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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 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Strategies for thorium fuel cycle transition in the SD-TMSR

Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/ 233 U fuel cycle with 233 U used as startup fissile material. Since 233 U does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GW d/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and 233 U. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of 233 U are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactorgrade Pu helps transition to the thorium fuel cycle within a relatively short time (≈ 4.5 years) compared to 26 years for 233 U startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time (≈ 40 years) without any external feed of 233 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Experimental and modeling studies of PR and ND oxalate solubility to high ionic strengths: Insight into actinide(III) oxalates

Actinide oxalates are chemical compounds important to nuclear industry, ranging from actinide separation in waste reprocessing, to production of specialty actinides, and to disposal of high level nuclear waste (HLW) and spent nuclear fuel (SNF). In this study, the solubility constants for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O by performing solubility experiments in HNO 3 and mixtures of HNO 3 and H 2 C 2 O 4 at 23.0 ± 0.2 °C have been determined. The targeted starting materials, Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O, were successfully synthesized at room temperature using PrCl 3 , NdCl 3 and oxalic acid as the source metrials. Then, we utilized the targeted solubility-controlling phases to conduct solubility measurements. There was no phase change over the entire periods of experiments, demonstrating that Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O were the solubility-controlling phases in our respective experiments. Based on our experimental data, we have developed a thermodynamic model for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O in the mixtures of HNO 3 and H 2 C 2 O 4 to high ionic strengths. The model for Pr 2 (C 2 O 4 ) 3 •10H 2 O reproduces well the reported experimental data for Pu 2 (C 2 O 4 ) 3 •10H 2 O, which are not utilized for the model development, demonstrating that Pr(III) is an excellent analog for Pu(III). Similarly, the model for Nd 2 (C 2 O 4 ) 3 •10H 2 O reproduces the solubility of Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The Pitzer model was used for the calculation of activity coefficients. Based on the published, well established model for dissociation constants for oxalic acid and stability constants for actinide-oxalate complexes [i.e., AmC 2 O 4 + , and Am(C 2 O 4 ) 2 - ] to high ionic strengths, we have obtained the solubility constants (log 10 K 0 ) for the following reactions at 25 °C, Pr 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Pr 3+ + 3C 2 O 4 2- + 10H 2 O(l). Nd 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Nd 3+ + 3C 2 O 4 2- + 10H 2 O(l). to be -30.82 ± 0.30 (2σ), and - 31.14 ± 0.35 (2σ), respectively. These values for can be directly applied to Pu 2 (C 2 O 4 ) 3 •10H 2 O, Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The model established for actinide oxalates by this study provides the needed knowledge with regard to solubilities of actinide/REE oxalates at various ionic strengths, and is expected to find applications in many fields, including the geological disposal of nuclear waste and the mobility of REE under the surface conditions, as Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O can be regarded as the pure Pr and Nd end-members of deveroite, a recently discovered natural REE oxalate with the following stoichiometry, (Ce 1.01 Nd 0.33 La 0.32 Pr 0.11 Y 0.11 Sm 0.01 Pb 0.04 U 0.03 Th 0.01 Ca 0.04 ) 2.01 (C 2 O 4 ) 2.99 •9.99H 2 O. Regarding its importance in the geological disposal of nuclear waste, Am 2 (C 2 O 4 ) 3 •10H 2 O/Pu 2 (C 2 O 4 ) 3 •10H 2 O/Cm 2 (C 2 O 4 ) 3 •10H 2 O can be the source-term phase for actinides, as demonstrated by the instance in the disposal in clay/shale formations. This is exemplified by the stability of Am 2 (C 2 O 4 ) 3 •10H 2 O in comparison with Am(OH) 3 (am), Am(OH) 3 (s) and AmCO 3 (OH)(s) under the relevant geological repository conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Evaluation of corrosion coupons exposed to molten-chloride electrorefiner salts long term

Corrosion mitigation has long been considered a challenge in long-term service of both pyrochemical reprocessing and molten salt reactor applications. Corrosion coupons consisting of 2.25Cr 1Mo ASME SA387 Grade 22 Class 2 steel, were installed in a uranium electrorefiner containing molten chloride salt at 500°C. This electrorefiner went into service at Idaho National Laboratory in 1994. Seven of these coupons were removed in 2019 (i.e., submerged in molten salt for 25 years) for analysis and characterization including neutron imaging, optical microscopy, scanning electron microscopy, and wave/energy dispersive x-ray spectroscopy. The results showed that all seven coupons experienced between 30 to 175 µm of surface degradation without evidence of brittle fracture. Meaning there was no evidence of accelerated corrosion or fractures, and the average steel surface loss was of less than 5 µm per year. All indications were that the environmental conditions inside the electrorefiner were favorable for the steel alloy.

