Neutronic Analysis of the Laser Inertial Confinement Fusion–Fission Energy (LIFE) Engine Using Various Thorium Molten Salts

In this study, a neutronic performance of the Laser Inertial Confinement Fusion Fission Energy (LIFE) molten salt blanket is investigated. Neutronic calculations are performed by using XSDRNPM/SCALE5 codes in S₈–P₃ approximation. The thorium molten salt composition considered in this calculation is 75 % LiF—25 % ThF₄, 75 % LiF—24 % ThF₄—1 % ²³³UF₄, 75 % LiF—23 % ThF₄—2 % ²³³UF₄. Also, effects of the ⁶Li enrichment in molten salt are performed for all heavy metal salt. The radiation damage behaviors of SS-304 structural material with respect to higher fissionable fuel content and ⁶Li enrichment are computed. By higher fissionable fuel content in molten salt and with ⁶Li enrichment (20 and 50 %) in the coolant in form of 75 % LiF—23 % ThF₄—2 % ²³³UF₄, an initial TBR >1.05 can be realized. On the other hand, the 75 % LiF—25 % ThF₄ or 75 % LiF—24 % ThF₄—1 % ²³³UF₄ molten salt fuel as regards maintained tritium self-sufficiency is not suitable as regards improving neutronic performance of LIFE engine. A high quality fissile fuel with a rate of ~2,850 kg/year of ²³³U can be produced with 75 % LiF—23 % ThF₄—2 % ²³³UF₄. The energy multiplication factor is increased with high rate fission reactions of ²³³U occurring in the molten salt zone. Major damage mechanisms in SS-304 first wall stell have been computed as DPA = 48 and He = 132 appm per year with 75 % LiF—23 % ThF₄—2 % ²³³UF₄. This implies a replacement of the SS-304 first wall stell of every between 3 and 4 years.     

Radiation Damage and Tritium Breeding Study in a Fusion Reactor Using a Liquid Wall of Various Thorium Molten Salts

A new magnetic fusion reactor design, called APEX uses a liquid wall between fusion plasma and solid first wall to reach high neutron wall loads and eliminate the replacement of the first wall structure during the reactor’s operation due to the radiation damage. In this paper, radiation damage behavior of the inboard and outboard first walls made of a ferritic steel, 9Cr-2WVTa, in the APEX blanket using various thorium molten salts, 75% LiF-25% ThF₄, 75% LiF-24% ThF₄-1% ²³³UF₄ and 75% LiF-23% ThF₄-2% ²³³UF₄ was investigated. Furthermore, tritium breeding potential of these salts in such a blanket was also examined. Computations were carried out using the code Scale 4.3 by solving Boltzmann neutron transport equation. Numerical results brought out that only the liquid wall containing the molten salt, 75% LiF-23% ThF₄-2% ²³³UF₄ and having a thickness of ≥38 cm would be suitable to be used in the APEX reactor with respect to radiation damage criteria for the first wall structures and tritium self-sufficiency for the (DT) fusion driver.     

Neutronic performance of HYLIFE-II fusion reactor using various thorium molten salts

HYLIFE-II is one of the major inertial fusion energy reactor design concepts in which a thick molten salt layer (Flibe = Li₂BeF₄) is injected between the reaction chamber walls and the explosions. Molten salt coolant eliminates the frequent replacement of solid first wall structure during reactors lifetime by decreasing intense neutron flux. This study presents the neutronic analysis of HYLIFE-II fusion reactor using various liquid wall coolants, namely, 75% LiF–25% ThF₄, 75% LiF–24% ThF₄–1% ²³³UF₄ or 75% LiF–23% ThF₄–2% ²³³UF₄. Neutron transport calculations for the evaluation of neutron spectra were conducted with the help of Scale 4.3 by solving the Boltzmann transport equation in S₈–P₃ approximation. The effects of flowing liquid wall thickness and type of coolant on the neutronic performance of the reactor were investigated. Furthermore, radiation damage calculations at the first wall structure with respect to type and thickness of liquid wall were carried out. Numerical results showed that using the flowing liquid wall containing the molten salt, 75% LiF–23% ThF₄–2% UF₄ with a thickness of 70 cm maintained tritium self-sufficiency of the (DT) fusion driver and extended the first wall lifetime to the reactors lifetime (30 full power years). In addition significant amount of high quality fissile fuel was bred through (n, γ) reaction of ²³²Th. Moreover, energy multiplication factor (M) was increased to 12 by high rate fission reactions of ²³³U occurring in the flowing wall. On the other hand, it was concluded that using the other two coolants, 75% LiF–25% ThF₄ or 75% LiF–24% ThF₄–1% ²³³UF₄, as liquid wall did not satisfy the radiation damage and the tritium sufficiency criteria together at any thickness, so that these two coolants were not suitable to improve neutronic performance of HYLIFE-II reactor.     

Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium–tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li₂BeF₄ as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF₄ or UF₄) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S ₈–P ₃ approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF₄ or ThF₄ and 4% UF₄. In addition, higher energy multiplication values were found for the salt with UF₄ compared to that with ThF₄. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.     

On the radiation damage characterization of candidate first wall materials in a fusion reactor using various molten salts

Evaluating radiation damage characteristics of structural materials considered to be used in fusion reactors is very crucial. In fusion reactors, the highest material damage occurs in the first wall because it will be exposed to the highest neutron, gamma ray and charged particle currents produced in the fusion chamber. This damage reduces the lifetime of the first wall material and leads to frequent replacement of this material during the reactor operation period. In order to decrease operational cost of a fusion reactor, lifetime of the first wall material should be extended to reactor’s lifetime. Using a protective flowing liquid wall between the plasma and first wall can decrease the radiation damage on first wall and extend its lifetime to the reactor’s lifetime. In this study, radiation damage characterization of various low activation materials used as first wall material in a magnetic fusion reactor blanket using a liquid wall was made. Various coolants (Flibe, Flibe + 4% mol ThF₄, Flibe + 8% mol ThF₄, Li₂₀Sn₈₀) were used to investigate their effect on the radiation damage of first wall materials. Calculations were carried out by using the code Scale4.3 to solve Boltzmann neutron transport equation. Numerical results brought out that the ferritic steel with Flibe based coolants showed the best performance with respect to radiation damage.     

Dissolution behavior of (U,Zr)O2-based simulated fuel debris in nitric acid

To explore the possibility of dissolving fuel debris into nitric acid as a potential pre-treatment for waste treatment in which the U and Pu are removed from the inventory, dissolution tests of U_1?xZr_xO₂ and (U,Pu)_1?xZr_xO₂ were carried out in 6 M HNO₃ at 353 K. At the end of the dissolution test (after 4 h), the ratio of dissolved uranium decreased with an increase in the Zr contents, x. While the dissolution of U-rich samples was congruent, a preferential leaching of U was observed with Zr-rich samples. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). The IDR decreased from 10^?5 down to 10^?10 mol cm^?2 min^?1 as x increased from 0 to 0.95. From these findings, dissolution with HNO₃ is expected to be only applicable in U-rich part of fuel debris (x < 0.3) if the dissolution in 6 M HNO₃ at 353 K is assumed. Application of complexing acids, such as mixture of HNO₃ and HF, should be considered to increase the dissolution rate of the Zr-rich part.     

Investigation of systems of fused salts based on thorium fluoride

The system ThF₄-NaCl-KCL, which is significant in the selection of an electrolyte for the production of thorium, by electrolysis, was investigated by thermal and x-ray phase-analytical methods. The salts were melted in an atmosphere of argon. The following structural diagrams were constructed: the system NaCl-ThF₄-an eutectic type with complete insolubility in the solid state (the eutectic lies at 712 °C and 52 mole % ThF₄); the system KCl-ThF₄-the same type (eutectic at 704 °C and 23 mole % ThF₄); and a section of the ternary system NaCl-KCl-ThF₄ along the line (1NaCl: 1KCl)-ThF₄ with the point of intersection with the eutectic crystallization curve at 628 °C and 40% (by weight) ThF₄.     

Theoretical Study of the Endogenous Production of N-13 in 115 kJ Plasma Focus Device Using Methane Gas

Mather type plasma focus device with the bank energy of 115 kJ (40 kV, 144μF) was studied for induced activity of N-13; a short-lived radioisotope β⁺ emitter with 511 keV of gamma rays and has a half-life of t_1/2 = 9.93 min through ¹²C (d, n)¹³N nuclear reaction. N-13 radioisotope is used in Positron Emission Tomography (PET) for imaging and treatment. In this paper endogenous production of ¹³N is considered. It is shown by adding 3–4 % CH₄ to the chamber, the induced activity of N-13 has increased about 4 %. Our study is representative of producing 10⁶ ? 10⁹ Bq induced activity of this SLR in IR-MPF-100 device.     

