Thorex Process Research Articles

Thorium as an abundant source of nuclear energy and challenges in separation science

Abstract Today about 440 nuclear power plants, with total installed capacity of 390 GW(e) are in operation worldwide generating around 10% of global electricity which are largely fuelled by enriched uranium oxide [Nuclear Power in the World Today. https://world-nuclear.org/information-library/current-and-future-generation/nuclear-power-in-the-world-today.aspx]. Thorium is 3–4 times more abundant than uranium and needs to be exploited by countries with limited stock of uranium. Historically, there have been several attempts to develop Th based reactors, but none has reached commercial scale. In recent years, High Temperature Reactor based on thorium has gained prominence for production of hydrogen with long term goal of complete carbon neutrality. However, unlike natural uranium, which contains ∼0.7% fissile 235U isotope, natural thorium does not contain any ‘fissile’ material and is made up exclusively of the ‘fertile’ 232Th which can be converted to ‘fissile’ 233U, thereby enlarging the fissile material resources. However, there is a need to develop robust closed fuel cycle to address to the challenges of high gamma dose due to the presence of decay products of 232U. It is necessary to gain more experience with promising THOREX process to achieve the desired recovery and D.F. of 233U from the irradiated 232Th. There is also scope to have a close look at some of the alternative extractants and emerging separation techniques for the reprocessing of spent Th based fuels. In view of the distinct advantages of non aqueous reprocessing over aqueous reprocessing, there is need to intensify the efforts to develop the former on a commercial scale. Molten Salt Breeder Reactor (MSBR) is particularly promising in this context. There is a need to investigate Th as energy amplifier under Accelerator Driven Sub-critical System (ADSS) which has potential to burn long lived radio nuclides, considered as a threat to environment over million of years.

Ultra-safe nuclear thermal rockets using lunar-derived fuel

Rocket launch failure rate is slightly higher than five percent. Concerned citizens are likely to protest against private-sector launches involving fission reactors. Yet, fission reactors can power long-duration lunar operations for science, observation, and in situ resource utilization. Furthermore, fission reactors are needed for rapid transport around the solar system, especially considering natural radiation doses for crews visiting Mars or an asteroid. A novel approach is to create nuclear fuel on the Moon. In this way, a rocket launched from the earth with no radioactive material can be fueled in outer space, avoiding the risks of spreading uranium across Earth's biosphere. A solution is to harvest fertile thorium on the lunar surface, then transmute it into fissile uranium using the gamma ray fog which pervades the deep sky. It is only at lunar orbit, at the very edge of cislunar space, that the Earth-launched machine becomes a nuclear thermal rocket (NTR). Thorium is not abundant, but can be concentrated by mechanical methods because of its very high specific density relative to the bulk of lunar regolith. Thorium dioxide (ThO2) has an extremely high melting point, such that skull crucible heating can be used to separate it from supernatant magma. When filled into a graphite-lined beryllium container (brought from Earth) and set out on the lunar surface, high-energy gamma rays will liberate neutrons from the Be. After moderation by the graphite, these thermal neutrons are captured by the thorium nucleus, which is transmuted into protactinium (Pa91). This element can be extracted using the THOREX process, and will then decay naturally into U-233 within two or three lunar days. The uranium is oxidized and packed into fuel pellets, ready to be inserted into a non-radioactive machine, which now becomes an NTR. Additionally, hydrogen can be extracted from deposits in permanently-shadowed regions on the Moon, providing reaction mass for the NTR. A novel method of solid-state hydrogen storage, which can be entirely fabricated using in situ resources, can deliver said hydrogen to the fission reactor to provide high and efficient propulsive thrust. These combined operations lead to an ultra-safe (for the Earth) means for private sector, commercial transport and power generation throughout the Solar System. With the hydrogen storage material used as radiation shielding for crewed spacecraft, and greatly-reduced transit times relative to chemical rocketry, this innovative approach could fundamentally transform how humans work, play, and explore in outer space.

