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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Fission Product Yields from 232Th, 238U, and 235U Using 14 MeV Neutrons

Neutron-induced fission yield studies using deuterium-tritium fusion-produced 14 MeV neutrons have not yet directly measured fission yields from fission products with half-lives on the order of seconds (far from the line of nuclear stability). Fundamental data of this nature are important for improving and validating the current models of the nuclear fission process. Cyclic neutron activation analysis (CNAA) was performed on three actinide targets–thorium-oxide, depleted uranium metal, and highly enriched uranium metal–at the University of Michigan's Neutron Science Laboratory (UM-NSL) using a pneumatic system and Thermo-Scientific D711 accelerator-based fusion neutron generator. This was done to measure the fission yields of short-lived fission products and to examine the differences between the delayed fission product signatures of the three actinides. The measured data were compared against previously published results for 89Kr, -90, and -92 and 138Xe, -139, and -140. The average percent deviation of the measured values from the Evaluated Nuclear Data Files VII.1 (ENDF/B-VII.1) for thorium, depleted-uranium, and highly-enriched uranium were -10.2%, 4.5%, and -12.9%, respectively. In addition to the measurements of the six known fission products, 23 new fission yield measurements from 84As to 146La are presented.

Pierson, Bruce D.↗
DOE OSTI2013

Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Sample Dissolution and Purification for 230 Th- 234 U Age-Dating

This is the standard operating procedure used by the Isotope Ratio Mass Spectrometry Group of the Chemical Sciences Division at LLNL for the preparation of a sample of uranium oxide or uranium metal for 230 Th- 234 U age-dating. The method described here includes the dissolution of a sample of uranium oxide or uranium metal, preparation of a secondary dilution, spiking of separate aliquots for uranium and thorium isotope dilution measurements, and purification of uranium and thorium aliquots for mass spectrometry. This SOP may be applied to uranium samples of unknown purity as in a nuclear forensic investigation, and also to well-characterized samples such as, for example, U 3 O 8 and U-metal certified reference materials. The sample of uranium is transferred to a quartz or PFA vial, concentrated nitric acid is added and the sample is heated on a hotplate at approximately 100°C for several hours until it dissolves. The sample solution is diluted with water to make the solution approximately 4 M HNO 3 and hydrofluoric acid is added to make it 0.05 M HF. A secondary dilution of the primary uranium solution is prepared. Separate aliquots for uranium and thorium isotope dilution measurements are taken and spiked with 233 U and 229 Th, respectively. The spiked aliquot for uranium isotope dilution analysis is purified using EiChrom UTEVA resin. The spiked aliquot for thorium isotope dilution analysis is purified by, first, a 1.8 mL AG1x8 resin bed in 9 M HCl on which U adsorbs and Th passes through; second, adsorbing Th on a 1 mL AG1x8 resin bed in 8 M HNO 3 and then eluting it with 9 M HCl followed by 0.1 M HCl + 0.005 M HF; and third, by passing the Th through a final 1.0 mL AG1x8 resin bed in 9 M HCl. The mass spectrometry is performed using the procedure 'Th and U Mass Spectrometry for 230 Th- 234 U Age Dating'.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2010

Fuel cycle analysis of once-through nuclear systems.

Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium (LEU) fuels. Examples of systems in this class include the small modular reactors being considered internationally; e.g. 4S [Tsuboi 2009], Hyperion Power Module [Deal 2010], ARC-100 [Wade 2010], and SSTAR [Smith 2008]. (2) Systems for Resource Utilization - In recent years, interest has developed in the use of advanced nuclear designs for the effective utilization of fuel resources. Systems under this class have generally utilized the breed and burn concept in which fissile material is bred and used in situ in the reactor core. Due to the favorable breeding that is possible with fast neutrons, these systems have tended to be fast spectrum systems. In the once-through concepts (as opposed to the traditional multirecycle approach typically considered for fast reactors), an ignition (or starter) zone contains driver fuel which is fissile material. This zone is designed to last a long time period to allow the breeding of sufficient fissile material in the adjoining blanket zone. The blanket zone is initially made of fertile depleted uranium fuel. This zone could also be made of fertile thorium fuel or recovered uranium from fuel reprocessing or natural uranium. However, given the bulk of depleted uranium and the potentially large inventory of recovered uranium, it is unlikely that the use of thorium is required in the near term in the U.S. Following the breeding of plutonium or fissile U-233 in the blanket, this zone or assembly then carries a larger fraction of the power generation in the reactor. These systems tend to also have a long cycle length (or core life) and they could be with or without fuel shuffling. When fuel is shuffled, the incoming fuel is generally depleted uranium (or thorium) fuel. In any case, fuel is burned once and then discharged. Examples of systems in this class include the CANDLE concept [Sekimoto 2001], the traveling wave reactor (TWR) concept of TerraPower [Ellis 2010], the ultra-long life fast reactor (ULFR) by ANL [Kim 2010], and the BNL fast mixed spectrum reactor (FMSR) concept [Fisher 1979]. (3) Thermal systems for resource extension - These systems were primarily considered during the INFCE/NASAP evaluations [NASAP 1979] and include various LWR designs for increasing resource utilization (both uranium and thorium). This class would include the Radkowsky seed-blanket concept. Also included in this class are the thermal reactor systems being considered for deployment as small modular reactors, such as IRIS [Carelli 2004], mPower [mPower], and NuScale [NuScale] that are all water cooled reactors. The purpose of this work is to provide relevant systems and fuel cycle information for some of these once-through fuel cycle systems. In this report, the intent is on providing information on most of the systems from open sources and from scoping studies recently done within the program. As there is insufficient fuel cycle information on the first class of systems, they are not discussed in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Separation of Radionuclides from a Rare Earth-Containing Solution by Zeolite Adsorption

The increasing industrial demand for rare earths requires new or alternative sources to be found. Within this context, there have been studies validating the technical feasibility of coal and coal byproducts as alternative sources for rare earth elements. Nonetheless, radioactive materials, such as thorium and uranium, are frequently seen in the rare earths’ mineralization, and causes environmental and health concerns. Consequently, there exists an urgent need to remove these radionuclides in order to produce high purity rare earths to diversify the supply chain, as well as maintain an environmentally-favorable extraction process for the surroundings. In this study, an experimental design was generated to examine the effect of zeolite particle size, feed solution pH, zeolite amount, and contact time of solid and aqueous phases on the removal of thorium and uranium from the solution. The best separation performance was achieved using 2.50 g of 12-µm zeolite sample at a pH value of 3 with a contact time of 2 h. Under these conditions, the adsorption recovery of rare earths, thorium, and uranium into the solid phase was found to be 20.43 wt%, 99.20 wt%, and 89.60 wt%, respectively. The Freundlich adsorption isotherm was determined to be the best-fit model, and the adsorption mechanism of rare earths and thorium was identified as multilayer physisorption. Further, the separation efficiency was assessed using the response surface methodology based on the development of a statistically significant model.

58 GEOSCIENCES↗
DOE OSTI2013

Start-up fuel and power flattening of sodium-cooled candle core

The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.

The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluate the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Synthesis and Characterization of Th2N2(NH) Isomorphous to Th2N3

Using a new, low-temperature, fluoride-based process, thorium nitride imide of the chemical formula Th{sub 2}N{sub 2}(NH) was synthesized from thorium dioxide via an ammonium thorium fluoride intermediate. The resulting product phase was characterized by powder X-ray diffraction (XRD) analysis and was found to be crystallographically similar to Th{sub 2}N{sub 3}. Its unit cell was hexagonal with a space group of P3m{bar 1} and lattice parameters of a = b = 3.886(1) and c = 6.185(2) {angstrom}. The presence of -NH in the nitride phase was verified by Fourier transform infrared spectroscopy (FTIR). Total energy calculations performed using all-electron scalar relativistic density functional theory (DFT) showed that the hydrogen atom in the Th{sub 2}N{sub 2}(NH) prefers to bond with nitrogen atoms occupying 1a Wyckoff positions of the unit cell. Lattice fringe disruptions observed in nanoparticle areas of the nitride species by high-resolution transmission electron microscopic (HRTEM) images also displayed some evidence for the presence of -NH group. As ThO{sub 2} was identified as an impurity, possible reaction mechanisms involving its formation are discussed.