Investigation on the Neutronic Performance of a Fusion Reactor Using Flibe with Heavy Metal Fluorides

Using liquid wall between the plasma and solid first wall in a fusion reactor allows to use high neutron wall loads and could eliminate frequent replacement of the first wall structure during reactor’s lifetime. Liquid wall should have a certain effective or optimum thickness to extend solid first wall lifetime to reactor’s lifetime and supply sufficient tritium for deuterium–tritium (DT) fusion driver. This study presents the effect of thickness of flowing liquid wall containing 90 mol % Flibe+10 mol % UF₄ or ThF₄ on the neutronic performance of a magnetic fusion reactor design called APEX. Neutron transport calculations were carried out with the aid of code Scale4.3. Numerical results brought out that optimum liquid wall thickness of ∼38 cm was found for the blankets using Flibe+10% UF₄ whereas, 56 cm for that with Flibe+10% ThF₄. Significant amount of high quality fissile fuel was produced by using heavy metal salt.     

Properties of Electrodeposited Tungsten Coatings on Graphite Substrates for Plasma Facing Components

Tungsten coating on graphite substrate is considered as one of promising candidate materials of plasma facing components. In this study, tungsten coatings on graphite substrate were successfully prepared by direct current (DC) and pulse current (PC) electrodeposition methods in Na₂WO₄–WO₃ molten salt under the air atmosphere. Pores were found on the surfaces of the tungsten coatings produced by DC electrodeposition method. For the coatings fabricated by PC method, compact and smooth tungsten coatings were successfully obtained. The crystal structure, morphology, density, microhardness, adhesive strength, oxygen content and the thermal conductivity of the coatings fabricated by PC method were investigated. The obtained tungsten coatings had a body centered cubic structure. After electro-deposition for 100 h, the thickness of the tungsten coating reached 810.02 ± 10.40 μm and the oxygen content was 0.03 wt%. The thermal conductivity of the tungsten coating was 134.29 W m^?1 K^?1. The density of the tungsten coating was 18.83 g cm^?3. The hardness of the coating was 492.0 ± 7.8 HV. After deuterium plasma irradiation, the tungsten coatings were prone to blistering.     

Structural investigation of thorium in molten lithium-calcium fluoride mixtures for salt treatment process in molten salt reactor

X-ray absorption fine structure (XAFS) measurements on thorium fluoride in molten lithium-calcium fluoride mixtures and molecular dynamics (MD) simulation of zirconium and yttrium fluoride in molten lithium-calcium fluoride mixtures have been carried out. In the molten state, coordination number of thorium (N_i) and inter ionic distances between thorium and fluorine in the first neighbor (r_i) are nearly constant in all mixtures. However the fluctuation factors (Debye-Waller factor (σ²) and C₃ cumulant) increase until xCaF₂ = 0.17 and decrease by addition of excess CaF₂. It means that the local structure around Th⁴⁺ is disordered until xCaF₂ = 0.17 and stabilized over xCaF₂ = 0.17. The variation of fluctuation factors is related to the number density of F⁻ in ThF₄ mixtures and the stability of local structure around Th⁴⁺ increases with decreasing the number density of F⁻ in ThF₄ mixtures. This tendency is common to those in the ZrF₄ and YF₃ mixtures. However, in the case of YF₃ mixtures, the local structure around Y³⁺ becomes disordered until xCaF₂ = 0.40 and it becomes stabilized by addition of excess CaF₂. The difference between ThF₄ mixtures and YF₃ mixtures is related to the difference of Coulumbic interaction between Th⁴⁺-F⁻ and Y³⁺-F⁻. Therefore, the variation of local structure around cation is related to not only number density of F⁻ in molten salts but also the Coulumbic interaction between cation and anion.     

Investigation of fused salt systems based on thorium fluoride

Thermographic, x-ray, and other analytical methods have been used to plot phase diagrams for the system NaF-ThF₄ with four chemical compounds (Na₄ThF₈; -Na₂ThF₆, -Na₂ThF₆, NaThF₅, NaTh₂F₉) and KF-ThF₄ with six chemical compounds (K₅ThF₉, K₃ThF₇, K₃Th₂F₁₁ KThF₅, KThF₉, KTh₆F₂₅). X-ray investigation of melts of the system NaF-KF-Th₄ showed the existence of the compound NaK(ThF₆) and its structure has been determined. Triangulation of the region NaF — Na₂ThF₆-KThF₅ — KF was carried out and the polythermic section for NaF-KThF₅ has been plotted.Commuvnication I (Consultants Bureau Translation page 561).     

