Influence of fission products on ruthenium oxidation and transport in air ingress nuclear accidents

In separate effect tests at 1000–1200 °C Ru oxidation rate and content of Ru in escaping air flow have been studied with special emphasis on effects of other fission product elements on the Ru oxidation and transport. The results showed that in the decreasing temperature section (1100–600 °C) most of the RuO₃ and RuO₄ (≈95%) decomposed and formed RuO₂ crystals; while the partial pressure of RuO₄ in the escaping air was in the range of 10^?6 bar. The re-evaporation of deposited RuO₂ resulted in about 10^?6 bar partial pressure in the outlet gas as well. Measurements demonstrated the importance of surface quality in the decreasing temperature area on the heterogeneous phase decomposition of ruthenium oxides to RuO₂. On the other hand water or molybdenum oxide vapour in air appears to decrease the surface catalyzed decomposition of RuO_x to RuO₂ and increases RuO₄ concentration in the escaping air. High temperature reaction with caesium changed the form of the released ruthenium and caused a time delay in appearance of maximum concentration of ruthenium oxides in the ambient temperature escaping gas, while reaction with barium and rare earth oxides extended Ru escape from the high temperature area.     

Molecular dynamics calculations of heat conduction in actinide oxides under thermal gradient

Thermal conductivities of UO₂, PuO₂ and (U_0.8,Pu_0.2)O₂ have been investigated by non-equilibrium molecular dynamics (NEMD) simulation between 300 K and 2000 K. The thermal conductivity was directly calculated by the temperature gradient on the system according to Fourier's law in NEMD simulation. The thermal conductivity obtained from the NEMD simulation decreases with a decrease of the supercell size, which means the phonon scattering occurs at the system boundaries in the microsystem. In addition, the present NEMD simulation, as well as previous EMD simulation studies, clearly shows that the Umklapp process causes the decrease of thermal conductivity at high temperatures. When comparison is made with literature data, the calculated results obtained from the relatively small supercell are in good agreement with the measured ones for the above actinide dioxides.     

The vaporisation behaviour of americium dioxide by use of mass spectrometry

The vaporisation behaviour of americium dioxide in vacuum at high temperatures up to 2400 K has been studied. A Knudsen cell coupled with a mass spectrometer was used to perform vapour pressure measurements. The ionisation efficiency curves of Am⁺, AmO⁺ and were simultaneously recorded. Appearance potentials of the key molecular species were determined by varying the energy of the ionising electrons at constant temperature. The partial and total vapour pressures above the americium oxide samples measured as a function of the temperature. The results on the vapour pressure of the pure americium dioxide samples were discussed together with the available literature data on plutonium dioxide containing small amounts of americium. Additional measurements have been performed on a mixed dioxide sample of plutonium containing 6.1 wt.% americium.     

Vapour pressure and standard enthalpy of sublimation of H3BO3

A horizontal thermal analysis instrument was adapted as a transpiration apparatus for the measurement of vapour pressure of solid boric acid, H₃BO₃. The experimental parameters necessary for establishing a dynamic isothermal congruent vapourisation equilibrium of H₃BO₃ were identified. Using these optimized transpiration experiments, the vapour pressures were measured in the temperature range 326-363 K. The temperature dependence of the measured values of vapour pressures could be expressed using the expression, log(p/Pa) = 26.83(±0.09) − 9094(±246)/T (K). The standard enthalpy of sublimation, , of H₃BO₃ was estimated to be 174.1 ± 4.7 kJ mol⁻¹ at the mean temperature of the present measurements, viz., 345 K.     

