Fission Gas Research Articles

NRC’s methodology to estimate fuel dispersal during a large break loss of coolant accident

Recent experimental findings indicate that under certain transient conditions, high-burnup fuel operating at sufficiently high power can fragment and escape from burst fuel cladding, as discussed in the Nuclear Regulatory Commission’s Research Information Letter (RIL) 2021–13. The escaped fuel fragments could subsequently be distributed throughout a reactor coolant system (RCS), potentially challenging core coolability during a loss of coolant accident (LOCA). Consequently, the NRC has developed a methodology to estimate the mass of fuel that could be dispersed during a large-break LOCA. The methodology involves several NRC-sponsored computer codes, including SCALE/Polaris for lattice physics, PARCS/PATHS for full-core neutronics, TRACE for core and RCS thermal hydraulics, and FAST for steady-state and transient fuel performance calculations. Output from FAST is combined with the models in RIL 2021–13 to estimate fuel dispersal to the coolant.NRC staff have exercised this methodology using a prospective Westinghouse four loop core design with high burnup and extended enrichment that was obtained from the Department of Energy’s NEAMS program. Results demonstrate that the NRC methodology can be exercised to provide fuel dispersal estimates that can subsequently be used to study the potential consequences of FFRD. However, this exercise has also identified several areas for improvements to both the codes and to the modeling approaches used here, including the pin power reconstruction methods in PARCS, the modeling approach used to connect the Cartesian and cylindrical vessel components in TRACE, transient fission gas release and fuel rod upper plenum models in FAST, and tighter coupling between FAST and TRACE.

Interpretation of the online measurement data for selected tests of the RISOE-III project

Two tests, AN3 and AN4 from the RISOE-III Fission Gas Release (FGR) project are analysed using the updated version of the FALCON MOD01 code coupled with the GRSW-A model. Based on comparison of the stand-alone code calculation with the measurement for centerline fuel temperature, the previous calibration and verification of the thermal analysis of the code is confirmed. Additional analysis with the FALCON-to-FRELAX (F2F) coupled code system is carried out and presented for simulation of the effects of delayed axial redistribution of the released fission gases in the rodlets, as an attempt to interpret the data on significant jumps in measured upper-plenum pressure at a power dip. The other feature of the data, addressed by the additional sensitivity study with F2F, is peaking that can be seen in the measured fuel temperature after each power rise. A conclusion can be proposed that the peculiarities in the online-measurement could, probably, result from the features of fabrication of the fuel segments for pre-irradiation in the NPP, and instrumentation of the rodlets to be tested in the research reactor. Therefore, these features can have only limited effect on evaluation of reliability and safety of the standard fuel during normal operation in power reactors.

The role of xenon/krypton adsorption on the growth of intra-granular bubbles in irradiated UO2

Under transient conditions, rapid fuel swelling in UO2 can occur through fission gas and vacancy capture by intra-granular bubbles. These swellings can lead to significant cladding deformations and clad failure. To understand the mechanisms of intra-granular swelling a number of key experiments have been conducted by the UK in conjunction with the Halden Reactor. Post-irradiation annealing enables the study of the thermal behaviour of the intra-granular bubbles in the absence of complicating processes such as irradiation-induced bubble destruction and fission gas resolution. Bubble-like features are frequently observed in straight lines and it has always been assumed that these were nucleated in the wake of energetic fission fragments and served as nuclei for bubble growth. However, most of these bubbles do not grow significantly and viable intra-granular bubbles are nucleated at much lower densities. At the temperature and pressure that these form, the xenon and krypton are above their critical points and the elements are supercritical fluids. However, bubble growth occurs at temperatures and pressures much greater than the critical points and the gases may be treated with simple equations of state such as those by Clausius or Van der Waals. Of greater importance is the fact that the large Van der Waals forces in these atoms generated by fluctuations in the electron dispositions lead to extensive adsorption on the inner surfaces of the bubbles and vacancy capture by the bubbles can lead to the diffusive loss of fission gas atoms resulting in bubble growth at slight over-pressure conditions. A diffusion model based on Langmuir Adsorption has been developed and shown to be consistent with the out-of-pile annealing studies.

