Pure Thorium Research Articles

Model Comparison of Passive Compact-Molten Salt Reactor Neutronic Design Using MCNP6 and Serpent-2

Passive Compact Molten Salt Reactor (PCMSR) is a thermal breeder molten salt reactor (MSR) developed in Universitas Gadjah Mada, Indonesia, run in thorium fuel cycle. Its design was initially developed using deterministic code SRAC2006 but has never been compared with other codes. This paper attempts to compare PCMSR neutronic design using Monte Carlo codes MCNP6 and Serpent-2 with ENDF B/VII.0 continuous neutron cross-section library. The reactor was run in a pure thorium fuel cycle with lithium fluoride as its carrier salt. The analyzed parameters were effective multiplication factor (k eff ), temperature coefficient of reactivity (TCR), void coefficient of reactivity (VCR), and conversion ratio (CR). The result shows that there are several important discrepancies between the original calculation and this research. The Monte Carlo calculations implied that PCMSR core was able to be critical using lower fissile concentration than previously designed, but failed to reach CR above unity. While the TCR value was found to be negative, the VCR value was positive up until the 10 % void fraction. The PCMSR core suffered from ineffective neutron moderation and high neutron leakage. These findings imply that the previous PCMSR neutronic design is inaccurate. For PCMSR to be able to operate as a thermal breeder MSR, geometrical modifications must be performed to improve neutron moderation and reduce neutron leakage.

Sequential separation and selective extraction of uranium and thorium from monazite sulfate leach liquor using dipropylamine extractant

Monazite mineral is the main source of uranium and thorium, which are mainly associated with rare earth elements (REEs). So, their separation is the crucial task to satisfy their application for nuclear fuel. Dipropylamine was used for the first time in the selective extraction and separation of uranium and thorium from Monazite sulfate leach liquor. The extraction of thorium (98%) was achieved using dipropylamine in kerosene at pH 1; it was followed by selective extraction of uranium (97%) by Dipropylamine at pH 3. The thermodynamic analysis demonstrated that uranium (VI) and thorium (IV) extraction by dipropylamine is an exothermic process. From the stripped liquor (1 M H2SO4), Th and U were precipitated as oxalate and peroxide, respectively, for technological application. Thus, marketable pure thorium and uranium products were prepared in addition to rare earth concentrate.

Commercial approach for highly pure thorium from Egyptian monazite mineral acid process

Thorium has a special interest in the future nuclear energy as a fuel especially the breeder reactors. The present research has been fulfilled throughout successive research schedule manner on separation and purification of thorium from Egyptian beach monazite mineral after its acid digestion. Monazite mineral was digested with sulfuric acid 96% at 220°C for 2.5h and dissolute with ice water (1:15), the clear solution precipitated with ammonia solution 16.5% at pH 1to produce thorium pyrophosphate with 45% thorium content in case of ideal optimized whole process. The purification was carried out after the caustafication and leaching with hydrochloric acid followed by the addition of oxalic acid at pH 0.3-0.4 to produce thorium oxalate with 91.68 % purity. Thorium oxalate was transformed to oxide by caustafication then washed and separated and dissolute by hydrochloric acid. Highly pure thorium salt was attained with the addition of thiosulphate salt on the produced faint chloride solution under 0.1 to maximum 0.5 phosphate content to produce 99% purity.

Radioactivity and mineralogical studies with an emphasis on recovery of thorium from mylonite rocks in the Abu Rusheid area, South Eastern Desert, Egypt

The Abu Rusheid area, part of the Arabian-Nubian Shield, lies about 90 km southwest of Marsa Alam City. The cataclastic rocks in this area include protomylonites, mylonites, ultramylonites, and silicified ultramylonites. Mylonite rocks make up about 65.0% by volume of the cataclastic rocks, and many of them have high concentrations of visible orange-colored thorium and yellow-colored uranium minerals. XRD and SEM analyses revealed the presence of thorite, kasolite, monazite, zircon, xenotime, hematite, and ilmenite. The mylonite rock studied here contained high concentrations of thorium: > 1% in some cases, and an average of 3962 ppm. A sample assaying 10,000 mg/L was leached at 50 °C for 1.5 h by 6 M HCl using a S/L ratio of 1/6, which dissolved about 95% of the thorium. The thorium was precipitated from the leach liquor by oxalic acid. Highly pure thorium oxide was then prepared and identified via qualitative (SEM/EDX) and quantitative (ICP-OES) analyses.

