Abstract

Sodium-cooled Fast Reactors (SFR) are one of the advanced nuclear reactor concepts to be commercially applied for electricity production. The benefits of SFR are well-known and include: the possibility of a closed fuel cycle, proliferation resistance, nuclear waste minimization via actinides burning, and fissile breeding capabilities. Metallic fuel used in SFR has well demonstrated irradiation performance. However, more studies are necessary to optimize and extend operational and safety limits for their commercialization and licensing. This could be achieved through a better understanding of fuel behaviors during transient and of fuel failure thresholds. This paper describes the experimental Research and Development (R&D) program aimed at providing the necessary data to support the development of SFR-optimized safety limits. This program integrates separate effects testing and integral effects testing, combined with advanced Modeling and Simulation (M&S). This R&D program, finally, focuses on delivering the science-based information necessary for supporting the licensing and utilization of SFR based on metallic fuel. In this paper we will describe the three research areas centered on fuel development and focused on separate effect testing, namely: (1) microstructural, chemistry, and material properties; (2) thermo-mechanical behavior; and (3) source term and fission product behavior. Preliminary results from these Separate Effect Tests (SET) studies and the current instruments and experimental plan are also presented.

1 Introduction

Development of new nuclear reactor concept respond to need of producing electricity via clean, safe, and reliable technology. Sodium cooled Fast Reactors (SFR) based on metallic fuel are an important candidate for advanced reactor deployment. Indeed, the irradiation performance of metallic fuel in SFR has been well demonstrated during both normal and off-normal operations [1]. Moreover, metallic fuel is considered a mature technology [2]. This fuel provides higher reactor performance to any other fuel form in fast reactors and displays high proliferation resistance characteristics [3]. Moreover, SFRs provide enhanced safety thanks to inherent favorable neutronic feedback in response to accident initiators and passive safety systems [4]. Finally, metal fuel offers the possibility of a close nuclear fuel cycle as it can transmute actinides, thus allowing for a lower inventory of long-lived waste products to be stored in a repository and can breed fissionable isotope creating a virtually unlimited supply of nuclear fuel [5]. Developing and demonstrating this advanced fuel form is thus in line with DOE-NE (Department of Energy-Nuclear Energy) missions of: (1) enhancing the performance and safety of the nation's current and future reactors; (2) enhancing proliferation resistance of nuclear fuel; (3) effectively utilizing nuclear energy resources; (4) and addressing the longer-term waste management challenges. Thus, large interest is currently posed in the licensing and development of metal fueled SFR in the U.S.

The reestablishment of a fuel safety research and development program for metallic fast reactor fuels is essential to develop data and models to reduce uncertainty factors in transient fuel behavior and fuel failure thresholds for SFR, thus providing expanded performance envelopes desirable for commercial applications and to support their licensing. DOE-NE has begun research to build on the historical legacy of metallic fuel performance that will enhance technology marketability. The proposed studies aim at addressing current knowledge gaps, leveraging on new advanced characterization techniques, which were not available since the last metal fuel safety test was performed (approximately 30 years ago) and utilizing the reinstate transient test capabilities (e.g., Transient Reactor Testing Facility-TREAT) at Idaho National Laboratory (INL) [6]. The revitalized TREAT facility will provide an important opportunity to further test the safety limits of SFR technology, as relevant in-pile safety and severe accident facilities are sparse in the modern international community, as reported by the Organization for Economic Co-operation and Development, Nuclear Energy Agency (OECD NEA) assessment [7]. The described Research & Development (R&D) program will couple in-pile and out-of-pile studies including approaches described as Separate Effect Tests (SET) and Integral Effect Tests (IET) in each. On one hand, SET provide unique opportunity for developing a deep understanding of the underlying mechanisms and physical processes, by separating the effect of one variable at time on the phenomena of interest. This is performed by designing tightly controlled boundary conditions in the experiments and focused measurement to reduce uncertainty in the results [2]. SET thus provide a good avenue for developing, informing, and validating Modeling & Simulation (M&S) tools (e.g., BISON and MARMOT). Complementary to SET are IET, as show in Fig. 1. These latter rely on large experiments and aim at technology demonstrations. IET serves two major purposes: (1) early phenomena identification experiments coupled with M&S predictions that elucidate fuel behaviors and failure mechanisms and can provide the basis for detailed testing strategies [9] and as mentioned (2) select confirmatory or qualification testing that will provide ultimate validation of M&S tools and support the regulatory process [2]. The described (SET-IET)- coupled approach together with recent advanced modeling and innovative experimental capabilities at INL, described in the following paragraphs, will permit to investigate the effects of a wide range of transients, and conditions of importance to further the deployment of this technology (shown in Fig. 2).

Fig. 1
Experimental approach used in this R&D program, which couples IET and SET and showing its characteristic. Figure is adapted from Ref. [8].
Fig. 1
Experimental approach used in this R&D program, which couples IET and SET and showing its characteristic. Figure is adapted from Ref. [8].
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Fig. 2
Transient regimes (in the time versus temperature domain) and key capabilities available at INL to investigate nuclear fuel behaviors. Failure threshold is also shown. Figure is adapted from Ref. [9]. In the figure SS stands for Steady-State irradiation, and VTR for the Versatile Test Reactor.
Fig. 2
Transient regimes (in the time versus temperature domain) and key capabilities available at INL to investigate nuclear fuel behaviors. Failure threshold is also shown. Figure is adapted from Ref. [9]. In the figure SS stands for Steady-State irradiation, and VTR for the Versatile Test Reactor.
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This paper will focus on SET part of the described R&D program for SFR safety, highlighting the current gaps needing investigation. The available capabilities, their development, and application for safety studies will also be described. Finally, preliminary results from these investigations will be presented.

1.1 Separate Effect Testing.

Three primary focus areas have been identified in the proposed R&D program. These are based on SET experiments and support fuel safety studies necessary for the licensing and commercialization of SFR. These areas can be summarized as following: (1) microstructure, chemistry, and material properties; (2) thermo-mechanical behavior; and (3) source term determination, as presented in Fig. 3. It should be highlighted that although separate experimental plans are being developed for each focus area, the phenomena which take place in reactor are cross-related to all areas (e.g., local material properties influence Fission Products (FPs) transport, and failure mechanism and accidents timeline affect FPs retention and release), thus attention is posed in strong collaboration in between the different areas and in cross-cutting R&D that cover multiple focal points [6].

