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Fast neutron irradiation has the capability of transforming fertile material, such as 238U or 232Th, into fissile material that can be used to manufacture fuel for conventional thermal light-water fission reactors. This would help both expand and extend the lifetime of nuclear power, which is currently constrained to natural uranium resources, which could only last between 50 and 150 years depending on the growth of the nuclear reactor fleet. In this paper, a design for a fast neutron source based on a low-aspect ratio tokamak is evaluated, based on two figures of merit: the tritium breeding ratio (TBR), a measure of tritium self-sufficiency, and the reactor support ratio (RSR), related to the speed of fissile material breeding. The system has neutron multiplying regions to ensure there are enough neutrons to breed both tritium and fissile material; it also has tritium breeding regions that contain 6Li which, when bombarded with neutrons, produce tritium. The MCNP code is used to perform neutronics simulations and obtain values for energy-resolved neutron fluxes for the relevant regions of the system, and the ORIGEN code is used to perform kinetic calculations for nuclear reactions between the neutrons and the materials within the regions. Simulations show that for the case of a spherical tokamak with aspect ratio of 1.8 at 120 MW of neutron power, the tritium breeding ratio can have values ranging 0.85 and 1.25, and with that power level the system can bring 336 assemblies to 5% 233U enrichment in irradiation times between 24 and 65 months. The values depend on both the selection of materials and key geometry parameters, such as the thickness of container walls, as well as the volumes for both neutron multiplying and tritium breeding regions. FLiBe was found to give the best performance as neutron multiplier, since it contributes to the tritium self-sufficiency as well; from the tritium breeding perspective, lithium oxide was found to be the best choice.

Molten-salt blankets that possess tritium-absorbing metal particles are promising emerging technical components in nuclear fusion reactors. In this study, we develop a numerical model and carry out a systematic analysis of the electromagnetic induction heating of metal particles (Ti, Pd, Mg, Zr, and V) in the molten salt for the extraction of tritium. We show that the maximum absorption power in the metal particles can be normalized in a form independent of the material combination of nonmagnetic metals and salts, and the peak power density per square of magnetic flux density per field frequency is 5.3 × 106 W m–3 T–2 s. Nevertheless, the steady-state temperature difference between the metal particles and the molten salt, a figure of merit of the selective electromagnetic heating scheme, is found to monotonically increase with the metal particle size, in contrast to the behavior of the absorbed power density, thus encouraging the use of larger metal particles. The attainable temperature difference between the Ti particles and the FLiBe blanket is estimated to be 110 °C for a particle diameter of 1 mm in a 2.45 GHz and 1 mT magnetic field, and it increases proportionally with the diameter, square root of frequency, and square of magnetic flux density around this condition.

Fluoride molten salt Flibe (2LiF + BeF2) is a promising candidate for the liquid blanket of a nuclear fusion reactor, because of its large advantages of tritium breeding ratio and heat-transfer fluid. Since its melting point is higher than other liquid candidates, another new fluoride molten salt Flinabe (LiF + NaF + BeF2) is recently focused on because of its lower melting point while holding proper breeding properties. In this experiment, hydrogen permeation behavior through the three molten salts of Flibe (2LiF + BeF2), Fnabe (NaF + BeF2) and Flinabe are investigated in order to clarify the effects of their compositions on hydrogen transfer properties. After making up any of the three molten salts and purifying it using HF, hydrogen permeability, diffusivity and solubility of the molten salts are determined experimentally by using a system composed of tertiary cylindrical tubes. Close agreement is obtained between experimental data and analytical solutions. H2 permeability, diffusivity and solubility are correlated as a function of temperature and are compared among the three molten salts.

It is important to investigate the characteristics of thermofluids for cooling blankets and divertors used in the high-temperature plasma of fusion reactors. Various coolant materials on the cooling problem such as a fusion reactor were widely considered from FLiBe to lithium. In the present study, we proposed pressurized water as the working fluid for the blanket design. We performed direct numerical simulation of pipe flow with Reynolds number of 2100 based on the pipe radius and friction velocity. The number of mesh points used were 4096 × 1024 × 1532 in the z−, r−, and φ-directions, respectively. The Prandtl number of pressurized water was set as 0.87. The results show that compared with air as a general coolant material, pressurized water has higher turbulent heat transfer.

Fusion-fission hybrid reactor (FFHR), a representative kind of subcritical nuclear reactor, has the advantage of better safety margin. There are less constraints for the loading of active fuel when burning spent nuclear fuel. A simplified three-dimensional model of tokamak blanket is constructed by the Serpent Monte Carlo Method Code to investigate the MA transmutation characteristics in the Flibe molten salt blanket. Based on the model with one homogenous fuel region, the effect of key design parameters: MA content, plutonium content, Li-6 enrichment in the initial fuel were determined, during which performance characteristics such as MA transmutation amount, keff, tritium breeding ratio (TBR), blanket energy multiplication factor M and conversion ratio were obtained. The results show that it is hard to improve the tritium breeding and meanwhile enhance the MA reducing in one homogenous fuel region. To balance these two functions, a double molten salt fuel regions model with separative MA transmutation zone and tritium breeding zone was proposed.

