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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Select Transient Metallic Fuel Bison Benchmarks

To explore the current capabilities of Bison with respect to transient metallic fuel modeling, four transient experiments are compared with their respective Bison model. The four experiments are the OPT-1 experiment of the Experimental Breeder Reactor II~(EBR-II) and the M5, M6, and M7 experiments of the Transient Reactor Test facility (TREAT). Each benchmark is a 2D-RZ geometry of a single pin employing coupled multiphysics. Bison models of the benchmark agree well and illustrate areas needing development. The areas of suggested development deal with better material models, better coolant channel module for transients, proper treatment of material melting, improved feedback with failure models, and an ability to set initial conditions from post irradiation examination (PIE) data. Two Fast Flux Test Facility (FFTF) pins are also modeled with M7 transient conditions to illustrate Bison being applied to investigate potential experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

TREAT Modeling and Simulation Strategy

This report summarizes a four-phase process used to describe the strategy in developing modeling and simulation software for the Transient Reactor Test Facility. The four phases of this research and development task are identified as (1) full core transient calculations with feedback, (2) experiment modeling, (3) full core plus experiment simulation and (4) quality assurance. The document describes the four phases, the relationship between these research phases, and anticipated needs within each phase.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Evaluation of oxygenation in the surface layers of biological tissues based on diffuse optical spectroscopy with automated calibration of measurements

To assess the oxygen saturation (oxygenation) in the surface layers of biological tissues (at a depth of 4 mm), a compact system based on diffuse optical spectroscopy (DOS) with reflective measurement geometry in the wavelength range of 520 – 590 nm is proposed. The experimental DOS device is based on a broadband LED source of probe radiation and a spectrometer using a fibre-optic radiation delivery system. To calculate the oxygenation, an original method is proposed based on measuring the intensity of scattered light at four different locations of the source and receiver on the surface of the object under study (four-measurement method). The developed method allows for the device hardware parameters (optical contact with the tissue, transient characteristics of the radiation delivery system) and does not require any additional calibration measurements. The results of testing the DOS system using the four-measurement method for monitoring the oxygen saturation of tissues during the artificial occlusion of the blood vessels agree with the published data of other research teams. (paper)

60 APPLIED LIFE SCIENCES↗
DOE OSTI2012

NWTC Controllable Grid Interface (Fact Sheet)

NREL's Controllable Grid Interface tests wind turbines off-line from the grid, verifies compliance with standards, and provides grid operators with the performance information they need for a faction of the time and cost it would take to test the turbine in the field.

17 WIND ENERGY↗
DOE OSTI2013

Stochastic joint inversion of hydrogeophysical data for salt tracer test monitoring and hydraulic conductivity imaging

The assessment of hydraulic conductivity of heterogeneous aquifers is a difficult task using traditional hydrogeological methods (e.g., steady state or transient pumping tests) due to their low spatial resolution. Geophysical measurements performed at the ground surface and in boreholes provide additional information for increasing the resolution and accuracy of the inverted hydraulic conductivity field. We used a stochastic joint inversion of Direct Current (DC) resistivity and Self-Potential (SP) data plus in situ measurement of the salinity in a downstream well during a synthetic salt tracer experiment to reconstruct the hydraulic conductivity field between two wells. The pilot point parameterization was used to avoid over-parameterization of the inverse problem. Bounds on the model parameters were used to promote a consistent Markov chain Monte Carlo sampling of the model parameters. To evaluate the effectiveness of the joint inversion process, we compared 8 cases in which the geophysical data are coupled or not to the in situ sampling of the salinity to map the hydraulic conductivity. We first tested the effectiveness of the inversion of each type of data alone (concentration sampling, self-potential, and DC resistivity), and then we combined the data two by two. We finally combined all the data together to show the value of each type of geophysical data in the joint inversion process because of their different sensitivity map. We also investigated a case in which the data were contaminated with noise and the variogram unknown and inverted stochastically. The results of the inversion revealed that incorporating the self-potential data improves the estimate of hydraulic conductivity field especially when the self-potential data were combined to the salt concentration measurement in the second well or to the time-lapse cross-well electrical resistivity data. Various tests were also performed to quantify the uncertainty in the inverted hydraulic conductivity field.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Candidacy Renewal Report for ICERR Designation

