Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Transient Testing”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Status and Availability of OPTD, the Out-of-Pile Transient Database

The DOE Advanced Reactor Technology program has supported knowledge preservation efforts to recover and preserve fuel data from the US sodium cooled fast reactor (SFR) program. Databases have been established containing experimental data generated during the Integral Fast Reactor program (from in-pile experiments at EBR-II, FFTF, and TREAT) and during related outof-pile examinations of EBR-II-irradiated fuels. The information in these databases is essential to support further development and licensing activities for advanced fast reactor designs. OPTD, the Out-of-Pile Transient Database, is an organized, searchable archive of records describing out-ofpile furnace transient tests conducted on metallic fuels. It includes records of over 150 tests on irradiated fuel pins conducted in furnace apparatuses constructed in the Alpha-Gamma Hot Cell Facility at ANL. This report presents the content, functions, and accessibility of the Out-of-Pile Transient Database (OPTD). It describes the information presently available in the database, the web interface created to provide access to this information, the activities undertaken to build OPTD, and the enabling of online access to the database for DOE users. Plans for FY2021 and beyond are also included.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Out-of-Pile Testing and Instrumentation Transient Water Irradiation System

Current initiatives to increase the burnup of conventional nuclear fuels past the approximate 62 GWd/t limit have been spurred on by direct savings to refueling and waste storage. The technical justification for a new license limit requires extensive qualification through experimental testing. Unlike beginning-of-life fuels, high-burnup fuels are more susceptible to fuel fragmentation, relocation, and dispersal (FFRD), therefore more data is needed to characterize fuels under key accident scenarios. The Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory has developed a testing apparatus architecture to test fuels and claddings at prototypic conditions. The Transient Water Irradiation System (TWIST) is the latest iteration of a testing device capable of conducting loss of coolant accidents (LOCAs) in TREAT. The Out-of-Pile Testing and Instrumentation TWIST (OPTI-TWIST) is an electrically heated device that is analogous to TWIST. OPTI-TWIST allows for detailed instrumentation and thermal-hydraulic characterization. TWIST ultimately aims to conduct the most advanced in-situ diagnostics to evaluate FFRD in a prototypic LOCA. Moreover, it will explore the phenomenological bifurcation of a decay-energy heat up driven LOCA and a stored-energy heat up driven LOCA. The instrumentation suite includes conventional thermocouples and pressure transducers in addition to an electro impedance sensor, an acoustic emission sensor, an optical pressure sensor, and an optical pyrometer. Characterizing these instruments in OPTI-TWIST eliminates complications of irradiation effects while preserving extreme thermal-hydraulic conditions. Finally, benchmarking both devices to a thermal-hydraulic code like the Reactor Excursion and Leak Analysis Program (RELAP)5-3D provides a unique opportunity for iteration. Pre-test predictions and post-test interpretations inform the physical designs, operational procedures, test conditions, and instrumentation types and positions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Foreword Special issue on the Kilopower Project, Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Test

This special issue of Nuclear Technology contains full-length, peer-reviewed papers describing the design work leading up to and the results of the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) test. KRUSTY was the centerpiece of the National Aeronautics and Space Administration (NASA) Kilopower Project to design, build, and test a space nuclear reactor. This test was the first such test since the end of the Space Nuclear Auxiliary Power (SNAP) project at the end of the 1960s. In this issue, the introduction paper presents the goals of the Kilopower Project and the potential missions this reactor concept could serve in NASA. Kilopower was intended to serve both human exploration needs on planetary surfaces as well as science needs for deep-space exploration. The design work for the experiment by Poston and the power conversion development by Gibson present the pre-work required to perform the eventual KRUSTY test. A paper on regulatory analysis follows, to show the path used to gain approval of the proposed experiment. Then, the early zero-power critical experiments are presented by Sanchez and Grove. These experiments were essential data used to enhance model predictions prior to the high-temperature test. Next, Poston presents the three experiments (warm criticals) that increase the temperature in an incremental fashion prior to the final experiment. These experiments were used to achieve final regulatory approval of the final high-temperature experiment. The last paper by Poston presents the results of the steady-state and transient testing of the reactor at full power and at the design temperature. These results show that the reactor design and as-built experiment met all of the requirements that NASA had developed for the system. Any experiment of this magnitude was accomplished only because of the hard work and dedication of a large number of people at multiple institutions, including the NASA Glenn Research Center, NASA Marshall Space Flight Center, Y-12 National Security Site, Los Alamos National Laboratory, and the Nevada National Security Site prime contractor (formerly National Security Technologies and now the Mission Support and Test Services). The project was jointly funded by the Space Technology Mission Directorate at NASA and the Criticality Safety Program at the National Nuclear Security Administration (NNSA). A special thanks to Jerry McKamy and Angela Chambers at the NNSA and Lee Mason and Don Palac at NASA for their support and guidance. For the authors and multitude of staff that performed this work, the passion for space nuclear reactors was the key to making this experiment a reality.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Comparison of Subdivided Versus Full-Length Fuel Pin Burnup in the ATF-2 Safety Analysis

