Comparison of PM-HIP to forged SA508 pressure vessel steel under high-dose neutron irradiation
Not Available
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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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A methodology for uncertainty quantification and sensitivity analysis for responses subject to Monte Carlo uncertainty with application to fuel plate characteristics in the ATRC
Not Available
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Nitrogen ion implantation into various materials using 28 GHz electron cyclotron resonance ion source
The installation of the 28 GHz electron cyclotron resonance ion source (ECRIS) ion implantation beamline was recently completed at the Korea Basic Science Institute. The apparatus contains a beam monitoring system and a sample holder for the ion implantation process. The new implantation system can function as a multipurpose tool since it can implant a variety of ions, ranging hydrogen to uranium, into different materials with precise control and with implantation areas as large as 1–10 mm{sup 2}. The implantation chamber was designed to measure the beam properties with a diagnostic system as well as to perform ion implantation with an in situ system including a mass spectrometer. This advanced implantation system can be employed in novel applications, including the production of a variety of new materials such as metals, polymers, and ceramics and the irradiation testing and fabrication of structural and functional materials to be used in future nuclear fusion reactors. In this investigation, the first nitrogen ion implantation experiments were conducted using the new system. The 28 GHz ECRIS implanted low-energy, multi-charged nitrogen ions into copper, zinc, and cobalt substrates, and the ion implantation depth profiles were obtained. SRIM 2013 code was used to calculate the profiles under identical conditions, and the experimental and simulation results are presented and compared in this report. The depths and ranges of the ion distributions in the experimental and simulation results agree closely and demonstrate that the new system will enable the treatment of various substrates for advanced materials research.
36 MATERIALS SCIENCE↗
Anomaly Detection for Online Monitoring of Thermocouple Sensors in the Advanced Test Reactor
This study explores data-driven anomaly detection methods to analyze sensor fail- ures in the Advanced Gas Reactor (AGR) nuclear fuel irradiation experiments. Specifically, we examine failures of thermocouples (TCs), which are critical for mon- itoring and controlling in-reactor temperatures during operation. Failures were pri- marily observed during abrupt power transitions and manifested as sensor drop-outs, drifts, or unexplained behavior. We applied three time-series analysis techniques— rolling mean smoothing, matrix profile, and vector auto-regression (VAR)—to de- tect anomalies in TC data prior to failure events. The rolling mean method effec- tively highlighted deviations aligned with reported failures, while the matrix profile provided partial early warning but sometimes flagged normal fluctuations during power-down periods. VAR shows potential in capturing multivariate dependencies but requires further calibration. A rare case of TC drift was also documented, which did not result in failure, underscoring the challenge of building predictive models with sparse positive examples. Our findings demonstrate that traditional statistical tools can aid anomaly detection but have limited predictive power without richer training data. We propose future directions including synthetic data generation, real- time surrogate modeling, and multi-modal feature integration. This work provides a foundation for applying robust anomaly detection frameworks to mission-critical sensor systems in experimental settings.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Feasibility study on phyto-remediation techniques for soil contaminated by the Fukushima Dai-Ichi nuclear power plant accident
Tottori University and the Japan Atomic Energy Agency carried out jointly the feasibility study on phyto-remediation techniques, which apply to soil contaminated by the TEPCO's Fukushima Dai-ichi NPP accident. This paper illustrates the results from experimental investigations. Experimental investigations include both water-culture tests and field tests. Several plants, mainly halophytes that can specifically absorb more Na than K, and others like sunflower demonstrated for other domestic large-scale tests, were water-cultured and examined for screening. Easily cultivated and harvested plants without harmful effects on subsequent cultivation were also considered. New Zealand spinach was selected as a candidate for demonstrations in fields. The field tests were carried out at two sites of different agricultural types in Minami-soma, Fukushima prefecture. Concentration of {sup 137}Cs in soil is about 4.5 Bq/g-dry as the average of 10 cm depth. The aims of the field tests are to confirm absorption ability and environmental adaptation of the test plants and to document the cost and performance of projects. In conclusion, the absorption of {sup 137}Cs activity per unit area (Bq/m{sup 2}) by New Zealand spinach could be approximately 0.5%. To achieve an effective result in removal of {sup 137}Cs from soil in around a decade, it is required to find the plant which has ten or more times higher absorption capacity than New Zealand spinach. From the consistency of both results in water-culture and field tests, the water-culture test can be valid for screening. In addition, applicable sites will be limited to fields which are too steep or too narrow to use mechanical diggers, and which are free from any restrictions to enter. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
RELAP-7 Software Verification and Validation Plan
This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.
