Two-phase modeling of a gas phase fluidized bed reactor for the fluorination of uranium tetrafluoride

A bubbling fluidized bed reactor for the fluorination of uranium tetrafluoride by fluorine gas was simulated employing two-phase models, with the bubble phase assumed to be in plug flow, and the emulsion phase in plug flow (P-P model) and in perfectly mixed flow (P-M model). The model calculations were compared with actual data in term of fluorine conversion. The comparison showed that the P-M model predicted the data satisfactorily. The P-M model was then applied to understand the influence of operating parameters on the reactor performance. A sensitivity study of various operating parameters showed the extent each parameter can influence the rates of fluorine conversion and uranium hexafluoride production.     

Dissolution of uranium metal without hydride formation or hydrogen gas generation

This study shows that metallic uranium will cleanly dissolve in carbonate-peroxide solution without generation of hydrogen gas or uranium hydride. Metallic uranium shot, 0.5–1 mm diameter, was reacted with ammonium carbonate–hydrogen peroxide solutions ranging in concentration from 0.13 M to 1.0 M carbonate and 0.50 M to 2.0 M peroxide. The dissolution rate was calculated from the reduction in bead mass, and independently by uranium analysis of the solution. The calculated dissolution rate ranged from about 4 × 10⁻³ to 8 × 10⁻³ mm/h, dependent primarily on the peroxide concentration. Hydrogen analysis of the etched beads showed that no detectable hydrogen was introduced into the uranium metal by the etching process.     

Kinetic parameters of a low enriched uranium fuelled material test research reactor at end-of-life

The kinetic parameters at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA’s 10 MW benchmark reactor. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time and effective delayed neutron fraction. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that in comparison with the beginning-of-life values, at end-of-life, the neutron flux increased throughout the core, the prompt neutron generation time increased by 3.68% while the effective delayed neutron fraction decreased by 0.35%.     

Reactivity feedback coefficients of a low enriched uranium fuelled material test research reactor at end-of-life

The reactivity feedback coefficients at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA’s 10 MW benchmark reactor. Simulations were carried out to calculate the different reactivity feedback coefficients including Doppler feedback coefficient, reactivity coefficient for change of water temperature and reactivity coefficient for change of water density. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitude of all the reactivity feedback coefficients increased at end of life of the reactor by almost 2–5%.     

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl_x–Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si–Al and U₃Si₂–Al) and oxide (U₃O₈–Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 °C to 50 °C and 100 °C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U₃O₈–Al was about 2% more than the original UAl_x–Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.     

Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl_x-Al) containing 4.40 gU/cm³ of an MTR was replaced with silicide (U₃Si-Al and U₃Si₂-Al) and oxide (U₃O₈-Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U₃Si₂-Al followed by 0.03% for U₃Si-Al, and 0.01% for U₃O₈-Al fuel. The U₃O₈-Al fueled reactor gave the maximum ρ_excess at BOL which was 21.67% more than the original fuel followed by U₃Si-Al which was 2.55% more, while that of U₃Si₂-Al was 2.50% more than the original UAl_x-Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U₃O₈-Al followed by U₃Si-Al and then U₃Si₂-Al fuel.     

Methodologies for hydrogen determination in metal oxides by prompt gamma activation analysis

Prompt gamma activation analysis (PGAA), available at The University of Texas at Austin (UT), has been employed for the direct determination of hydrogen content in a series of metal oxide materials typically used as cathodes in lithium ion battery systems. Special attention was given to the experimental setup including potential sources of error and system calibration for the detection of hydrogen. Spectral interference with hydrogen arising from cobalt was identified and corrected for. Limits of detection as a function of cobalt mass present in a given sample are also discussed. PGAA has proven to be a novel and precise technique for the determination of hydrogen in metal oxides. This type of investigation could provide valuable insight regarding the factors that limit the practical capacities of lithium ion oxide cathodes.     

