Investigations of the melting behaviour of the U–Zr–Fe–O system

During a severe nuclear accident, the UO₂ fuel rods, Zircaloy cladding, guide tubes, absorber and steel structural components inside the reactor pressure vessel overheat and a series of interactions between these elements and the steam atmosphere occur. These produce more heat in addition to the decay heat and result in a liquid corium of oxidic and metallic phases depending on the exact conditions and processes. A major systems resulting from this is the U–Zr–Fe–O system. High-temperature data for this system is important in order to be able to model these interactions. The Joint Research Centre, Institute for Transuranium Elements (JRC-ITU) has been examining the melting ranges for this system over the whole FeO range by means of a specialized laser flash technique that achieves very high temperatures and avoids crucible contamination. The melted zones were examined for their structure, composition and for estimation of the liquidus and solidus temperatures. The results showed that with FeO contents of over 20mol% there was a very large melting range that would permit long liquid cooling times and extend the relocation of fuel material within the reactor pressure vessel. Based on these results, the main phase regimes expected under severe accident conditions could be identified.     

Investigation of the radiation characteristics of the masonry in the ai uranium–graphite reactor

The results of investigations of the nuclear safety of the masonry in the AI uranium–graphite (Industrial Association Mayak) reactor are presented. It is concluded on the basis of these results that the masonry is nuclear safe and a radiation certificate is composed. The radiation examination made it possible to determine the level, composition, and distribution of the radioactive contamination of the masonry as well as the level and distribution of the neutron and γ radiation, and to construct a forecast of the change in the activity of radionuclides in graphite as a function of the holding time. These data are necessary for safety analysis and for making decisions about the subsequent stages of decommissioning of the reactor. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 266–269, November, 2008.     

Characteristics of hydriding and hydrogen embrittlement of the Ti–Al–Zr alloy

The characteristics of hydriding and hydrogen embrittlement of the Ti–Al–Zr alloy were evaluated. The amount of hydrogen absorbed into the alloy at 500 °C was continuously monitored using a hydrogen pressure measurement. The rate of decrease in hydrogen pressure indicated a high absorption rate of hydrogen into the alloy, following a linear rate law. X-ray diffraction studies showed the formation of δ-phase titanium hydride (TiH_1.97) after hydriding. At room temperature, the alloy showed much sensitivity to embrittlement in ductility by hydrogen. The δ-hydrides in the grain boundaries promoted the crack propagation in the presence of stress, leading to the cleavage failure mode. However, the tensile strengths were almost independent of the hydrogen content up to 1174 ppm. It is thus concluded that the δ-hydride acts to decrease the ductility without affecting tensile strengths.     

Calculation of the Reliability Indicators of the System Protected Object–Control and Protection System

A mathematical model is proposed for the reliability of an automated technological system consisting of a protected object, an automatic control system, and a safety system. The model is a superposition of alternating renewal processes. Relations are obtained for the probability of the first accident in the interval [0, t] and the probability of an accident both for nonrenewable and renewable components of the automated technological system. A study of the asymptotic properties of the mathematical model made it possible to write down the indicated reliability indicators in a simple form, specifically, in terms of stationary coefficients of readiness of the system components. Particular cases are examined, and the corresponding relations are presented for them.     

Thermodynamic assessments of the Al–Th and Th–Zn systems

The phase diagrams of the Al–Th and Th–Zn systems have been evaluated by using the Calculation of Phase Diagrams (CALPHAD) method with the experimental data including the phase equilibria and thermodynamic properties. The Gibbs free energies of the liquid, bcc and hcp phases were described by the subregular solution model with the Redlich–Kister equation, and those of the stoichiometric compounds of the Th₂Al, Th₃Al₂, ThAl, Th₂Al₃, ThAl₂, ThAl₃, Th₂Al₇, Th₂Zn, ThZn₂, ThZn₄ and Th₂Zn₁₇ were described by the two-sublattice model. The calculated phase equilibria and thermodynamic properties are in good agreement with the experimental data.     

Studies on the Characteristic of Zirconium–Tritium Reaction

The p–c–T curves of tritium absorption and desorption of zirconium were measured using the method of step equilibrium by stepping up the tritium quantity on an experimental apparatus of metal hydride. The p–c–T curves for tritium have one plateau at temperature range from 450 to 500°C and two plateaus at temperature above 600°C. The thermodynamic parameters of the different phases were determined according to the van’t Hoff equation. The hysteresis effect was observed in reversible process of tritium absorption and desorption of zirconium on our experimental condition. The tritium absorption behavior by zirconium in the temperature range from 450 to 620°C and desorption behavior of zirconium in the temperature range from 775 to 875°C have been investigated. A method of the reaction rate analysis was proposed and examined for determining the rate constant. The apparent activation energy obtained by this analysis for the absorption and the desorption were (−16.8 ± 0.8) kJ·mol⁻¹ and (57.7 ± 1.6) kJ·mol⁻¹, respectively.     

