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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Transmission electron microscopy characterization of the fuel-cladding chemical interactions in HT9 cladded U-10Zr fuel

Fuel cladding chemical interaction (FCCI) is a key phenomenon needs to be better understood to establish the design basis for U-10Zr metallic fuel performance. Characterizing the microstructure and chemical composition of FCCI at micron and sub-micron scale is critically important toward a more mechanistic understanding of FCCI phenomenon and its potential effects on cladding integrity and metallic fuel performance. This paper, by using transmission electron microscopy, investigated the FCCI region in HT9 cladded U-10Zr fuel irradiated to 5.7% FIMA burnup at a peak inner cladding temperature of 615 °C in Fast Flux Test Facility (FFTF). Four distinct layers are identified in the FCCI region. The migration of Fe into the fuel side leads to the formation of several U-Zr-Fe ternary phases, including ?-Fe0.5Zr0.32U0.18, e-Fe0.3Zr0.4U0.3, and ?-Fe0.06Zr0.23U0.71, mingled with UFe2, U6Fe, and U phase at various Fe penetration depth up to ~ 150 µm. On the cladding side, grain coarsening and significant lanthanides infiltration along grain boundaries are observed. Laves phase, (Fe,Cr)2(Mo,W), which typically does not exist in fresh HT9, is identified in a wide radial range in the cladding. The typical HT9 martensitic lath structure and pre-existing M23C6 precipitates disappear, partially or completely, depending on the radial distance from the fuel-cladding interface. Those microstructural and compositional changes could cause mechanical degradation in the HT9 cladding. The present characterization results will improve the understanding of FCCI phenomenon and facilitate the development of microstructure-informed FCCI modeling for metallic fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Helium pressure increase inside GPHS clads: implications for formation of distended clads

Distension of GPHS clads has generally been attributed to gas buildup inside the clad. The bulging vent frit cover can be caused only by gas pressure, since the vent cover is isolated from the fuel by the vent frit. Gas generation is the only change expected as the oxide fuel ages. However, several data suggest that gas pressure is not the origin of the distension of the clad body. Gas pressure sufficient to distend the iridium body of the clad would be on the order of 1500 psi. This high pressure is very unlikely to be reached inside a GPHS clad, because the vent frit cover weld cannot sustain high pressures. Every distended clad that has been examined has an open vent frit, suggesting that the vent frit cover eventually opens on all distended GPHS clads. No gas pressure has been observed in any distended clad. In addition, the vent frit cover is sufficiently thin, about 2 grains thick, that diffusion of helium through the intact cover is possible and almost certainly occurs. Conversely, no clad has ever been definitively observed to distend after venting. Removal of the vent cover allows free escape of any radiolytically generated helium. However, there is little data on the behavior of vented GPHS. The vent cover is seldom removed at LANL, and vent covers are not removed on clads that will be stored. The vent covers are known to have been removed from the four clads residing in a module which was stored for 8 years. Expansion of the fuel is more likely to provide sufficient force on the iridium clad to distort the iridium metal. Radiographs indicate that the fuel inside a distended GPHS clad is also distended, filling all or most of the interior volume of the GPHS. No radiograph shows the fuel expanded and in contact with the weld shield and the distorted side of the GPHS clad without distension. However, radiography is infrequently done on stored clads unless they exhibit distension, so the dimensions of fuel in normal clads are not generally known.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Innovative coating of nanostructured vanadium carbide on the F/M cladding tube inner surface for mitigating the fuel cladding chemical interactions

