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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

COR-0011 Rev 6 MARVEL Project Code of Record

This Code of Record identifies the codes, standards, and procedures necessary to design, develop, construct, and startup the Microreactor Applications Research Validation and Evaluation (MARVEL) Project at the Materials and Fuel Complex (MFC) Transient Reactor Test (TREAT) Facility and the Idaho National Laboratory (INL). The MARVEL Project is an INL test microreactor funded by the United States Department of Energy (DOE) via the Microreactor Program (MRP). The goal of the project is to establish an operational nuclear applications test bed that can generate combined heat and power to enable integration and R&D with end-user technologies, as well as allow microreactor technologists to test next-generation control systems. The microreactor is a thermal reactor utilizing Uranium Zirconium Hydride (UZrH) fuel with review and authorization by the Department of Energy Idaho Operations Office (DOE-ID) for National Environmental Policy Act (NEPA) compliance, safety review, and supplemental readiness assessments for startup and operation. To enable rapid deployment, the MARVEL reactor will reside in the Transient Reactor Test (TREAT) Facility and utilize the existing operating Category B reactor facility, approved facility safety basis, operating crews, and recent re-start experience.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2019

Second Annual Progress Report on Ultrasonic Sensors for TREAT Fuel Condition Measurement and Monitoring

Fuel deformation (both axial and radial) information provides insights into fuel performance and failure limits. Traditionally, fuel deformation has been quantified using post-irradiation examinations due to the lack of an appropriate sensor for monitoring deformation in situ. Recent developments in sensor and instrumentation design have led to the use of linear variable differential transformers (LVDT) for monitoring axial strain in fuel during irradiation. Size and operational constraints limit the use of LVDT sensors for radial deformation monitoring, and alternatives are needed. One possible alternative to LVDT sensors is an ultrasonic sensor. Ultrasonic sensors monitor the interaction of high-frequency stress waves with the test object. The resulting response may be analyzed to determine the properties of the material, including stress, strain, microstructure, presence of defects, and the elastic constants. While ultrasonic measurements have been successfully applied for monitoring strain in a number of applications, their use for in-core measurements has been limited. Largely, the limitations are due to the challenges associated with designing sensors that can operate under typical in-core conditions. The overall objective of this work is to design an ultrasonic sensor capable of rapid in-situ measurements of fuel rodlet dimensional changes during TREAT transient irradiation testing. This report summarizes the activities and research findings through the end of Year 2 of this three-year project.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Real-World 20-Hz IBR Subsynchronous Oscillations: Signatures and Mechanism Analysis

This article provides analyses of subsynchronous oscillations associated with solar photovoltaics (PVs) observed in real world. All have oscillation frequencies at approximately 20 Hz in voltage and current magnitudes. On the other hand, in phase voltage and current, the oscillations show very different signatures. Through constructing simplified linear models and conducting sensitivity analysis, we are able to associate dynamic phenomena with the underlying mechanism and speculate the possible causes of the 20-Hz oscillations. The analysis results have all been validated via simulation results from electromagnetic transient (EMT) test beds constructed in MATLAB/Simscape.

inverter-based resources↗
DOE OSTI2015

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the power coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.

Connaway, H. M.↗
DOE OSTI2023

A non-fueled nuclear-heated rod for in-pile transient boiling studies

Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10 B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (B nat ~ 2.05 wt %) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-a thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. In conclusion, these results give confidence that the BNHR design has successfully met experiment objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

TREAT HODOSCOPE – A Predictive Simulation Model

The Transient Reactor Test Facility (TREAT) reactor is being restarted after over two decades of standby operation. TREAT is designed to test the time-dependent behavior of nuclear fuel during simulated reactor accidents. One of the principle data collection systems of TREAT is the fast neutron hodoscope which is used to observe the motion of the fuel test specimen as it is being damaged, deformed, or melted over the course of the power transient. The hodoscope is a grid of ex-core collimator channels paired with a bank of fast neutron sensitive radiation detectors. These detectors have historically been calibrated using empirical methods utilizing measurement data from smaller power pulses. Given modern computational capabilities this work sets out to establish a predictive computational capability of the fast neutron flux field exiting the hodoscope to support these calibrations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Accelerated Climate Model for Energy (ACME): Ocean and Sea-Ice Processes