The International Atomic Energy Agency (IAEA) has established a designation for an International Centre based on Research Reactors (ICERR). The intention of this designation is to provide a vehicle for IAEA member states to access international research-reactor and ancillary nuclear research and development infrastructure. The U.S. Department of Energy (DOE) has made a commitment to world leadership in the development of advanced nuclear energy, science, and technology. To this end, DOE has established programs and initiatives to enhance this leadership role. ICERR designation constitutes an important step in achieving the DOE vision. DOE, represented by Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL), is submitting this application for re-designation as an IAEA ICERR. Both INL and ORNL have a decades-long and storied history that supports nuclear research, development, and deployment both nationally and internationally. Both have a history of safe and efficient nuclear operations and have demonstrated a track record of international collaboration and cooperation. INL and ORNL are home to two primary and two secondary research reactors. The primary reactors are Advanced Test Reactor at INL, and High-Flux Isotope Reactor at ORNL. Secondary reactors include: Neutron Radiography Reactor at INL, Transient Reactor Test (TREAT) Facility at INL, and INL Advanced Test Reactor: Critical Facility (ATRC). ICERR designation will also include numerous ancillary facilities at both laboratories, including the following: Multiple post-irradiation examination facilities at both INL and ORNL, Radiochemistry hot-cell facilities at ORNL, INL nuclear-fuel-fabrication and materials-science facilities, and INL and ORNL radiological and nuclear analytical capabilities.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Evaluation, Refurbishment, and Characterization of Fast Neutron Scintillators for the TREAT Fuel Motion Monitoring System

This paper details work being performed at the Idaho National Laboratory (INL) to evaluate, refurbish, and characterize proton-recoil scintillation (PRS) detectors utilized in a 360 channel array for the measurement of fast neutrons in a standalone system historically called a hodoscope located in the Transient Reactor Test Facility (TREAT). The TREAT hodoscope serves as the primary Fuel Motion Monitoring System for documenting real time fuel displacement during transient experiments. The detectors in the hodoscope have stood dormant for over 20 years and significant degradation must be addressed before the system can be brought back to operational status. The refurbishment process begins by decoupling the PRS from their matched photo multiplier tubes (PMTs) and treating the PRS with ethanol. This process cleans the surfaces, removing paint and legacy labels/ markings. During this procedure many physical variations can be observed. After cleaning, the PRS are physically measured using a digital caliper and the dimensions documented. Following this, a well-controlled 280nm LED light source is used to stimulate the PRS to produce an emission spectrum inside an integration sphere. This emission spectrum is then measured using a UV enhanced optical spectrometer. Spectral measurements are recorded from 185 nm to 580 nm in 0.4 nm steps. The PRS are then lightly sanded on the non-PMT mating surfaces and painted with commercially available reflective paint. A final emission spectrum is recorded for each PRS after painting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Analysis of Transient Groundwater Flow Near the Cheshire Test at Pahute Mesa, Nevada National Security Site

This report focuses on analysis and determination of the causes of transient water levels measured between 1964 and 2017 near the Cheshire underground nuclear test, which was detonated on February 14, 1976 in Area 20 (USDOE, 2015) of the Pahute Mesa (PM) testing areas of the Nevada National Security Site (NNSS) located in southern Nevada (Figure 1). Transient water levels near Cheshire indicate influences of a water-production well and yet-unexplained water level rise attributed to testing effects (Jackson and Fenelon, 2018; Jackson et al. 2021). Transient groundwater flow is relevant to the implementation of the hydrologic source term (HST), which prescribes the quantity and release rates of radionuclide source terms to be applied at individual test locations in assessment of regional-scale radionuclide transport (Carle and Tompson, 2021; Carle et al., 2021), as previously investigated for NNSS testing areas at PM (Pawloski et al., 2001), Frenchman Flat (Carle et al, 2007), Yucca Flat (Mcnab, 2009; Tompson, 2009), and Rainier Mesa (Carle, 2011; Tompson et al., 2011).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Analysis of Transient Groundwater Flow Near the Cheshire Test at Pahute Mesa, Nevada National Security Site

