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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 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Field Scoping Study of Gamma-Ray Measurement of UF 6 Cristillini Samples with CZT and HPGe Detectors

A field scoping study was conducted to evaluate the ability of Cd-Zn-Te (CZT) detectors and High Purity Germanium (HPGe) detectors to quantify the U-235 enrichment in UF 6 samples collected on alumina in P-10 tubes, similar to the Cristillini method of sample collection for analysis by mass spectrometry. With in-field gamma-ray spectroscopy methods, international safeguards and verification missions could benefit from quick in-field confirmation of declared enrichment values prior to or without the need for additional shipping, sample preparation and analysis. Samples of depleted (DU), low-enriched (LEU) and high-assay low-enriched (HALEU) uranium were collected on Cristillini-type alumina in P-10 tubes and measured with two H3D, Inc. M400 CZT detectors as well as a Canberra Falcon HPGe detector. The M400 detectors were able to confirm enrichment in the DU and LEU samples but not the HALEU sample in measurements 8-40 hrs after collection. The Falcon system was able to confirm enrichment in all three samples. Improvements in detector geometry, increase in number of detectors as well as increase sample mass loading may allow measurement of U-235 enrichment in these types of samples within 4-8hrs of collection with 5-10% uncertainty.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2022

An electrochemical mesoscale tool for modeling the corrosion of structural alloys by molten salt

Understanding the impact of microstructure on corrosion rates can aid the development of corrosion-resistant alloys for molten salt reactors. Here in this work, we develop an electrochemical phase-field model for capturing the microstructure-dependent corrosion of structural alloys by molten salts. As a demonstration problem, we apply this model to capture the selective depletion of Cr from Ni-Cr grain boundaries during corrosion in molten FLiBe salt. We perform sensitivity analysis and model verification on 1D simulations to confirm that the model predicts diffusion-limited kinetics. The model is validated using 1D, 2D, and 3D simulations against experimental data for Ni-5Cr and Ni-20Cr corrosion in molten FLiBe. The 1D simulations predict the corrosion behavior with reasonable accuracy when using an effective diffusion coefficient that accurately represents the grain boundary diffusion. 2D simulations that represent the grain structure underpredict the corrosion. 3D simulations that represent the grain structure predict the corrosion with reasonable accuracy. The corrosion rate predicted by the 3D simulations is proportional to the average grain size at the alloy/salt interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Charge collection efficiency in back-illuminated Charge-Coupled Devices

Low noise CCDs fully-depleted up to 675 micrometers have been identified as a unique tool for Dark Matter searches and low energy neutrino physics. The charge collection efficiency (CCE) for these detectors is a critical parameter for the performance of future experiments. We present here a new technique to characterize CCE in back-illuminated CCDs based on soft X-rays. This technique is used to characterize two different detector designs. The results demonstrate the importance of the backside processing for detection near threshold, showing that a recombination layer of a few microns significantly distorts the low energy spectrum. The studies demonstrate that the region of partial charge collection can be reduced to less than 1 micrometer thickness with adequate backside processing.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

MCS solutions for the TVA Watts Bar Unit 1 multi-physics depletion benchmark

The high-fidelity solution for exercises 1-3 of the recently developed TVA Watts Bar Unit 1 multi-physics and multi-cycle depletion benchmark using MCS was coupled with two different thermal-hydraulics (TH) codes: TH1D and COBRA-TF (CTF), and their solutions were compared. The solutions from MCS/CTF and MCS/TH1D for typical Pressurized Water Reactor (PWR) problems are quite close. And to evaluate MCS solutions for the benchmark, their results were also compared against measured data as well as publicly available high-fidelity solutions. The MCS criticality solutions at Hot Zero Power (HZP) deviate within 50 pcm against the reference solutions. While for the critical boron concentration (CBC) search calculations, the difference on the MCS calculated CBC is about 7 ppm, and the root-mean-squared (RMS) errors for the assembly power distribution and outlet coolant temperature are less than 0.5% and 1 Celsius degree respectively. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Modeling Molten Salt Reactor Fission Product Removal with SCALE

