Irradiated Fuels Research Articles

Manufacturing porous U10Zr fuels with controlled porosities by SPS and thermal properties

To accommodate the swelling of metallic fuels induced by the fission gas release for a sodium-free fuel option, advanced U-10Zr fuels with controlled porosities were designed and demonstrated by spark plasma sintering (SPS). U-10Zr fuel pellets with manufactured porosities varying from 35% up to fully dense fuel pellets have been fabricated by controlling mechanical attrition of the starting uranium powders, sintering temperature, pressure, and duration, and the correlation among the microstructure control – porosity – sintering conditions has been established. To further mimick the pore structure in irradiated fuels, different pore formers (NaCl and NH4HCO3) were used to control the pore size and distribution. Microstructure characterization indicates a lamellar reaction zone of enriched Zr and U composed of alpha Zr with 2 at% U and distorted alpha U with 13 at% of Zr, respectively, differing from the arc-melted U-10Zr as a result of rapid consolidation of SPS at lower temperatures and short durations. The thermal conductivity of U10Zr with different porosities were also measured. This work demonstrates the success in manufacturing new metallic fuel forms with controlled porosities and pore distribution, which can be used as model systems to investigate the thermal transfer behavior of metallic fuels in the reactor.

Simulating Araponga - The High-Resolution Diffractometer of Brazilian Multipurpose Reactor

The Brazilian Multipurpose Reactor (RMB) is a fundamental project that aims to turn Brazil into a self-sufficient country in the production of radioisotopes and radiopharmaceuticals to supply the Unified Health System (SUS) as much as the private institutions. In addition, the RMB project describes other applications as irradiation and testing of nuclear fuels and structural material analysis, for instance. There are many techniques in the project to study structural aspects of materials, where neutron diffraction represents one of the priorities for implementation. This technique will take place mainly on two diffractometers on Thermal Neutron Guide 1 (TG1), namely Araponga, a high-resolution diffractometer, and Flautim, a high-intensity diffractometer. In this work, we study the performance of the Araponga diffractometer through McStas simulations with input produced by the MCNP code of the RMB core. We investigate the neutron flux values considering a state-of-art high-resolution diffractometer, and the results are promising since some simulated scenarios present values compatible with high-intensity devices.

METHODOLOGICAL FEATURES TO JUSTIFY NUCLEAR SAFETY FOR IRRADIATED FUEL RODS TEMPORARY PLACEMENT IN HOT CELLS

METHODOLOGICAL FEATURES TO JUSTIFY NUCLEAR SAFETY FOR IRRADIATED FUEL RODS TEMPORARY PLACEMENT IN HOT CELLS

BISON fuel performance modeling optimization for experiment X447 and X447A using axial swelling and cladding strain measurements

With the recent need to qualify new reactor designs, such as the Versatile Test Reactor, researchers need fuel performance calculations to determine the safety criteria of proposed designs. In order to validate the fuel performance results obtained by the fuel performance code BISON for new reactor designs, legacy fuel from EBR-II and Fast Flux Test Facility MFF with post-irradiation examination (PIE) data need to be used as validation cases to benchmark models. In this work, BISON has been paired with the Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A. X447/X447A were assessed by implementing models for a fuel-cladding chemical interaction (FCCI) within BISON and optimizing the friction coefficient between the fuel surface and the cladding, the anisotropic swelling factor, and the HT9 first thermal creep scalar (which scales the first term in the HT9 creep equation) to best match the PIE axial fuel swelling height and cladding profilometry for all pins in X447/X447A. The optimal values were found using a generic algorithm developed to select different values for the three parameters until end criteria was met and error couldn’t be reduced further. The BISON-simulated cladding profilometry was evaluated using a standard error of the estimate to account for the profile shape of the cladding profilometry. Optimal values for the friction coefficient, anisotropic fuel swelling factor, and HT9 first thermal creep scalar were found to best fit the BISON simulation results to the PIE measurements found in IMIS and FIPD. Possible improvements to current models to account for the underprediction of fuel swelling at low burnups and the overprediction of fuel swelling at higher burnups that is observed for the axial fuel swelling height are discussed. Although two pins in EBR-II, X447/X447A (DP70 and DP75), were known to fail due to FCCI, none of the pins simulated in BISON reached a cumulative damage fraction above 0.008 with FCCI correlations coupled in the BISON simulations. The error estimate generated for all pins in X447/X447A using optimal values was 209 µm, which is deemed acceptable.

