In situ high-temperature EXAFS measurements on radioactive and air-sensitive molten salt materials.

The development at the Delft University of Technology (TU Delft, The Netherlands) of an experimental set-up dedicated to high-temperature in situ EXAFS measurements of radioactive, air-sensitive and corrosive fluoride salts is reported. A detailed description of the sample containment cell, of the furnace design, and of the measurement geometry allowing simultaneous transmission and fluorescence measurements is given herein. The performance of the equipment is tested with the room-temperature measurement of thorium tetrafluoride, and the Th-F and Th-Th bond distances obtained by fitting of the EXAFS data are compared with the ones extracted from a refinement of neutron diffraction data collected at the PEARL beamline at TU Delft. The adequacy of the sample confinement is checked with a mapping of the thorium concentration profile of molten salt material. Finally, a few selected salt mixtures (LiF:ThF4) = (0.9:0.1), (0.75:0.25), (0.5:0.5) and (NaF:ThF4) = (0.67:0.33), (0.5:0.5) are measured in the molten state. Qualitative trends along the series are discussed, and the experimental data for the (LiF:ThF4) = (0.5:0.5) composition are compared with the EXAFS spectrum generated from molecular dynamics simulations.

Short life fission products extracted from molten salt reactor fuel for radiopharmaceutical applications.

This work studies the potential of using short life fission product (AFp) radioisotopes e.g. 82Br, 86Rb, (90Sr) - 90mY, (99Mo) - 99mTc, 103Ru - 103mRh, 111Ag, 127Sb - 127(m)Te, 126I, 131I, 133Xe, 136Cs, 141Ce, 143Ce, 143Pr, 147Nd - 147Pm, 149Pm, 153Sm, 156Eu, 159Gd and 161Tb, extracted from a molten salt reactor and their separation using specific thermodynamic and radiochemical conditions. Their utilisation for coupled radiodiagnostics and radiotherapy is a key consideration. A molten salt reactor produces fission products during operation. These radioisotopes can be separated at line from the liquid fuel by evaporation/distillation, chemical reduction (using H2 doped gas), electro-deposition and/or chemical oxidation (using Cl2 doped gas). They can be refined and chemically treated for radiopharmaceutical use for imaging and radiodiagnostics utilising γ radioscopy or positron emission tomography, and potentially in radiotherapy to target specific cancers or viral diseases using ß- emitters. Some of the AFp isotopes are currently used for radiodiagnostics because they emit γ rays of energy 50-200 keV. However, some may also be used in parallel for radiotherapy utilising their ß- (EMean ≈ 100 keV) emission whose mean free pathway of c.a. 100 nm in biological tissue is much smaller than their penetration depth. Focus is given to 86Rb, 90Y, 99mTc, 131I and 133Xe as well as on the ALn isotopes (141Ce, 143Ce - 143Pr, 147Nd - 147Pm, 149Pm and 153Sm) because of their strong potential for complexation with bio-ligands (e.g. DOTA) or for their ability to form micro-nano-spheres, and because of their potential for dual radiodiagnostics and radiotherapy. It is shown that these radio-lanthanides could also replace 177Lu for the treatment of specific cancers.

Thermophysical Properties of FUNaK (NaF-KF-UF4) Eutectics.

General interest in the deployment of molten salt reactors (MSRs) is growing, while the available data on uranium-containing fuel salt candidates remains scarce. Thermophysical data are one of the key parameters for reactor design and understanding reactor operability. Hence, filling in the gap of the missing data is crucial to allow for the advancement of MSRs. This study provides novel data for two eutectic compositions within the NaF-KF-UF4 ternary system which serve as potential fuel candidates for MSRs. Experimental measurements include their melting point, density, fusion enthalpy, and vapor pressure. Additionally, their boiling point was extrapolated from the vapor pressure data, which were, at the same time, used to determine the enthalpy of vaporization. The obtained thermodynamic values were compared with available data from the literature but also with results from thermochemical equilibrium calculations using the JRCMSD database, finding a good correlation, which thus contributed to database validation. Preliminary thoughts on fluoride salt reactor operability based on the obtained results are discussed in this study.

Detection of Off-Gassed Products From Molten Salts Using Laser-Induced Breakdown Spectroscopy.

