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Nuclide inventory characterization of EBR-II MOX fuel test pins
Characterization of 3 mixed oxide (MOX) fuel pins from the Experimental Breeder Reactor II (EBR-II) SPA-2/-2B experiment program was performed in support of the Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) joint project between Idaho National Laboratory (INL) and the Japan Atomic Energy Agency (JAEA). The characterization efforts were performed to enable transient testing in the Transient Reactor Test (TREAT) facility at INL and will include post irradiation examination (PIE) and other analyses. MOX fuel pins UW02009, 02011, and 06024 were historically irradiated from 1989 to 1994 for a total of 822.42 effective full power days (EFPDs) and peak burnup of approximately 14.3 at.% (atomic percent of heavy metal atoms, ∼134.4 GWd/t). They had since been placed in storage until a future use was identified. Characterization of the fission products and activation of each fuel pin was necessary to enable removal from storage and subsequent irradiation testing or PIE activities. The initial fuel pin geometry and composition were established using historical documentation. Irradiation history details were utilized to compute irradiated and decayed fuel pin masses, atomic densities, and activities. A pellet-wise axial neutron flux was also computed to assist in evaluation of the axial distribution of {sup 235}U and {sup 239}Pu, per pellet, within each of the test fuel pins. The results computed herein supported experiment design for advanced high-burnup MOX fuel designs under slow transient overpower (MOXTOP) conditions, which was Phase I of the ARES project. (authors)
Transient flight flutter test of a wing with tip tanks
Wing flutter was encountered during flight testing of the F2H-2 airplane with full wing tip tanks. As a result, more refined theoretical analysis as well as flight flutter tests were initated to establish corrective measures and to experimentally verify the stability of the improved system. The results from the flight flutter tests, utilizing the transient response technique, are presented. The method of excitation consisted of abrupt deflections of the ailerons resulting from stick bangs and data were measured by wing tip accelerometers. A comparison of the results with theoretical predictions is presented and indicates that reasonably good correlation was obtained. The influence on wing flutter of tip tank fuel transfer cycle, which was incorporated to control the center of gravity range of the tank during defueling, is indicated by the measured results and compared with the theory. The final configuration utilized a transfer cycle which was proven stable as a result of flight flutter testing. It is concluded that transient response measurements resulting from stick bangs provide a reasonably reliable and safe technique of flight flutter testing for wings with external tanks or heavy stores.
Stability Test for Transient-Temperature Calculations
Graphical test helps assure numerical stability of calculations of transient temperature or diffusion in composite medium. Rectangular grid forms basis of two-dimensional finite-difference model for heat conduction or other diffusion like phenomena. Model enables calculation of transient heat transfer among up to four different materials that meet at grid point.
The Fuel Motion Monitoring System at TREAT - Current Status and Future Plans
An important component of the United States Nuclear Fuel Safety Transient Testing Program, the Fuel Motion Monitoring System (FMMS) at Idaho National Laboratory's Transient Reactor Test Facility (TREAT) is fast-neutron hodoscope capable of imaging the location, movement, and relocation of nuclear fuel experiments under simulated transient accident conditions. The FMMS was refurbished in parallel with the TREAT restart program starting in 2014, restoring 96-channels of fast-neutron detection. Since returning to operation in 2017 the FMMS has supported many fuel safety experiments supporting accident tolerant fuel development, light-water reactor safety, space thermal nuclear propulsion fuel development, and advanced reactor research and development. In Phase 2 of the FMMS restoration. work is now under way to expand the FMMS' field-of-view by adding an additional 96 channels of fast neutron detectors to the system's hodoscope, along with an expanded data acquisition system and associated transient timing electronics. An overview of the FMSS system and its fast-neutron detectors will be presented along with examples of current FMMS imaging performance and associated information.
Feasibility of Power Ramp Testing in the Advanced Test Reactor
For decades, the Halden Boiling Water Reactor (HBWR) in Norway was an international resource for assessing nuclear fuels and materials behavior and its unexpected shutdown in 2018 represented a significant loss in experimental capability for prototypical irradiation testing. In the aftermath of the closure, a study was performed to assess capability gaps related to the Accident Tolerant Fuels (ATF) program. It was concluded that the primary capability gaps left by the closure of the HBWR were the loss of prototypic in-pile Light Water Reactor (LWR) loops that provide operational transient testing and in-pile Loss-of-Coolant Accident (LOCA) testing capabilities. The study also concluded that the Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) likely have the necessary key capabilities to absorb the breadth of the HBWR mission gaps related to the ATF program, but additional investments in experimental infrastructure were needed. One such investment is the design and installation of additional pressurized water loops in the medium-I positions of the ATR, hereafter referred to as I-Loops, to support power ramp testing and testing in Boiling Water Reactor (BWR) conditions. This paper gives an overview of the planned ATR I-Loops and assesses the feasibility of performing power ramp testing inside such a loop.
