Fuel-cladding Chemical Interaction Research Articles

Fuel-Cladding Chemical Interaction of a Prototype Annular U-10Zr Fuel with Fe-12Cr Ferritic/Martensitic Cladding

As an alternative fuel form, annular metallic fuel design eliminated the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. Here, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom (FIMA). Compared to sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium bonded annular fuel. Instead, most of the lanthanides were retained in the newly formed UZr2 phase at the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal (U,Zr)6Fe phase (space group I4/mcm) and cubic (U,Mo)(Fe,Cr)2 phase (space group Fd3 m). The (U,Mo)(Fe,Cr)2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm3 m). At the interface of the interdiffusion zone and inner cladding, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body centered cubic (BCC) ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P42/mnm) occurred and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.

Experimental assessment of antimony (Sb) in pure uranium for immobilizing fission product lanthanides

The use of fuel additives is one of the concepts to mitigate fuel cladding chemical interaction (FCCI) for metallic fuel because fission product lanthanides are expected to be immobilized by the additive. Antimony (Sb) has been found to be a good additive candidate for UZr fuel. The present study focuses on the immobilization mechanism in pure uranium. U–4Sb alloy was fabricated to understand the behavior of Sb, while U–4Sb–4Ce was fabricated to simulate the condition when lanthanides are generated. The as-cast alloys were characterized by Scanning Electron Microscope (SEM) and Energy Dispersive Spectrometer (EDS). U–Sb precipitates are formed in U–4Sb alloy, while U–Sb and Ce–Sb precipitates were found in U–4Sb–4Ce alloy. The alloys that were heated under 650 °C for 500 h were also characterized by SEM/EDS and the results shows that the thermal exposure does not change the morphologies or chemical composition of the Sb-precipitates (U–Sb and Ce–Sb precipitates). In addition, diffusion couple tests between those alloys and cladding materials (Fe or HT9) under 650 °C for 500 h were performed and the samples were then analyzed using SEM/EDS. The reactions between cladding and uranium were found, while there was no reaction between Sb-precipitates and cladding materials.

Efficient computational search for lanthanide-binding additive dopants for advanced U-Zr based fuels

A rational design strategy combining theory and experiment is highly desirable for the development of U-Zr based fuels that are resistant to lanthanide-induced fuel-cladding chemical interaction (FCCI). We performed rapid computational screening of FCCI-mitigating (via strong lanthanide-binding) elements across the periodic table by utilizing three simple criteria and density functional theory (DFT) calculated total energies of relevant elements and compounds retrieved from high-throughput DFT databases. In addition to successfully identifying previously demonstrated additives, our search leads to the discovery of Bi as a highly effective element among all 68 elements screened, and its effectiveness was experimentally confirmed in this study.

Evaluation of Tellurium as a Fuel Additive in Neodymium-Containing U-Zr Metallic Fuel

Phase-stability in a U-Zr-Te-Nd multi-component metallic fuel for advanced nuclear reactors is systematically investigated by taking into account binary, ternary and quaternary interactions between elements involved. Historically, the onset of fuel-cladding chemical interactions (FCCI) greatly limits the burnup potential of U-Zr fuels primarily due to interactions between lanthanide fission products and cladding constituents. Tellurium (Te) is evaluated as a potential additive for U-Zr fuels to bind with lanthanide fission products, e.g. neodymium (Nd), negating or mitigating the FCCI effect. Potential fresh fuel alloy compositions with the Te additive, U-Zr-Te, are characterized. Te is found to completely bind with Zr within the U-Zr matrix. Alloys simulating the formation of the lanthanide element Nd within U-Zr-Te are also evaluated, where the Te-Nd binary interaction dominates and NdTe is found to form as a high temperature stable compound. The experimental observations agree well with the trends obtained from density functional theory calculations. According to the calculated enthalpy of mixing, Zr-Te compound formation is favored in the U-Zr-Te alloy whereas NdTe compound formation is favored in the U-Zr-Te-Nd alloy. Further, the calculated charge density distribution and density of states provide sound understanding of the mutual chemical interactions between elements and phase-stability within the multi-component fuel.

