Reactor Fuel Research Articles

Coupled Analysis of Thorium-based Fuels in the High-Performance Light Water Reactor Fuel Assembly

One of the six selected concepts to be part of Generation IV nuclear reactors is the Supercritical Light Water Cooled Reactor. The High-Performance Light Water Reactor (HPLWR) is the European version and it is a very promising design. In recent years, interest in the study of thorium-based fuel cycles has been renewed and its possibilities for current LWRs have been evaluated. The use of thorium-based fuels will be fundamental in the future sustainability of nuclear energy, since in addition to its abundance in nature, thorium has an important group of advantages. In this paper, performance of thorium-based fuels in the typical fuel assembly of the HPLWR reactor is evaluated, using a computational model based on CFD and Monte Carlo codes for the neutronic/thermal-hydraulic coupled analysis. The volumetric power density profiles, coolant temperature profiles, fuel temperature profiles and others are compared with those obtained for standard UO 2 fuel. When the thorium-based fuels are used, the obtained infinite multiplication coefficients are smaller than the value obtained when UO 2 is used, since the 232 Th isotope has a lower contribution to the multiplicative properties of the medium than 238 U. As a result, a difference of approximately 12 000 pcm was observed. The results verified that the HPLWR is a thermal reactor with a hard spectrum. There are no notable changes in the neutron spectrum if the mass fraction of thorium is slightly varied. With coupled analysis, the potential benefits of the utilization of thorium-based fuels were verified. Moreover, a significant temperature decrease by 136 K on the center line of the fuel elements was observed. When the mass fraction of thorium increases in the oxides mixture, the weighted average temperature on the fuel elements decreases .

Investigation of safety aspects during steady state operation of Tehran research reactor fuel test loop

In order to enhance making use of potential capabilities of Tehran research reactor (TRR) as a material test reactor (MTR), a 10-bar fuel test loop is designed to provide required infrastructure to perform irradiation tests on domestic fuels in Tehran research reactor. Tehran research reactor with appropriate neutron flux in irradiation positions in the core can provide required neutron fluence for irradiation tests on nuclear fuels. TRR fuel test loop with a wide range of adjustable coolant flow rate, pressure and temperature, within the range applicable in research reactors, has capability to simulate thermal-hydraulic conditions of various research reactors to perform irradiation tests on research reactors’ fuels under the same condition as the reactor core of concern. This paper presents an introductory to the systems and safety features of the TRR fuel test loop as well as the results of analyses performed to investigate neutronic and thermal-hydraulic safety parameters of TRR core during fuel test loop operation and the results of steady state analysis of the loop.

Quantification of effluents in the production of nuclear fuel

The Brazil with the purpose of becoming self-sufficient in the production of radioisotopes and radioactive sources used in nuclear medicine, agriculture and the environment developed the project of a multipurpose reactor of 30 megawatts of power to meet the national demand. In the Energy and nuclear research Institute (IPEN), Nuclear fuel Center (CCN) is responsible for the manufacture of fuel for the reactor IEA-R1 and, possibly, by the multipurpose reactor fuels. In order to meet the demand for the reactors was designed a new manufacturing plant with a maximum capacity of 60 fuel per year, which currently is 10. The increase in production as a consequence will increase the volume of effluents generated. The current concern with the environment it is necessary to draw up a management plan to make the process sustainable, which will result in environmental, economic and social benefits. The fuel production process generates various types of effluent containing uranium or not – being solid, gaseous and liquid with different physical and chemical characteristics. The objective of this work is to follow the process of nuclear fuel production and to identify, quantify and characterize the effluents, especially liquids, to subsequently draw up a management plan and eventually dispose of responsibly in the environment.

Performance Indicators for Benchmarking Solar Thermochemical Fuel Processes and Reactors

Concentrated solar energy offers a source for renewable high-temperature process heat that can be used to efficiently drive endothermic chemical processes, converting the entire spectrum of solar radiation into chemical energy. In particular, solar-driven thermochemical processes for the production of fuels include reforming of methane and other hydrocarbons, gasification of biomass, coal, and other carbonaceous feedstock, and metal oxide redox cycles for splitting H2O and CO2. A notable issue in the development of these processes and their associated solar reactors is the lack of consistent reporting methods for experimental demonstrations and modelling studies, which complicates the benchmarking of the corresponding technologies. In this work we formulate dimensionless performance indicators based on mass and energy balances of such reacting systems, namely: energy efficiency, conversion extent, selectivity, and yield. Examples are outlined for the generic processes mention above. We then provide guidelines for reporting on such processes and reactors and suggest performance benchmarking on four key criteria: energy efficiency, conversion extent, product selectivity, and performance stability.