The primary goal of this project was to understand the role of the ocean and sea-ice components in the evolution of the climate in both a high-resolution fully-coupled 1850 preindustrial control and present day transient climate simulations run using the Energy Exascale Earth System Model (E3SM), formerly known as the Accelerated Climate Model for Energy (ACME). Two distinct approaches to the initialization of the present day transients were tested: in the first, an ensemble of three transients used atmospheric reanalysis forced 0.1-degree ocean/sea-ice states, while a fourth transient was branched off the preindustrial control. Biases from the control were inherited by its successor transient, in particular, sea surface temperatures (SSTs) at high latitudes were too warm and too much northward meridional heat transport occurred in the North Atlantic Ocean. SSTs in the Northern Hemisphere and Atlantic meridional heat transport increased in the transients initialized from the ocean/sea-ice simulation over a 40-50 year period to largely resemble the transient branched off the control simulation. Excessive absorbed solar radiation (ASR) is considered to be responsible for the SST biases as well as the formation of mode waters in lighter density classes than in reality both in the Southern Ocean and the North Atlantic Ocean subtropical gyre. These mode waters ventilate the thermocline waters to the north of their formation sites explaining the excessive Atlantic meridional heat transport and some of the warming in the high-latitude North Atlantic Ocean after multiple decades.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Implementation of Depletion Architecture in the MAMMOTH Reactor Physics Application

Goal is to add functionality to MAMMOTH to perform depletion calculations. The purpose of this is to allow the use of the same mesh for fuel elements depleted in commercial reactors or the Advanced Test Reactor (ATR) before being placed in the Transient Reactor Test Facility (TREAT) for further experimentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Controlled Pyroshock Transients Can Be Used To Better Match Operational Shock Transients and Improve Shock Analysis and Testing

SRS analysis and testing to SRS is known to be conservative since SRS is generated from maxi-max shock transient data, maximum peaks are assumed to occur for each mode simultaneously, and the structural dynamics are not accounted for between the shock source and the test article. Knowing the structural dynamics of a shock test setup it is possible to subscribe a force shock transient applied at a particular strike location and direction to achieve a desired shock transient at a test reference accelerometer. The SRS at this test accelerometer reference location can be used to compare to the reference SRS, but the key reference is the shock transient not the SRS, which should lead to more realistic shock testing that better matches operational shock transient data and reduces artificial test failures that don’t occur in operational usage. The desired test shock transient should match operational shock transients. As such, the test shock transient should look similar to the operational shock transient and have similar positive and negative peaks, the same shape, and the same duration. Many shock transients can be created that satisfy an SRS, but better control of these shock transient parameters is needed to better match operational shock transients so that more realistic and less conservative shock testing can be performed. Three different shock transients are presented that satisfy a given SRS and the results are compared for an electronic board (test article) attached to a circular shock resonance test plate.