Thermodynamic modelling of the C–U and B–U binary systems

The thermodynamic modelling of the carbon–uranium (C–U) and boron–uranium (B–U) binary systems is being performed in the framework of the development of a thermodynamic database for nuclear materials, for increasing the basic knowledge of key phenomena which may occur in the event of a severe accident in a nuclear power plant. Applications are foreseen in the nuclear safety field to the physico-chemical interaction modelling, on the one hand the in-vessel core degradation producing the corium (fuel, zircaloy, steel, control rods) and on the other hand the ex-vessel molten corium–concrete interaction (MCCI). The key O–U–Zr ternary system, previously modelled, allows us to describe the first interaction of the fuel with zircaloy cladding. Then, the three binary systems Fe–U, Cr–U and Ni–U were modelled as a preliminary work for modelling the O–U–Zr–Fe–Cr–Ni multicomponent system, allowing us to introduce the steel components in the corium. In the existing database (TDBCR, thermodynamic data base for corium), Ag and In were introduced for modelling AIC (silver–indium–cadmium) control rods which are used in French pressurized water reactors (PWR). Elsewhere, B₄C is also used for control rods. That is why it was agreed to extend in the next years the database with two new components, B and C. Such a work needs the thermodynamic modelling of all the binary and pseudo-binary sub-systems resulting from the combination of B, B₂O₃ and C with the major components of TDBCR, O–U–Zr–Fe–Cr–Ni–Ag–In–Ba–La–Ru–Sr–Al–Ca–Mg–Si + Ar–H. The critical assessment of the very numerous experimental information available for the C–U and B–U binary systems was performed by using a classical optimization procedure and the Scientific Group Thermodata Europe (SGTE). New optimized Gibbs energy parameters are given, and comparisons between calculated and experimental equilibrium phase diagrams or thermodynamic properties are presented. The self-consistency obtained is quite satisfactory.     

Analytical developments in the Wong–Fung–Tam–Gao radiation model of thermal diffusivity

When the thermal diffusivity, χ, of a thin film on a substrate is measured by means of the mirage method, the photothermal deflection of the probe beam is determined by the heat radiation field contributed by the film and the substrate, heated by the pump beam. A two-dimensional algorithm is here presented in order to deduce the measure of the diffusivities of the film and the substrate in one set of mirage detection from the experimental data.     

The influences of Pu and Zr on the melting temperatures of the UO2–PuO2–ZrO2 pseudo-ternary system

Specimens of (U, Pu, Zr)O₂ were prepared as simulated corium debris that were assumed like debris generated in the severe accident of the Fukushima Daiichi Nuclear Power Plant and their melting temperatures were measured by the thermal arrest technique in order to evaluate the influence of plutonium and zirconium content on the melting temperature of the corium debris. From the evaluation, it was found that the influence of zirconium on the melting temperatures of both (U, Pu, Zr)O₂ and (U, Zr)O₂ was similar and that the melting temperature of (U, Pu, Zr)O₂ had a local maximum value in the Pu-content between 0 and 20 mol%. The UO₂–PuO₂–ZrO₂ pseudo-ternary phase diagram at 2900 and 3000 K was evaluated from the present experimental results and previously reported results.     

Towards the development of technical specifications for the production of lithium–lead alloys for the ITER HCLL TMB

The ITER and DEMO projects are developing new Test Blanket Modules (TBM), where the Pb–Li alloy plays a key role in the new commercial fusion reactors functionality. The Breeding Blanket (BB) has to perform several functions which are essential for the reactor operation. The HCLL TBM is one of the Breeding Blanket concepts to be tested in ITER. It is cooled by He and uses the eutectic liquid metal LLE (Lithium–Lead Eutectic) as breeder material (enriched at 90% in ⁶Li).Pb–Li eutectic alloy has no known uses outside of fusion technology, so the available databases of this material are currently incomplete. It is very important, within the material specifications, to have a complete characterization in order to define their chemical and physical properties, because any variation in the alloy composition has significant consequences in their behaviour, and therefore in their regenerative function inside the blanket.The chemical characterization methodology developed and presented in this paper (useful for both Pb–Li alloys as any Pb alloy) is a key tool that allows performing standard quality control procedures for base material and/or monitoring the alloy during the reactor operation. This report provides a procedure to perform a wide material chemical characterization, assessing the concentrations of major elements, as well as a review of trace level elements that can be found both in the eutectic alloy and in starting materials. In this determination plays an important role the ICP-MS technique because, as a highly sensitive technique, allows very low detection limits.     