Segregation behavior and embrittling effect of lanthanide La, Ce, Pr, and Nd at Σ3(111) tilt symmetric grain boundary in α-Fe**Project supported by the National Natural Science Foundation of China (Grant No. U1867217), the National Science and Technology Major Project of the Ministry of Science and Technology of China (Grant No. 2019ZX06004009), and the China National Nuclear Corporation Centralized Research and

The migration of lanthanide fission products to cladding materials is recognized as one of the key causes of fuel–cladding chemical interaction (FCCI) in metallic fuels during operation. We have performed first-principles density functional theory calculations to investigate the segregation behavior of lanthanide fission products (La, Ce, Pr, and Nd) and their effects on the intergranular embrittlement at Σ3(111) tilt symmetric grain boundary (GB) in α-Fe. It is found that La and Ce atoms tend to reside at the first layer near the GB with segregation energies of −2.55 eV and −1.60 eV, respectively, while Pr and Nd atoms prefer to the core mirror plane of the GB with respective segregation energies of −1.41 eV and −1.50 eV. Our calculations also show that La, Ce, Pr, and Nd atoms all act as strong embrittlers with positive strengthening energies of 2.05 eV, 1.52 eV, 1.50 eV, and 1.64 eV, respectively, when located at their most stable sites. The embrittlement capability of four lanthanide elements can be determined by the atomic size and their magnetism characters. The present calculations are helpful for understanding the behavior of fission products La, Ce, Pr, and Nd in α-Fe.

U-10Zr and U-5Fs: Fuel/cladding chemical interaction behavior differences

A fuel performance phenomenon that has previously limited the peak cladding temperature of metallic fast reactor fuel is fuel/cladding chemical interaction (FCCI). This study examined the differences in the phenomenon exhibited by the more recent fuel alloy, U-10Zr and a previously used fuel alloy, U-5Fs (Fs – fissium). Fs is comprised of Mo, Ru, Pd, Zr and Rh simulated fission products representing reprocessing carryover. Studies on the U-Zr alloy had shown the often controlling influence on FCCI of lanthanide (Ln) fission products (Ln: Ce, Nd, Sm, etc.) while a review of the U-Fs alloy behavior indicated a dominance of basic fuel/cladding interdiffusion (Fe, Ni with U). The recent studies on U-Zr fuel showed that migration of lanthanide fission products to the fuel/cladding interface facilitated the Ln dominance of FCCI behavior. This study was focused on understanding why Ln’s do not migrate in the U-5Fs alloy to the extent that they do in U-10Zr, or why they do not interact with the cladding like in U-10Zr fuel pins. A small fuel/cladding interaction layer was found containing the cladding and lanthanide components. However, it was also found that the lanthanide fission products are at least partially bound to agglomerates that also contained Pd, Rh and Zr, all minor components of Fs.

Diffusion behaviors between metallic fuel alloys with Pd addition and Fe

Fission product lanthanides in metallic fuels are known to cause adverse fuel-cladding chemical interaction (FCCI). Palladium (Pd) is being explored as a potential additive to reduce or mitigate FCCI by forming stable Pd-Ln compounds. The current study is an investigation of diffusion behaviors between UZrPd, UZrPdLn, UPuZrPd, and UPuZrPdLn alloys (Ln = 53Nd-25Ce-16Pr-6La wt. %) and iron (Fe). Diffusion couple tests were performed followed by microstructural analysis using scanning electron microscopy (SEM) and transmission electron microscopy (TEM). It is found that the diffusion of Fe into the fuel is reduced by PdZr2 precipitates randomly dispersed throughout the fuel matrix. Ln-Pd compounds form in UZrPdLn alloys, and (Ln,Pu)Pd forms in the UPuZrPdLn alloy. In both diffusion couples with Ln, small amounts of lanthanides are diffusing into the Fe. This could be due to small amounts of dissolved lanthanides in the precipitates, or could be due to the Ln-Pd compounds decomposing. In either case, diffusion caused by lanthanides is greatly reduced as compared to alloys without Pd.

