MARVEL Thermal-hydraulics and PCAT
Provide a brief overview of MARVEL thermal-hydraulics, focusing on MARVEL integral test faciliy PCAT.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
SEARCH · Search NASA
Results for “Nuclear Reactor 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.
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LOCA and LOFA Safety Margin Analysis for U-10Mo Monolithic LEU Fuel Conversion of the MURR{sup R} Reactor
The University of Missouri Research Reactor (MURR{sup R}) is a multi-disciplinary research and education facility providing a broad range of analytical and irradiation services to the research community and the commercial sector. The source of neutrons for the experimental facilities of MURR{sup R} is a compact 10 MW core fueled with highly-enriched uranium (HEU) fuel. The continued presence of HEU fuel in civilian installations such as research and test reactors poses a threat to national and international security. Therefore, most research and test reactors, both domestic and international, have started a program of conversion to the use of low-enriched uranium (LEU) fuel. MURR{sup R} is one of five U.S. high-performance research and test reactors (USHPRRs) that are actively collaborating with the U.S. National Nuclear Security Agency (NNSA) Office of Material Management and Minimization (M3) to develop a suitable LEU fuel replacement for the currently required HEU fuel. A new type of high-density LEU fuel based on an alloy of uranium and molybdenum is currently undergoing qualification testing, and is expected to allow the conversion of all USHPRRs. As a part of the LEU conversion of MURR{sup R}, an analysis of core behavior under a set of accident conditions for a core fueled with an LEU monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) has been conducted. This paper presents the newest results for the loss-of-coolant accidents (LOCAs) and loss-of-flow accidents (LOFAs). Other classes of accidents have also been investigated but were not included here. Namely, reactivity insertion accidents (RIAs), fuel failure during operations accidents (FFAs), and fuel handling accidents (FHAs)
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Radioactivity Calculator for Research and Test Reactor Neutron-Activated Materials
The Idaho National Laboratory Nuclear Science User Facilities program is proposing to launch a new online expert system that will assist users in planning experiments and writing proposals. The Combined Materials Experimentation Toolkit (CoMET) will include two nuclear calculators. The first of these, the Neutron Damage Calculator, will assist users in selecting reactor positions in order to obtain desired material damage in Damage Per Atom (DPA) and provide the required irradiation time to reach the damage levels. The second, the Radioactivity Calculator, will estimate the post-irradiation radioactivity levels of sample materials, which will inform technical staff of examination facility requirements and any potential scheduling delays. Both of these calculators were designed to simplify the information needed from users in order to properly plan and execute awarded proposals. This paper presents the primary purpose and methodologies implemented by these two calculators.
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Instrumentation Development for Transient Testing in TREAT
Development of nuclear fuels requires experimentation and modeling for behavior during transient power and cooling conditions. The Transient Reactor Test facility (TREAT) will provide unique capability in the U.S. to perform controlled experimentation of such conditions in a nuclear environment. To support experiment objectives and provide validation data for modeling and simulation tools, experiment instrumentation plays a critical role in monitoring test specimen behavior (temperature, dimensional changes, relocation, microstructural evolution, etc.) and specimen boundary conditions (thermal, mechanical, nuclear) at various length and time scales with the ultimate goal of reducing uncertainties in experiment conditions. This summary presents an overview of current instrument development objectives and progress for TREAT experiments with emphasis on near-term activities to support Reactivity Initiated Accident (RIA) experiments in LWR systems. (authors)
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Validation Testing for Molten Chloride Reactor Experiment Equipment Removal and Disposal Techniques
