Control and surveillance of redox potential for 233Pa dissolution in 2LiF-BeF2 molten salt.

233Pa, the precursor nuclide of 233U in the thorium-uranium conversion is prone to reductive deposition in 2LiF-BeF2 (66 : 34 mol%, FLiBe) molten salt. We explored the adjustment and control of the redox potential of the FLiBe melt to avoid the 233Pa reduction deposition. The experimental data indicated that the deposited 233Pa can be completely dissolved and reentered into the molten salt with the addition of oxidant NiF2, and the distribution and behaviour of uranium, thorium, neptunium, and most fission products did not have any significant change in the NiF2-oxidised FLiBe molten salt, showing the feasibility of this manner to make 233Pa exist stably in the melt. The effects of NiF2-addition on the behaviour of the fission products 95Nb and 131I in the FLiBe molten salt were also investigated. It was found that 131I could be used as a redox indicator to monitor the redox potential of the oxidation-enhanced FLiBe molten salt. All the information drawn from this study could provide significant support for the control and surveillance of the redox potential of the FLiBe molten salt in the upcoming thorium molten salt reactor (TMSR).

Effect of oxide impurities on the dissolution behavior of Th4+, Be2+ and U4+ in fluoride salts.

Oxides are one of the most important impurities in the fuel salt of molten salt reactors (MSRs), and excessive oxide impurities pose a risk to the safe operation of MSRs. This study focused on investigating the precipitation behavior between Th4+, U4+, and Be2+ with O2- in the 2LiF-BeF2 (FLiBe) eutectic salt system. The results showed that the solubility of UO2 was 5.52 × 10-3 mol kg-1, and the solubility product (Ksp) of UO2 was 6.14 × 10-7 mol3 kg-3 in FLiBe salt at 650 °C. It was also found that the O2- ion would firstly react with U4+ to form UO2, and then the excessive O2- would react with Be2+ to generate BeO in the FLiBe system. Despite conducting the solubility experiment of ThO2 and titration experiment of FLiBe-ThF4, the system failed to achieve the solubility and the Ksp of ThO2. The main reason for this was that O2- preferentially reacted with Be2+ over Th4+ to form precipitates, in other words, Be2+ exerted a protective effect against Th4+. Above all, this work experimentally demonstrated that in the FLiBe system, O2- preferentially combines with U4+ to form a precipitate, followed by Be2+, while Th4+ is relatively inert.

Activity Measurement of Iodine Fission Products in 2LiF-BeF2 Salt and Its Application in Redox Potential Surveillance for a Thorium Molten Salt Reactor.

The activities of 131I, 132I, 133I, and 135I produced by neutron-induced fission of 235U in 2LiF-BeF2 (FLiBe) eutectic salt and their dependence on the redox potential were studied. The dependence observed experimentally suggested that the activity ratio for 131I to 132I could be used as an indicator of the redox potential for FLiBe salt. Relying on the selective adsorption of iodine ions on the activated silver probe by ion exchange, a novel method for activity distribution measurement of the iodine isotopes in FLiBe salt was founded. The method is simple, fast, and easy to operate and would be suitable particularly to in situ monitor the redox potential of a thorium molten salt reactor, where the redox potential should keep at a high level to avoid possible safety risk induced by 233Pa deposition in the reactor.

Evaporation behavior of 233Pa in FLiBeZr molten salt.

In thorium molten salt reactors (TMSR), 233Pa is an important intermediate nuclide in the conversion chain of 232Th to 233U, its timely separation from the fuel salt is critically important for both the thorium-uranium (Th-U) fuel cycle and the neutron economy of the reactor. In this study, the evaporation behavior of 233Pa in the FLiBeZr molten salt was investigated during a vacuum distillation process. The separation characteristics between 233Pa and the major components of the fuel (salt and fission products) were evaluated in a calculation of the separation factors between these components. It was found that 233Pa5+ evaporated more readily than 233Pa4+ and the other components of the fuel, the relatively low temperature and medium pressure were much more beneficial to the separation of 233Pa5+ from FLiBeZr salt in the evaporation process, with the maximum value of the separation factor achieving more than 102. Results of distillation experiments also show that increasing the temperature and decreasing the ambient pressure enhances the separation between 233Pa5+ and most of the fission product nuclides due to the 233Pa5+ volatility more strongly depending on the process conditions. These results will be utilized to design a concept for a process for 233Pa separation from the fuel of a molten salt reactor.