Characterization of Thick Tungsten Coating on Cu Alloy Substrate Prepared by Electrodeposition

The W coating with the thickness over 1 mm was obtained by pulse electrodeposition on large CuCrZr alloy in a Na₂WO₄–WO₃ molten salt. The composition of this system keeps unchanged with the duration of electrodeposition. The coating’s structure, morphology and contents were investigated by XRD, XPS and SEM. The electrodepostion tungsten coating was compact and without void. The residual stress in surface of W coating was calculated to be a compressive stress with the value of 225 MPa. The W coating comprised an inner tooth-like layer and an outer columnar layer. The bonding strength between W coating and CuCrZr susbtrate was near 60 MPa; the thermal conductivity parallel to the crystal growth direction was measured to be 150.86 W m^?1 K^?1.     

A simulation of radiolysis of chloride solutions containing ferrous ion

ABSTRACT

To investigate the effect of dissolved species from steels on the radiolysis processes of Cl^?, radiolysis simulations of solutions containing both Cl^? and Fe²⁺ were carried out. The results showed that the generation of radiolytic products (H₂O₂, O₂ and H₂) increased mainly by the addition of Fe²⁺, and the concentrations of H₂O₂ and O₂ increased with increasing dose rate. Moreover, radiolysis of Fe²⁺ solutions also induced noticeable pH drop due to the hydrolysis of Fe³⁺. This pH drop enhanced the reactivity of Cl^? with ^?OH, which induced additional generation of H₂O₂ and O₂. These results show that low concentrations of Cl^? (1 × 10^?3 mol/dm³ = 35 mg/kg) in the presence of Fe²⁺ could influence the generation of H₂O₂ and O₂ during water radiolysis. On the other hand, it is considered that these effects of Fe²⁺ and Cl^? on water radiolysis are less important for corrosion of steels due to the low concentrations of H₂O₂ and O₂ generated if the concentrations of these additives and dose rate are sufficiently low. The other process, such as dissolution of iron enhanced by FeOOH, might predominantly induce corrosion under the conditions of solutions with low concentrations of H₂O₂ and O₂.     

Calcium Sulfate Activated by Lead and Manganese for Thermoluminescence Dosimetry

CaSO₄: Pb, Mn has been found to be free from the serious disadvantage of rapid fading possessed by CaSO₄: Mn, the most sensitive thermoluminescence phosphor available so far. A study has been made on the effect of lead and manganese content on CaSO₄, and it is concluded that lead in CaSO₄ produces new traps for radiation energy, resulting in improvement of the properties of the phosphors for dosimetry. The optimum content of such activator additives was found to be 0.2 mol/₀ of lead and 3 mol/₀ of manganese.

The improved phosphor thus obtained produces glow peaks at 160° and 190°C, and the energy yield of the thermoluminescence is about which is twice that of CaSO₄: Mn. The more significant properties of this phosphor from the viewpoint of application to radiation dosimetry include:

(1) Linear responce from 50μR to 10⁴R

(2) Minimum detectable dose of 50μR±25% by experimental reader

(3) Fading rate of 5% in a week     

Experimental study of pulse neutron irradiation damage in SiGe HBT

The response to transient irradiation of npn SiGe HBT (BG35 SiGe BiCMOS), i.e. device under test (DUT) was studied with online measurement of 1 MeV equivalent pulse neutron fluence of 0.8 × 10¹³n/cm². The differently biased DUT1 and DUT2 in test circuit were irradiated in the first day with neutron fluence (0.8 × 10¹³n/cm²) termed as Fluence1 and with an additional neutron fluence (0.8 × 10¹³n/cm²) in the second day to make Fluence2 equals to 1.6 × 10¹³n/cm². The experimental results show that pulsed neutron irradiation causes voltage surges in the DUTs exhibited by a negative and positive peak known to be radiation damage (RD). The RD in DUTs induced by pulse neutron Fluence1 initially created unstable displacement defects and the defects later reordered (cluster defects) to form more stable configurations via neutron Fluence2. The irradiated DUTs experienced online instantaneous annealing after 2.03 × 10^?9 s and offline measurement (i.e. without irradiation) of DUTs showed recovery to normal mode of operation after 24 h annealing. The level of pulse peaks in the base voltage terminals of DUT1 and DUT2 were compared as V_b2Fluence1 > V_b1Fluence1 and V_b2Fluence2 > V_b1Fluence2. A comprehensive analysis of RD region in DUTs with reference to Area (A₁, A₂), Peak (P₁, P₂), Height (H), and Full width at half maximum (FWHM) were investigated.     

