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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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Evaluation of Control Rod Designs for the Versatile Test Reactor
Different fractions of boron carbide were investigated for the control rod assemblies of the Versatile Test Reactor (VTR). Reducing the number of rods per assembly increases the overall volume fraction of B4C, thus increasing reactivity worth. Doing so also reduces the melting margin inside the pellets and the reactivity worth loss by EOL. As a result the control rod layouts with the most (CR61) and least (CR7) number of rods were deemed less favorable for each of those two reasons. Overall, CR19 was found to have sufficient safety margins with the second best reactivity worth. The design only required a 44% 10B enrichment to reach the conservative margins put forward. The arrangement is recommended for both the control and safety rods of the VTR.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
High-level fuel fabrication facility designs from discrete-event simulation
Like other industrial processes, the production of metallic nuclear fuels (MNF) requires that fabrication facilities be able to reliably meet production demands, operate efficiently, and adhere to federal and local safety regulations. In turn, the set of design variables employed by such a facility, such as operations policies, infrastructure and machinery purchased, and the type and number of staff hired, directly impact a facility’s ability to satisfy these goals. Therefore, facility designers must carefully determine which set of design variable values optimally satisfies these constraints. In this paper, we explore how values for these high-level design variables, namely hiring requirements, can be determined in the context of nuclear fuel manufacturing through the coupling of physics-based and discrete-event simulation technologies. Using the Versatile Test Reactor (VTR) program as a case study, we demonstrate how SCALE and MCNP nuclear physics model outputs can be integrated into ExtendSim discrete-event simulation (DES) models of the fuel fabrication process to determine the optimal number of staff hired to ensure fuel production goals are met, operations comply with effective dose limit regulations, and overall project costs are reduced.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NRIC DOME Crane Trade Study and Recommendation
The National Reactor Innovation Center (NRIC) is a national program that was established as part of the Nuclear Energy Innovation Capabilities Act (NEICA). NRIC’s mission is to accelerate the demonstration and deployment of advanced nuclear energy through its mission to inspire stakeholders and the public, empower innovators, and deliver successful outcomes through efficient coordination of partners and resources. NRIC is designed to bridge the gap between research, development, and the marketplace to help convert some of the nation’s most promising advanced nuclear reactors into commercial applications. The NRIC Demonstration of Microreactor Experiments (DOME) facility, formerly known as Experimental Breeder Reactor II (EBR II), located at the Materials and Fuels Complex (MFC) at the Idaho National Laboratory (INL) is intended to allow industrial and other partners the opportunity to test Advanced Microreactors up to 20MW thermal power. The 75-ton capacity polar crane located in the EBR II facility was rendered inoperable to support planned facility demolition in 2015; small holes were flame cut in the girders; hoists and cables were removed; oil was drained from gear boxes; trolley drive, and then filled with absorbent; and the electrical and control umbilical’s were disconnected, removed and disposed. Subsequently, the decision was made to convert the EBR II facility into the DOME test bed.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
Recrystallization driven softening and heating rate dependencies of FeCrAl nuclear fuel cladding during accident transients
A refined understanding of FeCrAl cladding behavior during rapid transients is critical for its potential deployment in light-water reactors. Current assessments focus on transient burst testing metrics such as balloon geometry, burst temperature, and hoop stress, often used as proxies for simpler conventional tensile properties. However, directly correlating isothermal tensile and creep data with accident transient scenarios remains a challenge, although it is essential for high fidelity model development. Recent modeling based on tensile tests up to 800 °C, conducted with both immediate loading and a 10-minute soak, showed that immediate loading better predicts experimental burst temperatures, indicating a thermal softening effect. Building upon this observation, the current study connects transient performance, microstructural evolution, and high-temperature tensile properties by leveraging results from C26M claddings burst tests performed at heating rates of 1–50 °C/s and hoop stresses from 25 to 100 MPa. At 25 MPa, rupture temperatures varied by only 6 °C, but at 100 MPa, the difference reached 116 °C, with faster heating yielding higher burst temperatures. Microstructural analysis identified recrystallization as the primary cause of heating rate-dependent softening, eliminating prior cold-working. In-situ thermomechanical data linked ballooning onset to localized instabilities, similar to ultimate tensile strength behavior in conventional tensile tests. High heating rates correlated with immediate loading tensile data, while lower rates matched soaked data. Furthermore, by linking burst performance to microstructural evolution and tensile properties, this work provides a foundation for more accurate modeling of FeCrAl claddings and potentially other Fe-based materials under accident conditions.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