Nowadays, Fusion Energy is one of the most important sources under study. During the last years, different designs of fusion reactors were considered. At the MIT, an innovative design was created: ARC, the Affordable Robust Compact reactor. It takes advantage of the innovative aspects of recent progress in fusion technology, such as high temperature superconductors, that permit to decrease the dimensions of the machine, reaching at the same time high magnetic fields. Our main goal is the low-activation analysis of possible structural materials for the vacuum vessel, which is designed as a single-piece placed between the first-wall and the tank that contains the breeding blanket. Due to its position, the vacuum vessel is subject to high neutron flux, which can activate it and cause the reduction of the component lifetime and decommissioning problems. The activation analysis was done also for the liquid breeder FLiBe, compared with Lithium-Lead. Codes used for the low-activation analysis were MCNP and FISPACT-II. The first one is based on a neutronics model and for each component a certain neutron flux is evaluated. For FISPACT-II, the main input is the composition of the analyzed material, the neutron flux and the irradiation time. Results from FISPACT-II are the time behavior of specific activity, contact dose rate. To assess suitable structural materials for the vacuum vessel, low-activation properties were considered. Vanadium alloys turn out to be one of the best alternatives to the present material, Inconel-718. Finally, isotopic tailoring and elemental substitution methods were applied. Here, the composition of each alloy is analyzed and critical isotopes or elements are eliminated or reduced. After the modifications, new simulations are done, and those leading to significant improvements in the final results are highlighted.

A sphere-packed pipe has been proposed as a heat transfer promoter for the first wall cooling in a Flibe blanket. In this study, the flow field in a sphere-packed pipe was well investigated by means of two-dimensional PIV method by matching refractive index of a channel material and working fluid. Three-dimensional flow structure was clarified by integrating the obtained data. The feature of the flow was tortuous high-velocity region formed near pebbles and large velocity fluctuation in the vicinity of the channel wall. And, to apply this flow structure to the actual first wall cooling, a new cooling system using finger-stacked structure was proposed and discussed.

Molten-salt used as a coolant in fusion reactors plays a significant role in the design of advanced reactors. Investigation of thermal behavior is necessary in an actual environment of a facility where heat transfer enhancement takes place under a high Pr number fluid flow such as in case of FLiBe. For the development of a facility, it is necessary to be able to monitor fluid motion of a basic heat transfer promoter such as a sphere-packed pipe (SPP). In the present study, to discern the complex flow structures in SPP, digital holographic PTV visualization is carried out by a refractive index-matching method using a sodium iodide (NaI) solution employed as a working fluid. Hologram fringe images of particles behind the spheres can be observed, and the particles’ positions can be reconstructed by a digital hologram. Consequently, 3-D velocity-fields around the spheres are obtained by the reconstructed particles’ positions. The velocity between pebbles is found to be faster than that in other regions.

A high-β spheromak reactor concept has been formulated with an estimated overnight capital cost that is competitive with conventional power sources. This reactor concept utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt (FLiBe) blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER-developed cryogenic pumping systems were implemented in this concept from the basis of technological feasibility. A tritium breeding ratio (TBR) of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%.

The inboard shielding of a fusion reactor can be a crucial issue due to the limited space available in a tokamak configuration. It is necessary to assess the inboard shielding capability of DEMO for its initial design. In this paper, 1D and 3D neutronics models were developed based on a reference design of the Chinese Fusion Engineering Testing Reactor (CFETR). The neutron wall load (NWL) is in the range of 1.5–3 MW/m2 and the inboard shielding thickness is constrained within 40–70 cm in order to achieve the tritium self-sufficiency of the reactor. Referring to the detailed design of the ITER Toroidal Field Coils (TFCs) and using radiation hardening technology developed for ITER, the inboard blanket shielding capability and nuclear responses of the TFC are investigated for both FLiBe and Li4SiO4 breeding blanket concepts. The impact of the gaps on shielding performance is discussed. Some suggestions on improving the inboard shielding performance for DEMO are also proposed.

A Flibe blanket has been proposed to be used in FFHR. Since Flibe has poor heat transfer performance, heat transfer promoter is required, and a sphere-packed pipe (SPP) has been proposed to enhance the heat transfer performance in the Flibe blanket. In this paper, the fluid flow and heat transfer characteristics in the SPP is evaluated numerically using a k–ɛ turbulent model for the flow field and an algebraic model for the thermal field. As a result, it was shown that bypass flows in the SPP play a significant role in heat transfer. Also it is thought that the turbulent energy can strongly affect heat transfer performance.

A key requirement for DEMO is the on-site breeding of tritium. In order to do this, a robust control system must be employed to ensure enough tritium is being bred to sustain the fusion reactor, whilst not breeding an amount which would exceed the plant's tritium inventory license. A tritium breeding method which is cost effective and reduces radioactive waste for disposal is that of the liquid metal breeder such as those based around LiPB and FLiBe. This paper focuses on the modeling of a simplified fusion reactor design with a LiPb blanket with linked radiation transport, nuclide burn-up and control theory. Two simple models were simulated using the FATI code which incorporated a PID (proportional integral derivative) controller that adjusted the Li6/Li7 ratio in order to increase/decrease tritium production based on the difference between the measured excess tritium inventory and the desired excess inventory. The modelling has initially demonstrated that a linear PID controller has the capability to manage tritium production within a LiPb liquid blanket.