Fuel cladding chemical interactions (FCCI) have been acknowledged as a critical issue in a metallic fuel/steel cladding system due to the formation of low melting intermetallic eutectic compounds between the fuel and cladding steel, resulting in reduction in cladding wall thickness as well as a formation of eutectic compounds that can initiate melting in the fuel at lower temperature. In order to mitigate FCCI, diffusion barrier coatings on the cladding inner surface have been considered. In order to generate the required coating techniques, pack cementation, electroplating, and electrophoretic deposition have been investigated. However, these methods require a high processing temperature of above 700 oC, resulting in decarburization and decomposition of the martensites in a ferritic/martensitic (F/M) cladding steel. Alternatively, organometallic chemical vapor deposition (OMCVD) can be a promising process due to its low processing temperature of below 600 oC. The aim of the project is to conduct applied and fundamental research towards the development of diffusion barrier coatings on the inner surface of F/M fuel cladding tubes. Advanced cladding steels such as T91, HT9 and NF616 have been developed and extensively studied as advanced cladding materials due to their excellent irradiation and corrosion resistance. However, the FCCI accelerated by the elevated temperature and high neutron exposure anticipated in fast reactors, can have severe detrimental effects on the cladding steels through the diffusion of Fe into fuel and lanthanides towards into the claddings. To test the functionality of developed coating layer, the diffusion couple experiments were focused on using T91 as cladding and Ce as a surrogate lanthanum fission product. By using the customized OMCVD coating equipment, thin and compact layers with a few micron between 1.5 µm and 8 µm thick and average grain size of 200 nm and 5 µm were successfully obtained at the specimen coated between 300oC and 500 oC, respectively. The coating layer contains both carbon and vanadium elements as quantified by WED, and the phases mainly consist of a mixture of V2C and VC, which was confirmed using X-ray diffraction patterns. In addition, the ratio between V and C varies with processing temperature, and it was observed that a higher temperature promotes the carbon adsorption and increases thickness of the coating. With optimized deposition conditions, we can apply the coating technique toward the actual T91 cladding materials, and provide the possibilities for the real application in sodium-cooled fast reactors (SFRs). Diffusion couple experiments were performed at both 550 oC and 690 oC, which corresponds to normal and aggressive operating temperatures, respectively. The results show that vanadium carbide coating with wider thickness (8 µm) and lower carbon concentration (27 at.%) reduced the width of the inter diffusion region, indicating that vanadium carbide coating can mitigate FCCI effectively. In specific, inter-diffusion between Fe and Ce was prohibited over most area, but Ce diffusion occurred toward the coating and the Fe substrate through thinner coating layer, which needs further optimization in terms of uniform coating thickness. Overall, it is concluded that this coating process can be successfully applied onto the inner surface of HT9 cladding tubes and the FCCI can be effectively mitigated if not totally eliminated.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Fuel cladding chemical interactions (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Here, through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the ferritic/martensitic (F/M) HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature reached to 650 °C. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe 2 Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the F/M HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ?).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectives of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Pellet cladding mechanical interaction as a potential failure mechanism during a control rod drop accident in a boiling water reactor

Boiling water reactors (BWRs) represent approximately one-third of the operating fleet in the United States, contributing significantly towards the global effort in reducing carbon emissions. Even though most of the operating fleet has been in operation for quite some time, continued advancements in new nuclear fuel (such as accident tolerant fuel) or operating regimes (such as power up-rates and higher burnup operation) necessitates similar advancements in modeling and simulation capabilities. Bison, a component of the Virtual Environment for Reactor Applications (VERA), is a high-fidelity fuel performance code able to explore the fuel performance of a wide variety of fuel types in one-, two-, and three-dimensions. Until recently, the code had not been used for analyses of BWRs. Modeling capabilities have been added for Gd-bearing UO{sub 2} and pure zirconium liners. New models have been added based upon the U.S. Nuclear Regulatory Commission (NRC) guide-lines for hydrogen pickup in Zircaloy-2 claddings and pellet-clad mechanical interaction (PCMI) failure during a reactivity insertion accident (RIA), known as a rod drop accident (CRDA) in BWRs. Implementing and/or improving Bison modeling capabilities extended its analytical reach to areas beyond its original intended purpose. This paper demonstrates one of these capabilities as a proof of concept. Recently developed models enable Bison to provide an alternate approach for cladding integrity determination in CRDA evaluations, which currently use bounding conservative estimates. As part of this demonstration, NRC guidance on hydrogen-pickup and PCMI failure models were utilized in this research. Even though more research is needed in establishing right inputs and process in this area, this paper demonstrates Bison's ability to determine cladding integrity in a CRDA evaluation. This first of a kind demonstration is a proof of concept in this area, which could potentially be extended to a number of other areas where a more accurate cladding integrity determination would be needed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Elucidating the effect of minor-actinide addition on fuel-cladding chemical interaction in an HT-9 clad U-Pu-Zr metallic fuel irradiated to 6.15 at.% burnup in EBR-II