A practical guide to characterizing irradiated nuclear fuels using FIB tomography

Focused ion beam (FIB) tomography with combined electron backscatter diffraction (EBSD) and energy dispersive x-ray spectroscopy (EDS) is a technique capable of statistically characterizing the microstructure and spatial compositional variation of nuclear fuel in three-dimensions (3D). The 3D visualization from FIB tomography provides a comprehensive picture of the interconnected microstructural and compositional features that can impact fuel performance. While these features are often characterized with surface examination, the complexity and relationship of fission products and grain boundary networks may not fully be captured by these 2D methods. This work presents a practical guide to FIB tomography that is tailored to nuclear fuel characterization. The steps used to collect and process the data are provided along with the scripts used to process the data. Additionally, suggestions for future characterization efforts utilizing this approach are given.

Identifying chemically similar multiphase nanoprecipitates in compositionally complex non-equilibrium oxides via machine learning

Characterizing oxide nuclear fuels is difficult due to complex fission products, which result from time-evolving system chemistry and extreme operating environments. Here, we report a machine learning-enhanced approach that accelerates the characterization of spent nuclear fuels and improves the accuracy of identifying nanophase fission products and bubbles. We apply this approach to commercial, high-burnup, irradiated light-water reactor fuels, demonstrating relationships between fission product precipitates and gases. We also gain understanding of the fission versus decay pathways of precipitates across the radius of a fuel pellet. An algorithm is provided for quantifying the chemical segregation of the fission products with respect to the high-burnup structure, which enhances our ability to process large amounts of microscopy data, including approaching the atomistic-scale. This may provide a faster route for achieving physics-based fuel performance modeling.

International case survey and literature review of maintenance strategy for decommissioning nuclear power plant during transition period

The Chinshan nuclear power plant in Taiwan has currently shutdown when the operating license expired in 2018. However, because the operation license of the independent spent nuclear fuel dry storage facility has not been approved yet, and the fuel storage racks of Spent Fuel Pool (SFP) has almost saturated, there is not enough space to accommodate all irradiated fuels from the reactor core during the transition period. Therefore, part of the mechanical systems, plant structures and electrical equipment of the plant still have to meet the functional requirements for maintaining the safety under the pre-decommissioning activities.In this article, we introduce the international cases of permanent shutdown but the fuel still in the reactor core, and also describe the actions of the Taiwan Power Company in response to this situation. For the providing regulatory advice, we have also proposed a framework of the maintenance strategy associated with important issues for the transition period since 2018, including a revised Aging Management Program (AMP) for keeping the safety of spent nuclear fuel. From the knowledge level and important issues related to the topic, the Phenomena Identification and Ranking Table (PIRT) is quoted to complete the understanding level of requirements during the transition period.

Nanoindentation testing of high burn-up fast reactor mixed oxide fuel

Continued research into the mechanical properties of nuclear fuel is necessary to improve modeling of the pellet-cladding mechanical interactions that occur in nuclear reactors during operation. Small scale mechanical testing with focused ion beam equipment can provide a means to study the mechanical properties of irradiated nuclear fuels by measuring their mechanical properties on minute volumes of material. In this work, room temperature nanoindentation measured the hardness and Young's modulus of fast reactor mixed oxide (MOX) fuel irradiated to 183 GWd/tHM. The hardness of the MOX fuel had a measured value of 11.2 ± 0.4 GPa. The Young's modulus of the MOX fuel had a value of 158 ± 6 GPa. The measured hardness and Young's modulus are discussed in the context of the available literature data. In addition, pathways toward elevated temperature testing and other mechanical properties are proposed.

