Natural Lithium Research Articles

AS33-01 - Natural lithium in drinking water and suicide mortality

There is evidence that natural levels of lithium in drinking water may have a protective effect on suicide mortality. To evaluate the association between local lithium levels in drinking water and suicide mortality in Austria. A nationwide sample of lithium measurements was examined for association with suicide rates across Austrian districts. Multivariate regression models were adjusted for socioeconomic factors. Sensitivity analyses and weighted least squares regression were used to challenge the robustness of the results. The overall suicide rate as well as the suicide mortality ratio were inversely associated with lithium levels in drinking water and remained significant after sensitivity analyses and adjustment for socioeconomic factors. In replicating and extending previous results, this study provides strong evidence that geographic regions with higher natural lithium concentrations in drinking water are associated with lower suicide mortality rates.

Characterization of (<i>ENR-50</i>)-Ionic Liquid Based Electrolyte System

Ionic liquid based on imidazolium cation; 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (EMITFSI) has been incorporated with epoxidized natural rubber-50 (ENR-50) and lithium bis(trifluoromethanesulfonyl)imide (LiTFSI) to obtain electrolyte material. Fourier transform infrared spectroscopy (FTIR) spectra showed evidence of complexation between ENR-50, EMITFSI and LiTFSI. Glass transition temperature, Tg displayed an increasing trend with increase in salt concentration. The incorporation of EMITFSI resulted in an increase in ionic conductivity. The increase in ionic conductivity was attributed to the role of ionic liquid which reduced Tg, thus, facilitated ion conduction in the system. The highest ionic conductivity at room temperature was 5.72 ´ 10-4 S cm-1 for sample containing 20 wt% of EMITFSI and 50 wt% of LiTFSI.

Lithium in drinking water and suicide mortality

There is some evidence that natural levels of lithium in drinking water may have a protective effect on suicide mortality. To evaluate the association between local lithium levels in drinking water and suicide mortality at district level in Austria. A nationwide sample of 6460 lithium measurements was examined for association with suicide rates per 100,000 population and suicide standardised mortality ratios across all 99 Austrian districts. Multivariate regression models were adjusted for well-known socioeconomic factors known to influence suicide mortality in Austria (population density, per capita income, proportion of Roman Catholics, as well as the availability of mental health service providers). Sensitivity analyses and weighted least squares regression were used to challenge the robustness of the results. The overall suicide rate (R(2) = 0.15, β = -0.39, t = -4.14, P = 0.000073) as well as the suicide mortality ratio (R(2) = 0.17, β = -0.41, t = -4.38, P = 0.000030) were inversely associated with lithium levels in drinking water and remained significant after sensitivity analyses and adjustment for socioeconomic factors. In replicating and extending previous results, this study provides strong evidence that geographic regions with higher natural lithium concentrations in drinking water are associated with lower suicide mortality rates.

Derivation of empirical equations for neutronic performance in a thorium fusion breeder with various coolants using regression analysis

In this paper, regression analyses (RA) are presented for the neutronic calculation of ThO 2 mixed 244CmO 2 fuel with different neutronic parameters for various coolants, natural lithium, Li 20Sn 80 and Flinabe, respectively. The tritium breeding ratio (TBR), energy multiplication factor ( M), total fission rate ( Σ f) and 232Th(n, γ) reaction is computed by XSDRNPM. In addition, this numerical results are estimated by RA depends on neutronic parameters and the empirical equations for neutronic performance are acquired. The results obtained by using XSDRNPM and the results of the RA, obtained empirical equations, are compared. The empirical equations indicate that RA can successfully be used for the prediction of the neutronic performance parameters in the hybrid reactor with a high degree of accuracy. In addition, correlation matrix is calculated to determined statistical relationships between variables TBR, M, Σ f, and 232Th(n, γ).

Growth and characterization of congruent lithium isotope niobate (7LiNbO3) single crystal

For further investigation on the defect structure of lithium niobate (LiNbO3—hereafter LN) crystals using the neutron scattering technique, a congruent lithium isotope niobate (C7LN) crystal was grown by the Czochralski method from a 48.4% 7Li melt. The X-ray powder diffraction (XRPD) results show that the as-grown crystal is a single-phased (R3c) LN. We observed 4A1(TO)+7E(TO) Raman modes. The crystal composition (Li mol%) determined from UV absorption edge and Raman measurements is 47.47% and 47.34%, respectively. The specific heat (Cp), thermal diffusion coefficients (λa, λc), thermal conductivities (κa, κc) and the average thermal expansion coefficients (αa, αc) were determined. The effect of Li vacancy content is analyzed by comparing the results with those obtained for stoichiometric lithium isotope niobate (S7LN) and congruent natural lithium niobate (CLN) crystals. The values of all thermal parameters of C7LN are smaller than those of S7LN and CLN, which may be related to either the difference of vacancy content or isotope substitution.