The general corrosion behavior of ENiCrFe-7 weld overlay cladding material in deoxygenated high-temperature and high-pressure water

The general corrosion behavior of Alloy ENiCrFe-7 in deoxygenated high-temperature and high-pressure water was investigated. The results showed that the precipitates of Alloy ENiCrFe-7 included niobium carbide and Al-Ti-O compounds, and the weight gain increased fast firstly before 2250 h, then the weight gain slowed down. There were obvious large particles spread on denser oxide film after 3000 h exposure. Ni was present at a single chemical metallic Ni state, Feⁿ⁺ content of the outer layer was close to 60%, which was much higher than that of the matrix. The oxide film consisted of an inner layer and an outer layer, the inner layer was mainly composed of Cr₂O₃ and the outer layer was mainly composed of Fe₃O₄ and FeCr₂O₄. Finally, it is found that the preferential corrosion location of pitting was niobium carbide precipitates by in same site observation, while Al-Ti-O compounds was not dissolved in deoxygenated high-temperature and high-pressure water for 1500 h exposure. The size and number of the pitting was not significantly changed with increasing exposure time.     

Effects of External Resonant Fields on the Tokamak Edge Plasma Fluctuations

In this contribution, the effects of external resonant electric and magnetic fields on the tokamak edge plasma fluctuations have been investigated. For this purpose, the radial and poloidal electric fields and ion saturation current have been measured by two arrays of the Langmuir probes. An external resonant electric field was applied with the limiter biasing system. The biased electric voltage has been restricted to 0 < V _bias  < +320 and it has been applied with the limiter that is fixed in the r/a = 0.9. Moreover, the power spectra of particle flux Γ _r , Γ _p , Reynolds stress, coherency between E _p and I _s have been calculated. Fourier-based techniques have been employed to analyze the frequency of the Reynolds stress and particle flux. The results show that after positive biasing application (V _bias  = +200v), the Reynolds stress increases about 50 % while it doesn’t change remarkability after positive biasing with V _bias  = +320v. The Reynolds stress power spectrum confirms these results. The effect of positive biasing on Γ _r has been displayed a decrease about 60, 20 % by V _bias  = +200v, V _bias  = +320v respectively. Γ _p and I _s increase about 30 and 4 % in the present of biased (V _bias  = +200v) while they don’t change remarkability in another voltage. The power spectrum of Γ _r decreases about 15 and 10 % while the power spectrum of Γ _p increases about 80 and 20 % after positive biasing (V _bias  = +200v, V _bias  = +320v). Consequently, a better confinement obtained for biased with V _bias  = +200v. It means that, the magnitude of biased is important factor in modifying and control plasma turbulence. Also, results were compared with the effects of external resonant magnetic field.     

Surface refractive indices and profile widths of He implanted LiNbO3 waveguides

He implanted LiNbO₃ waveguides have been investigated by dark and bright line spectroscopy. The refractive index profiles were reconstructed with an improved inverse WKB procedure. In the region of electronic damage the resulting profiles are very reliable. Likewise the surface-side flank of nuclear damage induced refractive index decrease is reproduced with high accuracy. In contrast with other reconstruction schemes we determine refractive indices close to the surface with an accuracy better than 2 × 10⁻⁴. However, the full range of nuclear damage cannot be explored. We discuss how the profile parameters depend on ion energy and irradiation dose. At 3.17 MeV the helium enriched layer seems to saturate for doses above 5 × 10¹⁵ cm⁻². Electronic damage increases the ordinary and decreases the extraordinary refractive index, more for higher doses and less for higher energies.     

Surface Irradiation Damage of 316 Stainless Steel,Vanadium and Molybdenum by Helium Ion Bombardment

Molybdenum, V and 316 stainless steel were irradiated with 50~150 keV He ions at the temperatures between 413 and 1,298K for total doses ranging 1× 10²²~10ײ³ m^?2, and the characteristics of the surface damage were compared. Severe exfoliation was observed in all of these materials for the irradiation at 413±110 and 748±25K. The number of exfoliated skins was larger than that in literature, and increased nearly in proportion with the total dose. It increased in the order Mo<316SS<. When the dose was low, the amount of erosion increased rapidly with total dose, but tended to be saturated for higher doses than 3×10²² m^?2. It increased in the order Mo<V<316SS at 413±110K, while in the order 316SS<Mo<V at 748±25K. At higher temperatures than 923 K, blisters and porous surface were formed and the exfoliation of skins ceased. The amount of erosion increased with increasing incident ion energy in the energy range between 50 and 150 keV at 413±110K for a total dose of 1×10²² m^?2.