Fig. 3
SET focus area for safety studies on SFR fuel, reproduced from Ref. [5]
Fig. 3
SET focus area for safety studies on SFR fuel, reproduced from Ref. [5]
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These SET focus areas are following described in better detail:

  1. Microstructural, chemistry, and material properties. This category encompasses detailed understanding of the impacts of irradiated fuel microstructure on transient performance and the evaluation of its evolution during off-normal event [2]. This focus area is deeply interlinked with the other two and is of especially relevant to mechanistic mesoscale modeling (e.g., MARMOT). SET in this area include the evaluation via advanced characterization of the chemical form of fission products and of the metallic matrix element, such as phase analyses. Microstructure evolution of the fuel such as porosity analysis, grain size, and texture, moreover, are of fundamental importance in this focus area. These parameters influence the physical properties of the fuel and thus its safe response in reactor during irradiation and transients' conditions. Great knowledge on this topic has been gained from Post Irradiation Examinations (PIE) based on fuel used in normal operation conditions (steady-state conditions) [10,11] while there is a gap on transient behavior analyses.

  2. Thermo-mechanical behavior is of concern to fuel and reactor designers, regulators, and utilities [2], as these properties strongly affect the response of fuel during a transient event. While IET provide the quantitative evaluation of these phenomena in reactor, the proposed SET experiments can offer important insights on the influence of different variable on the desired properties. The available advanced instrumentation at INL can provide local spatial evaluation of thermal and mechanical properties, linking such parameters to microstructural characteristics and local irradiation effects. These studies are fundamental because during irradiation, these properties vary and thus the characterization obtained from unirradiated fuel is insufficient. Thus, PIE and in situ evaluation of these properties is of essential importance in the proposed SET. Literature on thermo-mechanical properties of fresh metallic fuel is available [12,13]. However limited data are available for irradiated fuel [12].

  3. Source term determination (e.g., the evaluation of the types, quantities, and chemical forms of the radionuclides release) is necessary for the licensing process because it determines the response plan and the extent of the contamination area after an unwanted release event. While many studies in the past have focused on fission product release from irradiated oxide fuels (e.g., Refs. [14] and [15]), including large-scale integral tests [16,17], which lead to a general understanding of the release behavior during accident conditions for light water reactors [18]. Limited knowledge exists on FP release from metallic fuel for SFR [5,10]. A detailed mechanistic understanding of FP behavior and transport in metallic fuels during off-normal conditions (e.g., very high temperature) is lacking [1]. The availability at INL of legacy irradiated materials of high relevancy is a first step providing the opportunity to understand long-lived FP distribution and their response to transient conditions by utilizing new advanced characterization facilities [2], described in the following. While further IET studies will require new irradiated and safety tested (under transient) fuel, which could be obtained using the TREAT facility.

2 Experimental Plan and Facilities

2.1 Advanced Characterization for Microstructure, Chemical Analyses, and Mechanical Properties.

Advanced characterization instrumentation is available or actively being installed in the Irradiated Materials Characterization Laboratory (IMCL) and include scanning electron microscopy (SEM), transmission electron microscopy (TEM), focused ion beam (FIB), electron probe micro-analysis, X-ray diffractometry (XRD), atom probe tomography (APT) and microcomputed tomography (μ-CT). These instruments are equipped with state-of-the-art detectors ranging from energy dispersive X-ray spectroscopy (EDS), wavelength dispersive X-ray spectroscopy (WDX), electron backscatter diffractometry. These instruments are optimized to work with irradiated materials [19] and thus are a unique opportunity to evaluate elemental composition and microstructure evolution after irradiation and in-pile transient tests. Such tools enable detailed understanding of key fuel performance phenomena such as Fuel Cladding Chemical interaction (FCCI), phase evolution, and fission product chemistry transport. Moreover in situ capabilities are being developed including micro/nanomechanical accessory and heating stages, which will provide the opportunity to directly evaluate and correlate microstructural and material properties evolution in real-time.

First small-scale experiments have been conducted investigating microstructural evolution and FP chemical form in metallic fuel under transient and leveraging the described improved capabilities. These tests examined the effect of different transients on important fuel performance phenomena (such as FCCI, phase transformation, and grain behavior) and provide important insight to the modelers (such as fuel relocation and thermal expansion). These SET range from the investigation of microstructural evolution in transient single-phase specimens at low burn up, up to the characterization of the microstructure and FP behavior in a metallic fuel pin at high burn up which underwent an in-pile transient. The results of these studies are presented in detail in other publications [20,21]. The transients applied in these studies show to influence the microstructure of the specimens, however they did not strongly affect the safety performance of the metallic fuel. Further investigations are planned to analyze both fresh fuel and irradiated fuel microstructure evolution under different ramps covering a wide range of transients, and thus spanning over a wide range of techniques: such as in situ SEM/TEM heating, furnace testing, and in-pile reactor transients.

2.2 Out-of-Pile Steady State and Transient Furnaces Capabilities Development.

Advanced furnace instrumentations have been evaluated to conduct out-of-pile steady-state and transient tests on previously irradiated metallic nuclear fuel pins (or other SFR advanced fuel forms). These tests are crucial to complement the existing capabilities at the TREAT facility for short duration transients to longer duration, such as steady-state performance at the Advanced Test Reactor (ATR). As shown previously in Fig. 2, furnace testing is an essential capability for evaluating temperature performance over a continuum of time scales relevant to fuel performance. Historically, the Fuel Behavior Test Apparatus and Whole Pin Furnace previously located at the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory (ANL) performed similar tests and provided crucial transient data for defining fuel design limits [2225].