Liquid-fuel Molten Salt Reactors (MSRs) allow for continuous online fuel treatment and processing using a variety of subsystems. Among these subsystems, the off-gas system (OGS) continuously removes fission products from the primary fuel salt through a helium sparging process. Accurately modeling fission product behavior is paramount for the development of MSRs; while there are a variety of depletion modeling and simulation tools that allow for removal of material from a defined system, most of them work in a semicontinuous, batch-wise manner. The new tools implemented into TRITON, in the SCALE suite for reactor analysis and design, aim to address this issue by implementing truly continuous material feeds and removals. In addition, material accountancy is enhanced in these tools, which enable tracking of removed material in user-specified mixtures. These continuous depletion tools provide a way to model material transport throughout systems more efficiently and accurately. This report validates the new SCALE MSR modeling capabilities using data from the Molten Salt Reactor Experiment (MSRE). After a framework was established for deriving realistic removal rates based on design specifications, these rates were used by TRITON to continuously remove specific elements while performing burnup calculations. Results were compared to those obtained in MSRE, with a focus on the 135 Xe poison fraction. After these capabilities were validated, the learned concepts and best practices were applied to model the Molten Salt Breeder Reactor (MSBR) and the Molten Salt Demonstration Reactor (MSDR) to study the effects of continuous fission product removal in fuel salt reactivity, as well as the lifetime effects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Comparison of Modeled to Measured Spectra using MCNP and GADRAS to Benchmark and Contrast Modeling Limitations

The desire to improve nuclear material detection through portal monitors and other low resolution detectors has led to interest in benchmarking the performance of radiation transport codes. These codes can be used to generate a variety and quantity of spectra that may be cost prohibitive to measure directly. Particular characteristics of typical detection scenarios were isolated to compare the performance of radiation transport codes to laboratory experiments. These benchmark experiments were performed to validate simulations in a number of key areas. These experiments included high areal density configurations in both one- and three-dimensional configurations. In addition, off-angle measurements were conducted to investigate the impact of detector response assumptions on complex geometries, for example where an unknown source is not physically collocated with an apparent hot spot. Furthermore, by examining both simple shielding and backscatter configurations, isolation of the scatter emanating from the object and environmental background scatter contributions to the spectra could be elucidated. A series of experiments were performed in which the order of shielding layers and the thickness were evaluated to examine the ability of the simulations to address these variables. Experiments with increasing thickness of polyethylene around a neutron source also allows for examinations of the ability of the transport simulations to properly model n,γ reactions. Finally, depleted uranium (DU) measurements have more complicated spectra than the cobalt-60 and consequently allow for investigations into the ability of transport simulations to model self-shielding. In each case, the results of the experiments were compared to MCNP simulations to judge the performance of the code, as was done previously with GADRAS 18.7.9 simulations [7]. All of the benchmark measurements were taken with a liquid nitrogen cooled 140% high purity germanium (HPGe) detector with a bismuth collimator that had the front tin filter removed. The measurement location and detector configuration were subsequently used for all experimental configurations. The remainder of this report details the experimental measurements, Section 2 and the MCNP model, Section 3. Comparisons between the experimental results and MCNP-generated spectra are performed, Section 4. Section 5 compares the results obtained with MCNP with the previously reported results obtained with GADRAS [2]. Finally, Section 6 presents the conclusions.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Soft-hard framework with exact four-momentum conservation for small systems

A new framework, called x-scape, for the combined study of both hard and soft transverse momentum sectors in high-energy proton-proton (𝑝−𝑝) and proton-nucleus (𝑝−𝐴) collisions is set up. A dynamical initial state is set up using the 3d-Glauber model with transverse locations of hotspots within each incoming nucleon. A hard scattering that emanates from two colliding hotspots is carried out using the Pythia generator. Initial state radiation from the incoming hard partons is carried out in a new module called I-matter, which includes the longitudinal location of initial splits. The energy-momentum of both the initial hard partons and their associated beam remnants is removed from the hot spots, depleting the energy-momentum available for the formation of the bulk medium. Outgoing showers are simulated using the matter generator, and results are presented for both cases, allowing for and not allowing for energy loss. First comparisons between this hard-soft model and single inclusive hadron and jet data from 𝑝−𝑝 and minimum bias 𝑝−Pb collisions are presented. Single hadron spectra in 𝑝−𝑝 are used to carry out a limited (in number of parameters) Bayesian calibration of the model. Fair comparisons with data are indicative of the utility of this new framework. Theoretical studies of the correlation between jet 𝑝 𝑇 and event activity at mid and forward rapidity are carried out.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Depletion Analysis of a Generic Fast Spectrum Molten Salt Reactor Supporting Material Control and Accounting (Rev.1)

