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DOE OSTI2023

Post Irradiation Examination Results of Irradiated Yttrium Hydride

Department of Energy’s (DOE’s) Microreactor program (MRP) aims to provide the fundamental data to enable the development of microreactors. As such, material property data of critical materials for microreactor technologies are researched. Because substoichiometric yttrium dihydride (YHx, where x<2) is considered as a potential solid neutron moderator, its material property data has been combined in the Advanced Moderator Material Handbook which includes thermodynamic and thermophysical properties of YHx with the exception of irradiated material’s properties due to limited PIE. To fill the knowledge gap for the irradiated YHx, specimens and irradiation capsules were prepared at Los Alamos National Laboratory (LANL). Specimens were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory’s (INL’s). post-irradiation examination (PIE) was performed at INL’s Materials and Fuels Complex (MFC). This report compiles the PIE results of irradiated YHx specimens through fiscal years 2022 and 2023 (FY22-23) . The PIE data will directly be incorporated into the newer version of the Advanced Moderator Material Handbook. The main takeaways include that (i) the geometrical stability and mechanical integrity of YHx was intact with couple exceptions after high-temperature irradiations (600-800°C), (ii) hydrogen content variation due to manufacturing or irradiation in YHx caused visible surface discoloration, that is also related to the microstructural changes, (iii) qualitative comparisons of PIE methods implied that H retention was significantly higher at 600°C as compared to 800°C, (iv) thermal properties included signatures correlated with the H loss or re-gain, (v) importance of manufacturing readiness and initial as-manufactured specimens history was emphasized, (vi) the needs of targeted irradiations focusing on temperature and time parameters and very targeted PIE were specified.

08 HYDROGEN↗

DOE OSTI2024

Status Report on Fast Flux Test Facility Mechanistic Fuel Failure Experiment Analysis with BISON for Post Irradiation Examination Support

The renewed interest in metallic U-Zr nuclear fuel alloy has led to a drive for deeper understanding of the mechanisms driving the phenomena observed under irradiation conditions. The Department of Energy Advanced Fuel Campaign has developed infrastructure to support metallic fuel development, including Post Irradiation Examination (PIE) of legacy Fast Flux Test Facility (FFTF) Mechanistic Fuel Failure (MFF) experiments. The PIE performed on legacy FFTF MFF experiments gives insight on metallic fuel performance and can address the lack of knowledge and scarcity of reliable data identified in several studies over recent years. Unfortunately, PIE efforts can cost significant time and resources which can impede the progress of metallic U-Zr fuel development. Metallic U-Zr fuel performance modeling can be used to inform PIE efforts on regions of interest for relevant investigations and can help understand phenomena observed in PIE. This report demonstrates the current progress of FFTF MFF fuel performance simulations using the BISON fuel performance code and discusses the support provided by simulation to PIE efforts. Progress in temperature, profilometry, fission gas release, plenum pressure, and zirconium redistribution simulation results have been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Report on the Incorporation of Post-Irradiation Examination Results of Yttrium Hydride for Advanced Moderator Handbook

08 HYDROGEN↗

DOE OSTI2025

Recent Metallic Fuel Data Recovery in FIPD

The Metallic Fuels Irradiation and Physics Database (FIPD) [1] is an organized collection of metallic fuel test pin data (U-xPu-yZr, = 0 ~ 28; y = 2 ~ 10) and documentation available to industry. FIPD mainly contains three types of data: (1) Fuel pin fabrication data, including fuel slug diameter, fuel slug length, cladding diameter, smear density, etc. (2) Fuel pin operation conditions, including axial distributions for power, temperatures, fluences, burnup, and isotopic densities, etc. and (3) Fuel pin post-irradiation examination (PIE) data, including fission gas release and gas chemistry, profilometry, and neutron radiography, etc. The operating conditions for pins with PIE data available in FIPD span significant ranges across key parameters. The fuel peak burnup extends from less than 5% up to 20 at%. The cladding peak temperature varies from about 490°C to 660°C. Finally, the cladding peak DPA shows a wide range from less than 5 to 120. These broad ranges reflect the diverse testing conditions and operational parameters captured in the available PIE data. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). As reported in the ref. [3, 4], most of the PIE data generated during the IFR program [5] has been collected, reviewed, processed, and integrated into FIPD. The most recently added PIE data can be found in ref. [4], which shows the collection of over 95% of the PIE data by the time of this paper. The recent improvements to the database are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Nondestructive and destructive assay for forensics characterization of weapons-grade plutonium produced in LEU irradiated in a thermal neutron spectrum