INVESTIGATION ON INHERENT SAFETY OF ONE FLUID-MOLTEN SALT REACTOR (OF-MSR) WITH VARIOUS STARTING FUEL

Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

Production of High Purity Thorium Oxide from Complex Ores Leach Liquor

This paper investigates a method for separation and purification of thorium from leach liquors containing rare-earth elements (REE). Iron, which causes problems in the stage of thorium extracting, was first removed by using Adogen 464 solvent. Thorium was extracted with 30% D2EHPA-in kerosene with 1/3 organic/aqueous ratio and stoichiometry of the thorium-D2EHPA complex was calculated as 1:2.6 from slope analysis. After the extraction, LREE and HREE were scrubbed from the extracted organic with 0.25 M H₂SO₄ and 6 M HCl, respectively. Thorium oxalate was precipitated by oxalic acid and calcined at 1,050 °C. It was, after calcination, determined that purity of thorium oxide is 99.23%, based on the ICP-OES analysis.

Separation, purification and recovery of thorium from monazite leach liquors by counter-current extraction process

Separation and purification of thorium from monazite leach liquors was investigated by counter current extraction process. The Primene JM-T (mixture of highly branched C16 to C22 tertiary alkyl primary amine isomers) was used as an extractant and kerosene as a diluent for the present investigation. Extraction isotherm was plotted at the condition: 0.1 mol/L Primene JM-T, A:O = 3:1, which predicted on requirement of 3 stages for quantitative extraction of thorium. A 99.9% extraction of thorium from the leach liquor was achieved at the optimum conditions leaving 0.1 mg/L of the metal in the raffinate phase. The predicted three counter current stripping stages from the stripping were validated using 0.5 mol/L HCl + 0.5 mol/L NH4Cl as the stripping agent. XRD analysis proved the high purity thorium oxalate precipitated and recovered from monazite.

Extraction of Thorium(IV) with N-Methyl-N,N,N-trioctylammonium Chloride from Monazite Acidic Leach Liquor and Its Use for Spectrophotometric Determination

N-Methyl-N,N,N-trioctylammonium chloride was applied for selective extraction and separation of thorium. The Th extraction and stripping conditions were experimentally optimized. 0.75 M N-methyl-N,N,N-trioctylammonium chloride was found to be suitable for the Th separation. The extraction of Th(IV) form monazite is an exothermic process. The optimum parameters were applied to the spectrophotometric determination of Th using Thoron I at λmax = 540 nm with accurate results. From the stripped liquor (1 M HNO3), Th was precipitated as thorium oxalate for technological application. Thus, marketable pure thorium product was prepared in addition to uranium and rare earth concentrates.

Feasibility study of thorium-fueled molten salt reactor with application in radioisotope production

This paper explored the theoretical possibility of producing radioisotopes, especially Molybdenum-99 (Mo-99), in a reactor fueled with pure natural thorium. The challenge of reaching criticality with pure thorium fuel was discussed, and possible solutions were presented. Specifically, we proposed a solution involving high-intensity neutron source with varying strength over time. The numerical analyses showed that, under reasonable simplifications and assumptions, criticality could be achieved in an infinite medium reactor with the proposed neutron source. Additionally, a Mo-99 production scheme was presented, and an estimate on annual production was given.

Multiparametric Study of the Synthesis of ThSiO4 under Hydrothermal Conditions.

A multiparametric study of the hydrothermal synthesis of ThSiO4, thorite, was performed with the aim of determining the most efficient conditions to form single-phase samples. Among the experimental parameters examined, significant effects were found for the concentration of reactants in the starting mixture, pH of the reactive media, and temperature of the hydrothermal process. Such parameters affected both the rate of formation of thorite and the morphology of the final products synthesized. Precipitation of pure ThSiO4 was obtained over a wide range of pH on going from CHNO3 = 0.3 mol L-1 to pH 9.1 with a yield of over 95%. Temperatures higher than 160 °C favor the formation of thorite. Finally, thorium and silicon concentrations above 2.1 × 10-3 mol L-1 are required to obtain pure thorium silicate.