SRS, shock transient, frequency domain, FRF↗
DOE OSTI2016

Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup power density on the flow instability. The experimental startup transient results showed the existence of three different flow instability mechanisms, i.e., flashing instability, condensation induced flow instability, and density wave oscillations. In addition, the void-reactivity feedback did not have significant effects on the flow instability during the startup transients for NMR-50. ii Several initial startup procedures with different power ramp rates were experimentally investigated to eliminate the flow instabilities observed from the startup transients. Particularly, the very slow startup transient and pressurized startup transient tests were performed and compared. It was found that the very slow startup transients by applying very small power density can eliminate the flashing oscillations in the single-phase natural circulation and stabilize the flow oscillations in the phase of net vapor generation. The initially pressurized startup procedure was tested to eliminate the flashing instability during the startup transients as well. The pressurized startup procedure included the initial pressurization, heat-up, and venting process. The startup transient tests showed that the pressurized startup procedure could eliminate the flow instability during the transition from single-phase flow to two-phase flow at low pressure conditions. The experimental results indicated that both startup procedures were applicable to the initial startup of NMR. However, the pressurized startup procedures might be preferred due to short operating hours required. In order to have a deeper understanding of natural circulation flow instability, the quasi-steady tests were performed using the test facility installed with preheater and subcooler. The effect of system pressure, core inlet subcooling, core power density, inlet flow resistance coefficient, and void reactivity feedback were investigated in the quasi-steady state tests. The experimental stability boundaries were determined between unstable and stable flow conditions in the dimensionless stability plane of inlet subcooling number and Zuber number. To predict the stability boundary theoretically, linear stability analysis in the frequency domain was performed at four sections of the natural circulation test loop. The flashing phenomena in the chimney section was considered as an axially uniform heat source. And the dimensionless characteristic equation of the pressure drop perturbation was obtained by considering the void fraction effect and outlet flow resistance in the core section. The theoretical flashing boundary showed some discrepancies with previous experimental data from the quasi-steady state tests. In the future, thermal non-equilibrium was recommended to improve the accuracy of flashing instability boundary. As another part of the funded research, flow instabilities of a PWR-type SMR under low pressure and low power conditions were investigated experimentally as well. The NuScale reactor design was selected as the prototype for the PWR-type SMR. In order to experimentally study the natural circulation behavior of NuScale iii reactor during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. The three-level scaling method is used as well to obtain the scaling ratios derived from various non-dimensional numbers. The design of the ideally scaled facility (ISF) was initially accomplished based on these scaling ratios. Then the engineering scaled facility (ESF) was designed and constructed based on the ISF by considering engineering limitations including laboratory space, pipe size, and pipe connections etc. PWR-type SMR experiments were performed in this well-scaled test facility to investigate the potential thermal hydraulic flow instability during the blowdown events, which might occur during the loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) of the prototype PWR-type SMR. Two kinds of experiments, normal blowdown event and cold blowdown event, were experimentally investigated and compared with code predictions. The normal blowdown event was experimentally simulated since an initial condition where the pressure was lower than the designed pressure of the experiment facility, while the code prediction of blowdown started from the normal operation condition. Important thermal hydraulic parameters including reactor pressure vessel (RPV) pressure, containment pressure, local void fraction and temperature, pressure drop and natural circulation flow rate were measured and analyzed during the blowdown event. The pressure and water level transients are similar to the experimental results published by NuScale [51], which proves the capability of current loop in simulating the thermal hydraulic transient of real PWR-type SMR. During the 20000s blowdown experiment, water level in the core was always above the active fuel assemble during the experiment and proved the safety of natural circulation cooling and water recycling design of PWR-type SMR. Besides, pressure, temperature, and water level transient can be accurately predicted by RELAP5 code. However, the oscillations of natural circulation flow rate, water level and pressure drops were observed during the blowdown transients. This kind of flow oscillations are related to the water level and the location upper plenum, which is a path for coolant flow from chimney to steam generator and down comer. In order to investigate the transients start from the opening of ADS valve in both experimental and numerical way, the cold blow-down experiment is conducted. For the cold blowdown event, different from setting both reactor iv pressure vessel (RPV) and containment at high temperature and pressure, only RPV was heated close to the highest designed pressure and then open the ADS valve, same process was predicted using RELAP5 code. By doing cold blowdown experiment, the entire transients from the opening of ADS can be investigated by code and benchmarked with experimental data. Similar flow instability observed in the cold blowdown experiment. The comparison between code prediction and experiment data showed that the RELAP5 code can successfully predict the pressure void fraction and temperature transient during the cold blowdown event with limited error, but numerical instability exists in predicting natural circulation flow rate. Besides, the code is lack of capability in predicting the water level related flow instability observed in experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