08 HYDROGEN↗
DOE OSTI2012

The Indian Advanced Heavy Water Reactor (AHWR) and Non-Proliferation Attributes

This report describes the Indian Advanced Heavy Water Reactor (AHWR), focusing on the reactor design and operational features that are relevant to proliferation risk, for both the AHWR intended for domestic use and the AHWR-LEU for export. The AHWR is the key component of the final stage of the three-stage Indian Nuclear program, “Thorium Utilization for Sustainable Power Programme,” that has been in development since the 1950’s. The nuclear program in India has historically assumed a heavy reliance on the ultimate transition to a thorium-based fuel cycle due to the abundant domestic thorium resources and the scarcity of domestic uranium. Both the AHWR and AHWR-LEU use thorium-based oxide fuels, with the AHWR using both UO 2 /ThO 2 and PuO 2 /ThO 2 fuels simultaneously, while AHWR-LEU uses only UO 2 /ThO 2 fuel. The AHWRs are essentially derivatives of the advanced CANDU reactor, with the fuel oriented vertically instead of horizontally and refueling still occurring on-line, but only from the top of the reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 6 QUALIFICATION

Prior to initiating a new sludge batch in the Defense Waste Processing Facility (DWPF), Savannah River National Laboratory (SRNL) is required to simulate this processing, including Chemical Process Cell (CPC) simulation, waste glass fabrication, and chemical durability testing. This report documents this simulation for the next sludge batch, Sludge Batch 6 (SB6). SB6 consists of Tank 12 material that has been transferred to Tank 51 and subjected to Low Temperature Aluminum Dissolution (LTAD), Tank 4 sludge, and H-Canyon Pu solutions. Following LTAD and the Tank 4 addition, Liquid Waste Operations (LWO) provided SRNL a 3 L sample of Tank 51 sludge for SB6 qualification. Pu solution from H Canyon was also received. SB6 qualification included washing the sample per LWO plans/projections (including the addition of Pu from H Canyon), DWPF CPC simulations, waste glass fabrication (vitrification), and waste glass characterization and chemical durability evaluation. The following are significant observations from this demonstration. Sludge settling improved slightly as the sludge was washed. SRNL recommended (and the Tank Farm implemented) one less wash based on evaluations of Tank 40 heel projections and projections of the glass composition following transfer of Tank 51 to Tank 40. Thorium was detected in significant quantities (>0.1 wt % of total solids) in the sludge. In past sludge batches, thorium has been determined by Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS), seen in small quantities, and reported with the radionuclides. As a result of the high thorium, SRNL-AD has added thorium to their suite of Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES) elements. The acid stoichiometry for the DWPF Sludge Receipt and Adjustment Tank (SRAT) processing of 115%, or 1.3 mol acid per liter of SRAT receipt slurry, was adequate to accomplish some of the goals of SRAT processing: nitrite was destroyed to below 1,000 mg/kg and mercury was removed to below the DWPF target with 750 g of steam per g of mercury. However, rheological properties did not improve and were above the design basis. Hydrogen generation rates did not exceed DWPF limits during the SRAT and Slurry Mix Evaporator (SME) cycles. However, hydrogen generation during the SRAT cycle approached the DWPF limit. The glass fabricated with the Tank 51 SB6 SME product and Frit 418 was acceptable with respect to chemical durability as measured by the Product Consistency Test (PCT). The PCT response was also predictable by the current durability models of the DWPF Product Composition Control System (PCCS). It should be noted, however, that in the first attempt to make glass from the SME product, the contents of the fabrication crucible foamed over. This may be a result of the SME product's REDOX (Reduction/Oxidation - Fe{sup 2+}/{Sigma}Fe) of 0.08 (calculated from SME product analytical results). The following are recommendations drawn from this demonstration. In this demonstration, at the request of DWPF, SRNL caustic boiled the SRAT contents prior to acid addition to remove water (to increase solids concentration). During the nearly five hours of caustic boiling, 700 ppm of antifoam was required to control foaming. SRNL recommends that DWPF not caustic boil/concentrate SRAT receipt prior to acid addition until further studies can be performed to provide a better foaming control strategy or a new antifoam is developed for caustic boiling. Based on this set of runs and a recently completed demonstration with the SB6 Waste Acceptance Product Specifications (WAPS) sample, it is recommended that DWPF not add formic acid at the design addition rate of two gallons per minute for this sludge batch. A longer acid addition time appears to be helpful in allowing slower reaction of formic acid with the sludge and possibly decreases the chance of a foam over during acid addition.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2001

Monthly Variability in Upper Ocean Biogeochemistry Due to Mesoscale Eddy Activity in the Sargasso Sea