These tests will be used to extend the transient performance database, thus there should be some overlap in time versus temperature regime of the in-pile facilities. Such furnace studies are focused on characterizing and understand isolated fuel performance phenomena (e.g., SET) such as cladding overpressure to failure, fuel-cladding chemical interaction (FCCI), fuel-cladding mechanical interaction, and their mechanisms and thresholds. These tests will also provide valuable data for validation and improvement of M&S tools. Recent capability assessments have found this as a gap and efforts are beginning to reestablish the capability to test full length-irradiated pins (such as Experimental Breeder Reactor II-EBR-II-pins) or pieces of irradiated fuel pins (e.g., coming from EBR-II, Fast Flux Test Facility-FFTF, or the Advanced Test Reactor-ATR). These will provide an avenue complementary to IET in TREAT (Fig. 2) to evaluate failure thresholds in different transient regime. These studies are currently in the planning phase together with the development of the experimental facilities and will thus not presented in detail in this paper.

2.3 Thermal Properties Cell Instrumentation and Fission Products Release Studies.

New instruments to evaluate thermal properties of irradiated materials, with a various range of spatial resolution (from the mm to μm scale) are available at IMCL (e.g., thermoreflectance methods via a thermal conductivity microscope, and a pulse laser flash analysis). Moreover, a Mass Spectrometry (MS) system connected to a Thermogravimetric Analyzer/Differential Scanning Calorimeter (TGA/DSC) system is being installed. This could provide an important avenue for SET to investigate FP release behaviors at a local level. This system provides the possibility to analyze gaseous species and volatized materials released by temperature (up to 2273 K and under ramps up to 100 K/min), while measuring concurrently sample weight loss or gain. A schematic of the system is shown in Fig. 4. These techniques can provide thermal properties measurement and FPs behavior and release information and can be correlated to the microstructure evolution when combined with the PIE techniques mentioned in the previous paragraph. Such a system can thus enlighten on the response of materials and FPs during transient, providing complementary information to integral release data, such as the one described in Ref. [26]. The SET R&D program will try to integrate different length scale testing from the nano to the macroscale, such as microtesting of FP release from fuel fragments by TGA-DSC/MS together with SET TEM in situ studies to evaluate fission product behavior and release. These studies are indeed complementary to IET focused on transient testing in TREAT and will thus provide information on mechanism leading to the observed integral release. These studies are currently in the planning phase together with the development of the experimental facilities and thus will not be presented in detail in this paper.

Fig. 4
Sketch of the TGA-DSC/MS system from Netzsch (model STA 409 CD). The system is installed in the Thermal Properties Cell (TPC) in IMCL, a gloves box qualified to work with irradiated materials.
Fig. 4
Sketch of the TGA-DSC/MS system from Netzsch (model STA 409 CD). The system is installed in the Thermal Properties Cell (TPC) in IMCL, a gloves box qualified to work with irradiated materials.
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Currently, basic studies on fission gases and volatile FP behavior are being conducted. First experiments have focused on analyzing fission gas bubbles evolution in metallic fuel. Such analyses focus on the phenomenological understanding of fission gases behavior in metallic fuel and can thus support M&S of these phenomena important for source term determination and thus licensing. This paper will focus on presenting the preliminary results of this one-of-a-kind SET study, not presented elsewhere.

3 Fission Gases Set Results

These experiments monitored fission gas bubbles formation and their evolution via in situ techniques. Xenon (Xe) was used for these studies as one of the most abundant fission gases produced in reactor. Fission gases have a low solubility in the nuclear fuel and can either agglomerate into bubbles or be released from the fuel matrix into the cladding [25]. The first SET experiments were performed with the aim of evaluating separately the contributions of two variables (irradiation and temperature) on gas bubbles evolution in fuel matrix. The presented study was executed under NSUF-RTE (Nuclear Science User Facilities-Rapid Turnaround Experiments) program at the ANL Intermediated Voltage Electron Microscope (IVEM)—Tandem Facility [27,28]. The influence of temperature and radiation on fission gases (Xe) bubble distribution and dimension was investigated through in situ experiments at IVEM and by PIE via TEM at IMCL. This data could provide insight for mechanistic microstructural modeling of fission gas behavior in metallic fuel, as the data was acquired in well-defined systems.

3.1 Samples and Instruments.

Two specimens (α-U and α/δ U-10Zr) were investigated in this study to evaluate Xe bubble behavior and radiation damage in different phases present in metallic fuel (α and δ) [29]. Samples for these tests were produced in house via arc melting. Disks specimens of 3 × 10−3 m in diameter were punched from the specimen and electropolished to perforation with a solution of 7% perchloric acid in methanol. The electropolishing parameters for the two materials are given in Table 1. The in situ irradiation experiments were performed at IVEM Tandem Facility, which consists of a microscope (Hitachi 9000 NAR operated a 300 × 103 V) interfaced to a 500 × 103 V ion accelerator made by national electrostatics corporation. The microscope has a side entry stage and an expanded objective pole piece to accommodate the beam line [28]. Data was recorded by the Gatan OneView camera. Postirradiation TEM analyses were conducted at IMCL in Idaho National Laboratory using a FEI (Field Electron and Ion Company) Talos F200X field FEG (Field Emission Gun) STEM (Scanning Transmission Electron Microscope) and a FEI Titan Themis 200 STEM, both equipped with ChemiSTEM technology and operated at 200 × 103 V. EDS data were collected under STEM mode and STEM-EDS data analysis was done using the Thermo Fisher Scientific VeloxTM software and the Bruker ESPRIT (estimation of signal parameters via rotational invariant techniques) 2 software.

Table 1

Parameters used in the electropolishing process. The solution used was 7% Perchloric Acid in Methanol.

Sample typeAlpha uraniumUranium-10Zirconium
Voltage (V)11.515
Flow3025
Time (s)3030
Temperature (°C)−15−17
Sample typeAlpha uraniumUranium-10Zirconium
Voltage (V)11.515
Flow3025
Time (s)3030
Temperature (°C)−15−17

The experiments consisted of 300 × 103 eV (where 1 eV correspond to the energy gained by an electron when the electrical potential is increased by 1 V, 1.602 × 10−19 J) Xe ion implantation at room temperature (RT) followed by annealing with or without concurrent 1 × 106 eV Krypton (Kr) ion irradiation to evaluate gas bubble behavior. Irradiation with Kr was performed to evaluate separately the contributions of irradiation on gas bubbles evolution from temperature. The irradiation and annealing conditions are given in Table 2. Xe was chosen as a prototypical gaseous FP, while Kr was used to introduce damage in the material via the highest energy ions available (1 × 106 eV). A Gatan 652, double tilting heating holder (with capability to reach up to 1273 K) was used for the in situ irradiation and postirradiation annealing. Heating temperatures (523–723 K) were chosen to ensure the stability of alpha phase and to minimize oxidation, and as high as possible to permit bubble motion and coalescence. Similar experiments have been conducted for UO2 and U-10Zr fuel but using furnace annealing [3032]. While in situ approach is reported in Refs. [29,33], and [34] to monitor gas bubble formation/evolution and phase transformation.