TREAT LEU Conversion Analyses Performed 2016-2018

TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. After more than 20 years of non-operational standby, TREAT has been restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT from its current highly enriched uranium fuel to a low enriched uranium fuel is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization Reactor Conversion Program. This report documents additional analyses performed over the timeframe of 2016 to early 2018 to evaluate the LEU Conceptual Design for the TREAT reactor, following the initial design studies documented in the 2016 Conceptual Design Report. The analyses documented in this report include neutronic and thermal hydraulic sensitivity studies to evaluate the effect of key design parameters on predicted core behaviors, as well as a simple investigation into the effect of fuel thermal expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Realistic operation of two residential cordwood-fired outdoor hydronic heater appliances—Part 1: Particulate and gaseous emissions

This study investigated how heat demand and fuel loading affect the emissions from outdoor wood-fired hydronic heaters by testing two such appliances using an integrated-duty cycle test method. This test included transient operating conditions, such as cold and hot-starts and modulation between 15 and 100% of maximum rated output. Emission values indicate transient operating conditions produce higher emissions than steady state operation. Cold starts resulted in elevated particulate matter emission factors for both appliances; in one case the particulate matter emission factor for this period was >3500 mg/MJ, which represented emissions 20 times the average value. Additionally, when heat demand was cycled—elevated CO emission factors were measured, with values >5000 mg/MJ for both appliances and more than 3 times the appliance averages, respectively. It follows that the appliance average particulate matter emission factors and CO emission factors were not representative of the actual EF values during these transient periods. In contrast, methane emission factors were relatively stable throughout all tested combustion conditions; however, they were much higher than oil-fired appliances are therefore should not be ignored. These findings demonstrate that wood-fired hydronic heater emissions during transient operating conditions can be significantly greater than emissions during steady-state test conditions, such as those used in typical certification tests. Consequently, certification test values for particulate matter and CO emission factors may significantly underestimate the actual emissions of these appliances when operated in a home. Use of integrated duty-cycle test protocols that capture cold-starts and reloading are better for representing in-use operations of wood-fired hydronic heaters and provide more realistic emissions and delivered efficiency measurements.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2016

Improved Quasi-Static Method with Step Doubling Time Adaptation

The imminent restart of the Transient Reactor Testing Facility (TREAT) at Idaho National Laboratory (INL) has brought significant attention to the development of transient reactor modeling. TREAT was designed to subject fuels and other reactor components to various degrees of neutron pulses, from minor transients to accident scenario. Inherently, this task requires predicting pulse responses due to various control rod movements; which is a complicated endeavor. Originally, this was done using an iterative process of rod movements and measuring the resulting response. With the advances in computer modeling, the current goal, preceding the restart, is to use simulation to characterize responses before experiments take place. However, neutron transient modeling is extremely computationally expensive due to the required implicit time-stepping. Therefore, methods that mitigate computation time, with minimal detriment to accuracy, are highly desired. This paper will describe the improved quasi-static method (IQS), which hopes to decrease computation time significantly from traditional 'brute force' transient methods. IQS is a spatial kinetics method that involves factorizing the flux solution into time-dependent amplitude and space- and time-dependent shape. The impetus of the method is the assumption that shape is weakly time-dependent, so expensive space-dependent evaluations are only required on large time steps. While inexpensive amplitude evaluations are performed on much smaller time steps to retain accuracy. IQS has been tested extensively on kinetics benchmarks with constant time steps. However, to test it's viability for TREAT simulations, evaluating IQS's performance with time adaptation is required. The rest of this paper will briefly describe the derivation of IQS, the time adaptation technique used, and results from a kinetic benchmarks. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

