteristic Leach Procedure (TCLP) tests were conducted on two samples. The TCLP leachates were analyzed for RCRA metals, fluoride, and total cyanide. These chemical analyses were conducted to collect data on the waste form performance and to determine if the waste form will meet the disposal requirements of the Waste Control Specialists LLC (WCS) Federal Waste Disposal Facility (FWF) in Texas (WCS 2015) and the Integrated Disposal Facility at Hanford. The radionuclide results in the off-gas condensate were used to estimate concentrations in the Cast Stone waste forms; these estimates were then compared to Class C limits (30 TAC 336.362 Appendix E). With the possible exception of U-232, all measured radionuclides were determined to be below their respective Class C limits. Inductively coupled plasma-mass spectrometry (ICP-MS) was used to measure most of the radionuclides. It is not possible to differentiate the isotope measured at mass 232 between U 232 and Th-232 by mass spectroscopy. Class C limit comparisons were made by assuming two maximum possible scenarios; one in which all the isotope at mass 232 occurred as Th-232, and another assigning the entire mass contribution to U-232. If the isotope at mass 232 is all Th-232 then the waste form is well below its Class C limit; however, if the isotope at mass 232 is all U-232, then the Class C limit would be exceeded by more than a factor of 8. A chemical procedure to separate uranium from thorium followed by reanalysis by ICP-MS would be required to resolve this issue. A similar issue occurs at mass 238, which could be U-238 or Pu-238; however, in this case Pu-238 can be ruled out as a contributor to the isotope measured at mass 238 for the following reasons. Pu-238 has a half-life of 87.7 years and its daughter product (U-234) has a half-life of 2.5 × 105 years and a measured concentration of < 0.002 µg/L. If significant concentrations of Pu-238 were present in this waste, then the amount of measurable U 234 would have increased over the decades of storage in the Hanford tanks. Because this was not observed, it is concluded that the isotope measured at mass 238 is all U-238. A total of 68 organic compounds were targeted for analysis in the off-gas condensate. Of these compounds, only nine were measured above their respective minimum reportable concentration. These compounds were acenaphthene, acetone, n-butyl alcohol, p,p’-DDT, methyl ethyl ketone, nitrobenzene, o nitrophenol, pyridine, and toluene. Of the detectable compounds, only acenaphthene, acetone, o nitrophenol, and pyridine were above wastewater standards (40 CFR 268.48 - Universal Treatment Standards). None of the organic compounds exceeded the non-wastewater standards (40 CFR 268.48 - Universal Treatment Standards). Analysis of the TCLP leachates from the Cast Stone waste forms indicated that none of the analytes exceeded the Universal Treatment Standard and, in general, all were well below the standards.

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

Activation of Th-232 target using FLUKA simulation code

FLUKA is a general-purpose Monte Carlo simulation code used to model the interaction and transport of particles from a few KeV to thousands of TeV in arbitrary materials. It is built and maintained with the aim of including the best possible physical models in terms of completeness and precision [1]. In this study, we are using the USRBDX card to get the energy spectrum (photons/sec), RESNUCLEi card to score the residual isotopes and activities and USRBIN card to calculate the dose rate from the Thorium target irradiated in the BLIP.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Analysis of Sludge Batch 10 Samples from Tank 51H (HTF-51-18-94, -95)

Savannah River National Laboratory characterized sludge samples from Tank 51H that is being used to generate Sludge Batch 10. The composite slurry sample from Tank 51H was analyzed and determined to contain a concentrated salt solution with a sodium concentration of 5.79 M (1.33E+05 mg/L) and insoluble sludge solids content of 7.20 wt% (using all nine replicates from the 3 wt% total solids measurements). Hydroxide is the main anionic species in the salt solution with a concentration of 3.00 M followed by nitrate (0.728 M) and nitrite (0.657 M). The sum of the major cations versus the sum of the major anions from the analysis of the filtered supernatant liquid phase shows a difference of <5% indicating good data quality. The results from the analysis of the digestions of the well mixed composite slurry sample show sodium to be the main component of the total dried solids of the sample followed by aluminum, iron, manganese, and thorium. Additional measurements of the weight percent total solids and density using hand shaking of the sample bottle and recirculation through a peristaltic pump indicate both methods provide adequate mixing of the sludge slurry sample.

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

Technical Manual and User's Guide for MILDOS, Version 4.1

The MILDOS computer code is used to estimate the radiological impacts of airborne emissions from uranium mining and milling facilities. The code allows users to consider both conventional uranium ore operations and operations associated with in situ recovery (ISR) facilities. The code is used by license applicants and U.S. Nuclear Regulatory Commission staff to perform routine radiological impact and compliance evaluations for various uranium recovery operations. As part of a major code update, the following improvements and additions were made to the code: (a) revised ISR models, (b) support for ores containing thorium-232 (Th-232) and its daughter radionuclides in addition to the currently supported uranium-238 (U-238) and its daughter radionuclides, (c) a revised area source model, (d) the capability to perform sensitivity analysis on specific input parameters, (e) the capability to use current meteorological data provided by the National Climatic Data Center, and (f) an interactive results module.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Special Analysis for the Disposal of the Portsmouth Uranium Fluoride Solids Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This special analysis (SA) evaluates whether the Portsmouth Gaseous Diffusion Plant (PORT) Uranium (U) Fluoride Solids waste stream is suitable for disposal by shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site (NNSS). The PORT U Fluoride Solids waste stream consists of highly enriched uranium solids. The PORT U Fluoride Solids waste stream requires an SA to resolve whether it is acceptable to dispose of technetium-99 ( 99 Tc), uranium-234 ( 234 U), and thorium-230 ( 230 Th) with activity concentrations exceeding the NNSS Waste Acceptance Criteria Action Levels.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Special Analysis for the Disposal of Sandia National Laboratory Line Generated Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada, Revision 1.0

This special analysis report evaluates whether the Sandia National Laboratory (SNL) Line Generated Waste is suitable for shallow land burial at the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Disposal of the SNL Line Generated waste meets all DOE M 435.1-1 performance objectives. The SNL Line Generated waste is recommended for acceptance with the following conditions: 1. The total waste stream tritium (3H) inventory shall not exceed 2E15 Bq; 2. The total waste stream radium-226 (226Ra) inventory shall not exceed 2E8 Bq; 3. The total waste stream thorium-230 (230Th) inventory shall not exceed 3E8 Bq; 4. The total waste stream uranium-234 (234U) inventory shall not exceed 2E10 Bq.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