W coating by vacuum plasma spray process and Cr coating by chromizing process were performed on fusion low activation materials, JLF-1 ferritic steel and NIFS-HEAT-2 vanadium alloy. The present study discusses feasibility of the coatings as anti-corrosion coating against fluoridation in Flibe for fusion low activation materials. Coatings were characterized by microstructural analysis and examination on chemical stability by corrosion tests. The corrosion tests were conducted with H2O–47% HF solution at RT and He–1% HF–0.06 H2O gas mixture at 823 K to simulate fluoridation and oxidation in Flibe. The coatings presented suppression of fluoride formation compared with JLF-1 or NIFS-HEAT-2, however weight loss due to WF6 formation was induced, and much Cr2O3 was formed.

In Japan, development of Water Cooled Ceramic Breeder (WCCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) toward DEMO. Also development of high temperature Li–Pb blanket, Li/V blanket and Flibe blanket are being performed toward the advanced DEMO blanket. Regarding the WCCB TBM development, real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS), F82H and successfully tested by heat flux equivalent to ITER TBM condition, 0.5 MW/m 2 , 80 cycles with the same coolant condition as DEMO. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure are successfully fabricated. Furthermore, asemmbling test of the real scale FW plate mockup and SW plate mockup was performed to clarify the welding condition to form TBM box structure. All key technologies of blanket module fabrication has been addressed and almost achieved toward TBM. Regarding the development of liquid breeder blankets, all key technologies, such as, material compatibility and mass transfer, tritium recovery performance, insulation coating development, were covered, by using elementary experiments to convection loops of Li–Pb, Li and Flibe. High temperature heat exchange test up to 900 °C has been initiated by using Li–Pb test loop.

Transport of tritium (T) and heat is calculated to estimate the performance of liquid Li17Pb83 (Li–Pb) or Li2BeF4 (Flibe) as a T-breeder in a fusion reactor blanket. T is bred in such a way of low leak to facilities outside and continuous recovery by a removal system, and heat is transferred through structural walls to He coolant efficiently. In this paper, T permeation in a blanket composed of structural materials and liquid Li–Pb or Flibe is calculated based on data of previous experiment. The effects of T recovery ratio by the outside removal apparatus and fluid-film T diffusion resistance in the liquid blanket on overall T permeation rates are analytically clarified. Design of a liquid blanket with low T leak and high T recovery is discussed here. In addition, possibility in that microbubbles may be generated at interfaces between a liquid blanket and a structural wall is investigated.

Static corrosion tests were performed in molten salts, LiF–BeF2 (Flibe) and LiF–NaF–KF (Flinak), at 500 °C and 600 °C for 1000 h. The purpose is to investigate the corrosion characteristics of reduced activation ferritic steels, JLF-1 (8.92Cr–2W) in the fluids. The concentration of hydrogen fluoride (HF) in the fluids was measured by slurry pH titration method before and after the exposure. The HF concentration determined the fluoridation potential. The corrosion was mainly caused by dissolution of Fe and Cr into the fluids due to fluoridation and/or electrochemical corrosion. Carbon on the surface might be dissolved into the fluids due to the corrosion, and this resulted to the decrease of carbide on the surface. The corrosion depth of the JLF-1 specimen, which was obtained from the weight losses, was 0.637 μm in Flibe at 600 °C and 6.73 μm in Flinak at 600 °C.

This paper reports on the experimental investigation of tritium permeation in flibe (2LiF–BeF2) at the Safety and Tritium Applied Research facility of the Idaho National Laboratory. A stainless steel cell formed by two independent volumes separated by a 2-mm thick nickel membrane is maintained at temperatures between 500 and 700 °C. A controlled amount of T2 gas is flown in excess of argon in the source volume in contact with the bottom side of the nickel membrane, while a layer of molten salt is in contact with the top side. The tritium permeating above the liquid surface is carried by an argon flow to a diagnostic system comprised of a quadrupole mass spectrometer, a gas chromatographer and a proportional counter. Tritium permeability in flibe as a function of temperature is determined by the measured permeation flow rates reached in steady-state conditions, while the diffusivity is determined by fitting the transient process with the analytical solution for the diffusion process. As a result, the solubility of tritium in flibe as a function of temperature is also determined.

The best material for a wet wall and blanket of an inertial fusion reactor is selected among Li, eutectic alloys of Li–Pb, Li–Sn and a 2LiF + BeF2 molten salt mixture called Flibe, judging from their chemical, nuclear and heat-transfer properties. Li0.17Pb0.83 is found to be the most promising one because of low Li vapor pressure, moderate melting temperature, good heat-transfer properties under the condition of a KOYO-fast circulation loop operated between 300 and 500 °C. A counter-current extraction tower packed with metallic rashig rings is proposed to extract tritium generated and dissolved in the Li–Pb eutectic alloy. Mass-transfer parameters when He and Li–Pb flow counter-currently through the tower packed with the rings are determined to satisfy the two necessary conditions of a self-sufficient tritium generation rate of 1.8 MCi/day and a target tritium leak rate of 10 Ci/day. It is found that the height of a tower to achieve the 99.999% recovery is comparatively low because of the promising property of a large equilibrium pressure of tritium. In order to mitigate the disadvantage of its high density, which needs a large pumping power, a porous packing material that keeps good contact between He and Li–Pb should be developed in the future. It is found experimentally that D2 addition in He purge gas is effective to achieve a faster rate of tritium recovery from the Li–Pb flow. The rate-determining step of tritium permeation through a steam generator is determined as a function of a Li–Pb flow rate in a stainless-steel heat-transfer tube.