This report focuses on analysis and determination of the causes of transient water levels measured between 1964 and 2017 near the Cheshire underground nuclear test, which was detonated on February 14, 1976 in Area 20 (USDOE, 2015b) of the Pahute Mesa (PM) testing areas of the Nevada National Security Site (NNSS) located in southern Nevada (Figure 1). Transient water levels near Cheshire indicate influences of a water-production well and yet-unexplained water level rise attributed to testing effects (Jackson and Fenelon, 2018; Jackson et al. 2021). Transient groundwater flow is relevant to the implementation of the hydrologic source term (HST), which prescribes the quantity and release rates of radionuclide source terms to be applied at individual test locations in assessment of regional-scale radionuclide transport (Carle and Tompson, 2021; Carle et al., 2021), as previously investigated for NNSS testing areas at PM (Pawloski et al., 2001), Frenchman Flat (Carle et al, 2007), Yucca Flat (Carle et al., 2009; Mcnab, 2009; Tompson et al., 2009), and Rainier Mesa (Carle, 2011; Tompson et al., 2011).

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Conceptual Design of a Pumped Water Capsule for TREAT

This study performed mechanical design, thermal hydraulic, and fuel performance investigations to determine whether an existing large water capsule design could be modified to create forced convection boundary conditions on nuclear fuel specimens in the Transient Reactor Test facility. This concept was found to be viable and offers several advantages compared to a static water environment for a variety of transient power shape categories. The pumped capsule is not foreseen to be superior to a past concept for a larger water loop, but the pumped capsule approach is recommended for further engineering and deployment based on its high cost-to-value ratio primarily as a stop gap capability until a larger water loop can be realized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Plant-wide simulation model for transient studies on the XE-100 - 18464

This paper gives an overview of the plant-wide simulation model that is used by X-energy LLC for developing their plant design and control philosophy, as well as assessing plant upset conditions. The simulation model was created in Flownex{sup R} thermal-fluid system simulation software, which solves mass, momentum and energy equations for compressible, incompressible and homogeneous two-phase flow. The plant model encompasses all major components, from reactor to condenser, including main control loop elements. The paper pays particular attention to the validation of the two-phase flow framework used for the secondary steam cycle. Two steady-state validation cases are done using operational data from the THTR. Validation of transient operation is done by simulating density wave oscillations, a type of system-level boiling phenomenon. This is regarded as a thorough test of transient capability because of the challenges this phenomenon poses to several aspects of the calculation and to the consistency of the code as a whole. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Evaluation of Optical Fiber Bragg Gratings in a Nuclear Reactor

Optical fiber-based sensors are capable of distributed temperature and/or strain measurements with high accuracy and sub-cm spatial resolution. Optical fibers have a small diameter ({approx}100 {mu}m), a small mass (and therefore fast response time), and are fabricated from high temperature-tolerant, radiation-hard materials. These qualities make optical fiber-based instrumentation a promising candidate for in-pile sensing applications, monitoring fuel performance during irradiation in test reactors, or embedding fibers in other commercial reactor components or structures. The Department of Energy and Idaho National Laboratory are restarting the TREAT reactor (Transient Reactor Test facility) and are interested in updating the reactor's instrumentation regarding the analysis of the test fuel. The ability of optical fibers to make distributed temperature measurements coupled with its small footprint makes optical fiber a viable candidate to monitor fuel performance in the TREAT reactor. The work presented in this paper focuses on the radiation hardness of various types of optical fiber Bragg gratings. Previous experiments have shown that the inscription of Bragg gratings into optical fiber can extend the radiation survivability of distributed sensing in optical fibers, and so the goal is to determine which types of Bragg gratings survive the best in a reactor environment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Sensitivity analysis of in-pile critical heat flux experiments in TREAT for characterization of RIA power-transient effects