Fission accelerated steady-state post irradiation examinations - Part II

The Advanced Fuels Campaign's Fission Accelerated Steady State Testing (FAST) approach at Idaho National Laboratory creates a benchmark for evaluating accelerated irradiation via control rodlets and advanced metal fuel alloys for sodium-cooled fast reactors (SFRs). FAST experiments have been developed to generate prototypic temperature conditions during steady state irradiations of scaled geometric fuel pins. This approach helps to attain higher burn ups at a much faster rate than previous irradiation tests. For this study, the results from profilometry, fission gas release, and metallography of a FAST experiment are presented. Profilometry determined 0 % effective strain in the rodlets. The fission gas release fraction was measured from puncture/collection analysis. Constituent redistribution was observed in two specimens despite the peak fuel temperatures being below the normal ranges in which redistribution is expected. Metallography of the two higher temperature specimens showed typical swelling with the solid pin closing the fuel-cladding gap and the annular specimen having a fully closed annulus. Additionally, metallography indicated no swelling, no redistribution, and a homogenous microstructure for specimens with lower irradiation temperature. Post irradiation examination of FAST rodlets generally showed the expected representative behavior of metallic fuels within SFRs.

Multi-physical effects induced by spatiotemporal corrosion of T91 cladding in oxygen-controlled LBE environment

T91, a pivotal candidate material for cladding in lead-based fast reactors (LFRs), demonstrates considerable potential for corrosion resistance in liquid lead-bismuth eutectic (LBE) corrosion experiments. However, the multi-physics characteristics of T91 cladding corrosion in its operational environment remain unclear, posing an obstacle to its practical application. This study focuses on the spatiotemporal corrosion of T91 cladding in multi-physics coupling scenarios by means of COMMA (Cladding/Oxidation corrosion Multi-physics Modeling and Analysis code), in which, modules of cladding oxidation, cladding oxide removal, heat transfer, mechanical deformation, and fission gas release are integrated and collaboratively solved. The corrosion patterns of T91 cladding with different oxygen concentrations and the underlying multi-physics behavior of the corrosion processes under operational conditions simulating are summarized. For the lower oxygen concentration of 1.8e-9 wt% in LFRs, the results indicate a 0.2 m axial zone with base material exposed to LBE reveals the presence of the oxidation protection failure. An exacerbation of cladding heat transfer degradation is observed after oxygen concentration reaching the oxygen-dominant region, due to the enhanced temperature positive feedback of oxidation corrosion. For the high oxygen concentration of 1.4e-6 wt%, an axial offset of 0.24 m for the location of the minimum gap size is observed. This study identifies a trend of the reduction thickness of cladding that first decreasing and then increasing with an elevated oxygen concentration, attributed to a transition in corrosion patterns. The temperature and mechanical analyses indicate that the optimal oxygen concentration for corrosion inhibition and fuel safety is located around the critical oxygen-dominant concentration.

Assessing the PLEIADES/GERMINAL V3 fuel performance code against transmutation experiments in sodium fast reactors

We present the results of a first assessment of the multiphysics PLEIADES/GERMINAL V3 fuel performance code against integral irradiation experiments of americium-bearing MOX fuel pins irradiated in sodium fast reactors. The pins considered in this work cover a range of americium contents from 2 to 5 weight %, being representative for the homogeneous transmutation path. The code predictions are compared to several post-irradiation examination results on integral (e.g., fission gas release) and local quantities (e.g., actinides redistribution in the fuel pellet), and demonstrate a satisfactory agreement with experimental data. The deviations between calculated and experimental quantities shed some light on modeling developments needed to better calculate the behavior of americium-bearing mixed-oxides fuel pins under irradiation.

