6Li Enrichment Research Articles

Preliminary neutronic assessment of a helium-cooled Li8PbO6 breeding blanket design for DEMO

A preliminary neutronic assessment of the performances of a helium-cooled Li8PbO6 breeding blanket for the conceptual design of a DEMO fusion reactor is given. The study mainly focuses on TBR, power density responses and shielding factor optimization to estimate the feasibility of the design under the prescribed radiation deposition limits at TF-coils superconducting magnets. Computational analyses are based on three-dimensional 30° sector using the Monte Carlo code MCNPX 2.6. The scoping interest of helium-cooled Li8PbO6 blanket designs is based on a large potential minimization of the amount of Be required and the strong relaxation of 6Li enrichment requirements for this solution when compared to other solid breeder blanket options.

Pulse-shape analysis of Cs 2LiYCl 6:Ce scintillator for neutron and gamma-ray discrimination

Cs 2LiYCl 6:Ce (CLYC) is one of the most promising new scintillators for detecting both neutrons and gamma-rays. Its neutron and gamma-ray discrimination capability using pulse-shape analysis has drawn much attention, and there is significant interest in its use in field applications. For such applications, compact and low-power readout electronics capable of exploiting the pulse-shape discrimination (PSD) capabilities of CLYC will be essential. A readout system centered around a PSD-capable application specific integrated circuit (ASIC) that is well-suited for use with CLYC has been characterized, tested, and validated. As part of this study, automated analysis of CLYC data collected with a fast waveform digitizer extracted optimized charge integration windows for PSD. Additionally, several different CLYC samples were studied in order to gain understanding of the dependance of pulse shapes on parameters such as crystal size, 6Li enrichment level, crystal packaging, and choice of PMT. Extremely good PSD performance was obtained from CLYC scintillator and the ASIC-based readout system.

Fusion neutron flux of center hole in LiD assembly under CARR reactor

A LiD assembly was proposed to produce fusion neutrons from thermal neutrons in D2O region of the China Advanced Research Reactor (CARR). The transportation of produced fusion neutrons in the LiD assembly can obtain the fusion neutron flux in its central hole. The neutron flux and the distribution of 6Li(n,α) reaction rate in the LiD assembly were calculated by a computer simulation method with MCNP code. The calculation results show that the fusion neutron flux in the central hole of LiD assembly, Φ, is dependent on the size and structure of assembly, 6Li enrichment, tritium concentration in LiD rods and the physical state of LiD. In the case of LiD rods containing 5% tritium and 6Li enrichment 20%, Φ value reaches to 5.4×1013cm−2s−1, when the assembly was placed at the thermal neutron flux about 8.0×1014cm−2s−1 in CARR.

Preliminary studies of in-cell electrophoresis as 6Li enrichment technique

The development of a 6Li enrichment technique in terms of technical reliability and scalability to reach future 6Li requirement, represents a key step in the roadmap of fusion programmes worldwide. This paper presents the technique of liquid bath electrophoresis as one of the most promising method (simple, cheap and scalable to demands) for achieving the necessary enriched lithium production. The aim of this work is to carry out the process and conceive a prototype cell whose design allows the connection in series of several units in order to form a linear cascade. In this study we also present the possible materials for their construction (housing and electrodes) as well as the lithium ion conductors that can be used as electrophoretic membranes.

Basic technology for 6Li enrichment using an ionic-liquid impregnated organic membrane

The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 (6Li) in tritium breeding materials. However, natural Li contains only about 7.6at.% 6Li. In this paper, a new lithium isotope separation technique using an ionic-liquid impregnated organic membrane is proposed. In order to separate and concentrate lithium isotopes, only lithium ions are able to move through the membrane by electrodialysis between the cathode and the anode in lithium solutions. Preliminary experiments of lithium isotope separation were conducted using this phenomenon. Organic membranes impregnated with TMPA-TFSI and PP13-TFSI as ionic liquids were prepared, and the relationship between the 6Li separation coefficient and the applied electrodialytic conditions was evaluated using them. The results showed that the 6Li isotope separation coefficient in this method (about 1.1–1.4) was larger than that in the mercury amalgam method (about 1.06).