The control of tritium is essential for the performance of liquid blankets of fusion reactors. On-line hydrogen (isotopes) measurement is a key technology. The on-line hydrogen sensor made of proton conducting ceramics, CaZr0.95O3−aSc0.05, was designed to be used in reducible condition such as liquid blanket system. The CaZrO3 doped with Sc2O3 has higher chemical stability than those with the other dopant oxides such as In2O3 or Ga2O3. The evaluation of the expected performance of the sensor in molten salt LiF–BeF2 (Flibe), liquid metal lithium (Li) and lead–17 lithium (Pb–17Li) was carried out by means of the performance test in gas atmosphere at 700 °C with hydrogen partial pressures equivalent to those for the melts. The sensor performance in liquid metal with low oxygen potential was investigated by means of the test in molten Al. The sensor exhibited stable output and fast response when the hydrogen concentration in the gas or molten aluminum was changed.

An investigation of MHD effects on a Flibe (Li 2 BeF 4 ) simulant fluid has been conducted under the U.S.–Japan JUPITER-II collaboration program using the “FLIHY” pipe flow facility at UCLA. The present paper reports experimental results on turbulent pipe flow of an aqueous potassium hydroxide solution under magnetic field using particle image velocimetry (PIV) technique. The modification of turbulence was investigated by comparison of the experimental results with a direct numerical simulation (DNS) data base. The PIV measurements at Re = 11,300 were performed with variable Hartmann numbers, and the modification of the mean flow velocity as well as turbulence reduction was observed. A flat velocity profile in the pipe center and a steep velocity gradient in the near-wall region exhibit typical characteristics of wall-bounded MHD flows. The DNS was performed approximately the same conditions and the comparison of turbulence statistics between PIV and DNS shows good agreement for up to Ha = 10.

A theoretical model is developed to predict “rough estimates” of the isochors of the radiative cooling time of the high-temperature partially ionized Flibe gas at conditions relevant to the inertial fusion energy (IFE) chamber conditions. The model embodies the calculation of the occupational densities of all plasma species, calculation of the gas specific heat at constant volume, c v along with the gas opacity/emissivity and radiative cooling surface based on the National Institute of Standards and Technology (NIST) extensive compilation of atomic and spectroscopic data. Although the problem of calculating radiative cooling in IFE environment is a very complicated multi-physics and multi-scale problem, which includes self-consistent solution of the magneto-hydrodynamics, radiation transport (possibly in time-dependent formulation), plasma kinetic, and atomic systems of equations, the rough estimates presented herein will be beneficial for benchmarking and engineering purposes with some carefulness.

In order to evaluate neutron activation of materials relevant to the liquid blankets such as Lithium/vanadium-alloy and Flibe/vanadium-alloy blankets, irradiation experiments have been performed using the fusion neutronics source (FNS) at JAEA. Specimens of Er, Teflon (CF 2 CF 2 ) and NIFS-HEAT-2 (V–4Cr–4Ti) were selected in the experiments for the evaluations of the activation of Er in the MHD coating (Er 2 O 3 ) , F in Flibe (LiF–BeF 2 ) and V, Cr, Ti in the structural materials, respectively. The irradiation experiments were performed with two kinds of neutron field: D–T neutrons and Be moderated neutrons for the purpose of reaction with high and low energy neutrons, respectively. The neutron spectra calculations were carried out using Monte Carlo transport codes MCNP-4C and JENDL3.3 files. The neutron-induced activity measured with a high purity Ge detector was compared with calculations carried out by the FISPACT-2001 codes and EAF-2001 files. The results showed that the calculated activities of most of products through reaction with high energy neutrons are in agreement with the experiments within 20% uncertainty.

The second Japan/US Program on Irradiation Tests for Fusion Research (JUPITER-II) began on April 1, 2001. Part of the collaborative research centers on studies of the molten salt 2LiF2–BeF2 (also known as Flibe) for fusion applications. Flibe has been proposed as a self-cooled breeder in both magnetic and inertial fusion power plant designs over the last 25 years. The key feasibility issues associated with the use of Flibe are the corrosion of structural material by the molten salt, tritium behavior and control in the molten salt blanket system, and safe handling practices and releases from Flibe during an accidental spill. These issues are all being addressed under the JUPITER-II program at the Idaho National Laboratory in the Safety and Tritium Applied Research (STAR) facility. In this paper, we review the program to date in the area of tritium/deuterium behavior, Flibe mobilization under accident conditions and testing of Be as a redox agent to control corrosion. Future activities planned through the end of the collaboration are also presented.

We investigate the behavior of tritium in liquid Flibe (Li2BeF4) with molecular dynamics techniques. A major objective of the work is to estimate the diffusion coefficient of tritium. Preliminary results are reported in this paper. Ab initio molecular dynamics calculations yielded ionic distances of the molten salt in good agreement with values from the literature. A tritium ion arbitrarily introduced into the molten salt sample stayed bonded to a fluorine ion for the entire simulation run and BeF 4 2 − molecules were observed as a result of the calculation. The extracted diffusion coefficients from mean square displacements of Li, F, and T were lower than experimental values found in the literature.