97 MATHEMATICS AND COMPUTING↗
KiloPower Project - KRUSTY Experiment Nuclear Design
This PowerPoint presentation covers the following topics: Reference Kilopower configuration; Reference KRUSTY configuration; KRUSTY design sensitivities; KRUSTY reactivity coefficients; KRUSTY criticality safety and control; KRUSTY core activation/dose; and KRUSTY shielding, room activation/dose.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Historic American engineering record. Nevada national security site, Bren Tower Complex. Written historical and descriptive data and field records
The BREN (Bare Reactor Experiment, Nevada) Tower Complex is significant for its role in the history of nuclear testing, radiation dosimetry studies, and early field testing of the Strategic Missile Defense System designs. At the time it was built in 1962, the 1,527 ft (465 m) BREN Tower was the tallest structure west of the Mississippi River and exceeded the height of the Empire State Building by 55 ft (17 m). It remains the tallest ever erected specifically for scientific purposes and was designed and built to facilitate the experimental dosimetry studies necessary for the development of accurate radiation dose rates for the survivors of Hiroshima and Nagasaki. The tower was a key component of the Atomic Bomb Casualty Commission’s (ABCC) mission to predict the health effects of radiation exposure. Moved to its current location in 1966, the crucial dosimetry studies continued with Operation HENRE (High Energy Neutron Reactions Experiment). These experiments and the data they generated became the basis for a dosimetry system called the Tentative 1965 Dose or more commonly the T65D model. Used to estimate radiation doses received by individuals, the T65D model was applied until the mid-1980s when it was replaced by a new dosimetry system known as DS86 based on the Monte Carlo method of dose rate calculation. However, the BREN Tower data are still used for verification of the validity of the DS86 model. In addition to its importance in radiation heath effects research, the BREN Tower Complex is also significant for its role in the Brilliant Pebbles research project, a major component of the Strategic Defense Initiative popularly known as the “Star Wars” Initiative. Instigated under the Reagan Administration, the program’s purpose was to develop a system to shield the United States and allies from a ballistic missile attack. The centerpiece of the Strategic Defense System was space-based, kinetic-kill vehicles. In 1991, BREN Tower was used for the tether tests of the Brilliant Pebbles prototype vehicle at the earth’s surface prior to the more costly space testing program. The success of these tests established the Brilliant Pebbles program as an essential component of America’s space-based missile defense system even after the dismantling of the Soviet Union. Data from the Brilliant Pebbles research program continues to inspire current missile defense system research (Independent Working Group 2009).
54 ENVIRONMENTAL SCIENCES↗
TREAT Cask Stand Drawings for Unlimited Use
This report contains drawings of the HFEF-15 Cask stand in the TREAT Reactor Building (MFC-720) and the Cask Stand Tower on the south wall of the TREAT High Bay. Public release is sought so these can be sent to DOD-SCO and their contracted companies looking at options for fueling the PELE reactor prior to demonstration in the DOME or elsewhere at INL.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding
The primary challenge preventing the deployment of reactor designs that leverage wrought FeCrAl as an advanced light-water reactor (LWR) cladding is irradiation hardening and embrittlement. Wrought FeCrAl alloys experience a loss of post-irradiation ductility and fracture toughness under low-temperature neutron irradiation (< 350°C)resulting from the combined effects of dislocation loop formation and the irradiation-enhanced precipitation of Cr-rich alpha-prime precipitates throughout the microstructure. Over the past decade, significant improvements in our understanding regarding the effect of Cr and Al content have been enabled through neutron irradiations, but even optimized wrought alloys such as C26M remain vulnerable to brittle failure during storage, transportation, and handling following irradiation.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Safety of rift vehicle launch and flight operations
Nuclear radiation safety of rift vehicle launch & flight operations
NUCLEAR RADIATION↗
Nuclear power for surface effect vehicle and aircraft propulsion
Nuclear surface effect vehicle and subsonic aircraft for transoceanic cargo shipping, discussing mobile reactor safety tests under high speed impact conditions
Rom, F. E.↗
A risk-informed approach to safety margins analysis
The Risk Informed Safety Margins Characterization (RISMC) Pathway is a systematic approach developed to characterize and quantify safety margins of nuclear power plant structures, systems and components. The model has been tested on the Advanced Test Reactor (ATR) at Idaho National Lab.
