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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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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Analysis of inconel 600 oxidized under loss-of-coolant accident conditions: A multi-modal approach

Understanding the performance of cladding materials during a loss-of-coolant accident scenario is crucial for developing next-generation accident tolerant nuclear fuel rod claddings. We describe a multi-modal approach combining synchrotron-based diffraction, and Raman spectroscopy with nano-scale electron microscopy techniques for investigating oxidation of Inconel 600 (A600) in steam and air environments at 1200 °C for 2 h. We report that A600 exposed to steam develops a Cr 2 O 3 -enriched surface layer while a mixed phase oxide layer containing NiFe 2 O 4 , Cr 2 O 3 , Fe 3 O 4 and Fe (3–x) CrxO 4 is formed in air. Mechanism of oxidation for A600 in air and steam environments is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Public Health Response and Medical Management of Internal Contamination in Past Radiological or Nuclear Incidents: A Review

Following a radiological or nuclear emergency, workers, responders and the public may be internally contaminated with radionuclides. Screening, monitoring and assessing any internal contamination and providing necessary medical treatment, especially when a large number of individuals are involved, is challenging. Experience gained and lessons learned from the management of previous incidents would help to identify gaps in knowledge and capabilities on preparedness for and response to radiation emergencies. In this paper, eight largescale and five workplace radiological and nuclear incidents are reviewed cross 14 technical areas, under the broader topics of emergency preparedness, emergency response and recovery processes. The review findings suggest that 1) new strategies, algorithms and technologies are explored for rapid screening of large populations; 2) exposure assessment and dose estimation in emergency response and dose reconstruction in recovery process are supported by complementary sources of information, including ‘citizen science’; 3) surge capacity for monitoring and dose assessment is coordinated through national and international laboratory networks; 4) evidence-based guidelines for medical management and follow-up of internal contamination are urgently needed; 5) mechanisms for international and regional access to medical countermeasures are investigated and implemented; 6) long-term health and medical follow up programs are designed and justified; and 7) capabilities and capacity developed for emergency response are sustained through adequate resource allocation, routine nonemergency use of technical skills in regular exercises, training, and continuous improvement.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Heat capacity, entropy, formation energy and spin-fluctuation behavior of U 3 Si 5 from 2.4 to 397.4 K

U-Si intermetallic compounds are of considerable interest for their applications as accident-tolerant nuclear fuels. Here we present low-temperature heat capacity (LTHC) measurements of one of the U-Si phases, U 3 Si 5 , using a Quantum Design Physical Properties Measurement System (PPMS) from 2.4 to 397.4 K. We observed an upturn in C p /T (T) below 10 K and have attributed this behavior to potential spin-fluctuations (SF) with an SF temperature (T sf ) of 27 K. An enhancement of LTHC was also observed, as manifested by a large electronic heat capacity coefficient (γ el ) of 342.9 mJ/mol•K 2 . From the heat capacity data, the following thermodynamic parameters were determined: the characteristic Debye temperature (θ D ) over the temperature range 30 – 397 K is 177 ± 2 K, and the standard entropy ($Δ^{298.15}_0$$S^o$) is 283.3 ± 5.7 J•mol -1 •K -1 (equivalent to 35.4 ± 0.7 J•mol -1 •atom -1 •K -1 ). Combined with our previously measured formation enthalpy ($Δ_fH^°_{el}$) of U 3 Si 5 , the Gibbs free energy of formation of U 3 Si 5 from the elements ($Δ_fG^°_{el}$) was determined to be –45.2 ± 9.0 kJ•mol -1 •atom -1 .

36 MATERIALS SCIENCE↗
DOE OSTI2019

High temperature neutron diffraction investigation of U 3 Si 2

The U-Si system is actively undergoing studies due to its promise as a component of an accident tolerant nuclear fuel. In this work, the crystal structure of the U 3 Si 2 compound in the U-Si system was investigated as a function of temperature from room temperature to 1298 K for a sample of approximately 42.02 at.% Si composition and 1373 K for a sample with approximately 42.10 at.% Si using high temperature time-of-flight neutron diffraction on the High-Pressure Preferred Orientation (HIPPO) diffractometer at Los Alamos Neutron Science Center (LANSCE). The simultaneous Rietveld refinement of five histograms from the five HIPPO detector backs (40°, 60°, 90°, 120°, and 145°) provided datasets for the lattice parameters, anisotropic atomic displacement parameters, and atomic positions as a function of temperature. To explore the possibility of a homogeneity range, two sample compositions were analyzed, a 40.02 at.% Si (near stoichiometric) and a potentially hyperstoichiometric (40.10 at.% Si) sample. While minor differences in the anisotropic atomic displacement parameters between the two samples were observed, over the entire investigated temperature range no additional phases were found indicative of a single phase sample and the ability of the U 3 Si 2 compound to accommodate nonstoichiometry, suggesting that the U 3 Si 2 compound is not a line compound confirming previous determinations. However, the differences observed in the average coefficient of linear thermal expansion identified between the two compositions warrant future investigation.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Chromium-doped uranium dioxide fuels: A review