Studies on the Thoria Fuel Recycling Loop Using Triflic Acid: Effects of Powder Characteristics, Solution Acidity, and Radium Behavior

A convenient recycling loop is one of the key factors that will help to implement the thorium-based nuclear fuel cycle in the future. Herein, a study is presented concerning the recycling loop of thoria production scrap using dissolution with trifluoromethanesulfonic (triflic) acid and recovery by precipitation with oxalic acid. The kinetics of thoria powders and pellet dissolution was assessed and compared with the THOREX process. The effect of the triflic acid concentration on the recovery yield of thorium during oxalate precipitation was examined, as well as the behavior of radium during this reaction. Finally, the effect of the triflic acid concentration on the recovered thoria morphology was investigated. The triflic acid has minimal effect on the recovery yield during oxalate precipitation even in the case of high acid concentrations. In general, the solubility of radium(II) oxalate is higher than for thorium(IV) and has a maximum at a triflic acid solution of 4 M. An interesting observation was that a stable dihydrate of thorium(IV) oxalate is formed in solution of high triflic acid concentration. It might be explained by the hygroscopic medium during precipitation of the oxalate. All above mentioned observations confirm the benefits of usage of the triflic acid-based closed recycling loop for thorium dioxide-based nuclear fuel.

Enhancing decontamination of zirconium and ruthenium in the Thorex process using acetohydroxamic acid

The decontamination of some extractable fission products, such as Zr and Ru, is an important issue in the Thorex process. Acetohydroxamic acid (AHA) is an organic ligand composed of only C, H, N and O atoms. It can be used as complexing and reducing agent to inhibit the extraction of Zr and Ru. In this paper, the effect of AHA and preheating on the extraction of Th, U, Zr and Ru was studied. It is found the extraction of Zr and Ru would be inhibited in the presence of AHA and preheating, meanwhile there was no obvious effect on the extraction of Th and U. Afterwards, the countercurrent extraction using centrifugal extractors was performed to further test and verify the effect of AHA in a typical Thorex process. The results indicate that the decontamination of Zr and Ru could be enhanced evidently due to the effect of AHA, and the corresponding decontamination factor would be improved from 27.0 to 3.3 × 103 and 111.0 to 1.5 × 103, respectively. Thus, the use of AHA would be an efficient means for enhancing the decontamination of Zr and Ru in Thorex process.

Reprocessing Th-based spent fuels with di-1-methyl heptyl methyl phosphonate using centrifugal extractors

In this paper, a reprocessing Th-based spent fuels process using di-1-methyl heptyl methyl phosphonate (DMHMP) as extractant was proposed and tested by multistage countercurrent extraction using batch simulation and centrifugal extractors. The analysis results of process samples show that the recovery of Th and U in 1A (co-extraction of Th and U), 1B (separation of Th from U) and 1C (stripping of U from organic to aqueous phase) section was more than 99.8% and 99.9%, respectively. The separation factor of Th from U and U from Th in 1B section was 5.9×103 and 7.1×103, respectively. Compared with Thorex process using TBP as extractant, the process could be operated under lower acidity and flow ratio of O:A due to the strong extraction ability of DMHMP. Thus, DMHMP would be a promising alternative extractant for Th-based spent fuels process.

Di-1-methyl heptyl methylphosphonate (DMHMP): A promising extractant in Th-based fuel reprocessing

In this paper, the physical properties of tri-iso-amyl phosphate (TiAP), tri-sec-butyl phosphate (TsBP) and di-1-methyl heptyl methyl phosphonate (DMHMP) were studied and compared with tri-n-butyl-phosphate (TBP). Furthermore, the extraction behaviors of U(VI), Th(IV), nitric acid and fission products, and the extraction capacity of Th(IV) in different extractants were also investigated. The results indicated that the physical and extraction properties of DMHMP were better than that of other extractants and it is more appropriate to be used in thorium-based fuel reprocessing. A thorium-uranium co-decontamination process using DMHMP as extractant was proposed, optimized and verified by multistage countercurrent extraction using batch simulation and centrifugal extractors. By comparison with conventional Thorex process using TBP as extractant, it is found that DMHMP is more promising in thorium-based fuel reprocessing.