Transient release of radioactive iodine from the fission of UF4 in 2LiF-BeF2 salt.

In this study, the behavior of fission product iodine released from the melting process of a mixture consisting of UF4 irradiated with neutrons and 2LiF-BeF2 (FLiBe) salt was studied. The experiment showed that a large amount of iodine was released immediately during melting and captured by Ni metal foils. The transient release of iodine observed in this experiment is attributed to the redox reaction between the hot atoms of the fission product iodine that cumulated due to long-time irradiation. The effect of the redox status of the molten salt on the transient release of iodine was also investigated. Based on this investigation, it was proposed that the activity ratios of 131I to salt-seeking fission products in the fuel salt, as an effective diagnostic criterion, may be used for the surveillance of the redox potential of fuel salts in a molten salt reactor.

Distribution and behaviour of 233Pa in 2LiF-BeF2 molten salt.

Distribution and behavior of 233Pa, essential in the thorium-uranium nuclear fuel cycle, were studied in 2LiF-BeF2 (66 : 34 mole%, FLiBe) molten salt by γ-ray spectrometry. The experiments showed that 233Pa deposited slightly on the surface of graphite crucible. The addition of Hastelloy and metallic lithium decreased the 233Pa specific activity in the salt by 1 to 2 orders of magnitude rapidly. Analysis indicated that reductive deposition of 233Pa was responsible for the rapid decrease of 233Pa specific activity in the salt. Additional experiments strongly supported the mechanism of reductive deposition of 233Pa induced by Hastelloy and metallic lithium. In view of the large deposition of 233Pa on Hastelloy, the possible influence of fissile nuclide 233U produced from 233Pa decay on the operation of thorium-based molten salt reactor was discussed.

Evaporation behavior of 2LiF-BeF2-ZrF4 molten salt with irradiated nuclear fuel.

The evaporation behaviours of various components were investigated by using a low pressure distillation method in a 2LiF-BeF2-ZrF4 mixture containing irradiated ThF4 and UF4. The experiment showed that BeF2 and ZrF4 were found to mainly condensate at the outer cover, the coolest zone, and their relative volatilities vs. LiF were 9.8 and 32.2, respectively, while for ThF4 and UF4, at four different temperature zones the values were almost constant, at 0.1 and 0.3. The radioactivity of various nuclides was further detected using gamma spectrometer analysis. 137Cs was hardly observed due to long half-time decay. 233Pa was found to co-evaporate with the carrier salt, while 239Np mainly remained in the residual salt as 237U. In different temperature zones, the decontamination factors of rare earth in receiver salts ranged from 10 to 103. On the basis of the investigation, it was proposed that the distribution of various nuclides after distillation, may be helpful to design the feasible condensate system to recover the carried salt in a molten salt reactor.

Amidoxime-grafted hydrothermal carbon microspheres for highly selective separation of uranium.

A new amidoxime-functionalized carbonaceous sorbent has been successfully prepared using hydrothermal carbon microsphere as solid matrix and diaminomaleonitrile as precursor of amidoxime ligand. Effects of pH, sorbent dosage, contact time, temperature, initial U(VI) concentration and ionic strength on U(VI) sorption were investigated in detail through batch experiments. Sorption of U(VI) on the sorbent was pH-dependent. Sorption equilibrium was reached in 5 min. Distinctively, higher temperature was beneficial to the sorption of U(VI) in the range of 15-60 degrees C, high ionic strength up to 1 mol L(-1) NaNO3 had almost no effect on the sorption, and the maximum U(VI) sorption capacity of 466 mg g(-1) was observed under the conditions tested. The as-synthesized sorbent exhibited a high selectivity for U(VI) over other 12 competing ions coexisting in a simulated nuclear industrial effluent sample and the U(VI) sorption amount reached up to 1.09 mmol g(-1), accounting for about 52% of the total sorption amount.

Preparation and adsorption performance of 5-azacytosine-functionalized hydrothermal carbon for selective solid-phase extraction of uranium.