UO 2 doped with parts per million CR 2 O 3 powder is considered a potential near term accident tolerant fuel candidate. Here, the results of decades of industry and academic research into Cr-doped UO 2 are analyzed and their shortcomings are critiqued. Focusing on the incorporation mechanisms of Cr into the fuel matrix, we explore a mechanistic understanding of the characteristic properties of Cr-doped UO 2 , notably, enhanced fission gas retention attributed to enlarged grain sizes following sintering, along with marginal improvements in the thermophysical properties. The findings of recent X-ray Adsorption Near Edge Spectroscopy studies were compared and put into conversation with historic data regarding the incorporation of Cr in UO 2 . On the basis of defect mechanisms, the case is made for the substitutional incorporation of Cr governing the lattice solubility but not the enhanced U diffusivity. Instead, Cr/CR 2 O 3 redox chemistry in a well-defined oxygen potential explains the differences in the U diffusivity and O/M ratio. The primary mechanism of doping enhanced grain growth is found to be liquid assisted sintering due to a CRO (1) eutectic phase at the grain boundaries. The role of inhomogeneities in Cr concentration in UO 2 at various length scales across the materials microstructure is highlighted and connected to promising experimental and modeling work to fill in the gaps in the current understanding of Cr-doped UO 2 . The review considers both the open scientific questions and engineering applications to illustrate the deep connections between the practice and theory in the design of accident tolerant nuclear fuels. In conclusion, the review ends with an outline of future works that combine meticulous irradiation studies and high resolution experiments with next generation modeling and simulations techniques empowered by machine learning advances to accelerate the fabrication and adoption of Cr-doped UO 2 light water reactors.

Cleveland, Mack Wesley [Massachusetts Inst. of Tec↗
DOE OSTI2020

Mutational signatures in tumours induced by high and low energy radiation in Trp53 deficient mice

Lonising radiation (IR) is a recognised carcinogen responsible for cancer development in patients previously treated using radiotherapy, and in individuals exposed as a result of accidents at nuclear energy plants. However, the mutational signatures induced by distinct types and doses of radiation are unknown. Here, we analyse the genetic architecture of mammary tumours, lymphomas and sarcomas induced by high ( 56 Fe-ions) or low (gamma) energy radiation in mice carrying Trp53 loss of function alleles. In mammary tumours, high-energy radiation is associated with induction of focal structural variants, leading to genomic instability and Met amplification. Gamma-radiation is linked to large-scale structural variants and a point mutation signature associated with oxidative stress. The genomic architecture of carcinomas, sarcomas and lymphomas arising in the same animals are significantly different. Our study illustrates the complex interactions between radiation quality, germline Trp53 deficiency and tissue/cell of origin in shaping the genomic landscape of IR-induced tumours.

Science & Technology - Other Topics↗
DOE OSTI2025

Irradiation Vehicles for Evaluating SiC/SiC Cladding Bowing Under Neutron Flux Gradients

Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are among the most promising candidates for long term accident-tolerant nuclear fuel cladding. A key challenge related to their deployment is lateral bowing caused by differential radiation-induced swelling under dose or temperature gradients, which could obstruct coolant flow or interfere with control rod/blade movements. Although previous modeling efforts have predicted bowing behavior in light-water reactor (LWR) environments, experimental validation remains limited, especially at prototypic LWR temperatures. This study addresses that gap by irradiating six reduced-length SiC/SiC cladding tubes (~600 mm) in the High Flux Isotope Reactor (HFIR) at ~300°C, which is representative of LWR conditions. The tubes will be housed in a sealed vessel with an inert gas gap to maintain target temperatures and prevent direct coolant contact. Arranged in three pairs, each set will receive a different radiation dose (low, medium, high), with the central pair receiving ~0.1 displacements per atom (dpa)—the expected dose for peak bowing. The experiment will determine the dose-dependent bowing behavior and validate predictive finite element models. In this work, the tubes are freely suspended from pins to allow for unconstrained bowing; however, we present a concept for introducing localized constraints to represent grid spacer effects. Post-irradiation examination will include swelling measurements and profilometry to assess bowing and compare the results with model predictions. This work aims to confirm the conditions under which maximum bowing occurs so as to improve the reliability of SiC/SiC performance models in nuclear applications.