Cyclic Voltammetric Experiment-Simulation Comparisons of the Complex Zr4+ to Zr0 Reduction Mechanism at a Molybdenum Electrode in LiF-CaF2 Eutectic Molten Salt

Nuclear energy represents an important option for generating largely clean CO2-free electricity. Zirconium is a fission product in the nuclear reaction that needs to be extracted from irradiated fuels used in Gen-IV molten salt reactors. The present investigation addresses the electrochemical reduction of solution soluble Zr4+ (soln) to surface confined Zro (s−c) at a molybdenum electrode in a LiF—CaF2 eutectic molten salt at 840 °C using DC cyclic, square-wave and AC voltammetry. Cyclic voltammograms simulated by the reaction scheme: Zr4+ (soln) + 4e− → Zro (soln); Zro (soln) ↔ Zr*(s−c); Zr*(s−c) → Zr4+*(s−c) + 4e−; Zr4+*(s−c) → Zr4+ (soln); Zr*(s−c) + Zr4+ (soln) ↔ 2Zr2+ (soln); Zr2+ (soln) ↔ Zr2+*(s−c) and Zr2+*(s−c) → Zr4+*(s−c) + 2e− provided excellent agreement with experimental data over the scan rate range of 50 to 500 mV s−1. The interpretation of the simulation is that the reduction of Zr4+ (soln) to Zro (metal) takes place via a transiently soluble Zro (soln) in an overall 4-electron essentially reversible diffusion-controlled process having a reversible formal potential (Eo f) of −1.22 V (vs Pt). A minor oxidation process observed at −0.455 V (vs Pt) on the reversing the potential scan direction is simulated via the reaction step Zr(s−c) + Zr4+ (soln) ↔ 2Zr2+ (s−c) followed by Zr2+ (s−c) → Zr4+ (sol) + 2e−. The sharply rising initial component where reduction of Zr4+ (sol) commences, contains evidence of a nucleation-growth mechanism associated with the electrocrystallisation of zirconium metal. This initial rapid growth of current is not fully accommodated in the simulations, but all features found beyond the peak potential are supported by the theory. A comparison with theory based on a direct reduction of Zr4+ (soln) to the metallic state having unit activity is provided. It is proposed that an analogous mechanism applies at a Ni electrode, except that a Ni-Zr alloy formation occurs instead of Zr metal.

Application of the SCIANTIX fission gas behaviour module to the integral pin performance in sodium fast reactor irradiation conditions

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their “pre-INSPYRE” versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the “post-INSPYRE” code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

Understanding fission gas bubble distribution, lanthanide transportation, and thermal conductivity degradation in neutron-irradiated α-U using machine learning

U 10Zr based metallic nuclear fuel is the leading candidate for next-generation sodium-cooled fast reactors in the United States. US research reactors have used and tested this fuel type since the 1960s and accumulated considerable experience and knowledge about the fuel performance. Most of the knowledge, however, remains empirical. The lack of mechanistic understanding of fuel performance puts a large burden on proof through experimental verification for the qualification of U 10Zr fuel for commercial use. This paper proposes an image data-driven machine learning approach, coupled with domain knowledge provided by advanced post irradiation examination, to provide unprecedented quantified insights into the morphology, size, density and the connectivity of fission gas bubbles and their effects on the fission product transportation and thermal conductivity. Specifically, we developed a method to automatically detect, extract statistics, and classify ~19,000 fission gas bubbles into different categories, and quantitatively link the data to lanthanide transportation through connected bubbles and degradation of thermal conductivity along the radial temperature gradient in a neutron irradiated U 10Zr annular fuel. Results indicate the approach can be modified to study other irradiation effects, such as secondary phase redistribution and gaseous fuel swelling in other irradiated nuclear fuels. • This study showcases a novel method utilizing an image data-driven machine learning approach for advanced post irradiation examination. • The features and statistics of 19,000 fission gas bubbles are derived by the decision tree algorithm. • The bubble statistics provide quantified insights into microstructural features of irradiated fuel and how they relate to fuel performance.