Table 2

Experimental matrix of the in situ tests conducted at IVEM. RT indicates room temperature. Kr irradiation was performed at 1 × 106 eV, and Xe irradiation was performed at 300 × 103 eV. Where 1 eV correspond to the energy gained by an electron when the electrical potential is increased by 1 V, 1.602 × 10−19 J.

Implantation conditionsIn situ heatingXenon fluence
Sample nameIon/temperaturetemperature (K)Irradiation while heatingSET variable investigatedIon/m2
Alpha UXe/RT523NoTemperature effect1 × 1020
Alpha U-KrXe/RT523KrRadiation effect1 × 1020
U-10ZrXe/RT523NoTemperature effect5 × 1019
U-10Zr-2Xe/RT773NoTemperature effect5 × 1019
U-10Zr-KrXe/RT523KrRadiation effect5 × 1019
Implantation conditionsIn situ heatingXenon fluence
Sample nameIon/temperaturetemperature (K)Irradiation while heatingSET variable investigatedIon/m2
Alpha UXe/RT523NoTemperature effect1 × 1020
Alpha U-KrXe/RT523KrRadiation effect1 × 1020
U-10ZrXe/RT523NoTemperature effect5 × 1019
U-10Zr-2Xe/RT773NoTemperature effect5 × 1019
U-10Zr-KrXe/RT523KrRadiation effect5 × 1019

In this study to quantify bubble evolution, an image analyses method was developed, which is described in the following paragraph. This was necessary due to the large number of bubbles formed during irradiation that impeded manual operator quantification. Moreover, analyses of the samples before irradiation were performed to identify morphology and crystal structure and compare to its evolution after irradiation.

The study was performed with the aim of evaluating separately the contributions of these two phenomena to gas bubbles evolution of importance to the M&S of such phenomena.

3.2 Image analyses Method.

To identify and quantify bubbles within the fuel microstructure, an image analyses method was created. This compared the under and over focus image as a mean to identify bubbles in the samples from other artifacts. The method is based on Matlab 2019b image processing, computer vision, and statistical toolboxes. Four general steps were required in these analyses: (1) registration, (2) creation of a difference image, (3) identification of regions of interest for analysis, and (4) application of analysis routines. As described in the following and summarized in Fig. 5.

Fig. 5
Schematic of step processes applied in the image analyses method and example of results obtained
Fig. 5
Schematic of step processes applied in the image analyses method and example of results obtained
Close modal
  1. As in general, there was a small spatial variation between corresponding over- and under-focused micrographs, a Speeded-Up Robust Features blob detection algorithm was used to identify precipitate features for alignment of the micrographs. This utilized the built in Matlab function “detectSURFFeatures.” Matching features within the two micrographs were determined utilizing Matlab's “extractFeatures” algorithm. The match threshold was set to 9.5% to significantly increase the strength/quality of corresponding matches between micrographs and a max ratio of 0.095% to help exclude small ambiguous (noise) features from the matching process. Based on the resulting matches, the position information for matching pairs in the respective micrographs were used to map the over-focused micrograph to the under-focused image by estimating and implementing a geometric transform. An affine transform was used. Both translational and rotational manipulation was allowed.

  2. Once a version of the over-focused image corresponding to the spatial coordinates of under-focused was determined, the two micrographs could be directly compared. Both micrographs were filtered using a fine 3 × 3 kernel wiener filter to remove noise from the micrographs without significantly degrading features such as edges. Histogram matching was also applied to roughly align the gray scale values of features within the two micrographs. This step helped assure that differences observed between images were truly related to fission gas bubbles seen as dark and light bubbles in the over- and under-focused micrographs, respectively. A difference image was produced by direct pixel by pixel subtraction of the over-focused micrograph from the under-focused micrograph. This difference image was the basis for quantitative analysis. Small, erroneous, features were removed from the difference image using morphological opening with a square 3 × 3 structure element.

  3. The analyses of interest included an estimate of the average fission gas bubble diameter and the corresponding number density of fission gas bubbles observed. The precipitate phase could however significantly skew either result as the difference image information in these regions were significantly lower quality. To avoid this, a mask was created to remove regions with a precipitate from the analysis. The mask was created by first morphologically closing the filtered under-focused micrograph with a circular structure element having a radius of 20 pixels. This was followed by application of a threshold determined via Otsu's method. Otsu's method uses the objective criterion of maximized intensity variance to select threshold value. The resulting threshold typically corresponded to a value of 22,500 for an unsigned 16-bit gray scale micrograph or nominally 34% of the entire gray scale range. Compared to direct application of a threshold, this approach also masked regions surrounded by precipitates leaving only regions that were further distanced from precipitates for analysis.

  4. To obtain an average fission gas bubble size and the observed distribution, a simple morphological granulometry technique was used. One significant advantage of this methodology is size distribution can be estimated without explicitly detecting (segmenting) each individual object. Thus, bubbles that appear to overlap in the transmission image are not necessarily counted as a single bubble. Granulometry estimates object sizes in many ways similar to powder sieving with successively smaller sieve sizes. In granulometry, a structure element (a disk of radius R for this work), is used to morphologically open the difference image. When a bubble is morphologically opened by a disk smaller than the bubble, it remains relatively unchanged. When a bubble is morphologically opened by a disk larger than the bubble, the bubble is removed, and the corresponding bubble pixels take on the value of the surrounding background signal. By recording the total image intensity for successively larger structure element sizes, information can be obtained regarding the size of objects. Plotting the 1st derivative of image intensity against structure element radius gives an estimated bubble size distribution. To approximate a number density of bubbles, the masked difference image was segmented, again using Otsu's method. Upon segmentation, a connective components algorithm was used to group pixels into corresponding objects. The number of objects was then counted and normalized to the unmasked area of the image

