Search NASA⌕ Search

SEARCH · Search NASA

Results for “NATRIUM”

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.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Natrium Demonstration Reactor Support [Abstract]

TerraPower, LLC (TerraPower, Participant) and other private industry partners endeavor to design, license, construct, and operate a sodium-cooled fast-spectrum nuclear reactor technology demonstration plant called Natrium. This demonstration plant is supported by the U.S. Department of Energy (DOE) through the Advanced Reactor Demonstration Program (ARDP; DE-FOA-0002271). TerraPower, together with its technology co-developer GE Hitachi Nuclear Energy (GEH) and engineering and construction partner Bechtel, submitted a proposal under the program’s Advanced Reactor Demonstration Pathway for its Natrium reactor and energy system and recently received an award. TerraPower is partnering with Battelle Memorial Institute, the Management and Operating Contractor of Pacific Northwest National Laboratory (PNNL, Contractor) under the Natrium project to provide critical research outcomes necessary to demonstrate the reactor technology. Over the expected five-year timeframe of the project, PNNL will provide TerraPower and its partners with vital support in the areas of post-irradiation examination (PIE) of specimens irradiated in test reactors. These efforts will be combined and managed as a program titled “Natrium Demonstration Reactor Support.”

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Stochastic Optimization and Uncertainty Quantification of Natrium-based Nuclear-Renewable Energy Systems for Flexible Power Applications in Deregulated Markets

Rapid integration of variable renewable energy sources (VRES) has made modeling and stochastic optimization of hybrid energy systems crucial for studying their long-term performance and viability. However, most studies have focused on just historical data, which may be unreliable for capturing short-term fluctuations, rare events, and long-term patterns of energy demand, price, and the variability of renewable energy sources. For this study, optimal synthetic time series models were developed using Wasserstein distance. The models were validated by comparing the key statistical measures against those of the historical data. They were then used to optimize the integrated Natrium-style advanced energy systems and their long-term (30 years) economics. The stochastic model performs bi-level optimization to find the optimal sizes for the balance of plant and thermal energy storage, while also optimizing energy dispatch to achieve the maximum net present value. In studies of two deregulated markets (California ISO and the Electric Reliability Council of Texas), the integrated Natrium-style system performed better in CAISO than in ERCOT, given higher and more consistent electricity prices during peak-demand periods. The potentially enlarged cost associated with the variable operation and maintenance of the TES system also plays a significant role in driving the system sizing, thus its impacts on the system are investigated in detail through comparison against a baseline case. The study also finds that the bi-level optimization results based on stochastic gradient descent closely match the grid search results. The uncertainty quantification of the stochastic signals provides further NPV-related insights and probability distributions for the case studies. The normal standard error of the mean of NPV for the case with and without TES VOM for CAISO were found to be 7.73M (plus-minus sign) 1.09M USD and 104.99M (plus-minus sign) 1.25M USD, respectively based on a 95% confidence. Given the relatively small NPV variance based on 150 samples, the analysis affords the most robust possible prediction of the techno-economic performance of the integrated Natrium-style energy systems.

25 ENERGY STORAGE↗
DOE OSTI2022

ARDP Natrium Neutronic Methodology: Argonne Neutronic Assessment of ABR-1000

The Natrium Sodium-cooled Fast Reactor (SFR) concept developed by TerraPower, LLC was selected as one of the advanced reactor designs for demonstration under the Advanced Reactor Demonstration Program Natrium Demo project. In collaboration with TerraPower, the Argonne National Laboratory (ANL) team provide independent reviews of the neutronic methodology being used for the Natrium design. As part of this activity, both the TerraPower and ANL teams agreed to perform independent neutronics analyses of the ABR-1000 reactor problem as a demonstration on the usage of the neutronics methodologies. Comparison of the results from TerraPower and ANL would expose any methodological differences in the modeling approaches. This report describes and presents the Argonne analysis work on the ABR-1000, and will be provided to TerraPower, along with data files, to complete the comparison study. The comparison will cover the methodology to evaluate the reactivity coefficients for a safety analysis, the shutdown margin, and the reconstructed pin power distribution. The methods and models used to calculate the kinetic parameters and the reactivity coefficients (density, Doppler, axial/radial expansion, etc.) are described in detail with recommendations of specific method options and modelling techniques. The control system reactivity worth and the excess reactivity due to temperature defect are calculated for the shutdown margin analysis. The method to evaluate the excess reactivity is discussed in detail. The pin power reconstruction methodology used in SE2ANL, SE2RCT, and DASSH are discussed in this report. The pin power distribution evaluated by different methods are compared and discussed in the report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Verification of the ENDF/B-VII.1 Based MC 2 -3 Library Rev.1