36 - MATERIALS SCIENCE↗
DOE OSTI2015

Rare earth elements and critical metal content of extracted landfilled material and potential recovery opportunities

Highlights: • Samples from multiple core drills were obtained from 4× landfill sites in the UK. • Each sample analysed for rare earth elements, critical metals and valuable metals. • Two stage microwave digestion method ensuring high yield. • High quantities of copper and aluminium were observed in the soil layers of landfill. • Across 4× landfills aluminium and copper present has a value of around $400 million. - Abstract: Rare earth elements (REEs), Platinum group metals (PGMs) and other critical metals currently attract significant interest due to the high risks of supply shortage and substantial impact on the economy. Their uses in many applications have made them present in municipal solid waste (MSW) and in commercial and industrial waste (C&I), since several industrial processes produce by-products with high content of these metals. With over 4000 landfills in the UK alone, the aim of this study was to assess the existence of these critical metals within landfills. Samples collected from four closed landfills in UK were subjected to a two-step acid digestion to extract 27 metals of interest. Concentrations across the four landfill sites were 58 ± 6 mg kg{sup −1} for REEs comprising 44 ± 8 mg kg{sup −1} for light REEs, 11 ± 2 mg kg{sup −1} for heavy REEs and 3 ± 1 mg kg{sup −1} for Scandium (Sc) and 3 ± 1.0 mg kg{sup −1} of PGMs. Compared to the typical concentration in ores, these concentrations are too low to achieve a commercially viable extraction. However, content of other highly valuable metals (Al and Cu) was found in concentrations equating to a combined value across the four landfills of around $400 million, which increases the economic viability of landfill mining. Presence of critical metals will mainly depend on the type of waste that was buried but the recovery of these metals through landfill mining is possible and is economically feasible only if additional materials (plastics, paper, metallic items and other) are also recovered for reprocessing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Chemically and Mechanically Recyclable Vitrimers from Carbon Dioxide-Based Polycarbonates

Designing thermoset materials with dynamic crosslinks is an important strategy to mitigate rising global carbon dioxide emission levels. The development of polymers from sustainable feedstocks, with efficient manufacturing methods, for high-value applications, and with circular end-of-use solutions is essential for advancing material technologies. One approach involves exploiting carbon dioxide itself as feedstock to create high performance, sustainable materials, by enchaining 50 mol% CO2 via copolymerization with epoxides to yield polycarbonates. This work describes the synthesis, end-functionalization, and curing of poly(propylene carbonate) (PPC) and poly(cyclohexene carbonate) (PCHC) into beta-hydroxy ester vitrimers. These vitrimers demonstrate the ability to be mechanically reprocessed up to 3 times with retention of the material’s properties through dynamic transesterification exchange reactions. The polycarbonate vitrimers with gel fractions exceeding 90 % exhibit high tensile strength (> 50 MPa) and Young’s modulus (> 2 GPa), achieved by varying repeat unit structure in the polymer backbone from the low Tg PPC to the more rigid high Tg PCHC structures. Owing to an entropically favorable chain back-biting mechanism, the network chains can be cleaved and depolymerized into cyclic small molecules. In the case of PCHC, this process enables repolymerization back to polycarbonates with 69 wt.% CO2 retention through life-cycles. The promising mechanical performance and recyclability of these CO2-based polycarbonate vitrimers indicate their potential for sustainable, high-performance materials, paving the way for future innovations in circular polymer technologies and carbon capture utilization.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Probabilistic risk analysis toward cost-effective 3S (safety, safeguards, security) implementation