Fabrication and characterization of silicon nitride-based inert matrix fuels sintered with magnesium silicates

The effects of sintering additives of magnesium silicates, i.e. enstatite (MgSiO₃), steatite (MgSiO₃) and forsterite (Mg₂SiO₄), on the sintering behaviors and characteristics of the silicon nitride ceramics-based inert matrix fuels (IMFs) were experimentally investigated. Fabrication tests and characterizations of Si₃N₄-based IMFs with the sintering additives were carried out using cerium oxide (CeO₂) to represent minor actinide oxides. Sintered bodies were characterized in terms of their densities and thermal conductivities. In addition, a solubility test with hot nitric acid was carried out for evaluation of applicability to the existing reprocessing technology. The densification of sintered bodies was enhanced by using additives of magnesium silicates at relatively low sintering temperature. In particular, the relative density of Si₃N₄-based IMFs with Mg₂SiO₄ was above 90% at 1723 K. The thermal conductivities of Si₃N₄-based IMFs varied according to sintering temperature, and for the IMFs sintered at 1923 K were above 34 W/m K. The solubility test results revealed that only grain boundary phases in Si₃N₄-based IMFs can be dissolved into hot nitric acid.     

Chlorination of UO2, PuO2 and rare earth oxides using ZrCl4 in LiCl-KCl eutectic melt

A new chlorination method using ZrCl₄ in a molten salt bath has been investigated for the pyrometallurgical reprocessing of nuclear fuels. ZrCl₄ has a high reactivity with oxygen but is not corrosive to refractory metals such as steel. Rare earth oxides (La₂O₃, CeO₂, Nd₂O₃ and Y₂O₃) and actinide oxides (UO₂ and PuO₂) were allowed to react with ZrCl₄ in a LiCl-KCl eutectic salt at 773 K to give a metal chloride solution and a precipitate of ZrO₂. An addition of zirconium metal as a reductant was effective in chlorinating the dioxides. When the oxides were in powder form, the reaction was observed to progress rapidly. Cyclic voltammetry provided a convenient way of establishing when the reaction was completed. It was demonstrated that the ZrCl₄ chlorination method, free from corrosive gas, was very simple and useful.     

Effects of working fluid and shaft rotation speed on the performance of HTR plants and the size of CBC turbo-machine

Investigated are the effects of the molecular weight of the working fluid, reactor exit temperature, and shaft rotation speed on the size and number of stages of the turbo-machine as well as the performance of high temperature reactor (HTR) plants with actively cooled reactor pressure vessel and direct or indirect Closed Brayton Cycles (CBCs). The present analyses for working fluids of helium (4 g/mol) and the 15 g/mol He–Xe and He–N₂ binary mixtures are performed for a reactor thermal power of 600 MW, shaft rotation speed of 3000–9000 rpm, and reactor exit temperature from 973 K to 1223 K. For the plants with indirect CBCs, the analyses assume a temperature pinch of 50 K in the IHX. Results show that the CBC compression ratio is relatively low (2.6 for He and He–Xe and 3.2 for He–N₂), increases very little with increasing the reactor exit temperature and near the maximum thermal efficiency of the plant. For the plants with a direct helium CBC, the thermal efficiency increases from 42% to 51% as the reactor exit temperature increases from 973 K to 1223 K, respectively, versus 37% to 47% for the plants with indirect He-CBC. The HTR plants with indirect He–Xe and He–N₂ CBCs and operating at a turbine inlet temperature of 1123 K have slightly higher thermal efficiencies (45.9% and 45.8%) than the He plant with indirect CBC (45.6%), generating 1.6 MWe more electrical power. The molecular weight of the working fluid has a very small effect on the plant thermal efficiency, but significantly reduces the size and number of stages of the CBC turbo-machine. Increasing the shaft rotation speed also decreases the size and number of stages of the CBC turbo-machine.     

Densification of magnesia-based inert matrix fuels using asbestos waste-derived materials as a sintering additive

A new concept for densification of minor actinide-containing inert matrix fuels (IMFs) using asbestos waste-derived materials was proposed for the effective utilization of resources and health protection of the general public. In this concept, magnesium silicates, which are mainly generated by the decomposition of asbestos in low temperature heat-treatments, are used as a sintering additive to achieve high density magnesia (MgO) -based IMFs at relatively low sintering temperature. In the present study, preliminary fabrication tests of MgO-based IMFs with magnesium silicates were carried out using cerium oxide (CeO₂) as a representative of minor actinide oxides. The sintered densities of MgO-based IMFs increased with use of the additives. The sintering behavior of MgO-based IMFs with magnesium silicate additives was discussed from the viewpoints of the effects of magnesium silicates on the densification of the MgO and CeO₂.     