The MC 2 -3 code, developed by Argonne National Laboratory under the DOE-NE NEAMS program, is a multigroup cross section generation code for fast reactor applications. Last year, the ENDF/B-VII.0 (E70) MC 2 -3 library, which has been extensively used, verified, and validated over a long period, was intensively reverified and updated to support the commercial grade dedication (CGD) requirement of the TerraPower Natrium project. This year, the ENDF/B-VII.1 (E71) MC 2 -3 library, the preliminary version of which was generated several years ago, was regenerated and rigorously verified to support the Natrium project as well as the completion of verification of the E71 library. The E71 library was verified using the process developed during the verification of the E70 library, including comparisons of cross sections with the NJOY-generated cross sections, comparisons of the resolved resonance cross sections with those using the PEDNF library, and comparison of total cross sections with the sum of partial cross sections. Additional verifications were conducted to ensure that the benchmark problem solutions with the E71 library are reasonable compared to the corresponding Monte Carlo solutions. Furthermore, the E71 gamma library was generated, which includes data for prompt gamma, delayed gamma, and delayed beta as well as neutron and gamma heating. The gamma library was verified at the level of individual isotopes. The EBR-II core solutions from MC 2 -3/ DIF3D and MCNP were compared, demonstrating that those solutions in terms of k-effective and assembly powers were in good agreement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Nuclear waste attributes of near-term deployable small modular reactors

The nuclear waste attributes of near-term deployable SMRs were assessed using established nuclear waste metrics, which are the DU mass, SNF mass, volume, activity, decay heat, radiotoxicity, and decommissioning LLW volumes. Metrics normalized per unit electricity generation were compared to a reference large PWR. Three SMRs, VOYGR, Natrium, and Xe-100, were selected because they represent a range of reactor and fuel technologies and are active designs deployable by the decade’s end. The SMR nuclear waste attributes show both some similarities to the PWR and some significant differences caused by reactor-specific design features. The DU mass is equivalent to or slightly higher than the PWR. Back-end waste attributes for SNF disposition vary, but the differences have a limited impact on long-term repository isolation. SMR designs can vary significantly in SNF volume (and thus heat generation density). However, these differences are amenable to design optimization for handling, storage, transportation, and disposal technologies. Nuclear waste attributes from decommissioning vary depending on design and decommissioning technology choices. Given the analysis results in this study and assuming appropriate waste management system and operational optimization, there appear to be no major challenges to managing SMR nuclear wastes compared to the reference PWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Ionic-Based Electrochemical Gas Sensors for Low-Cost, High-Sensitivity SO2 Detection

Sulfur dioxide (SO2) is a toxic gas associated with adverse health and environmental effects that necessitate reliable monitoring techniques. Here, we report the development of an all-solid-state electrochemical sensor utilizing a lithium borate (Li3BO3) solid electrolyte capable of subppm of SO2 detection. While subppm of SO2 sensing has been previously demonstrated in other solid-state electrolyte systems─such as stabilized zirconia, natrium super ionic conductors (NASICON) under mixed-potential conditions─here we establish Li3BO3 as an alternative solid electrolyte enabling equilibrium potentiometric sensing in an all-solid architecture. This sensor demonstrates a detection limit of at least 0.25 ppm, surpassing the human-olfactory threshold and meeting the rigorous requirements for industrial and personal monitoring applications. The sensing mechanism relies on the formation of Li2SO4 on the electrode surface, as evidenced by multimodal characterization techniques, including Raman spectroscopy, scanning electron microscopy (SEM), and scanning transmission electron microscopy (STEM). The strong linear correlation between the open-circuit potential (OCV) and the logarithm of SO2 concentration between 0.25 and 2 ppm indicates that the response is Nernstian in nature.

