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At least 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Performance Analysis and Simulaion of the Hydraulic Scram System in TREAT Reactor

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) serves a vital role in nuclear fuel safety research, enabling transient experiments that simulate reactivity excursions and accident scenarios. Central to these operations is the transient control rod drive system (TCRDS), which drives rapid motion of the transient control rods such that TREAT can simulate rapid power changes typical of reactor accidents. The reliability and performance of this system are critical for protecting both fuel specimens and reactor infrastructure. This study presents the initial phase of a two-year investigation into the dynamics and reliability of the TREAT hydraulic TCRDS. Conducted in collaboration with INL, the research employs a combined computational and experimental approach to analyze the system's response time, pressure transients, and potential failure modes. Emphasis is placed on understanding how fluid characteristics influence the TCRDS’s ability to achieve both rapid power changes and mechanical stability. The TRDS and the skid that powers it will be analyzed throughout this investigation. Computational modeling using computational fluid dynamics (CFD) will simulate the hydraulic response under varying conditions. In parallel, experimental testing planned at INL will validate these models and capture key performance metrics. This paper outlines the system design, analytical framework, and modeling strategies that form the foundation for later testing. Ultimately, this work aims to support improvements to the TCRDS’s design and reliability, contributing to the broader goal of enhancing nuclear fuel safety and sustaining TREAT’s mission as a premier nuclear fuel test facility.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2012

AGR-2 Data Qualification Report for ATR Cycles 149B, 150A, 150B, 151A, and 151B

This report provides the data qualification status of AGR-2 fuel irradiation experimental data from Advanced Test Reactor (ATR) cycles 149B, 150A, 150B, 151A, and 151B), as recorded in the Nuclear Data Management and Analysis System (NDMAS). The AGR-2 data streams addressed include thermocouple temperatures, sweep gas data (flow rate, pressure, and moisture content), and fission product monitoring system (FPMS) data for each of the six capsules in the experiment. A total of 3,307,500 5-minute thermocouple and sweep gas data records were received and processed by NDMAS for this period. There are no AGR-2 data for cycle 150A because the experiment was removed from the reactor. Of these data, 82.2% were determined to be Qualified based on NDMAS accuracy testing and data validity assessment. There were 450,557 Failed temperature records due to thermocouple failures, and 138,528 Failed gas flow records due to gas flow cross-talk and leakage problems that occurred in the capsules after cycle 150A. For FPMS data, NDMAS received and processed preliminary release rate and release-to-birth rate ratio (R/B) data for the first three reactor cycles (cycles 149B, 150B, and 151B). This data consists of 45,983 release rate records and 45,235 R/B records for the 12 radionuclides reported. The qualification status of these FPMS data has been set to In Process until receipt of QA-approved data generator reports. All of the above data have been processed and tested using a SAS®-based enterprise application software system, stored in a secure Structured Query Language database, and made available on the NDMAS Web portal (http://ndmas.inl.gov) for both internal and external VHTR project participants.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

University of Illinois Temperature Sensors

This document summarizes background information and presents results related to temperature measurements in the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) University of Illinois Project 29609 irradiation. The objective of this test was to assess the radiation performance of ferritic alloys for advanced reactor applications. The FeCr-based alloy system is considered the lead alloy system for a variety of advanced reactor components and applications. Irradiations of FeCr alloy samples were performed using the Hydraulic Shuttle Irradiation System (HSIS) in the B-7 position and in a static capsule in the A-11 position of the ATR.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

High-Fidelity Modeling of Fuel-To-Coolant Thermomechanical Transport Behaviors Under Transient Conditions