Integrated infrastructure initiatives for material testing reactor innovations

The key goal of the European FP6 project MTR+I3 was to build a durable cooperation between Material Testing Reactor (MTR) operators and relevant laboratories that can maintain European leadership with updated capabilities and competences regarding reactor performances and irradiation technology.The MTR+I3 consortium was composed of 18 partners with a high level of expertise in irradiation-related services for all types of nuclear plants.This project covered activities that foster integration of the MTR community involved in designing, fabricating and operating irradiation devices through information exchange, know-how cross-fertilization, exchanges of interdisciplinary personnel, structuring of key-technology suppliers and professional training. The network produced best practice guidelines for selected irradiation activities.This project allowed to launch or to improve technical studies in various domains dealing with irradiation test device technology, experimental loop designs and instrumentation. Major results are illustrated in this paper. These concern in particular: on-line fuel power determination, neutron screen optimization, simulation of transmutation process, power transient systems, water chemistry and stress corrosion cracking, fission gas measurement, irradiation behaviour of electronic modules, mechanical loading under irradiation, high temperature gas loop technology, heavy liquid metal loop development and safety test instrumentation.One of the major benefits of this project is that, starting from a situation of fragmented resources in a strongly competitive sector, it has created a framework opening the way to further collaboration between the involved partners regarding the development and utilization of irradiation devices. In addition, the courses developed in the different institutes within the framework of professional training can continue after the end of the project.     

Absorption of optical radiation by nanoclusters of sputtered metallic uranium and its oxides

Thin films of uranium-containing materials are used as energy-releasing elements in nuclear-pumped lasers. They serve as a source of fission products which ionize and excite the gaseous medium. In this pumping method, a uranium layer is sputtered by its own fission products. This is one of the factors that limit the service life of energy-releasing elements. The products of sputtering can enter the gas and affect the optical and kinetic properties of laser media. In the present work, calculations are performed of the electrodynamic cross-sections for photoabsorption of optical radiation by nanoparticles of metallic uranium and its oxides. The main space-time relaxation parameters are determined. The effect of sputtered uranium-containing nanoparticles on the absorption of optical radiation in a laser element is evaluated. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 98–102, August, 2007.     

A thin-layer reference material for hydrogen analysis

In many areas of material sciences, hydrogen analysis is of particular importance. For example, hydrogen is most abundant as impurity in thin film materials - depending on the deposition process - and has great influence on the chemical, physical and electrical properties of many materials. Existing bulk reference materials (RMs) are not suited for surface sensitive analytical methods like elastic recoil detection analysis (ERDA) or nuclear reaction analysis (NRA). To overcome this serious lack of (certified) thin-layer reference materials for the determination of hydrogen in the near-surface region (1-2 μm depth), we produced stable, homogeneous amorphous silicon layers on Si-wafers (aSi:H-Si) by means of chemical vapour deposition (CVD), while about 10% of hydrogen was incorporated in the Si-layer. Homogeneity and stability were proved by NRA whereas traceability of reference values has been assured by an international interlaboratory comparison.     

Passive modular gas safety system for a reactor

Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center, RNTs. Translated from Atomnaya Énergiya, Vol. 75, No. 1, pp. 8–13, July, 1993.     

Development of oxides dispersion strengthened steels for high temperature nuclear reactor applications

By introducing a dispersion of nanosized yttrium oxides particles into a steel matrix, the upper temperature limit in mechanical creep strength can be enhanced in temperature by 100 K at least. Production routes for the production of a new class of oxides dispersion strengthened (ODS) steels are investigated within this work. Preliminary results obtained when doping pure iron matrix phase with two types of yttrium oxides (Y₂O₃) nanoparticles (commercial as well as laboratory fabricated nanopowder) are presented. The twofold purpose of this work is firstly to obtain a comparative analysis between the commercial and the laboratory fabricated Y₂O₃ nanopowder used to produce the doped iron, and secondly to demonstrate the feasibility of new production route by observing the nanostructure of the first test batches with pure iron. Observations are carried out with transmission electron microscopy (TEM) to determine the size distribution of the particles in the powder, while glow discharge optical emission spectroscopy (GDOES) and high resolution-scanning electron microscopy (HR-SEM) are used to analyze the chemical composition and the homogeneity of the produced doped iron. It is demonstrated, that even with small size particles nanopowder fabricated in the laboratory, the distribution is fairly homogeneous compared to the one obtained with a relatively large particles commercial nanopowder, confirming the feasibility of the new production route.     