Lagunas, Francisco (ORCID:000000026377683X)↗
DOE OSTI2025

Assessment of RELAP5-3D Code with Molten Salt Heat Transfer Experiments

The accurate thermal-hydraulic assessment of thermal storage systems is crucial for addressing the integrity and performance of the storage system design. This is particularly important for thermal storage systems using molten salts as a heat transport and storage medium, as the chemical corrosion and erosion by the high-temperature molten salts can cause serious damage to the structural components of the storage system. To understand these effects for TerraPower’s Natrium® Demonstration Reactor design, the system thermal-hydraulic analysis code, RELAP5-3D, is utilized to analyze the thermal storage system. This study evaluates two heat transfer correlations implemented in the RELAP5-3D code—the Dittus-Boelter and Gnielinski correlations—using selected benchmark cases from two molten salt heat transfer experiments conducted at Xi’an Jiao Tong University and the German Aerospace Center. The Nusselt numbers calculated using the Gnielinski correlation agree with the experimental data within ±5% of the relative deviations at 300°C and 400°C. With the increasing Reynolds numbers, the Dittus-Boelter correlation underestimates the Nusselt numbers by up to 22% compared to the Gnielinski correlation. Comparison of the RELAP5-3D calculation results with experimental data at a high temperature of 550°C revealed that the thermo-chemical characteristics of solar salt at high temperatures surpassing the decomposition temperatures of nitrate salts reduce the accuracy of heat transfer model of the RELAP5-3D code. Specifically, it has been observed that the temperatures of the wall surface and liquid near the wall can surpass the decomposition temperatures, even though the bulk temperature of molten salts remains significantly lower than these decomposition temperatures. In summary, this study recommends the use of the Gnielinski correlation for the heat transfer analysis of molten salt, rather than the Dittus-Boelter correlation.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

Materials Data on NaC64 by Materials Project

NaC64 is Boron Nitride-like structured and crystallizes in the hexagonal P6_3/mmc space group. The structure is two-dimensional and consists of two natrium molecules and sixteen C sheets oriented in the (0, 0, 1) direction. In each C sheet, C+0.02- is bonded in a trigonal planar geometry to three equivalent C+0.02- atoms. There is one shorter (1.42 Å) and two longer (1.43 Å) C–C bond length.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on RbNaH2 by Materials Project

NaRbH2 crystallizes in the tetragonal P4/mmm space group. The structure is three-dimensional and consists of one natrium molecule and one RbH2 framework. In the RbH2 framework, Rb1+ is bonded to six H1- atoms to form a mixture of edge and corner-sharing RbH6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Rb–H bond lengths are 2.81 Å. There are two inequivalent H1- sites. In the first H1- site, H1- is bonded in a square co-planar geometry to four equivalent Rb1+ atoms. In the second H1- site, H1- is bonded in a linear geometry to two equivalent Rb1+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na2AgSbCl6 by Materials Project

(Na)2AgSbCl6 is High-temperature superconductor-derived structured and crystallizes in the cubic Fm-3m space group. The structure is three-dimensional and consists of eight natrium molecules and one AgSbCl6 framework. In the AgSbCl6 framework, Ag1+ is bonded to six equivalent Cl1- atoms to form AgCl6 octahedra that share corners with six equivalent SbCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Ag–Cl bond lengths are 2.68 Å. Sb3+ is bonded to six equivalent Cl1- atoms to form SbCl6 octahedra that share corners with six equivalent AgCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Sb–Cl bond lengths are 2.65 Å. Cl1- is bonded in a linear geometry to one Ag1+ and one Sb3+ atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na2SmCuCl6 by Materials Project

(Na)2SmCuCl6 is High-temperature superconductor-derived structured and crystallizes in the cubic Fm-3m space group. The structure is three-dimensional and consists of eight natrium molecules and one SmCuCl6 framework. In the SmCuCl6 framework, Sm3+ is bonded to six equivalent Cl1- atoms to form SmCl6 octahedra that share corners with six equivalent CuCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Sm–Cl bond lengths are 2.68 Å. Cu1+ is bonded to six equivalent Cl1- atoms to form CuCl6 octahedra that share corners with six equivalent SmCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Cu–Cl bond lengths are 2.51 Å. Cl1- is bonded in a linear geometry to one Sm3+ and one Cu1+ atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na2PrCuCl6 by Materials Project

(Na)2PrCuCl6 is High-temperature superconductor-derived structured and crystallizes in the cubic Fm-3m space group. The structure is three-dimensional and consists of eight natrium molecules and one PrCuCl6 framework. In the PrCuCl6 framework, Pr3+ is bonded to six equivalent Cl1- atoms to form PrCl6 octahedra that share corners with six equivalent CuCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Pr–Cl bond lengths are 2.73 Å. Cu1+ is bonded to six equivalent Cl1- atoms to form CuCl6 octahedra that share corners with six equivalent PrCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Cu–Cl bond lengths are 2.51 Å. Cl1- is bonded in a linear geometry to one Pr3+ and one Cu1+ atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na3TmCl6 by Materials Project

