Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Nuclear Test”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

The international research reactor conversion effort and its contribution to the validation of reactor physics and thermal-hydraulics codes

Full text of publication follows. With about 200 reactors in operation, civilian Research and Test Reactors (RTRs) represent approximately a third of the global nuclear fleet. RTRs make use of core designs that are drastically different from commercial power plants to perform a wide variety of non-power applications that greatly benefit society. Because they often rely on high neutron fluxes, RTRs are designed with relatively compact cores and as a result, prior to the 1980's, were often deployed using Highly Enriched Uranium fuel (HEU, {sup 235}U/U = 20 wt. %). Due to proliferation concerns, the international community aims at eliminating the use of HEU in civilian facilities and favor instead the use of Low Enriched Uranium fuel (LEU, {sup 235}U/U < 20 wt. %). A program to support conversion of the world's RTRs to LEU fuel has been initiated in 1978 by the U.S. Department of Energy (DOE). This program is still alive today and has achieved more than 103 conversion metrics. Today, the program focuses heavily on the conversion of so-called high-performance RTRs, which are far more challenging than previous conversions as they require new fuel element designs and the use of new, higher density LEU fuel forms. Development and qualification of new LEU fuel elements is ongoing and requires extensive engineering analysis and testing. Both activities require the development and validation of codes and methods for reactor physics, thermal-hydraulics, and multi-physics, which in turn rely on experiments performed in RTRs or other experimental facilities. This talk will present the collection of RTR experimental data and benchmark analyses from the international conversion program that contribute to the validation of computer codes and methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Radiation Testing for High-Resolution Radiation-Hardened Camera System (Final CRADA Report)

The nuclear energy industry needs higher-image quality and higher rad-hard vision systems for refueling and inspection operations that are required every 12-18 months for all operating commercial reactors. During refueling operations, the serial numbers of fuel assemblies need to be visually verified in a challenging high radiation environment with dose rates on the order of 1kGy/hr. at the top of the core (approximately 10 days after shutdown). Vega Wave Systems has developed and built a radiation hard vision system that: 1. has high radiation hardness, 2. demonstrated best in the industry image quality, 3. has small size and weight, and 4. does not suffer from radiation-induced noise. ANL-1060 (02/13/2023) The proprietary Enduray vision system from Vega Wave Systems, Inc. is designed to withstand more than 400x the radiation level of the Vidicon-based systems, currently the highest radiation tolerant system in use today. The Van De Graaff (VDG) electron accelerator at Argonne National Laboratory is ideally suited for testing of Enduray vision system. VDG provides a broad band (up to 3 MeV) x-ray source that reasonably approximates the energy and spectrum of the gamma radiation expected at the top of a nuclear core 10 days after shutdown (~1kGy/hr.), is easily controlled for variable tests, and will not activate the parts under radiation for ease of timely and close inspection and characterization of the system in case of operational issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Physics with CEBAF at 12 GeV and future opportunities

Here, we summarize the ongoing scientific program of the 12 GeV Continuous Electron Beam Accelerator Facility (CEBAF) and give an outlook into future opportunities. The program addresses important topics in nuclear, hadronic, and electroweak physics, including nuclear femtography, meson and baryon spectroscopy, quarks and gluons in nuclei, precision tests of the standard model and dark sector searches. Potential upgrades of CEBAF and their impact on scientific reach are discussed, such as higher luminosity, the addition of polarized and unpolarized positron beams, and doubling the beam energy.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Contribution of nuclear excitation electromagnetic form factors in C 12 and O 16 to the Coulomb sum rule

We report on empirical parametrizations of longitudinal and transverse nuclear excitation electromagnetic form factors in 12 C and 16 O. We extract the contribution of nuclear excitations to the normalized inelastic Coulomb sum rule [S L (q)] as a function of momentum transfer q and find that it is significant (0.29 ± 0.030 at q = 0.22 GeV). The total contributions of nuclear excitations to S L (q) in 12 C and 16 O are found to be equal within uncertainties. Since the cross sections for nuclear excitations are significant, the radiative tails from nuclear excitations should be included in precise calculations of radiative corrections to quasielastic electron scattering at low q and deep-inelastic electron scattering at large energy transfers ν. The parametrizations also serve as a benchmark in testing theoretical modeling of cross sections for excitation of nuclear states in electron and neutrino interactions on nuclear targets at low energies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Accelerating Nuclear Fuels and Materials Qualification-Learning from MeV Summer School 2022

This presentation focused on the general overview of the lecture content that related to nuclear fuels and material qualification. The content of this presentation is taken from the lecture notes of MeV Summer School 2022. The sources are cited as per lecture notes. The focus of this presentation is to review the learning from the Modeling, Experiment, and Validation (MeV) Summer School (July 18-29, 2022) for early career researchers and scientists. This year's school was hosted by Oak Ridge National Laboratory (ORNL) and was focused on Accelerating nuclear fuels, and materials qualification by combining high through-put materials irradiation and testing, advanced PIE, and Multiphysics modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