Phase-field simulation of the effect of grain boundary on fission gas migration in UO2 fuel

Grain boundaries are widely recognized as line defect sinks. During reactor operation, vacancies and fission gas atoms within UO2 fuel grains migrate to the grain boundaries, forming bubbles at these locations. In order to better understand the effect of grain boundaries on the migration of fission gas atoms, this study employed a phase-field model to simulate the nucleation and growth process of fission gas bubbles within the UO2 fuel grains and in the vicinity of grain boundaries. This study also investigated the degree to which grain boundaries affect fission gas atoms under different temperature conditions. The results indicated that in models containing grain boundaries, the nucleation of fission gas bubbles occurred earlier, as compared to models without grain boundaries. A noticeable bubble denuded zone was also observed adjacent to grain boundary interfaces. Furthermore, with increasing temperature, the bubble denuded zone becomes thicker. The effects of irradiation, vacancy diffusion, Ostwald ripening, as well as grain boundary trapping were discussed.

Transient analysis of temperature field and fission gas about the arc plate fuel element

Obtaining the temperature field and fission gas release of fuel elements are crucial to the study on fuel behaviors and ensure the safety of elements. In order to enhance fuel efficiency, providing a design idea to make the fast reactor with high flux and multi-function, we have designed a core of the lead-based fast reactor with arc plate fuel elements. Based on above, a transient performance analysis program of arc plate fuel elements (TPAPAFE) was developed in this paper, for analyzing the thermal and fission gas module of arc plate fuel element by calculating transient variation in the maximum pellet temperature, internal and external coolant outlet temperature, and fission gas release share. The modular verification was done by comparing the results of TPAPAFE with others. Besides, the result of the arc plate fuel analyzed by TPAPAFE shows that the code is feasible and rational, and the single arc plate is so thin that coolant outlet temperatures on both sides of the plate are almost the same. Meanwhile, the error in temperature data between TPAPAFE and the steady-state program is only 5%. Furthermore, fission gas release fraction increased linearly with pellet temperature and bubble growth.

Impact of grain boundary and surface diffusion on predicted fission gas bubble behavior and release in UO2 fuel

In this work, we quantify the impact of grain boundary (GB) and surface diffusion on fission gas bubble evolution and fission gas release in UO2 nuclear fuel using simulations with a hybrid phase field/cluster dynamics model. We begin with a comprehensive literature review of uranium vacancy and xenon atom diffusivity in UO2 through the bulk, along GBs, and along surfaces. In our model we represent fast GB and surface diffusion using a heterogeneous diffusivity that is a function of the order parameters that represent bubbles and grains. We find that the GB diffusivity directly impacts the rate of gas release via GB transport, and that the GB diffusivity is likely below 104 times the lower value from Olander and van Uffelen (2001). We also find that the surface diffusivity impacts bubble coalescence and mobility, and that the bubble surface diffusivity is likely below 10−4 times the value from Zhou and Olander (1984).

CRUD deposition induced multi-physics effects analysis in PWR

In PWR, the deposition of corrosion products from the steam generator leads to the formation of Chalk River Unidentified Deposits (CRUD) on the surface of fuel rods. CRUD deposition can cause issues such as CRUD induced localized corrosion, axial power anomaly and deterioration of cladding heat transfer. A new multi-physics program named CIFF (CRUD Induced eFFects analysis) is developed to analyze the effects caused by CRUD deposits. CIFF is a 1.5D model and can be used to simulate the variations in surface temperature, cladding corrosion and fuel deformation caused by CRUD deposition. A three-node CRUD deposition model is developed based on mass transfer theory and chemical equilibrium. The calculations of heat transfer, mechanical deformation, oxide layer growth, CRUD deposition and fission gas behavior are fully coupled based on finite element analysis and numerical solution technology. Axisymmetric element is used to solve the deformation of fuel rod under thermal expansion, gas pressure, creep and radiation swelling. The performance of CIFF is validated through IFA-513, ANO-2, Oconee and PCCL cases. The DBN-1 case is used to explore the influence of CRUD. The results indicate that CRUD deposition and CRUD induced local corrosion deteriorate the heat transfer of cladding, resulting in a three-stage increase of cladding temperature and accelerated growth of the oxide layer. CRUD deposition also leads to a relative increase in the gap pressure exceeding the coolant pressure by 8.4 %, a relative increase in the fuel rod strain by 10 % and a relative decrease in the gap size by 11 %, which indirectly leads a premature pellet-cladding mechanical interaction by 90 days.

