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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 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Advanced Test Reactor Neutron Dosimetry Report - Cycle 175C: Results of Neutron Dosimetry Measurements for the Advanced Test Reactor Cycle 175C

Results of neutron dosimetry measurements in ATR for cycle 175C are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2017 Status Report

This report contains a summary of irradiation testing of FCRD ATF-1 experiments performed at the INL in FY 2017. ATF-1 irradiation testing work performed in FY 2017 included fabrication of one ATF-1B drop in capsule ATF-1 series experiment and irradiation testing of ATF-1 capsules in the ATR. Additionally, five capsules met the sponsor-specified burnup targets, three of which were shipped from the ATR to the HFEF at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Common Cause Analysis Report for 2011 Fueled Vessel Drain, 2015 Defueled Vessel Drain, and 2015 EFIS Switch Misalignment

This causal analysis was initiated by ATR Management and committed to in INR 5238 for the August 24, 2015 primary coolant system (PCS) leak. This analysis sought common causal factors between the August 24, 2015 PCS leak, the July 23, 2015 emergency firewater injection system (EFIS) switch mispositioning, and the 2011 ATR fueled vessel level loss. This common cause analysis was conducted at the level of effort of an apparent cause.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report

The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach to create a temperature profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. This requires, necessarily, that the power generation profile inside of test fuel rodlets also be nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by both supporting analyses and by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The analyses and comparisons presented in this report show that ATR irradiations performed using cadmium shrouding are sufficiently prototypic that they can be used with confidence in the development and testing of fast reactor fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Advanced Test Reactor Neutron Dosimetry Report – 174 A & B

Results of neutron dosimetry measurements in ATR for cycle 174 A B are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Electrochemical Flow Reactor Design Allows Tunable Mass Transport Conditions for Operando Surface Enhanced Infrared Absorption Spectroscopy

Abstract In situ attenuated total reflection surface enhanced infrared absorption spectroscopy (ATR‐SEIRAS) is often used to investigate the near‐surface electrocatalytic reaction environment. However, there is a gap in directly correlating the near‐surface reaction environment with electrocatalytic reaction rates. To that end, we designed an electrochemical flow reactor for operando electrochemical ATR‐SEIRAS and demonstrate its capability with the CO 2 reduction reaction (CO 2 RR). Roughened gold catalyst thin films are prepared on ATR silicon crystals as a model system to probe local species under CO 2 RR conditions in 0.1 M KHCO 3 . We measured changes in the interfacial CO 2 concentration as a function of applied potential and electrolyte flow rate in operando , allowing us to correlate the changes in reaction rates with the observed CO 2 concentration. Including the choice of the catalyst and electrolyte, coupling hydrodynamic control with ATR‐SEIRAS in this platform enables investigations of how the local microenvironment affects the activity and selectivity of electrochemical reactions.

Avilés Acosta, Jaime E.↗
DOE OSTI2018

A Statistical Approach to Safety Margin Evaluation of Irradiation Experiments in the ATR

The current safety basis of the INL Advanced Test Reactor requires that irradiation experiments meet minimum safety margins to critical heat flux and flow instability during anticipated reactor fault conditions. Compliance to these requirements may be demonstrated by calculating safety margins using safety factors that account for uncertainties, or using a statistical approach to quantify the uncertainty in the safety margins. Safety margins obtained from the first approach are conservative since the individual events leading to the cumulative safety factor are assumed to occur simultaneously at their maximum uncertainty. The statistical approach is less conservative in the case where the uncertainties are not too large, and may reduce the conservatism resulting from an overabundance of safety factors. This report describes the procedure for safety margin evaluation using a statistical approach and identifies the important thermal-hydraulic variables to consider in an uncertainty analysis. In particular, safety margins to critical heat flux and flow instability are evaluated for three recent high power fueled experiments (MP-1, FSP-1 and AFIP-6) using RELAP5 coupled to RAVEN to perform statistical analysis and uncertainty quantification. The results demonstrate adequate safety margins to critical heat flux and flow instability during an anticipated loss of commercial power to the reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Nuclear Fuels & Materials Spotlight Volume 4

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Summary of LWRS Research in Addressing RPV Research Gaps in NRC EMDA Report

Reactor Pressure Vessels (RPVs) are critical components in nuclear reactors, housing the reactor core and coolant under extreme conditions of temperature, pressure, and radiation. These harsh environments contribute to the degradation of RPV materials over time, presenting challenges for extending reactor operations beyond their original design lifespans. The NRC Expanded Materials Degradation Assessment (EMDA) report volume 3 have been instrumental in guiding research to support extending the operational life of light water reactors (LWRs) up to 80 years. It provides a comprehensive framework to address technical challenges related to aging and degradation mechanisms in RPVs. The LWRS program has played a key role in advancing this research, supporting projects such as the UCSB ATR-2 Experiment, material testing from Zion and Palisades reactors, and the development of advanced mini-compact tension testing techniques. These efforts have been crucial in identifying and addressing gaps in our understanding of RPV aging, contributing to the successful subsequent license renewals of eight LWR units in the U.S. The EMDA report volume 3, built on the Phenomena Identification and Ranking Table (PIRT) analysis from earlier versions of the EPRI Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs), provides a detailed assessment of RPV degradation mechanisms. However, as EPRI has updated the MDM and IMT, it is important to revisit research priorities and methodologies to reflect these changes. The revised MDM and IMT may introduce new factors affecting long-term RPV performance and safety, highlighting the need for continued research and updated guidance to ensure the reliable and safe operation of reactors beyond 80 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Initial Characterization of ATR End Box Samples

