A Measurement Platform Scheme and Data Post-processing Method for Thermal Conductivity of Li<Subscript>4</Subscript>SiO<Subscript>4</Subscript> Pebble Bed

The effective thermal conductivity of tritium breeder pebble bed is an important thermal parameter and must be known for the thermo-mechanical design of solid tritium breeder blankets. In order to obtain the parameter, experimental measurement is an effective method. A measurement platform was designed by University of Science and Technology of China for CFETR solid blanket scheme to measure the immediate thermal conductivity data and study the effect of pebble bed temperature, the purge gas pressure and pebble deformation on the thermal conductivity of pebble bed. Measurements were performed based on about 1 mm diameter Li₄SiO₄ pebbles in the temperature range between 100 and 800 °C, with purge gas pressure ranging from 0.1 to 0.3 MPa. This paper described a measurement platform scheme by thermal probe method. On the other hand, for the sake of increasing the precision of thermal conductivity data transformed from temperature data, some improvements for the data post-processing using Monte Carlo inversion method were made in this paper too.     

Study on the Chemical Compatibility Study Between Li<Subscript>2</Subscript>TiO<Subscript>3</Subscript> Pebbles and 14Cr-ODS Steel

In order to investigate the chemical compatibility between tritium breeder Li₂TiO₃ pebbles and tritium breeder blanket material oxide dispersion strengthened (ODS) steel, the contact interface between Li₂TiO₃ pebbles and ODS steel heated in argon atmosphere at 500, 600 and 700 °C for 300 h was studied. It was found that the ions of pebbles could diffuse and corrode with the cladding material after a long-time reaction at high temperature. The corrosion area formed on the surface of Li₂TiO₃ pebbles was small. With the increase of temperature, a zone with enriched iron was found on the surface of the pebble. This part of the surface was the direct contact surface between the pebble and the steel. At the same time, the relative density of the pebbles increased and the crush load was decreased to 30 N. In addition, a slight corrosion phenomenon was found on the surface of ODS steel. It has been proved that the main components of the corrosion products were the complex oxide containing Fe and Cr and the complex oxide containing Li and Fe.     

Fabrication of Al<Subscript>2</Subscript>O<Subscript>3</Subscript>/FeAl Coating as Tritium Permeation Barrier on Tritium Operating Component on Quasi-CFETR Scale

Limited CFETR-scale experience of engineering preparation techniques of tritium permeation barrier (TPB) exists up to date. Aimed at processing some real components that are usually tubular components sealed in one end, in the tritium cycling systems of China Fusion Engineering Test Reactor (CFETR), an Al₂O₃/FeAl coatings as TPB was prepared on tubular components of 321 type stainless steel components with a length of 400 mm and an external diameter of 150 mm, by Al-electroplating followed by heat treating and selective oxidation. The ability to construct TPB coated components on quasi-CFETR scale was demonstrated, with fabricating a TPB of Al₂O₃/FeAl coating with a double-layered structure, consisted of an outer γ-Al₂O₃ layer with a thickness of 0.3 µm and an inner (Fe,Cr,Ni)Al/(Fe,Cr,Ni)₃Al layer of 40 µm in thickness. The tritium permeation reduction factors of the Al₂O₃/FeAl TPB on component were 229 and 96 at 500 and 600 °C respectively. Finally, signatures and gaps of TPB mass process on CFETR-scale were discussed.     

Investigation of the Y<Subscript>2</Subscript>O<Subscript>3</Subscript> distribution in the weld joints of dispersion-hardened steel cladding of fast-reactor fuel elements

The results of investigations of the yttrium oxide distribution in the weld joints of dispersion-hardened steel cladding, manufactured by the powder metallurgy method, are presented. It is shown that when using the methods of fusing welding of thin-walled fuel-element cladding, the content and uniformity of the yttrium oxide distribution in the metal of a seam changes as compared with the cladding metal. The concentration and uniformity of the yttrium oxide distribution in the section of a weld joint obtained by pulsed laser welding is higher than that obtained with argon-arc welding. __________ Translated from Atomnaya énergiya, Vol. 102, No. 6, pp. 348–351, June, 2007.     