Fuel Particle Heat Transfer Part 1: Convective Cooling of Irradiated Fuel Particles

ABSTRACT Wildfire spread requires fuel particles heated to ignition but the roles of radiation and convection heat transfer have not before been examined in detail. This paper reports on laboratory experiments and numerical modeling of wood particle response when subjected to a fixed radiant flux. The experiments used a propane-fueled radiant panel with an elliptical mirror system to focus a constant radiant flux on fuel particles of different sizes and cross-sectional shapes. Particle temperature was measured with fine-wire thermocouples embedded on the surface. Particle heating during free and forced convection with ambient air was found to depend on particle size, specifically the convective length of the irradiated surface. Ambient convection cooled 1 mm square particles sufficiently to prevent ignition but not 3 mm particles and larger. Surface area-to-volume ratio did not govern particle surface heating but did determine a particle’s thermal response rate. A finite difference, numerical method was developed to solve a two-dimensional, transient conduction model with radiative and convective boundary conditions. Model results were reliable when compared to experimental results and confirmed particle surface length (convection) as a principal determining factor governing fuel particle heating and fuel particle ignition.

Serpent-Openfoam Coupled Calculations of Irradiated Nuclear Fuel

Serpent-Openfoam Coupled Calculations of Irradiated Nuclear Fuel

Postirradiation characterization of palladium as an additive for fuel cladding chemical interaction mitigation in metallic fuel

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Thus, the efficacy of using this additive needs further investigation.

Twenty years of research on spent nuclear fuel in the context of final disposal at KIT-INE in multinational collaborative projects

Abstract. Disposal of spent nuclear fuel (SNF) in deep geological repositories is considered a preferential option for the management of such wastes in many countries with nuclear power plants. With the aim to permanently and safely isolate the radionuclide inventory from the biosphere for a sufficient time, a multibarrier system consisting of technical, geotechnical and geological barriers is interposed between the emplaced waste and the environment. In safety assessments for deep underground repositories, access of water, followed by failure of canisters and finally loss of the cladding integrity is considered in the long-term. Hence, evaluating the performance of SNF in deep geological disposal systems requires process understanding of SNF dissolution and rates as well as quantification of radionuclides release from SNF under reducing conditions of a breached container. In order to derive a radionuclide source term, the SNF dissolution and alteration processes can be assigned to two steps: (i) instantaneous release of radionuclides upon cladding failure from gap and grain boundaries and (ii) a long-term release that results from dissolution of the fuel grains itself (Ewing, 2015). In this context, research at KIT-INE has focused for more than 20 years on the behavior of SNF (irradiated UO2 and MOX fuels) under geochemical conditions (pH, redox and ionic strength) representative of various repository concepts, including the interaction of SNF with backfill material, such as bentonite as well as the influence of iron corrosion products, e.g. magnetite and radiolytic reactions on SNF dissolution mechanisms. Since 2001, KIT-INE has contributed with experimental and theoretical studies on the behavior of SNF under repository relevant conditions to six Euratom projects viz SFS (2001–2004), NF-PRO (2004–2006), MICADO (2006–2009), RECOSY (2007–2011), FIRST-Nuclides (2012–2014) and DISCO (2016–2021). Moreover, since 2007, overall 4 consecutive projects for the Belgian waste management organization, ONDRAF-NIRAS, were performed on the behavior of SNF under conditions representative of the Belgian “Supercontainer” concept. In this contribution, we summarize major achievements of theses research projects to understand and quantify the radionuclide release from dissolving SNF under repository conditions. In particular, the dependence of radionuclide release on the chemical composition of the aqueous and gaseous phase in the proximity of repositories in different types of host rock is discussed.

SUPERFACT: A Model Fuel for Studying the Evolution of the Microstructure of Spent Nuclear Fuel during Storage/Disposal.