Scanning and transmission electron microscopy (S/TEM) were used to characterize the local fuel-cladding chemical interaction (FCCI) in one cross-section taken from a HT-9 clad U-20.3Pu-10Zr-1.2Am-1.3Np (in wt.%) fuel irradiated to 6.15 at.% burnup with inner cladding temperatures ranging between 460–490 °C. Results showed that the total interaction thickness between fuel and cladding was <10 µm. Fe infiltrated the fuel to form U-Zr-Fe phases while fuel elements or lanthanides did not infiltrate into the cladding. Np was not involved in the formation of any phases in the examined locations; however, Am played a role by forming a ∼2 µm thick homogeneous Fe-Pu-Am planar front at the inner cladding wall. An oxidized Na layer existed in the fuel-cladding gap with Fe and lanthanide particles dispersed within, suggesting Na could facilitate the transport of fuel and cladding constituents. Secondary phases, including an FCC Zr-rich phase, lanthanide phases, and α’-Cr(Fe) were identified in the outer fuel and FCCI regions. Furthermore, this study suggests that, for the irradiation conditions specific to this cross-section, minor actinides have little impact on FCCI behavior beyond what would be observed in typical HT-9 clad U-Pu-Zr fuel pins systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Advanced Characterization of Fuel-cladding Chemical Interaction between U-10Zr Fuel and HT9 Cladding Tested in Fast Flux Test Facility

Fuel cladding chemical interaction (FCCI) can greatly accelerate the cladding failure. However, due to the limited space in a fuel cladding assembly, it is historically challenging to gain an mechanistical understanding of the formation mechanism of FCCI and its influence on fuel and cladding performance. With the imminent need to qualify U-10Zr based metallic fuel cladded by HT-9 for advanced reactors demonstration project, it is of vital importance to use advanced characterization method to study FCCI in a unprecedent detailed manner and gain better mechanism understanding of FCCI. Mechanistic Fuel Failure (MFF) series of prototypic fuel elements irradiated in FFTF [1, 2] provides the best samples to study FCCI since the MFF-series assemblies had an axial fuel height the same as proposed length by industry partners. Jason et al. [3] has performed preliminary post-irradiation examination on a MFF fuel pin sample, which was extracted from the HT9 cladded U-10at.%Zr MFF-3 pin MFF-3 pin (#193045) at an axial location of X/L = 0.98. This sample has a peak burnup of 5.7 at.% and peak inner cladding temperature (PICT) of around 615 °C during in-core testing. Scanning electron microscope examination has identified visible FCCI region on more than half of the HT9 circumference [3]. The most striking feature are grain boundary attacking by apparently lanthanides (Lns) rich phase. However, SEM cannot provide accurate assessment of phase and concentration of grain boundary phases and prevented a better understanding of the formation mechanism of such attach. This study, by pairing transmission electron microscope (TEM) characterization and atom probe tomography (APT) techniques with in-situ micro-tensile testing in scanning electron microscope (SEM), aims at gaining in-depth understanding on the formed FCCI region. The identified FCCI region roughly consists of multilayers as illustrated in Figure 1 (c). The main findings are: (1) layer-B shows observable lanthanides (Lns) infiltration along grain boundaries and mechanical softening due to FCCI- and irradiation-induced microstructural and microchemistry changes, particularly the recovery of martensitic lath structure and dissolution of pre-existing M23C6 together with the formation of coarsened Laves phases, (Fe, Cr)2(Mo, W); (2) layer-C is Fe depleted but Lns significantly enriched, becoming very brittle; (3) layer-D is mainly composed of UFe2 and Lns; (4) three FCCI-induced intermetallic U-Fe-Zr phases, ? (Fe0.5Zr0.32U0.18), e (Fe0.3Zr0.4U0.3), ? (Fe0.06Zr0.23U0.71), were identified near layer-E; (5) the ? (Fe0.5Zr0.32U0.18) phase was characterized to be a face centered cubic (FCC) crystal structure. These results will help to better understanding the governing mechanism of FCCI and facilitating the development of theoretical model for assessing the performance of metallic fuel and cladding integrity.