Neutronics analysis of the conceptual design of a component test facility based on the spherical tokamak

One of the crucial aspects of fusion research is the optimisation and qualification of suitable materials and components. To enable the design and construction of DEMO in the future, ITER is taken to demonstrate the scientific and technological feasibility and IFMIF will provide rigorous testing of small material samples. Meanwhile, a dedicated, small-scale components testing facility (CTF) is proposed to complement and extend the functions of ITER and IFMIF and operate in association with DEMO so as to reduce the risk of delays during this phase of fusion power development. The design of a spherical tokamak (ST)-based CTF is being developed which offers many advantages over conventional machines, including lower tritium consumption, easier maintenance, and a compact assembly. The neutronics analysis of this system is presented here. Based on a three-dimensional neutronics model generated by the interface programme MCAM from CAD models, a series of nuclear and radiation protection analyses were carried out using the MCNP code and FENDL2.1 nuclear data library to assess the current design and guide its development if needed. The nuclear analyses addresses key neutronics issues such as the neutron wall loading (NWL) profile, nuclear heat loads, and radiation damage to the coil insulation and to structural components, particularly the stainless steel vessel wall close to the NBI ports where shielding is limited. The shielding of the divertor coil and the internal Poloidal Field (PF) coil, which is introduced in the expanded divertor design, are optimised to reduce their radiation damage. The preliminary results show that the peak radiation damage to the structure of martensitic/ferritic steel is about 29 dpa at the mid-plane assuming a life of 12 years at a duty factor 33%, which is much lower than its ∼150 dpa limit. In addition, TBMs installed in 8 mid-plane ports and 6 lower ports, and 60% 6Li enrichment in the Li 4SiO 4 breeder, the total tritium generation is ∼3.7 kg in the whole lifetime of CTF.

Parameter study of the LIFE engine nuclear design

LLNL is developing the nuclear fusion based Laser Inertial Fusion Energy (LIFE) power plant concept. The baseline design uses a depleted uranium (DU) fission fuel blanket with a flowing molten salt coolant (flibe) that also breeds the tritium needed to sustain the fusion energy source. Indirect drive targets, similar to those that will be demonstrated on the National Ignition Facility (NIF), are ignited at ∼13 Hz providing a 500 MW fusion source. The DU is in the form of a uranium oxycarbide kernel in modified TRISO-like fuel particles distributed in a carbon matrix forming 2-cm-diameter pebbles. The thermal power is held at 2000 MW by continuously varying the 6Li enrichment in the coolants. There are many options to be considered in the engine design including target yield, U-to-C ratio in the fuel, fission blanket thickness, etc. Here we report results of design variations and compare them in terms of various figures of merit such as time to reach a desired burnup, full-power years of operation, time and maximum burnup at power ramp-down and the overall balance of plant utilization.

Estimation of Neutronic Performance in a Hybrid Reactor with Regression Analysis

This study presents regression analysis method used for prediction and investigation of neutronic performance in a hybrid reactor using UO2 fuel and Flibe (Li2BeF4) coolant. The 235U fraction is increased gradually from 0 to 4% stepped by 1% and the 6Li fraction within the Flibe coolant is enriched gradually to 30, 60 and 90% from 7.5%. Relations between 235U fuel fraction and lithium (6Li) enrichment are investigated for the estimation of neutronic performance as the tritium breeding ratio (TBR), energy multiplication factor (M), total fission rate (Σf), 238U (n,γ) reaction and fissile fuel breeding (FFB) in the hybrid reactor. Regression analysis by results obtained by using the code (XSDRNPM/SCALE5) for TBR, M, Σf, 238U (n,γ) and FFB are performed. The results of the regression analysis and the values obtained by using the code (XSDRNPM/SCALE5) are compared with respect to the TBR, M, Σf, 238U (n,γ) and FFB of the reactor. The values calculated from the obtained formulations with regression analysis are found to be in good agreement with results obtained by using the code (XSDRNPM/SCALE5). It is observed that the derived equations from regression analysis could provide an accurate computation of the neutronic performances so that these equations could use for the prediction of TBR, M, Σf, 238U (n,γ) and FFB. In addition, correlation matrix is calculated to determine the degree of relationship between variables as TBR, M, Σf, 238U (n,γ) and FFB.

Measurement system for in-pile tritium monitoring from Li 2TiO 3 ceramics at WWRK reactor

A 220-day irradiation of Li 2TiO 3 ceramics with 96% enrichment of isotope 6Li was carried out at WWRK reactor. One of the study goals was to examine tritium release behavior during Li burn-up. To achieve this goal three types of ceramics samples were examined simultaneously using a system for in-pile tritium monitoring: one (pebbles) – under constant temperature of 650 °C, and two (pebbles and pellets) – within temperature change ranges from 500 to 900 °C. Flows of tritium release from ceramics during the various reactor campaigns, as well as tritium generation rates for each ampoule are presented in the paper. The main result is justification of burn-up in 6Li reaching up to 20% of Li 2TiO 3 ceramics.