A post-irradiation examination (PIE) of an enriched uranium sample was performed to advance a nuclear forensics methodology, based on intra-elemental nuclide ratios. The PIE was carried out on a few milligrams of low enriched uranium dioxide (LEUO 2 ) sample, irradiated to a burnup of 1 GWd/MTU. The PIE of the LEUO 2 sample provided the concentration of selected plutonium and fission product nuclides ( 239 Pu, 240 Pu, 241 Pu, 91 Y, 95 Zr, 95 Nb, 103 Ru, 133 Cs, 134 Cs, 135 Cs, 137 Cs, 136 Ba, 138 Ba, 140 Ba, 140 La, 141 Ce, 144 Ce, 149 Sm, 150 Sm, 152 Sm, 153 Eu, and 154 Eu). The PIE supported validation of the MCNP (Monte Carlo N-Particle neutronics simulation code) predicted concentration of these nuclides in the irradiated LEUO 2 sample. This study showed that most of the MCNP predicted nuclide concentrations were within 15% and several within 10% compared to the PIE results. Furthermore, gamma and mass spectrometry of nuclides in irradiated LEUO 2 enabled the determination of uranium burnup and time since irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1990

Validation environment for AIPS/ALS: Implementation and results

The work is presented which was performed in porting the Fault Injection-based Automated Testing (FIAT) and Programming and Instrumentation Environments (PIE) validation tools, to the Advanced Information Processing System (AIPS) in the context of the Ada Language System (ALS) application, as well as an initial fault free validation of the available AIPS system. The PIE components implemented on AIPS provide the monitoring mechanisms required for validation. These mechanisms represent a substantial portion of the FIAT system. Moreover, these are required for the implementation of the FIAT environment on AIPS. Using these components, an initial fault free validation of the AIPS system was performed. The implementation is described of the FIAT/PIE system, configured for fault free validation of the AIPS fault tolerant computer system. The PIE components were modified to support the Ada language. A special purpose AIPS/Ada runtime monitoring and data collection was implemented. A number of initial Ada programs running on the PIE/AIPS system were implemented. The instrumentation of the Ada programs was accomplished automatically inside the PIE programming environment. PIE's on-line graphical views show vividly and accurately the performance characteristics of Ada programs, AIPS kernel and the application's interaction with the AIPS kernel. The data collection mechanisms were written in a high level language, Ada, and provide a high degree of flexibility for implementation under various system conditions.

Segall, Zary↗

DOE OSTI2023

AGR-5/6/7 Irradiation Disassembly and Metrology First Look

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of high-temperature gas-cooled reactors (HTGRs), which are graphite-moderated nuclear reactors cooled with helium. This work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials (e.g., graphite), safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product transport and source term determination. The ART AGR program has conducted four distinct fuel irradiation experiments in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The first of these irradiation tests, designated AGR-1, began in ATR in December of 2006 and ended in November 2009. This experiment was primarily to act as a shakedown test of the multi-capsule test train design and to provide early data on fuel performance that would be used in fuel fabrication process development. AGR-1 fuel kernels were produced on an engineering scale, but the TRISO coatings and cylindrical fuel compacts were fabricated on a laboratory scale. The AGR-1 PIE was completed and the final report was published in 2015. The second irradiation test, AGR-2, started in ATR in June 2010 and ended in October 2013. The AGR-2 irradiation test was designed to provide fuel performance data for coated particles fabricated on an engineering-scale pilot line using a coater with an internal chamber diameter of 150 mm (6 in.). The final PIE report was published in 2021. AGR-3/4, a single irradiation that combined what were originally conceived as the third and fourth tests, was to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport. PIE of the AGR-3/4 experiment is still in progress as of this writing. The subject of this report is AGR-5/6/7, the final qualification test of AGR TRISO fuel made entirely at the engineering scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data