Advanced reactors are of interest for a variety of use scenarios, and there are numerous advanced reactor designs being considered, which includes molten salt reactors (MSRs). Depending on the design, advanced reactors may have more extensive material control and accounting (MC&A) processes than current Light Water Reactor (LWR) designs. This project evaluated the use of Monte Carlo simulations for generating information that could guide the development of MC&A approaches for MSRs. In particular, the current capabilities of MCNP version 6.3 internally coupled with CINDER’90 were assessed for a fast spectrum liquid fueled MSR with online fission product removal and refueling. The fission product removal and refueling was performed in batches, and a Python wrapper was developed to control the simulations and update the fuel composition.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Mechanical and Thermophysical properties of low enriched uranium-10wt% molybdenum rolled foils

In this study, low-enriched (~19.75% 235 U) U-10Mo(wt%) foils were evaluated for both mechanical and thermophysical properties to establish baseline characteristics prior to irradiation testing. Cast U-10Mo ingots were sequentially hot and cold rolled to two different final foil thicknesses, resulting in different levels of cold work being present in the finished foils. The mechanical response of the U-10Mo foils was evaluated in the temperature range of 293–623 K. Increasing test temperatures resulted in decreases of both yield and ultimate tensile strength. Comparison of results to earlier work performed on depleted U-10Mo rolled foils indicates that the material in this study has lower strength and increased ductility. However, a comparison of the two different foil thicknesses in this study found that they were nearly identical in yield and ultimate tensile strength, differing by only 10–20 MPa (<2%). Thermal conductivity was calculated from the measurements of specific heat, linear thermal expansion, and thermal diffusivity from 323 to 1273 K on both thicknesses of the foils. These data were approximately 5–10% lower than previously published data above 573 K and were similar to data for temperature ranges below 573 K. Differences are likely related to differences in grain size and/or impurity content and variation in fabrication history.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Feasibility Study for Cathodic Protection of Waste Tanks at Hanford

The Hanford Nuclear Reservation site contains approximately 211 million liters of radioactive and chemically hazardous waste arising from nuclear weapons production, beginning with World War II, and continuing through the Cold War. The waste is stored in 177 carbon-steel underground storage tanks, of which 149 are single-shell tanks (SSTs) and the remaining are double-shell tanks (DSTs). The mission of an ongoing River Protection Project is to retrieve the waste from the underground storage tanks and then treat and immobilize (i.e., vitrify) it for disposal. Waste from the older SSTs is being progressively retrieved into the newer DSTs for storage pending treatment, immobilization, and disposal. Waste chemistry controls are in place to mitigate general corrosion, pitting corrosion, and stress corrosion cracking (SCC) and minimize corrosion risks to the carbon steel tank liners of DST at the Hanford Site. The chemistry control program focuses on preserving liner integrity while balancing the need for caustic/nitrite additions that have large downstream impacts on vitrification. Some of the DSTs store salt cake or sludge waste layers contain out-of-specification interstitial liquids that contact the primary tank bottom or sidewall. Remediation of such layers through chemical additions is not practical. At the present time, the most significant leak integrity concern at Hanford is for corrosion of the tank bottoms of DSTs where the solids layer is out of specification because of hydroxide ion depletion and consequently, pH values that are less than 12. Cathodic protection (CP) has been proposed as a possible way to protect the liners of such tanks against corrosion. The objective of this project was to investigate the feasibility of applying CP to Hanford DSTs to mitigate internal corrosion of the carbon steel tank primary liners. The scope of the project included (1) a literature review on the application of CP to storage tank systems, in general, and to waste tanks at the Savannah River Site (SRS) and Hanford, (2) a review of the finite element analysis (FEA) model recently developed to evaluate potentials for a DST, and (3) the development of a boundary element analysis (BEA) model for CP for a DST. Tank AN-107, one of the Hanford DSTs, was selected for this work because (1) the bottommost waste layer in the tank is out-of-specification, and (2) electrochemical and resistivity data, based on recent waste sampling and testing, are available for this tank. The overarching conclusion of this study is that, based on the available information, it is feasible to apply CP to mitigate internal corrosion of the primary liners of the Hanford DSTs. However, significant additional information and work are required before a CP system can be designed. Recommendations for next steps in the design of the CP system will be presented.