The Accident Tolerant Fuels (ATF) campaign, under the Fuel Cycle Research and Development (FCRD) Advanced Fuels program, was developed after the Fukushima Daiichi Nuclear Power Plant disaster in 2011. Under ATF, fuel and cladding concepts are to be developed and tested that may more readily withstand an accident similar to the one that occurred at Fukushima Daiichi. ATF is a combination of industry, universities, and national laboratories in an effort to develop advanced accident tolerant cladding and fuel concepts. The end goal of the ATF program is to insert a lead test rod (LTR) or lead test assembly (LTA) into an operating commercial power light water reactor by 2022. In order to meet this 2022 insertion goal, ATF is split into three separate phases: ATF-1, ATF-2, and ATF-3. The first phase of the ATF program, ATF-1, is the concept feasibility study and early irradiation testing. Feasibility, fabrication, and out-of-pile testing are happening at universities and laboratories all over the United States. Irradiation testing of these concepts is happening at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Early irradiation testing involves testing clad and/or fuel in drop-in capsules in the ATR. Currently, these capsules are in the relatively low-flux positions in the ATR, but some newer concepts and concepts that require higher burnup will be moved to the higher-flux positions later on during irradiation. Currently, Oak Ridge National Laboratory (ORNL), Westinghouse Electric Company, Areva, and General Electric (GE) have inserted capsules for ATF-1, although more concepts are planned. The second phase of ATF, ATF-2, is the focus of this paper. ATF-2 is the loop testing of fuel pins with new cladding and fuel concepts to simulate pressurized water reactor (PWR) and boiling water reactor (BWR) environments. ATF-2 will concurrently be happening at the ATR and at the Halden Reactor in Norway. The ATF-2 irradiation at the ATR will only simulate a PWR environment. ATF-2 will be inserted into the ATR pressurized water loop, Loop 2A, in the Center Flux Trap (CFT), where chemistry, temperature, and pressure can all be controlled separately from the primary coolant system and other loops in ATR. Loop 2A will be borated, pressurized to 2200 psi nominal, and have an average water temperature of 300 deg. C. A borated loop with fueled rodlets is extremely rare for Nuclear Energy experiments in the ATR, so ATF-2 will demonstrate new and unique capabilities for INL. ATF-3 is the testing of fresh and irradiated ATF concepts at the Transient Reactor Test (TREAT) Facility. The TREAT Reactor is planned for restart, re-establishing U.S. transient testing capabilities after nearly 20 years of non-operations. TREAT will test ATF concepts under transient conditions, such as a reactivity insertion accident. In addition to fresh fuel pins, TREAT will use from ATF2 and do transient testing on those irradiated pins. The information TREAT will provide to the vendors to support fuel licensing will be valuable, showing that these concepts do enhance the accident tolerance of the fuel in commercial reactors. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preliminary Results for the OECD/NEA Time Dependent Benchmark using Rattlesnake, Rattlesnake-IQS and TDKENO