A comparison of monthly biogeochemical measurements made from 1993 to 1995, combined with hydrography and satellite altimetry, was used to observe the impacts of nine eddy events on primary productivity and particle flux in the Sargasso Sea. Measurements of primary production, thorium-234 flux, nitrate+nitrite, and photosynthetic pigments made at the US JGOFS Bermuda Atlantic Time-Series Study (BATS) site were used. During the three years of this study, four out of six high thorium-234 flux events over 1000 dpm/sq m/d occurred during the passage of an eddy. Primary production nearly as high as the spring bloom maximum was observed in two mode-water eddies (May 1993 and July 1995). The 1994 spring bloom at BATS was suppressed by the passage of an anticyclone. Distinct phytoplankton community shifts were observed in mode-water eddies, which had an increased percentage diatoms and dinoflagelletes, and in cyclones, which had an increased percentage cyanobacteria (excluding Prochlorococcus). The difference in the observations of mode-water eddies and cyclones may result from the age of the eddy, which was very important to the biological response. In general, eddies that were one to two months old elicited a large biological response; eddies that were three months old may show a biological response and were accompanied by high thorium flux measurements; eddies that were four months old or older did not show a biological response or high thorium flux. Our conceptual model depicting the importance of temporal changes during eddy upwelling and decay fit the observations well in all seven upwelling eddies. Additional information is needed to determine the importance of deeper mixed layers and winter mixing to the magnitude of the eddy impacts. Also, sampling generally captured only the beginning, end, and/or edge of an eddy due to the monthly to semi-monthly frequency of the measurements made at BATS. Lagrangian studies, higher resolution time-series, and/or more spatial coverage is needed to provide additional information for improved C and N budgets in the Sargasso Sea and to complete our understanding of the temporal changes that occur in an eddy.

Sweeney, Erin N.↗
NASA NTRS2001

Monthly Variability in Upper Ocean Biogeochemistry due to Mesoscale Eddy Activity in the Sargasso Sea

A comparison of monthly biogeochemical measurements made from 1993 to 1995, combined with hydrography and satellite altimetry, was used to observe the impacts of nine eddy events on primary productivity and particle flux in the Sargasso Sea. Measurements of primary production, thorium-234 flux, nitrate+nitrite, and photosynthetic pigments made at the US JGOFS Bermuda Atlantic Time-series Study (BATS) site were used. During the three years of this study, four out of six high thorium-234 flux events over 1000 dpm/sq m/d occurred during the passage of an eddy. Primary production nearly as high as the spring bloom maximum was observed in two modewater eddies (May 1993 and July 1995). The 1994 spring bloom at BATS was suppressed by the passage of an anticyclone. Distinct phytoplankton community shifts were observed in mode-water eddies, which had an increased percentage diatoms and dinoflagelletes, and in cyclones, which had an increased percentage cyanobacteria (excluding Prochlorococcus). The difference in the observations of mode-water eddies and cyclones may result from the age of the eddy, which was very important to the biological response. In general, eddies that were one to two months old elicited a large biological response; eddies that were three months old may show a biological response and were accompanied by high thorium flux measurements, eddies that were four months old or older did not show a biological response or high thorium flux. Our conceptual model depicting the importance of temporal changes during eddy upwelling and decay fit the observations well in all 7 upwelling eddies. Additional information is needed to determine the importance of deeper mixed layers and winter mixing to the magnitude of the eddy impacts. Also, sampling generally captured only the beginning, end, and /or edge of an eddy due to the monthly to semi-monthly frequency of the measurements made at BATS. Lagrangian studies, higher resolution time-series, and/or more spatial coverage is needed to provide additional information for improved C and N budgets in the Sargasso Sea and to complete our understanding of the temporal changes that occur in an eddy.

Sweeney, Erin N.↗
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↗
DOE OSTI2019

The application of poorly crystalline silicotitanate in production of 225 Ac

Actinium-225 ( 225 Ac) can be produced from a Thorium-229/Radium-225 ( 229 Th/ 225 Ra) generator, from high/low energy proton irradiated natural Thorium or Radium-226 target. Titanium based ion exchanger were evaluated for purification of 225 Ac. Poorly crystalline silicotitanate (PCST) ion exchanger had high selectivity for Ba, Ag and Th. 225 Ac was received with trace amounts of 227 Ac, 227 Th and 223 Ra, and the solution was used to evaluate the retention of the isotopes on PCST ion exchanger. Over 90% of the 225 Ac was recovered from PCST, and the radiopurity was >99% (calculated based on 225 Ac, 227 Th, and 223 Ra). The capacity of the PCST inorganic ion exchange for Barium and 232 Th was determined to be 24.19 mg/mL for Barium and 5.05 mg/mL for Thorium. PCST ion exchanger could separate 225 Ac from isotopes of Ra and Th, and the process represents an interesting one step separation that could be used in an 225 Ac generator from 225 Ra and/or 229 Th. Capacity studies indicated PCST could be used to separate 225 Ac produced on small 226 Ra targets (0.3–1 g), but PCST did not have a high enough capacity for production scale Th targets (50–100 g).