A Krylov–Schur solution of the eigenvalue problem for the neutron diffusion equation discretized with the Raviart–Thomas method

Mixed-dual formulations of the finite element method were successfully applied to the neutron diffusion equation, such as the Raviart–Thomas method in Cartesian geometry and the Raviart–Thomas–Schneider in hexagonal geometry. Both methods obtain system matrices which are suitable for solving the eigenvalue problem with the preconditioned power method. This method is very fast and optimized, but only for the calculation of the fundamental mode. However, the determination of non-fundamental modes is important for modal analysis, instabilities, and fluctuations of nuclear reactors. So, effective and fast methods are required for solving eigenvalue problems. The most effective methods are those based on Krylov subspaces projection combined with restart, such as Krylov–Schur. In this work, a Krylov–Schur method has been applied to the neutron diffusion equation, discretized with the Raviart–Thomas and Raviart–Thomas–Schneider methods.     

Engineering-scale test on the thermal–hydro–mechanical behaviors in the clay barrier of a HLW repository

The thermo–hydro–mechanical (T–H–M) behaviors of a clay barrier are of importance from a performance and safety viewpoint of the engineered barrier system (EBS) for a high-level waste (HLW) repository. An engineering-scale test was carried out to investigate the T–H–M behaviors in the buffer of the Korean reference disposal system (KRS). The test started on May 31, 2005 and is still in operation. The experimental data obtained allowed a preliminary and qualitative interpretation of the T–H–M behavior in bentonite blocks. The temperature was higher as it became closer to the heater, while it became lower as it was farther away from the heater. The water content had a higher value in the part close to the hydration surface than that in the heater part. The relative humidity data suggested that a hydration of the bentonite blocks might occur by different drying–wetting processes, depending on their position. The total pressure was continuously increased by the evolution of the saturation front in the bentonite blocks and thereby the swelling pressure. There was also a contribution of the thermal expansion of the bentonite blocks near the heater and the capillary force in the dry bentonite blocks which the water did not reach from the hydration surface.     

Coupled model analysis of the ablative Rayleigh–Taylor instability

The weakly nonlinear stage of the ablative Rayleigh–Taylor instability(ARTI) is investigated by expanded hydrodynamic equations in which the third-order corrections of the two-mode perturbations are considered. In the present coupling model, two linear perturbations are simultaneously added near the ablation front at the initial moment, and we have derived the first three coupling harmonics. Furthermore, the coupling model analysis is studied via direct numerical simulation as well. When the ori...     

On the electrochemical formation of Pu–Al alloys in molten LiCl–KCl

Properties of Pu–Al alloys were investigated in connection with development of pyrochemical methods for reprocessing of spent nuclear fuel. Electroseparation techniques in molten LiCl–KCl are being developed in ITU to group-selectively recover actinides from the mixture with fission products. In the process, actinides are electrochemically reduced on solid aluminium cathodes, forming solid actinide–aluminium alloys. This article is focused on electro-chemical characterisation of Pu–Al alloys in molten LiCl–KCl, on electrodeposition of Pu on solid Al electrodes and on determination of chemical composition and structure of the formed alloys. Cyclic voltammetry and chronopotentiometry were used to study Pu–Al alloys in the temperature range 400–550 °C. Pu is reduced to metal in one reduction step Pu³⁺/Pu⁰ on an inert W electrode. On a reactive Al electrode, the reduction of Pu³⁺ to Pu⁰ occurs at a more positive potential due to formation of Pu–Al alloys. The open circuit potential technique was used to identify the alloys formed. Stable deposits were obtained by potentiostatic electrolyses of LiCl–KCl–PuCl₃ melts on Al plates. XRD and SEM–EDX analyses were used to characterise the alloys, which were composed mainly of PuAl₄ with some PuAl₃. In addition, the preparation of PuCl₃ containing salt by carbochlorination of PuO₂ is described.     