To understand the interactions between molten Flibe and Be, a metallic Be rod was immersed in molten Flibe at 803 K under He atmosphere for 210 h. The Be rod was significantly eroded during immersion in molten Flibe, and the Flibe changed from a clear crystal to a brownish-gray marble-like appearance. The concentration of Be 0 in Flibe was evaluated by dissolving salt samples in acid solutions. This dissolution test is based on the reaction of Be 0 with proton ions in the acid solutions to generate H 2 . Hydrogen gas was generated from Flibe contacted with Be under flowing He. The amounts of gas generated corresponded to mole fractions of [Be 0 ]/[Flibe] ranging from 9.9 × 10 −4 to 7.6 × 10 −3 . On the other hand, no H 2 was generated from Flibe that was not exposed to Be or which was exposed to Be and then given significant H 2 –HF–He bubbling. These observations showed that Be does, indeed, dissolve in Flibe as Be 0 . The fact that no Be 0 was detected after bubbling H 2 –HF–He into the salt indicates that Be 0 is an effective redox agent for reacting HF.

The application of Flibe (2LiF + BeF2 mixed molten salt) to a blanket material of force-free helical reactor (FFHR-2) was discussed based on our recent experiments on redox control of Flibe by Be and on hydrogen permeation through Flinak (0.465LiF + 0.115NaF + 0.42 KF mixed molten salt) as a fluid simulant of Flibe. The two experimental works were carried out under the collaboration work of JUPITER-II. To maintain Flibe in the blanket loop under reduction condition is a key issue to transform TF to T2 with a fast reaction rate. In order to achieve that, proper rates of Be dissolution and reduction reaction of TF to T2 in Flibe have to be established in the self-cooled blanket. The overall material balance of tritium in the FFHR-2 system was discussed in terms of the tritium permeation through Flibe and the tritium recovery rate.

The reduced activation ferritic steel (RAFS) has been selected as structural material for a variety of blanket systems for ITER test blanket modules (TBM). In the evaluation of integrated performance of ferritic steels as structural components of blanket systems, there are unique issues as well as common issues for each blanket system. One of the unique issues for each system is the compatibility of ferritic steels with the coolant materials. The corrosion rate of ferritic steels in hot water, super critical pressurized water (SCPW), humid air, Pb–17Li, lithium and Flibe at various temperatures is reviewed in this work. Efforts to improve corrosion resistance have been made, taking the alloy design into account. A dispersion of yttria was effective to improve corrosion resistance of a RAFS. The compatibilities of RAFSs with hot water, Pb–17Li, lithium and Flibe are considered to be good enough for the TBM applications. The liquid metal embrittlement (LME) is considered to be a critical issue for the utilization of RAFSs for the lithium systems. Several issues towards DEMO and beyond are shown from the compatibility point of view.

Neutronics calculations were carried out to characterize self-cooled Li and Flibe blankets with vanadium alloy (V–4Cr–4Ti) structure, as advanced options for the Force Free Helical Reactor (FFHR-2). In the present paper, the parameter dependence of Tritium Breeding Ratio (TBR) and neutron shield performance was examined for a Li blanket (FFHR-LV) and a Flibe blanket (FFHR-FV). The TBR was independent from the Flibe composition between 30 and 45 mol% of BeF 2 in the case of the FFHR-FV. Due to the increase in the temperature range of the coolant Flibe as a result of the change of the structural materials from JLF-1 (low activation ferritic steel) to V–4Cr–4Ti, the fraction of BeF 2 can be changed from 43 to 31 mol% without changing the margin to coagulation of Flibe. The compositional change of Flibe resulted in a significant decrease in its viscosity. The decrease of TBR by the effects of MHD insulator coating with Er was negligibly small with the expected doping level of Er in the FFHR-LV. The increase in the B 4 C in the shielding material resulted in the decrease in the fast neutron flux at the superconducting magnet by 60–70% in both FFHR-LV and FFHR-FV. Optimum neutronics performance for the two blanket systems was derived by the calculation.

Z-pinch driven IFE (Z-IFE) requires the design of a repetitive target insertion system that allows coupling of the pulsed power to the target with adequate standoff, and a chamber that can withstand blast and radiation effects from large yield targets. The present strategy for Z-IFE is to use high yield targets (∼2–3 GJ/shot), low repetition rate per chamber (∼0.1 Hz), and 10 chambers per power plant. In this study, we propose an alternative power plant configuration that uses very high yield targets (20 GJ/shot) in a single chamber operating at 0.1 Hz. A thick-liquid-wall chamber is proposed to absorb the target emission (X-rays, debris and neutrons) and mitigate the blast effects on the chamber wall. The target is attached to the end of a conical shaped recyclable transmission line (RTL) made from a solid coolant (e.g., frozen flibe), or a material that is easily separable from the coolant (e.g., steel). The RTL/target assembly is inserted through a single opening at the top of the chamber for each shot. This study looks at the RTL material choice from a safety and environmental point of view. Materials were assessed according to waste disposal rating (WDR) and contact dose rate (CDR). Neutronics calculations, using the TART2002 Monte Carlo code from Lawrence Livermore National Laboratory (LLNL), were performed for the RTL and Z-IFE chamber, and key results reported here.