43 PARTICLE ACCELERATORS↗
DOE OSTI2011

Fusion-fission hybrid for fissile fuel production without processing

Two scenarios are typically envisioned for thorium fuel cycles: “open” cycles based on irradiation of 232 Th and fission of 233 U in situ without reprocessing or “closed” cycles based on irradiation of 232 Th followed by reprocessing, and recycling of 233 U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in the critical reactors. This combination consumes about 20% of the thorium initially loaded in the hybrid reactor (~200 GWd/tHM), partially during hybrid operation, but mostly during operation in the critical reactor. The plant support ratio is low compared to the one attainable using continuous fuel chemical reprocessing, which can yield a plant support ratio of about 20, but the resulting fuel cycle offers better proliferation resistance as fissile material is never separated from the other fuel components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Elucidating Actinide–Pertechnetate and Actinide–Perrhenate Bonding via a Family of Th–TcO 4 and Th–ReO 4 Frameworks and Solutions

Technetium-99, a β-emitter produced from 235 U fission, poses a challenge for the nuclear industry due to co-extraction of pertechnetate (TcO 4 – ) with the actinides (An) during nuclear fuel reprocessing. Previous studies suggested that direct coordination of pertechnetate with An plays an important role in the coextraction process. However, few studies have provided direct evidence for An–TcO 4 – bonding in the solid state, and even fewer in solution. The present study describes synthesis and structural elucidation of a family of thorium(IV)-pertechnetate/perrhenate (ReO 4 – , nonradioactive surrogate) compounds, which is obtained by dissolution of thorium oxyhydroxide in perrhenic/pertechnic acid followed by crystallization, with or without heating. For reaction ratios of 3:1, 4:1, and 6:1 MO 4 – /Th(IV) (M = Tc, Re), the crystallized compounds reflect the same ratio, suggesting facile and flexible coordination. Furthermore, nine structures reveal 1-dimensional and 2-dimensional frameworks with varying topologies. While a multitude of compounds isolated from 4:1 (and 6:1) reaction solutions feature Th monomers linked by MO 4 – , the 3:1 reaction solution yielded the well-known dihydroxide-bridged thorium dimer, linked, and capped by MO 4 – . Density functional theory calculations on ReO 4 – /TcO 4 – isomorphs suggest similar bonding characteristics in the solid state, but experimental solution characterization noted differences. Specifically, small-angle X-ray scattering studies suggest the bonding of Th–TcO 4 – persists in solution, while Th–ReO 4 – bonding is less apparent.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2011

Exploration Of Activity Measurements And Equilibrium Checks For Sediment Dating Using Thick-Window Germanium Detectors

Activity measurements on sediment samples for trapped-charge geological dating using gamma-ray spectroscopy are an important verification of the field-site dose rate determination. Furthermore gamma-ray spectroscopy can check if the natural decay series are in secular equilibrium which is a crucial assumption in such dating. Typically the activities of leading members of the Thorium and Uranium decay series are measured, which requires Germanium detectors with thin windows and good energy resolution in order to effectively detect the associated low energy gamma-rays. Such equipment is not always readily available. The potential of conventional Germanium detectors with thick entrance window has been explored towards routine gamma-ray spectroscopy of sediment samples using higher energy gamma-rays. Alternative isotopes, such as Ac-228 and Pb-212 for the Thorium series, and Pa-234m, Ra-226 and Bi-214 for the Uranium series, have been measured in order to determine the mass-specific activity for the respective series and possibly provide a check of secular equilibrium. In addition to measurements of the K-40 activity, with the alternative approach, the activities of both decay series can be accurately determined. The secular equilibrium condition may be tested for the Thorium series. Measurement accuracy for Pa-234m is, however, not sufficient to permit also a reliable check of equilibrium for the Uranium series.

07 ISOTOPE AND RADIATION SOURCES↗