This report summarizes the work completed under NEUP project number 21-24006. The objectives of this project are to advance the high-fidelity modeling capabilities and important phenomena that is important for high-burnup UO 2 and accident tolerant fuels (ATF) during transient conditions. Accurate modeling of the time-dependent phenomena that impact material performance must be used to determine the figures of merit and safety margin. Phenomena such as fuel fragmentation, cladding oxidation, pellet-clad interaction, clad ballooning, and clad rupture are examples that pose challenges to modeling during these transients. This project focused on the development of high-fidelity tightly coupled multiphysics models that can capture the time-dependent material response and associated thermal hydraulic conditions during these events. These models can then be validated against existing separate effects tests and in-pile integral experiments and will be used to model Transient Reactor Test facility (TREAT) loss-of-coolant accidents (LOCA) experiments. To achieve the project objective, we used a combination of NEAMS and NRC codes to model various LOCA test sets for the separate effects and in-pile integral experiments. BlueCRAB tool set, which can accurately predict material response at a sub-fuel pin level, as well as modeling the entire reactor system response to these events. Fuel performance was modeled using BISON (various versions) and FAST (version 1.2.1). BISON and FAST can model on a sub-fuel pin level the fuel performance under transient conditions. Both have simplified thermal hydraulic models that are capable of providing basic coolant boundary conditions. To better capture the thermomechanical interaction between the fuel, clad, and coolant, more sophisticated thermal hydraulic models are necessary.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of Advanced Instrumentation for Transient Testing

A fuel safety research program centered on in-pile transient testing experiments is being developed to support assessment and qualification of advanced nuclear fuel systems using the recently restarted Transient Reactor Test (TREAT) facility at the Idaho National Laboratory (INL). While resumption of transient testing at TREAT is crucial to enable these programs, full recovery and cutting-edge transient testing capability also requires a well-coordinated and innovative instrumentation development and qualification program to support near-term and future objectives. This paper summarizes the experimental approach of transient testing to focus on measuring the response of nuclear fuel to off-normal (or power-cooling mismatch) conditions for modern and advanced reactor environments requiring capabilities extending over wide measurement and environment conditions. It also highlights unique attributes of transient testing of importance to in-pile instruments including relatively low total neutron fluence, high gamma heating, and the need for well-defined and possibly short time response. Historical approaches to instrumentation for transient testing are also reviewed to provide context to the modern instrument strategy. The instrumentation needs of a modern transient testing are detailed while summarizing several ongoing R&D activities are supporting the development of state-of-the-art and ‘advanced’ measurement technologies that will provide baseline capability for Light Water Reactor (LWR) and Sodium-Cooled Fast Reactor (SFR) experiment objectives, while extending to other advanced reactor needs and advanced sensing technology opportunities. Examples of specific sensors planned for near-term deployment with ongoing development includes prompt-response self-powered neutron detectors (SPND), miniature fission chambers, optical-fiber-coupled infrared pyrometer, cladding surface thermocouples, electrical-impedance based boiling detector, and Linear Variable Differential Transformer (LVDT) based sensors for fuel elongation and pressure measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

AGR-2 Irradiation Experiment Fission Product Mass Balance

The Advanced Gas Reactor (AGR) program is in the process of qualifying tristructural isotropic (TRISO)-coated uranium carbide/uranium oxide (UCO) fuel for use in high-temperature gas-cooled reactors (HTGRs) in the United States. The second AGR irradiation (AGR-2) started in the Advanced Test Reactor (ATR) in June 2010 and was completed in October 2013. The first AGR irradiation (AGR-1) tested fuel produced on laboratory scale equipment, but the AGR-2 TRISO-coated fuel particles were produced on an industrial scale and then overcoated and compacted into cylindrical compacts using laboratory-scale processes. Both UCO and UO2 fuels were irradiated in AGR-2 so that their performance could be compared. The UCO fuel was in Capsules 2, 5, and 6, and the UO2 fuel was in Capsule 3. (Capsules 1 and 4 were French and South African fuel, respectively. These fuels will not be discussed.) UCO burnups ranged from 7.3 to 13.2 % fissions per initial metal atom (FIMA), and compact time-averaged, volume-averaged (TAVA) irradiation temperatures ranged from about 1000°C to 1300°C. UO2 compact burnups ranged from 9.0 to 10.7 % FIMA, and compact TAVA temperatures ranged from about 1000 to 1060°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