The transmutation of minor actinides (in particular, Np and Am), which are among the main contributors to spent fuel α-radiotoxicity, was studied in the SUPERFACT irradiation. Several types of transmutation UO2-based fuels were produced, differing by their minor actinide content (241Am, 237Np, Pu), and irradiated in the Phénix fast reactor. Due to the high content in rather short-lived alpha-decaying actinides, both the archive, but also the irradiated fuels, cumulated an alpha dose during a laboratory time scale, which is comparable to that of standard LWR fuels during centuries/millenaries of storage. Transmission Electron Microscopy was performed to assess the evolution of the microstructure of the SUPERFACT archive and irradiated fuel. This was compared to conventional irradiated spent fuel (i.e., after years of storage) and to other 238Pu-doped UO2 for which the equivalent storage time would span over centuries. It could be shown that the microstructure of these fluorites does not degrade significantly from low to very high alpha-damage doses, and that helium bubbles precipitate.

Thermodynamics and the Separation Factor of Dy/U Couple on Ga and Ga–Al Electrodes in Fused LiCl–KCl Eutectic

The radiological danger of spent nuclear fuel (SNF) mainly depends on the presence of minor actinides (Np, Am, Cm) and certain long-lived fission products (Cs, I, Tc). If these elements are effectively separated from SNF and converted into short-lived or stable isotopes, the radiological danger of SNF storage can be significantly reduced. Pyrochemical reprocessing of spent nuclear fuel represents a promising alternative to the existing hydrometallurgical technology (solvent extraction PUREX-process) with subsequent transmutation of long-lived radionuclides. Pyrochemical separation methods in molten salts have some potential advantages over hydrometallurgical processes for separating actinides from fission products present in irradiated fuels. High radiation stability of molten salts and liquid metals as well as absence on neutron moderators make possible reprocessing highly active fuel after a short cooling time.In this study the principal thermodynamic properties and the separation factor of U/Dy couple were determined on gallium and gallium-aluminum electrodes in molten LiCl–KCl eutectic at 723–823 K using open-circuit potentiometry.The equilibrium electrode potentials of metallic dysprosium, uranium and their alloys with Ga and Ga-Al eutectic in molten LiCl–KCl–MeCl3 solutions were determined by open-circuit potentiometry. The potential-time dependencies of Me–Ga–Al and Me–Ga alloys (Me = Dy, U) after cathodic polarization are presented in Fig. 1.The electrochemical behavior of the alloys is characterized by the apparent standard potential of the alloys, . Temperature dependencies of the apparent standard potentials of Me(Ga–Al) and Me(Ga) alloys were fitted to the following expressions using the Origin Pro software (ver. 7.5):E** Dy(Ga-Al) = –(2.705±0.009) + (3.5±0.1) ·10–4∙T ±0.005 V [1]E** U(Ga-Al) = –(2.688±0.013) + (6.4±0.1) ·10–4∙T ±0.005 V [2]E** Dy(Ga) = –(2.726±0.007) + (3.4±0.2) ·10–4∙T ±0.009 V [3]E** U(Ga) = –(2.565±0.004) + (4.2±0.1) ·10–4∙T ±0.003 V [4]Activity coefficients of Dy and U in liquid Ga and Ga–Al alloys were calculated from the following equation:log(γMe(alloy)) = [nF(E* Me(III)/Me – E** Me(alloy))]/2.303RT [5]Values of Dy(III)/Dy and U(III)/U equilibrium electrode potentials were taken from our previous works (1, 2).E* Dy(III)/Dy = –(3.401±0.009) + (6.2±0.1) ·10–4∙T ±0.007 V [6]E* U(III)/U = –(3.051±0.006) + (7.6±0.1) ·10–4∙T ±0.003 V [7]Activity coefficients of solid U and Dy in liquid Ga and Ga–Al alloys as a function of temperature were fitted to the following equations:log(γDy(Ga)) = 4.24 – 10224/T ±0.12 [8]log(γDy(Ga-Al)) = 4.09 – 10542/T ±0.12 [9]log(γU(Ga)) = 5.15 – 7361/T ±0.14 [10]log(γU(Ga-Al)) = 1.16 – 4962/T ±0.17 [11]Partial excess Gibbs free energy change of Dy and U in liquid Ga and Ga–Al alloys was calculated according to eq. [13] and previously obtained expressions [8–11]:ΔGex Me(alloy) = ΔHMe(alloy) – TΔSex Me(alloy) [12]ΔGex Me(alloy) = 2.303RTlog(γMe(alloy)) [13]ΔGex Dy(Ga) = –195.4 + 0.081· T ±2.8 kJ/mol [14]ΔGex Dy(Ga-Al) = –201.5 + 0.078· T ±2.8 kJ/mol [15]ΔGex