A new solid-phase extraction adsorbent was prepared by employing a two-step "grafting from" approach to anchor a multidentate N-donor ligand, 5-azacytosine onto hydrothermal carbon (HTC) microspheres for highly selective separation of U(VI) from multi-ion system. Fourier-transform infrared and X-ray photoelectron spectroscopies were used to analyze the chemical structure and properties of resultant HTC-based materials. The adsorption behavior of U(VI) onto the adsorbent was investigated as functions of pH, contact time, ionic strength, temperature, and initial U(VI) concentration using batch adsorption experiments. The U(VI) adsorption was of pH dependent. The adsorption achieved equilibrium within 30 min and followed a pseudo-second-order equation. The adsorption amount of U(VI) increased with raising the temperature from 283.15 to 333.15K. Remarkably, high ionic strength up to 5.0 mol L(-1) NaNO(3) had only slight effect on the adsorption. The maximum U(VI) adsorption capacity reached 408.36 mg g(-1) at 333.15K and pH 4.5. Results from batch experiments in a simulated nuclear industrial effluent, containing 13 co-existing cations including uranyl ion, showed a high adsorption capacity and selectivity of the adsorbent for uranium (0.63 mmol U g(-1), accounting for about 67% of the total adsorption amount).

Selective solid-phase extraction of uranium by salicylideneimine-functionalized hydrothermal carbon.

A new salicylideneimine-functionalized hydrothermal-carbon-based solid-phase extractant was developed for the purpose of separating uranium selectively for sustainability of uranium resources. The resulting adsorption material was obtained via hydrothermal carbonization, calcination at mild temperature (573.15K), amination, and grafting with salicylaldehyde in sequence. Both Fourier transform infrared spectra and elemental analysis proved the successful grafting of salicylideneimine onto hydrothermal carbon matrix. Adsorption behaviors of the extractant on uranium(VI) were investigated by varying pH values of solution, adsorbent amounts, contact times, initial metal concentrations, temperatures, and ionic strengths. An optimum adsorption capacity of 1.10 mmol g(-1) (261 mg g(-1)) for uranium(VI) was obtained at pH 4.3. The present adsorption process obeyed pseudo-second-order model and Langmuir isotherm. Thermodynamic parameters (ΔH=+8.81 kJ mol(-1), ΔS=+110 J K(-1)mol(-1), ΔG=-23.0 kJ mol(-1)) indicated the adsorption process was endothermic and spontaneous. Results from batch adsorption test in simulated nuclear industrial effluent, containing Cs(+), Sr(2+), Ba(2+), Mn(2+), Co(2+), Ni(2+), Zn(2+), La(3+), Ce(3+), Nd(3+), Sm(3+), and Gd(3+), showed the adsorbent could separate uranium(VI) from those competitive ions with high selectivity. The adsorbent might be promising for use in certain key steps in any future sustainable nuclear fuel cycle.

Sorption of uranium(VI) using oxime-grafted ordered mesoporous carbon CMK-5.

A new sorbent for uranium(VI) has been developed by functionalizing ordered mesoporous carbon CMK-5 with 4-acetophenone oxime via thermally initiated diazotization. The sorption of U(VI) ions onto the functionalized CMK-5 (Oxime-CMK-5) was investigated as a function of sorbent dosage, pH value, contact time, ionic strength and temperature using batch sorption techniques. The results showed that U(VI) sorption onto Oxime-CMK-5 was strongly dependent on pH, but to a lesser extent, on ionic strength. Kinetic studies revealed that the sorption process achieved equilibrium within 30 min and followed a pseudo-first-order rate equation. The isothermal data correlated with the Langmuir model better than the Freundlich model. Thermodynamic data indicated the spontaneous and endothermic nature of the process. Under current experimental conditions, a maximum U(VI) sorption capacity was found to be 65.18 mg/g. Quantitative recovery of uranium was achieved by desorbing the U(VI)-loaded Oxime-CMK-5 with 1.0 mol/L HCl and no significant decrease in U(VI) sorption capability of Oxime-CMK-5 was observed after five consecutive sorption-desorption cycles. The sorption study performed in a simulated nuclear industry effluent demonstrated that the new sorbent showed a desirable selectivity for U(VI) ions over a range of competing metal ions.