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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Final Report

Uranium-containing and thorium-containing compounds have been produced using a laser ablation source. Spectral transitions from these compounds in the 6 GHz to 18 GHz frequency region have been recorded using advanced techniques in Fourier transform microwave spectroscopy. The pure rotational spectrum of thorium (II) oxide is particularly strong and rotational transitions have been observed in highly excited vibrational states. These measurements have allowed a further characterization of the molecules potential energy well.

Heavy Element Chemistry↗
DOE OSTI2021

Modeling 233 Pa Generation in Thorium-fueled Reactors for Safeguards

Thorium has been considered as a possible alternative to uranium for nuclear fuel for many decades. It is three to four times more abundant in the earth than uranium and produces significantly less long-lived transuranic nuclear waste. Some claim thorium poses fewer proliferation concerns than other fuel types largely due to 232 U buildup (and associated high energy gamma-emitting decay products) in the irradiated thorium fuel. However, to fully explore potential proliferation concerns, generation and subsequent decay of 233 Pa produced in the reactor core still must be studied. With its half-life of 27 days, 233 Pa decays to 233 U, which is an International Atomic Energy Agency (IAEA) defined special fissionable material that can be used for nuclear weapons production. With more research being dedicated to thorium-fueled reactors, and several of these reactor designs possessing online fuel processing (allowing for on-site protactinium separation), it is important to understand this potential proliferation pathway. In particular, it is theoretically possible to extract protactinium from the irradiated fuel salt before it decays into 233 U. This hypothetical potential diversion can become an even greater proliferation concern if the extracted protactinium is purified through a second separation of protactinium approximately ten days later to remove the short half-life decay products of 232 Pa and 234 Pa, thus resulting in a higher concentration of the 233 Pa isotope, which decays into weapons usable 233 U with hardly any 232 U or 234 U in it. To estimate the concern of this potential proliferation challenge of thorium, different nuclear material accountancy techniques were reviewed for their viability to quantify 233 Pa if extracted from used thorium fuel. Characteristics of interest included technology maturity, cost, precision, and time taken to acquire results. Some technologies, like hybrid K-edge densitometry and passive gamma spectroscopy, appear to be viable techniques based on current literature. Due to the limited scope of this project, only passive gamma spectroscopy was further investigated. Three different reactor types (PWR, CANDU, MSR) were modeled with mixed thorium-uranium oxide fuels that were burned until the fuel was spent. The protactinium in the used fuel was extracted at the time of shutdown and the change in isotopic content of the protactinium quantified. Gamma spectroscopy simulations were performed for the protactinium isotopes and their decay products at various decay times. Given the simplicity of the models and large assumptions made (e.g. no background, no shielding, no self-attenuation), the initial results indicate that though 233 Pa is detectible for all the reactor types modeled at all decay times (0 to 300 days), more work should be done with higher fidelity models.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

In this paper, transmutation of transuranic (TRU) nuclides from LWR spent fuels is studied by using LWR fuel assemblies which consist of UO{sub 2}-ThO{sub 2} fuel pins and FCM (Fully Ceramic Microencapsulated) fuel pins. TRU from LWR spent fuel is loaded in the kernels of the TRISO particle fuels of FCM fuel pins. In the FCM fuel pins, the TRISO particle fuels are distributed in SiC matrix having high thermal conductivity. The loading patterns of fuel pins and the fuel compositions are searched to have high transmutation rate and feasible neutronic parameters including pin power peaking, temperature reactivity coefficients, and cycle length. All studies are done only in fuel assembly calculation level. The results show that our fuel assembly designs have good transmutation performances without multi-recycling and without degradation of the safety-related neutronic parameters. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Extreme Spectroscopy: In situ nuclear materials behavior from optical data

