Nuclear Fuel Research Articles

Corrosion products deposition and its effects on conjugate heat transfer for helical cruciform fuel assembly

The corrosion products in the primary loop of the nuclear reactor can be deposited on the surface of the fuel rod, forming a deposition layer called CRUD layer, introducing the heat transfer resistance. The CRUD layer will result in an increase in the temperature of the cladding and accelerating the corrosion of the cladding. In this work, the corrosion product deposition and thermal resistance models of the deposition layer were established. The deposition behavior of corrosion products on the surface of a novel Helical-Cruciform Fuel (HCF) assembly and its influence on fluid-solid conjugate heat transfer were studied by coupling the deposition models with the flow and heat transfer models for fuel rods and coolant. The results show that the deposition rate of particles is affected by the wall shear force; the deposition rate of impurity ions is affected by the wall shear force and the temperature of the coolant in the viscous layer. The erosion rate of the deposited layer is related to the wall shear force and the sum of the particle and ion deposition rates. The thickness of CRUD layer on the surface of HCF assembly increases with the decrease of coolant flow rate and the increase of fuel power. Furthermore, the CRUD layer thickness varies in a period of 90° along the circumference of the fuel and reaches a peak at the root of the fuel ribs.

Issues Associated with Yellow Phase Formation and Other Disruptions in Japanese High-Level Waste Vitrification and Their Improvement

The Rokkasho Reprocessing Plant, located in northern Japan, includes facilities for reprocessing spent nuclear fuels and immobilizing high-level waste (HLW) into a stable-glass waste form based on borosilicate. The vitrification process consists of two large liquid-fed, joule-heated, ceramic-lined melters (LFCMs). During the active test campaigns at the Rokkasho Vitrification Facility, unstable melting performance and difficulty in glass operation were observed in the LFCM operation. The operating protocols that had been developed in the earlier full-scale mock-up testing with inactive waste simulants were found to be inadequate to manage a stable LFCM operation. We needed to understand correctly the fundamental causes for the operational difficulties experienced in the active test. First, we studied and summarized the troublesome element behaviors and disturbances in the HLW vitrification process, primarily on the LFCM. Second, we proposed an empirical model, including the transition of the phenomena that appeared in the active LFCM operation. Third, we analyzed the operating data and the trend of operation indexes to find out the fundamental causes for the unstable LFCM operation based on the results obtained in the described model. It was revealed that the presence of a significant amount of molybdenum and sulfur, together with the noble metals (NMs) Ru, Rh, and Pd in the Rokkasho HLW stream could lead to the formation of molybdate and sulfate salt phases [yellow phases (YPs)], and subsequently, the sedimentation of NMs at the bottom of the melter. The YPs may concentrate at the melter bottom and clog the pouring nozzle prior to the bottom accumulation of NM sludge, which can disrupt power deposition and cause difficulties during glass pouring operation. Based on these investigated results, we provide insights into short-term countermeasures, including numerical control of the heat balance in the melter, dilution of HLW feed with inactive waste simulants, and periodic rinsing of the bottom glass. We also provide an evaluation of their validities on the active operation through full-scale mock-up testing. However, the periodic rinsing and dilution operation produce additional glass-filled canisters and increase the total backend cost. Finally, we see a couple of important directions for future research and development. An advanced LFCM system has been developed for the long-term countermeasures to introduce recent advances in the LFCM design with YP and NM compatibilities. Furthermore, the modified or new glass formulations, which can incorporate more waste oxides into the glass matrix, known as high-waste loading glasses, also have been developed in parallel.

A science‐based mixed oxide property model for developing advanced oxide nuclear fuels

AbstractHerein, a science‐based uranium and plutonium mixed oxide (MOX) property model is derived for determining the properties of MOX fuel and analyzing their performance as functions of Pu content, minor‐actinide content, oxygen‐to‐metal ratio, and temperature. The property model is constructed by evaluating the effect of phonons and electronic defects on the heat capacity and thermal conductivity of MOX fuels. The effect of phonons was evaluated based on experimental datasets related to lattice parameters, thermal expansion, and sound speeds. Moreover, the effect of electronic defects was determined by analyzing oxygen‐potential data based on defect chemistry. Furthermore, the model evaluated the effect of the Bredig transition on the thermal properties of MOX fuel by analyzing the irradiation test results. The derived property model is applied to the performance code to analyze fast reactor fuel pins.

An integrated statistical-thermodynamic model for fission gas release and swelling in nuclear fuels

We propose a new model for burst fission gas release induced by microcracking in ceramic nuclear fuels such as uranium dioxide. The model stipulates that the densities of defects in the fuel material, such as microcracks and fission gas bubbles on grain boundaries, evolve in accordance with the second law of thermodynamics. Central to the model is the notion of an effective temperature, conjugate to the configurational entropy of the fuel material, and directly linked to the burnup. The model predicts that microcracking, driven by the internal stress state of the fuel material, reduces the bubble storage capacity of grain boundaries, and accounts for burst fission gas release during rapid temperature transients that simulate power transients, reactor startup, and loss-of-coolant accident conditions.