The molten salt, Flibe, has been recommended as a candidate coolant material in the blanket system of the FFHR fusion reactor though it is high Prnadtl-number (Pr) fluid that leads to low heat transfer performance. This paper, describes the results of numerical simulations performed in order to estimate the effects of cylinders as obstructions for heat-transfer enhancement in high-Pr fluid duct flow. Two-dimensional thermofluid simulations were performed for cases with 44, 24 and 48 cylinders, respectively, inserted perpendicularly to the fluid flow, and acting as heat transfer enhancers between parallel plates. From these analyses, the flow contraction created by the cylinders causes a high-vorticity around the heated wall. This high-vorticity plays an important role in the heat-transfer enhancement. In the high-vorticity region, the momentum perpendicular to a wall has a large gradient along the stream direction. In fact, the fluid flows down while rotating and “washing” the heated wall. This effect is also governed by the arrangement of cylinders. A staggered arrangement is adopted in the case with 44 cylinders, while square arrangement is employed in the cases with 24 and 48 cylinders. The enhancement of perpendicular flow is very effective when using a staggered arrangement, procuring a higher heat transfer downstream of the cylinders. The estimated pressure drop for high-Pr fluid flow was larger for the with 44 cylinders than for the cases with 24 and 48 cylinders. This result indicates that the heat transfer of high-Pr fluid flow strongly depends on the effect of flow stirring caused by obstructions.

Thermofluid research issues relating to self-cooling liquid blanket system for fusion reactors are discussed to find ways to realize the system. In this paper, liquid Li and Flibe molten salt are chosen as the blanket coolants. For the Li blanket system, there exists some possibility to overcome MHD problem by using three-surface coated channel with multi-layer structure. The material properties in terms of electrical conductivity required for the innermost metal layer seems achievable together with new concept that the coated material works as the structural component of the innermost thin layer. In the case of Flibe coolant, which shows very small MHD pressure drop, electrolysis occurs to result in generation of fluorine and tritium. The numerical results show that this electrolysis can be suppressed by optimizing the channel geometry. Numerical and experimental results indicate that heat transfer enhancement using pebble beds is expected when the flow velocity is relatively small to reduce the MHD effect.

In order to investigate the viability of using Be as a redox agent in a molten Flibe (2LiF–BeF 2 ) blanket, a series of kinetics experiments were performed in which HF was bubbled through Flibe with varying concentrations of dissolved Be. The feed gas consisted of 910–1800 ppm HF and 0.1–0.2 vol.% H 2 , with the balance comprised of He. A cylindrical rod of Be was contacted with the salt for periods of time ranging from 600 to 3600 s, resulting in mole fractions in the salt ranging from 4.3 × 10 −5 to 2.6 × 10 −4 . Initially, high HF conversion levels in excess of 90% were observed. As HF reacted with the Be, the conversion levels slowly dropped over a period of several hours to a few days. A simple kinetic model, which is first order in both HF and Be concentration has been coupled with a non-mixed reactor model to yield a good fit to the data. Application of this model indicates that Be should be suitable for keeping the TF concentration in the salt below 0.02 ppb.

The Advanced Power Extraction (APEX) project has been exploring concepts for power producing blankets that can enhance the potential of fusion energy. The pursued concepts cover both liquid and solid wall designs. The solid wall blanket branch of the project has been concentrating on the use of nano-composited ferritic (NCF) steel structure coupled with Flibe (Li 2 BeF 4 ) coolant. The present blanket is a solid wall design with an innovative coolant scheme, which allows part of the coolant to be re-circulated in order to enhance the outlet temperature, and thus improve the power cycle efficiency. The structure is 12YWT, an oxide dispersion strengthened (ODS) ferritic steel, which has a maximum operating temperature of 800 °C and a compatibility with Flibe up to 700 °C. Several methods for fabricating the complex geometry of the first wall (FW) and blanket are presented. Preliminary coolant routing is proposed with solutions offered for minimizing heat losses and simplifying assembly and maintenance. Even though this blanket is somewhat complicated, its forward looking aims, that of maximizing nuclear performance to achieve a high thermal conversion efficiency, are well worth striving for.

The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn–Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied.

In this article, we explore some of the safety issues associated with two advanced ferritic steel (AFS) molten salt blanket designs from the Advanced Power Extraction (APEX) design study [M.A. Abdou, The APEX Team, On the exploration of innovative concepts for fusion chamber technology, Fus. Eng. Des. 54 (2000) 181]. In particular, we examine radiological inventories, decay heat, waste disposal ratings, and toxic chemical inventories of these design concepts. In addition, we predict the thermal response of these blanket designs during accident conditions, and the mobilization of the radiological inventories and site boundary dose from the release of this mobilized material during a worst-case confinement-boundary-bypass accident. The molten salts being proposed for these blanket concepts are Flibe and Flinabe, and the structural material is a nano-composite strengthened ferritic steel alloy called 12YWT. The estimated dose at the site boundary is less than the no-evacuation limit of 10 mSv for a ground level release during conservative weather conditions if plant isolation occurs within 5 days.

As an element in the U.S. Advanced Power Extraction (APEX) program, the solid first wall and blanket design team assessed innovative design configurations with the use of advanced nano-composite ferritic steel (AFS) as the structural material and FLiBe as the tritium breeder and coolant. The goal for the assessment is to search for designs that can have high volumetric power density and surface heat-flux handling capability, with assurance of fuel self-sufficiency, high thermal efficiency and passive safety for a tokamak power reactor. We selected the re-circulating flow configuration as our reference design. Based on the recommended material properties of AFS we found that the reference design can handle a maximum surface heat flux of 1 MW/m2, and a maximum neutron wall loading of 5.4 MW/m2, with a gross thermal efficiency of 47%, while meeting all the tritium breeding, structural design and passive safety requirements. This paper will cover the results of the following areas of assessment: material design properties, FW/blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal-hydraulics analysis including MHD effects, structural analysis; molten salt and helium closed cycle power conversion system; and safety and waste disposal of the re-circulating coolant first wall and blanket design.