(Na)2NaTmCl6 is High-temperature superconductor-derived structured and crystallizes in the cubic Fm-3m space group. The structure is three-dimensional and consists of eight natrium molecules and one NaTmCl6 framework. In the NaTmCl6 framework, Na1+ is bonded to six equivalent Cl1- atoms to form NaCl6 octahedra that share corners with six equivalent TmCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Na–Cl bond lengths are 2.70 Å. Tm3+ is bonded to six equivalent Cl1- atoms to form TmCl6 octahedra that share corners with six equivalent NaCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Tm–Cl bond lengths are 2.58 Å. Cl1- is bonded in a linear geometry to one Na1+ and one Tm3+ atom.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na3ErCl6 by Materials Project

(Na)2NaErCl6 is High-temperature superconductor-derived structured and crystallizes in the cubic Fm-3m space group. The structure is three-dimensional and consists of eight natrium molecules and one NaErCl6 framework. In the NaErCl6 framework, Na1+ is bonded to six equivalent Cl1- atoms to form NaCl6 octahedra that share corners with six equivalent ErCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Na–Cl bond lengths are 2.70 Å. Er3+ is bonded to six equivalent Cl1- atoms to form ErCl6 octahedra that share corners with six equivalent NaCl6 octahedra. The corner-sharing octahedral tilt angles are 0°. All Er–Cl bond lengths are 2.59 Å. Cl1- is bonded in a linear geometry to one Na1+ and one Er3+ atom.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Nuclear Waste Attributes of SMRs Scheduled for Near-Term Deployment

The purpose of this study is to evaluate the nuclear waste attributes of Small Modular Reactors (SMRs) scheduled for deployment within this decade using available data and established nuclear waste metrics, with the results compared to a reference large Pressurized Water Reactor (PWR). The current fleet of commercial nuclear reactors in the U.S. is composed of 92 large Light Water Reactors (LWR) with an average electricity generating capacity of over 1,000 MWe each. These large LWRs built on-site in massive construction projects have been the mainstay of the industry for the last 50 years. However, new construction soon is expected to include several designs of smaller reactors primarily fabricated in factories and installed in the field in modules. Some of these SMRs will also be LWRs, while some will use other coolants such as liquid metals, molten salts or gases, and different types of fuels. The technologies and economics of SMRs have been the focus of many studies, but there has been only minimal information published on the amount of nuclear waste different types of SMRs are expected to generate and no reports focused on near-term-deployable designs. In this study, the nuclear waste attributes of three small reactors scheduled for near-term-deployment, VOYGR TM (from NuScale Power), Natrium TM ab (from TerraPower), and Xe-100 (from X-energy), were assessed by comparing nuclear waste metrics with those of a reference large Pressurized Water Reactor (PWR).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Report on Fuel Cycle Facility Requirements for Deployment of Demonstration Reactors and Potential Evolutionary Fuel Cycle Scenarios

A series of fuel cycle scenarios studies were performed to inform on fuel cycle capacities and facilities needed for large-scale deployment of the Advanced Reactor Demonstration Program (ARDP) reactors and potential future evolutionary fuel cycle scenarios. The reactor deployment and evolutionary fuel cycle scenarios from the present to 2100 were developed based on the following assumptions: 1) achievement of a net-zero emissions economy in the United States by 2050, which requires a nuclear energy generation capacity of ~250 GWe by 2050, 2) the U.S. economic growth of 1% per year from 2051 to 2100, which results in ~340 GWe of nuclear energy capacity in 2100, and 3) commercial-scale recycling and high burnup fuel technologies are available after 2050. Thus, evolutionary fuel cycles with those advanced nuclear technologies start after 2050. A single once-through fuel cycle scenario was assumed from the present to 2050 to achieve a net-zero emissions economy in the United States, and the following four evolutionary fuel cycle scenarios from 2051 to 2100 were considered, 1) Once through fuel cycle with ARDP reactors (Natrium and Xe-100), 2) Once-through fuel cycle with Breed-and-Burn (B&B) fast reactors, 3) Recycling fuel cycle of used metallic fuel in fast reactors, and 4) Recycle fuel cycle of both used uranium oxide and metallic fuels in fast reactors. The projected front-end and back-end fuel cycle capacity demands are compared with the current domestic and global (if needed) fuel cycle capacities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Pros and Cons Analysis of HALEU Utilization in Example Fuel Cycles