ANS Winter 2024 Summary: Optimizing the ATF-2Ramp Power Profile

When the Halden Boiling Water Reactor closed down in 2018, a need to restore the capability for in-reactor power ramp testing arose. Such testing is valuable for studying pellet-clad interaction phenomena in nuclear fuels. The data from these studies is of great interest to a number of research programs, including the accident-tolerant fuel (ATF) program at Idaho National Laboratory (INL). In 2022, Woolstenhulme et al. proposed several power ramp testing ideas using facilities at INL, including irradiation in the Transient Reactor Test Facility (better known as TREAT) and the Advanced Test Reactor (ATR) [1]. Worrall et al. [2] and Labossiere-Hickman et al. [3] subsequently performed feasibility studies for the ATR testing options in 2023. This summary further investigates the three-pin trefoil design (Fig. 1) for the proposed ATF-2Ramp Experiment discussed in Labossiere-Hickman et al. [3]. ATF-2Ramp is designed to operate in the center flux trap (CFT) of the ATR during a powered axial locator mechanism (PALM) cycle: a short, variable-powered cycle with an asymmetric power distribution. Previously, it was shown that tailoring the thickness of the hafnium (Hf) neutron shields (“mini-shrouds”) surrounding each pin offered a degree of control sufficient to achieve the programmatic linear heat generation rate (LHGR) targets for ATF-2Ramp during the high-power period of a PALM cycle. New work involves shortening the experiment test train for consistency with the fuel pins in ATF-2D [4] and then shaping the axial power profile of the three test pins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The Virtual Test Bed (VTB) Repository: A Library of Multiphysics Reference Reactor Models using NEAMS Tools

With the next generation of nuclear reactors under development, modeling and simulation (M&S) tools are being developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in order to support their design, licensing, and future operation. Mirroring the physical test beds currently under construction (i.e., EBR-II and ZPPR), the Virtual Test Bed (VTB) was launched by the National Reactor Innovation Center (NRIC) in collaboration with NEAMS to support the advanced reactor community. This collaborative effort, which involves multiple teams at both Idaho National Laboratory and Argonne National Laboratory aims to use NEAMS tools to model a wide range of reactor designs. Those models are automatically tested to ensure their continued functionality as the tools are further developed. Examples are extensively documented, each acting as a tutorial for applying the relevant NEAMS tools to that reactor design. Currently, five advanced reactor types (with a total of eight specific design variants) are simulated by a variety of different models. These models range from steady-state, core multiphysics simulations to integrated plant analysis during loss-of-flow transients.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

High Explosive Testing Capabilities at Nevada National Security Site (NNSS) [Slides]

LANL is a non-profit Federally Funded Research & Development Center (FFRDC). LANL’s Integrated Weapons Experiments (J) Division addresses national security challenges by executing largescale, integrated, focused experiments, and tests of engineered devices and systems. J-NV plans, fields, and executes dynamic high explosive experiments at NNSS. This includes sub-critical experiments involving special nuclear material, as well as small- and largescale non-nuclear experiments. J-NV generates, analyzes, and disseminates unique, high-quality data.

42 ENGINEERING↗
DOE OSTI2024

Development of Accelerated High Temperature Mechanical Testing Techniques

The Advanced Materials and Manufacturing Technologies (AMMT) Program focuses on advancing materials and manufacturing techniques for nuclear energy applications, particularly in the qualification of materials for high-temperature structural use. This report presents work on refining the creep testing of small specimen geometries. Efforts include the development of a new specimen geometry for sub-sized specimens, which were subjected to uniaxial creep tests. The results contribute to the understanding of material behavior under stress at elevated temperatures and offer potential improvements in creep data collection methods. These findings support ongoing advancements in material qualification processes essential for nuclear reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

ALLOY SELECTION AND C-276 CODE DESIGN VALUE EXTENSION FOR ADVANCED MOLTEN SALT TECHNOLOGY TEST FACILITIES EXPERIMENTATION

Molten halide salts are being considered as working fluids for nuclear and concentrated solar power applications. High temperature molten fluoride and chloride salts are known to preferentially attack and deplete Cr in alloys, which leads to the use of high-Ni lower-Cr alloys in test facilities for advanced Molten Salt Technology. Alloys 600 and C-276 are two commercially available Ni alloys that have moderate Cr contents and are qualified to the maximum temperature of 649°C and 677°C (1,200°F and 1,251°F), respectively, in the Boiler and Pressure Vessel Code. Both alloys have good corrosion resistance to acids, are resistant to stress-corrosion cracking, and have long track records of use in the chemical industry. Comparatively, Alloy C-276 has a similar Cr content but much higher allowable stresses for temperatures greater than 600°C (1,112°F). Therefore, it has been considered as a structural material for test facilities that require operations at 700°C (1,292°F) or greater to develop high-temperature Molten Salt Technology. To meet the requirements, the current code design values were extended based on some experimental data. Analysis showed that above current Codified maximum temperature, strength of the alloy is mainly controlled by creep rupture life under the average stress, although the Sc creep rate criterion is close to the Favg.Savg rupture criterion. This paper presents the intended test facilities and the design requirements, alloy selection considerations, literature review, data analysis, and proposed allowable stress extension based on some creep test data for C-276 at temperatures greater than 677°C (1,251°F). Further research activities are also briefly mentioned.