SPC-70783 Rev 0 Seismic Detector and Switch Survey and Verification Specification
A. Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core and Stirling engines under the Microreactor Application Research Validation and Evaluation (MARVEL) program. The MARVEL microreactor Reactor Protection System (RPS) will have a seismic trip function. The selected part for this design is ETNA2 manufactured by Kinemetrics, Inc. B. Acceptance of the items delivered under this scope of work can be performed through a direct evaluation of the supplier’s controls and a factory acceptance test.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
Testing Instrument Extremes in the TREAT Facility
TREAT is a transient power-shaping reactor whose primary mission is to enhance safety performance by testing nuclear fuels and materials under thermodynamic and neutronic conditions ranging from off-normal to extreme. Transient testing of nuclear fuels is analogous to car crash testing, in that the dramatic changes commonly seen when comparing the test specimens' initial and final state points necessitate in-situ measuring in order to interpret and understand the evolution of the experiment. TREAT’s experiment design strategy utilizes a highly reconfigurable and accessible reactor core that affords flexibility when installing experiment devices, thanks to the lack of a primary coolant boundary and the inclusion of multiple access points in the reactor bioshield. Within the experiment devices, the test specimens (which are integrated with their test environments), desired instrumentation diagnostics, and safety considerations are all contained in a single engineered package. The experiments encompass various types of environments (e.g., water, sodium, and gas at temperature and pressure), as well as a range of control options (e.g., static or flowing coolants). The experimental conditions are uniquely representative of advanced reactors, providing challenges to sensor performance but also affording a unique opportunity for development and qualification of sensors suited to these environments.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Improved Creep Testing Approach for Bentonite EBS
In the United States, the nuclear reactor contributes 18.7 percent of the nation’s total electricity generation and produces a significant amount of spent nuclear fuel (SNF) and high-level waste. Because SNF is being stored for longer periods than initially envisioned, the U.S. Department of Energy Office of Nuclear Energy, Office of Spent Fuel and Waste Science and Technology is assessing the technical performance of the SNF storage systems after extended durations. The concept of a sustainable geologic repository for SNF disposal relies heavily on the safety provided by multiple barriers, such as repository host rock, overlying rock formation, and human-made engineered barrier systems (EBS). One of the major components of EBS is bentonite-based buffer materials that isolate the nuclear-waste canisters from the host rock and fill the void left in the horizontally drilled boreholes. The resaturated bentonite has a swelling characteristic, which is useful for self-sealing the microcracks in the host rock and supporting the canister’s heavy weight. However, factors such as nonuniformity in the saturation, swelling pressure of bentonite, and the high temperature at the canister-bentonite interface may lead to instability of the waste canister inside the borehole. Therefore, evaluating the long-term deformation of bentonite-based material is necessary. RESPEC Company, LLC recently completed a preliminary study for the U.S. Department of Energy (DE SC0022804) that is attempting to understand the time-dependent deformation in the consolidated sand/bentonite (SB) specimens through triaxial creep experiments and improve the understanding of EBS performance under anticipated repository conditions. The project included developing a standard testing procedure for preparing the consolidated core specimen from a mixture of sand and bentonite in a 50:50 ratio by weight, fabricating a new pressure vessel in-house, modifying existing creep equipment to perform triaxial creep experiments on consolidated SB specimens, and performing six long-term triaxial creep experiments under low-deviatoric stress at two different temperature conditions, which is representative of hypothetical conditions that might be encountered in a geologic repository within a reasonable rate of success in completing creep experiments. The consolidated SB specimen was prepared using the novel specimen preparation procedure and had physical characteristics (e.g., moisture content and bulk density) reasonably comparable with the properties of bentonite-based buffer, which is proposed to be used in the EBS in the Swedish KBS-3H design and the Swiss design of SNF disposal in the horizontal drifts. The newly fabricated pressure vessel could withstand the confining pressure of up to 15 megapascals (MPa) while maintaining the test temperature of up to 90 degrees Celsius (°C) for long-term creep experiments. The preliminary results from the triaxial creep experiments suggested that the consolidated SB specimen may experience time-dependent deformation at a low-deviatoric stress state and room temperature condition. Based on the deformation recorded in the SB specimen from axial and radial linear variable differential transformers (LVDTs) and physical characteristic changes, it is apparent that the time-dependent deformation is a combination of consolidation and creep, which is influenced by the level of deviatoric stresses and temperature. The systematic creep testing of bentonite-based specimens helped develop a standard testing procedure applicable for analyzing the similar characteristics of shale or clay-like materials in different geologic conditions outside repository science, including in civil engineering and infrastructure projects, underground and open pit mining, and deep drilling.