Analysis of temperature distribution, burn-up and breeding parameters in nuclear fuel rod in fusion-fission reactor system fueled with mixed ThO2-UO2 fuel

Temperature distribution in nuclear fuel rod, burn-up (BU), possibility of nuclear fuel rejuvenation and breeding parameters are investigated for different coolants under various first wall loads (Pw=2, 5 and 7 MW m−2) in a deuterium–tritium driven fusion–fission reactor system, fueled with mixed UO2-ThO2 fuel. The fuel mixture is considered to be mixed with various mixture fractions. A (D-T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. In the tritium breeding zone of the blanket Li2O is used and blanket is reflected by graphite for neutron economy. The fissile fuel zone is considered to be cooled with three different coolants [gas (He, CO2), Flibe (Li2BeF4) and natural lithium (Li)] with the volumetric ratio of coolant-to-fuel [e=(Vcoolant/Vfuel)=1:1 (45.515% coolant, 45.515% fuel and 8.971% clad)] for nuclear heat transfer. The behavior of the blankets fueled with mixed fuel mentioned above are observed for 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. The fissile fuel breeding occurs through the neutron capture reaction in the 232Th (ThO2), in the 238U (UO2) isotopes. As a result of the calculations, fuel mixtures having the best performance parameters have been obtained as follows for different coolants and under the 7 MW m−2 first wall load during operation period without reaching the melting point: • The blanket fueled with 70 wt% UO2−30 wt% ThO2 for Flibe (Li2BeF4) coolant • The blanket fueled with 80 wt% UO2−20 wt% ThO2 for Gas (He, CO2) coolant • The blanket fueled with 90 wt% UO2−10 wt% ThO2 for natural lithium (Li) coolant Clad surface temperature Tc controlled by coolant velocity is taken between certain operation temperatures (Tc=200, 300, 400, 500 and 600°C). The best Cumulative Fissile Fuel Enrichment (CFFE) grade of the nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials (233U, 235U, 239Pu, and 241Pu) is obtained in Flibe (Li2BeF4) coolant blankets, followed by Gas (He, CO2), whereas natural lithium (Li) coolant shows a poor rejuvenation performance in all fuels. CFFE reaches the maximum value (CFFE=10.6200%) in blanket cooled with Li2BeF4 coolant, followed by gas coolant with 9.0319% and natural lithium coolant with 7.9863% without reaching the melting point (Tmax=Tm>2830°C) of the fuel materials. At this point, the maximum temperature in centerline of the fuel rod (Tm) reach to 2727.57°C in blanket cooled with Li2BeF4 coolant. However, in the blanket fueled with pure UO2 fuel and cooled with Flibe, CFFE and Centerline Temperature of the fuel rods (Tm) are reached to 9.7277% and 2811.13°C respectively at the end of 39 months. In the blanket fueled with pure UO2 and cooled with gas, CFFE and Centerline Temperature of the fuel rods (Tm) are reached to 7.8115% and 2814.57°C at the end of 38 months and 7.2777% and 2724.34°C at the end of the 43 months in the blanket cooled with natural Li. In the blanket fueled with pure ThO2, CFFE reaches the maximum value with 9.2843% in the blanket cooled with Flibe. The enrichment is sufficient for LWR reactor. The temporal change of the fuel composition during the hybrid reactor plant operation and the most pertinent nuclear reactions been considered in the course of BU. The BU reaches the maximum value (72 749.1 MWd t−1 (tone)) in the gas coolant blanket, followed by Flibe coolant blanket with 67 641.0 MWd t−1 natural lithium coolant blanket with 63 897.5 MWd t−1 during the operation period without reaching the melting point. Copyright © 2010 John Wiley & Sons, Ltd.