PERFORMANCE IMPROVEMENTS FOR THORIUM-BASED FUELS IN PRESSURE-TUBE HEAVY-WATER REACTORS

Lattice physics sensitivity studies have been performed with WIMS-AECL to quantify the impact of various design and operating parameters on the performance characteristics of thorium-based fuel concepts in pressure-tube heavy-water reactors. Fuels modeled included 37-element bundles with natural uranium oxide (for comparison), pure thorium oxide (blanket-type fuel) and 35-element bundles of mixed oxide with thorium and U-233. Key performance parameters evaluated included the lattice reactivity, exit burnup, coolant void reactivity (CVR), and fissile concentration. The effects of various design/operational parameters were evaluated, including calandria tube radius, moderator purity, coolant purity, zirconium enrichment, and temporary out-of-core fuel storage at zero power. Results demonstrated that removing the moderator around the blanket fuel can harden the neutron energy spectrum and increase the discharge fissile content from ∼1 wt% U-233 to ∼2 wt% U-233 at a low discharge burnup (5 MWd/kg). Increasing...

Theoretical analysis of uranium-doped thorium dioxide: Introduction of a thoria force field with explicit polarization

Thorium dioxide is used industrially in high temperature applications, but more insight is needed into the behavior of the material as part of a mixed-oxide (MOX) nuclear fuel, incorporating uranium. We have developed a new interatomic potential model including polarizability via a shell model, and commensurate with a prominent existing UO2 potential, to conduct configurational analyses and to investigate the thermophysical properties of uranium-doped ThO2. Using the GULP and Site Occupancy Disorder (SOD) computational codes, we have analyzed the distribution of low concentrations of uranium in the bulk material, where we have not observed the formation of uranium clusters or the dominance of a single preferred configuration. We have calculated thermophysical properties of pure thorium dioxide and Th(1−x)UxO2 which generated values in very good agreement with experimental data.

Formation of Thorium-disodium Fluoride Crystals by Electron-beam Evaporation

The technique of thin film forming of NaThxFy compounds on the Si (111) surface by electron-beam evaporation was developed. It is shown that the band gap of pure thorium fluoride is 6.7eV; of sodium fluoride – 9.5eV, while in compounds of thorium-sodium fluoride NaTh1.3F6.2 it takes the value 7.1eV. It was found that prolonged electron beam exposure leads to the recovery of thorium fluoride followed by oxidation of thorium atoms.

Preparation of high-purity thorium by solvent extraction with di-(2-ethylhexyl) 2-ethylhexyl phosphonate

The extraction behavior of thorium(IV) by di-(2-ethylhexyl) 2-ethylhexyl phosphonate (DEHEHP, B) from nitric acid media has been investigated. The influence factors including the concentration of HNO3, salting-out reagents, temperature, the concentration of metal ions and DEHEHP have been examined systematically. A possible extraction mechanism is proposed and the extracted species as Th(NO3)4·2B (o) is confirmed by the slope analysis method. The extraction equilibrium constants (K ex) and thermodynamic parameters (ΔG, ΔH and ΔS) were calculated under the present experimental conditions. DEHEHP shows a high selectivity of thorium(IV) over rare earths(III). Stripping study indicates that thorium can be completely stripped by distilled water from the Th-loaded DEHEHP. Furthermore, a solvent extraction process including six extraction stages, six scrubbing stages, and six stripping stages was designed for the preparation of highly pure thorium from thorium concentrate with DEHEHP as extractant in laboratory scale, and finally thorium product can be obtained with a purity of 99.999 % and a yield of 98 %.

Measurements and model calculations of activation cross sections for 232Th( n,2 n) 231Th reaction between 13.57 and 14.83 MeV neutrons

In this study, activation cross sections were measured for the reaction of 232Th( n,2 n) 231Th ( T 1/2 = 25.5 h) by using neutron activation technique at six different neutron energies from 13.57 and 14.83 MeV. Neutrons were produced via the 3H( 2H, n) 4He reaction using SAMES T-400 neutron generator. Irradiated and activated high purity Thorium foils were measured by a high-resolution γ-ray spectrometer with a high-purity Germanium (HpGe) detector. In cross section measurements, the corrections were made for the effects of γ-ray self-absorption in the foils, dead-time, coincidence summing, fluctuation of neutron flux, low energy neutrons. For this reaction, statistical model calculation, which the pre-equilibrium emission effects were taken into consideration, were also performed between 13.57 and 14.83 MeV energy range. The cross sections were compared with previous works in literature, with model calculation results, and with evaluation data bases (ENDF/B-VII, ENDF/B-VI, JEFF-3.1, JENDL-4.0, JENDL-3.3, and ROSFOND-2010).