3.3 Results and Discussion.

The characterization of the sample before the analyses showed that α-uranium samples present large grains of about 10 × 10−3 m (Fig. 6), while the U-10Zr showed the typical α/δ lamellar structure, as the one presented in Ref. [29] (Fig. 7). After irradiation, grain restructuring may have occurred in the U-10Zr sample as observed by the growth and change in the lamellar structure (Fig. 7). Gas bubbles formation was observed already at room temperature because of the high ion fluence. For the alpha uranium samples, Xe bubbles could be observed already at 5 × 1019 ion/m2, while for U-10Zr they seem to appear already at 3 × 1019 ion/m2. All bubbles had a small dimension under 1 × 10−9 m, due to the implantation occurring at room temperature, limiting thus coalescence and growth. The bubbles seemed to be distributed mostly intergranular (Figs. 8 and 9). During the annealing process, the bubbles were observed to grow (Table 3, Figs. 8 and 9). However, they did not seem to migrate to the boundaries, this may be related to the low temperature range and short time of annealing (2 h). Moreover, in the U-10Zr samples the bubble seems to be larger indicating possibly a higher mobility, even with lower irradiation doses (Table 3). The higher mobility of bubbles in the α/δ lamellar structure may be motivated by increased diffusion due to high density of grain boundaries or higher local defects concentration facilitating motion. In both samples (Table 3, Figs. 8 and 9), bubbles were observed to grow under both annealing and irradiation. From the data it may be inferred that the temperature seems to have a stronger influence than irradiation in the tested conditions. When 723 K was used, the growth of the gas bubble was significant, indicating diffusion and coalescence. This may indicate that defect creation rate by Kr irradiation was not high enough during the annealing/irradiation step. Defect analysis was difficult due to the complex microstructure and high presence of bubble.

Fig. 6
TEM analyses on alpha uranium before irradiation (top), and after irradiation (bottom) under focused image showing the presence of gas bubbles (sample shown is Alpha-U from Table 2)
Fig. 6
TEM analyses on alpha uranium before irradiation (top), and after irradiation (bottom) under focused image showing the presence of gas bubbles (sample shown is Alpha-U from Table 2)
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Fig. 7
TEM analyses of U-10Zr before irradiation (top), and after irradiation (bottom) under focused image showing the presence of gas bubbles (sample shown is U-10Zr from Table 2)
Fig. 7
TEM analyses of U-10Zr before irradiation (top), and after irradiation (bottom) under focused image showing the presence of gas bubbles (sample shown is U-10Zr from Table 2)
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Fig. 8
TEM analyses of alpha U showing gas bubble evolution before and after annealing
Fig. 8
TEM analyses of alpha U showing gas bubble evolution before and after annealing
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Fig. 9
TEM analyses of U-10Zr showing gas bubble evolution before and after annealing in the different tests
Fig. 9
TEM analyses of U-10Zr showing gas bubble evolution before and after annealing in the different tests
Close modal
Table 3

Results obtained from the image analyses, showing gas bubble size and density. In the table # stand for number of gas bubble, is the uncertainty based on 2 times the standard deviation, and N/A for not available when only one measurement was available for the analyses

Size (Angstrom = 1 × 10−10 m)Density (#/nm2)
SamplesAverageAverage
Alpha-Uranium
Starting conditions Xe 1 × 1020 ion/m25.920.200.480.26
Alpha U8.71.40.240.04
Alpha U-Kr9.24.40.250.26
Uranium-Zirconium
Starting conditions Xe 5 × 1019 ion/m26.41.10.500.32
U-10Zr7.08N/A0.43N/A
U-10Zr-28.32.00.320.20
U-10Zr-Kr7.31N/A0.43N/A
Size (Angstrom = 1 × 10−10 m)Density (#/nm2)
SamplesAverageAverage
Alpha-Uranium
Starting conditions Xe 1 × 1020 ion/m25.920.200.480.26
Alpha U8.71.40.240.04
Alpha U-Kr9.24.40.250.26
Uranium-Zirconium
Starting conditions Xe 5 × 1019 ion/m26.41.10.500.32
U-10Zr7.08N/A0.43N/A
U-10Zr-28.32.00.320.20
U-10Zr-Kr7.31N/A0.43N/A

The only other study which applied (He) ion implantation on similar samples [33] observed the formation of larger mobile bubbles (up to 70 × 10−9 m in the α phase) and that the δ phase was identified to be more irradiation resistance than α, in contrast with the presented observation. It must be noted that the bubbles analyzed in the study [33] were formed under higher fluences over 1 × 1020 ion/m2 (up to 5 × 1020 ion/m2), and the sample preparation was different. Samples irradiated in Ref. [33] were in the form of foil, which were following punched and electropolished after the irradiation to perform the desired TEM characterization. Further analyses are necessary to identify the depth of diffusion of Xe in the different phases, such as SIMS (Secondary Ion Mass Spectrometer) measurements described in Ref. [30], which will permit a quantification of the diffusion phenomena. These data are, indeed, relevant to source term determination. A FIB-nano-SIMS setup is currently under installation at IMCL, which will provide an important avenue for such analyses. Such data could be used to determine gas diffusion coefficients in fuel matrix. Moreover, further experiment with lower fluence and implantation at higher temperature, together with defect analyses is suggested to better understand the phenomena taking place in these tests. The lamellar (α/δ) structure showed to be very difficult to analyzed, thus the preparation of one phase material is suggested for further similar studies [33].

4 Conclusions and Summary

This paper describes a new R&D program coupling SET and IET and aimed at supporting the development of SFR-optimized safety limits, in view of future licensing need. While IET provide the quantitative evaluation of the safety limit, SET studies advance the phenomenological understanding of the variables influencing fuel performance. Thus, the basic science insight collected from the SETs are relevant to support the development and validation of modeling tools to describe transient metallic fuel performance.