Research Progress in Ionic Functional Materials Used for Energy Saving and Emission Reduction

The application of ionic functional materials for energy saving and emission reduction in China was introduced in this paper. Firstly, the application of mine materials in the field of architectural energy saving, including heat insulation materials, combustion energy saving materials and enhancing heat emission materials,etc, was introduced. Secondly, the application of ionic functional materials in the field of automobile energy saving was introduced. The product technology of promoting energy saving through activating fuel, air and cycling water by activative bed technology was also introduced, aimed at industry boiler. Finally, the application in battery field of lithium-ion storage materials, including natural pyrite and lithium iron phosphate materials was introduced.

The Impact of Depleted 6Li on the Standard Atomic Weight of Lithium

Li (lithium) is one of a handful of elements whose stable isotopic ratio varies in natural terrestrial samples to the extent that the resulting atomic weight variation exceeds the measurement uncertainty on the value. As a result, the standard atomic weight of lithium is more accurately characterized as a range of atomic weight values from 6.9387 to 6.9959. Lithium has become the least accurately known atomic weight because of the existence and the distribution in the distant past of some chemical reagents, which were depleted in the 6 Li isotope of natural lithium. This background story brings to light an interesting page of history.

6Li Atom Percentage Determination by Atomic Absorption–Emission Spectrometry Using a Natural Lithium Hollow Cathode Lamp

A simple method has been developed for the determination of 6Li atom % using combined atomic emission-absorption spectrometry employing a commonly available natural lithium hollow cathode lamp. Unlike in previous practice, there is no need for specially fabricated and high cost 6Li and 7Li monoisotopic lamps in this method. The method requires adjustment of total lithium contents of the sample, i.e., 6Li + 7Li, to 2 microg x mL(-1) based upon atomic emission spectroscopy (AES) (C(aes)) against a 2 microg x mL(-1) natural lithium standard. The concentration of the sample was then analyzed by atomic absorption spectroscopy (AAS) measurements (C(aas)). The difference between the concentration measured by AES and AAS, i.e., C(aes)-C(aas), was calculated. The magnitude of the difference was found to be a function of 6Li fraction in the sample. A calibration curve was constructed by plotting 6Li atom % versus [(C(aes)-C(aas))/C(aes)] x 100. 6Li atom % of an unknown sample can be evaluated by putting its [(C(aes)-C(aas))/C(aes)] x 100 value in the calibration curve. The method is fast, convenient, and precise.

The thermodynamic characteristics of natural iron-lithium micas

A thermochemical study of natural lithium micas, iron-containing polylithionite and lepidolite, was performed on a high-temperature heat-flux Calvet microcalorimeter (Setaram, France). Melt solution calorimetry was used to measure the enthalpies of mineral formation from the elements Δf H°el (298.15 K), −5989.3 ± 9.6 and −5981.3 ± 6.3 kJ/mol, respectively. The drop method was used to determine the enthalpy increments heating of the micas over the temperature interval 444–973 K. The equations for the temperature dependences of the heat capacities and enthalpies of Fe-polylithionite and Fe-lepidolite were obtained. The S° (298.15 K) and Δf G°el (298.15 K) values were estimated. The thermodynamic functions of the micas were calculated over the temperature range 298.15–1000 K.

Monte Carlo calculation for various enrichment lithium coolant using different data libraries in a hybrid reactor

The main objective of this study is to compare the effect of the natural lithium and lithium with different enrichments between 10% and 90% on neutronic parameters; such as tritium breeding ratio (TBR), displacement per atom (DPA) and gas production using the different data libraries (ENDF/B.V, ENDF/B.VI and CLAW IV). Therefore, the natural lithium and different enrichment lithium were used as a moderator in an experimental hybrid reactor for the calculation of the nuclear parameters. Neutronic calculations were performed by recent Monte Carlo Neutron-Particle Transport code MCNP5 version 1.40 for a 14.1 MeV (D, T) fusion driver under a neutron wall load of 2.25 MW/m 2 (10 14 n/s). TBR values in the blanket for all investigated cases were obtained greater than the minimum requirement (TBR > 1.05). Considering radiation damage limits (100 DPA and 500 appm/FPY) for structural materials, the FW replacement will be needed every 2.1 and 3.5 years for DPA and He-production, respectively.