Aluminum samples taken from discarded irradiated Advanced Test Reactor fuel element end boxes (upper and lower adapter ) were characterized to determine the chemical composition and hydration state of the surface oxides. The samples were analyzed using scanning electron microscope imaging and chemical identification, x-ray diffraction to evaluate crystalline morphology, and thermogravitational analysis to determine mass loss by dehydration. The samples were acquired from the discarded fuel element components stored in the ATR fuel handling canal and transferred to the Materials and Fuels Complex for analysis. The measurements indicate that the samples have limited oxide thickness, indicating that little corrosion has occurred beyond the initial passive layer that was intentionally formed prior to irradiation. The presence of minimal aluminum oxyhydroxide suggests that there will be minimal risk to store these fuels in air with limited risk for alteration of the base passive layer. These observations will be compared against samples that will be taken of the same type of fuel that has been in dry storage for more than 10 years.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

At-power subcritical multiplication in the advanced test reactor

The degree to which a given physical subdivision of a critical reactor acts as a neutron source for other lobes is not analytically characterized for general application. ATR and ATRC both operate at power, and individual power-producing lobes rely on each other as neutron sources in order to maintain constant power, which in general requires either exactly critical multiplication within a reactor or an external neutron source. Here, this work shows that lobe powers in ATR or ATRC can be related with subcritical multiplication theory, demonstrating that the lobes are not individually critical. This explanation is significant for ATR due to the desire to irradiate a large variety of experiments simultaneously, each having its impact on the core neutron population. This work also establishes the lack of relationship between experiment reactivity worth and cross-sections calculated from fundamental nuclear data. For any physical subdivision of any other critical reactor, it is likewise true that the subdivision undergoes only subcritical multiplication.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Experimental quantification of the background neutron flux in the Advanced Test Reactor fuel storage canal - 25562

The Advanced Test Reactor at the Idaho National Laboratory is one of only a few high-power, versatile, research reactors of its type in the world. The ATR fuel storage canal features a gamma irradiation facility that has various uses as well. It is composed of a simple dry tube that extends down into one of the fuel storage rack positions. This tube is surrounded by used fuel elements that can be selected from the inventory to produce the desired gamma radiation level and, to some extent, spectrum within the dry tube. The gamma field within the dry tube is accompanied by a small background neutron flux, produced primarily by photoneutron production in the small fraction of heavy water (D{sub 2}O) that is naturally present in light water. Neutron activation spectrometry was employed to quantify the background neutron flux in the dry tube in a configuration where the gamma exposure rate was approximately 5x10{sup 6} Roentgens per hour. The measurement technique used is very sensitive and provided self-consistent results in terms of raw activation data quality. The background neutron fluxes in the gamma facility were very low, but still detectable. The total neutron flux in the energy region above 0.414 eV and the thermal neutron flux (below 0.414 eV) were estimated from the activation measurements to be 7.5x10{sup 2} and 1.0x10{sup 2} n/cm{sup 2}-s, respectively. High uncertainties (factors of approximately 20-30% in the above-thermal range and 100% in the thermal range) apply to these results. These high uncertainties are primarily due to the limited number of activation interactions that could be used to detect such small fluxes and the resulting propagation of uncertainties associated with the various input parameters involved in the spectral unfolding process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Westinghouse Accident Tolerant Fuel Program

The PowerPoint presentation includes a timeline for the program implementation, the technical approach taken, a summary of testing results, a status of task 2, what has been done to share the results so far, and a summary of upcoming work.

ATR, HIFR, advanced test reactor, high flux isotop↗
DOE OSTI2022

NuScale Physics ECAR and Hardware Pictures

NuScale is requesting information on their Consolidated Innovative Nuclear Research (CINR) experiment. The CINR was awarded through the Nuclear Science User Facilities. Photos include the NuScale basket, insert, and instrumentation holder.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Improvements to In-Canal Assay of High Specific Activity 60Co at the ATR

Improvements to the quality of measurements in the assaying of High Specific Activity (HSA) cobalt 60 produced at the Advanced Test Reactor at the Idaho National Laboratory are presented. Such assays are conducted regularly between cycles at the ATR, and due to the radiological safety concerns, they are performed within the ATR canal. A standard ion chamber detector was placed in a collimated assay tube within the canal. Another tube, parallel to the detector, was used to position various cobalt 60 standards and targets. The collimated port between these tubes allows for the targets to be pulled across via a winch system. Previous measurements were performed at discrete 1.5-inch steps using a hand-cranked winch system. Charge was collected and integrated for 10 seconds at each discrete step during a scan. Two substantial improvements to the assay include an electronically controlled, motor-driven winch system and a software controlled continuous charge collection. Combined, these changes improved the quality of measurements as well as streamline the assay process. Further examination of these charge profiles indicated a reasonable amount of photon scattering, or shine, before a line-of-sight is established. This created inaccuracies in the reported activity of the cobalt targets. A series of MCNP-6.3 calculations were performed in an effort to quantify and correct for this shine effect. The results of the calculations and how they were employed to a create shine correction factor for the in-canal cobalt assay at the ATR are presented. The overall bias showed an overestimation of the HSA cobalt 60 activity by around 7-8%.

07 - ISOTOPES AND RADIATION SOURCES↗