Performance of solid oxide fuel cell with chemical looping gasification products as fuel

Chemical looping gasification (CLG) can achieve the utilization of solid fuels for syngas production. The CLG system integrated with solid oxide fuel cell (SOFC) is a promising energy conversion way. In this work, an integration system of CLG and SOFC is evaluated via the implementation of a multi-field coupling modelling, where the products from the CLG are directly transported into the SOFC as the fuel and the coke deposition effect on the cell performance is evaluated. The results reveal that SOFC temperature using pure hydrogen as fuel has an increase of around 4 K compared to that with gas mixture as fuel owing to the inhibition of carbon deposition. It is found that the arrangement of anode and cathode in the countercurrent mode can promote the overall uniformity of current density compared to that in the cocurrent flow. Moreover, the impact of operating parameter of the CLG system on the SOFC performance is also examined. The results demonstrate that the increase of fuel reactor (FR) temperature and H2O/C molar ratio in the CLG system is beneficial to the inhibition of carbon deposition and the enhancement of the SOFC performance.

Radiation Tolerant 3D Laser Scanner for Structural Inspections in Nuclear Reactor Vessels and Fuel Storage Pools

Accurate and timely assessment of displacements and/or structural damages in nuclear reactor vessels’ components is a key action in routine inspections for planning maintenance and repairs but also in emergency situations for mitigating consequences of nuclear incidents. Nevertheless, all these components are maintained underwater and reside in high-radiation fields thus imposing harsh operative conditions to inspection devices which must cope with effects such as Cerenkov radiation background, Total Ionizing Radiation (TID), and occlusions in the detectors’ field of view. To date, ultrasonic techniques and video cameras are in use for inspection of components’ integrity and with measurements of volumetric and surface crack opening displacements, respectively. The present work reports the realization of a radiation tolerant laser scanner and the results of tests in a nuclear research reactor vessel for acquisition of 3D models of critical components. The device, qualified for underwater operation and for withstanding up to 1 MGy of TID, is based on a 515 nm laser diode and a fast-scanning electro-optic unit. To evaluate performances in a significant but controlled environment, the device has been deployed in the vessel of a research reactor operated by ENEA in the Casaccia Research Centre in Rome (Italy). A 3D model of the fuel rods assembly through a cooling water column of 7 m has been acquired. The system includes proprietary postprocessing software that automatically recognizes components of interest and provides dimensional analysis. Possible application fields of the system stretch to dimensional analysis also in spent nuclear fuel storage pools.

Influence of Stove, Fuel, and Oxidation Flow Reactor Conditions on Aging of Laboratory-Generated Cookstove Emissions

Influence of Stove, Fuel, and Oxidation Flow Reactor Conditions on Aging of Laboratory-Generated Cookstove Emissions

Methodological Issues and Results of Experimental and Computational Studies of Critical Heat Flows in RITM-200 Reactor Fuel Assemblies for Small NPP

Methodological Issues and Results of Experimental and Computational Studies of Critical Heat Flows in RITM-200 Reactor Fuel Assemblies for Small NPP

Uniform corrosion of FeCrAl cladding tubing for accident tolerant fuels in light water reactors

Iron-Chromium-Aluminum (FeCrAl) alloys are candidate materials for light water reactor accident tolerant fuel (ATF) rod cladding because of the unparalleled resistance of these alloys to attack by superheated steam in the case of a loss of coolant accident. Since FeCrAl alloys were never used before in a reactor environment, it was important to characterize their corrosion resistance under simulated reactor normal operation conditions. Tests were conducted for APMT and C26M alloys in three ~300°C water systems (two containing hydrogen and one containing oxygen). Results show that after 12 months of immersion the FeCrAl alloys lose a small amount of mass. The resistance to corrosion is mostly given by the development on the surface of a protective oxide mostly rich in Cr and Al. The mass loss was lower in APMT than in C26M because of the higher Cr content in the former. After 12 months residence in the ~300°C water, the FeCrAl specimens maintained their unequaled resistance to attack by 1200°C steam.