A variety of experiments and unique examinations capabilities have been presented in this paper, together with R&D plan. The R&D program can be divided in three research areas: (1) microstructural, chemistry, and material properties; (2) Thermo-mechanical behavior; and (3) Source term and fission product behavior. Preliminary results from SET studies aimed at analyzing fission products transport in metallic fuel were presented in this paper. Although this study is at an initial stage and further examinations are necessary to evaluate the effective depth and diffusion of gaseous bubbles. It can be concluded that the methodology can be used to understand gas behavior in fuel matrix. We observed, moreover, that the moderate temperatures and limited time applied in these studies did not strongly influence gas bubble evolution. Further studies are necessary in this area with one phase material to quantify the contribution of each phase and with carefully prepare sample permitting the evaluation of defects influence on gas diffusion, also a lower fluence. Results from the transient studies can be found in Refs. [20] and [21] these showed that metallic fuel offers a safe performance also under moderate transients. More studies are under the planning phase, as described in the focus areas, and include release studies via TGA/DSC-MS, SET on microstructure evolution on more complex systems and evaluation of material properties evolution after irradiation and transients.

The presented SET studies are finally an important avenue to collect relevant data, which will help to understand the behavior of metallic fuel in integral tests in the Transient REactor Test facility (TREAT) and can provide insight for modeling of SFR fuel.

Acknowledgment

Dr. T. Yao, Dr. B.D. Miller, Dr. D. Murray, Mrs. K. Wright, and Dr. F. Teng are thanked for their scientific and technical support. All INL and ANL support staff for sample preparation, coordination, and scientific insight. B. Frickey and C. Anderson are acknowledged for the extensive support at the met box. Mr. P. Baldo from ANL is acknowledged for technical insight and support at the IVEM facility. Dr. J. Harp and Mr. S. Ashby are thanked for sample fabrication.

This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

Funding Data

  • This work was supported by the U.S. Department of Nuclear Energy under DOE-NE Idaho Operations Office Contract DE-AC07-05ID14517. Part of this work was supported by the Nuclear Science User Facility (NSUF) under the Rapid Turnaround Experiment (RTE) FY19 1621. This paper was authored by a contractor for the U.S. Government.

