TiN-based coatings on fuel cladding tubes for advanced nuclear reactors

Titanium nitride (TiN) thin films are coated on HT-9 and MA957 fuel cladding tubes and bars to explore their mechanical strength, thermal stability, diffusion barrier properties, and thermal conductivity properties. The ultimate goal is to implement TiN as an effective diffusion barrier to prevent the inter-diffusion between the nuclear fuel and the cladding material, and thus lead to a longer lifetime of the cladding tubes. Mechanical tests including hardness and scratch tests for the samples before and after thermal cycle tests show that the films have a high hardness of 28GPa and excellent adhesion properties despite the thermal treatment. Thermal conductivity measurements demonstrate that the thin TiN films have very minimal impact on the overall thermal conductivity of the MA957 and HT-9 substrates, i.e., the thermal conductivity of the uncoated HT-9 and MA957 substrates was 26.25 and 28.44 W m-1 K-1, and that of the coated ones was 26.21 and 28.38W m-1 K-1, respectively. A preliminary Ce diffusion test on the couple of Ce/TiN/HT-9 suggests that TiN has excellent material compatibility and good diffusion barrier properties.

High stress monitoring of prestressing tendons in nuclear concrete vessels using fibre-optic sensors

Maintaining the structural health of prestressed concrete nuclear containments is a key element in ensuring nuclear reactors are capable of meeting their safety requirements. This paper discusses the attachment, fabrication and characterisation of optical fibre strain sensors suitable for the prestress monitoring of irradiated steel prestressing tendons. The all-metal fabrication and welding process allowed the instrumented strand to simultaneously monitor and apply stresses up to 1300 MPa (80% of steel's ultimate tensile strength). There were no adverse effects to the strand's mechanical properties or integrity. After sensor relaxation through cyclic stress treatment, strain transfer between the optical fibre sensors and the strand remained at 69%. The fibre strain sensors could also withstand the non-axial forces induced as the strand was deflected around a 4.5 m bend radius. Further development of this technology has the potential to augment current prestress monitoring practices, allowing distributed measurements of short- and long-term prestress losses in nuclear prestressed-concrete vessels. © 2014 Elsevier B.V.

Experiments to assess the transport of fibrous insulation debris

Loss of coolant accidents (LOCA) in the primary cooling circuit of a nuclear reactor may result in damage to insulation materials that are located near to the leak. The insulation materials released may compromise the operation of the emergency core cooling system (ECCS). Insulation material in the form of mineral wool fibre agglomerates (MWFA) maybe transported to the containment sump strainers mounted at the inlet of the emergency cooling pumps, where the insulation fibres may block or penetrate the strainers. In addition to the impact of MWFA on the pressure drop across the strainers, corrosion products formed over time may also accumulate in the fibre cakes on the strainers, which can lead to a significant increase in the strainer pressure drop and result in cavitation in the ECCS. Thus, knowledge of transport characteristics of the damaged insulation materials in various scenarios is required to help plan for the long-term operability of nuclear reactors, which undergo LOCA. An experimental and theoretical study performed by the Helmholtz-Zentrum Dresden-Rossendorf and the Hochschule Zittau/Görlitz1 is investigating the phenomena that maybe observed in the containment vessel during a LOCA. The study entails the generation of fibre agglomerates, the determination of their transport properties in single and multi-effect experiments and the long-term effect that corrosion of the containment internals by the coolant has on the strainer pressure drop. The focus of this presentation is on the experiments performed that characterize the horizontal transport of MWFA, whereas the corresponding CFD simulations are described in an accompanying contribution (see abstract of Cartland Glover et al.). The experiments were performed a racetrack type channel that provided a near uniform horizontal flow. The channel is 0.1 wide by 1.2 m high with a straight length of 5 m and two bends of 0.5 m. The measurement techniques include particle imaging (both wide-angle and macro lens), concurrent particle image velocimetry, ultravelocimetry, laser detection sensors to sense the presence of absence of MWFA and pertinent measurements of the MWFA concentration and quiescent settling characteristics. The transport of the MWFA was observed at velocities of 0.1 and 0.25 m s-1 to verify numerical model behaviour in and just beyond expected velocities in the containment sump of a nuclear reactor.