MITR DDE End Fitting Structural Rigidity Analysis

As a part of the design of the MITR Design Demonstration Element (DDE), a new design for the end fittings must be considered to meet specific experiment needs, e.g., compatibility with the geometrical restrictions of the DDE testing location (BR2 test reactor located in Mol, Belgium), and to allow for proper inspection of these test elements during the experiments. The design of the MITR DDE end fittings is substantially different from the design of the end fittings of the MITR LEU fuel elements, with the DDE end fittings considerably smaller in overall size and weight. Consequently, the stiffness of the end fittings as well as of the entire fuel element is affected. The end fittings in the MITR LEU fuel element are designed primarily for proper handling and flow distribution. The stiffness of the end fittings also constitutes a fraction of the total stiffness of the entire fuel element. It is debatable to what extent the end fittings provide the structural rigidity necessary for the fuel element to properly respond to potential loadings. It is presumed that the bulk of the overall stiffness of the element is derived from swaging multiple fuel plates into the side plates. However, the contribution of stiffness from various components of the MITR LEU fuel element and MITR DDE has not been determined so far. The primary objective of the work presented in this report was to assess the extent to which the stiffness of the end fitting in the MITR DDE contributes to the stiffness of the entire fuel element, and how it compares to the equivalent stiffness of the end fitting in the MITR LEU fuel element. Several simplified load cases were analyzed for that purpose using the geometries of the MITR LEU fuel element, MITR DDE, as well as that of the MITR DDE without the top end fitting.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Contributions to the mechanistic understanding of the microstructural evolution in irradiated U-Mo dispersion fuel

Here, advanced microstructural characterization techniques, such as scanning electron microscopy (SEM) and scanning transmission electron microscopy - energy dispersive x-ray spectroscopy (STEM-EDS), were used to interpret the fuel microstructure evolution and fission products behavior in U-Mo dispersion fuel irradiated in the Advanced Test Reactor (ATR) as part of the European Mini-Plate Irradiation Experiment (EMPIrE) test. The larger as-fabricated fuel grain size achieved by heat-treating the U-Mo powder resulted in slower high burnup structure (HBS) development and reduced fission gas porosity. Slower HBS kinetics was observed at the fuel kernels’ periphery, which contained smaller and less fission gas bubbles at all fission densities (FDs) investigated and was attributed to a locally reduced damage density and fission products concentration, as corroborated with Monte Carlo simulations. The non-refined grains at the fuel kernel periphery hosted a perfectly ordered fission Gas Bubble Superlattice (GBS) up to 6.3 × 10 21 fissions/cm 3 . Nano-scale STEM-EDS analysis presented in this study provided useful information on the GBS characteristic morphology and evolution in U-Mo fuel. The concentration of fission gas in the GBS progressively increased with FD, pointing to an evolution of the nanobubble pressure status with irradiation. A possible connection between the GBS collapse and HBS onset is proposed for which there exists a threshold in the misorientation of the refined sub-grains above which the GBS stability during irradiation is no longer preserved, resulting in the GBS collapse.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Test Plan Proposal for an International Joint Project for Transient Testing of Fast Reactor Fuels

EXECUTIVE SUMMARY An international joint project for transient testing of fast reactor fuels is proposed. Input from several countries is provided regarding their advanced/fast reactor programs, and the objectives and methods are given based on both historical projects and modern tools and methods. The full package would include irradiation testing using TREAT, use of existing facilities for non destructive exams/evaluations (NDE) and post irradiation exams (PIE), and integration of current state-of-the-art modeling and simulation (M&S) tools to aid in both experiment design, and prediction of experimental results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1981

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Tong, H.↗
DOE OSTI2017

Flux in a 1D-Reactor Cell by Response Matrix

Legacy computational reactor physics software tools and protocols currently used for support of the Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. Thus in late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity transport theory based computational software and models to replace the legacy methods, with appropriate (V&V) protocols according to applicable national standards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Oklo and BWXT Short Decks