Nomenclature

     
  • AFC =

    Advance Fuel Campaign

  •  
  • ANL =

    Argonne National Laboratory

  •  
  • APT =

    Atom Probe Tomography

  •  
  • ATR =

    Advanced Test Reactor

  •  
  • DOE =

    Department of Energy

  •  
  • DSC =

    differential scanning calorimeter

  •  
  • EBR-II =

    Experimental Breeder Reactor II

  •  
  • EDS =

    Electron Dispersive X-ray Spectroscopy

  •  
  • ESPRIT =

    estimation of signal parameters via rotational invariant techniques

  •  
  • eV =

    electron volt

  •  
  • FCCI =

    fuel cladding chemical interaction

  •  
  • FEG =

    field emission gun

  •  
  • FEI =

    Field Electron and Ion company

  •  
  • FFTF =

    Fast Flux Test Facility

  •  
  • FIB =

    focused ion beam

  •  
  • FIMA =

    fission for initial metal atom

  •  
  • FP =

    fission products

  •  
  • IET =

    integral experiment testing

  •  
  • IMCL =

    Irradiated Material Characterization Laboratory

  •  
  • INL =

    Idaho National Laboratory

  •  
  • IVEM =

    Intermediate Voltage Electron Microscope

  •  
  • M&S =

    modeling and simulation

  •  
  • MS =

    mass spectroscopy/spectrometry

  •  
  • N/A =

    not available

  •  
  • NE =

    nuclear energy

  •  
  • NEA =

    Nuclear Energy Agency

  •  
  • NSUF =

    Nuclear Science User Facility

  •  
  • OECD =

    Organization for Economic Co-operation and Development

  •  
  • PIE =

    post irradiation examination

  •  
  • R&D =

    Research and development

  •  
  • RT =

    room temperature

  •  
  • RTE =

    rapid turnaround experiments

  •  
  • SET =

    separate effect testing

  •  
  • SEM =

    scanning electron microscopy

  •  
  • SFR =

    sodium fast reactor

  •  
  • SIMS =

    secondary ion mass spectrometer/spectrometry

  •  
  • STEM =

    scanning transmission electron microscopy

  •  
  • SS =

    steady-state

  •  
  • TEM =

    transmission electron microscopy

  •  
  • TGA =

    thermal gravimeter analyzer

  •  
  • TPC =

    thermal properties cell

  •  
  • TREAT =

    Transient Reactor Test facility

  •  
  • U.S. =

    United States of America

  •  
  • VTR =

    versatile test reactor

  •  
  • WDS =

    wavelength dispersive spectroscopy

  •  
  • XRD =

    X-ray diffractometry

  •  
  • μ-CT =

    microcomputer tomography

  •  
  • σ =

    standard deviation

  •  
  • # =

    number of gas bubble

References

1.
Carmack
,
W. J.
,
Porter
,
D. L.
,
Chang
,
Y. I.
,
Hayes
,
S. L.
,
Meyer
,
M. K.
,
Burkes
,
D. E.
,
Lee
,
S. L.
,
Mizuno
,
T.
,
Delage
,
F.
, and
Somers
,
J.
,
2009
, “
Metallic Fuels for Advanced Reactors
,”
J. Nucl. Mater.
,
392
(
2
), pp.
139
150
.10.1016/j.jnucmat.2009.03.007
2.
Jensen
,
C.
, and
Wachs
,
D.
,
2018
, “
Fuel Safety Research Plan for Metallic Fast Reactor Fuels
,” Idaho National Laboratory, Idaho Falls, ID, Technical, Report No. INL/INT-18-51467.
3.
Ogata
,
T.
,
2012
, “
Chapter 3.01 Metal Fuel
,”
Comprehensive Nuclear Material
(Elsevier Science),
R. J. M.
Konings
, ed.,
Elsevier
, Amsterdam, The Netherlands, pp.
1
40
.
4.
Sofu
,
T.
,
2015
, “
A Review of Inherent Safety Characteristics of Metal Alloy Sodium-Cooled Fast Reactor Fuel Against Postulated Accidents
,”
J. Nucl. Eng. Technol.
,
47
(
3
), pp.
227
239
.10.1016/j.net.2015.03.004
5.
Middleton
,
B. D.
,
Parma
,
E. J.
,
Olivier
,
T. J.
,
Phillips
,
J.
, and
LaChance
,
J. L.
,
2011
, “
The Development of a Realistic Source Term for Sodium-Cooled Fast Reactors: Assessment of Current Status and Future Needs
,” Sandia National Laboratory, Albuquerque, NM, Report No. SAND2011-3404.
6.
Di Lemma
,
F. G.
,
Jensen
,
C.
, and
Wachs
,
D.
,
2019
, “
Fuel Safety Research Plan for Metallic Fast Reactor Fuel R&D Focus Area: Fission Product Transport for Source Term Determination
,” Idaho National Laboratory, Idaho Falls, ID, Report No. INL/INT-19-53083.
7.
Task Group on Advanced Reactors Experimental Facilities (TAREF)
,
2011
, “
Experimental Facilities for Sodium Fast Reactor Safety Studies
,” OECD NEA, Issy-les-Moulineaux, France, Report No. NEA/CSNI/R(2010)12.
8.
Rider
,
W.
,
Mousseau
,
V.
,
Williams
,
B.
,
Adams
,
B.
,
Smith
,
R.
,
Hooper
,
R.
,
Belcourt
,
N.
,
Ougouag
,
A.
,
Abdel-Khalik
,
H.
,
Copps
,
K.
, and
Sieger
,
M.
,
2014
, “
Validation and Uncertainty Quantification (VUQ) Strategy, Revision 1
,” Report No. CASL-U-2014-0042-001.
9.
Bauer
,
T. H.
,
Kramer
,
J. M.
,
Tilbrook
,
R. W.
, and
Goldman
,
A. J.
,
1989
, “
A Program to Resolve the Safety Implications of Fuel Damage in the Operation of Advanced Metal-Fueled Reactors
,” Argonne National Laboratory, Argonne National Laboratory, Argonne, IL, Report No. ANL-IFR-103.
10.
Lee
,
C. B.
,
Kim
,
D. H.
, and
Jung
,
Y. H.
,
2001
, “
Fission Gas Release and Swelling Model of Metallic Fast Reactor Fuel
,”
J. Nucl. Mater.
,
288
(
1
), pp.
29
42
.10.1016/S0022-3115(00)00718-2
11.
Grabaskas
,
D.
,
Bucknor
,
M.
, and
Jerden
,
J.
,
2016
, “
Regulatory Technology Development Plan Sodium Fast Reactor-Mechanistic Source Term- Metal Fuel Radionuclide Release
,” Argonne National Laboratory, Argonne, IL, Report No. ANL-ART-38.
12.
Ogata
,
T.
,
2002
, “
Irradiation Behavior and Thermodynamic Properties of Metallic Fuel
,”
J. Nucl. Sci. Technol.
,
39
(
3
), pp.
675
681
.10.1080/00223131.2002.10875558
13.
Janney
,
D. E.
,
Papesch
,
C.
, and
Middlemas
,
S. C.
,
2016
, “
FCRD Advanced Reactor (Transmutation) Fuels Handbook
,” Idaho National Laboratory, Idaho Falls, ID, Report No. INL/EXT-15-36520.
14.
Colle
,
J.-Y.
,
Hiernaut
,
Y.-P.
,
Wiss
,
T.
,
Beneš
,
O.
,
Thiele
,
H.
,
Papaioannou
,
D.
,
Rondinella
,
V. V.
,
Sasahara
,
A.
,
Sonoda
,
T.
, and
Konings
,