Materials qualification through the Nuclear Science User Facilities (NSUF): a case study on irradiated PM-HIP structural alloys

This article presents neutron irradiation and post-irradiation examination (PIE) capabilities available to the nuclear materials research community through the US Department of Energy’s Nuclear Science User Facilities (NSUF). The pressing need to deploy advanced nuclear reactors to combat climate change requires qualification of new fuels and materials. Among advanced manufacturing processes, powder metallurgy with hot isostatic pressing (PM-HIP) is nearest to becoming qualified for nuclear applications. This article provides examples from a recent irradiation and PIE program on a series of structural alloys fabricated by PM-HIP to illustrate how NSUF capabilities can be used to generate qualification data. The neutron irradiation experiments are described, and a sampling of results from tensile testing, nanoindentation, transmission electron microscopy, and atom probe tomography are presented, showing the favorable performance of PM-HIP alloys compared to their cast or forged counterparts under irradiation. This article provides a perspective on leveraging NSUF for future nuclear fuels and materials testing and qualification.

Improved correlations of the fuel/cladding liquid penetration rate with the out-of-pile transient database

Utilizing data from the Out-of-Pile Transient Database (OPTD), we developed advanced correlations for the cladding liquid penetration rate in metallic fuels used in sodium-cooled fast reactors (SFRs). Based on the available OPTD data, our research highlights the key role of Pu concentration in influencing the liquid penetration behavior, showing different threshold fuel-cladding interaction temperatures, with burnup playing a substantial role, particularly in fuels with a 19% Pu concentration. The penetration rate exhibits a parabolic, diffusion-limited behavior, thereby facilitating predictions of maximum liquid penetration depth over time. These correlations are significantly more accurate, under a variety of conditions, compared to existing single-parameter models, and enhance our understanding of SFR metallic fuel behavior, promoting safer and more efficient reactor operations.

Accelerated Formation of Oxide Layers on Zircaloy-4 Utilizing Air Oxidation and Comparison with Water-Corroded Oxide Layers.

For the dry storage of Canada Deuterium Uranium (CANDU) spent nuclear fuels, the integrity of Zircaloy-4 fuel cladding has to be verified. However, the formation of ~10 µm-thick oxide layers in typical CANDU reactor operating conditions takes several years, which makes sample preparation a slow process. To overcome such limitations, in this study, an accelerated formation of an oxide layer on Zircaloy-4 cladding tube was developed with a combination of high-temperature water corrosion (HT-WC) and air oxidation (AO). First, Zircaloy-4 tubes were corroded in oxygenated (2 ppm dissolved oxygen) high-temperature water (360 °C/19.5 MPa) for 500 h. Then, the tubes were air-oxidized at 500 °C for 30 h. Finally, the tubes were corroded again in HT-WC for 500 h to produce ~10 µm-thick oxide layers. The morphology and characteristics of the oxide layer in each step were analyzed using X-ray diffraction, scanning and transmission electron microscopy. The results showed that the oxide layer formed in the accelerated method was comparable to that formed in HT-WC in terms of morphology and oxide phases. Thus, the accelerated oxide formation method can be used to prepare an oxidized Zircaloy-4 cladding tube for CANDU fuel integrity analysis.

Image reconstruction algorithm for momentum dependent muon scattering tomography

Nondestructive radiography using cosmic ray muons has been used for decades to monitor nuclear reactor and spent nuclear fuel storage. Because nuclear fuel assemblies are highly dense and large, typical radiation probes such as x-rays cannot penetrate these target imaging objects. Although cosmic ray muons are highly penetrative for nuclear fuels as a result of their relatively high energy, the wide application of muon tomography is limited because of naturally low cosmic ray muon flux. This work presents a new image reconstruction algorithm to maximize the utility of cosmic ray muon in tomography applications. Muon momentum information is used to improve imaging resolution, as well as muon scattering angle. In this work, a new convolution was introduced known as M-value, which is a mathematical integration of two measured quantities: scattering angle and momentum. It captures the objects’ quantity and density in a way that is easy to use with image reconstruction algorithms. The results demonstrate how to reconstruct images when muon momentum measurements are included in a typical muon scattering tomography algorithm. Using M-value improves muon tomography image resolution by replacing the scattering angle value without increasing computation costs. This new algorithm is projected to be a standard nondestructive radiography technique for spent nuclear fuel and nuclear material management.