Selection of coolant used in the fuel zone of a fusion–fission (hybrid) reactor affects the neutronic performance of the blanket much. Recently, two coolants namely, Flinabe and Li20Sn80 have been investigated to use in fusion reactors as tritium breeder and energy carrier due to their advantages of low melting point, low vapor pressure. In this study, neutronic performance of these coolants in a hybrid reactor using Canada Deuterium Uranium Reactor (CANDU) spent fuel was investigated for an operation period of 48 months. And also that of natural lithium and Flibe was also examined for comparison. Neutron transport calculations were conducted on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S8–P3 approximation.

New results on neutron intensities and energy spectra in structural wall materials versus time after inertial fusion (IFE) target emission are presented, showing differences between two IFE chamber protections (LiPb, Flibe). Key parameters and mechanisms are: density, moderation, and threshold reactions such as (n, 2n) and (n, 3n). Using computed time-dependent neutron intensities in the structural wall, we present a Multiscale Modeling study of pulse (1–10 Hz) irradiation in Fe, up to the level of defect microscopic characterization depending on time irradiation. Final responses of the microscopic structure after irradiation to 10−3 dpa are reported, and the differences with a continuous irradiation, for a still low irradiation fluence, are remarked. A new code based on tight binding molecular dynamics has been developed for studying SiC; and its first applications to different temperature-dependent situations demonstrates a reliable proof of principle of the new model. An efficient Multiscale Modeling systematic approach of SiC is lacking; starting with the absence of an appropriate description of defects and its diffusion. That goal can be obtained using this TBMD accurate tool.

Nuclear aspects of candidate molten salts, namely a mixture of LiF and BeF2 (FLIBE) and a mixture of LiF, NaF, and BeF2 (FLINABE), were investigated for application as blanket coolants in tokamak fusion power plants. Tritium breeding, blanket energy multiplication, and neutron transmutation of these salts were assessed. Neutron activation of FLIBE and FLINABE was evaluated and site-boundary dose due to a worst-case loss of vacuum accident was estimated. Formation of F2, TF and T2 during power plant operation was analyzed and issues relevant to corrosion of structural materials due to the fluorine and fluoride species was assessed. Mechanism to control the corrosion of structural materials due to TF has been identified. Depletion of LiF, BeF2, and NaF in the salts was calculated and quantities of the make-up fluorides to be added into the salts were estimated.

The force-free helical reactor (FFHR) is a demo relevant helical-type D–T fusion reactor based on the large helical device (LHD). In this study, the possibilities of fissile fuel breeding (FFB) and incineration of curium oxide discharged from spent fuel of pressured water reactor (PWR) have been investigated in the FFHR. For this purpose, FFB zone containing mixed fuel has been located in the blanket of the FFHR by using two different design shapes (models 1 and 2). Four different mixtures, namely, 244CmO2–UC, 244CmO2–UO2, 244CmO2–UN and 244CmO2–U3Si2, have been used as fuel. The mixed fuels have been spherically prepared, and cladded with SiC to prevent fission products from contaminating coolant and fuel–coolant contact. The prepared fuel spheres have been replaced in the FFB zone as ten rows in radial direction, and two different coolants, helium (for models 1a and 2a) and flibe (for models 1b and 2b), have been selected for the nuclear heat transfer in the FFB zone. Calculations of neutronic data per DT fusion neutron have been performed individually for each of 16 different cases created by using four different mixed fuels in each of four different models. In the models 1 and 2, the case of helium coolant has higher M (the blanket energy multiplication ratio), which is one of main parameters in a fusion–fission hybrid reactor than the case of flibe coolant. M is quite high, and varies in the ranges of 4.747–5.895 and 2.453–2.784 in the models 1 and 2, respectively, depending on the fuel and coolant types. The fissile fuel breeding ratio (FFBRs) are also quite high in the model 1a with respect to model 1b, although coolant type does not change the FFBRs substantially in the model 2. Total FFBR (239Pu+245Cm breeding ratio) varies from 0.551 to 0.650 and from 0.170 to 0.184, in the models 1 and 2, respectively, depending on the fuel and coolant types. Values of tritium breeding ratio (TBR), which is another important parameter in a fusion–fission hybrid reactor, are about 1.2 for all investigated cases so that tritium self-sufficiency is maintained for (D,T) fusion driver. Values of peak-to-average fission power density ratio, Γ are in the range of 1.108–1.249 in all investigated cases, depending on the fuel and coolant types. Furthermore, values of neutron leakage out of the blanket for all models are quite low due to B4C shielding. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.

A new cooling concept of free surface flow is proposed as an alternative candidate for the first wall of self-cooling molten-salt Flibe blanket in a helical-type fusion reactor force-free-like helical fusion reactor. Numerical analyses on a capillary flow show that the free surface balanced between surface tension and the forces of wall shear stress and gravity is feasible even in helical systems, where a spiral flow is formed and drastically enhances the heat transfer from the coolant channel wall.