36 MATERIALS SCIENCE↗
DOE OSTI2025

TEM characterization of two variants of fuel cladding chemical interaction in a HT-9 Clad U-10Zr Fuel. Variant 1: FCCI with a Zr Rind

Here, this study investigated the fuel cladding chemical interaction (FCCI), a key factor that limits operational temperature and burnup, in an HT-9 clad U-10Zr nuclear fuel sample irradiated to a high burnup of 13.1 at.% at a time-averaged peak inner cladding temperature (PICT) of 530 °C. Previous results showed this fuel sample exhibited two distinct levels of FCCI at d. This paper analyzed the FCCI at an azimuthal position showing an interdiffusion layer of <10 µm using transmission electron microscopy to examine chemical and crystallographic nature of phases at the fuel-cladding interface at the nanoscale level. A ZrC layer and a Zr 3 Si phase were identified at the interface; these, along with the relatively low local temperature, potentially contributed to limit interdiffusion, behaving as inhibitors for deleterious interactions. Lanthanides (Ln) partially consumed the ZrC layer and interacted with Fe, forming a Zr-Ln compound and a (Zr,Ce)Fe 2+x phase while also infiltrating up to 4 µm into the cladding. Neither U nor Zr were observed in the cladding, whereas Fe diffused up to 3–5 µm in the fuel. Fe infiltration formed a ternary U-Zr-Fe ε-phase and likely promoted the precipitation of a Cr-rich α’ phase on the cladding interface. Additionally, a Cr-rich χ-phase, likely formed by the dissociation of pre-existing M 23 C 6 carbide precipitates, was identified about 2–5 µm from the fuel-cladding interface. Irradiation-induced nano-voids were also observed in the HT-9 bulk. These findings provide critical insights into FCCI mechanisms at representative irradiation conditions, essential for developing models simulating in-pile metallic fuel behaviors for next-generation reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Mechanical Response of Cr-Coated Zircaloy-4 Cladding to Pellet-Cladding Interactions

Chromium-coated zircaloy is a promising accident tolerant fuel (ATF) cladding design which can help mitigate high temperature oxidation. The mechanical response of the cladding during and after loading may be altered by the Cr coating throughout anticipated operational and accident occurrences such as a pellet-cladding interaction (PCI). In this work, the mechanical behavior of Cr-coated and uncoated cold worked stress-relieved (CWSR) Zircaloy-4 (Zry-4) cladding to PCI loading conditions are studied. Stress relaxation tests are performed on CWSR Zry-4 cladding using full-tube axial tension and internal pressurization to investigate mechanical anisotropy and rate-limiting deformation mechanisms of the uncoated Zry-4. These results are compared to internal pressure testing results of similar cladding with physical vapor deposition (PVD) Cr-coating to reveal how the Cr-coating affects the cladding mechanical response. Findings are discussed in the context of ATF cladding performance during operation and storage.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel

Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate the tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Coating of nuclear fuel cladding materials, method for coating nuclear fuel cladding materials