In the nuclear industry, materials are regularly exposed to high temperature or/and irradiation and a better knowledge and understanding of their behavior under such extreme conditions is a key-point for improvements and further developments. Nowadays, Raman spectroscopy begins to be well known as a promising technique in the post mortem and remote characterization of nuclear materials exposed to extreme conditions. On this topic, at ANIMMA 2013 conference, we have presented some results about its implementation in the study of model or real nuclear fuel. However, the strength of Raman spectroscopy as in situ characterization tool is mainly its ability to be implemented remotely through optical fibers. Aware of this, implementation of other optical techniques can be considered in order to gain information not only on the structural dynamics of materials but also on the electronic charge carrier populations. In this paper, we propose to present our last advances in Raman characterization of nuclear materials and enlarge to the in situ use of complementary optical spectroscopies. Emphasis will be made on the information that can be gained to the behavior of the model fuel depleted UO{sub 2} under extreme conditions of high temperature and ionic irradiation: - In Situ Raman identification of the radiolysis alteration products of UO{sub 2} in contact with water under ionic irradiation. - In Situ Raman recording of the damaged dynamic of UO{sub 2} under inert atmosphere. - In Situ Raman and photo-luminescence study of virgin and damaged UO2 at high temperature. - In Situ study of electronic charge carriers' behavior in U{sub x}Th{sub 1-x}O{sub 2} solid solutions by mean of Iono- and Thermo- luminescence under and post- ionic irradiation. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Evaluation of B and W UO{sub 2}/ThO{sub 2} VIII Experimental Core: Criticality and Thermal Disadvantage Factor Analysis

In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock and Wilcox (B and W) Spectral Shift Control Reactor (SSCR), critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO{sub 2}-ThO{sub 2} fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated. Benchmark criticality results using different cross-section libraries are reported. ENDF/B-VI.6 calculation was run using the S(α, β) model based on ENDF/B-VI.3 data while ENDF/B-VII.0 and ENDF/BVII.1 used an S(α, β) model based on ENDF/B-VII.0 data. JEFF calculations were run using the S(α, β) model based on JEFF 3.1.1 data except for the aluminum isotope, for which the S(α, β) model was missing. The maximum absolute deviation from the benchmark model was obtained using the JEFF-3.1.1 cross sections while closer results were obtained using ENDF/ B-VII.0 and ENDF/B-VII.1. Only ENDF/B-VII.1 and ENDF/B-VII.0 results are inside the 3σ uncertainty band (+2.8σ and 2.4σ, respectively), having deviations on the order of 300 pcm. The maximum deviation is -827 pcm (-6.1σ), using JEFF.3.1.1. Sample calculations of the TDF benchmark were also performed, studying the effects of the main nuclear data libraries. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Atomistic Simulations in Support of UO{sub 2} Fuel Performance Modeling

Large fission gas atoms such as Xe have very low solubility in the UO{sub 2} fuel matrix, which drives precipitation and growth of intra-granular fission gas bubbles, segregation of gas atoms to grain boundaries and evolution of intergranular bubbles. The inter-granular bubbles develop percolation networks leading to release of the fission gas. Fission gas bubbles induce fuel swelling and the fuel thermal conductivity is reduced by both fission gas bubbles and dispersed gas atoms. When fission gas is released from the fuel to the plenum, the pressure on the clad walls increases and the gap thermal conductivity decreases. All of these detrimentally affect fuel performance. The thermal conductivity of UO{sub 2} is low across a wide temperature range compared to iso-structural materials such as ThO{sub 2}, which is caused by the coupling between phonons carrying heat and uranium spins or magnetism providing a scattering mechanism with very low mean free path. For in-pile conditions the thermal conductivity is further reduced due to, for example, radiation damage, generation of fission products and fission gases and fuel microstructure evolution. This results in a complex burn-up dependent thermal conductivity. Thermal conductivity and fission gas evolution are two of the most important properties determining UO{sub 2} nuclear fuel performance. Both properties have significant uncertainty. In order to reduce this uncertainty, new models are being developed based on multi-scale simulations that attempt to improve the description of material properties and also couple them to the microstructure state of the material. In this paper we highlight atomistic calculations and simulations used to develop meso-scale models of the evolution of fission gases and thermal conductivity in UO{sub 2}. Literature examples of the application of these results and the corresponding models to fuel performance simulations will also be given. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Synthesis of ThN using a Carbothermic Reduction to Nitridation Process