A study of constituent redistribution in U–Zr fuels using quantitative phase-field modeling and sensitivity analysis

Constituent redistribution in U–Zr fuels produces radially-distributed phase fields, each with different material properties and behaviors. The location and composition of the phase fields evolve dynamically due to the influence of swelling, fission gas release, and sodium infiltration on the thermal conductivity of the fuel. Gaps in the understanding of these processes limit the ability to model higher-level behaviors such as fuel-cladding chemical interaction and perform design and safety analyses with confidence. In the current work, we developed a quantitative phase-field model of macroscale constituent redistribution in the U–Zr system. Model parameters were optimized, and the model was validated against an independent dataset. No modification of the phase transition temperatures was necessary, but diffusion needed to be enhanced to reproduce the observed behaviors. Sensitivity analysis was then used to investigate gaps in the understanding of the system and identify the most impactful parameters and behaviors. The mobile interface between the γ and β+γ phase fields was found to be particularly influential in forming the Zr bathtub. The β and γ kinetic parameters scale the overall magnitude of constituent redistribution, while parameters associated with swelling and sodium infiltration influence the size, shape, and location of the Zr bathtub. These findings were used to justify specific recommendations for improving U–Zr modeling capability, which will expedite fuel development and qualification.

Microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln

Palladium is being investigated as a potential additive to metallic fuel to bind fission product lanthanides, with the goal of reducing or preventing fuel-cladding chemical interactions (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study is the microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln in wt. %, where Ln = 53Nd-25Ce-16Pr-6La, using scanning electron microscopy (SEM), transmission electron microscopy (TEM), and X-ray diffraction (XRD). In both alloys, the matrix is comprised of ζ-U0.4Pu0.6 and δ-(U,Pu)Zr2. Based on the matrix compositions, modifications to the room temperature extrapolated U-Pu-Zr ternary phase diagram are suggested. In U-20Pu-10Zr-3.86Pd, there is very little δ phase, due to formation of PdZr2. In U-20Pu-10Zr-3.86Pd-4.3Ln, a lanthanide-rich phase is present, although it does not have the crystal structure for 53Nd-25Ce-16Pr-6La. This phase cannot be identified based on known compounds. (Ln,Pu)Pd is the primary lanthanide phase, with Pu substituting into the crystal structure in place of Nd.

Diffusion behavior of lanthanide-additive compounds (Ce4Sb3, Ce2Sb, and CeTe) against HT9 and Fe

Antimony and tellurium have been identified as promising additives in metallic fuel, which can immobilize free-lanthanide fission products into stable intermetallic compounds in order to mitigate the fuel-cladding chemical interaction. Ce4Sb3, Ce2Sb, and CeTe are the primary compounds formed by Sb or Te with the lanthanide Ce present in the fuel. If these compounds are present at the outer periphery of the fuel, they will come in contact with and react with the cladding after the fuel swells. The present study investigates the reactivity of these compounds with two cladding materials, HT9 and Fe. The diffusion couple tests between these compounds and HT9 or Fe were conducted at 853 K. Scanning electron microscopy and transmission electron microscopy were used to analyze the morphology, microstructure, and phase distribution of the diffusion region. It was observed that the diffusion region thickness formed by the three compounds was significantly reduced compared to free Ce. There was no observed diffusion or reaction between Ce4Sb3 or Ce2Sb with either HT9 or Fe. CeTe was found to diffuse and react with HT9, forming Cr3Te4 and TeFe at the diffusion region, as well as to penetrate into Fe, mostly by intergranular diffusion.

Fuel clad chemical interaction of U-Mo fast reactor fuel

U-33 at.% Mo (16.8 wt.%) metallic fuel is a candidate material for metallic fuel Fast Reactor. One of the life-limiting factors of fast breeder reactor clad is the fuel clad chemical interaction due to formation of low melting eutectics between U and Fe. This chemical interaction should be avoided or minimized to increase the fuel burn-up. Fuel clad chemical interaction between U-33 at.% Mo metallic fuel with T91 (9 Cr-1Mo) clad has been studied at 650 °C, 675 °C and 700 °C, for different soaking time, through diffusion couple experiments. Development of microstructures, phase constituents and compositions during thermal treatment were examined by scanning electron microscopy and X-ray energy dispersive spectroscopy. In the fuel side, U6Fe phase is formed along with bcc-Mo in lamellar morphology through cellular precipitation. Due to slower diffusion of Mo compared to uranium from U-33 at.% Mo fuel, a Mo rich layer is formed at the slug surface. This Mo rich layer subsequently acts as a diffusion barrier layer and minimizes further growth of clad-wastage zone and slug-interaction zone. Multi-phase layer growth constants and activation energies have been calculated and compared with interactions reported in literature of other U-based fuels and cladding elements. The study indicates that U-33 at.% Mo/T91 fuel reduces fuel clad chemical interaction significantly compared with U-23Zr/Fe and U-23Zr/Fe-Cr.