Low-additive uranium alloys for the fuel elements of the KS-150 reactor

Conclusion The investigations showed a high resistance to swelling in an alloy of uranium with small amounts of aluminium and chromium at a burnup value of 1 at.%. The hot-working scheme for the alloy, including hardening of the solid solution, annealing (aging), and hardening, ensured the formation of a fine-grained isotropic structure with a finely dispersed distribution of second-phase particles, which facilitated the effective utilization of the alloying additives in limiting the cavitation and vacancy swelling of the uranium.The results obtained showed that an alloy of uranium with aluminum and chromium will make it possible to obtain high burnup values in the KS-150 reactor.Translated from Atomnaya Énergiya, Vol. 42, No. 6, pp. 452–456, June, 1977.     

Effects of high density dispersion fuel loading on the dynamics of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the dynamics of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different properties affecting the reactor in different ways, fuels U–Mo (9w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to determine the reactor performance under reactivity insertion and loss of flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It is observed that during the fast reactivity insertion transient, the maximum reactor power is achieved and the energy released till the power reaches its maximum increases by 45% and 18.5%, respectively, as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved during the transient, by 27.7 K, 19.7 K and 7.9 K, respectively. The time required to reach the peak power decreases. During the slow reactivity insertion transient, the maximum reactor power achieved increases slightly by 0.3% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³ but the energy generated till the power reaches its maximum decreases by 5.7%. The temperatures of fuel, clad and coolant outlet remain almost the same for all types of fuels. During the loss of flow transients, no appreciable difference in the power and temperature profiles was observed and the graph plots overlapped each other.     

Predicting the probability for fission gas resolution into uranium dioxide

Classical molecular dynamics simulations, using a set of previously established pair potentials, have been used to predict the minimum energy needed for krypton and xenon atoms to be resolved into uranium dioxide across a perfect (1 1 1) surface. The absolute minimum energy, E_min, is 53 eV for krypton and 56 eV for xenon atoms, significantly less than the 300 eV value often assumed in fuel modelling as the minimum energy required for gas resolution. The present values are, however, still sufficient to preclude thermal resolution at normal reactor temperatures. The discrepancies between the present and previous resolution energies are due to the significant variation in probabilities of absorption at different impact points on the crystal surface; we have mapped out the probability distribution for various impact sites across the crystal surface. The value of 300 eV corresponds to an 85% chance of resolution.     

Control rod drive for high temperature gas cooled reactor

This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.     

Noble gas binary mixtures for gas-cooled reactor power plants

This paper examines the effects of using noble gases and binary mixtures as reactor coolants and direct closed Brayton cycle (CBC) working fluids on the performance of terrestrial nuclear power plants and the size of the turbo-machines. While pure helium has the best transport properties and lowest pumping power requirement of all noble gases and binary mixtures, its low molecular weight increases the number of stages of the turbo-machines. The heat transfer coefficient for a He–Xe binary mixture having a molecular weight of 15 g/mole is 7% higher than that of helium, and the number of stages in the turbo-machines is 24–30% of those for He working fluid. However, for the same piping and heat exchange components design, the loop pressure losses with He–Xe are 3 times those with He. Consequently, for the same reactor exit temperature and pressure losses in piping and heat exchange components, the higher pressure losses in the nuclear reactor decrease the net peak efficiency of the plant with He–Xe working fluid (15 g/mole) by a little more than 2% points, at higher cycle compression ratio than with He working fluid.