R. J. M.
,
2013
, “
Fission Product Release and Microstructure Changes of Irradiated MOX Fuel at High Temperatures
,”
J. Nucl. Mater.
,
442
(
1–3
), pp.
330
340
.10.1016/j.jnucmat.2013.09.022
15.
Clément
,
B.
, and
R. Zeyen
,
R.
,
2013
, “
The Objectives of the Phébus FP Experimental Programme and Main Findings
,”
Ann. Nucl. Energy
,
61
, pp.
4
10
.10.1016/j.anucene.2013.03.037
16.
Gallais-During
,
A.
,
Bonnin
,
J.
,
Malgouyres
,
P.-P.
,
Morin
,
S.
,
Bernard
,
S.
,
Gleizes
,
B.
,
Pontillon
,
Y.
,
Hanus
,
E.
, and
Ducros
,
G.
,
2014
, “
Performance and First Results of Fission Product Release and Transport Provided by the VERDON Facility
,”
J. Nucl. Eng. Des.
,
277
, pp.
117
123
.10.1016/j.nucengdes.2014.05.045
17.
Haste
,
T.
,
Steinbrück
,
M.
,
Barrachin
,
M.
,
de Luze
,
O.
,
Grosse
,
M.
, and
Stuckert
,
J.
,
2015
, “
A Comparison of Core Degradation Phenomena in the CORA, QUENCH, Phébus SFD and Phébus FP Experiments
,”
J. Nucl. Eng. Des.
,
283
, pp.
8
20
.10.1016/j.nucengdes.2014.06.035
18.
2012
, “
Chapter 5—Fission Product Release and Transport
,”
Nuclear Safety in Light Water Reactors, Severe Accident Phenomenology
,
B. R.
Sehgal
, ed.,
Academic Press
, Cambridge, MA, pp.
425
517
.
19.
Miller
,
B. D.
,
Wright
,
K.
,
Hernandez
,
B. J.
,
Harp
,
J.
,
Meyer
,
M. K.
, and
Gan
,
J.
,
2017
, “
Current Status of the Irradiated Materials Characterization Laboratory at INL With Limited PIE Microstructural Characterization
,”
Conference Proceeding Hot Lab Conference
, Mito, Japan, Sept. 17–22, Paper No.
INL/CON-17-42391-Revision-0
.https://inldigitallibrary.inl.gov/sites/sti/sti/Sort_2411.pdf
20.
Di Lemma
,
F. G.
,
Liu
,
X.
,
Holschuh
II
,
T. V.
,
Folsom
,
C. P.
,
Murray
,
D. J.
,
Teng
,
F.
, and
Jensen
,
C. B.
,
2020
, “
Investigation of the Microstructure Evolution of Alpha Uranium After in Pile Transient
,”
J. Nucl. Mater.
,
542
, p.
152467
.10.1016/j.jnucmat.2020.152467
21.
Di Lemma
,
F. G.
,
Wright
,
K. E.
,
Capriotti
,
L.
,
Zabriskie
,
A. X.
,
Winston
,
A. J.
,
Jensen
,
C. B.
, and
Wachs
,
D. M.
,
2021
, “
Investigation of the Microstructure of the Top of a Metallic Fuel Pin After a Reactor Transient
,”
J. Nucl. Mater.
,
544
, p.
152711
.10.1016/j.jnucmat.2020.152711
22.
Liu
,
Y. Y.
,
Tsai
,
H.
,
Billone
,
M. C.
,
Holland
,
J. W.
, and
Kramer
,
J. M.
,
1993
, “
Behavior of EBR-II Mk-V-Type Fuel Elements in Simulated Loss-of-Flow Tests
,”
J. Nucl. Mater.
,
204
, pp.
194
202
.10.1016/0022-3115(93)90217-M
23.
Liu
,
Y. Y.
,
Tsai
,
H. C.
,
Donahue
,
D. A.
,
Pushis
,
D. O.
,
Savoie
,
F. E.
,
Holland
,
J. W.
,
Wright
,
A. E.
,
August
,
C.
,
Bailey
,
J. L.
, and
Patterson
,
D. R.
,
1990
, “
Whole-Pin Furnace System: An Experimental Facility for Studying Irradiated Fuel Pin Behavior Under Potential Reactor Accident Conditions
,”
Conference Proceeding: International Topical Meeting on Fast Reactor Safety
, Snowbird, UT, Aug. 12–16, Paper No. CONF-900804-204.
24.
Cohen
,
A. B.
,
Tsai
,
H.
, and
Neimark
,
L. A.
,
1993
, “
Fuel/Cladding Compatibility in U-19Pu-10Zr/HT9-Clad Fuel at Elevated Temperatures
,”
J. Nucl. Mater.
,
204
, pp.
244
251
.10.1016/0022-3115(93)90223-L
25.
Tsai
,
H.
,
1989
, “
A Versatile Apparatus for Studying Irradiated Fuel Behavior
,”
Conference Proceeding: American Nuclear Society Winter Meeting
, San Francisco, CA, Nov. 26–30, Paper No. CONF-891103-54.
26.
Harp
,
J. M.
,
Capriotti
,
L.
, and
Chichester
,
H. J. M.
,
2019
, “
Postirradiation Examination of FUTURIX-FTA Metallic Alloy Experiments
,”
J. Nucl. Mater.
,
515
, pp.
420
433
.10.1016/j.jnucmat.2018.12.051
27.
Li
,
M.
,
Kirk
,
M. A.
,
Baldo
,
P. M.
,
Xu
,
D.
, and
Wirth
,
B. D.
,
2012
, “
Study of Defect Evolution by TEM With In Situ Ion Irradiation and Coordinated Modeling
,”
Philos. Mag., Part A: Mater. Sci.
,
92
(
16
), pp.
2048
2078
.10.1080/14786435.2012.662601
28.
Kirk
,
M. A.
,
Baldo
,
P. M.
,
Liu
,
A. C. Y.
,
Ryan
,
E. A.
,
Birtcher
,
R. C.
,
Yao
,
Z.
,
Xu
,
S.
,
Jenkins
,
M. L.
,
Hernandez-Mayoral
,
M.
,
Kaoumi
,
D.
, and
Motta
,
A. T.
,
2009
, “
In Situ Transmission Electron Microscopy and Ion Irradiation of Ferritic Materials
,”
Microsc. Res. Tech.
,
72
(
3
), pp.
182
186
.10.1002/jemt.20670
29.
Yao
,
T.
,
Wagner
,
A. R.
,
Liu
,
X.
,
Ei-Azab
,
A.
,
Harp
,
J. N.
,
Gan
,
J.
,
Hurley
,
D. H.
,
Benson
,
M. T.
, and
He
,
L.
,
2020
, “
On Spinodal-Like Phase Decomposition in U–50Zr Alloy
,”
Materials
,
9
, p.
100592
.10.1016/j.mtla.2020.100592
30.
Marchand
,
B.
,
Moncoffre
,
N.
,
Pipon
,
Y.
,
Bérerd
,
N.
,
Garnier
,
C.
,
Raimbault
,
L.
,
Sainsot
,
P.
,
Epicier
,
T.
,
Delafoy
,
C.
,
Fraczkiewicz
,
M.
,
Gaillard
,
C.
,
Toulhoat
,
N.
,
Perrat-Mabilon
,
A.
, and
Peaucelle
,
C.
,
2013
, “
Xenon Migration in UO2 Under Irradiation Studied by SIMS Profilometry
,”
J. Nucl. Mater.
,
440
(
1–3
), pp.
562
567
.10.1016/j.jnucmat.2013.04.005
31.
Hocking
,
W. H.
,
Verrall
,
R. A.
, and
Bushby
,
S. J.
,
1999
, “
A New Technique to Measure Fission-Product Diffusion Coefficients in UO2 Fuel
,” Chalk River, ON, Canada, Report No. IAEA-TECDOC-1122.
32.
Kolay
,
S.
,
Shirsat
,
A. N.
,
Ali Basu
,
M.
, and
Das
,
D.
,
2018
, “
Xenon Transport in Uranium-10 wt% Zirconium, Material Science (S36)
,” BARC Newsletter No. 331, Bhabha Atomic Research Centre, Mumbai, India, pp.
13
18
.
33.
Ahn
,
S.
,
Irukuvarghula
,
S.
, and
McDeavitt
,
S. M.
,
2016
, “
Microstructure of α-U and δ-UZr2 Phase Uranium-Zirconium Alloys Irradiated With 140 keV He+ Ion-Beam
,”
J. Alloys Compd.
,
681
, pp.
6
11
.10.1016/j.jallcom.2016.04.219
34.
Miao
,
Y.
,
Harp
,
J.
,
Mo
,
K.
,
Zhu
,
S.
,
Yao
,
T.
,
Lian
,
J.
, and
Yacout
,
A. M.
,
2017
, “
Bubble Morphology in U3Si2 Implanted by High-Energy Xe Ions at 300 °C
,”
J. Nucl. Mater.
,
495
, pp.
146
153
.10.1016/j.jnucmat.2017.07.066