First-principles study of electronic structure and insulating properties of uranium and plutonium dioxides

First-principles density functional theory calculations were carried out to investigate the electronic structure and the degree of 5f states localization of the Mott-Hubbard type insulators UO₂ and PuO₂. We used the fully relativistic cluster discrete variational method (RDV) with the local exchange-correlation potential. The energies of one-electron transition between occupied and vacant 5f^5/2 states of neighboring actinide atoms were evaluated on the base of the ground state and the excited state calculations. It is found that in UO₂ and PuO₂ the energy difference between 5f^5/2 levels of nearest metal sites in the lattice are close to 1.0 eV and 0.9 eV, despite the results of conventional band structure approach predicting that both oxides are good conductors.     

Deuterium diffusion in LiOH–water-corroded oxide layer of zirconium alloys

In order to clarify the hydrogen diffusion mechanism in the oxide layer of zirconium alloys, in situ hydrogen isotope diffusion in the oxide layer has been examined. The zirconium alloys used were Zircaloy-2, GNF-Ziron (Zircaloy-2 type alloy with high iron content) and VB (zirconium-based alloy with high iron and chromium contents). They were corroded in 1 or 0.1 M LiOH-containing water at 563 K, producing oxide layers of 1.1–2.1 μm in thickness. The diffusion experiments were carried out in the temperature range from 488 to 633 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis for D (³He,p)⁴He reaction. From the transient deuterium profiles in the oxide layers, it was concluded the LiOH–water-corroded oxides had a single-layer structure, which was in contrast to the double-layer structure previously observed in steam-corroded oxide layers. The diffusion coefficients in the 1 M LiOH–water-corroded oxides evaluated from the deuterium profiles were smaller in the order of Zircaloy-2 > GNF-Ziron > VB at 573 K. For the 0.1 M LiOH–water-corroded oxide of GNF-Ziron, the diffusivity was lower than that of the 1 M LiOH–water-corroded oxide by a factor of 1/4. The present diffusion coefficients of the 1 M LiOH–water-corroded oxides of GNF-Ziron and VB were approximately 7 times larger than the previous data of the corresponding steam-corroded oxides. The deuterium diffusion properties in the oxides of the three alloys obtained in the in situ experiment were roughly consistent with their hydrogen absorption performances in the LiOH–water-corrosion tests, as well as in the previous steam corrosion tests.     

Reaction of neptunium with molecular and atomic oxygen: Formation and stability of surface oxides

The surface reactions of thin films of Np metal with molecular and atomic oxygen were investigated by X-ray and Ultra-Violet Photoelectron Spectroscopy (XPS and UPS, respectively). Goal of this work was to study the entire range of oxides, starting with the very early reaction stages, in presence of metal, up to the highest possible oxides, reached at saturation under highly reactive, oxidative conditions. Emphasis was given to the surface layers, whose properties often differ from the bulk, and which are directly involved in corrosion processes of solids. Molecular O₂ reacts readily with the metallic neptunium surface to form the sesquioxide and dioxide. The sesquioxide is observed as thin ‘bulk’ species of up to nine monolayers thickness. A higher oxide, identified as Np₂O₅, is formed when the NpO₂ surface is exposed to atomic oxygen. It is stable under UHV conditions up to a temperature of about 200 °C. The high oxide, Np₂O₅, is still capable of chemisorbing further oxygen. This is shown in UPS spectra by the additional O-2p line at 5-6 eV BE, superimposing onto the valence band (VB). The formation of both a lower (Np₂O₃) and higher oxide (Np₂O₅) besides the dioxide is discussed in the framework of ongoing 5f localization throughout the actinide series.     