Isotope Exchange Capacity on Li4SiO4 and Comparison of Tritium Inventory in Various Solid Breeder Blankets

It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of a breeder material, absorption of water vapor into bulk of the grain, adsorption of water on surface of the grain, and exchange capacity of tritium to be trapped to surface of the grain together with two types of isotope exchange reactions for evaluation of the tritium inventory in a solid breeder blanket under various conditions. The isotope exchange capacity on the Li₄SiO₄ surface is experimentally obtained in this study. Most of the properties required for evaluation of the tritium inventory for various blanket materials have been already quantified by the present authors. Then it has become possible to compare the tritium inventory in solid breeder blankets packed with either Li₂O, LiAlO₂, Li₂ZrO₃, Li₂TiO₃ or Li₄SiO₄ using the calculation model previously presented by the present authors.     

Effect of CH<Subscript>4</Subscript> Ion Implantation in Pure Aluminium

In this paper, aluminium samples with 99.96% purity were exposed to ion beam, extracted from CH₄ plasma. Implantation of ions were performed for 50 keV energy and various doses ranging from 1 × 10¹⁷ to 6 × 10¹⁷ ions/cm². Morphology of surfaces, roughness and its evolution during variation of ion dose has been studied by atomic force microscopy (AFM). Microstructure of the modified surfaces after ion implantation has been obtained by X-ray diffraction technique and Raman spectroscopy. Formation of aluminium carbide (Al₄C₃) was confirmed by XRD results at implantation doses of 3 × 10¹⁷ and 6 × 10¹⁷ ions/cm². In addition, it was observed that when the ion dose is increased, orientation of aluminium planes change from (2 2 0) to (2 0 0). Corrosion test was performed and compared for implanted and un-implanted samples. The results showed that corrosion resistivity increase by accumulation of ion dose.     

Results of U<Subscript>0.55</Subscript>Pu<Subscript>0.45</Subscript>N and U<Subscript>0.4</Subscript>Pu<Subscript>0.6</Subscript>N mixed mononitride fuel tests in a bor-60 reactor to burnup 12% h.a.

The results of post-reactor studies of U_0.55Pu_0.45N and U_0.4Pu_0.6N mixed mononitride fuel elements (density 85% of the theoretical value) and a helium sublayer are presented. The fuel elements are irradiated in a BOR-60 reactor to burnup 9.4 and 12.1% h.a., respectively, with power density 430 and 540 W/cm. All fuel elements remained hermetic; the ChS-68 steel cladding (20% cold deformation) retained excess plasticity. The maximum zone of interaction between the cladding and the fuel and fission products did not exceed 15 μm. The swelling rate of U_0.4Pu_0.6N and U_0.55Pu_0.45N fuel was 1.1 and 0.68%/% burnup, respectively. The gas release did not exceed 19.3 and 19%. The steel damage dose was 43 dpa. The character of the porosity distribution in the fuel affects the swelling and gas release.     

Investigation of the influence of the initial state of powder,pore-forming additions,and U<Subscript>3</Subscript>O<Subscript>8</Subscript> on the microstructure and strength of fuel pellets

It is shown that the initial state of uranium dioxide powder has no effect on the density, microstructure, and strength of pellets. Pore-forming agents and U₃O₈ used in fabrication lower the pellet strength because their particles are not spherical. To increase pellet strength, it is recommended that U₃O₈ be subjected to special processing to spheroidize the particles before mixing for uranium dioxide powder.     