Sump simulations

Mineral wool insulation material applied to the primary cooling circuit of a nuclear reactor maybe damaged in the course of a loss of coolant accident (LOCA). The insulation material released by the leak may compromise the operation of the emergency core cooling system (ECCS), as it maybe transported together with the coolant in the form of mineral wool fiber agglomerates (MWFA) suspensions to the containment sump strainers, which are mounted at the inlet of the ECCS to keep any debris away from the emergency cooling pumps. In the further course of the LOCA, the MWFA may block or penetrate the strainers. In addition to the impact of MWFA on the pressure drop across the strainers, corrosion products formed over time may also accumulate in the fiber cakes on the strainers, which can lead to a significant increase in the strainer pressure drop and result in cavitation in the ECCS. Therefore, it is essential to understand the transport characteristics of the insulation materials in order to determine the long-term operability of nuclear reactors, which undergo LOCA. An experimental and theoretical study performed by the Helmholtz-Zentrum Dresden-Rossendorf and the Hochschule Zittau/Görlitz1 is investigating the phenomena that maybe observed in the containment vessel during a primary circuit coolant leak. The study entails the generation of fiber agglomerates, the determination of their transport properties in single and multi-effect experiments and the long-term effects that particles formed due to corrosion of metallic containment internals by the coolant medium have on the strainer pressure drop. The focus of this presentation is on the numerical models that are used to predict the transport of MWFA by CFD simulations in the containment sump. Two dispersed phases were conditions to determine the influence of entrained air from a jet on the transport of fibre agglomerates through the sump. The strainer model of A. Grahn was implemented to observe the impact that the accumulation of the fibres have on the pressure drop across the strainers. The geometry considered is similar to the containment sump configurations found in Nuclear Power Plants.

Simulations of agglomerate sedimentation and suspension

Mineral wool insulation material applied to the primary cooling circuit of a nuclear reactor maybe damaged in the course of a loss of coolant accident (LOCA). The insulation material released by the leak may compromise the operation of the emergency core cooling system (ECCS), as it maybe transported together with the coolant in the form of mineral wool fiber agglomerates (MWFA) suspensions to the containment sump strainers, which are mounted at the inlet of the ECCS to keep any debris away from the emergency cooling pumps. In the further course of the LOCA, the MWFA may block or penetrate the strainers. In addition to the impact of MWFA on the pressure drop across the strainers, corrosion products formed over time may also accumulate in the fiber cakes on the strainers, which can lead to a significant increase in the strainer pressure drop and result in cavitation in the ECCS. Therefore, it is essential to understand the transport characteristics of the insulation materials in order to determine the long-term operability of nuclear reactors, which undergo LOCA. An experimental and theoretical study performed by the Helmholtz-Zentrum Dresden-Rossendorf and the Hochschule Zittau/Görlitz is investigating the phenomena that maybe observed in the containment vessel during a primary circuit coolant leak. The study entails the generation of fiber agglomerates, the determination of their transport properties in single and multi-effect experiments and the long-term effects that particles formed due to corrosion of metallic containment internals by the coolant medium have on the strainer pressure drop. The focus of this presentation is on the numerical models that are used to predict the transport of MWFA by CFD simulations. A number of pseudo-continuous dispersed phases of spherical wetted agglomerates can represent the MWFA. The size, density, the relative viscosity of the fluid-fiber agglomerate mixture and the turbulent dispersion all affect how the fiber agglomerates are transported. In the cases described here, the size is kept constant while the density is modified. This definition affects both the terminal velocity and volume fraction of the dispersed phases. Application of such a model to sedimentation in a quiescent column and a horizontal flow are examined. The scenario also presents the suspension and horizontal transport of a single fiber agglomerate phase in a racetrack type channel.