The Molten Chloride Reactor Experiment (MCRE) will be the first reactor featuring a fast-spectrum molten chloride circulating nuclear fuel system in the world. Planning for equipment removal and disposal (ERD) of MCRE has identified several technology gaps due to the unique environment of this nuclear experiment. Some of the gaps arise from the application of existing disassembly and/or sizing methods to novel material forms or in novel configurations. Others arise from unknown material behavior. This paper summarizes proposed test plans for ERD validation experiments to address these complicated or unknown equipment removal procedures. At the Waste Management Symposia in 2024, the Idaho National Laboratory (INL) MCRE ERD team presented the challenges associated with hosting multiple nuclear experiments in series with only brief transition periods between systems. Such difficulties include higher dose rates, the presence of radioisotopes infrequently encountered in reactor decommissioning and radioactive waste management, lack of intrinsic remote-operations infrastructure in the test bed, space constraints in the test bed, and contamination minimization requirements. To address these challenges, remote or semi-remote technologies are planned to be implemented in a non-hot cell environment with limited space availability. The team also discussed how a systems engineering approach is being used for conceptual development and design of equipment removal systems to address these challenges. For example, to reduce constraints for the removal of more difficult components, non-activated, noncontaminated elements are planned to be taken out first where possible. Still, there are complexities associated with the remaining components. In this work, the operational framework for MCRE ERD was reviewed for technical gaps and open questions, and test plans were drafted to address these areas. The tests plans were written for the following categories: vision systems, pipe cutting, drill/grout/filler, flush salt, and miscellaneous, with the miscellaneous group consisting of tests like techniques for removing bearings and reflector bricks. The test plans explore material, infrastructure, and staffing requirements needed for test execution. The test plans additionally focus on the evaluation of success. Determining the outcome of a test is imperative - as these explorative actions have the potential to rearrange or re-scope planned ERD activities. Success criteria identified thus far include required tool output, required area(s), debris production and mitigation, and repeatability. Test plans are an essential aspect of the systems engineering approach to MCRE ERD. They are used as the beginning steps in defining use cases for the ERD system. Performance of the validation tests is expected to begin in the summer of 2025 and will take approximately 9 to 12 months to complete. Execution of these plans will be expedited by specifying test needs ahead of time, facilitating efficient interactions with any subcontractors tasked with running the requested tests. Evaluating the outcomes of these tests will inform MCRE ERD procedures and timing and will also identify additional technical constraints for the MCRE ERD System. This upfront process optimization effort will help the project save time and resources at the end of the experiment.
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A Kuiper Belt Object Orbiter Enabled By 10 kW Kilopower Electric Propulsion
Recently, the New Horizons spacecraft flew by the Kuiper Belt Object Ultima Thule 13 years after launch. While flybys are a necessary 'first look' at the object in question a more comprehensive evaluation of the object will require an orbiter. Due to these object’s low mass a chemical propulsion system is not a viable option for entering their shallow gravity wells. The Dawn mission showed that solar electric propulsion is significantly better at the task of reaching low mass objects, but at Kuiper belt distances solar power is not a viable choice. A small nuclear reactor based on the recent kilopower ground test could provide 1-10 kWe of power for an electric propulsion system. The NASA Compass Team developed a Nuclear Electric Propulsion Kuiper Belt Object Orbiter to explore what a vehicle would look like to orbit these deep space objects.
Kuiper belt Object↗
A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing
Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. Thirteen additional tests would be performed in the CABRI reactor over the next decade on fuel rods with higher burnups [1]. Additionally, a larger testing program at the NSRR reactor in Japan with high burnup fuel would uncover a similar trend of pre-DNB ruptures in high burnup test rods at lower-than-expected peak enthalpies [2]. Numerous out of pile testing programs involving a variety of innovative mechanical testing techniques have been employed in an attempt to better quantify the failure thresholds of corroded zirconium alloy cladding in these rapid heating and loading conditions [3][4][5]. While previously interim guidance had been issued, in June of 2020 the NRC officially published updated regulatory guidance to account of these pre-DNB failures in Regulatory Guide 1.236 [6]. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.