Russell, Nick [ORNL] (ORCID:0000000296099820)↗
DOE OSTI2023

Optimizing chemistry for designing oxidation resistant FeCrAl alloys

Abstract Traditionally, FeCrAl alloys played an important role in high-temperature applications due to their ability to form a passive Al oxide film at temperatures above ~ 800 °C. Recently, FeCrAl alloys became of interest for the application of accident tolerant nuclear fuel cladding. This study covers work done at GE Research for better understanding the role of Al, Cr, and Mo in oxidation kinetics and thermodynamics. Several models and commercial prototype alloys have been tested in hydrothermal corrosion autoclave loops, at low temperature steam exposure (~ 400 °C), high temperature steam exposure (~ 1000 °C or higher), and high temperature air exposures. The results provide insights on how chromium and aluminum play a significant role in both high temperature and low temperature oxidation of FeCrAl. Additionally, machine learning tools are used to gain further insights on both predicting future optimized chemistries for balancing the properties of hydrothermal corrosion, low and high temperature steam oxidation, and thermal aging (which is exacerbated due to radiation in a nuclear reactor environment). GE plans to use this framework to further optimize the FeCrAl alloy system for use in nuclear reactor environments. Graphical abstract

Roy, Indranil (ORCID:0000000336124323)↗
DOE OSTI2026

MACCS User Guide - Version 5.2

MACCS is used by the Nuclear Regulatory Commission (NRC) and various national and international organizations for probabilistic consequence analysis of nuclear power accidents. This user guide is intended to assist analysts in understanding the MACCS/MACCS-UI User Interface (UI) model and to provide information regarding the code. This user guide version describes MACCS Version 5.2, model history, explains how to set up and execute a problem, and informs the user of the definition of various input parameters and any constraints placed on those parameters. This report is part of a series of reports documenting MACCS. Other reports include the MACCS Theory Manual, MACCS Verification Report, Technical Bases for Consequence Analyses Using MACCS, as well as documentation for preprocessor codes including SecPop, MelMACCS, and COMIDA2.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI1994

Nuclear Safety [Vol. 35, No. 2, July-December 1994]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 179 Consideration of Postaccident Consequences in the Determination of Safety Objectives for Future Nuclear Power Plants in France, D. Queniart, A. Sugier, and J. Lochard; ACCIDENT ANALYSIS: 187 Nuclear Safety Research: The Phebus FP Severe Accident Experimental Program, P. von der Hardt, A. V. Jones, C. Lecomte, and A. Tattegrain; 205 Containment Performance Analysis of the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory, S. H. Kim, R. P. Taleyarkhan, and V. Georgevich; 213 Assessment of Fission Product Deposits in the Reactor Coolant System: The DEVAP Program, G. Le Marois and M. Megnin; 222 Erratum to “A Review of the Available Information on the Triggering Stage of a Steam Explosion," Vol. 35, No. 1, CONTROL AND INSTRUMENTATION: 223 Effects of Normal Aging on Calibration and Response Time of Nuclear Plant Resistance Temperature Detectors and Pressure Sensors, H. M. Hashemian; DESIGN FEATURES: 235 Defense in Depth Against the Hydrogen Risk—A European Research Program, F. Fineschi; ENVIRONMENTAL EFFECTS: 246 Technical Note: A Preliminary Analysis of the Risks to Hong Kong Resulting from Potential Accidents of Daya Bay Nuclear Power Plant, Z. Shi and X. Wei; OPERATING EXPERIENCES: 253 Reactor Shutdown Experience, Compiled by J. W. Cletcher; SPECIAL SECTION ON TMI-2 VESSEL INVESTIGATION PROJECT: 256 Three Mile Island—New Findings 15 Years After the Accident A. M. Rubin and E. Beckjord; 269 Relocation of Molten Material to the TMI-2 Lower Head, J. R. Wolf, D. W. Akers, and L. A. Neimark; 280 Insight Into the TMI-2 Core Material Relocation Through Examination of Instrument Tube Nozzles, L. A. Neimark; 288 Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head, D. W. Akers and B. K. Schuetz; 301 Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head, D. R. Diercks and G. E. Korth; 313 Margin-to-Failure Calculations for the TMI-2 Vessel, J. Rempe, L. Stickler, S. Chavez, G. Thinnes, R. Witt, and M. Corradini; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 328 1993 Accident Sequence Precursor (ASP) Program Results, L. N. Vanden Heuvel, J. W. Cletcher, D. A. Copinger, J. W. Minarick, B. W. Dolan, and P. D. O’Reilley; RECENT DEVELOPMENTS: 339 Reports, Standards, and Safety Guides, D. S. Queener; 345 Proposed Rule Changes as of June 30, 1994; ANNOUNCEMENTS: 234 Thirty-First Annual Meeting of the National Council on Radiation Protection and Measurements; 356 1995 International Incineration Conference; 358 Fifth International Controls and Instrumentation Conference; 359 ANS International Topical Meeting on Safety of Operating Reactors; 359 Fifth International Conference on Nuclear Criticality Safety; 360 International Conference on Probabilistic Safety Assessment Methodology and Applications; 351 The Authors; 357 Reviewers of Nuclear Safety, Vol. 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Effects of transient fission gas release on rod balloon burst behavior during a loss-of-coolant accident

