铀化合物与不同氟化剂反应的高温热力学

高温氟化物挥发法是一种用于分离和获取高纯度铀的核燃料干法分离工艺。该工艺技术成熟,分离得到的铀品质好,但因高温氟化的反应条件对设备材质要求苛刻,该方法在乏燃料后处理领域未能得到很好的发展。本文通过比较不同氟化剂与铀化合物发生氟化反应时的反应热及反应平衡常数,着重探讨了采用对设备材料腐蚀相对较弱的NF3替代F2的理论可行性。结果表明,NF3替代F2作为氟化剂在反应热力学上是可行的;氟化反应时放出的反应热也不会对熔融盐反应体系的温度带来剧烈影响。     

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.     

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.     

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.     

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.     

Recognition of uranium oxides in soil particulate matter by means of μ-Raman spectrometry

Soil samples from an abandoned uranium mine have been investigated in order to determine the molecular phases of uranium compounds. The experiments were carried out with soil particulate matter, collected randomly from the area of the formerly exploited ore. To select the particles rich with uranium, scanning electron microscopy with energy-dispersive X-ray attachment (SEM/EDX) was applied first. Afterwards, the particles were relocated and measured by μ-Raman spectrometry (MRS). Residues of the main deposit, uraninite UO₂, were detected, along with its alteration products. In terms of Raman scattering properties, uranium oxides are quite sensitive to the laser beam wavelength, which results in very specific features of their Raman spectra. In this paper the Raman spectra of uranium oxides of different origin and oxidation states, measured with 514 and 785 nm lasers, are also presented.     

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.     

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.     

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.     

Adapting the deep burn in-core fuel management strategy for the gas turbine – modular helium reactor to a uranium–thorium fuel

In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine – modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium–thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: ²³⁵U, which represents the 20% of the fresh uranium, ²³³U, which is produced by the transmutation of fertile ²³²Th, and ²³⁹Pu, which is produced by the transmutation of fertile ²³⁸U. In order to compensate the depletion of ²³⁵U with the breeding of ²³³U and ²³⁹Pu, the quantity of fertile nuclides must be much larger than that one of ²³⁵U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of ²³⁵U. At the same time, the amount of ²³⁵U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k_eff and mass evolution, reaction rates, neutron flux and spectrum at the equilibrium of the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium–thorium fuel.     

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.