Numerical simulation of flow and heat transfer in fuel assembly of floating nuclear power platform

ABSTRACT Floating nuclear power platform (FNPP), unlike onshore nuclear power plants, is subjected to periodic rolling loads at sea. These loads will have a more obvious impact on the flow characteristics and temperature distribution inside the fuel assembly (FA). This paper first establishes a refined model of the 7 × 7 core fuel assembly. Then, the flow and heat transfer characteristics of assembly with pure fuel bundles and assembly with guide tubes are analyzed using a segmented grid partitioning method, taking into consideration both accuracy and efficiency in calculation. Finally, the flow and heat transfer characteristics inside the assembly with guide tubes under rolling conditions are simulated and calculated, and compared with the previous simulation results. The calculation results show that the flow and heat transfer characteristics of the assembly with guide tubes are quite different from those of the pure fuel rod bundle assembly, and the rolling environment has obvious adverse effects on the flow field characteristics and temperature distribution in the fuel assembly of the offshore reactor, which must be considered in the design and construction of the floating nuclear power platform. The research results have significant reference value for the design and optimization of floating nuclear power platform assemblies under rolling conditions.

Novel hydrophobic Ti-PVDF energetic thin films with excellent mechanical and ignition properties and their reaction mechanisms

With the development of traditional aluminum(Al)-based nanothermites, some problems on account of the metal Al powder gradually emerged. Therefore, extensive research is devoted to exploring novel metal fuels to achieve better energy release effects. Titanium (Ti) powder is a potential metal fuel theoretically considering of reducibility and calorific value, but it has rarely been reported in the field of nanothermites. This paper put forward the design strategy by selecting as the metal fuel and polyvinylidene fluoride (PVDF) as oxidant. According to the hydrophilicity of Ti powders, the PVDF film encapsulation can effectively improve hydrophobicity and storage stability. Another advantage of PVDF film state is that it imparts excellent mechanical properties (yield strength of 30.5 MPa and elongation at break of 202.8 %). According to the results of thermal analysis, Ti increased the thermal stability of PVDF while accelerated its thermal decomposition rate. In ignition properties, Ti-PVDF films have the shortest ignition delay time (217 ms) lower than results reported in the literature of other nanothermites under the same conditions. In addition, the energy release process of ignition combustion is analyzed and verified in this paper. Therefore, Ti-PVDF energetic ignition material with diverse properties is expected to be applied in more fields not limited to nanothermites.

A model of fuel pellet swelling and its relation with grain growth

In this paper, an efficient model (from a computational point of view) is implemented to evaluate the gaseous and solid swelling in nuclear fuels under irradiation conditions. The model contemplates the precipitation of fission gases as intragranular and intergranular bubbles in the form of two coupled systems. For the intragranular gas, the diffusion equation is solved taking into account the processes of gas production, as well as the nucleation, growth and destruction of bubbles. In the case of the intergranular gas, bubbles are considered to have a lenticular shape and grow due to diffusion-controlled vacancy absorption. The results obtained and their comparison with experimental data indicate that the model is capable of performing realistic simulations of the pellet swelling processes, both under constant power conditions and fast power transient conditions. After that, the model was coupled with a solid swelling model and a grain growth model in order to analyze its performance as a whole.

First-Principle Studies on Local Lattice Distortions and Thermodynamic Properties in Non-Stoichiometric Thorium Monocarbide

Thorium monocarbide (ThC) is interesting as an alternative fertile material to be used in nuclear breeder systems and thorium molten salt reactors because of its high thermal conductivity, good irradiation performance, and wide homogeneous composition range. Here, the influence of carbon vacancy site and concentration on lattice distortions in non-stoichiometric ThC1−x (x = 0, 0.03125, 0.0625, 0.125, 0.1875, 0.25, or 0.3125) is systematically investigated using first-principle calculations by the projector augmented wave (PAW) method. The energy, mechanical parameters, and thermodynamic properties of the ThC1-x system are calculated. The results show that vacancy disordering has little influence on the total energy of the system at a constant carbon vacancy concentration using the random substitution method. As the concentration of carbon vacancies increases, significant lattice distortion occurs, leading to poor structural stability in ThC1−x systems. The changes in lattice constant and volume indicate that ThC0.75 and ThC0.96875 represent the boundaries between two-phase and single-phase regions, which is consistent with our experiments. Furthermore, the structural phase of ThC1−x (x = 0.25–0.3125) transforms from a cubic to a tetragonal structure due to its ‘over-deficient’ composition. In addition, the elastic moduli, Poisson’s ratio, Zener anisotropic factor, and Debye temperature of ThC1-x approximately exhibit a linear downward trend as x increases. The thermal expansion coefficient of ThC1−x (x = 0–0.3125) exhibits an obvious ‘size effect’ and follows the same trend at high temperatures, except for x = 0.03125. Heat capacity and Helmholtz free energy were also calculated using the Debye model; the results showed the C vacancy defect has the greatest influence on non-stoichiometric ThC1−x. Our results can serve as a theoretical basis for studying the radiation damage behavior of ThC and other thorium-based nuclear fuels in reactors.