Using the BISON fuel-performance code, simulations were conducted of an automated process to read initial and operating conditions from the Pacific Northwest National Laboratory (PNNL) database and reports, which contain metallic-fuel data from the Fast Flux Test Facility (FFTF) MFF Experiments. This work builds on previous modeling efforts involving 1977 EBR-II metallic fuel pins from experiments. Coupling the FFTF PNNL reports to BISON allowed for all 338 pins from MFF-3 and MFF-5 campaigns to be simulated. Each BISON simulation contains unique power and flux histories, axial power and flux profiles, and coolant-channel flow rates. Fission-gas release (FGR), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON and compared to available post-irradiation examination (PIE) data. Cladding profilometry, FGR, and fuel axial swelling simulation results for full-length MFF metallic pins were found to be in agreement with PIE measurements using FFTF physics and models used previously for EBR-II simulations. The main two peaks observed within the cladding profilometry were able to be simulated, with fuel-cladding mechanical interaction (FCMI), fuel-cladding chemical interaction (FCCI), and thermal and irradiation-induced creep being the cause. A U-Pu-Zr hot-pressing model was included in this work to allow pore collapse within the fuel matrix. This allowed better agreement between BISON-simulated cladding profilometry and PIE measurements for the peak caused by FCMI. This work shows that metallic fuel models used to accurately represent fuel performance for smaller EBR-II pins may be used for full-length metallic fuel, such as FFTF MFF assemblies and the Versatile Test Reactor (VTR). As new material models and PIE measurements become available, FFTF MFF assessment cases will be reassessed to further BISON model development.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Initial demonstration of automated fuel performance modeling with 1977 EBR-II metallic fuel pins using BISON code with FIPD and IMIS databases

Using the BISON fuel performance code, simulations were conducted using an automated process to read initial and operating conditions from the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor materials information system (IMIS) database, which contains metallic fuel data from the Experimental Breeder Reactor-II (EBR-II). This work demonstrates use of an integrated framework to access the vast majority of EBR-II experimental fuel pin data to support rapid development of fuel performance models for next-generation metallic fuel systems. With this capability, validation for fuel qualification can be performed rapidly. Between IMIS and FIPD, there is enough information to conduct 1977 unique EBR-II metallic fuel pin histories from 24 different experiments, at varying levels of detail between the two databases. Each of these histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures. Fission gas release (FGR), cumulative damage fraction (CDF), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON. The results were compared to post-irradiation examination (PIE) results for the initial demonstration of automated BISON modeling. BISON simulations conducted with IMIS and FIPD were in rough agreement with PIE measurements and calculations. Cladding profilometry, FGR, and fuel axial swelling were found to be in rough agreement with PIE measurements, depending on the physics used within the BISON input files. Here, the mechanical contact solver chosen was found to significantly impact axial fuel swelling and cladding strain predictions. CDF values were assessed to see whether pin failure may have been predicted (CDF ≥ 1). This work suggests that continued development of an automated tool for BISON should focus on inclusion of the Fast Flux Test Facility (FFTF) experimental data for a larger database for metallic fuel, improved physical models to better capture fuel performance, such as fuel-cladding interactions, and a more detailed comparison with available PIE data to further the BISON model development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Enhancement of Uranium Ionization Efficiencies Using Zn-MOF-74 Derived Nanoporous Ion Emitters for Thermal Ionization Mass Spectrometry