C<Subscript>60</Subscript>-Fullerene Hyper-Velocity High-Density Plasma Jets for MIF and Disruption Mitigation

We present an innovative idea to use hyper-velocity (>30 km/s) high-density (>10¹⁷ cm⁻³) plasma jets of D-T/H and C₆₀-fullerene for magneto-inertial fusion (MIF), high energy density laboratory plasma (HEDLP), and disruption mitigation in magnetic fusion plasma devices. The mass (~1–2 g) of sublimated C₆₀ and hydrogen (or D-T fuel) produced in a pulsed power source is ionized and accelerated as a plasma slug in a coaxial plasma accelerator. For MIF/HEDLP we propose to create a magnetized plasma target by injecting two high-Mach number high-density jets with fuel (D-T) and liner (C₆₀/C) structure along the axis of a pulsed magnetic mirror. The magnetized target fusion (MTF) plasma created by head-on collision and stagnation of jets is compressed radially by a metallic liner (Z-pinch) and axially by the C₆₀/C liner. For disruption mitigation, the C₆₀ plasma jets were shown to be able to provide the required impurity mass (J Fusion Energy 27:6, 2008).     

Study of the properties of ammonium polyuranate precipitates and UO<Subscript>2</Subscript> powders and pellets obtained with different technologies

The results of a differential-thermal analysis are used to compare the properties of ammonia polyuranate precipitates, UO₂ powders and pellets, obtained by different methods as well as metallic uranium. It is found that the phase NH₃·3UO₃·5H₂O forms in regular precipitation of ammonium polyuranate. When using nanotechnology, the phases NH₃·2UO₃·3H₂O and 4NH₃·6UO₃·8H₂O are also present in the precipitate. UO₂ powder prepared from such precipitate has high activity, since all phase transformations in it occur at a lower temperature. Modified fuel pellets of uranium dioxide, which are obtained by means of nanotechnology or mechanical addition of ammonia-containing reagents to powder, differ from the standard powders by a lower rate and more complex mechanism of oxidation, similar to metallic uranium. Modified UO₂ fuel pellets fabricated at the Physics and Power-Engineering Institute, are now undergoing tests in the BOR-60 reactor. After tests on the irradiated new modified fuel have been completed, it will be possible to judge its reliability.     

Effect of O<Subscript>2</Subscript>/Ar Mixture on the Structural and Optical Properties of ZnO Thin Films Fabricated by DC Cylindrical Magnetron Sputtering

Direct Current Cylindrical Magnetron Sputtering Setup was used to deposit ZnO thin films on BK7 substrates. The effects of changing O₂/Ar reactive gas mixtures on the structural and optical properties of films were studied. Crystallinity and structure of films were obtained by X-ray diffraction (XRD). Preferential crystalline growth orientation of ZnO films detected by XRD was always along the (002) orientation. The thickness of films was measured by surface profilometer which showed thickness increasing from 68.7 to 80.8 nm for 3–6% O₂ amount respectively. The morphology and roughness of the films were investigated by Atomic Force Microscopy (AFM). As oxygen gas amount was increased, the roughness and the grain size were decreased and the deposition rate was increased. The optical transmittance of ZnO films are obviously affected by the changing of O₂/Ar reactive gas mixtures. All films exhibit a transmittance higher than 70% in the visible region. The optical band gap of films was measured by Tauc’s method. The results show that by increasing the amount of O₂ in reactive gas mixture, the optical band gap of deposited films increases.     

Properties of the final product of thermochemical reprocessing of irradiated reactor graphite wastes in the Al-TiO<Subscript>2</Subscript>-C system

Laboratory investigations of the strength and chemical resistance of the final product of thermochemical reprocessing of reactor graphite wastes in the Al-TiO₂-C system are presented. The ¹³⁷Cs and ⁹⁰Sr leaching rate, which is determined for samples synthesized from a charge with real irradiated graphite from an AM research reactor, does not exceed 10⁻⁶ g/(cm²·day) at the 28th day. __________ Translated from Atomnaya énergiya, Vol. 104, No. 4, pp. 224–227, April, 2008.     