One goal of the MAMMOTH M&S project is to validate the analysis capabilities within MAMMOTH. Historical data has shown limited value for validation of full three-dimensional (3D) multi-physics methods. Initial analysis considered the TREAT startup minimum critical core and one of the startup transient tests. At present, validation is focusing on measurements taken during the M8CAL test calibration series. These exercises will valuable in preliminary assessment of the ability of MAMMOTH to perform coupled multi-physics calculations; calculations performed to date are being used to validate the neutron transport solver Rattlesnake\cite{Rattlesnake} and the fuels performance code BISON. Other validation projects outside of TREAT are available for single-physics benchmarking. Because the transient solution capability of Rattlesnake is one of the key attributes that makes it unique for TREAT transient simulations, validation of the transient solution of Rattlesnake using other time dependent kinetics benchmarks has considerable value. The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently developed a computational benchmark for transient simulations. This benchmark considered both two-dimensional (2D) and 3D configurations for a total number of 26 different transients. All are negative reactivity insertions, typically returning to the critical state after some time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Exhaust particle characterization for lean and stoichiometric DI vehicles operating on ethanol-gasoline blends

Gasoline direct injection (GDI) engines can offer better fuel economy and higher performance over their port fuel-injected (PFI) counterparts, and are now appearing in increasingly more U.S. and European vehicles. Small displacement, turbocharged GDI engines are replacing large displacement engines, particularly in light-duty trucks and sport utility vehicles, in order for manufacturers to meet the U.S. fuel economy standards for 2016. Furthermore, lean-burn GDI engines can offer even higher fuel economy than stoichiometric GDI engines and have overcome challenges associated with cost-effective aftertreatment for NOx control. Along with changes in gasoline engine technology, fuel composition may increase in ethanol content beyond the current 10% due to the recent EPA waiver allowing 15% ethanol. In addition, the Renewable Fuels Standard passed as part of the 2007 Energy Independence and Security Act (EISA) mandates the use of biofuels in upcoming years. GDI engines are of environmental concern due to their high particulate matter (PM) emissions relative to port-fuel injected (PFI) gasoline vehicles; widespread market penetration of GDI vehicles may result in additional PM from mobile sources at a time when the diesel contribution is declining. In this study, we characterized particulate emissions from a European certified lean-burn GDI vehicle operating on ethanol-gasoline blends. Particle mass and particle number concentration emissions were measured for the Federal Test Procedure urban driving cycle (FTP 75) and the more aggressive US06 driving cycle. Particle number-size distributions and organic to elemental carbon ratios (OC/EC) were measured for 30 MPH and 80 MPH steady-state operation. In addition, particle number concentration was measured during wide open throttle accelerations (WOTs) and gradual accelerations representative of the FTP 75. Fuels included certification gasoline and 10% (E10) and 20% (E20) ethanol blends from the same supplier. The particle mass emissions were approximately 3 and 7 mg/mile for the FTP75 and US06, respectively, with lower emissions for the ethanol blends. The data are compared to a previous study on a U.S.-legal stoichiometric GDI vehicle operating on the same ethanol blends. The lean-burn GDI vehicle emitted a higher number of particles, but had an overall smaller average size. Particle number per mile decreased with increasing ethanol content for the transient tests. For the 30 and 80 mph tests, particle number concentration decreased with increasing ethanol content, although the shape of the particle size distribution remained the same. Engine-out OC/EC ratios were highest for the stoichiometric GDI vehicle with E20, but tailpipe OC/EC ratios were similar for all vehicles.

02 PETROLEUM↗
DOE OSTI2019

Core-to-Specimen Energy Coupling Results of the First Modern Fueled Experiments in TREAT

For this study, a series of experiments was conducted in the Transient Reactor Test (TREAT) Facility following a decades-long operational hiatus. A series of physics tests were then performed emphasizing relevant transients for Light Water Reactor (LWR) fuels research. A series of fueled experiments were then conducted using LWR-type sub-length specimens in inert gas capsules with the primary objective to quantify the ratio of nuclear heat deposited in the core to that deposited in the specimens. This type of data can be difficult to measure but is highly important in the design and interpretation of transient tests. Five fueled capsules were irradiated and energy coupling factors (which express the ratio of nuclear heat deposited per mass in the test specimen to that deposited in the reactor core in units of J/gUO 2 MJ TREAT ) were observed ranging from 2.0 to 2.2 over a wide range of transient energies with good agreement between measurement methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Heat Source Characterization in a TREAT Fuel Particle Using Coupled Neutronics-BCMC Calculations

This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. The macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Drop Testing of a Container for the Storage, Transport and Disposal of Intermediate Level Waste - 18078