WIERSMA, BRUCE↗
DOE OSTI2022

Verification and validation of the Alternative Nonlinear Two-phase Subchannel (ANTS) code

The Alternative Nonlinear Two-phase Subchannel solver (ANTS) code was written to provide a fast-running, steady-state, pin-resolved modeling and simulation tool for analysis of common boiling water reactor (BWR) geometry and common operating conditions. ANTS has been integrated into the Virtual Environment for Reactor Application (VERA) core simulator software, where it can be used to provide a thermal/hydraulic (T/H) subchannel solution that is then used to provide neutronic feedback as well as perform the fuel depletion and temperature solution. Herein, this paper presents the rigorous analysis performed on the ANTS code, which includes both code and solution verification testing, benchmarking with the existing two-phase subchannel capability in VERA, CTF, and validation testing using popular two-phase experiments such as PWR Sub-channel and Bundle Tests (PSBT), BWR Full-size Fine-mesh Bundle Tests (BFBT), Risø, and FRIGG. This assessment was used to qualify ANTS for its intended applications before its use for core-scale, multiphysics BWR simulations. In general, it was found that agreement with experimental data was good; errors were within the range of experimental data uncertainty. Furthermore, code and solution verification confirmed that the governing equations and the most important closure terms were correctly implemented and behaving as expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1995

Hot bottom burning in asymptotic giant branch stars and its effect on oxygen isotopic abundances

A self-consistent calculation of asymptotic giant branch (AGB) evolution was carried out, including nucleosynthesis at the base of the convective envelope (hot bottom burning). Hot bottom burning was found to occur for stars between approximately 4.5 and approximately 7 solar mass, producing envelopes with O-18/O-16 less than or equal to 10(exp -6) and 10(exp -3) approximately less than or equal O-17/O-16 approximately less than or equal to 10(exp -1). The O-17 abundance depends sensitively on the nuclear O-17-destruction rate; this rate is only loosely constrained by the requirement that first and second dredge-up models match O-isotope observations of red giant branch (RGB) stars (Boothroyd, Sackmann, & Wasserburg 1994). In some cases, high mass-loss rates can terminate hot bottom burning before further O-17 enrichment takes place or even before all O-18 is destroyed. These predictions are in accord with the very limited stellar observations of J type carbon stars on the AGB and with some of the circumstellar Al2O3 grains from meteorites. In contrast, precise data from a number of grains and data from most low-mass S and C AGB stars (approximately less than 1.7 solar mass) lie in a region of the O-18/O-16 versus O-17/O-16 diagram that is not accessible by first and second dredge-up or by hot bottom burning. We conclude that for AGB stars, the standard models of stellar evolution are not in accord with these observations. We surmise that an additional mixing mechanism must exist that transports material from the cool bottom of the stellar convective envelope to a depth at which O-18 is destroyed. This 'cool bottom processing' mechanism on the AGB is similar to extra mixing mechanisms proposed to explain the excess C-13 (and depleted C-12) observed in the earlier RGB stage of evolution and the large Li-7 depletion observed in low-mass main-sequence stars.

Boothroyd, Arnold I.↗
DOE OSTI2023

Improved Fast Reactor Capability of Griffin in FY23

Griffin is a MOOSE based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE-NE NEAMS program. In FY23, we enhanced capabilities required for fast reactor analysis. This effort included primarily updating the cross-section generation workflow using MC2-3 for various reactor configurations, such as homogeneous, duct-heterogeneous, ring-heterogeneous, and fully-heterogeneous geometries. In addition, we initiated the implementation of a multi-cycle depletion and shuffling capability. To support fast reactor simulation capabilities, we significantly improved the performance of the DFEM-SN-based R-Z transport solver to efficiently solve ultrafine group (over 1000 groups) transport problems. Additionally, the performance of HFEM-PN was improved by introducing red-black iteration, the cmfd acceleration technique, and various optimizations. We also completed the pin power reconstruction capability to support multiphysics simulations while identifying and addressing issues associated with SPH equivalence parameter approach. These enhanced capabilities for fast reactor core simulations, specially HFEM-PN and pin power reconstruction features, were applied to benchmark problems involving ABTR and ABR-1000. These applications showcased excellent agreement with Monte Carlo and other code solutions in terms of eigenvalue, control rod worth, and assembly and pin powers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Sub-GeV dark matter and neutrino searches with Skipper-CCDs: status and prospects

Thick fully-depleted charge-coupled devices (CCDs) with high-resistivity silicon are used in a wide range of scientific applications, from particle detection to astronomical imaging. Their low noise and high charge collection efficiency allow us to reach unprecedented sensitivity to physical processes with low-energy transfers. The newly-developed Skipper-CCD enhances this sensitivity by reducing the read-out noise reaching a sub-electron resolution. In this work, we summarize the fundamentals of the skipper-CCD operation and the prospects for both sub-GeV dark-matter searches and the detection of coherent elastic neutrino-nucleus scattering. A brief discussion of the challenges associated with the construction of the foreseen detectors with multi-kilogram target mass is also presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1992