The Effect of Applied Voltage and Operating Pressure on Emitted X-Ray from Nitrogen (N<Subscript>2</Subscript>) Gas in APF Plasma Focus Device

In the present work, the effect of applied voltage and operating pressure on behaviour of X-rays emitted from nitrogen gas (N₂) used in APF plasma focus facility is investigated. It was found that the optimum conditions for high emissions of SXR and HXR from the plasma focus (PF) are different. At four applied voltages of 10, 11, 12, and 13 kV, the optimum pressures for SXR and HXR emissions of nitrogen gas (N₂) were obtained. At lower voltages, 10, and 11 kV optimum pressure for SXR emission was 3.5 torr while for HXR emission was 2.5 torr. At higher voltages, 12, and 13 kV, the optimum pressures shift to higher values at 4 and 3 torr for SXR and HXR emissions, respectively. Among the applied voltages, the least intensity of both SXR and HXR was at voltage 10 kV and the most intensity was for 13 kV which confirm with increasing voltage, the intensity of X-ray emission increases. Also the results obtained by images of pin-hole camera were in compatible with the results of detected signals by different filtered Pin-diodes and Scintillation detector. Our results illustrate that the voltage and the pressure are effective parameters in X-ray emission from the PF.     

Effect of the salt melt composition,temperature, and interaction time on the contact exchange reaction in the MCl-PbCl<Subscript>2</Subscript>-MeN systems

The interaction of titanium, zirconium, and aluminum nitrides with a mixture of fused alkali-metal chlorides with lead chloride at 870–1070 K is studied. It is shown that as a result of the contact exchange interaction, which occurs in no more than 5 h, a large fraction of the powdered nitrides transforms into a soluble state, and lead precipitates in a metallic form. The same behavior is also characteristic for compact hot-pressed samples of nitrides. The method proposed for transferring insoluble nitrides into a state which is soluble in salt systems can be used to recover mixed uranium-plutonium mononitride fuel and to fractionate fission products into high-, medium-, and low-level radioactive wastes. __________ Translated from Atomnaya énergiya, Vol. 104, No. 6, pp. 343–348, June, 2008.     

Investigation of the effect of B<Subscript>4</Subscript>C on the interaction of oxide melt and steel in a VVER-1000 vessel during a severe accident

The results of investigations of the interaction of U-Zr-B-C-O melts with steel, which are performed as part of the OECD Masca international program, are presented. It is found that, as a result of the interaction, boron and carbon become concentrated predominately in the metallic phase of the melt. As the initial mass ratio m_Fe/m_melt increases, the effect of the addition of B₄C on the melt-iron interaction decreases because the metallic phase is diluted with iron. It is concluded on the basis of a comparison of the results of the STFM-B Nos. 3, 7 experiments with the STFM-Fe Nos. 3, 7 experiments performed previously without the participation of boron carbide that the effect of boron carbide on the interaction of the oxide melt with iron decreases as the degree of oxidation of zirconium increases. Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 63–67, August, 2008.     

Use of γ Radiation from Short-Lived Fission Products for Detecting <Superscript>235</Superscript>U and <Superscript>239</Superscript>Pu

It is suggested that γ radiation with E _γ > 4900 keV from short-lived fission products produced by thermal neutrons be used to detect ²³⁵U and ²³⁹Pu in samples. A time regime is substantiated: 120 sec irradiation, 60 sec holding time, and 120 sec measurement time. The contribution of the reaction (n, p) on fast neutrons is studied.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 365–370, May 2005.     

Effect of <Superscript>3</Superscript>He and <Superscript>6</Superscript>Li accumulation in beryllium blocks on the neutron-physical characteristics of the MIR reactor

Reactions resulting in the accumulation of ³He and ⁶Li, whose thermal neutron capture cross-section is large, occur under the action of neutron radiation in the beryllium blocks of the MIR reactor core. When a neutron absorber accumulates in the moderator of a reactor, important physical characteristics change: reactivity access, efficiency of safety and control rods, and reactivity effects; in addition, energy release is redistributed. An algorithm for calculating ³H, ³He, and ⁶Li in each beryllium block of the core has been developed and implemented. This algorithm makes it possible to follow the change in the concentration of these nuclides during reactor operation and shutdown. The ³He and ⁶Li concentrations are used as initial data for calculating the neutron-physical characteristics of the MIR reactor using the MCU and BERCLI programs. The computational results for the effect of the accumulation of the nuclides indicated on the neutron-physical characteristics of the core are presented. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 84–88, February, 2008.