Results of the CHF-SERTTA In-Pile Transient Boiling Experiments at TREAT
To advance the current understanding of heat transfer between nuclear fuel rod cladding and the surrounding coolant under transient scenarios involving critical heat flux (CHF), the CHF-Static Environment Rodlet Transient Test Apparatus (CHFSERTTA), an in-pile experiment, was developed for and irradiated in the Transient Reactor Test (TREAT) facility at Idaho National Laboratory. The CHF-SERTTA experiment consists of a hollow, borated stainless steel rod submerged in a static water pool. The rod, heated via the (n, a) reaction in boron-10, was designed for investigating in-pile transient heat transfer between the rod surface and the coolant, avoiding the complexity of including the heat transfer between the fuel, gas gap, and cladding in typical light-water reactor (LWR) fuel rods. The instrumentation package for CHF-SERTTA focused on capturing those data that best describe the heat transfer behaviour between the rod and coolant. This paper outlines the CHFSERTTA experiment design and presents the initial results of the experiment campaign.
Structure and Process of Managing the UO2 Dry In-Pile Fracture Test Irradiation Experiment
The management and structure of a novel research and development (R&D) project can be hard to navigate. This is especially true when it involves nuclear fuel. Familiarity with institutional safety and procedural requirements, a well defined scope, and accurate cost estimates can all help ensure successful project management from start to finish. The Dry In-Pile Fracture Test (DRIFT) experiments performed at the Transient Test Reactor Facility are an excellent example of managing an R&D project within its scope and budget constraints. The objective of the DRIFT test was to develop fracture propagation data in a manner consistent with light-water reactors to validate and improve the fracture propagation models of uranium oxide fuels in the MOOSE (Multiphysics Object-Oriented Simulation Environment)-BISON-Marmot code framework [1].
Mo-100 Disk Structural Integrity Tests II
This report is a continuation of the work performed in the report, Mo-100 Disk Structural Integrity Tests I [1]. The NorthStar pressed and sintered 29 mm diameter, 0.5 mm thick Mo-100 disks will be subjected to large temperature gradients when in-beam during production. These large temperature gradients cause high stress and deformation within the disks that could prove detrimental to the target assembly’s cooling performance. Disk fracture is also a possibility that must be understood. Therefore, disk structural integrity tests were performed to understand the thermal induced behavior of these disks. In order to replicate the in-beam conditions, the Mo-100 disks were subjected to large temperature gradients using a 30kW induction heater and coil. A series of transient tests were conducted in a similar fashion as the previous work, as well as an attempt to create steady state conditions for more accurate temperature gradients. The transient thermal test setup and results are described in this report.
Development Testing and Analysis of the Integrated Gateway-ESPRIT Bipropellant Refuelling System
The Gateway will be humanity’s first space station orbiting the Moon completed by NASA in partnership with ESA and other US and international partners. The Gateway will provide critical support to sustainable human exploration on the moon through the Artemis program. To enable the Lunar Gateway to complete its mission, on orbit refuelling is essential. The ESPRIT Refuelling Module (ERM) will provide the capability to transfer propellants, MMH and MON-3, through the Habitation and Logistics Outpost (HALO) to the Power and Propulsion Element (PPE). The transfer of these propellants carries known risks and hazards. These hazards include overpressure of the propellant lines during priming sequences between modules and during refuelling pause operations. To support the early system development and to mitigate these risks, a collaborative test program between NASA, ESA and Thales Alenia Space was completed at the Thales Alenia Space test facility in Harwell, UK. This test program integrated fluidic breadboards of the ERM, HALO and PPE modules. The objectives of the test program were to demonstrate and characterise critical performance and transient operations, inform refuelling concept of operations and to calibrate and validate numerical models of the refuelling subsystem in EcosimPro. To support the completion of this final objective a detailed model of the integrated breadboard was developed in EcosimPro and key steady-state and transient test cases were simulated. As an industry first, a National Institute of Standards and Technology (NIST) database correlation for Hydrofluoroether (HFE-7000) was used as a mixed oxides of nitrogen (MON-3) simulant with EcosimPro and European Space Propulsion System Simulation (ESPSS) libraries to result in a more accurate analysis for transient phenomenon such as priming. The test and simulation data showed good agreement validating the model for further system analysis as the ERM design progresses.
FUEL PERFORMANCE STUDIES AT IDAHO NATIONAL LABORATORY MAKING USE OF THE BYRON FUEL SHIPMENT
In December of 2023 a shipment of commercially irradiated fuel rods from the Byron Generating Station in Illinois was successfully shipped to the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The make-up of the rods includes a mix of cladding types from traditional ZirloTM, advanced zirconium alloys, and chrome coated ZirloTM. Burnups range from regular end of life values to over 70 GWd/MTU rod average. The R&D plan for the rods involves multiple projects from developing licensing data for new claddings to integral transient tests to support burnup extension efforts in the United States. The R&D began in early 2024 with the nondestructive examinations of the rods after which they will be sectioned for microscopy, mechanical testing, and analytical chemistry. Additionally, many rod segments will be refabricated into new test pins and inserted into a static water capsule for integral Reactivity Initiated Accident (RIA) and Loss of Coolant Accident (LOCA) testing at the TREAT reactor.
Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility
The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.
Preliminary results of testing a single tube potassium boiler for the advanced Rankine system
Performance and transient testing of single tube potassium boiler for Rankine nuclear power plant system
CFD Simulations of Supersonic Highly Swirling Flow Exiting a Turbine Vane Row Compared with Experimental Observations
Marshall Space Flight Center (MSFC) has been heavily involved in developing the J2-X engine. The Center has been testing a Work Horse Gas Generator (WHGG) to supply gas products to J2-X turbine components at realistic flight-like operating conditions. Three-dimensional time accurate CFD simulations and analytical fluid analysis have been performed to support WHGG tests at MSFC. The general purpose CFD program LOCI/Chem was utilized to simulate flow of products from the WHGG through a turbine manifold, a stationary row of turbine vanes, into a Can and orifice assembly used to control the back pressure at the turbine vane row and finally through an aspirator plate and flame bucket. Simulations showed that supersonic swirling flow downstream of the turbine imparted a much higher pressure on the Can wall than expected for a non-swirling flow. This result was verified by developing an analytical model that predicts wall pressure due to swirling flow. The CFD simulations predicted that the higher downstream pressure would cause the pressure drop across the nozzle row to be approximately half the value of the test objective. With CFD support, a redesign of the Can orifice and aspirator plate was performed. WHGG experimental results and observations compared well with pre-test and post-test CFD simulations. CFD simulations for both quasi-static and transient test conditions correctly predicted the pressure environment downstream of the turbine row and the behavior of the gas generator product plume as it exited the WHGG test article, impacted the flame bucket and interacted with the external environment.
High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage
Chromium-coated zirconium alloys are one of the promising candidates for accident-tolerant fuel cladding (ATF) tubes for light water reactors (LWRs). In this study, the high temperature oxidation and degradation of two types of Cr coatings (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on prototype rod segment samples filled with ZrO 2 pellets and tightly sealed with welded end caps. Isothermal tests at 1100 and 1200 °C were terminated by quenching with water; transient tests were performed up to 1500–1600 °C until complete coating failure. The positive effect of both types of Cr coatings was observed in all tests. Pre-damaged specimens showed only locally increased oxidation of the scratched zone, but no negative effects on the adjacent Cr coating. Finally, the behavior of the two types of coatings is compared and the degradation mechanisms are discussed based on hydrogen release data, extensive metallographic post-test examinations, and the current state of global research.
Code validation of SAM using natural-circulation experimental data from the compact integral effects test (CIET) facility
The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) experimental loop. SAM is a modern system analysis code being developed at Argonne National Laboratory for safety analysis of designs for advanced non-light water reactors (non-LWRs), such as sodium-cooled fast reactors, high-temperature gas-cooled reactors, and fluoride salt-cooled high-temperature reactors (FHRs). To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments highly relevant to these reactor concepts. Additionally, the CIET facility, which was designed to reproduce the thermal-hydraulics response of FHRs under both forced- and natural-circulation conditions, has been identified and selected as one of the benchmark test facilities for SAM code validation. In this study, two sets of available CIET tests were selected for SAM code validation purposes, namely, power step change transient tests and steady-state natural-circulation tests. For all selected tests, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.
Testing in Support of Fission Surface Power System Qualification
The strategy for qualifying a FSP system could have a significant programmatic impact. The US has not qualified a space fission power system since launch of the SNAP-10A in 1965. This paper explores cost-effective options for obtaining data that would be needed for flight qualification of a fission system. Qualification data could be obtained from both nuclear and non-nuclear testing. The ability to perform highly realistic nonnuclear testing has advanced significantly throughout the past four decades. Instrumented thermal simulators were developed during the 1970s and 1980s to assist in the development, operation, and assessment of terrestrial fission systems. Instrumented thermal simulators optimized for assisting in the development, operation, and assessment of modern FSP systems have been under development (and utilized) since 1998. These thermal simulators enable heat from fission to be closely mimicked (axial power profile, radial power profile, temperature, heat flux, etc.) and extensive data to be taken from the core region. For transient testing, pin power during a transient is calculated based on the reactivity feedback that would occur given measured values of test article temperature and/or dimensional changes. The reactivity feedback coefficients needed for the test are either calculated or measured using cold/warm zero-power criticals. In this way non-nuclear testing can be used to provide very realistic information related to nuclear operation. Non-nuclear testing can be used at all levels, including component, subsystem, and integrated system testing. FSP fuels and materials are typically chosen to ensure very high confidence in operation at design burnups, fluences, and temperatures. However, facilities exist (e.g. ATR, HFIR) for affordably performing in-pile fuel and materials irradiations, if such testing is desired. Ex-core materials and components (such as alternator materials, control drum drives, etc.) could be irradiated in university or DOE reactors to ensure adequate radiation resistance. Facilities also exist for performing warm and cold zero-power criticals.
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.