Impact tests were performed, as part of a corresponding container's substantiation, during design development of a shielded container proposed for use at Sellafield for waste retrieval from the First Generation Magnox Storage Pond (FGMSP) in the UK. The mechanical test program comprised a 9 m free drop test onto an unyielding target in a container long lid edge down orientation at ambient conditions. Further, a 0,5 m free drop test onto a punch target was performed. Here, the container was orientated with the lid downwards in a way that the punch, a mild steel bar, impacts a filter lid. The test specimen was instrumented with strain gauges and accelerometers for the drop tests. Transient strains at selected points of the inner and outer container walls, at the shielding lid, as well as at the lid bolts were measured during the container's impact. Furthermore, decelerations of the container body, container lid, and the skip were measured. The complex geometrical changes of the container due to impact were determined by optical 3d- deformation measures using the projected fringes method in combination with multi-image photogrammetry. This paper summarizes the performance of the drop tests and various drop test results in context with the design development as well as aspects regarding the associated Finite Element (FE) analyses and post-test evaluation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Sizing and Selection of Pressure Relief Valves for High-Pressure Thermal–Hydraulic Systems

This study covers the critical concerns related to the sizing, selection, installation, maintenance, and testing of pressure safety valves (PSVs). The aim is to ensure the safety of pressurized systems, hydrostatic transmission systems, and hydraulic plants, including process plants, thermal power plants, and nuclear reactor systems. PSVs are devices that ensure the safety and reliability of pressurized vessels, lines, and systems during overpressure events. The task of selecting which PSV features are of greatest value for a specific purpose is complex—especially in the design of a high-pressure experimental thermal–hydraulic facility for hydrostatic and transient testing of the reactor system—when the systems are in the design and development phases and require qualification and demonstration to prove that they have reached a given level of technological readiness. The present study highlights the required steps for users to follow the associated rules, guidelines, and recommendations. As a part of this research, case studies are presented to help readers better understand the applicable strategy and standards. A discussion and a review of PSV performance degradation and failure are summarized to provide a better understanding of varied process applications and conditions, including fluid flow dynamics, boundary-layer formation and pressure drops, gas bubble formation and collapse, geometric configurations, inlet/outlet piping, abrupt pressure fluctuations, and acoustic resonance. Moreover, this study discusses the servicing and testing of PSVs in a multiphase pressurized system. Overall, it provides a basic overview of how PSVs ensure the safety of pressurized systems, supported by case studies and industrial practices.

02 PETROLEUM↗
DOE OSTI2016

Neutronics Analysis Supporting Baseline Fuel-Rodlet Irradiation Design in TREAT Multi-SERTTA Vehicle

The Transient Reactor Test (TREAT) Facility is a thermal-spectrum test reactor designed to evaluate reactor fuels and structural material under simulated nuclear excursion and transient power/cooling mismatch situations in nuclear reactors. TREAT was historically operated from 1959 through 1994. There are ongoing activities to resume its transient testing capabilities to support advanced reactor designs and validate computational predictions of fuel and core behavior. Such validation will be an important factor for review and approval of next generation power reactor design and safety evaluations. Historic methods used in TREAT were very approximate and served to estimate a predicted transient power shape. A number of calibration test experiments were then required prior to the actual transient test to better estimate the power that would have been deposited in the experiment during a transient pulse. Much of the experience in the use of these legacy methods has also been lost over the past two decades. Monte Carlo criticality codes are being utilized for calculation of the initial critical state, control rod reactivity worth, and power coupling between core and experiments. Current efforts primarily support neutronics characterization with safety and design analyses for the Multi-SERTTA (Static Environment Rodlet Transient Test Apparatuses) experiment test vehicle, which can accommodate up to four concurrent rodlets under separate environmental conditions. Ongoing efforts are necessary in preparation of transient testing support for Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. The power coupling factor (PCF) is one of two key parameters utilized in designing and characterizing the performance of a TREAT fuel irradiation experiment. The second key parameter is the Transient Correction Factor (TCF) which accounts for time-dependent effects. The PCF is the effective quantity of fission-generated energy per mass of fuel specimen per total core energy. The quantity of energy deposited into a fuel rodlet under various test environments can be used to evaluate overall fuel performance, including failure mechanisms and thresholds. Baseline fuel rodlet calculations have been performed to estimate the effectiveness of the Multi-SERTTA test vehicle and optimization of PCF for various test conditions. The baseline serves as reference fuel pin conditions against which calibration and baseline experiments can be performed prior to actual ATF transient testing. Baseline experiments further serve to validate computational analyses and experimental test methods prior to design planning, experimentation, and evaluation of the ATF candidates. This paper serves to document activities performed to evaluate a reference Pressurized Water Reactor (PWR) rodlet design in the TREAT Multi-SERTTA test vehicle. (authors)