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Development of The DOME Shield Model For The NRIC Virtual Test Bed
As several advanced reactor concepts are maturing, test beds are needed to accelerate the demonstration and deployment of these advanced nuclear technologies. The National Reactor Innovation Center (NRIC) is building new or enhancing existing US Department of Energy infrastructure to support testing of components and systems. Demonstration of Microreactor Experiments (DOME) will utilize the Experimental Breeder Reactor-II (EBR-II) dome containment structure to host reactor demonstrations. A reactor supplemental shielding is needed so that DOME dose requirements are met. To accelerate the confirmatory analysis required for the reactor demonstration, the NRIC Virtual Test Bed (VTB) is developing a virtual model of the DOME shield that will be made available on the VTB public repository. This will allow developers to leverage advanced modeling and simulation tools to ensure their reactor demonstration concept will meet dose requirement and the limit concrete temperature in the shield during steady state and transient operation conditions. This paper presents the model developed for the DOME shield using open-source tools: MOOSE heat transfer module, Monte Carlo code OpenMC, and Cardinal to calculate the DOME shield temperature distribution during steady state
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In-pile measurements of fuel rod dimensional changes utilizing the test reactor loop pressure for motion - 359
Different types of materials are being considered for fuel, cladding, and structures in existing and advanced nuclear reactors. These materials can undergo significant dimensional changes during irradiation. Currently in the US, such changes are measured with the 'cook and look' method (i.e. repeatedly irradiating a specimen for a specified period of time, and then removing it from the reactor for evaluation) - which leads to costly and time consuming experiments. In addition, such techniques provide limited data, and handling may disturb the phenomena of interest. Therefore, in-pile detection of geometric changes is needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). The developmental results of an advanced Linear Variable Differential Transformer (LVDT)-based test rig capable of detecting real-time changes in diameter or length of fuel rods or other material samples during irradiation in US MTRs are presented. The LVDT is being developed at the Idaho National Laboratory (INL), and will provide experimenters with the capability to measure dimensional changes associated with fuel and clad swelling, pellet-clad interaction, and crud buildup. The LVDT design utilizes the MTR's loop pressure to drive the test specimen at a constant rate. This leads to a simplified construction which only requires three pressure boundary penetrations. Further, the LVDT pressure boundary is a leak tight hydraulic ram capable of actuating under low differential pressure (e.g. 1 psid), and is able to survive the high irradiation tests at Pressurized Water Reactor (PWR) conditions. (authors)
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Status Report for NEET In-Pile Sensor Irradiation Capabilities at TREAT
The recently-established in-pile sensor development project under the Nuclear Energy Enabling Technology (NEET) program is focused on the development of measurement technologies to enable acquisition of data focused on the performance, properties, and boundary conditions of nuclear fuels/materials under irradiation in nuclear reactors. These sensors can be used in steady state irradiation tests to measure parameters of interest under conditions representing normal nuclear plant operation, or in transient irradiation tests to determine specimen performance under postulated unplanned conditions. In either case, development of suitable sensor technologies must include instrument irradiations to help guide sensor material/design selection, verify instrument performance, and demonstrate deployment viability. The Transient Reactor Test facility (TREAT) while just a part of the overall capabilities portfolio needed to enable this sensor development mission, has unique capabilities and characteristics that make it well poised for a number of instrument irradiation data objectives including sensor survivability screening, response characterization, qualification tests, and in-situ properties measurements. TREAT’s brilliantly-basic design, when married with appropriate irradiation hardware and analytic capabilities, fills a crucial capability gap between out-of-pile testing and irradiation in steady state high power research reactors. Leveraging a recently developed approach to irradiation testing in TREAT, the present work demonstrates that cost-effective and scientifically-impactful instrument irradiation capabilities for TREAT are a well-placed capability in the NEET in-pile sensor development project.