A new generation of uranium coincidence fast neutron collars for assay of LWR fresh fuel assemblies

The active uranium neutron coincidence collar provides a means of non-destructively assaying the fissile linear density of Light Water Reactor fresh fuel assemblies containing low enriched uranium. These neutron collars can operate in two modes: a thermal and a fast mode. In fast mode, a neutron collar has an added cadmium (Cd) liner in the sample cavity of the detector to reduce the impact of the burnable poison (thermal neutron absorber) on the detector signal (doubles). The main advantage for operating in fast mode is a detected signal that is less dependent of the burnable neutron poison content and thus less dependent on facility operator declarations. The drawback is that operating in fast mode requires a longer measurement time (∼hour vs tens of minutes for thermal mode) to achieve the statistically needed precision in the measurements. The trend in the modern reactor fuel assemblies is moving to higher burnup by using higher initial enrichment and, consequently, a higher number of burnable poison rods to compensate the initial neutron reactivity. The increase of the burnable poison loading has motivated the development of a new generation of high efficiency fast neutron collars to allow practical measurements in-field by nuclear inspectors. This paper describes the development and performance evaluation of a new generation of neutron collars, for both boiling water reactor (BWR) and pressurized water reactor (PWR) fuels, jointly developed between the US Department of Energy, through Los Alamos National Laboratory, and the Euratom Safeguards Directorate of the European Commission. We present here calibrations with reference fuel assemblies at Los Alamos National Laboratory as well as the results of in-field measurement campaigns in fuel fabrication plants with modern commercial fuel assemblies. The experimental results show that a typical PWR verification can be made in a total time of 30 min with an uncertainty in the measured mass of 2% at one standard deviation (1σ). A BWR verification can be made in 47 min with an uncertainty in the measured mass of 1.9% at 1σ, or a total time of 20 min with 1σ uncertainty in the measured mass of 2.5%.

Instability of U3Si2 in pressurized water media at elevated temperatures

Following the Fukushima Daiichi accident, significant efforts from industry and the scientific community have been directed towards the development of alternative nuclear reactor fuels with enhanced accident tolerance. Among the proposed materials for such fuels is a uranium silicide compound (U3Si2), which has been selected for its enhanced thermal conductivity and high density of uranium compared to the reference commercial light water reactor (LWR) nuclear fuel, uranium oxide (UO2). To be a viable candidate LWR fuel, however, U3Si2 must also demonstrate that, in the event of this fuel coming in contact with aqueous media, it will not degrade rapidly. In this contribution, we report the results of experiments investigating the stability of U3Si2 in pressurized water at elevated temperatures and identify the mechanisms that control the interaction of U3Si2 under these conditions. Our data indicate that the stability of this material is primarily controlled by the formation of a layer of USiO4 (the mineral, coffinite) at the surface of U3Si2. The results also show that these layers are destabilized at T > 300 °C, leading to the complete decomposition of U3Si2 and its pulverization due to its full oxidation to UO2.

Mechanism analysis and simulation of methyl methacrylate production coupled chemical looping gasification system

Nowadays, the efficient and cleaner utilization of coal have attracted wide attention due to the rich coal and rare oil/gas resources structure in China. Coal chemical looping gasification (CCLG) is a promising coal utilization technology to achieve energy conservation and emission reduction targets for highly pure synthesis gas. As a downstream product of synthesis gas, methyl methacrylate (MMA), is widely used as raw material for synthesizing polymethyl methacrylate and resin products with excellent properties. So this paper proposes a novel system integrating MMA production and CCLG (CCLG-MMA) processes aiming at “energy saving and low emission”, in which the synthesis gas produced by CCLG and purified by dry methane reforming (DMR) reaction and Rectisol process reacts with ethylene for synthesizing MMA. Firstly, the reaction mechanism of CCLG is investigated by using Reactive force field (ReaxFF) MD simulation based on atomic models of char and oxygen carrier (Fe2O3) for obtaining optimum reaction temperature of fuel reactor (FR). Secondly, the steady-state simulation of CCLG-MMA system is carried out to verify the feasibility of MMA production. The amount of CO2 emitted by CCLG process and DMR reaction is 0.0028 (kg CO2)−1·(kg MMA)−1. The total energy consumption of the CCLG-MMA system is 45521 kJ·(kg MMA)−1, among which the consumption of MMA production part is 25293 kJ·(kg MMA)−1. The results show that the CCLG-MMA system meets CO2 emission standard and has lower energy consumption compared to conventional MMA production process. Finally, one control scheme is designed to verify the stability of CCLG-MMA system. The CCLG-MMA integration strategy aims to obtain highly pure MMA from multi-scale simulation perspectives, so this is an optimal design regarding all factors influencing cleaner MMA production.