The detection of off-gassed sodium from molten sodium nitrate (NaNO3) at temperatures between 330 °C and 505 °C and off-gassed calcium from molten lithium chloride-potassium chloride eutectic (LKE) mixtures at 510 °C with laser-induced breakdown spectroscopy (LIBS) was demonstrated. NaNO3 and LKE samples were melted in a custom-built crucible that promoted the generation of off-gassed products from the molten sample. The off-gassed products were analyzed with a LIBS system designed to probe the high-temperature environment. Na D emission lines, Na(I)588.99 nm and Na(I) 589.59 nm, were detected from the NaNO3 samples after reaching a temperature threshold, which indicated the occurrence of phase change. In LKE mixtures, the detection of Ca impurities at a concentration of 78 mg/kg was possible using the emission lines Ca(II) 393.66 nm and Ca(II) 395.85 nm. This work demonstrates the real-time monitoring capabilities of LIBS in high-temperature environments that simulate the conditions of molten salt reactors.

Monitoring Noble Gases (Xe and Kr) and Aerosols (Cs and Rb) in a Molten Salt Reactor Surrogate Off-Gas Stream Using Laser-Induced Breakdown Spectroscopy (LIBS).

This study with surrogate materials shows that laser-induced breakdown spectroscopy (LIBS) is a robust tool with promising capability toward monitoring gaseous (Xe and Kr) and aerosol (Cs and Rb) species in an off-gas stream from a molten salt reactor (MSR). MSRs will continually evolve fission products into the cover gas flowing across the reactor headspace. The cover gas entrains Xe and Kr gases, along with aerosol particles, before passing into an off-gas treatment system. Univariate models of Xe and Kr peaks showed a strong correlation to concentration indicated by their coefficients of determination of 0.983 and 0.997, respectively. Multivariate models were built for all four analytes using partial least squares regression coupled with preprocessing steps including normalization, trimming, and/or genetic algorithm derived filters. The models were evaluated by predicting the concentrations of the analytes in four validation samples, in which all calibration models were successfully validated at a confidence interval of 99.9%. Lastly, pressure controllers were used to regulate the mass flow rate of Kr flowing into the measurement cell in sinusoidal and stepwise waveforms to test the real-time monitoring capabilities of the regression models. Both univariate and partial least squares Kr models were able to successfully quantify the gas concentration in the real-time evaluation. The root mean squared error of prediction (RMSEP) values for these real-time tests were calculated to be 0.051, 0.060, and 0.121 mol% demonstrating the measurement systems' capability to perform online monitoring with acceptable accuracy.

Design and operation of a molten salt electrochemical cell.

Molten salts such as 2LiF-BeF2 (FLiBe) have been proposed as coolants for advanced nuclear fission and fusion reactors. Critical to the design, licensing and operation of these reactors is characterization and understanding of the chemical behavior and mass transport of activation and fission products, corrosion products, and other solutes in the coolant. Electrochemical techniques are a powerful suite of tools for probing these phenomena. The design of an experimental cell for molten salt electrochemistry is described herein. As a demonstration of this design, details of the experimental methods used to conduct electrochemical experiments with molten FLiBe with addition of LiH are provided. Decommissioning of the cell is considered from the point of view of decontamination and waste generated. Main features of the cell include:•Suitable for operation up to 800 °C; suitable for operation inside and outside of a glovebox.•Enables sweep gas, gas sampling and analysis; enables addition of solid and liquid materials during operation.•Supports a variety of electrode materials and arrangements.

Investigation of the Quasi-Binary Phase Diagram FLiNaK-NdF3.

The NdF3 solubility in molten eutectic FLiNaK, which is a conceivable medium for a molten salt reactor (MSR), was determined by the quasi-binary phase diagram FLiNaK-NdF3. The eutectic mixture FLiNaK was prepared by direct melting of components LiF, NaF and KF·HF. The acidic anhydrous salt (KF·HF) was used instead of the hygroscopic KF. The NdF3 was sintered by hydrofluorination of Nd2O3. The oxygen impurity in the prepared eutectic FLiNaK, determined by an oxygen analyzer LECO OH836, was 0.036 wt.%, whereas the NdF3 contained 0.04 wt.% of oxygen. A part of the FLiNaK-NdF3 quasi-binary phase diagram was obtained using two thermal analysis techniques: differential thermal analysis (DTA) and differential scanning calorimetry (DSC). The FLiNaK-NdF3 phase diagram in the region of 0-30 mol.% NdF3 contains one eutectic at 2 mol.% NdF3 and 450 °C and two peritectic points: 8 mol.% NdF3 at 500 °C and 22 mol.% NdF3 at 575 °C. The region of the FLiNaK-NdF3 phase diagram below the liquidus line is rather complicated due to the complex structure of the multicomponent system in its molten state, as in its solid state. The NdF3 solubility in FLiNaK is about 5 mol.% at 490 °C and 29 mol.% at 700 °C; this means that the process of the MA transmutation in the MSR can be carried out in molten FLiNaK with a content of actinides as high as 15-20 mol.% in the temperature range of 550-650 °C.

