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[en] The performance of amended micrometric iron particles (Fe0) in trimetallic systems for the removal of aqueous diclofenac (DF) was investigated. Tested additive metals were Pd, Cu and Ni. Trimetallic systems were shown superior to Fe0 for aqueous DF removal both under oxic and anoxic conditions. The extent of DF removal varied with (i) the combination of the additives, (ii) the order of plating on Fe0 surface, (iii) the availability of dissolved oxygen and (iv) the presence of MnO2. In comparison with previous results for bimetallic systems: (i) DF reduction products were observed under anoxic conditions for all trimetallic systems and only in some bimetallic systems; and (ii) under oxic conditions, DF reaction products were only observed in CuPdFe system. This was accredited to the intrinsic and electronic properties of the involved metals. Rapid DF removal was mainly attributed to the capacity of additive metals to accelerate Fe0 corrosion and thus producing more atomic hydrogen within the porous iron oxide layers. This promoted catalytic hydrodehalogenation was clearly apparent in systems where reductive transformation products were identified (e.g. PdCuFe). Recycling experiments using PdCuFe showed a continuous reactivity even after 4 successive cycles under anoxic conditions, however, under oxic conditions, a serious reactivity loss was noticed.(author)
[en] Joints between carbon steel and Alloy 600, containing Alloy 82 weld metal, were exposed to two steam-hydrogen environments considered to simulate exposure to primary water conditions in a Pressurized Water Reactor (PWR) or Canada Deuterium Uranium (CANDU) reactor. The welds were found to have elevated and variable iron contents due to dilution by carbon steel during welding. This gave the Alloy 82 weld, near the inner surface of the component, an iron content approaching that of Alloy 800. A potentially protective external iron oxide film formed on the inner surface of the weld. However, the chromium content throughout the weld is below that which would form an external chromium oxide. The results indicate that low chromium content causes internal oxidation throughout the weld and potentially below the external iron oxide which could lead to Primary Water Stress Corrosion Cracking (PWSCC). (author)
[en] This thesis is dedicated to the development of a Monte Carlo neutron transport solver based on the subgroup (or multiband) method. In this formalism, cross sections for resonant isotopes are represented in the form of probability tables on the whole energy spectrum. This study is intended in order to test and validate this approach in lattice physics and criticality-safety applications. The probability table method seems promising since it introduces an alternative computational way between the legacy continuous-energy representation and the multigroup method. In the first case, the amount of data invoked in continuous-energy Monte Carlo calculations can be very important and tend to slow down the overall computational time. In addition, this model preserves the quality of the physical laws present in the ENDF format. Due to its cheap computational cost, the multigroup Monte Carlo way is usually at the basis of production codes in criticality-safety studies. However, the use of a multigroup representation of the cross sections implies a preliminary calculation to take into account self-shielding effects for resonant isotopes. This is generally performed by deterministic lattice codes relying on the collision probability method. Using cross-section probability tables on the whole energy range permits to directly take into account self-shielding effects and can be employed in both lattice physics and criticality-safety calculations. Several aspects have been thoroughly studied: • The consistent computation of probability tables with a energy grid comprising only 295 or 361 groups. The CALENDF moment approach conducted to probability tables suitable for a Monte Carlo code. • The combination of the probability table sampling for the energy variable with the delta-tracking rejection technique for the space variable, and its impact on the overall efficiency of the proposed Monte Carlo algorithm. • The derivation of a model for taking into account anisotropic effects of the scattering reaction consistent with the subgroup method. In this study, we generalize the Discrete Angle Technique, already proposed for homogeneous, multigroup cross sections, to isotopic cross sections on the form of probability tables. In this technique, the angular density is discretized into probability tables. Similarly to the cross-section case, a moment approach is used to compute the probability tables for the scattering cosine. • The introduction of a leakage model based on the B1 fundamental mode approximation. Unlike deterministic lattice packages, most Monte Carlo-based lattice physics codes do not include leakage models. However the generation of homogenized and condensed group constants (cross sections, diffusion coefficients) require the critical flux. This project has involved the development of a program into the DRAGON framework, written in Fortran 2003 and wrapped with a driver in C, the GANLIB 5. Choosing Fortran 2003 has permitted the use of some modern features, such as the definition of objects and methods, data encapsulation and polymorphism. The validation of the proposed code has been performed by comparison with other numerical methods: • The continuous-energy Monte Carlo method of the SERPENT code. • The Collision Probability (CP) method and the discrete ordinates (SN) method of the DRAGON lattice code. • The multigroup Monte Carlo code MORET, coupled with the DRAGON code. Benchmarks used in this work are representative of some industrial configurations encountered in reactor and criticality-safety calculations: • Pressurized Water Reactors (PWR) cells and assemblies. • Canada-Deuterium Uranium Reactors (CANDU-6) clusters. • Critical experiments from the ICSBEP handbook (International Criticality Safety Benchmark Evaluation Program). (author)