The recent introduction of nano-porous ion emitters (nano-PIEs) formed from metal organic frameworks (MOFs) has demonstrated the potential to enhance sensitivity for thermal ionization mass spectrometry (TIMS). Nano-PIEs take advantage of the parent MOF’s chemical and structural tunability to form scaffolds for ion emitters. A study by Barpaga et al. 2023 on MOF-74 as the parent material found that with high volatility metals in their framework uranium sample utilization efficiency (SUE) increases by up to nine-fold (e.g., Zn-MOF-74) compared to that of the analyte on a bare filament (~0.05%). Here, in this study, we investigate the performance of Zn-MOF-74 to maximize uranium efficiencies at the trace level (= 10 -12 g) by altering the parent MOF morphology and chemistry (i.e., MOF crystal sizes and thermal degradation) and optimizing its integration with TIMS (i.e., MOF mass on a filament and ramp conditions). We observed improvement in SUE up to 20 times (≤1.0%) that of a bare filament load when nano-PIEs derived from nanocrystals of Zn-MOF-74 were heated under a specific current ramp condition. This demonstrates that rates of nano-PIE structural collapse during TIMS analysis and the subsequently formed nanomaterials (and their features) can be tuned to control analyte ionization.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2026

Dual Representations and H ∞ -Optimal Control of Partial Differential Equations

We consider H ∞ -optimal state-feedback control of the class of linear Partial Differential Equations (PDEs) which admit a Partial Integral Equation (PIE) representation. While linear matrix inequalities are commonly used for optimal control of Ordinary Differential Equations (ODEs), the absence of a universal state-space representation and suitable dual form prevents such methods from being applied to optimal control of PDEs. Specifically, for ODEs, the controller synthesis problem is defined in state-space, and duality is used to resolve the bilinearity of that synthesis problem. Recently, the PIE representation was proposed as a universal state-space representation for linear PDE systems. In this paper, we show that any PDE system represented by a PIE admits a dual PIE with identical stability and I/O properties. This result allows us to reformulate the stabilizing and optimal state-feedback control problems as convex optimization over the cone of positive Partial Integral (PI) operators. Operator inversion formulae then allow us to construct feedback gains for the original PDE system. The results are verified through application to several canonical problems in optimal control of PDEs and indicate the resulting bounds on H ∞ norm are not conservative.

42 ENGINEERING↗

DOE OSTI2022

Progress Report on SFR Metallic Fuel Data Qualification

This report summarizes the progress of SFR metallic fuel qualification related activities, which are focused on providing quality assurance relevant information applicable to experiments irradiated during the Integral Fast Reactor (IFR) program. A background of metallic fuel performance data and the associated databases, including the EBR-II Fuels Irradiation & Physics Database (FIPD) and Out-of-Pile Transient Database (OPTD), is included. The legacy data in the databases, including as-built, post-irradiation examination (PIE), operating parameters, and out-of-pile experiment data are introduced. The SFR metallic fuel Quality Assurance Program Plan (QAPP) and its implementation to qualify these legacy data is described in detail. Important PIE data QA documents and the specifications of four types of PIE measurements (contact profilometry, laser profilometry, neutron radiography and gamma scan) are provided. Examples of the implementation of the QAPP to qualify each of those types of PIE data are provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Progress Report on SFR Metallic Fuel Data Qualification

This report summarizes the progress of SFR metallic fuel qualification related activities, which are focused on providing quality assurance relevant information applicable to experiments irradiated during the Integral Fast Reactor (IFR) program. An overview of the metallic fuel performance data and the associated databases, including the EBR-II Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT Experimental Relational Database (TREXR) is included. The legacy data in the databases, including as-built, post-irradiation examination (PIE), operating parameters, and out-of-pile experiment post-test data are introduced. The SFR metallic fuel Quality Assurance Program Plan (QAPP) and its implementation to qualify these legacy data is described in detail. Important PIE data QA documents and the specifications of seven types of PIE measurements (contact profilometry, laser profilometry, neutron radiography, gamma scan, fission gas release fission gas chemistry, and metallography) are provided. Examples of the implementation of the QAPP to qualify each of those types of PIE data are provided.