Behavior of the Pb–Li alloy impurities by ICP-MS

The ITER and DEMO projects are developing new test blanket modules (TBM), such as HCLL where the Li–Pb alloy plays a key role in the new commercial fusion reactors functionality. Lithium–lead eutectic alloy has no known uses outside of fusion technology, so the available databases of this material are currently incomplete. It is very important, within the material specifications, to have a complete characterization in order to define their chemical and physical properties, because any variation in the alloy composition has significant consequences in their behavior, and therefore in their regenerative function inside the blanket.This report provides a procedure to perform a wide material characterization, assessing the concentrations of major elements, as well as a review of trace level impurities that can be found both in the eutectic alloy as in starting materials. In this determination inductively coupled plasma mass spectrometry (ICP-MS) technique plays an important role, because as a highly sensitive technique it allows very low detection limits.     

Study of spatial resolution of the associated alpha particle imaging–time-of-flight method

Associated alpha particle imaging based on the time-of-flight(API-TOF) technique is an advanced neutron analysis method, which is capable of discriminating material nuclides and three-dimensional imaging of the spatial distribution of material nuclei. In this paper, the spatial resolution of API-TOF and its effects are studied using mathematical analysis and Monte Carlo numerical simulation. The results can provide guidance and assist in designing of API-TOF detection devices. First, a mathematical analysis of the imaging principles of the API-TOF was carried out, and the calculation formulas of the spatial resolution of API-TOF were deduced. Next, the relationship between the device layout and the spatial resolution of the API-TOF detection device was studied. The concept of a typical API-TOF detection device with an optimized structure was proposed. Then, the spatial distribution of the spatial resolution of the typical API-TOF detection device was analyzed, and the effects of the time resolution and the neutron emission angle resolution on the spatial resolution were studied. The results show that spatial resolutions better than 1 cm can be achieved by improving the time resolution and the neutron emission angle resolution to appropriate levels. Finally, a Monte Carlo numerical simulation program was developed for the study of the APITOF and was used to calculate the spatial resolutions of the API-TOF. The comparison of the results shows that thespatial resolutions calculated based on the Monte Carlo numerical simulation are in good agreement with those calculated based on the mathematical analysis. This verifies the mathematical analysis and the evaluation of the effects of the spatial resolution of the API-TOF in this study.     

Effect of iodine partial pressure on the iodine stress corrosion cracking behaviors of Zr–Sn–Nb alloy using the ring tensile test

In order to evaluate the effect of iodine partial pressure on the iodine stress corrosion cracking (I-SCC) behaviors of a Zr–Sn–Nb alloy, ring tensile tests were conducted at 350 °C and in atmosphere without iodine and with iodine partial pressure of 10², 10³, and 10⁴ Pa, respectively. Results show that the maximum load, fracture displacement, tensile strength, and fracture energy of the Zr–Sn–Nb specimens decrease monotonically with the increase of iodine partial pressure. The fracture morphology of specimen with 10² Pa iodine exhibits two different fracture regions. One is the mixture form of ductile and brittle fracture, and the other is only ductile fracture. For the specimen with 10³ Pa, stair-shaped fracture surface is formed as a result of the alternative propagation of transgranular cracks and intergranular cracks. The critical iodine partial pressure of the Zr–Sn–Nb alloys is lower than 10² Pa under the present conditions.     

Influence of precipitate density on the nodular corrosion resistance of Zr–Sn–Fe–Cr alloys at 500°C

In a Zr–1.3% Sn base alloy, both the addition of increasing amounts of iron and chromium, conserving a constant Fe/Cr ratio, and the reduction of the cumulative annealing parameter ΣA have beneficial effects on the corrosion resistance in 500°C steam. It is shown that these two observations can be rationalized by considering that the important metallurgical factor is the number of precipitates per unit volume rather than their size.     

Three-dimensional FE analysis of the thermal–mechanical behaviors in the nuclear fuel rods

In order to implement numerical simulation of the thermal–mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal–mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal–mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm³, the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm² K to 0.01 W/mm² K, while increases up to 54.7% when h decreases from 0.01 W/mm² K to 0.005 W/mm² K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal–mechanical behaviors in the fuel rod; when the gap size varies from 0.03 mm to 0.1 mm, the highest temperature in the pellet increases by 19.7%, and the maximum first principal stress at the outer pellet surface decreases by 17.4%; it is critical to optimize the gap size in order to reduce the pellet–cladding mechanical interaction and avoid their contact at early stage. This study lays a foundation for the further research on the irradiation-induced mechanical behaviors in the fuel rods.