Laser isotope separation of lithium by two-step photoionization

Lithium isotope separation has been achieved employing the two-step photoionization technique along with a narrow band dye laser in conjunction with a time of flight mass spectrometer. The demonstrated method yields a high degree of selectivity by tuning the dye laser at the resonance levels of Li6 and Li7. It is inferred that the concentration of the natural abundance of the Li6 isotope gets enhanced up to over 47% as the exciter dye laser is tuned to the P1∕22 of Li6 even if the linewidth of the exciter laser is not sufficiently narrow to excite the isotopic level. It is also noticed that the much higher energy density of the exciter laser limits the resolution of the fine structure levels of the lithium isotopes that leads to a loss in the enrichment of Li6 due to the power-broadening effect. Measurements of the photoionization cross section of the lithium isotopes from the 2pP1∕2,3∕22, excited states for Li6 and Li7 and the corresponding number densities are reported.

An ESR and ENDOR study of irradiated 6Li-formate

Lithium formate ( 6LiOOCH·H 2O), 95% 6Li enrichment, combined with an exchange of crystallization water with D 2O was investigated. The ESR spectrum of the radiation induced free radicals stable at room temperature consists of a singlet with a narrow line width, 0.92 mT. 6Li has smaller magnetic moment and nuclear spin, which resulted in the narrower line width accompanied with an increase in peak amplitude. In comparison with lithium formate with natural isotopic composition, 6Li (7.5%, I = 1) and 7Li (92.5%, I = 3/2), the sensitivity was increased by a factor of two. With optimised spectrometer settings 6Li formate had seven times higher sensitivity compared to alanine. Therefore this material is proposed as a dosimeter material in a dose range down to 0.1 Gy. The g and the 13C-hyperfine (hf) tensors of the CO 2 − radical anion, major paramagnetic products, were evaluated to be g = (2.0037, 1.9975, 2.0017), and A( 13C) = (465.5, 447.5, 581.3) MHz for polycrystalline samples at room temperature. Furthermore, the 1H-hf and 6Li-hf tensors observed for the surroundings of CO 2 − by ENDOR technique were in fairly good agreement with DFT calculations. The CO 2 − radicals are found to be so stable that the formate is applicable to the ESR dosimetry, because of fully relaxing in a fully relaxed geometrical structure of the CO 2 − component and remaining tight binding with the surroundings after the H atom detachment from HCO 2 −.

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a 6Li enrichment of 90%. Assuming a 6Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.

Nuclear analyses for two “look-alike” helium-cooled pebble bed test blanket sub-modules proposed by the US for testing in ITER

The US is proposing two “look-alike” sub-modules, based on helium-cooled pebble bed (HCPB) ceramic breeder, to be tested in the same test blanket module (TBM) that will occupy a quarter of a port in ITER and placed next to the Japanese TBM. The TBM has a toroidal width of 73 cm, a radial depth of 60 cm and a poloidal height of 91 cm. The ceramic breeder is made of Li 4SiO 4 with 75% Li-6 enrichment (60% packing factor) and beryllium is used as the multiplier. The two sub-modules are arranged in two configurations, namely a layered configuration and an edge-on configuration. In the present work, we analyze these two sub-modules using two-dimensional discrete ordinates transport codes in R-θ model that accounts for the presence of the ITER shielding blanket and the surrounding frame of the port. The objectives are: (1) to examine the profiles of heating and tritium production rates in the two sub-modules, both in the radial and toroidal direction, in order to identify locations where neutronics measurements can be best performed with least perturbation from the surroundings, (2) to provide both local and integrated values for nuclear heating rates required for subsequent thermo-mechanics analysis, and (3) to compare the tritium production capabilities of two variants for the HCPB blanket concept, mainly the parallel and the edge-on configurations. We present the main findings from this study in this paper.

Evaluation of contact strength of Li 2TiO 3 pebbles with different diameters

Lithium titanate (Li 2TiO 3) pebbles are considered to be a candidate material of the tritium breeders for fusion reactor from viewpoints of good tritium recovery, chemical stability and other good properties. The pebbles of 0.3–2 mm in diameter have been fabricated by a wet process and its characterization have been performed. Mechanical properties such as compressive strength are necessary for the design of fusion blanket packing the small pebbles. In this study, collapse loads of Li 2TiO 3 pebbles with different diameters were measured by an unconfined compression tester, and contact strength of these pebbles was evaluated by Hertzian contact theory. In the results, the contact strength of Li 2TiO 3 pebbles decreased with increasing the radius. The radial strength of Li 2TiO 3 pebbles was about 1.7 times large as the evaluated compression strength. On the other hand, the contact strength of Li 2TiO 3 pebbles was also independent of 6Li enrichment.