The thin Convective Liquid Flow First Wall (CLiFF) concept is one of the liquid FW concepts investigated in the Advanced Power Extraction (APEX) study for high power density application. Liquid tin has been suggested as the 2 cm thick front flowing liquid layer because of its low vapor pressure. Two choices were selected for the conventional blanket that follows the thin liquid wall (LW), namely: (1) LiPb/SiC blanket, and (2) Flibe/SiC blanket. Lithium is enriched to 90% Li-6 in the first blanket option and to 25% Li-6 in the second option (with 10 cm-thick beryllium front zone). Due to the superior attenuation characteristics of Flibe over LiPb, this impacted the lifetime of the SiC structure used in both options. In this paper, we assessed the lifetime of the SiC structure in the FW/Blanket and the shield in both blanket options. The end-of-life limit of 200 dpa is assumed (corresponding to ∼3% burn-up). The frequency of replacement of each component is estimated based on 30-years plant lifetime. Comparison is made for the waste volume of replaced components in each option. It is shown that the shield can last the plant lifetime in the Flibe blanket while part of the shield in the LiPb blanket will require replacement. The frequency of replacing the FW/Blanket with the LiPb blanket option is twice as much as in the Flibe blanket option. This is translated to a total volume of disposed structure at plant end-of-life from the entire FW/B/shield system that is larger by ∼60%.

Successful development of heavy-ion fusion (HIF) will require scientific and technology advances in areas of targets, drivers and chambers. Design work on heavy-ion targets indicates that high gain (60–130) may be possible with a ∼3–6 MJ driver depending on the ability to focus the beams to small spot sizes. Significant improvements have been made on key components of heavy-ion drivers, including sources, injectors, insulators and ferromagnetic materials for long-pulse induction accelerator cells, solid-state pulsers, and superconducting quadrupole magnets. The leading chamber concept for HIF is the thick-liquid-wall HYLIFE-II design, which uses an array of flibe jets to protect chamber structures from X-ray, debris, and neutron damage. Significant progress has been made in demonstrating the ability to create and control the types of flow needed to form the protective liquid blanket. Progress has also been made on neutron shielding for the final focus magnet arrays with predicted lifetimes now exceeding the life of the power plant. Safety analyses have been completed for the HYLIFE-II design using state-of-the-art codes. Work also continues on target fabrication and injection for HIF. A target injector experiment capable of >5 Hz operation has been designed and construction will start in 2002. Methods for mass-production of hohlraum targets are being evaluated with small-scale experiments and analyses. Progress in these areas will be reviewed.

Among the advantages offered by the deployment of thick liquid walls (LW) in high power density reactors is the substantial reduction in radwaste volume and hazard that is mainly attributed to the extended lifetime of structural materials. In this paper, we quantitatively estimate the volume of the generated waste when different thick LWs are used. In particular, we make a comparison of the radwaste volume between lithium and Flibe as potential candidates for deployment as the thick LWs in high power density reactors (10 MW/m2). In addition, the volume of the generated waste is compared with the corresponding volume in two conventional solid wall (SW) blankets with low wall load (LWL, 5 MW/m2) and high wall load (HWL, 10 MW/m2). In this assessment exercise, the blankets under consideration were optimized first such that adequate tritium breeding ratio (TBR) is obtained and the same level of magnet protection against radiation damage is reached. This initial optimization step was necessary to arrive at consistent comparisons. It is shown that using LWL conventional SW blanket generates ∼10% more waste volume per unit height than that generated with the lithium LW concept while ∼35% more waste volume is generated if the HWL conventional SW blanket is deployed. On the other hand, it is shown that the thick Li LW option generates total waste whose volume is a factor of ∼1.3 larger than the one with the thick Flibe option which has superior neutron moderating capabilities compared with lithium.

Experimental research on molten salt thermofluid technology using a high-temperature molten salt loop (MSL) is described in this paper. The MSL was designed to be able to use Flibe as a coolant, however, a simulant, heat transfer salt (HTS) has to be used alternatively since Flibe is difficult to operate under avoiding a biohazard of Be. Experiment on heat-transfer enhancement, that is required for applying to cool the high heat flux components of fusion reactors, is ongoing. Preliminary experimental results showed that an internal structure of a mixing chamber in the MSL was important to obtain accurate bulk temperatures under severe thermal conditions. For operating the loop, careful handling are needed to proceed how to melt the salt and to circulate it in starting the operation of the MSL. It is concluded that several improvements proposed from the present experiences should be applied for the future Flibe operation.

New concept of remountable first wall using liquid flow has been proposed as a different approach of liquid wall concepts like APEX (Abdou, the APEX team, Fusion Eng. Des. 54 (2001) 181). In this study, the numerical analysis of MHD flow has been carried out to evaluate effect of coating on the MHD pressure drop. In the proposed concept the liquid layer is sandwiched by two metal plates with two side ribs to compose the flow channel. The numerical results indicate that coating the whole rib surface has good performance as three surface coating to reduce the MHD pressure drop. In both cases of flibe and lithium coolants, the solid metal (HT-9) wall can be used by coating with the insulator whose electric conductivity ratio ( σ coating / σ HT-9 ) is less than 10 −6 under 6 T.

Flibe–tritium experiment in the Japan–US joint project (JUPITER-II) was initiated in 2001. H/D isotopic exchange experiments were conducted to select a Flibe-facing material. Because of hydrogen isotope interactions with carbon, Ni crucibles were selected for Flibe/tritium behavior experiments. A Flibe–tritium pot with two Ni (or Cu) permeable probes was designed. The rate of the overall tritium permeation through the Flibe-facing Ni or Cu was estimated by numerical simulation using TMAP4 code. Diffusion in bulk Flibe was found to be the rate-determining step for purified Flibe.