Physical and Chemical Beneficiation of the Egyptian Beach Monazite

AbstractThe Egyptian black sands contain several economic minerals, especially ilmenite. Monazite can reach concentrations up to 0.6 wt.%. The majority of monazite grains have light to deep canary and lemon yellow colors, whereas enigmatic monazite grains have brown, red, resinous, yellow and colorless to pinkish colors. The behavior of monazite with the electrostatic field setting of the roll‐type electrostatic separator was studied for the different variables of the separator. Most of the monazite grains are reversible negative and are attracted towards the positively charged static electrode. Using wet gravity concentration, both low and high intensity magnetic separation, and electrostatic separation techniques, a high grade concentrate assaying 97 wt. % monazite with a recovery of 76.8% can be obtained.The Egyptian beach monazite contains high concentrations of Ce, La and Nd in addition to minor amounts of Y, Pr, Sm, Gd, Dy and Eu. It is particularly rich in the lighter rare earth elements (cerium subgroup). By a new method of ion exchange technique after chemically dissolving the mineral with sulfuric acid, both of REEs, Th and U, could be individually separated. The efficiency of the separation has been tested with sulfuric acid concentration at 2M H2SO4 where the thorium and uranium could be obtained with a cation exchange synthetic resin column, while rare earth metal ions are adsorbed and then individually separated. A highly pure thorium product could be obtained by oxalate precipitation followed by uranium precipitation as diuranate using NaOH.

Monte Carlo Study of Quantitative Electron Probe Microanalysis of Monazite with a Coating Film: Comparison of 25 nm Carbon and 10 nm Gold at E0= 15 and 25 keV

Carbon (25–30 nm in thickness) is the most common coating material used in the electron probe microanalysis (EPMA) of geological samples. A gold coating is also used in special cases to reduce the surface damage by electron bombardment. Monte Carlo simulations have been performed for monazite with a 25 nm carbon and a 10 nm gold coating to understand the effect of a coating film in quantitative EPMA at E0= 15 keV and 25 keV. Simulations showed that carbon‐coated monazite gave the same depth distribution of the generated X‐rays in the monazite as uncoated monazite, whilst gold‐coated monazite gave a distorted depth distribution. A 10 nm gold coating was 1.06 (15 keV) and 1.05 (25 keV) times higher in k‐ratio between monazite and pure thorium than a 25 nm carbon coating at an X‐ray take‐off angle of 40 degrees. Thus, a 10 nm gold coating is a possible factor contributing to inaccuracy in quantitative EPMA of monazite, while a 25 nm carbon coating does not have a significant effect.

Synthesis and characterization of low-temperature precursors of thorium–uranium (IV) phosphate–diphosphate solid solutions

Several compositions of new precursor of thorium–uranium (IV) phosphate–diphosphate solid solutions (Th4−xUx(PO4)4P2O7, called β-TUPD) were synthesized in closed PTFE containers either in autoclave (160°C) or on sand bath (90–160°C). All the samples appeared to be single phase. From XRD data and TEM observations, the diffraction lines matched well with that of pure thorium phosphate–hydrogenphosphate hydrate (TPHPH), Th2(PO4)2(HPO4)·H2O, which confirmed the preparation of a complete solid solution between pure thorium and uranium (IV) compounds. TGA/DTA experiments showed that samples of thorium–uranium (IV) phosphate–hydrogenphosphate hydrate (TUPHPH) prepared at 150–160°C were monohydrated leading to the proposed formula Th2−x/2Ux/2(PO4)2(HPO4)·H2O. The variation of the XRD diagrams versus the heating temperature showed that TUPHPH remained crystallized and single phase from room temperature to 200°C. After heating between 200°C and 800°C, the presence of diphosphate groups in the solid was evidenced. In this range of temperature, the solid was transformed into the low-temperature monoclinic form of thorium–uranium (IV) phosphate–diphosphate (α-TUPD). This latter compound finally turned into well-crystallized, homogeneous and single-phase β-TUPD (orthorhombic form) above 930–950°C for x values lower than 2.80. For higher x values, a mixture of β-TUPD, α-Th1−zUzP2O7 and U2−wThwO(PO4)2 was obtained. By this new chemical route of preparation of β-TUPD solid solutions, the homogeneity of the samples is significantly improved, especially considering the distribution of thorium and uranium.