U(Ga) = –140.7 + 0.098· T ±3.1 kJ/mol [16]ΔGex U(Ga) = –140.7 + 0.098· T ±3.1 kJ/mol [17]When the separation of lanthanides and actinides in the spent nuclear fuel is studied, the effectiveness of electrochemical separation methods is usually described by the value of the distribution or separation factor. The separation factor for uranium and dysprosium on liquid gallium and gallium–aluminum based alloys was calculated using the following equation:log(θ) = [(n–m)FE + mFE** Dy(alloy) – nFE** U(alloy)]/2.303RT [18]The temperature dependence of the apparent standard potentials of dysprosium [1, 3] and uranium [2, 4] alloys were used to derive the following expressions for the separation factor of Dy/U couple:log(θ(Ga)) = 1.21 + 2438/T ±0.04 [19]log(θ(Ga-Al)) = 4.39 + 275/T ±0.6 [20]The separation factor of Dy/U couple in molten LiCl–KCl eutectic calculated according to the above formula indicates that dysprosium will be concentrated in the molten salt phase and uranium in the liquid metal phase. Analysis of experimental data shows that these systems are interesting for application in pyrochemical methods of nuclear waste disposal.The study was carried out with the financial support of the RFBR in the framework of the scientific project # 20-03-00743.References А. Novoselova, V. Smolenski, and V. A. Volkovich, J. Electrochem. Soc. 167, 112510 (2020). A. Novoselova and V. Smolenski, J. Radioanal. Nucl. Chem., 326, 621 (2020). Fig. 1. The potential-time dependencies of Me–Ga and Me–Ga–Al alloys (Me = Dy, U) after cathodic polarization in LiCl–KCl–MeCl3 solutions at 726 K. Polarization current 50–150 mA, duration 20–40 s. [U(III)] = 4.7·10-2 mol·kg-1. [Dy(III)] = 5.8·10-2 mol·kg–1.1 – The equilibrium potential of U(Ga-Al) alloy. 2 – The equilibrium potential of U(Ga) alloy.3 – The equilibrium potential of Dy(Ga-Al) alloy. 4 – The equilibrium potential of Dy(Ga) alloy. Figure 1

Microstructural and Micro-Chemical Evolutions in Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles

AGR-1 and AGR-2 tristructural-isotropic (TRISO) fuel particles were fabricated using slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. Extensive microstructural and analytical characterizations were conducted to correlate those differences with the fuel kernels’ responses to neutron irradiations in terms of irradiated fuel microstructure, fission products’ chemical and physical states, and fission gas bubble evolutions. The studies used state-of-the-art transmission electron microscopy (TEM) equipped with energy-dispersive x-ray spectroscopy (EDS) via four silicon solid-state detectors with super sensitivity and rapid speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen in order to create representative irradiation conditions with fuel burnup in the range of 10.8 to 18.6% fissions per initial metal atom (FIMA) and time-average volume-average temperatures varying from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% and 14.03 wt.% for the AGR-1 and AGR-2 fuel kernels, respectively. The TEM results showed significant microstructural reconstructions in the irradiated fuel kernels from both the AGR-1 and AGR-2 fuels. There are four major phases: fuel matrix of UO2 and UC, U2RuC2, and UMoC2—in the irradiated AGR-2 fuel kernel. Zr and Nd form a solid solution in the UC phase. The UMoC2 phase often features a detectable concentration of Tc. Pd was mainly found to be located in the buffer layer or associated with fission gas bubbles within the UMoC2 phase. EDS maps qualitatively show that rare-earth fission products (Nd et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR-1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there was a high number of rod-shaped precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The differences in irradiated fuel kernel microstructural and micro-chemical evolution when comparing AGR-1 and AGR-2 TRISO fuel particles may result from a combination of irradiation temperature, fuel geometry, and chemical composition. However, irradiation temperature probably plays a more deterministic role. Limited electron energy-loss spectroscopy (EELS) characterizations of the AGR-2 fuel kernel show almost no carbon in the UO2 phase, but a small fraction of oxygen was detected in the UC/UMoC2 phase.