Grafted natural rubber-based polymer electrolytes: ATR-FTIR and conductivity studies

Attenuated total reflectance–Fourier transformed infrared spectroscopy measurement is employed to study the interactions between the components of 30% methyl-grafted natural rubber (MG30), lithium trifluromethanesulfonate (LiCF3SO3 or LiTF), and propylene carbonate (PC). Vibrational spectra data of LiTF reveals that the νs(SO3) at 1,045 cm−1, δs(CF3) at 777 cm−1, and C=O stretching mode at 1,728 cm−1 for MG30 have shifted to lower wave numbers in MG30–LiTF complexes indicating that complexation has occurred between MG30 and LiTF. The solvation of lithium ion is manifested in Li+ ← O=C interaction as shown by the downshifting and upshifting of C=O mode at 1,788 to 1,775 cm−1 and νas(SO3) at 1,250 to 1258 cm−1, respectively, in LiTF–PC electrolytes. There is no experimental evidence of the interaction between MG30 and PC. Competition between MG30 and PC on associating with lithium ion is studied, and the studies show that the interaction between MG30–LiTF is stronger than that of the PC–LiTF in plasticized polymer–salt complexes. The effect of PC on the ionic conductivity of the MG30–LiTF system is explained in terms of the polymer, plasticizer, and salt interactions. The temperature dependence of conductivity of the polymer films obeys the Vogel–Tamman–Fulcher relation. Values of conductivity and activation energy of the MG30-based polymer electrolyte systems are presented and discussed.

Use of Nuclear Data Sensitivity and Uncertainty Analysis for the Design Preparation of the HCLL Breeder Blanket Mockup Experiment for ITER

An experiment on a mockup of the test blanket module based on helium-cooled lithium lead (HCLL) concept will be performed in 2008 in the Frascati Neutron Generator (FNG) in order to study neutronics characteristics of the module and the accuracy of the computational tools. With the objective to prepare and optimise the design of the mockup in the sense to provide maximum information on the state-of-the-art of the cross-section data the mockup was pre-analysed using the deterministic codes for the sensitivity/uncertainty analysis. The neutron fluxes and tritium production rate (TPR), their sensitivity to the underlying basic cross-sections, as well as the corresponding uncertainties were calculated using the deterministic transport codes (DOORS package), the sensitivity/uncertainty code package SUSD3D, and the VITAMINJ/ COVA covariance matrix libraries. The cross-section reactions with largest contribution to the uncertainty of the calculated TPR were identified to be (n,2n) and (n,3n) reactions on lead. The conclusions of this work support the main benchmark design and suggest some modifications and improvements. In particular this study recommends the use, as far as possible, of both natural and enriched lithium pellets for the TRP measurements. The combined use is expected to provide additional and complementary information on the sensitive cross-sections.

Experimental Study on Corrosion of Ceramic Materials in Natural Convection Lithium Loop

Corrosion of candidate materials of insulating coating was investigated in natural convection lithium (Li) loop. The loop was constructed for present study and bulk specimens of Er2O3, Y2O3 and AlN were placed at several positions in the loop. The operation temperature was 400 °C in the low temperature region and 500 °C in the high temperature region. After the exposure, specimens were cleaned and the corrosion characteristics were investigated. The results indicated that the corrosion characteristics of the bulk specimens were similar to those obtained by the static test in the previous study.

Axisymmetric Tandem Mirror Magnetic Fusion Energy Power Plant with Thick Liquid-Walls

A fusion power plant is described that utilizes a new version of the tandem mirror device including spinning liquid walls. The magnetic configuration is evaluated with an axisymmetric equilibrium code predicting an average beta of 60%. The geometry allows a flowing molten salt, (flibe–Li2BeF4), which protects the walls and structures from damage arising from neutrons and plasma particles. The surface of the liquid facing the burning plasma is heated by bremsstrahlung radiation, line radiation, and by neutrons. The temperature of the free surface of the liquid is calculated, and then the evaporation rate is estimated from vapor-pressure data. The allowed impurity concentration in the burning plasma is taken as 1% fluorine, which gives a 17% reduction in the fusion power owing to D/T fuel dilution, with F line-radiation causing minor power degradation. The end leakage power density of 0.6 MW/m2 is readily handled by liquid jets. The tritium breeding is adequate with natural lithium. The simple geometry and the use of liquid walls promise the cost of power competitive with that from fission and coal.