Assessment of Te as a U-Zr fuel additive to mitigate fuel-cladding chemical interactions

Tellurium was investigated as a potential additive to metallic fuel (e.g. U-10Zr) to mitigate fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is the lanthanide fission products migrating to the fuel periphery and interacting with cladding, Te therefore was used to bind the lanthanides into stable compounds in the U-10Zr. The present work investigates the microstructures of as-cast and annealed U-10Zr-4Te and U-10Zr-4Te-4Ce alloys (weight percent) with scanning electron microscope and transmission electron microscopy. Tellurium was found to bind Ce and form CeTe compound, which will mitigate lanthanides migration to the cladding.

Studies on fuel-cladding chemical interaction between U−10 wt%Zr alloy and T91 steel

The chemical interaction of U−Zr based fast reactor fuel with T91 grade steel cladding was investigated by employing DTA (differential thermal analysis), diffusion couple experiment and SEM/EDS, SEM/EBSD analyses. The DTA results revealed a strong exothermic reaction just below the solidus temperature of the U−Zr alloy. The eutectic liquefaction temperature between U−Zr alloy and T91 steel was found to be at 995 K. The interdiffusion between U–10Zr and T91 resulted in the formation of UFe2-type phase, a Zr-rich layer and a Zr-depleted layer at the interface. The growth kinetics of the reaction zone was also determined using the total width of UFe2 and Zr-rich reaction layers. The activation energy for the reaction was found to be 170 kJ mol−1. The Zr-rich layer at the interface acts as diffusion barrier which enhances the fuel-cladding chemical compatibility at reactor operating temperature. However, the U−Zr fuel is no more compatible with the cladding above the eutectic liquefaction temperature (995 K).

Study on reaction of 15-15Ti with fission products under various oxygen potential

ABSTRACT The effect of oxygen potential on the Fuel-Cladding Chemical Interaction (FCCI) of 15-15Ti cladding material was studied in this paper. In the test, the materials have been exposed to simulated fission products at 600◦C for 120 h under the oxygen potential in equilibrium with Mo/MoO2 and Cr/Cr2O3.The general corroded depth was roughly 30 µm. Furthermore, the results also show that corroded depth on the cladding surface increase with the increase of oxygen potential. No intergranular attacks were observed, which might be caused by the absence of the high-temperature gradient. A single layer of CrTe is assumed to be formed under low oxygen potential, whereas cesium and oxygen may play an important role under high oxygen potential. Moreover, FCCI may impose a great threat to the integrity of China designed 15-15Ti material.

Behaviors of Ce, Pr, and Nd in liquid cesium by ab initio molecular dynamics simulations

The fuel cladding chemical interaction (FCCI) phenomenon is potentially the main factor restricting the application of metallic fuels in liquid sodium cooled fast reactors. The understanding of the lanthanide (Ln) transport behaviors in liquid Cs filled pores in U-Zr fuel is essential for understanding FCCIs. By using ab initio molecular dynamics, fundamental properties of the metallic system Cs-Ln, such as density of states and coordination number, have been studied. Then, the Ln diffusivities in liquid cesium and the solution viscosity were calculated. For validating the model, the viscosity of the pure liquid Cs which has been well measured is also calculated at three temperatures, which indicates the present model has a high accuracy in calculation of viscosity and self-diffusivity of Cs in liquid Cs.

Characterization of U-10Zr-2Sn-2Sb and U-10Zr-2Sn-2Sb-4Ln to assess Sn+Sb as a mixed additive system to bind lanthanides

Tin and antimony are being investigated as potential additives to metallic fuel to control fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction can lead to wastage of the cladding and, given enough time or burn-up, eventually to a cladding breach. The current study involves a microstructural characterization of as-cast and annealed U-10Zr-2Sn-2Sb and U-10Zr-2Sn-2Sb-4Ln, where Ln = 53Nd-25Ce-16Pr-6La, all in wt %, as alloys which could prevent FCCI. Scanning electron microscopy (SEM) analysis and X-ray powder diffraction indicates the additive-Zr compound in U-10Zr-2Sn-2Sb is Zr5(Sn,Sb)3, while the additive-Ln compound being formed in U-10Zr-2Sn-2Sb-4Ln is Ln5(Sn,Sb)4, with significantly more Sb than Sn (37 at. % versus 6 at. %, respectively). The bulk of the Sn, with a small amount of Sb, remained as Zr5(Sn,Sb)3 precipitates. The potential benefits of a mixed additive system, along with relative stabilities of the intermetallic compounds, are discussed.