The Release of C-14 of HTR Graphite Spheres during High Temperature Oxidation

As the key components of the HTGR fuel, the intact TRISO-coated particles contain almost all fission products under the normal and accident conditions in the reactor. On the other hand, the high temperature oxidation of irradiated fuel pebbles is one of feasible processing methods for HTR spent fuel pebbles. The knowledge of the C-14 release from irradiated fuels and the migration behavior during high temperature oxidation is of great importance for the post-irradiation examinations (PIE) and long-term storage and reprocessing of HTR-PM fuel pebbles. In this work, the C-14 release behavior from one irradiated HTR graphite sphere, which was used as a substitute for a irradiated HTR fuel, was determined by out-of-pile examination. The irradiated HTR fuel was oxidized in the hot lab of INET, and the substantial migration of fission products from graphite matrix was adsorbed and systematically studied by γ-scanning. Specifically, the C-14 release during oxidation was analyzed and illustrated. This work will provide theory and experiment basement for spent fuel disposal of HTGR, as well as a new sample preparation method for C-14 measurement by liquid scintillation counter (LSC).

Comparison of Irradiated TRISO Fuel Radioactivity from Multiple Advanced Reactor Designs

The SCALE code system was used to model, deplete, and compare several different tri-structural isotropic (TRISO)–fueled reactor designs: a helium-cooled prismatic reactor, a helium-cooled pebble-bed reactor (PBR), and a fluoride-lithium-beryllium (FLIBE) molten-salt-cooled PBR. The purpose of this comparison was to understand how differences in the reactor designs affect the radioactivity of the fuel after discharge and whether those differences are significant. First, the various reactor designs were built and depleted in the TRITON module for each design and fuel enrichment. Then, the TRITON outputs were used to create burnup-dependent reactor libraries. These libraries were then used by ORIGEN to determine the activities of discharged fuel for each reactor, which were compared to generic Westinghouse 17 × 17 fuel. Overall, the results showed that short-term activities are dominated by reactors with higher operating powers, and the reactor type, initial fuel enrichment, and maximum burnup are of only secondary importance. Although this analysis only focuses on activities in Becquerels, these dependencies are consistent with the expected behavior of decay heat. However, analysis of long-term time periods post irradiation shows that the reactor type and maximum burnup have strong impacts on the activities; initial fuel enrichment has a secondary impact while operating power is inconsequential. These results would be useful for analyses, such as dose assessment and modeling in postrelease scenarios, normal fuel handling operations, and spent fuel transport, storage, and disposal. Of particular interest, the results in this technical note show that analyses that focus on spent nuclear fuel of advanced reactors need to consider each parameter carefully. Unsurprisingly, if the correct operating power is not used in short-term analyses, the results will not be correct. Perhaps unexpectedly, however, if the correct reactor type is not used, then the long-term results will also be incorrect, especially for areas such as permanent disposal. Even though this technical note focuses on the total activity of nuclear fuel, it provides initial results on the effects of various input parameters and also provides a framework to extend the work into other analyses of spent fuel from advanced reactors, especially those employing TRISO fuel.