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Phase II Development of the Surveillance Test Articles to Improve the Design, Fabrication, and Testing
Advanced reactors, such as the molten salt reactor (MSR), require materials that will withstand harsher environments than the materials used for lower temperature water reactors. Materials used in MSR construction need not only withstand elevated temperature, temperature cycling, and neutron radiation, but must be able to withstand the corrosive molten salt environment. Significant materials research has been driven by the materials needs of the MSR due to the harsh environment and the material data requirements to support licensing. Limited operation experience with MSRs has made this challenging work. Information is very limited on materials degradation due to irradiation, molten salt corrosion, elevated temperature, and the resulting fatigue, creep, and creep-fatigue loading during operation. While efforts are underway to better understand the effects of this harsh environment on construction materials, the Advanced Reactors Technologies (ART) Program has been working to develop materials surveillance test articles that could be used in a materials surveillance program and allow for the collection of information on the materials degradation during plant operation and could support timely licensing of these advanced reactors.
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Completion of a Chlorination Test Using 250 grams of High-Burnup Used Fuel Cladding from a North Anna Pressurized Water Reactor
A second large -scale chlorination test was completed in February 2017 by chlorinating 250 g of used nuclear fuel (UNF) cladding from a North Anna pressurized water reactor (PWR). Analytical results and a Zr mass balance from this test are described and compared with previous results obtained by the Zirconium Recovery and Recycle Project.
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Primer for OTERR Core Reloading Optimization
OTERR (Optimization of Test Reactor Reloading) is a software tool which assists in finding optimal fuel assembly reloading patterns for nuclear reactors. It was created specifically to support the Versatile Test Reactor (VTR) program, but its functionality is general enough to be applied to most fast spectrum reactors which use hexagonal prismatic fuel elements. This document is meant to be a primer for new users of OTERR to walk through example workflows for core reloading optimization cases. Simple cases are followed step-by-step to discuss what needs to be done to complete a reloading optimization sequence. Note that the intent of this document is to provide practical examples for users to follow along with so they can quickly start using OTERR and then make changes to fit their own modeling needs. Detail is limited in terms of addressing additional features not used in these examples and especially lacking in discussion of theory used in the code. To better address these points, users are highly encouraged to refer to the OTERR User Manual and the OTERR Theory Manual.
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Modeling and Simulation to Support Current and Advanced Reactors : Thermal Hydraulics
Presentation on thermohydraulic modeling and simulation capabilities in support of current and advanced reactors. The presentation includes information on the capabilities of RELAP5-3D, MOOSE and the experimental testbed for this purpose. This presentation is intended for a Taiwan delegation that will be visiting INL in August 2024.
42 ENGINEERING↗
Long-term Prediction of Emissivity of Structural materials for High Temperature Reactor Systems
The purpose of this research was to extend the recent short-term studies of emissivity of reactor materials performed independently at the University of Missouri and University of Wisconsin to tests aimed at longerterm predictions of emissivity. The University of Missouri measured hemispherical emissivity, which provides an integrated measurement, while the University of Wisconsin measured spectral emissivity (2 to 10 μm wavelength) from which the integrated quantity can be obtained. Hemispherical emissivity is needed for MELCOR, RELAP and FLUENT type heat transfer code calculations, while spectral emissivity is needed for a detailed Monte Carlo Transport theory computation of heat transfer and fundamental modeling verification. The objectives of the study were: Objective 1: Defining materials for long-term emissivity in cross-cutting reactor applications - procurement, preparation, and analyses of materials. Objective 2: High temperature materials testing in various environments and post-test surface analysis Objective 3: Independent hemispherical and spectral emissivity testing. Objective 4: Development of models to gain a predictive capability for long-term emissivity and application and validation in current RCCS research programs. The research has been performed at two universities over a four-year term with the participation of students, scientific staff, and faculty. The report describes progress in realizing the objectives, and the accomplishments at the two universities. There was considerable progress with respect to Objectives 1-3. The model development was redirected towards modeling of interrelationships between the spectra and hemispherical emissivities and uncertainty analysis of total hemispherical emissivity measurements.