Preparation of highly pure thorium nitrate via thorium sulfate and thorium peroxide

A simple alternative and selective method has been proposed for the preparation of thorium nitrate using as raw material a hydrated crystalline technical grade thorium sulfate. This salt was produced industrially [1] from the alkaline chemical processing of monazite sand in S. Paulo. From a crude thorium hydroxide as the raw material containing uranium and rare earth elements was prepared a technical grade thorium sulfate [Th(SO 4) 2.9H 2O]. A typical analysis of this salt had shown the following composition: ThO 2 47–53%, R 2O 3 (rare earth oxides) 0.5–0.9%, and UO 3 0.01% and minor amounts of iron, titanium, phosphorus, and silicon oxides. The thorium sulfate is dissolved into water and the thorium peroxide is precipitated from acidic medium by addition of hydrogen peroxide. The thorium peroxide is filtered, washed and finally dissolved with nitric acid. Samples of the thorium peroxide were dried and fired to ThO 2 and aliquot of the thorium nitrate solution was evaporated to dryness and calcined to obtain the ThO 2. Both ThO 2 have high purity (>99.5%). The content of total rare earth elements is very low (0.23% as oxide). Uranium was not detected. The great majority of the RE is precipitated and recovered from the filtrate by oxalic acid.

Kinetic and thermodynamic studies of the dissolution of thorium–uranium (IV) phosphate–diphosphate solid solutions

The dissolution of thorium–uranium (IV) phosphate–diphosphate solid solutions (TUPD) was studied as a function of the temperature and leachate acidity. The dependence of the normalized dissolution rate on the temperature leads to an activation energy equal to about 40 kJ mol −1, close to that obtained for the pure thorium phosphate–diphosphate ( 42±3 kJ mol −1 ) and for thorium–plutonium (IV) phosphate–diphosphate solid solutions ( 41±1 kJ mol −1 ). The normalized dissolution rate of TUPD slightly increases with the leachate acidity. The partial order related to the proton concentration, n, is equal to 0.40±0.02 while the apparent normalized dissolution rate constant, k ′ T,I, reaches (2.8±0.7)×10 −4 g m −2 d −1 at 90°C and for [ H 3 O +]=1 M . When the saturation of the leachate is reached, the concentration of thorium, uranium and phosphate ions measured in the solution are controlled by the precipitation of the uranyl phosphate pentahydrate (UO 2) 3(PO 4) 2·5H 2O and the thorium phosphate–hydrogenphosphate Th 2(PO 4) 2(HPO 4)·H 2O. Both solids were extensively characterized using XRD, infrared and UV–visible spectroscopies or electron probe microanalysis (EPMA). Their solubility products, K ° S,0, were determined and extrapolated to I=0. They are equal to 10 −55.2±0.5 and 10 −66.6±1.2, respectively. All the samples leached were characterized using EPMA, SEM and TEM. These techniques showed that during the dissolution process, thorium and uranium are completely separated as (UO 2) 3(PO 4) 2·5H 2O, on one hand, and Th 2(PO 4) 2(HPO 4)·H 2O, on the other hand. In the first days of leaching tests, an amorphous additional phase, identified as Th 2(PO 4) 2(HPO 4)· nH 2O was also observed. Several leaching tests performed on sintered TUPD samples revealed that the dissolution rates measured in 10 −1 M HNO 3 is very low (6.5×10 −5 g d −1) by comparison to other ceramics studied in the same objective. In these conditions, the thorium phosphate–diphosphate (TPD) appears as a promising ceramic for the immobilization of tetravalent actinides like uranium, neptunium or plutonium.