A reactivity-initiated accident (RIA) is one type of postulated design basis accident (DBA) that can cause a departure from nucleate boiling (DNB) event in pressurized-water reactors (PWRs). A DNB occurrence and its consequences depend on the thermophysical properties of the fuel components and coolant, characteristics of the transient energy insertion into the fuel rod, and the onset of the critical heat flux (CHF) phenomenon. To leverage the restart of the Transient Reactor Test (TREAT) Facility, an effort is currently underway to better understand the cladding-to-coolant heat transfer mechanisms and the CHF phenomenon under fast-transient irradiation conditions. This paper characterizes the impact of power transients on the thermal-hydraulic behavior of a TREAT Facility reactor heater rodlet CHF experiment to provide the priority of parameters that need to be investigated for an improved CHF model. Sobol sensitivity analysis methods and the Reactor Excursion and Leak Analysis Program (RELAP5-3D) code were used to identify key input parameters on the uncertainty in the prediction of peak outer- and inner-surface temperatures of the heater tube, as well as the time of the DNB event. A series of sensitivity analyses revealed the total energy deposition on the tube and the transient effects of power pulse had large impacts on the maximum temperatures. The CHF multiplier had the largest impact on the time occurrence of CHF. The overall results show the energy deposition rate in the tube is the most influencing factor to the manifestation of CHF and the resulting thermal-hydraulic behaviors of the tube. The multiplier for the CHF, which is interpreted as the predicted CHF value, has the largest Sobol indices for the time of the CHF in all cases, since it directly determines the occurrence of CHF. It is inferred that the uncertainties in the thermal-hydraulic behaviors of fuels increase with respect to the key parameters as the power pulse becomes broader, and an accurate estimation of the energy deposition rate is required to reduce the uncertainty in the evaluation of the integrity of fuel if the CHF is expected to occur near the peak power. Therefore, the outputs are expected to provide rigorous interpretation of ongoing in-pile CHF experiments in the TREAT Facility reactor regarding thermal-hydraulic behavior of the fuel system aiming for a new transient in-pile CHF model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Effect of reactor radiation on the thermal conductivity of TREAT fuel

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

NCERC-CL TREAT Free Field Characterization Foil Reaction Rate Results

This report summarizes the analysis and preliminary MCNP reaction rate modeling of free field characterization measurements performed in the Transient Reaction Test Facility (TREAT) at Idaho National Laboratory (INL) between 12/18/2023 and 2/22/2024. A series of seven irradiations were performed using the Big-BUSTER (Broad Use Specimen Transient Experiment Rig) core configuration. Foil sets including 19.5% LEU-Zr alloy wire, S, Au, Fe, Ni, Co, Ti, and Zr were deployed in each irradiation. Identical foil sets were supplied to Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and INL. All foils were supplied by INL. Additional details on the experiment can be found in [2]. The foil sets were shipped to the National Criticality Experiments Research Center Counting Laboratory (NCERC-CL) branch at LANL (NCERC-CL NISC) and were received on 3/25/2024. Additional details on receipt and counting are reported in [3]. All reactor dosimetry measurements were performed in adherence to the ASTM standards applicable to reactor dosimetry. Measured reaction rates from Au, Co, Fe, Ni, and Ti are reported. Select reaction rate ratios and simulated reaction rates and ratios are discussed. Reaction rates for all measured reaction products were modeled with MCNP 6.3.1 and a TREAT input deck supplied by Edward Lum. Neutron emission estimates calculated using the measured and simulated reaction rates show excellent agreement between the three fast-threshold n-p reactions with the percent differences being < 9% between the three reactions. The capture reactions had poorer, but still reasonable, agreement with < 20% percent differences between the neutron emission estimates of the three capture reactions. The percent difference between the fast-threshold and capture reactions is ∼ 100% for each reaction. This supports the TREAT model is not accurately modeling the neutron spectrum and additional measurements are required to characterize TREAT and match measurements to simulation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Development of the TREAT M2 experiment as a transient benchmark