42 ENGINEERING↗
DOE OSTI2011

Materials-Enabled High-Efficiency (MEHE) Heavy-Duty Diesel Engines

The purpose of this Cooperative Research and Development Agreement (CRADA) between UTBattelle, Inc. and Caterpillar, Inc. was to improve diesel engine efficiency by incorporating advanced materials to enable higher combustion pressures and temperatures necessary for improved combustion. The project scope also included novel materials for use in advanced components and designs associated with waste-heat recovery and other concepts for improved thermal efficiency. Caterpillar initially provided ORNL with a 2004 Tier 2 C15 ACERT diesel engine (designed for on-highway use) and two 600 hp motoring dynamometers. The first year of the CRADA effort was focused on establishing a heavy-duty experimental engine research cell. First year activities included procuring, installing and commissioning the cell infrastructure. Infrastructure components consisted of intake air handling system, water tower, exhaust handling system, and cell air conditioning. Other necessary infrastructure items included the fuel delivery system and bottled gas handling to support the analytical instrumentation. The second year of the CRADA focused on commissioning the dynamometer system to enable engine experimentation. In addition to the requirements associated with the dynamometer controller, the electrical system needed a power factor correction system to maintain continuity with the electrical grid. During the second year the engine was instrumented and baseline operated to confirm performance and commission the dynamometer. The engine performance was mapped and modeled according to requirements provided by Caterpillar. This activity was further supported by a Work-for-Others project from Caterpillar to evaluate a proprietary modeling system. A second Work-for-Others activity was performed to evaluate a novel turbocharger design. This project was highly successful and may lead to new turbocharger designs for Caterpillar heavy-duty diesel engines. During the third (and final) year of the CRADA, a novel valve material was evaluated to assess high temperature performance and durability. A series of prototype valves, composed of a unique nickel-alloy was placed in the engine head. The engine was aggressively operated using a transient test cycle for 200 hours. The valve recession was periodically measured to determine valve performance. Upon completion of the test the valves were removed and returned to Caterpillar for additional assessment. Industrial in-kind support was available throughout the project period. Review of the status and research results were carried out on a regular basis (meetings and telecons) which included direction for future work activities. A significant portion of the industrial support was in the form of information exchange and technical consultation.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2018

Summary and Assessment of Metal Fuel Capabilities in Bison

The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-18IN0201012 on Release Bison with Initial Metallic Fuel Performance Capability. Metallic fuel development began in Bison a few years after light water reactor development began (~ 2009). The initial development was funded by the Advanced Fuels Campaign. With recent interest from industry, the Nuclear Regulatory Commission, and the Versatile Test Reactor program, the Nuclear Energy Modeling Simulation Program has increased funding for metallic fuel development and validation in FY18. The material models with the most mature development are binary uranium-zirconium and ternary uranium-plutonium-zirconium alloys for fuel and HT9 for cladding. The first material models were incorporated from open literature sources, which include mechanical and thermal material properties that are functions of temperature, porosity, and zirconium concentration, swelling, fission gas release, zirconium diffusion, creep, and sodium coolant channel boundary conditions. These material models have been utilized to simulate ERB-II fuel pins and the results compared to measurements. The EBR-II experiment measurements mostly consist of cladding strain and zirconium redistribution. As such, comparisons to EBR-II measurements are currently limited to these two figures of merit. Simulations and comparisons to experiment measurements have also been done for transient testing of EBR- II fuel pins in TREAT, which included temperature measurements. Preliminary comparisons between Bison calculations and experiment measurements are favorable. These comparisons have also highlighted the importance of the fuel swelling and fission gas release models on cladding strain. Also, Bison metallic fuel simulations are currently being run in support of ATR experiments and VTR exploratory design calculations. Further development and evaluation of individual material models and fuel system models are planned for FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Simulation of Peach Bottom turbine trip transients TT1, TT2 and TT3 with SIMULATE-3K