Mo, Kun [Argonne National Laboratory (ANL), Argonn↗

DOE OSTI2023

Evaluation of BISON metallic fuel performance modeling against experimental measurements within FIPD and IMIS databases

Simulations were conducted using the BISON fuel performance code on an automated process to read initial and operating conditions from two databases—the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor Materials Information System (IMIS) database. These databases contain metallic fuel data from the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). The work demonstrates use of an integrated framework to access EBR-II fuel pin data for evaluating fuel performance models contained within BISON to predict fuel performance of next-generation metallic fuel systems. Between IMIS and FIPD, there is enough information to conduct 1,977 unique EBR-II metallic fuel pin histories from 29 different experiments, and 338 pins from FFTF MFF-3 and MFF-5 with varying levels of details between the two databases. Each of these fuel performance histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures, and new model developments in BISON since the initial demonstration of this integrated framework. Fission gas release (FGR), cumulative damage fraction, fuel axial swelling, FCCI wastage thickness, cladding profilometry, and burnup were all simulated in BISON and compared to post-irradiation examination (PIE) results to evaluate BISON fuel performance modeling. Implementation of new fuel performance models into a generic BISON input file coupled with IMIS and FIPD yielded results with a better representation of physics than the initial evaluation of the integrated framework. Cladding profilometry, FGR, and fuel axial swelling were found to be in good agreement with PIE measurements for most of the pins simulated. The chosen mechanical contact solver was found to significantly impact the axial fuel swelling and cladding strain predictions when used in conjunction with the U-Pu-Zr hot-pressing model since it bound the fuel to prevent further swelling and increased hydrostatic stresses. This work suggests that fuel performance modeling in BISON under steady-state conditions represents the PIE data well and should be reassessed when new PIE data become available in IMIS and FIPD databases and when improved physical models to better capture fuel performance are added to BISON.

Paaren, Kyle M.↗

DOE OSTI2020

Effects of Neutron Irradiation on the Micro/Nano Scale Structure and Fission Product Distribution of TRISO Coated Particle Fuel Kernels from AGR Experiments

The Advanced Reactor Technologies (ART) Program at Idaho National Laboratory (INL) includes the Advanced Gas Reactor (AGR) fuel development and qualification program that consists of fuel fabrication, experiment irradiations, post-irradiation examination (PIE) and safety testing to assess tristructural isotropic (TRISO) fuel performance during normal irradiation and under potential accident conditions. Advanced microscopy work on selected AGR-1 and AGR-2 unirradiated and irradiated fuel specimens is performed as part of the PIE effort. PIE work on fuel from the first experiment irradiation, AGR-1, began at INL in April 2010, and AGR-2 PIE began at INL in July 2014. This work scope includes University of Florida (UF) performing advanced electron microscopy examination and analysis using facilities at the Center for Advanced Energy Studies (CAES), INL or UF’s electron microscopy facilities. Electron microscopic examination and analysis may include scanning transmission electron microscopy (STEM), transmission electron microscopy (TEM), selected area diffraction (SAD), electron energy loss spectroscopy (EELS), electron dispersive spectroscopy (EDS) and atom probe tomography (APT) on the fuel kernels of TRISO-coated fuel particles. The electron microscopy lamellae will be provided by INL and will be available at CAES or Irradiated Materials Characterization Laboratory (IMCL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Advanced Characterization Capabilities for Nuclear Materials via Nuclear Science User Facilities (NSUF)

Advanced post-irradiation examination (PIE) techniques are required to design new or improved nuclear materials, characterize, and understand in-core behavior of fuel and materials, and support the qualification of new reactor materials. The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear science user facility. NSUF provides researchers access to PIE capabilities at Idaho National Laboratory and at a diverse mix of university, national laboratory and industrial partner institutions. The PIE capabilities include novel destructive and non-destructive techniques for radiation damage characterization, such as advanced diffraction techniques (X-ray, electron, or neutron) coupled to extreme environments; in-situ observation of microstructural evolution under irradiation; in-situ irradiation to monitor corrosive attack in coolant environments; in-situ irradiation and mechanical testing; and test methods for synergistic effects of superimposed extreme environments (temperature, irradiation, stress, corrosion) on materials behaviors. This talk will provide an overview of NSUF PIE capabilities.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance

Here, the development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.

BISON↗