[en] Study of quench cooling is very important in nuclear reactor safety for limiting the extent of core damage during the early stages of severe accidents after Loss of Coolant Accidents (LOCA). Quench of a hot dry surface involves the rapid decrease in surface temperature resulting from bringing the hot surface into sudden contact with a coolant at a lower temperature. The quench temperature is the onset of the rapid decrease in the surface temperature and corresponds to the onset of destabilization of a vapor film that exists between the hot surface and the coolant. Re-wetting the surface is the establishment of direct contact between the surface and the liquid at the so-called re-wetting temperature. Re-wetting is characterized by the formation of a wet patch on the surface which then spreads to cover the entire surface. Situations involving quench andre-wetting heat transfer are encountered in a number of postulated accidents in Canada Deuterium Uranium (CANDU) reactors, such as re-wetting of a hot dry calandria tube in a critical break LOCA. This accident results in high heat transfer from the calandria tube to the surrounding moderator liquid which can cause the calandria tube surface to experience dryout and a subsequent escalation in the surface temperature. If the calandria tube temperature is not reduced by initiation of quench heat transfer, then this may lead to subsequent fuel channel failure. In literature very limited knowledge is available on quench and re-wetting of hot curved surfaces like the calandria tubes. In this study, a Water Quench Facility (WQF) has been constructed and a series of experiments were conducted to investigate the quench and re-wetting of hot horizontal tubes by a vertical rectangular water multi-jet system. The tubes were heated to a temperature between 380-800°C in a controlled temperature furnace then cooled to the jet temperature. The temperature variation with time in the circumferential and the axial directions of the tubes has been measured. The two-phase flow behavior and the propagation of the re-wetting front around and along the tubes were simultaneously observed by using a high-speed camera. The effects of several parameters on the cooling process have been investigated. These parameters include: initial surface temperature, water subcooling (in the range 15- 80°C), jet velocity (in the range 0.15-1.60 m/s), tube solid material (brass, steel and Alumina), surface curvature, tube wall thickness, jet orientation and number of jets. The variables studied include the re-wetting delay time (time to quench after initiating the cooling process), there-wetting front propagation velocity, the quench and re-wetting temperatures, the quench cooling rates and the boiling region size. The quench and the re-wetting temperatures as well as the re-wetting delay time were found to be a strong function of water subcooling. The quench and re-wetting temperatures increase with increasing water subcooling. The rewetting delay time decreases with increasing the water subcooling, decreasing initial surface temperature, increasing liquid velocity and decreasing the surface curvature. There-wetting front velocity is mainly dependent on the initial surface temperature and water subcooling. The re-wetting velocity increases by decreasing the initial surface temperature and by increasing the water subcooling. Decreasing the surface curvature was found to also increase the re-wetting front velocity. Correlations of the phenomena studied have been developed and provided good prediction of the experimental data collected in this study and data available from literature. The. results of this study provide novel knowledge and an experimental database for mechanistic modeling of quench heat transfer on calandria tube surfaces that experience dryout and film boiling. (author)
[en] Several instrumentation equipment (pressure transmitters, flow and level) are used in a nuclear power plant. To make sure of the correctness of their reading, it is essential to calibrate periodically these equipment. These systematic interviews represent a very high workload, generate exposures, radiological information for the staff and represent a risk of error in the execution. Based on the experience gained over the years by several nuclear power plants, it has been shown that the majority of calibrations performed on these equipment s are not required since they usually keep their calibration. However, any transmitter can possibly develop errors and it is preferable to to quickly detect this failure. Given these facts, the nuclear industry has developed new technologies that enable online monitoring of instrumentation equipment. This surveillance is based on non-intrusive techniques that evaluate the performance of the equipment, allowing among others to detect the drift of a transmitter. By using these technologies, it is thus possible to focus maintenance efforts on equipment for which observations are deemed necessary. Despite a number of online monitoring systems (SSL) are already in use at some plants (Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) ), this technology is still not widely used in the nuclear industry. It is also noted that such systems are not widely used in CANDU power plants. The objective of this project is to validate the implementation of an SSL in a nuclear power plant (CANDU 6) in order to increase the interval for calibration of instrumentation equipment. An application project has first been done to evaluate the performance of an SSL. In the frame of this project, the mathematic algorithm Instrumentation and Calibration Monitoring Program (ICMP) developed by Electric Power Research Institute (EPRI) has been adapted and tested with data from a power station (CANDU 6). The results showed that when a transmitter presented a drift, SSL was able to detect in the vast majority of cases studied.