N,N-Dialkyl amides as extractants for spent fuel reprocessing: an overview

Reprocessing of spent nuclear fuel is vital for the long-term global nuclear power growth and is the major motivation for developing novel separation schemes. Conventionally, PUREX and THOREX processes have been proposed for the reprocessing of U and Th based spent fuels employing tri-n-butyl phosphate (TBP) as extractant. However, based on the experiences gained over last five–six decades on the reprocessing of spent fuels, some major drawbacks of TBP have been identified. Evaluation of alternative extractants is, therefore, desirable which can overcome at least some of these problems. Extensive studies have been carried out on the evaluation of N,N-dialkyl amides as extractants in the back-end of the nuclear fuel cycle for addressing the issues related to the reprocessing of U and Th based spent fuels. Under advanced fuel cycle scenario, efforts are also being made by countries with a developed nuclear technological base to provide safe nuclear power to other countries and to minimize proliferation concerns worldwide. This paper presents an overview of studies carried out in our laboratory on different aspects of reprocessing of U and Th based spent fuels employing N,N-dialkyl amides as extractants.

Estimation of 90Sr in reprocessing streams using pre-packed extraction chromatographic column of 4,4'(5')-bis(tert-butylcyclohexano)-18-crown-6 impregnated on amberlite XAD-7

This paper describes a method for the estimation of 90Sr in various PUREX and THOREX process streams. Radiostrontium present in the sample is loaded on an extraction chromatographic column containing 4,4’(5’)-bis(tert-butylcyclohexano)-18-crown-6 (DtBuCH18C6) in 1-octanol impregnated on amberlite XAD-7 after diluting it appropriately in about 3.0 M HNO3. The column is then washed using 3 M HNO3 to remove the impurities completely and finally 90Sr is eluted in 5 mL of water and subjected to beta counting without time delay. Recovery of 90Sr separation is studied in presence of varying quantity of uranium and thorium in the sample. The method is used to estimate 90Sr in actual process samples containing 90Sr activity in wide range of activity varying from Bq/mL to Ci/L. The results are compared with those obtained by conventional radiochemical precipitation method and found to agree within ±5 %.

A fast and sensitive method for the determination of uranium in the Thorex Process streams

This paper presents a simple, rapid and sensitive radiometric method for the determination of uranium in Thorex Process stream containing large amount of thorium. This method involves the extraction of uranium into 0.05M tri-n-octyl phosphine oxide (TOPO) in xylene at 2M HNO3. The extraction of thorium is prevented by masking them with suitable quantity of fluoride ions. The optimum experimental parameters for extraction of 233U were evaluated and using the most suitable experimental conditions the extracted uranium is radiometrically determined by α-counting in proportional counter with a prior knowledge of specific activity of uranium. Simultaneously in the same sample uranium was determined by spectrophotometric method using 2-(5bromo-2 pyridylazo)-5-diethylaminophenol (Bromo-PADAP) as chromogenic reagents. Simulated as well as actual samples of dissolver, conditioner and raffinate tank of Thorex stream have been analyzed by both these methods. The method was tested for as low as 0.15 μg of uranium and the results of these analyses were found to be satisfactory within the experimental limits.

Cation-exchange separation of uranium from thorium in nitric acid medium</p> </p>

Sorption behavior of Th and U on cation-exchange resins was investigated from nitric acid medium by both batch and column methods. The cation-exchange studies involved the sorption of UO22+ and Th4+ and their cationic complexes onto Dowex 50Wx8 and Dowex 50Wx4 resins (50-100 mesh). The batch data yielded a separation factor (Kd,Th/Kd,U) value of >100 for the cation-exchanger, Dowex 50Wx4 at 1-2M HNO3. Separation of uranium from thorium was also carried out by column method in nitric acid medium using cation-exchangers, Dowex 50Wx4 as well as Dowex 50Wx8. While uranium elution was possible at 1M HNO3, Th could be eluted only at higher concentration of nitric acid (>6M). The stripped solution emanating from a mixer settler employing di-2-ethyl hexyl isobutyramide as extractant and feed solution similar to THOREX process comprising 350 mg/l U and 380 mg/l Th in 0.75M HNO3 was loaded on the column and the decontamination factor value for U in the product was >1000.