99 GENERAL AND MISCELLANEOUS↗
Assessment of Space Nuclear Thermal Propulsion Facility and Capability Needs
The development of a Nuclear Thermal Propulsion (NTP) system rests heavily upon being able to fabricate and demonstrate the performance of a high temperature nuclear fuel as well as demonstrating an integrated system prior to launch. A number of studies have been performed in the past which identified the facilities needed and the capabilities available to meet the needs and requirements identified at that time. Since that time, many facilities and capabilities within the Department of Energy have been removed or decommissioned. This paper provides a brief overview of the anticipated facility needs and identifies some promising concepts to be considered which could support the development of a nuclear thermal propulsion system. Detailed trade studies will need to be performed to support the decision making process.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Interfacial Area Transport through Vertical-Downward Elbows in Air-water Two-phase Flow
In the study of two-phase flow, models are often developed by the data acquired in a test section with fixed orientation and with no flow restrictions. In view of practical nuclear reactor systems, however, two-phase flow through flow restrictions is important. Flow restrictions and orientations can alter both global and local two-phase flow parameters such as: cause additional pressure loss; alter the flow regime transition; change bubble distributions; affect the bubble velocity and the bubble interaction mechanisms. However, there is insufficient database and predictive models that can account for such effects. In view of above, the objectives of the current study are: (1) to establish database for local and global two-phase flow parameters downstream of a horizontal-to-vertical glass elbow with a non-dimensional radius of curvature, R{sub c}/D, of three, (2) to develop predictive models in the elbow region, and (3) to apply the developed model to the interfacial area transport equation (IATE) and evaluate it in elbow regions. (authors)
42 ENGINEERING↗
The Annular Core Research Reactor (ACRR) Description and Capabilities
The Annular Core Research Reactor (ACRR) is a unique, one-of-a-kind nuclear reactor facility operated by Sandia National Laboratories for the National Nuclear Security Administration that is considered by many to be a national treasure. There is no other research reactor in the world that has the attributes and capabilities comparable to that of the ACRR. The ACRR was specifically designed to meet the irradiation testing needs of the U.S. nuclear weapons program. The ACRR has four major attributes that make it unique: 1) a large dry central cavity; 2) an epithermal neutron flux; 3) a large pulsing capability; and 4) a fueled-ring external cavity with a larger dry cavity. This report presents the unique capabilities of the ACRR and documents some of the more important metrics associated with its operation. ACKNOWLEDGEMENTS The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
FUEL PERFORMANCE MODELING OF REACTIVITY-INITIATED ACCIDENTS USING THE BISON CODE
The Fukushima Daiichi nuclear accidents in 2011 sparked considerable interest in the U.S. to develop new nuclear fuel with enhanced accident tolerance. Throughout the development of these new fuel concepts they will be extensively modeled using specialized computer codes and experimentally tested for a variety of different postulated accident scenarios. One accident scenario of interest, reactivity-initiated accident, is a nuclear reactor event involving a sudden increase in fission rate that causes a rapid increase in reactor power and temperature of the fuel which can lead to the failure of the fuel rods and a release of radioactive material. The focus of this work will be on the fuel performance modeling of reactivity-initiated accidents using the BISON code being developed at Idaho National Laboratory. The overall goal of this work is to provide the best possible modeling predictions for future experimental tests. Accurate predictive capability modeling using BISON is important for safe operation of these tests and provides a cheaper alternative to the expensive experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility
Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