Performance and properties evolution of near-term accident tolerant fuel: Cr-doped UO2

Chromium-doped UO2 fuel has received significant interest due to the ability for chromium to produce pellets with large average grain size (>30 μm), which has shown to increase fission gas retention during operation. Sintering of chromium-doped UO2 pellets was pursued with oxygen potential and sintering atmosphere controlled to tailor the final microstructure of the material. Chromium additions in this study ranged from 750 to 7800 ppm. Cr concentrations were studied pre and post sintering using Inductively Coupled Plasma Optical Emission Spectroscopy (ICP-OES). Effects of chromium content on lattice parameter and microstructure were examined with X-ray diffraction (XRD) and scanning electron microscopy (SEM). Contraction of the UO2 lattice parameter was observed, as well as enlargement of grain size with increasing chromium content up to 4900 ppm Cr2O3. SEM indicated Cr incorporation within the matrix and the formation of chromium oxide precipitates throughout the microstructure at high Cr concentrations. Evaluation of thermophysical properties of Cr-doped UO2 pellets were conducted up to 1200 °C to illustrate their evolution with increased dopant concentration and microstructural changes. The results show that grain size is maximized at 52 μm with Cr2O3 concentration equal to 4900 ppm; however, grain size decreases at higher Cr2O3 concentrations. No significant changes were observed in specific heat capacity, linear thermal expansion, and coefficient of thermal expansion compared to undoped UO2. The thermal conductivity also decreased through the incorporation of Cr2O3 dopants above 750 ppm and is shown to be ∼15 % lower than reported UO2 values.

Structural optimization of gas-solid separator used in TMSR-SF based on computational fluid dynamics

The fission gas removal system plays a critical role in the Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF). A gas–solid separator is used to separate the waste gas and solid particles in the gas removal system. As a key component of the system, the gas–solid separator directly determines the system energy efficiency. Nowadays, a new combined gas–solid separation equipment combining axial flow blade cyclone separator and filter for TMSR-SF have been studied. However, the primary separation efficiency of the new separation equipment is low, resulting in a large number of particles entering the secondary filter and causing wear and tear of the filter material. Based on the above considerations, a new combined gas–solid separator consisting of a stable chamber with swirl louver structure and a filter is developed for the TMSR-SF in this study. The velocity and pressure fields of two phases in this new combined separator are calculated and analyzed by using commercial software FLUENT, and the separation performance for this structure is also discussed. Then the structure is optimized considering the influence on pressure and separation efficiency. The results show that the RNG k-ε is feasible and accurate to simulate the flow fields of the new combined separator. Coarse particles are separated in the first-stage by mechanical separation and fine particles are separated in the second-stage filter. The mechanical separation efficiency of the first stage separator with the optimum structure can reach 72.49%. The unit pressure drop separation efficiency of the new combined separator is increased by 30.32% and much better than the original combined separator.

Cradle to grave: the importance of the fuel cycle to molten salt reactor sustainability

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective.