58 GEOSCIENCES↗
NEP Space Test Program Objective
The objective of the Nuclear Electric Propulsion (NEP) space test program is to launch an NEP satellite powered by the Russion Topaz 2 reactor by Dec. 1995. The primary goals of the NEP space test program are as follows: (1) demonstrate the feasibility of launching a space nuclear power system; (2) demonstrate and orbit adjust capability using NEP; (3) evaluate the in-orbit performance of the Topaz 2 reactor and selected electric thrusters; and (4) measure, analyze, and model the NEP self-induced environment. The discussion is presented in vugraph form.
Source record↗
Report Summarizing Boiling Transition (Dryout) Testing of FeCrAl Cladding
Nuclear fuel and cladding in a light water reactor must maintain integrity under all normal operational and anticipated operational occurrences. In a Boiling Water Reactor (BWR) some Anticipated Operational Occurrences (AOOs) and Anticipated Transients Without SCRAM (ATWS) events have been known to result in cladding damage due to dryout with conventional Zircaloy-2 claddings. This work presents an assessment of commercially-fabricated, accident tolerant iron-chromium-aluminum (FeCrAl) alloy C26M cladding in such events with regard to the ballooning failure mode for unirradiated cladding In the severe accident test station (SATS), an experiment has been developed to gauge accident tolerance of fuel cladding materials under simulated cyclic dryout conditions. Thermal cycling from 300°C to 650°C was implemented on C26M and Zircaloy-2 tubes. The C26M tube did not fail after 54 cycles. The Zircaloy- 2 tube run under the same parameters as the C26M tube failed during the fourth cycle, consistent with expectations from the literature. Due to thermal gradients present in the furnace, both tubes experienced a maximum temperature at ~780°C. Due to differences in tube geometry, the hoop stress of the C26M tube was significantly higher than that of the Zircaloy-2 tube, 76 MPa at the maximum peak temperature compared to 55 MPa at burst, respectively. The cyclic dryout experiments show an increased performance of C26M over Zircaloy-2 under simulated conditions, an indication of C26M possessing superior high temperature oxidation resistance, mechanical properties and potentially resistance to fatigue.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Advanced Instrumentation for Transient Reactor Testing
Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Contribution of IAEA coordinated research projects to light water reactors advanced technology fuel testing and simulation
The Fukushima-Daiichi accident in 2011 highlighted the pressing need for enhanced design and safety analysis of nuclear fuels, especially under accident conditions in nuclear power plants (NPPs). To address this need, the international nuclear fuel community has developed several concepts of Accident Tolerant and Advanced Technology Fuels (ATFs) for light water reactors. The International Atomic Energy Agency (IAEA), at the request of its Member States, has initiated a series of three Coordinated Research Projects (CRPs) to aid in the development and testing of these ATFs. Further, these projects, known as FUMAC, ACTOF, and ATF-TS, focus on the experimentation and simulation of ATFs under various accident conditions, including design basis accidents and design extension conditions. This paper provides a comprehensive overview of the main activities, as well as the achieved or anticipated results of these three CRPs. The insights and advancements gained from these projects will enable IAEA Member States to further refine nuclear fission technology, thereby contributing to efforts aimed at mitigating climate change.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
SMR Re-Scaling and Modeling for Load Following Studies
This study investigates the creation of a new set of scaling parameters for the Oregon State University Multi-Application Small Light Water Reactor (MASLWR) scaled thermal hydraulic test facility. As part of a study being undertaken by Idaho National Lab involving nuclear reactor load following characteristics, full power operations need to be simulated, and therefore properly scaled. Presented here is the scaling analysis and plans for RELAP5-3D simulation.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Development of Ultra Small Scale Mechanical Testing and Localized He Implantation for Nuclear Applications
Mechanical properties degradation of components deployed in nuclear applications has been a concern since the deployment of first generation nuclear power plants. Radiation damage along with environmental degradation, elevated temperatures, and long term deployment of nuclear structures pose a challenging environment for all materials in these applications. Traditionally, the mechanical properties are assessed by standardized macroscopic testing methods. While those are providing engineering relevant data these techniques do have limitations, especially on reactor irradiated materials, due to the large size of the samples. Limitations on the amount of radioactivity of large samples must be applied for research applications as they are conducted at universities. In addition it can be difficult to retrieve sufficiently large amounts of material from an operating reactor component. Therefore smaller sample sizes are desirable and reduce the amount of radioactive material which needs to be handled or material which needs to be retrieved. In addition, the number of tests which can be performed on materials irradiated in a test reactor program is limited due to limited space in specific dose and temperature positions. Downsizing of the mechanical test can overcome this