The automation of resonant thermometer probe measurements

An ultrasonic thermometer has been developed for high temperature measurement over a wide temperature range. It is particularly suitable for use in measuring nuclear fuel rod centerline temperatures in advanced liquid metal and high flux nuclear reactors. The thermometer which was designed to determine fuel temperature up to the fuel melting point, utilizes the temperature dependence of the ultrasonic propagation velocity (related to the elastic modulus} in a thin rod sensor as the temperature transducing mechanism. A pulse excitation technique has been used, where the mechanical resonator at the remote end of the acoustic·line is madto vibrate. Its natural frequency is proportional to the ultrasonic velocity in the material. This is measured by the electronic instrumentation and enables a frequency­ temperature or period-temperature calibration to be obtained. A completely digital automatic instrument has been designed, constructed and tested to track the resonance frequency of the temperature sensors. It operates smoothly over a frequency range of about 30%, more than the maximum working range of most probe materials. The control uses the basic property of a resonator that the stored energy decays exponentially at the natural frequency of the resonator.The operation of the electronic system is based on a digital multichannel transmitter that is capable of operating with a predefined number of cycles in the burst. this overcomes a basic defect in the previous deslgn where the analogue time-delayed circuits failed to hold synchronization and hence automatic control could be lost. Development of a particular type of temperature probe, that is small enough to fit into a standard 2 mm reactor tube has made the ultrasonic thermometer a practicable device for measuring fuel temperature. The bulkiness of previous probes has been overcome, the new design consists of a tuning fork, integral with a 1mm line, while maintaining a frequency of no more than 100 kHz. A magnetostrictive rod, acoustically matched to the probe is used to launch and receive the acoustic oscillations. This requires a magnetic bias and the previously used bulky magnets have been replaced by a direct current coil. The probe is supported by terminating the launcher with a short heavy isolating rod which can be secured to the reactor structure. This support, the bias and launching coil and the launcher are made up into a single compact unit. On the material side an extensive study of a wide range of refractory materials identified molybdenum, iridium, rhenium and tungsten as satisfactory for a number of applications but mostly exhibiting to some degree a calibration drift with thermal cycling. When attention was directed to ceramic materials, Sapphire (single crystal alumina) was found to have numerous advantages, particularly in respect of stability of calibration which remained with ±2°C after many cycles to 1800oC. Tungsten and thoriated tungsten (W - 2% Tho2) were also found to be quite satisfactory to 1600oC, the specification for a Euratom application.

Transfer coefficients over inclined tube banks

The thesis presents experimental results for shell-side transfer coefficients and pressure drops across four different tube banks, using small-scale models, with yawed tubes, as found in many types of heat exchangers, boilers and nuclear reactors. The tube banks investigated have a staggered tube layout on a rotated square pitch, with a 1.25 pitch-to-diameter ratio. The angle of attack was varied between 45o and 90o. An extensive range of Reynolds number, i.e. 0.5. to 12,600, covering so-called laminar, transition and turbulent flows, was investigated. A diffusion-controlled electrochemical mass transfer technique has been employed to measure mass transfer coefficients. The heat transfer coefficients may be then readily obtained from the mass transfer values by applying the well-established Chilton-Colburn analogy. The results for the normal tube bank, which forms the base case for the study on inclined tube banks, show close agreement with previous work. The transfer coefficients and pressure drops of the inclined tube banks are compared with results from the ideal normal tube bank to examine the effect of inclination angle on heat transfer and pressure drop variations. The variation of the transfer coefficients row-by-row and the entrance and exit effects have also been investigated. An auxilary investigation has been carried out on the role of natural convection. A preliminary correlation of transfer coefficients and pressure drops against the variation in the yaw angle has been attempted. The results are discussed in the light of the few existing theoretical treatments and experimental data for these situations, and recommendations made for future work.

Nano-indentation measurement of oxide layers formed in LBE on F/M steels

Ferritic/martensitic (F/M) steels (T91, HT-9, EP 823) are candidate materials for future liquid lead or lead bismuth eutectic (LBE) cooled nuclear reactors. To understand the corrosion of these materials in LBE, samples of each material were exposed at 535 °C for 600 h and 200 h at an oxygen content of 10 wt%. After the corrosion tests, the samples were analyzed using SEM, WDX and nano-indentation in cross section. Multi-layered oxide scales were found on the sample surfaces. The compositions of these oxide layers are not entirely in agreement with the literature. The nano-indentation results showed that the E-modulus and hardness of the oxide layers are significantly lower than the values for dense bulk oxide materials. It is assumed that the low values stem from high porosity in the oxide layers. Comparison with in-air oxidized steels show that the E-modulus decreases with increasing oxide layer thickness. © 2008 Elsevier B.V. All rights reserved.

Numerical models used for the modelling of the transport of fibrous insulation debris