The thorium fuel cycle has several advantages when compared to the uranium fuel cycle. Not only is thorium more abundant than uranium, it also avoids many of the disadvantages which are inherent to the uranium fuel cycle. For example, the thorium fuel cycle does not result in significant formation of long lived, highly toxic transuranics that are inevitably formed in the uranium fuel cycle. The fissile U-233 that is created through neutron capture of Th-232 is also more proliferation resistant than the Pu-239 created from the uranium fuel cycle, because of the buildup of U-232, a source of extremely hazardous gamma emitters. Additionally, thermal reactors which utilize the thorium fuel cycle may be used to destroy transuranics through transmutation as opposed to the fast reactors that are typically required for the uranium fuel cycle. The result is that not only does the thorium fuel cycle avoid the production of transuranic actinides, it may also be used to destroy the inventory of these isotopes within existing reactors. The work presented here investigates the fabrication of ThN so that its properties may be studied to determine its feasibility as a nuclear fuel within a thorium fuel cycle. While both ThO{sub 2} and ThN have melt points higher than the metal, one of the potential advantages of using ThN instead of ThO{sub 2} is that it has a 28% higher fertile density. Additionally, previous investigations have suggested that ThN has substantially higher thermal conductivity when compared to ThO{sub 2} fuel. This suggests that if a (Th,U)N solid solution can be shown to behave similarly to (U,Pu)N, the nitride fuel form would possess a significant advantage over (Th,U)O{sub 2} which has poor thermal conductivity. A carbothermic reduction to nitridation (CTR-N) process has been utilized extensively at LANL for the production and study of UN, but this type of process has not been used for the fabrication of high purity ThN. The work presented here describes the development of a similar process which will be used to fabricate bulk ThN for subsequent experimental study. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Assessment of Near-Breeding Heterogeneous Seed/Blanket Cores in Pressure-Tube Heavy Water Reactors with Thorium-Based Fuels

To help extend uranium resources, thorium, which is three to four times as abundant as uranium, can be exploited as an alternative fertile nuclear fuel. The pressure tube heavy water reactors (PT-HWR) is an excellent mainstream technology for harnessing the energy potential in thorium due to its high neutron economy enabled by using heavy water as a moderator and coolant, and by refueling online. The PT-HWR is also capable of flexibility in fuel management, allowing the use of multiple fuel types with different refueling rates. One approach to using thorium-based fuels in PT-HWRs is to use a homogeneous core with a once-through thorium (OTT) fuel cycle, where a fissile material is mixed with thorium (such as (LEU,Th)O{sub 2} or (Pu,Th)O{sub 2}), burned in a high-conversion reactor, and then put into storage for several decades until it is economically attractive to reprocess and recycle. An alternative approach is to use a heterogeneous seed/blanket core, where the seed/driver fuel is made of ({sup 233}U,Th)O{sub 2}, and the blanket fuel is make of ThO{sub 2}, and attempt to achieve a very-high conversion ratio (CR) core, and perhaps net breeding with a self-sufficient equilibrium thorium (SSET) fuel cycle. The objective of this study was to assess the performance of various heterogeneous seed/blanket cores in a 700-MWe-class PT-HWR using ({sup 233}U,Th)O{sub 2} fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Opportunities to Reuse and Recycle Redundant Radioisotopes - 16023

Sixty plus years of nuclear research and operation in the UK has produced significant quantities of nuclear material. To those involved in decommissioning a site or facility, this material represents a liability that needs to be managed and eventually disposed of. However to those in the wider nuclear industry and supply chain, the radioisotopes represent a potential resource of valuable, and in some cases irreplaceable material for future work. There is therefore an opportunity to create a win-win by identifying material that could be recycled or reused, removing it from being a liability and creating value from what is considered a waste. The UK is currently focused on the clean up of nuclear sites and the safe disposal of legacy nuclear material. Whilst the bulk of legacy material on these sites is mixed wastes in various solid and liquid forms, some of it consists of material produced for research purposes in the form of radioisotopes or exotic fuels. This has resulted in a plethora of unirradiated and irradiated materials in store across the UK including: Depleted and natural Uranium, Enriched Uranium, Mixed U/Pu oxides, Uranium carbide and mixed U/Pu carbides, Thorium as metal and oxides, Research radioisotopes. The material exists in almost every conceivable state from pure metal and oxides to solutions. Some material is well characterised and free from extraneous debris, other material is poorly characterised and contaminated with chemicals, experimental residues and sweepings. In addition, to reduce the risk and potential hazard across the UK there has also been a period of consolidation of obsolete radioactive sources from a wide variety of industrial or medical facilities to secure storage within a licensed site. The waste hierarchy is an established principle for the management of wastes, radioactive or otherwise. Its principles are set out and it gives top priority to prevention in the first place. Where waste is produced, it gives priority to preparation for re-use, then recycling before consideration for disposal as a last resort. The term wastes may only be applicable to the owner of the material where they may view the material as unwanted, unusable, or a by-product. It does not necessarily follow that the material is a waste to all potential owners. Therefore it is important to consider the potential value of material before it is considered for disposal since there is a significant and growing market for radioisotopes across the world. This paper will outline some of the opportunities in the UK for the recycle of surplus radioisotopes and orphan material. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Metallic Coating of Cerium Oxide Microspheres