Neutronic design optimisation of modular HCPB blankets for fusion power reactors

Neutronic design optimisation analyses based on three-dimensional Monte Carlo calculations have been conducted with the MCNP code as part of the development work for the modular helium cooled pebble bed (HCPB) blanket for a fusion power reactor of the model B-type of the European power plant conceptual study (PPCS). Design variants with breeder pebble beds parallel and perpendicular to the first wall have been investigated with the objective to assess and optimise their tritium breeding efficiency. Although both configurations are feasible, the latter design solution was shown to provide a superior performance in terms of the required blanket thickness, the beryllium inventory and the 6Li enrichment due to a better utilisation of the space available for the breeder and the beryllium pebbles.

A neutronic investigation of He-cooled liquid Li-breeder blankets for fusion power reactor

Configuration of a He-cooled liquid lithium blanket has been optimized in terms of the tritium breeding ratio (TBR) and the neutron shielding. For a model blanket, fundamental neutronic characteristics are analyzed with a Monte Carlo code. To improve the TBR, various neutron reflectors are compared, which are graphite, TiC and ZrH 2, and the optimal application of the reflector is investigated, either lumped or layered. The TBR is analyzed as a function of the Li-6 enrichment for the reflected blanket, and the results are used in determining the lifetime of a natural lithium breeder. In addition, various types of moderated shields are assessed to improve the neutron shielding performance.

Neutronic analysis for the EU water-cooled lithium–lead test blanket module in ITER-FEAT

Comprehensive neutronic analyses using the MCNP Monte Carlo code have been performed to predict the neutronics performance of the EU water-cooled lithium–lead test blanket module (TBM) integrated in ITER-FEAT. The analyses have been performed using models representing the complex ITER and TBM structure close to reality. Tritium generation and parameters relevant to the lifetime performance of the TBM such as helium, hydrogen and atomic displacement production have been calculated. Assuming 90% 6Li enrichment and 22% duty cycle, the estimated tritium production amounts to 3.8738 × 1016 T atoms s−1 (16.62 mg d−1). The maximum radiation damage obtained, based on 0.3 MWa m−2 neutron fluence, is in the first wall first steel layer of the breeder container totalling 2.77 dpa of Fe in Euro steel, 19.3 appm He production and 103 appm H generation. The calculated helium production in the demountable hydraulic connections of TBM is below the required limit for reweldability of steel structure.

Numerical Study on Effects of Fuel Mixture Fraction and Li-6 Enrichment on Neutronic Parameters of a Fusion–Fission Hybrid Reactor

This study presents the effects of mixture fractions of nuclear fuels (mixture of fissile–fertile fuels and mixture of two different fertile fuels) and 6Li enrichment on the neutronic parameters (the tritium breeding ratio, TBR, the fission rate, FR, the energy multiplication ratio, M, the fissile breeding rate, FBR, the neutron leakage out of blanket, L, and the peak-to-average fission power density ratio, Γ) of a deuterium–tritium (D–T) fusion neutron-driven hybrid blanket. Three different fertile fuels (232Th, 238U and 244Cm), and one fissile fuel (235U) were selected as the nuclear fuel. Two different coolants (pressurized helium and natural lithium) were used for the nuclear heat transfer out of the fuel zone (FZ). The Boltzmann transport equation was solved numerically for obtaining the neutronic parameters with the help of the neutron transport code XSDRNPM/SCALE4.4a. In addition, these calculations were performed by also using the MCNP4B code. The sub-limits of the mixture fractions and 6Li enrichment were determined for the tritium self-sufficiency. The considered hybrid reactor can be operated in a self-sufficiency mode in the cases with the fuel mixtures mixed with a fraction of equal to or greater than these sub-limits. Furthermore, the numerical results show that the fissile fuel breeding and fission potentials of the blankets with the helium coolant are higher than with the lithium coolant.

Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

Abstract At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation.

Neutronic Comparison of Tritium-Breeding Performance of Candidate Tritium-Breeding Materials

Tritium self-sustainment, which will meet the fuel requirement of fusion reactor, is one of the key issues of fusion power development. The tritium breeding performances of various tritium-breeding materials are compared based on a series of neutronics calculations using three-dimensional Monte Carlo neutron-photon transport code MCNP/4C with the IAEA FENDL-2 data library. The effects of the dimensions of the tritium-breeding zone and the enrichment of 6Li on Tritium Breeding Ratio (TBR) are analyzed. The effects of Be as a neutron multiplier on TBR are also calculated.