Probabilistic Risk Analysis (PRA) has been introduced for several decades in safety and nuclear advanced countries have already used this methodology in their own regulatory systems. However, PRA has not been developed in safeguards and security so far because of inherent difficulties in intentional and malicious acts. In this paper, probabilistic proliferation and risk analysis based on random process is applied to hypothetical reprocessing process and physical protection system in nuclear reactor with the Markov model that was originally developed by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) in Generation IV International Framework (GIF). Through the challenge to quantify the security risk with a frequency in this model, integrated risk notion among 3S to pursue the cost-effective installation of those countermeasures is discussed in a heroic manner.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Bulk characterization of (U, Pu) mixed carbide fuel for distribution of plutonium

Homogeneous distribution of plutonium in (U, Pu) mixed fuels is important from fuel performance as well as reprocessing point of view. Radiation imaging and assay techniques are employed for the detection of Pu rich agglomerates in the fuel. A simulation study of radiation transport was carried out to analyse the technique of autoradiography so as to estimate the minimum detectability of Pu agglomerates in MC fuel with nominal PuC content of 70% using Monte Carlo simulations.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2015

A two-dimensional, finite-difference model of the oxidation of a uranium carbide fuel pellet

The oxidation of spent uranium carbide fuel, a candidate fuel for Generation IV nuclear reactors, is an important process in its potential reprocessing cycle. However, the oxidation of uranium carbide in air is highly exothermic. A model has therefore been developed to predict the temperature rise, as well as other useful information such as reaction completion times, under different reaction conditions in order to help in deriving safe oxidation conditions. Finite difference-methods are used to model the heat and mass transfer processes occurring during the reaction in two dimensions and are coupled to kinetics found in the literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Development and evaluation of a silver mordenite composite sorbent for the partitioning of xenon from krypton in gas compositions

A new engineered form composite sorbent for the selective separation of xenon from krypton in simulant composition off-gas streams resulting from the reprocessing of used nuclear fuel has been developed and evaluated. A sodium mordenite powder was incorporated into a macroporous polymer binder, formed into spherical beads and successfully converted to a 9 wt.% silver form composite sorbent. The final engineered form sorbent retained the characteristic surface area indicative of sodium mordenite powder. The sorbent was evaluated for xenon adsorption potential with capacities measured as high as 30 millimoles of xenon per kilogram of sorbent achieved at ambient temperature and 460 millimoles of xenon per kilogram sorbent at 220 K. Xenon/krypton selectivity was calculated to be 22.4 with a 1020 µL/L xenon, 150 µL/L krypton in a balance of air feed gas at 220 K. Furthermore, adsorption/desorption thermal cycling effects were evaluated with results indicating sorbent performance was not significantly impacted while undergoing numerous adsorption/desorption thermal cycles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

NO 2 -mediated voloxidation for iodine separation from cesium iodide surrogates

Heterogeneous NO 2 -mediated oxidation of uranium, also known as advanced voloxidation, is a proposed head-end reprocessing method for used nuclear fuel. An advantage of advanced voloxidation is the removal of volatile fission products, which complicate downstream separation and containment challenges leading to increased processing economics. Iodine, one of the volatile species of interest, has exhibited varied results in this process. Using CsI as a surrogate material, this work mimics the effect of NO 2 -based voloxidation on iodine and sheds light on the factors that influence the solid–gas phase reaction. Solid-state analysis using Fourier transform infrared attenuated total reflectance spectroscopy and scanning electron microscopy with energy-dispersive X-ray spectroscopy confirmed the conversion of CsI to CsNO 3 . Iodine separation ranged from 46% to 100% across multiple tests. Iodine separation was most effective when multiple recharges of NO 2 were administered. In conclusion, at the bench scale, liberating iodine from CsI appears to occur within 1 h, but the presence of surface H 2 O and the composition of the NO x reagent mixtures greatly influence its success.