The ability to remove heat is paramount to nuclear fuel performance and longevity. Retaining fission product and separating fuel from reactor coolant and the environment is also necessary to prevent radiological contamination. Conventional nuclear fuel for commercial light water reactors and radioisotope power systems (RPS) is composed of oxide powders pressed into a pellet (cm-scale) and then sealed into a metal cladding to confine the fuel. What typical fuels lack is a method to surround each particle of nuclear fuel in metal, thus providing a more intimate protection layer for accident tolerance and boosting the thermal extraction from the fuel element. In such a way, metal-coated fuel particles increase heat extraction efficiency over clad-pellet designs while increasing the accident tolerance of the fuel. Metal oxide microspheres have wide-ranging applications, including the realm of fuels for nuclear reactors and RPS. Microspheres of uranium oxide/uranium carbide, mixed uranium/plutonium oxides, transuranics, and thorium fuels have been extensively studied. Pacific Northwest National Laboratory has also demonstrated the production of 238 PuO 2 microspheres for RPS applications. Metal-coated oxide microsphere fuels may also be attractive for other applications such as nuclear thermal rockets, future nuclear reactor designs, and catalysts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Solid-state syntheses and single-crystal characterizations of three tetravalent thorium and uranium silicates

Colorless crystals of ThSiO{sub 4} (huttonite) (1) and (Ca{sub 0.5}Na{sub 0.5}){sub 2}NaThSi{sub 8}O{sub 20} (2) have been synthesized by the solid-state reactions of ThO{sub 2}, CaSiO{sub 3}, and Na{sub 2}WO{sub 4} at 1073 K. Green crystals of (Ca{sub 0.5}Na{sub 0.5}){sub 2}NaUSi{sub 8}O{sub 20} (3) have been synthesized by the solid-state reactions of UO{sub 2}, CaSiO{sub 3}, and Na{sub 2}WO{sub 4} at 1003 K. All three compounds have been characterized by single-crystal X-ray diffraction. Compound 1 adopts a monazite-type three-dimensional condensed structure, which is built from edge- and corner-shared ThO{sub 9} polyhedra and SiO{sub 4} tetrahedra. Compounds 2 and 3 are isostructural and they crystallize in a steacyite-type structure. The structure consists of discrete pseudocubic [Si{sub 8}O{sub 20}]{sup 8−} polyanions, which are connected by An{sup 4+} cations into a three-dimensional framework. Each An atom coordinates to eight monodentate [Si{sub 8}O{sub 20}]{sup 8−} moieties in a square antiprismatic geometry. Na{sup +} and Ca{sup 2+} ions reside in the void within the framework. Raman spectra of 1, 2, and 3 were collected on single crystal samples. 1 displays more complex vibrational bands than thorite. Raman spectra of 2 and 3 are analogous with most of vibrational bands located at almost the same regions. - Graphical abstract: A Raman spectrum and crystal structures of (Ca{sub 0.5}Na{sub 0.5}){sub 2}NaAnSi{sub 8}O{sub 20} (An=Th, U), which contain pseudocubic [Si{sub 8}O{sub 20}]{sup 8−} polyanions and eight-coordinate An{sup 4+} cations. - Highlights: • Single crystal growth of three tetravalent actinide silicates from melts. • Single-crystal structures and Raman spectra of (Ca{sub 0.5}Na{sub 0.5}){sub 2}NaAnSi{sub 8}O{sub 20} (An=Th, U). • First report of Raman spectrum of huttonite on single crystal samples.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Analysis of Synergistic Fuel Cycle Options with Thorium and Heavy Water Reactors