Postirradiation examination results of several metallic fuel alloys and forms from low burnup AFC irradiations

Several different U based metallic nuclear fuel alloys and forms were irradiated as part of the Advanced Fuels Campaign (AFC) series of irradiation experiments. Results from postirradiation examinations of the AFC-3A and AFC-3B are presented here. In this irradiation, U alloyed with Mo, Zr, and Zr with Pd were investigated. Different geometric forms (solid, annular), smear densities and bonding media (sodium, helium) were also irradiated and evaluated. All fuel forms were irradiated in ferritic-martensitic HT-9 cladding. Examination of the irradiated fuel included neutron radiography, evaluation of the distribution of gamma-ray emitting radionuclides, dimensional inspection, fission gas release analysis, optical microscopy, and chemical analysis. No in-pile failure was observed with this fuel. Neutron radiography, gamma spectrometry, and dimensional inspections indicated the fuel largely performed acceptably. Optical microscopy revealed significant fuel-cladding interaction between the Mo based alloys and the cladding. The Zr based alloys generally performed better although some forms revealed behaviors that present specific fabrication challenges such as tighter fabrication tolerances that will need to be better controlled to match historical U-Zr fuel performance. The addition of Pd as an additive to prevent fuel cladding chemical interaction (FCCI) also appears to present some challenges to fuel performance.

Evaluation of Fuel-Clad Chemical Interaction in PFBR MOX test fuel pins

Fuel Clad Chemical Interaction (FCCI) is one of the life limiting issues in the MOX fuel pins of fast breeder reactors. Clad wastage due to FCCI coupled with stress arising from fission gas pressure and Fuel Clad Mechanical Interaction (FCMI) due to fuel swelling can lead to fuel pin failure. Non-destructive evaluation (NDE) techniques such as Gamma Scanning, Eddy current testing and Neutron Radiography have been successfully used on MOX fuel pins of a test sub-assembly irradiated in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GW d/t to detect signatures of FCCI. The results obtained by the various NDE techniques have been correlated and verified through metallographic examination on fuel pin cross-sections. The measured clad wastage of 85 μm agrees well with a model developed for MOX fuel with high Pu content. The results of examinations have enabled validation of the model and given confidence to the designer that PFBR MOX fuel can safely attain the target burn-up of 100 GW d/t.

Microstructural characterization of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln

Palladium is being investigated as a potential additive to metallic fuel to bind fission product lanthanides, with the goal of reducing or preventing fuel-cladding chemical interactions (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study is a scanning electron microscopy (SEM) and transmission electron microscopy (TEM) investigation of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln in wt. %, where Ln = 53Nd-25Ce-16Pr-6La. In U-20Pu-10Zr-3.86Pd, PdZr2 is forming, along with a possible ternary phase between Pu, Zr, and Pd. Pu is also present in the Pd-Ln precipitates formed in U-20Pu-10Zr-3.86Pd-4.3Ln. In the LnPd phase, Pu appears to be substitutional, forming (Ln,Pu)Pd. The other prominent phase, which appears to be Ln7Pd3, has a fine, lamellar structure. The lanthanides remain essentially constant across this fine structure, but Pu and Pd alternate as to which has the higher concentration.

Young's modulus evaluation of high burnup structure in UO2 with nanometer resolution

The mechanical properties of a well-characterized irradiated light water reactor fuel pin with an average burnup of 72 MWd/kgU were investigated by utilizing nano-indentation technique. Young's modulus and hardness are reported as functions of radial positions covering from mid radial position in the urania pellet to high burnup structure, fuel-cladding chemical interaction zone and zirconium-based cladding. By probing microstructure of the high burnup zone, the mechanical properties of sub-grains and the interaction between them were examined at the nanometer scale. Force-displacement curves from each of these individual sub-grains suggest they resemble single crystal UO2. However, these restructured grains are weakly bonded to their neighbors.