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium silicide

The challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breech of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium silicide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UC and UB 2 . Presented here are the literature for the U 3 Si 2 degradation modes, thermodynamics, and oxidation performance of the pure compound and its reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation/corrosion behavior for high uranium density fuels for use in LWRs.

Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

Investigation of biomass alkali release in a dual circulating fluidized bed chemical looping combustion system

Chemical looping combustion (CLC) of biomass is a promising technology for power generation with integrated carbon capture. In CLC, alkali content of biomass poses potential issues of bed agglomeration, as well as heat exchanger fouling and corrosion. The fate of biomass alkalis was investigated in a dual-interconnected circulating fluidized bed CLC system. Experiments were conducted in oxygen carrier aided combustion (OCAC) and CLC modes. Ilmenite and braunite oxygen carriers and three biomass fuels (wood pellets, wood char, straw pellets) were tested. Flue gas alkali emissions in the air reactor (AR) and fuel reactor (FR) were measured with a surface ionization detector (SID). Results showed that CLC operation yields gas-phase alkali emissions that are up to 15 times higher than in comparable OCAC operation. Results analysis concluded that increased alkali emissions in CLC arise from the steam atmosphere in the FR, whereby steam accelerates the decomposition of alkali compounds in the biomass. Retention of alkalis in the condensed phase was found to be >97% for ilmenite and >92% for braunite CLC operation. Up to 60–80% of the retention was attributed to fuel ash formation. The residual retention was attributed to absorption of alkalis by the oxygen carriers. Absorption likely occurred mainly through formation of alkali manganates and silicates in braunite, and formation of alkali silicates, aluminosilicates, manganates, and titanates in ilmenite. Gas-phase alkali emissions in the AR, although less than in the FR, were found to occur due to combustion of unconverted fuel carried over from the FR to the AR.

Multi-scale modeling and control of chemical looping gasification coupled coal pyrolysis system for cleaner production of synthesis gas

The cleaner and efficient utilization of coal is of great significance to achieve energy conservation enhancement and CO2 emission reduction targets. Coal pyrolysis (CP) and chemical looping gasification (CLG) are both new coal utilization technologies. A cleaner system integrating CP and CLG processes is put forward in this paper to produce high purity synthesis gas with lower CO2 emission for the sustainable development of coal industry, with the aid of multi-scale modeling means including molecular dynamics (MD) simulation, computational fluid dynamics (CFD) simulation, and process simulation. First, MD simulation is executed for CP process with a coal model consisting of 3860 atoms to obtain the optimum operation parameters by mechanism analysis. The obtained coal density of 1.2 ± 0.11 g/cm3 is consistent with the experimental result of 1.28–1.33 g/cm3, ensuring the optimized coal geometry model below 1100 K. Second, the CLG process is simulated by CFD method. CuO is selected as oxygen carrier based on its better heat transfer performance before its residence time is determined based on the fuel reactor (FR) model. Then the CP-CLG system is simulated with Aspen Plus software on the basis of above MD and CFD results. The process simulation result indicates that the purity of synthesis gas produced by CP-CLG system is about 20% higher than single coal CLG system as the latter generates much more CO2. Finally, one control scheme is designed to keep the operational stability of CP-CLG based cleaner synthesis gas production system with the aid of Aspen Dynamics software.