The US nuclear energy industry is investigating strategies to increase the reactor operating cycle to 24 months, which would result in rod average burnups exceeding the current limit of 62 GWd/tU. To support this goal, multiphysics simulations tools and methodologies are being developed to predict the effects of this change during transient events such as loss-of-coolant accidents (LOCAs). In this work, two cladding burst correlations were used to predict cladding failure. Here, these were coupled with three transient fission gas release (tFGR) models and were implemented in the BISON fuel performance code to quantify any changes in cladding burst behavior during a large-break LOCA. First, a simple linear model was used to perform a sensitivity analysis of a single fuel rod. Second, a tFGR model available in BISON was applied to 281 fuel rods throughout the core. The third model examined was an empirical correlation based on data from the literature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

State-of-the-Art Reactor Consequence Analysis Project: Sensitivity Analysis Using Dakota

The State-of-the-Art Reactor Consequence Analyses (SOARCA) project has focused on best estimate analyses and uncertainty analysis for postulated accidents at specific nuclear power plants. The consequences of these accidents are estimated using the simulation tools MELCOR and MACCS. To understand which uncertain input variables are important to determining these consequences, analysts have performed sensitivity analyses. The tool used to perform these sensitivity analyses in previous SOARCA work, CompModSA, is no longer supported. Therefore, the current work focuses on migrating these analyses to another tool and evaluating its performance. Dakota, which is a tool developed at Sandia National Laboratories, is used in this work. Sensitivity results are created for three analyses from the SOARCA Surry UA. Though CompModSA and Dakota vary slightly in their algorithms and implementation, their sensitivity results generally agree, which gives confidence in the Dakota approach and increases confidence in the original analyses. It is likely that this methodology is extendable to the rest of SOARCA analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Risk-Informed Initiating Events and Accident Response

Emerging nuclear technologies and advanced reactor developments have brought proposed changes to the regulatory framework which can be leveraged to improve safety implementation and regulatory oversight processes at existing operating nuclear power plants. Specifically, opportunities exist to make plants’ regulatory compliance processes more efficient. This can be done by decreasing reliance on purely deterministic and prescriptive approaches and expansion of the use of risk informed and performance-based approaches to demonstrate reactor safety while achieving economic gains and efficiencies. In this report, we researched possibilities of economic benefits potentially available from the application of concepts associated with a modernized regulatory framework developed for advanced reactors to the existing light water reactors. We used timing factors to demonstrate the complexity of the existing risk assessments and described how these complexities affect regulatory compliance activities at the plants. We also investigated regulatory framework for both existing and new reactors and provided an overview of potential improvements. Lastly, we evaluated various areas of existing plant operations where modernization of compliance activities can offer substantial benefits.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Scoping Analysis of MACCS Modeling Improvements for the Study of Protective Action Recommendations

In late 2004, the U.S. Nuclear Regulatory Commission (NRC) initiated a project to analyze the relative efficacy of alternative protective action strategies in reducing consequences to the public from a spectrum of nuclear power plant core melt accidents. The study is documented in NUREG/CR-6953, “Review of NUREG-0654, Supplement 3, ‘Criteria for Protective Action Recommendations for Severe Accidents,’” Volumes 1, 2, and 3. The Protective Action Recommendations (PAR) study provided a technical basis for enhancing the protective action guidance contained in Supplement 3, “Guidance for Protective Action Strategies,” to NUREG-0654/FEMA-REP-1, Rev. 1, “Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, ” dated November 2011. In the time since, a number of important changes and additions have been made to the MACCS code suite, the nuclear accident consequence analysis code used to perform the study. The purpose of this analysis is to determine whether the MACCS results used in the PAR study would be different given recent changes to the MACCS code suite and input parameter guidance. Updated parameters that were analyzed include cohorts, keyhole evacuation, shielding and exposure parameters, compass sector resolution, and a range of source terms from rapidly progressing accidents. Results indicate that using updated modeling assumptions and capabilities may lead to a decrease in predicted health consequences for those within the emergency planning zone compared to the original PAR study.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

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↗