Chemical transport during formation and alteration of Martian impact and volcanic deposits

Much of the surface of Mars, including volcanic and cratered terrains, probably experienced alteration and degassing processes. These processes may have depleted or enriched many important elements in surface materials, including bedrock, dust, and soils. The composition of the martian soil may represent the best estimate, for some elements, of the average composition of the martian crust, similar to the composition of loess created by glacial action on the Earth. The martian soil may represent the only convenient, globally or regionally averaged sample of the martian crust. In order to understand the composition of the source material for the soil, however, we need to understand the contributions of volcanic vs. impact sources for this material and the chemical fractionations involved in its production. The processes to be addressed include degassing of volcanic deposits, as observed in the Valley of Ten Thousand Smokes at Katmai, Alaska, and degassing of meltbearing impact ejecta as inferred for suevite ejecta sheets at the Ries Crater, and alteration or palagonitization of volcanic deposits, as documented for volcanos in British Columbia and many other volcanic terrains, and impact crater deposits. The process of palagonitization has been the subject of several studies with reference to Mars, and palagonite is a good analogue for the spectroscopic properties of the martian dust. The role of impact in cratering has not been as well studied, although other researchers have established that both degassing and alteration are common features of impact crater deposits. Other relevant sources of experimental data include the extensive literature on the corrosion of nuclear waste glass and leaching of shocked materials.

Newsom, H. E.↗
DOE OSTI2025

Advanced Analysis of X-ray Fluorescence Measurements of the Interfacial Density of Eu Ions at Liquid–Liquid Interfaces for Solvent Back-Extraction

X-ray fluorescence near total reflection (XFNTR) is the primary technique used to measure the element-specific interfacial density of ions at liquid−liquid interfaces. Fluorescence from ions in the bulk liquids can complicate the determination of the interfacial density; consequently, measurements have been previously limited to samples without ions in the upper phase and with concentrations on the order of 10 μM or less in the lower phase. We modify the analysis of XFNTR data to account for ions in both bulk phases, then demonstrate its use in the context of rare-earth separations processes. In a model of ion stripping (i.e., back-extraction), dodecane solutions of di(2-ethylhexyl)- phosphoric acid (HDEHP) loaded with Eu(III) at the 1 mM level are placed in contact with either pure water or aqueous solutions of nitric or citric acid at pH 3. XFNTR measurements of equilibrated, quiescent samples reveal that citric acid solutions produce a disproportionately large depletion of ions from the interface compared to that from the bulk organic solution, whereas stripping by pure water or nitric acid solutions is negligible. As a result, this advance in the methodology of XFNTR may have broad applicability to the investigation of metal ions in chemical and biological processes at liquid−liquid interfaces.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Exponential time differencing scheme for mass transport and depletion in molten salt reactors

This work extends the capability previously shown for addressing the problem of computing depletion and mass transport calculations in molten salt reactors (MSRs) by calculating matrix exponentials. Additional algorithms are implemented to compute the matrix exponential and the action of the matrix exponential on a matrix. These algorithms include two methods based on the Pade approximation, a Taylor series method, and three methods based on Cauchy's integral formula. In addition to the added matrix exponential solvers, a variable-order total variation diminishing scheme is applied to the convective flux approximation to provide enhanced accuracy. Finally, a simplified MSR problem is shown for each of the exponential time differencing solvers along with classical backwards differencing integrators. The results show excellent convergence for exponential time differencing methods. Computation time is a key element for selecting the optimal solver in these problems, and this work shows that Pade and Cauchy-based solvers may provided the fastest and most accurate solutions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Assessment of Self-interrogation Safeguards Signatures for Pebble Bed Reactor Fuel

New reactor designs, such as pebble-bed reactors, present challenges for the safeguarding of fissile material. Due to the multi-pass fuel circulation design and the non-uniform path a pebble may traverse through a pebble-bed reactor, there will be variations in the irradiation history of spent fuel pebbles. The standard approach of estimating actinide quantities based on fission gamma spectra or neutron emission paired with depletion modeling may yield uncertainties too large for safeguards and material accountancy. This project investigated, through modeling and experimentation, the potential of neutron self-interrogation of spent fuel pebbles as an innovative method to implement materials accountability. As an example, our feasibility studies indicate that the mass of U-235, U-238, Pu-239 and Pu-241 can be predicted to 4.1, 0.86, 13 and 13 % accuracy, respectively, when measuring 100 closely packed end-of-life spent fuel pebbles over approximately 12 days using a 4π counting geometry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