Sulphate-reducing bacteria (SRB) in interim storage of spent nuclear fuel

Interim Storage of Spent Nuclear Fuel (ISSF) is an installation for temporarily storing spent nuclear fuels. Nuclear fuel storage pools must be free from contamination by corrosion-causing microorganisms such as Sulphate Reduction Bacteria (SRB). This research aims to detect SRB on the pool’s wall, floor, and spent nuclear fuel racks. The measured parameters consisted of physicochemical, total bacteria, total SRB, and detection of SRB on the wall, floor, and spent fuel rack in the pool using the SRB kit. The results showed that the quality of the water chemistry in the pool was within limits according to the provisions of the International Atomic Energy Agency, except for the total bacteria (> 1000 CFU/ml). In the water pools around the racks of spent fuels, SRB was also detected at 30 CFU/ml, while the level of SRB on the walls, floors, and spent nuclear fuel racks showed varied values from moderate to severe. This research is important especially to identify corrosion-causing microbes for water and biofilm samples. This is the first step in understanding the problem of biofilm formation and contamination of microorganisms in spent fuel storage pools.

Computational and Experimental Method for Determination of the Inertia of Thermocouples within the Framework of Pre-Reactor Tests of Experimental Fuel Rods

Computational and Experimental Method for Determination of the Inertia of Thermocouples within the Framework of Pre-Reactor Tests of Experimental Fuel Rods

Assessment of Applicability of Bentonite Clay-Based Ceramics for Immobilization of Radioactive Waste from Pyrochemical Processing of Spent Nuclear Fuel

Assessment of Applicability of Bentonite Clay-Based Ceramics for Immobilization of Radioactive Waste from Pyrochemical Processing of Spent Nuclear Fuel

Comparative Evaluations of Nuclear Fuel Breeding Rate in Hybrid Fission-Fusion Reactor with U and Th

Comparative Evaluations of Nuclear Fuel Breeding Rate in Hybrid Fission-Fusion Reactor with U and Th

Secondary Flows and Transverse Exchange in Cylindrical Fuel Element Bundles

Secondary Flows and Transverse Exchange in Cylindrical Fuel Element Bundles

Peculiarities of Changes in the Isotopic Composition of Pilot Fuel Elements of a VVER-SKD Reactor under Successive Irradiation in a Fast and Thermal Neutron Spectrum

Peculiarities of Changes in the Isotopic Composition of Pilot Fuel Elements of a VVER-SKD Reactor under Successive Irradiation in a Fast and Thermal Neutron Spectrum

Experience in Gas Dynamic Tests of Model Bundles of Self-Separating Helical Fuel Elements Made Using Additive Technologies

Experience in Gas Dynamic Tests of Model Bundles of Self-Separating Helical Fuel Elements Made Using Additive Technologies

Equibiaxial flexural strength determination of UO2 using a ball-on-ring test

To increase nuclear fuel performance and reliability, their mechanical properties require an accurate and statistically relevant assessment. Biaxial flexural strength tests provide an alternative to bend bar techniques for assessing mechanical behavior; namely, the transverse rupture strength (TRS) of ceramic samples. Biaxial test samples require simple geometries and minimal surface preparation, reducing fabrication costs and handling hazards. This study investigated the TRS of polycrystalline UO2 fuel forms at room temperature using a ball-on-ring test fixture. Pellets were fabricated from UO2 powder using conventional powder processing and sintering techniques. The TRS and Weibull parameters were obtained through Weibull statistics on over 60 UO2 samples tested under equibiaxial flexure. The larger sample size in this study enabled a more robust Weibull statistical analysis than alternative test methods, which may not capture the stochastic failure of typical ceramics. Furthermore, two different loading ball diameters were employed to assess the impact of contact damage on fracture strength. While Hertzian contact damage was observed with the smaller loading ball, the fracture strength remained unaffected. A fracture analysis of the tested UO2 samples indicated a mixture of intergranular and transgranular fracture that transitions to transgranular fracture with increasing distance from the fracture origin. The characteristic strength of the combined data sets was determined to be 148 MPa, and the Weibull modulus was determined to be 9.1. The TRS values and Weibull parameters were close to values found in the literature for alternative testing techniques using samples with similar microstructure and density. The findings in this study validate the ball-on-ring method used to obtain the TRS of UO2 with a sample geometry more representative of nuclear fuels. Additionally, experimental TRS results from this study can be implemented in modeling codes to predict fuel performance, which is critical to fuel burnup extension and advanced nuclear fuel technologies.