[en] The Economic Simplified Boiling Water Reactor (ESBWR) is GEH’s next evolution of advanced BWR technology. There are 1132 fuel bundles in the core and the thermal power is 4500 MWt. As part of design simplification it uses natural circulation flow with no recirculation pumps or their associated piping. The control blades are the primary control mechanism to address the need for performing reactivity adjustments (using fine-motion drives) at or near rated steady state power. This introduces the potential for duty-related fuel failure, which has to be rigorously addressed as part of reliable design and operation. As means to mitigate this potential for duty-related fuel failure and also to support a simplified ESBWR operation, this study investigates the feasibility of a fuel cycle core design strategy. The objective is to design fuel bundles, and to use them for developing a core design, that minimizes (but does not eliminate) the use of control blades during operation. The reduction in use is envisioned in their number as well as movement in the core. In such a strategy, the effect of the burnable poison in the fuel (that largely drives the core reactivity) is enhanced, and operationally the control blades react modestly to maintain the core critical. While the logic is simple, challenges exist in developing such a design because it needs to balance the requirement for having enough blade inventory in the core to address design/operational constraints and uncertainties. The strategy is conceptualized as “minimum hot excess (reactivity)” design. It reduces the number of blades in the core during normal operation by 50% in comparison to a similar fuel cycle core design with regular inventory of control blades. Because of the increased burnable poison, the minimum hot excess core design strategy comes at a cost of fuel cycle efficiency. This cost is determined in terms of an increased enrichment for the fresh fuel batch fraction.
[en] Since the thermophysical properties of water change dramatically at near-critical and pseudocritical point, the instability of natural circulation at supercritical pressure may occur in a loop. To predict the region of instability of natural circulation at supercritical pressure, a test loop was built at Tsinghua University. The paper presents the information of the test loop and a numerical analysis model for the loop. The paper verified the numerical analysis code by experiment results and using the code to analyze the instability of the loop. The paper concludes conclusion that there will be no Ledinegg instability occurring at supercritical pressure in the loop.
[en] The Rod Ejection Accident (REA) belongs to the Reactivity-Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA at Hot Zero Power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS v2.7 a REA in Almaraz NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions.
[en] This paper will present a high-fidelity system analysis of advanced nuclear energy systems approaching a zero-nuclear-waste limit. The focus will be on achievable performance characteristics. The applied high-fidelity system modeling approach will be discussed with illustrations using obtained physics simulation results. The near-term deployable dedicated small 14 MeV-neutron waste incineration back-end clean-up facilities are envisioned as advanced incinerators that are used in combination with thermal reactors (Westinghouse AP1000 and Generation IV VHTR are considered in the present study) to minimize the remaining long-lived radioactive species. These 14 MeV transmuters are viewed as small facilities designed to reduce waste inventories prior to reprocessing stages minimizing spent fuel activity levels and repository requirements. It is envisioned that under sustainable nuclear energy scenario small clean-up facilities will eventually evolve into self-consistent complexes producing electricity and serving as energy sources for industrial applications. The paper discusses physics features of these advanced systems and challenges of their conclusive evaluations.
[en] Operability of Very High Temperature Reactor (VHTR) hydrogen cogeneration systems in response to abnormal transients initiated by the hydrogen production plant is one of the important concerns from economical and safety points of views. The abnormal events in the hydrogen production plant could initiate load changes and induce temperature variations in a primary cooling system. Excessive temperature increase in the primary cooling system would cause reactor scrams since the temperature increase in the primary cooling system is restricted in order to prevent undue thermal stresses from reactor structures. Also, temperature decrease has a potential propagation path for reactor scrams by reactivity insertions as a consequence of the reactivity feedbacks. Since suspensions of reactor operation and electricity generation should be avoided even in case of abnormal events in the hydrogen production plant from an economical point of view, an establishment of a control scheme against abnormal transients of hydrogen production plant is required for plant system design. In the present study, basic controls and their integration for the GTHTR300C, a VHTR cogeneration system designed by JAEA with a direct Brayton cycle power conversion unit and thermochemical Iodine–Sulfur process hydrogen production plant (IS hydrogen production plant), against abnormal transients of IS hydrogen production plant are presented. Transient simulations for selected load change events in the IS hydrogen production plants are performed by an original system analysis code which enables to evaluate major phenomena assumed in process heat exchangers of the IS hydrogen production plant. It is shown that abnormal load change events are successfully simulated by the system analysis code developed. The results demonstrated the technical feasibility of proposed controls for continuous operation of the reactor and power conversion unit against load change events in the IS hydrogen production plant.