The Systems Analysis and Integration campaign assessed the pros and cons of high-assay low-enriched uranium (HALEU) utilization in advanced reactors and associated fuel cycles. The assessment was done for three example fuel cycles at equilibrium states: once-through, limited recycle, and continuous recycle (CR) starting with HALEU. Front- and back-end fuel cycle parameters and the Levelized Cost of Fuel (LCF), which is the Levelized Cost of Electricity excluding reactor cost, of the three example fuel cycles were calculated using a single Analysis Example Reactor. The pros and cons of HALEU utilization were assessed by normalizing the fuel cycle parameters and LCF to a unit of electricity generation (GWe-year) and comparing them with a Basis of Comparison. In this study, a sodium-cooled reactor with sodium-bonded metallic fuel having a burnup of ~100 GWd/t was used as the Analysis Example Reactor because its technology readiness level is high, and the burnup and fuel enrichment are in the middle of those ranges of advanced reactor concepts that are under development. The current once-through Light Water Reactors (OT-LWRs) with <5% low-enriched uranium and 50 GWd/t burnup were used as the Basis of Comparison. In addition, a series of sensitivity analyses was conducted by varying burnup, enrichment, fuel forms, and reactor types to capture the design variations in two once-through Advanced Reactor Demonstration Program (ARDP) reactors, Natrium with sodium-free metallic fuel having a burnup of ~150 GW/t and Xe-100 with Tristructural-Isotropic (TRISO) pebble fuel having a burnup of ~168 GWd/t.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Gripper Test Assembly Status Report (FY2020)

Refueling systems for fast reactors are designed to handle fresh and used core assemblies (fuel, reflector, and shield core assemblies) within the reactor vessel in an opaque coolant environment without visual reference. These refueling machines are designed to work in a sodium (or other fast reactor coolant) and argon vapor space environment and are engineered with the rotatable plug system to allow for the movement of fresh and spent fuel into and out of the reactor core. The refueling machines are a critical component in any reactor and thus need to undergo extensive testing in a prototypic environment to ensure that they will meet all of the system functions and requirements. Argonne has developed an innovative compact refueling system design for the Advanced Fast Reactor-100 that is based upon some mechanisms used in previous reactor designs, such as the U.K.’s Prototype Fast Reactor (PFR) and some mechanisms that have not been used in sodium. This compact refueling machine supports the reduction in size of the AFR-100’s reactor vessel, and if fully developed, would support and inform the development of the in-vessel refueling machines for such commercial reactors as the GEH PRISM reactor plant, the ARC Clean Energy’s ARC-100 reactor, and the Natrium reactor, among others. This refueling system is a vital component of a fast reactor that supports reducing the cost of the reactor and increasing its reliability. During the development of the compact fuel handling machine conceptual design for the AFR-100, a lack of testing data for many mechanical components in sodium under typical in-reactor loads and conditions was discovered. The reduction in lifetime of the various mechanical components in the liquid sodium environment needs to be quantified versus the calculated component lifetimes under normal conditions in the testing while they are subjected to typical loading profiles experienced in the past. The Gripper Test Assembly discussed here includes a full-size gripper device with appropriate mechanical features that will be tested in sodium to provide this testing data. The Gripper Test Assembly is used to test various mechanical fuel handling components submerged in high temperature liquid sodium. These mechanisms are gears, bearings, gripper jaws and head, universal joints and shafts, ball screws, among others. These components will be tested under the typical sodium environmental conditions experienced during refueling operations with appropriate loading conditions that simulate the removal and insertion of core assemblies into a fast reactor grid plate structure. This Gripper Test Assembly is the second in a series of refueling system mechanisms developed for testing in sodium. The first test assembly is the Gear Test Assembly (GTA) which was used to test the performance of gears and bearings operating in sodium. Because of the successful testing conducted with the GTA, it was decided to continue with the development of the Gripper Test Assembly which uses the same gears and bearings tested in GTA. Using the data collected during operation of this gripper test assembly, lifetime reduction factors of the various mechanical components can be calculated for the material combinations selected. These lifetime reduction factors can be used in the design of future mechanical systems which operate in these environments to accurately predict component end of life. In addition, an understanding of the ability of these components and mechanisms to operate under-sodium with the chosen materials will be accomplished. Again, this Gripper Test Assembly is the follow-on test article to the Gear Test Assembly which was testing the ability of gears and bearings operating in a sodium environment. Once fabricated and qualified, it will be tested in the Mechanism Engineering Test Loop (METL) facility located in Building 308 at Argonne National Laboratory.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