The critical influencing factors responsible for the particle cracking in UMo/Zr dispersion fuel plates during post-irradiation anneal tests

The post-irradiation anneal tests for dispersion fuel plates are commonly employed to determine the blistering threshold temperature in order to evaluate their resistance to accidents. In this study, a three-dimensional simulation method for UMo/Zr dispersion fuel plates is developed to predict the normal stresses in the fuel skeleton and some related variables during in-pile irradiation and out-of-pile anneal tests. The effects of the intrusion of extra fission gas atoms into gas bubbles are investigated, and the contributions of the creep performances of fuel particles and the matrix are examined. The predictions reveal the following insights: (1) the thermo-mechanical behaviors during anneal tests are strongly dependent on the intrusion of extra fission gas atoms into gas bubbles and the additional fission-gas induced swelling during post-irradiation anneal tests; the maximum tensile stress in the fuel skeleton increases significantly, from 57.4 MPa to 346.0 MPa, as the fraction of generated fission gas atoms in gas bubbles shifts from approximately 23.4 % to 100 % in the annealing process; (2) the thermo-mechanical responses during anneal tests are greatly influenced by the creep performances of fuel particles and the matrix; the significantly weakened creep ability of fuel particles is found to be the critical factors responsible for particle cracking in the annealing process; the fuel particle fracture will hardly occur when the irradiation creep rate model is maintained during the considered annealing scheme, because the skeleton tensile stress does not appear. It is important to note that the particle cracking and the subsequent blister behaviors under in-pile neutron irradiation conditions are possible to occur if a rapid temperature increase takes place under accident conditions. Consequently, the out-of-pile annealing tests are of significance.

Review of fission gas release in liquid metal reactor fuel cladding failure accident

The release of gas fission products from liquid metal fast reactors in the event of cladding failure accidents is a significant focus during the development and improved design of such reactors. Researchers worldwide have devoted considerable attention to this topic. But studies aiming at liquid metal reactors have not yet reached the same level as studies of fission gas release in water-cooled reactors. To address this gap, this paper reviews the research progress of related fission gas release from the aspects of mechanism model, experimental research, numerical simulation, and measurement. Considering the potential essential similarity of the release process of fission gas in water-cooled reactors and liquid metal cooled reactors, the researches on water-cooled reactors are also included. This review helps to understand the behavior and characteristics of fission gas release in liquid metal cooled reactors and provides some guidance for subsequent numerical simulation and experimental research on fission gas release based on liquid metal cooled reactors.

Phase-field simulation on fission gas release behavior of large grain UO<sub>2</sub> fuel

In order to predict the release behavior of fission gas in large grain UO<sub>2</sub> fuel and provide support for the development of accident tolerant fuel, a phase-field model is used to simulate the release behavior of fission gas in the microstructure of UO<sub>2</sub> polycrystalline in this work. This model adopts a set of coupled Cahn-Hilliard equations and Allen-Cahn equations, using conserved field variables to represent the distribution of fission gas and vacancies, and distinguishing bubble phase from matrix phase by using order parameters. This model focuses on investigating the effects of different grain sizes, temperature conditions, and diffusion coefficients on the release behavior of fission gas, demonstrating the nucleation, growth, and fusion behavior of bubbles. Simulation results are obtained for fuel porosity, bubble coverage on grain boundaries, and average bubble radius at a certain degree of burnup. The results show that temperature and diffusion coefficient have a significant influence on porosity and bubble coverage on grain boundaries. When the diffusion coefficient is high, grain size also has a significant influence on fission gas release behavior. And when the diffusion coefficient is low, the influence of grain size is not significant. In addition, the distribution of fission gas bubbles under high burnup obtained through this model is also in good agreement with experimental result. The model can predict the behavior of fission gas release in large grain UO<sub>2</sub> fuel.