advanced voloxidation↗
DOE OSTI2015

DETERMINING THE COVERING FACTOR OF COMPTON-THICK ACTIVE GALACTIC NUCLEI WITH NuSTAR

The covering factor of Compton-thick (CT) obscuring material associated with the torus in active galactic nuclei (AGNs) is at present best understood through the fraction of sources exhibiting CT absorption along the line of sight (N{sub H} > 1.5 × 10{sup 24} cm{sup −2}) in the X-ray band, which reveals the average covering factor. Determining this CT fraction is difficult, however, due to the extreme obscuration. With its spectral coverage at hard X-rays (>10 keV), Nuclear Spectroscopic Telescope Array (NuSTAR) is sensitive to the AGNs covering factor since Compton scattering of X-rays off optically thick material dominates at these energies. We present a spectral analysis of 10 AGNs observed with NuSTAR where the obscuring medium is optically thick to Compton scattering, so-called CT AGNs. We use the torus models of Brightman and Nandra that predict the X-ray spectrum from reprocessing in a torus and include the torus opening angle as a free parameter and aim to determine the covering factor of the CT gas in these sources individually. Across the sample we find mild to heavy CT columns, with N{sub H} measured from 10{sup 24} to 10{sup 26} cm{sup −2}, and a wide range of covering factors, where individual measurements range from 0.2 to 0.9. We find that the covering factor, f{sub c}, is a strongly decreasing function of the intrinsic 2–10 keV luminosity, L{sub X}, where f{sub c} = (−0.41 ± 0.13)log{sub 10}(L{sub X}/erg s{sup −1})+18.31 ± 5.33, across more than two orders of magnitude in L{sub X} (10{sup 41.5}–10{sup 44} erg s{sup −1}). The covering factors measured here agree well with the obscured fraction as a function of L{sub X} as determined by studies of local AGNs with L{sub X} > 10{sup 42.5} erg s{sup −1}.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2015

PS1-10jh CONTINUES TO FOLLOW THE FALLBACK ACCRETION RATE OF A TIDALLY DISRUPTED STAR

We present late-time observations of the tidal disruption event candidate PS1-10jh. UV and optical imaging with Hubble Space Telescope/WFC3 localize the transient to be coincident with the host galaxy nucleus to an accuracy of 0.023 arcsec, corresponding to 66 pc. The UV flux in the F225W filter, measured 3.35 rest-frame years after the peak of the nuclear flare, is consistent with a decline that continues to follow a t{sup −5/3} power-law with no spectral evolution. Late epochs of optical spectroscopy obtained with MMT ∼ 2 and 4 years after the peak, enable a clean subtraction of the host galaxy from the early spectra, revealing broad helium emission lines on top of a hot continuum, and placing stringent upper limits on the presence of hydrogen line emission. We do not measure Balmer Hδ absorption in the host galaxy that is strong enough to be indicative of a rare, post-starburst “E+A” galaxy as reported by Arcavi et al. The light curve of PS1-10jh over a baseline of 3.5 years is best modeled by fallback accretion of a tidally disrupted star. Its strong broad helium emission relative to hydrogen (He iiλ4686/Hα > 5) could be indicative of either the hydrogen-poor chemical composition of the disrupted star, or certain conditions in the tidal debris of a solar-composition star in the presence of an optically thick, extended reprocessing envelope.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2015

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m{sup 3}Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Evaluation of the response of tritium-in-air instrumentation to HT in dry and humid conditions and to HTO vapor

Nuclear plant operators (power generation, decommissioning and reprocessing operations) are required to monitor releases of tritium species for regulatory compliance and radiation protection purposes. Tritium monitoring is performed using tritium-in-air gas monitoring instrumentation based either on flow-through ion chambers or proportional counting systems. Tritium-in-air monitors are typically calibrated in dry conditions but in service may operate at elevated levels of relative humidity. The NPL (National Physical Laboratory) radioactive gas-in-air calibration system has been used to study the effect of humidity on the response to tritium of two tritium-in-air ion chamber based monitors and one proportional counting system which uses a P10/air gas mixture. The response of these instruments to HTO vapour has also been evaluated. In each case, instrument responses were obtained for HT in dry conditions (relative humidity (RH) about 2%), HT in 45% RH, and finally HTO at 45% RH. Instrumentation response to HT in humid conditions has been found to slightly exceed that in dry conditions. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

ORNL experience and perspectives related to processing of thorium and 233 U for nuclear fuel

Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233 U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