An efficient pathway for Li6 isotope enrichment

Separation of lithium isotopes has been achieved using two-step laser photoionization in conjunction with an atomic beam and in-house built time of flight (TOF) mass spectrometer. We present an efficient pathway for the enrichment of Li6 isotope by tuning the exciter laser to the 3p 2P1/2, 3/2 excited state of Li6. A concentration of up to 60% is demonstrated from a natural isotopic abundant lithium sample. In addition, the first measurement of the absolute photoionization cross-section of the 3p excited state of Li6 and Li7 are reported as 26.8±4 Mb and 25.5±3.8 Mb, respectively.

Active detection of small quantities of shielded highly-enriched uranium using low-dose 60-kev neutron interrogation

Active interrogation with low-energy neutrons provides a search technique for highly-enriched uranium concealed in cargo. We describe the technique and show initial results using a low-dose 60-keV neutron beam. This technique produces a clear induced fission signal in the presence of small quantities of 235U. The technique has been validated with low-Z and high-Z cargo materials. The technique uses a forward-directed beam of 60-keV neutrons to induce fission in 235U. Detection of the fast fission neutrons with pulse-shape discriminating scintillators is then the signature for 235U. The beam of neutrons is generated with a 1.93MeV proton beam impinging on a natural lithium target. The proton beam is produced by a radio-frequency quadrupole LINAC. The 60keV neutron beam is forward-directed because the 7Li(p,n) reaction is just above threshold for a proton energy of 1.93MeV.

Burning of Reactor Grade Plutonium Mixed with Thorium in a Hybrid Reactor

Significant amount of plutonium have been discharged and accumulated from the conventional LWRs and CANDU reactors. Reducing this reactor grade (RG) plutonium is very important because it may be misused and/or released accidentally into the environment. Fusion-fission (hybrid) reactors have strong potential on burning plutonium effectively. This study presents the burning of RG plutonium mixed with thorium in a hybrid reactor for an operation period of 24 months. The effect of various fuel mixtures (98% ThO2 + 2% RG-PuO2, 94% ThO2 + 6% RG-PuO2 and 90% ThO2 + 10% RG-PuO2) and coolants (Flinabe, natural lithium and Li20Sn80) on the reactor’s performance was investigated. Numerical results showed that utilization of RG plutonium in the mixed fuel in such a hybrid reactor not only enhanced the reactor’s performance but also reduced its 239Pu content significantly.

Resonantly enhanced three-photon excitation spectra of lithium

We present the first measurement on the resonantly enhanced three-photon excitation spectra of natural lithium using a Nd:YAG laser pumped dye laser in conjunction with a thermionic diode ion detector. Exploiting the linear and circular polarizations, the np2P3/2(8⩽n⩽11) and nf 2F7/2 (8⩽n⩽38) series have been observed via three-photon excitation from the ground state. The measured level energies reveal a dynamic shift from calculated values, which increases with an increase of the principal quantum number n. The ac stark shift and line broadening mechanisms are studied as a function of laser intensity. It is noted that the width increases and the line center shifts towards the higher energy side as the laser intensity is increased. The maximum observed shift for the 12f2F7/2 line is 0.33cm−1 corresponding to the laser intensity variation from 1.34×1012W/m2 to 1.03×1013W/m2, whereas its width increases from 0.36cm−1 to 0.82cm−1.

Investigation on tritium breeding capability of solid-liquid breeders in a (DT) fusion driven hybrid reactor

In design of deuterium-tritium (DT) fusion driven hybrid reactors, maintaining tritium self-sufficiency is very important to decrease operational costs. A (DT) fusion driven hybrid reactor must produce its own tritium to annihilate external investment on tritium production. Tritium Breeding Ratio (TBR) should be greater than 1.05 to maintain tritium self-sufficiency of the (DT) fusion plasma. Tritium is an artificial isotope that can be bred by using the breeding reactions of 6Li and 7Li isotopes. Therefore use and selection of lithium bearing coolants and blanket materials are very important with respect to tritium self-sufficiency of the (DT) fusion plasma. In this study, tritium breeding potential of various solid breeders Li2TiO3, Li8ZrO6, Li4SiO4, Li2ZrO3 or Li2O with different coolants Li2BeF4 (Flibe), NaF.LiF.BeF2 (Flinabe) or natural lithium in the hybrid blanket using different nuclear carbide fuels was investigated. Computations were carried out by using Scale4.3 solving Boltzmann transport equation. According to numerical results, among the solid breeders Li2O and Li8ZrO6 and among the liquid breeders, natural lithium had the highest contribution to TBR. Furthermore addition of uranium to thorium fuel improved tritium production.