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NSUF 2016 Annual Report
The Nuclear Science User Facilities (NSUF) and its partner facilities represent a prototype laboratory that utilizes a distributed partnership, with each facility bringing exceptional capabilities to the relationship including reactors, beamlines, state-of-the-art instruments, hot cells, and equally important, expertise. These capabilities and people together create a nationwide infrastructure that allows the best ideas to be tested using the most advanced capabilities. Through the NSUF, researchers and their collaborators are building on current knowledge to better understand the complex behavior of materials and fuels in a nuclear reactor. The NSUF’s partnership program in 2016 included eight universities, one research and education consortium, three national laboratories, and one industry partner. The avenues opened through these partnerships facilitate cooperative research across the country, matching people with capabilities and students with mentors. The NSUF in 2016 included Idaho National Laboratory (INL) and the following institutions: • Argonne National Laboratory • Center for Advanced Energy Studies (a research and education consortium between Boise State University, INL, Idaho State University, University of Idaho, and University of Wyoming) • Illinois Institute of Technology • Massachusetts Institute of Technology • North Carolina State University • Oak Ridge National Laboratory • Pacific Northwest National Laboratory • Purdue University • University of California, Berkeley • University of Michigan • University of Nevada, Las Vegas • University of Wisconsin • Westinghouse Materials Center of Excellence. This report contains details on new capabilities brought into the NSUF in 2016, the evidence of the increasing research facilitated by the NSUF, and new tools and activities to propel nuclear research and development forward to meet energy reliability and security needs. Learn about the NSUF Nuclear Fuels and Materials Library and the Nuclear Energy Infrastructure Database, and how the robust NSUF website continues to evolve to support new tools. Read about how the NSUF supports Department of Energy (DOE) missions through efforts like the 2016 NSUF Ion Beam Options Workshop, which led to a prioritiza-tion of ion beam facilities according to their ability to support Department of Energy Office of Nuclear Energy (DOE-NE) missions.
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Perspectives on Tailoring Neutron Energy Spectra in Material Test Reactors
Material test reactors (MTRs) are used to irradiate nuclear fuels and materials to develop data for how they endure neutron bombardment in or near reactor cores. Most historic MTRs, and all that remain operational in the US today, are water-cooled types and produce a thermalized neutron flux. New irradiation facilities are needed which can produce neutron energy spectra relevant to fast and fusion reactor environments. Construction of these facilities will take several years of steadfast funding to complete, which poses a formidable schedule challenge for current fast and fusion reactor developers. Irradiation designs which modify the neutron energy spectra delivered to test specimens in thermal spectrum MTRs, an approach referred to as “spectral tailoring”, can be used to approximate several relevant phenomena in the materials needed to enable fast and fusion reactor technologies. This approach is imperfect, but still valuable in the present situation. Here, the two highest flux MTRs operational in the United States, the Advanced Test Reactor (ATR) and High Flux Isotope Reactor (HFIR), have rich histories, ongoing developments, and new potentials for spectral tailoring that will be reviewed and discussed in this paper.
36 MATERIALS SCIENCE↗
Preliminary neutronic analysis of a cavity test reactor
A reference configuration was calculated for a cavity test reactor to be used for testing the gascore nuclear rocket concept. A thermal flux of 4.1 x 10 to the 14th power neutrons per square centimeter per second in the cavity was provided by a driver fuel loading of 6.4 kg of enriched uranium in MTR fuel elements. The reactor was moderated and cooled by heavy water and reflected with 25.4 cm of beryllium. Power generation of 41.3 MW in the driver fuel is rejected to a heat sink. Design effort was directed toward minimization of driver power while maintaining 2.7 MW in the cavity during a test run. Ancillary data on material reactivity worths, reactivity coefficients, flux spectra, and power distributions are reported.
Whitmarsh, C. L., Jr.↗