The Peach Bottom 2 Turbine Trip transients tests TT1, TT2 and TT3 are simulated with SIMULATE-3K. This work provides the initial basis for evaluation of the capabilities and accuracy of coupled 3D neutronics/thermal hydraulics best estimate calculations for support of safety analysis. Selected results, obtained with SIMULATE-3K, are compared to available measurements and benchmark results. The comparisons are analyzed and the application of SIMULATE-3K for best estimate analysis approach within the safety analysis methodology is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Investigation of Treat M2 and M3 calibration experiments for benchmark analysis

The Transient Reactor Test Facility (TREAT) located at Idaho National Laboratory is a graphite moderated and reflected, air cooled core. This reactor was originally designed and operated to experimentally test fuel under extreme transient conditions. To this end, TREAT successfully performed many series of steady-state and transient tests until operations were discontinued in 1994. With the restart of the reactor planned for 2018, accurate models of previous reactor experiments are necessary to verify the ability to predict conditions for planned upcoming tests. In this work, the methodology for modeling the M2 and M3 calibration experiments (M2/M3CAL) will be described along with results from the variation of key parameters including the boron concentration. Additionally, this model will be used to demonstrate the correlation between reactor behavior and the implementation of 'nuclear graphite' thermal scattering law (TSL) data to capture the behavior of thermal neutrons in the TREAT core.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

High Burnup Transient Fission Gas Release Testing and Post-Test Examinations

The contribution of transient fission gas release (tFGR) to the burst behavior of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant additional transient fission gas release in excess of the steady state fission gas release from normal operation. An addition to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) was created to explore the different conditions that impact transient fission gas release. Four tests on commercially irradiated zirconium clad uranium dioxide nuclear fuel were performed to support transient fission gas release study. These experiments and their post-test analyses are discussed. The first test on high burnup fuel revealed a LOCA relevant transient fission gas release of 5.3% for an unpressurised segment. Subsequent tests have measured transient fission gas release at different conditions that span from 4.8% to 14.4 % transient fission gas release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Effect of natural fractures on determining closure pressure

Abstract In Utah FORGE reservoir, eight pressure transient tests (microhydraulic fracturing and DFIT™) show natural fracture/pressure-dependent leakoff. This behavior may lead to misinterpretation of the closure pressure (proxy for minimum principal stress). The closure pressures obtained from DFIT™ (or microhydraulic fracturing) testing may reflect shear failure along natural fractures or discontinuities rather than tensile failure and lead to inaccurate estimates of the minimum principal stress. In pressure tests conducted at the Utah FORGE site showed that reactivation or opening of natural fractures intersecting the wellbore and were suggested by multiple closure events in DFIT™ tests. In addition, comparisons between the pre- and post-well-tests FMI logs show that there are populations of induced fracture and reactivated natural fractures which are mostly vertical and sub-vertical. In this study, DFIT™ (or microhydraulic fracturing) test analysis and numerical simulations were used to suggest that hydraulic shearing of critically stressed natural fractures can contribute to multiple closure signatures and possibly lead to incorrect determination of the minimum principal stress. As other authors have previously advocated, the DFIT™ test analyses and numerical simulations suggest that better estimations of minimum principal stress may be derived by injecting at relatively high rate and pressure and insuring that tensile breakdown is reached. In addition, extended shut-in period is required to determine accurate reservoir characteristics and fluid flow regime.

Nadimi, S.↗
DOE OSTI2023

Microstructure and phase evolution in the U-10Zr fuel investigated by in situ TEM heating experiments

The development of U-Zr metallic nuclear fuel for fast spectrum reactors is impacted by a lack of mechanistic understanding of the fuel behavior evolution under thermal irradiation conditions, despite previous works providing substantial fuel performance data. This work uses in-situ transmission electron microscopy heating experiments to study phase and microstructural evolution in several unirradiated U-10Zr specimens during rapid heating ramps (from room temperature to 1000 °C). The starting a-U + bcc-(Zr, U) eutectic microstructure began to decompose above 600 °C. The decomposition initiated from the bcc (U,Zr) phase. Similar results were observed for all specimens even when fabricated by different routes (e.g., cold rolled or annealed). As a result, the impact of observed microstructure and phase evolutions at high temperatures on fuel fabrication and in-pile fuel transient test was also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