VALIDATION OF A RELAP5 POINT KINETICS MODEL OF TREAT

Development of advanced nuclear fuels requires testing of fuel specimens under various power-cooling mismatch conditions. The Transient Reactor Test (TREAT) facility at the Idaho National Laboratory will be used to provide transient nuclear heating of fuel specimens. Experiments will be conducted to support assessment of accident tolerant fuel designs being proposed for light water reactor applications. The experiments to be conducted at TREAT will provide data for the assessment of advanced, multi-physics computer codes. However, before the experiments can be performed, calculations are required to demonstrate that they will meet program objectives and that they can be performed safely in TREAT in the event of an unplanned reactivity insertion accident that occurs during the experiment. The safety calculations for the first experiments need to be performed in the next year and therefore will be performed with an existing computer code, RELAP5-3D. A point kinetics model of TREAT was developed for RELAP5-3D. The purpose of the model is to calculate the reactor power during experiments and to support the accident analyses of those experiments. The RELAP5-3D model was validated using data from nine TREAT experiments conducted previously. The RELAP5-3D model generates results that are in reasonable agreement with measured values of maximum core power, energy deposition, and maximum fuel temperature for a wide range of reactivity insertions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Fuel Motion Monitoring System at TREAT – FY2017 Instrumentation Summary

This document is a report summarizing FY2017 efforts related to the Transient Reactor Test Facility (TREAT) Fuel Motion Monitoring System, "hodoscope," refurbishment campaign. TREAT is an air-cooled, graphite-moderated, heterogeneous test facility designed to evaluate reactor fuels and structural materials under conditions that simulate various types of transient overpower and under-cooling situations in a nuclear reactor. Fuel meltdowns, metal-water reactions, thermal interaction between overheated fuel and coolant, and the transient behavior of ceramic fuel for high-temperature systems can be studied. A key instrument that monitors fuel motion as these events take place is the fast neutron hodoscope. The hodoscope is designed to allow for pre-, during-, and post- transient imaging of fuel within a test loop in the center of the reactor core. The preferred technique for imaging is detection of fast neutrons produced by fission in the test fuel, travelling unmoderated through an empty row of core fuel assemblies and then through a multi-slot collimator to a 360-detector array. Due to the age of the instrument, a refurbishment of the hodoscope’s fast neutron detection capability is necessary to support real-time, fuel-position experimental data collection. One particular challenge in this regard is related to the refurbishment of the fast-neutron-sensing detector elements in the detector array. This report summarizes major activities and accomplishments related to the FMMS refurbishment in FY2017. Of particular note was work related to identifying options for new proton recoil scintillator wafer and initial examination of the TREAT proton recoil proportional counter detectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Comparison of Single-Event Transients in an Epitaxial Silicon Diode Resulting from Heavy Ion-, Focused X-Ray-, and Pulsed Laser-Induced Charge Generation

Heavy ion, focused x-ray, and pulsed laser single event transient experiments are performed on a silicon epitaxial diode. Collected charge, transient rise times, and transient fall times are calculated and compared between the different sources. It is observed that these transient shape characteristics depend on the source (ion, x-ray, or laser), even when similar amounts of charge are generated. Finally, the observed differences are examined and explained in terms of basic charge collection mechanisms.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Resistance heater rod and method of making such

A resistive heater capable of delivering heat loads on the same order as those produced by in-pile nuclear fuel experiments. The heater rod provides the energy for high-temperature steady-state testing, as well as the power needed to simulate the transient pulse in the Transient Reactor Test Loop (TRTL) system. The resistive heater includes a removable housing, two or more thermal conductors in the housing; and one or more stabilizers coupled to the two or more thermal conductors to keep the two or more thermal conductors separated to avoid shorting, wherein the two or more thermal conductors are coupled to the housing via an inert gas (e.g., Helium). The two or more thermal conductors comprise a material with substantially zero infrared spectrum (e.g., sapphire, silica, or glass).

LaBrier, Daniel↗