Building a DFT+U machine learning interatomic potential for uranium dioxide

Despite uranium dioxide (UO2) being a widely used nuclear fuel, fuel performance models rely extensively on empirical correlations of material behavior, leveraging the historical operating experience of UO2. Mechanistic models that consider an atomistic understanding of the processes governing fuel performance (such as fission gas release and creep) will enable a better description of fuel behavior under non-prototypical conditions such as in new reactor concepts or for modified UO2 fuel compositions. To this end, molecular dynamics simulation is a powerful tool for rapidly predicting physical properties of proposed fuel candidates. However, the reliability of these simulations depends largely on the accuracy of the atomic forces. Traditionally, these forces are computed using either a classical force field (FF) or density functional theory (DFT). While DFT is relatively accurate, the computational cost is burdensome, especially for f-electron elements, such as actinides. By contrast, classical FFs are computationally efficient but are less accurate. For these reasons, we report a new accurate machine learning interatomic potential (MLIP) for UO2 that provides high-fidelity reproduction of DFT forces at a similar low cost to classical FFs. We employ an active learning approach that autonomously augments the DFT training data set to iteratively refine the MLIP. To further improve the quality of our predictions, we utilize transfer learning to retrain our MLIP to higher-accuracy DFT+U data. We validate our MLIPs by comparing predicted physical properties (e.g., thermal expansion and elastic properties) with those from existing classical FFs and DFT/DFT+U calculations, as well as with experimental data when available.

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

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

An efficient instance segmentation approach for studying fission gas bubbles in irradiated metallic nuclear fuel

Gaseous fission products from nuclear fission reactions tend to form fission gas bubbles of various shapes and sizes inside nuclear fuel. The behavior of fission gas bubbles dictates nuclear fuel performances, such as fission gas release, grain growth, swelling, and fuel cladding mechanical interaction. Although mechanical understanding of the overall evolution behavior of fission gas bubbles is well known, lacking the quantitative data and high-level correlation between burnup/temperature and microstructure evolution blocks the development of predictive models and reduces the possibility of accelerating the qualification for new fuel forms. Historical characterization of fission gas bubbles in irradiated nuclear fuel relied on a simple threshold method working on low-resolution optical microscopy images. Advanced characterization of fission gas bubbles using scanning electron microscopic images reveals unprecedented details and extensive morphological data, which strains the effectiveness of conventional methods. This paper proposes a hybrid framework, based on digital image processing and deep learning models, to efficiently detect and classify fission gas bubbles from scanning electron microscopic images. The developed bubble annotation tool used a multitask deep learning network that integrates U-Net and ResNet to accomplish instance-level bubble segmentation. With limited annotated data, the model achieves a recall ratio of more than 90%, a leap forward compared to the threshold method. The model has the capability to identify fission gas bubbles with and without lanthanides to better understand the movement of lanthanide fission products and fuel cladding chemical interaction. Lastly, the deep learning model is versatile and applicable to the micro-structure segmentation of similar materials.

A fine pore-preserved deep neural network for porosity analytics of a high burnup U-10Zr metallic fuel

U-10 wt.% Zr (U-10Zr) metallic fuel is the leading candidate for next-generation sodium-cooled fast reactors. Porosity is one of the most important factors that impacts the performance of U-10Zr metallic fuel. The pores generated by the fission gas accumulation can lead to changes in thermal conductivity, fuel swelling, Fuel-Cladding Chemical Interaction (FCCI) and Fuel-Cladding Mechanical Interaction (FCMI). Therefore, it is crucial to accurately segment and analyze porosity to understand the U-10Zr fuel system to design future fast reactors. To address the above issues, we introduce a workflow to process and analyze multi-source Scanning Electron Microscope (SEM) image data. Moreover, an encoder-decoder-based, deep fully convolutional network is proposed to segment pores accurately by integrating the residual unit and the densely-connected units. Two SEM 250 × field of view image datasets with different formats are utilized to evaluate the new proposed model’s performance. Sufficient comparison results demonstrate that our method quantitatively outperforms two popular deep fully convolutional networks. Furthermore, we conducted experiments on the third SEM 2500 × field of view image dataset, and the transfer learning results show the potential capability to transfer the knowledge from low-magnification images to high-magnification images. Finally, we use a pre-trained network to predict the pores of SEM images in the whole cross-sectional image and obtain quantitative porosity analysis. Our findings will guide the SEM microscopy data collection efficiently, provide a mechanistic understanding of the U-10Zr fuel system and bridge the gap between advanced characterization to fuel system design.