Preparation, sintering and leaching of optimized uranium thorium dioxides

Mixed actinide dioxides are currently studied as potential fuels for several concepts associated to the fourth generation of nuclear reactors. These solids are generally obtained through dry chemistry processes from powder mixtures but could present some heterogeneity in the distribution of the cations in the solid. In this context, wet chemistry methods were set up for the preparation of U_1−xTh_xO₂ solid solutions as model compounds for advanced dioxide fuels. Two chemical routes of preparation, involving the precipitation of crystallized precursor, were investigated: on the one hand, a mixture of acidic solutions containing cations and oxalic acid was introduced in an open vessel, leading to a poorly-crystallized precipitate. On the other hand, the starting mixture was placed in an acid digestion bomb then set in an oven in order to reach hydrothermal conditions. By this way, small single-crystals were obtained then characterized by several techniques including XRD and SEM. The great differences in terms of morphology and crystallization state of the samples were correlated to an important variation of the specific surface area of the oxides prepared after heating, then the microstructure of the sintered pellets prepared at high temperature. Preliminary leaching tests were finally undertaken in dynamic conditions (i.e. with high renewal of the leachate) in order to evaluate the influence of the sample morphology on the chemical durability of the final cohesive materials.     

Effects of different surfaces on the transport and deposition of ruthenium oxides in high temperature air

In order to understand the behaviour of ruthenium oxides in the reactor coolant system during an air ingress accident, new tests were performed in the frame of the RUSET (RUthenium Separate Effect Test) experimental program. These aimed to ascertain the effects of different surfaces (quartz, stainless steel (SS), zirconium alloy, alumina, oxidised metal, and surfaces with Mo or Cs deposits) on the transport and decomposition of ruthenium oxides in air stream along the temperature gradient zone (1100–100 °C). The results demonstrated that the heterogeneous phase decomposition of RuO₃ and RuO₄ to RuO₂ is catalysed more efficiently by the quartz surface than by the SS or alumina surfaces. The presence of MoO₃ layers decreased the RuO_x precipitation extent on all investigated surfaces. The trapping effect of Cs deposit on Ru in the temperature gradient zone was proved in the case of the SS surface. On the contrary, presence of Cs precipitate on alumina and especially on quartz surfaces was found to decrease their catalytic effect on the decomposition of ruthenium oxides, and thus increased the RuO₄ concentration in the outlet air. Similarly to the effect observed for Cs deposition, the presence of other fission products in the evaporation area (at 1100 °C) decreased the partial pressure of RuO₄ in the outlet air at the SS surface and increased it at quartz and alumina surfaces. When zirconium (E110) cladding material was placed in the temperature gradient zone, no Ru transmittance occurred until the high temperature end of the zirconium tube was completely oxidised. After the intense oxidation of E110, Ru release occurred only in the presence of other fission product species. Pre-oxidation of SS surfaces in steam had no significant effect on the Ru passage.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.     

Transient gasification in an NBG-18 coolant channel of a VHTR prismatic fuel element

This paper investigates the transient gasification of NBG-18 nuclear graphite with atmospheric air ingress in a 0.8-m long coolant channel of a prismatic Very High Temperature Reactor fuel element. Analysis varied the initial graphite and air inlet temperature, T_o, from 800 to 1100 K at air inlet Reynolds number, Re_in = 5, 10 and 20. The analysis employs a Generic Interface that couples a multi-species diffusion and flow model to readout tables of the CO and CO₂ production fluxes. These fluxes are functions of the graphite local surface temperature, oxygen partial pressure and graphite weight loss fraction and calculated using a chemical-reactions kinetics model for the gasification of nuclear graphite. The analysis accounts for the heats of formation of CO and CO₂ gases, the heat conduction in the graphite sleeve, and the change in the oxygen partial pressure in the bulk gas flow mixture along the channel. Transient calculations performed up to a weight loss fraction of 0.10 at the entrance of the channel, t₁₀. They include the local graphite surface temperature and composition of bulk gas flow, the local and total graphite weight losses and the local and total production rates of CO and CO₂ gases. The heat released in the exothermic production reactions of these gases increases the local graphite surface temperature, accelerating its gasification. At the end of the calculated gasification transient, t = t₁₀, the graphite weight loss is highest at the channel entrance and decreases rapidly with axial distance into the channel, to its lowest value where oxygen in the bulk gas flow is depleted. Increasing T_o decreases t₁₀ and the total graphite loss, while increasing Re_in decreases t₁₀ but increases graphite loss.     

Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies.In this paper, two fast reactor systems are discussed—the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries.First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW_e) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems.We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating minor actinides. The tight pin lattice SFRs (P/D = 1.2) appears to have a better neutron economy than wide channel LFRs (P/D = 1.6), resulting in larger BOL actinide inventories and lower burn-up swings for LFRs. The reactivity burn-up swing of an LFR self-breeder employing BeO moderator pins could be limited to 1.3$ in 1 year. For a 600 MW_e LFR burner, LWR-to-burner support ratio was about two for (U, TRU)O₂-fuelled system, while it increased to approximately 2.8 when (Th, TRU)O₂ fuel was employed. The corresponding figures for an SFR were somewhat lower. The calculations revealed that LFRs have an advantage over SFRs in coping with the investigated severe accident initiators (ULOF, ULOHS, TLOP). The reason is better natural circulation behavior of LFR systems and the much higher boiling temperature of lead. A ULOF accident in an LFR only leads to a 220 K coolant outlet temperature increase whereas for an SFR the coolant may boil. Regarding the economics, the LFR seems to have an advantage since it does not require an intermediate coolant circuit. However, it was also proposed to avoid an intermediate coolant circuit in an SFR by using a supercritical CO₂ Brayton cycle. But in an LFR, the reduced concern about air and water ingress may decrease its cost further.     

Irradiation effects of swift heavy ions in actinide oxides and actinide nitrides: Structure and optical properties

Actinide oxides have been used as nuclear fuels in the majority of power reactors working in the world and actinide nitrides are under investigation for the fuels of the future fast neutron fission reactors developed in Forum Generation IV. Radiation damage in actinide oxides UO₂, (U_0.92Ce_0.08)O₂, and actinide nitride UN has been characterized after irradiation with swift heavy ions. Fluences up to 3 × 10¹³ ions/cm² of heavy ions (Kr 740 Mev, Cd 1 GeV) available at the CIRIL/GANIL facility were used to simulate irradiation in reactors by fission products and by neutrons. The macroscopic effects of irradiation remains very weak compared with those seen in other ceramic oxides irradiated in the same conditions: practically no swelling can be measured and no change in colour can be observed on the irradiated part of a polished face of sintered disks. The point defects in irradiated actinide compounds have been characterized by optical absorption spectroscopy in the UV–Vis–NIR wavelength range. The absorption spectra before and after irradiation are compared, and unexpected stability of optical properties during irradiation is shown. This result confirms the low rate of formation of point defects in actinide oxides and actinide nitrides under irradiation. Actinide oxides and nitrides studied are >40% ionic, and oxidation state of the actinides seems to be stable during irradiation. The small amount of point defects produced by radiation (<10¹⁶ cm⁻²) has been identified from differences between the absorption spectrum before irradiation and the one after irradiation: point defects in oxygen or nitrogen lattices can be observed respectively in oxides and nitrides (F centres), and small amounts of U⁵⁺ would be present in all compounds.     

Influence of tungsten–carbon mixed layer and irradiation defects on deuterium retention behavior in tungsten

The D₂⁺ fluence dependence on deuterium (D) retention was studied to clarify the D retention mechanism in tungsten. The additional D desorption stage was observed around 660 K in the TDS spectrum for a sample implanted with D₂⁺ up to the fluence of 10²³ D⁺ m^?2, which desorption stage was not observed the D₂⁺ implanted sample with the fluence less than 10²² D⁺ m^?2. The TEM observation showed that the highly dense voids were formed in tungsten by D₂⁺ implantation with the fluence of 10²³ D⁺ m^?2, considering that the D would be trapped by voids. To understand the D trapping by voids in C⁺ implanted tungsten, C⁺–D₂⁺ sequential implantation experiments at various C⁺ implantation temperatures were performed. It was found that the amount of D desorbed around 560 K was increased by increasing the C⁺ implantation temperature. The formation of the voids was observed with increasing the C⁺ implantation temperature by TEM, indicating that the increase of D desorption around 560 K was caused by the formation of voids. However, the desorption temperature of D trapped by voids in C⁺ implanted sample was lower than that in D₂⁺ implanted one. TEM observation and XPS measurement indicated that this difference was caused by the increase of void size and/or the presence of implanted carbon.