Two abstracts, one for each deck: Oklo is planning on producing fuel using spent EBR-II material and demonstrating an NRC licensed reactor. This presentation gives a high level overview of Oklo's progress in this work. The BWXT Advanced Nuclear Reactor Experiment 1 (BANR-1) is an irradiation experiment at INL's Advanced Test Reactor (ATR) that will irradiate novel Tristructural Isotropic (TRISO) fuel for potential future use by BWXT. This presentation gives a program overview and progress.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Combined TREAT-LOC & SATS Integral LOCA Experiment Plan

The Transient Reactor Test Facility (TREAT) loss-of-coolant (LOC) and highburnup (HBu) experiment series, along with the Severe Accident Test Station (SATS) HBu experiment series, are integral LOC accident (LOCA) experiments planned under the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program, which aims to support burnup extension needs by addressing identified R&D priorities in order to achieve an improved understanding of fuel fragmentation, relocation, and dispersal (FFRD) of HBu fuel during LOCA events. Priorities have been identified by the Electric Power Research Institute (EPRI)’s Collaborative Research on Advanced Fuel Technologies (CRAFT) Fuel Performance and Testing Technical Experts Group (FPTTEG). The data produced under this plan will be used to further validate and confirm existing models and inform future R&D and model development. The experimental program was specifically developed to address data gaps and opportunities identified via detailed review of the existing public knowledge base on LOCA FFRD, as well as reviewing specific experimental development activities regarding prototypic LOCA conditions for light-water reactor (LWR) systems. The test program relies on a unique combination of in- and out-of-pile experimental approaches to provide a clear tieback to the existing integral and semi-integral LOCA experiment database, using state-of-the-art facilities. More importantly, the program will systematically investigate the impacts of prototypic HBu fuel/cladding thermomechanical behaviors under postulated LWR LOCA conditions not yet fully investigated. These conditions correspond with prototypic decay-energy heatup (DEH) and stored-energy heatup (SEH) conditions. First, TREAT’s unique capability will enable the first evaluation of the impact of SEH conditions on HBu fuels. The test program will emphasize the development of an improved mechanistic understanding of key phenomena through independent experimental systems, development of a database to support fuel performance modeling tools, world-leading advanced materials characterization, and the most advanced approach to in situ diagnostics ever deployed to evaluate FFRD. The results will represent a significant leap forward in evaluating prototypic conditions and novel data to support modeling development and validation, as well as to inform the technical basis for LOCA-induced FFRD.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Qualification of CASMO5 / SIMULATE-3K against the SPERT-III E-core cold start-up experiments

SIMULATE-3K is a three-dimensional kinetic code applicable to LWR Reactivity Initiated Accidents. S3K has been used to calculate several international recognized benchmarks. However, the feedback models in the benchmark exercises are different from the feedback models that SIMULATE-3K uses for LWR reactors. For this reason, it is worth comparing the SIMULATE-3K capabilities for Reactivity Initiated Accidents against kinetic experiments. The Special Power Excursion Reactor Test III was a pressurized-water, nuclear-research facility constructed to analyze the reactor kinetic behavior under initial conditions similar to those of commercial LWRs. The SPERT III E-core resembles a PWR in terms of fuel type, moderator, coolant flow rate, and system pressure. The initial test conditions (power, core flow, system pressure, core inlet temperature) are representative of cold start-up, hot start-up, hot standby, and hot full power. The qualification of S3K against the SPERT III E-core measurements is an ongoing work at Studsvik. In this paper, the results for the 30 cold start-up tests are presented. The results show good agreement with the experiments for the reactivity initiated accident main parameters: peak power, energy release and compensated reactivity. Predicted and measured peak powers differ at most by 13%. Measured and predicted reactivity compensations at the time of the peak power differ less than 0.01 $. Predicted and measured energy release differ at most by 13%. All differences are within the experimental uncertainty. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

TREAT neutronics analysis and design support, Part I: Multi-SERTTA

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis of a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.73 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

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.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