issue and perform more tests on the same amount of irradiated materials, reducing the statistical uncertainties. In addition, smaller scale samples allow access to specific regions of interest in a material or component, be it the zone in front of a crack tip or the sampling of very small components like springs or welded areas. In addition ion beam irradiation has become an integral part of nuclear materials research to enhance the understanding of radiation damage in materials in a time and cost effective fashion. The drawback of ion beam irradiation is the small penetration depth of the ions into the material making localized microstructural characterization as well as small scale mechanical testing necessary. Another more fundamental motivation is the fact that since the beginning of electron microscopy, one has been concerned about imaging and quantifying defects in materials. Multiscale tools to image a large range of radiation induced defects have been established starting from atom probe tomography to transmission electron microscopy, up to scattering techniques. Mechanical testing, however, has always been performed on a large scale, and an attempt to match the same scale where a defect is observed with the materials' mechanical property response has not yet been realized. However, the question of separate effects testing on specific radiation induced defects remains and is an additional driving force for small scale mechanical testing. In this work we present an approach for small scale mechanical testing in irradiated materials and the variety of possible test methods developed today. We also present recent efforts of performing localized ion beam implantation to achieve a number of different doses in the same region of interest enhancing comparability and quantification of observed phenomena. It is the intention to give the reader an overview of the different microscale mechanical testing techniques possible today as well as an introduction of a novel ion beam irradiation method using a newly established capability. In this paper we can't discuss all aspects of all small scale mechanical testing of irradiated materials However, It does give an overview over the techniques enabled due to recent effort and enhanced instrument capability. While indentation or micro compression testing has been performed previously tensile testing and ion beam irradiation using the He ion beam microscope are novel approaches not attempted previously. (authors)
36 MATERIALS SCIENCE↗
High-temperature irradiation-resistant thermocouple instability model for in-pile reactor use
This article presents an instability model for the high-temperature irradiation-resistant thermocouple (HTIR-TC). Here the term instability defines the superposition of both drift and inhomogeneity of TC thermoelements occurring simultaneously. The HTIR-TC is an advanced thermocouple (TC) that uses the refractory metals niobium and molybdenum as sensing thermoelements for generating electromotive force (EMF) in a field of neutrons and at temperatures upward of 1,600°C. In the Advanced Gas Reactor (AGR) 5/6/7 tests conducted at Idaho National Laboratory’s Advanced Test Reactor (ATR), the HTIR-TCs showed low to moderate instability throughout the life of the test. The instability model reveals that HTIR-TCs can, when the operating temperature of the reactor fuel is normal, maintain performance throughout an 18-month refueling cycle typical of nuclear power plants, reflecting an instability of less than ±1%. The HTIR-TC is also qualified for incorporation into a test fixture during the testing of new fuels.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
LANDSAT-4 image data quality analysis for energy related applications
No useable LANDSAT 4 TM data were obtained for the Hanford site in the Columbia Plateau region, but TM simulator data for a Virginia Electric Company nuclear power plant was used to test image processing algorithms. Principal component analyses of this data set clearly indicated that thermal plumes in surface waters used for reactor cooling would be discrenible. Image processing and analysis programs were successfully testing using the 7 band Arkansas test scene and preliminary analysis of TM data for the Savanah River Plant shows that current interactive, image enhancement, analysis and integration techniques can be effectively used for LANDSAT 4 data. Thermal band data appear adequate for gross estimates of thermal changes occurring near operating nuclear facilities especially in surface water bodies being used for reactor cooling purposes. Additional image processing software was written and tested which provides for more rapid and effective analysis of the 7 band TM data.
Wukelic, G. E.↗
Software Requirements for a Point Kinetics Neutronics Code for Simulating TREAT Reactor Transients
This document provides the software requirements for a computer code to perform point kinetics-based simulations of nuclear reactor transients. These requirements address all aspects of code development and use, including input and output parameters and structure, software capabilities, and supporting documentation. This software is intended to replace the existing point kinetics code TREKIN, which is currently used for the analysis of the Transient Reactor Test Facility (TREAT).
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Sample Selection Report for the Irradiation and Post Irradiation Examination of Ultra High Burn up Fuel
This report summarizes the available data for fuel pins available to the project and selects samples for integral LOCA tests, heating tests, and irradiation in the Advanced Test Reactor.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