Modeling of fission and activation products in molten salt reactors and their potential impact on the radionuclide monitoring stations of the International Monitoring System.

Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to model the production of activation and fission products and analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios: (1) a sealed reactor with releases only during operational shutdown, and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (Boiling Water Reactor, Pressurized Water Reactor, High-Power Channel-Type Reactor). In cases where continuous reprocessing of the fuel salt occurred, both the radioxenon and radioiodine signatures were nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 1015-8×1016 Bq/d for 133Xe, which may adversely affect nuclear explosion monitoring, if no abatement is used. An assessment was made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1).

Analyses of production capacity of 89Sr and 90Sr in the 2 MW molten salt reactor.

The production capacity of 89Sr and 90Sr in the 2 MW MSR are evaluated. The gaseous 89Kr and 90Kr are extracted from the core through the helium bubbling system, and then decay to 89Sr and 90Sr, respectively. In order to improve purity of 89Sr product, two cooling devices are adopted in the 89Sr and 90Sr production system. The annual yields of 89Sr and 90Sr are about 9000 Ci and 32 Ci, respectively, and the impurity of 89Sr product is less than 2 ppm which can meet the medical requirement.

Supply of I-131 in a 2 MW molten salt reactor with different production methods.

Four I-131 production methods including irradiated TeO2 target and uranium target in the irradiation channel, batch-wise extracted iodine from the fuel salt, and online extracted solid tellurium through the by-pass loop system have been assessed in a 2 MW molten salt reactor. The latter method can produce a large annual yield of I-131 (about 155,000 Ci). The radioactivity shielding demand of the latter method is much smaller than the other I-131 production methods under the identical annual yield of I-131.

Sustainable supply of 99Mo source in a 2 MW molten salt reactor using low-enriched uranium.

The 99Mo production in a 2 MW molten salt reactor using liquid low-enriched uranium (LEU) fuel has been evaluated. The batch-wise extraction period of 99Mo is optimized to be one day corresponding to 9415 6-day Ci/week of the 99Mo production rate. The required amount of uranium is only 4.77 kg annually. The required chemically reprocessed amount of FPs is about 58.4 g annually, accounting for only 4.9% of the solid LEU target method under the identical production capacity of 99Mo.

The Effect of Milling Time on the Microstructural Characteristics and Strengthening Mechanisms of NiMo-SiC Alloys Prepared via Powder Metallurgy.

A new generation of alloys, which rely on a combination of various strengthening mechanisms, has been developed for application in molten salt nuclear reactors. In the current study, a battery of dispersion and precipitation-strengthened (DPS) NiMo-based alloys containing varying amounts of SiC (0.5-2.5 wt %) were prepared from Ni-Mo-SiC powder mixture via a mechanical alloying (MA) route followed by spark plasma sintering (SPS) and rapid cooling. Neutron Powder Diffraction (NPD), Electron Back Scattering Diffraction (EBSD), and Transmission Electron Microscopy (TEM) were employed in the characterization of the microstructural properties of these in-house prepared NiMo-SiC DPS alloys. The study showed that uniformly-dispersed SiC particles provide dispersion strengthening, the precipitation of nano-scale Ni3Si particles provides precipitation strengthening, and the solid-solution of Mo in the Ni matrix provides solid-solution strengthening. It was further shown that the milling time has significant effects on the microstructural characteristics of these alloys. Increased milling time seems to limit the grain growth of the NiMo matrix by producing well-dispersed Mo2C particles during sintering. The amount of grain boundaries greatly increases the Hall-Petch strengthening, resulting in significantly higher strength in the case of 48-h-milled NiMo-SiC DPS alloys compared with the 8-h-milled alloys. However, it was also shown that the total elongation is considerably reduced in the 48-h-milled NiMo-SiC DPS alloy due to high porosity. The porosity is a result of cold welding of the powder mixture during the extended milling process.

Novel waste printed circuit board recycling process with molten salt.

The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going.