Quenching Experiments with CrAl-coated Zircaloy Cladding in Reflooding Water Flows

A quenching experiment is performed to investigate the heat transfer characteristics and cooling performance of CrAl-coated Zircaloy (Zr) cladding in a water flow. The CrAl-coated Zr cladding is one of the accident tolerant fuels for light water reactors. The uncoated Zr cladding is also used in this quenching experiment for comparison. This experiment simulates reflood quenching of fuel rod during loss of coolant accident (LOCA) in nuclear power plant. The test conditions were determined to represent the peak cladding temperature, the coolant subcooling and the reflood velocity in the event of LOCA. The flow visualization showed the film boiling during early stage of reflood quenching and the transition to nucleate boiling. The film layer decreases as the coolant subcooling increases and becomes wavy as the reflood velocity increases. The CrAl-coated Zr cladding showed more wavy and thinner film than the uncoated Zr cladding. The rewetting temperature increases as the initial wall temperature and/or the coolant subcooling increases. The quench front velocity increases significantly as the coolant subcooling increases. The reflood velocity has a negligible effect on rewetting temperature and quench front velocity.

Solid-solution strengthening by Al and Cr in FeCrAl oxide-dispersion-strengthened alloys

FeCrAl oxide-dispersion-strengthened (ODS) alloys are promising materials for accident tolerant fuels for light water reactors (LWRs). In these alloys, Al and Cr are key elements with important synergistic effects: enhancement of the formation of oxidation-resistant Al 2 O 3 phase by Cr addition and suppression of the formation of the embrittling Cr-rich α′ phase by Al addition. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The Al and Cr contents were systematically varied from 9 to 16 at.% and 10–17 at.%, respectively, and tensile tests were conducted at 298 K, 573 K, and 973 K. The solid-solution strengthening increased linearly, 20 MPa per 1 at.% Al and 5 MPa per 1 at.% Cr, at the typical LWR operational temperature of 573 K. The conventional Fleischer–Friedel and Labusch theories cannot explain this level of solid-solution strengthening. It was originally clarified that the solid-solution strengthening by Al and Cr in the FeCrAl-ODS alloys can be reasonably predicted by modifying shear modulus difference for the Suzuki's double kink theory for screw dislocations. The inverse dependency on the absolute temperature and linear dependency on the Al and Cr content can be successfully realized by this approach.

BISON: A Flexible Code for Advanced Simulation of the Performance of Multiple Nuclear Fuel Forms

BISON is a nuclear fuel performance application built using the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element library. One of its major goals is to have a great amount of flexibility in how it is used, including in the types of fuel it can analyze, the geometry of the fuel being modeled, the modeling approach employed, and the dimensionality and size of the models. Fuel forms that can be modeled include standard light water reactor fuel, emerging light water reactor fuels, tri-structural isotropic fuel particles, and metallic fuels. BISON is a platform for research in nuclear fuel performance modeling while simultaneously serving as a tool for the analysis of nuclear fuel designs. Recent research in BISON includes techniques such as the extended finite element method for fuel cracking, exploration of high-burnup light water reactor fuel behavior, swelling behavior of metallic fuels, and central void formation in mixed-oxide fuel. BISON includes integrated documentation for each of its capabilities, follows rigorous software quality assurance procedures, and has a growing set of rigorous verification and validation tests.

Optimization of synthesis gas production in the biomass chemical looping gasification process operating under auto-thermal conditions

Biomass Chemical Looping Gasification (BCLG) is a promising route to obtain N2 free high purity syngas. In this work, heat and mass balances were solved to determine the auto-thermal operation conditions that maximize the syngas yield in a BCLG system. A Fe-based oxygen carrier and pine wood as fuel were considered for the simulation. Two methods to control the transference of oxygen between the air reactor (AR) and fuel reactor (FR) were analysed. Syngas yield was higher controlling the oxygen fed to the AR by the air flow (OCM-1) than controlling the oxygen supplied to the FR by the oxygen carrier circulation flow (OCM-2).The influence of different operating parameters, such as oxygen carrier transport capacity, preheating of gases fed to the system, steam/biomass ratio, fuel reactor temperature, was also analysed. It is noteworthy that working with OCM-2 it is necessary to optimize the amount of active phase in synthetic oxygen carriers or to dilute the natural oxygen carriers (ores, wastes) with an inert material to maintain realistic temperature difference values between reactors. Therefore, it is recommended to operate with the OCM-1 as it has the advantages of a more flexible operation and the possibility of obtaining pure N2.