Mineral wool insulation material applied to the primary cooling circuit of a nuclear reactor maybe damaged in the course of a loss of coolant accident (LOCA). The insulation material released by the leak may compromise the operation of the emergency core cooling system (ECCS), as it maybe transported together with the coolant in the form of mineral wool fiber agglomerates (MWFA) suspensions to the containment sump strainers, which are mounted at the inlet of the ECCS to keep any debris away from the emergency cooling pumps. In the further course of the LOCA, the MWFA may block or penetrate the strainers. In addition to the impact of MWFA on the pressure drop across the strainers, corrosion products formed over time may also accumulate in the fiber cakes on the strainers, which can lead to a significant increase in the strainer pressure drop and result in cavitation in the ECCS. Therefore, it is essential to understand the transport characteristics of the insulation materials in order to determine the long-term operability of nuclear reactors, which undergo LOCA. An experimental and theoretical study performed by the Helmholtz-Zentrum Dresden-Rossendorf and the Hochschule Zittau/Görlitz1 is investigating the phenomena that maybe observed in the containment vessel during a primary circuit coolant leak. The study entails the generation of fiber agglomerates, the determination of their transport properties in single and multi-effect experiments and the long-term effects that particles formed due to corrosion of metallic containment internals by the coolant medium have on the strainer pressure drop. The focus of this presentation is on the numerical models that are used to predict the transport of MWFA by CFD simulations. A number of pseudo-continuous dispersed phases of spherical wetted agglomerates can represent the MWFA. The size, density, the relative viscosity of the fluid-fiber agglomerate mixture and the turbulent dispersion all affect how the fiber agglomerates are transported. In the cases described here, the size is kept constant while the density is modified. This definition affects both the terminal velocity and volume fraction of the dispersed phases. Only one of the single effect experimental scenarios is described here that are used in validation of the numerical models. The scenario examines the suspension and horizontal transport of the fiber agglomerates in a racetrack type channel. The corresponding experiments will be described in an accompanying presentation (see abstract of Seeliger et al.).

Ion irradiation effects in nanocrystalline TiN coatings

Nitride materials and coatings have attracted extensive research interests for various applications in advanced nuclear reactors due to their unique combination of physical properties, including high temperature stability, excellent corrosion resistance, superior mechanical property and good thermal conductivity. In this paper, the ion irradiation effects in nanocrystalline TiN coatings as a function of grain size are reported. TiN thin films (thickness of 100 nm) with various grain sizes (8-100 nm) were prepared on Si substrates by a pulsed laser deposition technique. All the samples were irradiated with He ions to high fluences at room temperature. Transmission electron microscopy (TEM) and high resolution TEM on the ion-irradiated samples show that damage accumulation in the TiN films reduces as the grain size reduces. Electrical resistivity of the ion-irradiated films increases slightly compared with the as-deposited ones. These observations demonstrate a good radiation-tolerance property of nanocrystalline TiN films. © 2007 Elsevier B.V. All rights reserved.

Verification and validation of numerical models of the transport of insulation debris

A combination of the two-fluid and drift flux models have been used to model the transport of fibrous debris. This debris is generated during loss of coolant accidents in the primary circuit of pressurized or boiling water nuclear reactors, as high pressure steam or water jets can damage adjacent insulation materials including mineral wool blankets. Fibre agglomerates released from the mineral wools may reach the containment sump strainers, where they can accumulate and compromise the long-term operation of the emergency core cooling system. Single-effect experiments of sedimentation in a quiescent rectangular column and sedimentation in a horizontal flow are used to verify and validate this particular application of the multiphase numerical models. The utilization of both modeling approaches allows a number of pseudocontinuous dispersed phases of spherical wetted agglomerates to be modeled simultaneously. Key effects on the transport of the fibre agglomerates are particle size, density and turbulent dispersion, as well as the relative viscosity of the fluid-fibre mixture.

Chemical reactions at nuclear fuel-clad interfaces

This thesis presents a study of the chemical reactions that may occur at the fuel- clad interfaces of fuel elements used in advanced gas-coooled reactors (A.G.R.) The initial investigation involved a study of the inner surfaces of irradiated stainless steel clad and evidence was obtained to show that fission products, in particular tellerium, were associated with reaction products on these surfaces. An accelerated rate of oxidation was observed on the inner surfaces of a failed A.G.R. fuel pin. It is believed that fission product caesium was responsible for this enhancement. A fundamental study of the reaction between 20%Cr/25%Ni/niobium stabilised stainless steel and tellerium was then undertaken over the range 350 - 850 degrees C. Reaction occurred with increasing rapidity over this range and long term exposure at ≤ 750 degrees resulted in intergranular attack of the stainless steel and chromium depletion. The reaction on unoxidised steel surfaces involved the formation of an initial iron-nickel-tellerium layer which subsequently transformed to a chromium telluride product during continued exposure. The thermodynamic stabilities of the steel tellurides were determined to be chromium telluride > nickel telluride > iron telluride. Oxidation of the stainless steel surface prior to tellerium exposure inhibited the reaction. However reaction did occur in regions where the oxide layer had either cracked or spalled.