A transient benchmark based on the Transient Reactor Test Facility (TREAT) M2 Calibration experiment (M2-CAL) is under development. TREAT, at Idaho National Laboratory, is a graphite moderated air-cooled research reactor which has been used extensively for fuel material testing under extreme and accident conditions. Accurate benchmark models are a beneficial component in the operations, experimental planning, and development of TREAT. In this work, we present a transient benchmark model for the M2-CAL experiment core loading. The benchmark model incorporates the coupling between the Monte Carlo Code SERPENT and the computational fluid dynamics code OpenFOAM to capture the temperature feedback mechanism. The M2-CAL transient 2580 was simulated in this work. A pre-transient analysis was performed to determine the optimum core composition and conditions before the beginning of the transient. The analysis was validated against historic TREAT kinetic measurements and the worth of the transient rod T-2. The axial power distribution in the flux wire for the M2-CAL experiment was determined and contrasted with the experiment. The model was then used to simulate the M2-CAL transient 2580 experiment based on the reported pre-transient and transient conditions. Several transient observables were calculated and compared to the experiment. The model shows good agreement with the experimental power traces, period, and average increase of power at the power ramp (time > 7.8 s). Inverse point kinetics analysis was performed during the period of the power ramp for the model and the experiment based on that, the temperature feedback component was isolated and contrasted with the experimental feedback. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Analysis of Silver Release from Furnace-Tested TRISO Particles

High temperature testing of intact TRISO particles previously irradiated in the AGR-5/6/7 experiment was performed in the Furnace for Irradiated TRISO Testing (FITT) to directly confirm diffusive release of silver and europium from intact TRISO particles. Testing was conducted from 1,100–1,600°C for exposure times up to 100 h to directly confirm silver through layer release below safety testing temperatures. The FITT analysis showed highest levels of silver release at 1,300–1,400°C which confirms athermal release behaviors previous observed in step-wise and transient safety tests. Additionally, release was non-uniform with some particles releasing a majority of their inventory while others did not appear to release silver under identical testing conditions, which was consistent with historic observations. An assessment of the effective silver diffusion coefficient in the SiC layer was conducted and indicated maximum values in the 1,300–1,400°C range. The magnitude of the calculated diffusion coefficients also exceeded currently accepted diffusion coefficients. The release behavior of europium from intact particles was also analyzed in FITT for at 1,450–1,550°C for 500 h to 750 h. Direct confirmation of europium release below safety testing temperatures was confirmed absent contributions from matrix release. The analysis indicated europium release follows a general Arrhenius behavior and suggests general uniform release and indicates irradiation conditions influence observed release response. Calculated diffusion kinetics agree well with historic experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multiscale Thermal-hydraulic analysis of the MARVEL micro-reactor using coupled MOOSE Subchannel (SCM) and SAM

MARVEL is a natural-convection-cooled sodium-potassium microreactor that is anticipated to generate 85 kilowatts of thermal energy. It will operate within Idaho National Laboratory’s Transient Reactor Test Facility and is being developed by the DOE Microreactor Program. MARVEL will be used to test microreactor applications, evaluate systems for remote monitoring, and develop autonomous control technologies. A thermal-hydraulic computational model of this facility is a valuable tool to study important transients and calculate the safety limits of the micro-reactor design. For this purpose, the authors have chosen to use a multiscale coupled simulation: SCM for modeling the reactor core and SAM for the reactor’s primary cooling system. SCM is MOOSE physics module for subchannel analysis, which was designed to model single-phase flows through liquid-metal cooled, wire-wrapped fuel pin sub-assemblies, ordered in a triangular lattice. The SCM code was modified to be able to model MARVEL’s unique geometry. SAM is a systems analysis module based on the MOOSE framework. It aims to provide fast-running, whole-plant transient analyses capability with improved-fidelity for various advanced reactor types. The coupling between the two SCM and SAM for MARVEL modeling is done implementing a domain over-lapping approach. The resulting coupled simulation can model transients such as reactor startup/shutdown and provide an intermediate fidelity picture of the temperature field and other variables, in the core. Results for the steady-state simulations are presented in the article as well as flow blockage transient.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