The invention provides a nuclear reactor cladding, wherein at least one layer of coating is deposited on the exterior surface of the cladding. A nuclear reactor cladding, wherein at least one layer of coating is deposited on the interior surface of the cladding. A method of coating a nuclear reactor cladding, with the steps of selecting the cladding and depositing at least one layer of a first coating on the cladding.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding

The high resistance of cladding to plastic deformation and burst failure is one of the most essential properties of accident tolerant fuel (ATF) for maintaining structural integrity during a loss-of-coolant accident (LOCA) since the deformation and burst behavior governs the cooling efficiency of flow channels and process of fission product release. To simulate and evaluate such deformation and burst process of thin-walled cladding, an in-situ testing and evaluation method has been developed on the basis of visual imaging and image analysis techniques. The method uses a specialized optics system consisted of a high-resolution video camera, light filtering unit, and monochromatic light sources, and the in-situ testing is performed using a 50 mm long pressurized thin-walled tubular specimen set in a programmable furnace. In this study eleven (11) candidate cladding materials for ATF, i.e., 6 FeCrAl alloys and 5 nanostructured steels, were tested using the newly developed method, and the time-dependent images were analyzed to produce detailed deformation and burst data such as true hoop stress, strain (creep) rate, and failure stress. Relatively soft FeCrAl alloys deformed and burst below 800°C while negligible strain rates were measured for higher strength alloys and/or for relatively thick wall specimens.

Gussev, Maxim N.↗
DOE OSTI2025

Transmission Electron Microscopy Characterization of Fuel Cladding Chemical Interaction between Minor Actinides bearing U-Pu-Zr Fuel and AIM1 Cladding

Minor actinides (MA) significantly contribute to the long-term radiotoxicity of spent nuclear fuel (SNF). Separating MA from SNF and incorporating it into metallic fuels for fast reactor transmutation is a potential method to reduce this radiotoxicity. Here, this study focuses on transmission electron microscopy characterization of two samples from the fuel cladding chemical interaction (FCCI) region of an americium (Am) and neptunium (Np)-bearing (MA-bearing) uranium-plutonium-zirconium (U-Pu-Zr) fuel irradiated in the Phenix fast reactor to 9.5 % FIMA burnup at approximately 550 °C cladding temperature. The results show that despite the complex chemical interactions between MA and AIM1 cladding elements, excessive FCCI was not induced, and Am penetration depth in the cladding limited to less than 4 µm. Np remained mostly inside fuel. The Zr-rich compounds layer effectively limited the accumulation of lanthanide on the inner cladding surface. Overall, the FCCI behavior between investigated MA-bearing U-Pu-Zr fuel and AIM1 cladding is benign.

Chemical interaction↗
NASA NTRS1973

Testing of uranium nitride fuel in T-111 cladding at 1200 K cladding temperature

Two groups of six fuel pins each were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a tantalum alloy clad. The first group of fuel pins was irradiated for 1500 hours to a maximum burnup of 0.7-atom-percent uranium. The second group of fuel pins was irradiated for about 3000 hours to a maximum burnup of 1.0-atom-percent uranium. The average clad surface temperature during irradiation of both groups of fuel pins was approximately 1200 K. The postirradiation examination revealed the following: no clad failures or fuel swelling occurred; less than 1 percent of the fission gases escaped from the fuel; and the clad of the first group of fuel pins experienced clad embrittlement whereas the second group, which had modified assembly and fabrication procedures to minimize contamination, had a ductile clad after irradiation.

Rohal, R. G.↗
DOE OSTI2021

Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding

As an alternative fuel form, the annular metallic fuel design eliminates 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. In this work, 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. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr 2 phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal ( U ,Zr) 6 Fe phase (space group I4/mcm) and cubic ( U ,Zr)( Fe ,Cr) 2 phase (space group Fd 3 ¯ m). The ( U ,Zr)( Fe ,Cr) 2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm 3 ¯ m). At the interdiffusion zone and inner cladding interface, 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 ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4 2 /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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