Reprocessing of AHWR spent-fuel: challenges and strategies

Reprocessing of advanced heavy water reactor (AHWR) spent-fuel involves separation of Th, 233U and Pu, from the fission products and from one another. A proper combination of Purex and Thorex processes is required. The technology development for a reprocessing facility is extremely complex owing to high fissile content, high levels of irradiation, presence high of levels of 232U, difficulty in thoria dissolution, presence of thorium as the major constituent, problems due to third phase formation with Th, etc. It demands for development of suitable dissolution, solvent extraction, criticality control, U-Pu partitioning, and other equipments and/or techniques. Process modelling, simulation and optimisation are crucial in predicting behaviour of equipments/cycles, and in arriving at safe and optimum flowsheet. A significant success in this field has been achieved. This paper describes the reprocessing aspects pertaining to AHWR spent-fuel, indicating the major technological challenges, strategies to be followed and development requirements. A schematic flowsheet is proposed for Th-233U-Pu separation.

DISTRIBUTION BEHAVIOUR OF U(VI), Th(IV) AND Pa(V) FROM NITRIC ACID MEDIUM USING LINEAR AND BRANCHED CHAIN EXTRACTANTS

ABSTRACT The extraction behaviour of 1M solutions of tri-2-ethylhexyl phosphate (TEHP), di-2-ethyl hexyl isobutyramide (D2EHIBA), tri-n-butyl phosphate (TBP) and di-n-hexyl hexanamide (DHHA) in n-dodecane towards U(VI), Th(IV) and Pa(V) in the presence of 220 g/L of Th from nitric acid medium has been studied. The limiting organic concentrations (LOC) of thorium (g/L) for 1 M TBP and 1 M DHHA are evaluated as 31, 20 ( at 1 M HNO3) and 25,13 (at 4 M HNO3) respectively. The distribution ratio (D) values of U(VI), Th(IV) and Pa(V) in the presence of thorium (220 g/L) at. 1 M HNO3 suggest that branching in the alky group of amides suppresses the extraction considerably. In view of the selective extraction of U over Th by 5 % TBP in THOREX process at 4 M HNO3, distribution behaviour is also studied employing a lower concenfration (0·18 M) of extractant for comparison purpose, Separation factor (S. F.) values for U(VI) over Th(IV) under different experimental conditions consistently varied in the order: D2EHIBA > DHHA > TEHP > TBP. The quantitative extraction of 233U from a synthetic mixture containing 233U (10−5 M). 233Pa (10−11 M) and thorium (220 g/L) at 1 M HNO3 using 1 M solution of D2EHIBA in n-dodecane is achieved in three stages, Stripping and reusability studies of D2EHIBA have also been carried out.

Inductively coupled plasma atomic emission spectrometric studies for the determination of 233U in the Thorex process

Inductively coupled plasma atomic emission spectrometric (ICP-AES) studies carried out for the determination of 233U in the presence of large amounts of thorium are described. Thorium was separated by precipitation as oxalate in 1 mol l–1 HNO3 medium (after ascertaining the thorium content) and maintaining an excess oxalic acid concentration of 0.01 mol l–1 over the stoichiometric requirement. The supernatant was analysed by ICP-AES for 233U. A sensitive 238U spectral line of 385.958 nm with an isotopic shift of 0.008 nm for 233U and free from 232Th spectral line interference, was used for the analysis. A high resolution sequential spectrometer with an inductively coupled argon plasma excitation source was employed. The RSD was ±5% at 0.1 µg ml–1233U. Based on these studies, a method can be developed for the routine monitoring of 233U at ppb levels in waste streams of the Thorex process.