Fuel cycle options (FCOs) with thorium-based fuels have characteristics that may have benefits for meeting certain objectives, such as transuranic actinide inventory management and overall natural resource utilization. Neutron irradiation of thorium and uranium-233 results in negligible production of transuranic actinides. This advantage is compromised, to a degree, when thorium fuels are used in conjunction with uranium-238; however, strategic reprocessing and fuel use (e.g., recycle of plutonium and/or minor actinides) can be used to manage this problem. Furthermore, thorium FCOs with reprocessing can extend the availability of natural nuclear resources by breeding new fissile material, recovering and recycling unused material, and achieving high burnups. In instances where natural thorium is more readily available than uranium, FCOs which use thorium may be preferable from sustainability and energy security perspectives. However, many hybrid uranium-thorium FCOs struggle to perform well in this regard, since large amounts of external fissile material (and therefore natural uranium) are required to breed the required uranium-233. DOE's evaluation and screening effort reports that for every metric ton (MT) of U-233 bred in a light water reactor for reuse, about 1000 MT of natural uranium are required to support such breeding, much of which is used inefficiently due to enrichment. The combination of HWRs and thorium-based fuels has the potential to reduce this consumption of natural uranium resources. Because heavy water moderates neutrons without significant absorption, fissile material requirements are lower compared to those of LWRs. This has enabled conventional 'once-through' HWRs to operate with only slightly-enriched or even natural uranium. Also, HWRs can be combined with LWRs in multistage FCOs where used LWR fuel can be processed and the remaining fissile content is sufficient to support use as fuel in HWRs in spite of the presence of neutron poisons (e.g., uranium-236); this concept has been demonstrated at a research scale in Korea. Thus, additional energy is extracted from the same initial natural uranium feed and uranium enrichment effort. Another consideration with using thorium-based fuels in HWRs is the current limited operational experience of the reactors and associated fuel cycle facilities using these fuels. However, work continues internationally towards gaining further experience. For example, at equilibrium the third stage of India's planned Three-Stage Nuclear Power Program uses thorium/uranium-233 or thorium/plutonium fuels with a fissile supply from second-stage fast reactors. Further, China plans to experiment with low-enriched uranium (LEU)-Th CANFLEX fuel bundles in a once-through configuration in its HWRs and a long-term objective is to transition to thorium use in a closed HWR FCO. Vanderbilt University (VU) and Oak Ridge National Laboratory (ORNL) are characterizing multi-stage, thermal-reactor FCOs that incorporate thorium and examining their potential performance. This paper presents insights from the examination of FCOs which consider the synergy of HWRs and thorium. Two multi-stage FCOs which incorporate thorium-based fuels in HWRs have been examined for their potential, one which uses recycled PWR uranium oxide fuel in conjunction with thorium in an HWR, and another that uses recycled HWR fuel in an MSR which also includes thorium. Both of the systems studied offer resource utilization improvements as compared to a once-through LWR fueled with low-enrichment uranium and highlight the potential synergy of thorium and HWRs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Neutron Coincidence Measurements of Uranium-233 Oxide

Renewed international interest in thorium-fueled advanced reactors has challenged the safeguards community to address future proliferation concerns. Thorium-based technology presents many benefits but does not eliminate the proliferation risks associated with producing and processing fissile material. A byproduct of thorium-fueled reactors is uranium-233, which is classified as a direct-use material. As a result, the development of new or improved methods to characterize and measure materials containing 233U must mirror the pace of development of reactors and facilities that produce such material. Research is underway to assess, develop, and test approaches for safeguarding nuclear materials within the thorium fuel cycle. Neutron signatures from the nondestructive assay (NDA) of materials containing 233U are being quantified to inform the potential characterization of these materials. Using a traditional neutron coincidence counter and a series of well-documented 233U oxide samples, initial measurements have been made to assess the feasibility of 233U characterization and discrimination from other uranium isotopes, primarily 235U, using a combination of measurement techniques and analysis methods. Data acquisition is performed in list mode, allowing for a variety of analyses to be performed on the raw data that is not available using traditional shift register technology. Measurements were performed in passive and active configurations to quantify the strength of signal and to validate simulations in support of this work. This paper presents and discusses the results of the initial measurements of 233U oxide performed at Oak Ridge National Laboratory.

Lockhart, Madeline↗