Determination of microgram amounts of uranium in thorium by differential-pulse polarography

A differential-pulse polarographic (DPP) method for the determination of uranium(VI) in the presence of a large excess of thorium without any prior separation of the latter is described. Disodium ethylenediaminetetraacetate and sodium carbonate–hydrogencarbonate buffer (pH 10) was used as the supporting electrolyte. The interference of several impurities expected in the Thorex process while processing irradiated thorium was overcome by using the above supporting electrolyte. A relative standard deviation of 2.4% for the determination of uranium at the 5 ppm level was obtained. The uranium(VI) values obtained by the proposed DPP method were compared with those obtained by inductively coupled plasma atomic emission spectrometry.

Crud Formation in the Purex and Thorex Processes

In the solvent extraction processes for reprocessing spent nuclear fuel, operational flow problems may be caused by the formation of stable emulsions, commonly called ''crud''. Our experiments have demonstrated that crud is generated by finely dispersed solids. Such fines can be produced by precipitation of zirconium with degradation products of tributyl phosphate. Experiments show that crud can also be produced by other solids, however, viz., finely divided silica and bentonite. Therefore, fines present in the aqueous feed solution might also produce crud. A special problem arises in the reextraction step in the Thorex process, where a precipitate of thorium dibutyl phosphate causes crud formation.

A Study of Pulse Columns for Thorium Fuel Reprocessing (I)

Two 5-m pulse columns with the same cartridge geometries have been installed for Thorex process studies. First, the differences of the aqueous and the organic continuous modes of operation have been investigated. Flow characteristics and extraction efficiencies of acetic acid and thorium nitrate have been obtained. Then, the uranium and thorium coextraction and partitioning have been carried out. The concentration profiles of thorium, uranium, and nitric acid along columns have been determined for the coextraction and the partitioning process and the Height Equivalent Theoretical Stage values have been calculated from these data.

Interpretation of the Extraction Mechanism of the Purex and Thorex Processes from Kinetics Data

Abstract The rate of transfer of species between phases in solvent extraction is expected to be controlled by chemical reaction kinetics, diffusion, or a combination of these steps. Transfer kinetics data for uranium(VI), thorium(IV), and zirconium(IV) indicate that the chemical reaction is the rate-controlling step for most dispersed systems. The evidence for this conclusion is: (1) the transfer rate constants for the extraction of uranium were the same for Lewis cell, single drop, and Kenics mixer tests; (2) the forward transfer rate constants increase with increasing TBP concentration (the viscosity of the organic phase also increases) in all cases; (3) the reverse transfer rate varies dramatically with changes in nitrate concentration in the aqueous phase (same phase viscosities); and (4) the difference in the activation energies of the forward and reverse reactions for uranium(VI) transfer is equal to the heat of reaction of uranium with TBP. A proposed mechanism for the solvent extraction kinetics which is consistent with the experimental results includes the formation or decomposition of a charged interfacial complex as the rate-controlling step and equilibration of the interfacial complex with the bulk organic complex as the rapid step.

Use of the SEPHIS-MOD4 Program for Modeling the Purex and Thorex Solvent Extraction Processes

Abstract The SEPHIS-MOD4 computer program simulates solvent extraction in the Purex and Thorex processes. The program employs simple mass balance equations to describe the movement of solutes through the idealized mixer-settler stages which are used in the model. These differential mass-balance equations are integrated by an iterative scheme that allows the user to observe transient changes and the resulting steady state profile in the mixer-settler system. Correlations describe the equilibrium concentrations in the aqueous and organic phases. Kinetic rate equations are used to simulate the reduction of plutonium by hydroxylamine nitrate and uranium(IV). The value of the SEPHIS-MOD4 program lies primarily in its application to the design and testing of solvent extraction flowsheets. Given a set of operating conditions, the program can predict the solute concentration profile at steady state and during the approach to steady state. Changes in the operating conditions can be used to determine the influence ...

Laboratory Development of the Acid Thorex Process For Recovery of Thorium Reactor Fuel

The Acid Thorex process has been developed on a laboratory scale for recovery of uranium and thorium from spent fuel solutions. The thorium and uranium are extracted by tributyl phosphate (TBP) wit...