Ren, Weiju↗
DOE OSTI2020

Simulations of the High Temperature Test Facility using SAM

Under the support from the U.S. Department of Energy Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an effort is being pursued to support the modeling and simulation needs of high-temperature gas-cooled reactor (HTGR) technology development. There is a particular need for advanced modeling and simulation methods and tools to predict thermal-fluid behavior in the nuclear reactor primary system during safety-related transients. This report focuses on one such activity related to HTGR: developing a model of the High Temperature Test Facility (HTTF) at Oregon State University using the system-level code SAM, and using the model to understand thermal response behavior in the facility. Note this activity is coordinated with the DOE-NE’s Advanced Reactor Technology Gas-Cooled Reactor Program

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Chemical durability assessment of enhanced low-activity waste glasses through EPA method 1313

In this work, we report the progress of the Glass Leaching Assessment for Durability (GLAD) program on the implementation of the United States Environmental Protection Agency (EPA) Leaching Environmental Assessment Framework pH-dependent leach test (EPA Method 1313) to low-activity nuclear waste (LAW) glasses. The GLAD program seeks to develop new strategies to understand the chemical durability of nuclear waste glasses for the disposal in near-surface conditions. A series of 16 high-waste loading LAW glasses, currently under development, were selected using machine learning methods to study the corrosion behavior using EPA Method 1313. Reacted glass powders were examined using scanning electron microscopy and the eluate compositions were examined using inductively coupled plasma-optical emission spectroscopy. Compositional modeling was used to fit the measured elemental releases from EPA Method 1313. The compositional models demonstrated that elements such as Si reduce elemental release while B can increase elemental release (consistent with elemental modeling of the Product Consistency Test and Vapor Hydration Test) while other elements, such as Fe, exhibit pH-dependent behavior. The amount of acid added during the EPA testing was found to significantly impact the observed result, which was only apparent after preforming the present matrix study. The overall titration curves were able to be compositionally modeled for future process optimization.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Improving robustness for model discerning synthesis process of uranium oxide with unsupervised domain adaptation

The quantitative characterization of surface structures captured in scanning electron microscopy (SEM) images has proven to be effective for discerning provenance of an unknown nuclear material. Recently, many works have taken advantage of the powerful performance of convolutional neural networks (CNNs) to provide faster and more consistent characterization of surface structures. However, one inherent limitation of CNNs is their degradation in performance when encountering discrepancy between training and test datasets, which limits their use widely. The common discrepancy in an SEM image dataset occurs at low-level image information due to user-bias in selecting acquisition parameters and microscopes from different manufacturers. Therefore, in this study, we present a domain adaptation framework to improve robustness of CNNs against the discrepancy in low-level image information. Furthermore, our proposed approach makes use of only unlabeled test samples to adapt a pretrained model, which is more suitable for nuclear forensics application for which obtaining both training and test datasets simultaneously is a challenge due to data sensitivity. Through extensive experiments, we demonstrate that our proposed approach effectively improves the performance of a model by at least 18% when encountering domain discrepancy, and can be deployed in many CNN architectures.

scanning electron microscopy↗
DOE OSTI2022

Development and Test Results of Thermocouples Used in the TRISO-Fuel Irradiation Experiment AGR-5/6/7

High-temperature gas reactor irradiation experiments create unique challenges for thermocouple-based temperature measurements. High-temperature industrial thermocouples suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. For lower-temperature applications, Type K and Type N thermocouples are affected by neutron irradiation only to a limited extent. But until recently, the use of these nickel-based thermocouples was limited when the temperature exceeded 1050°C due to drift related to phenomena other than nuclear irradiation. Certain portions of the AGR-5/6/7 experiment experienced temperatures higher than any of the previous AGR tests, up to 1500°C. Recognizing the limitations of existing thermometry to measure such high temperatures, the sponsor of the AGR-5/6/7 test supported a development and testing program for thermocouples capable of low-drift operation at temperatures above 1100°C. This program included additional development of high-temperature irradiation-resistant thermocouples based on molybdenum/niobium thermoelements, which have been studied at Idaho National Laboratory since circa 2004. A step change in accuracy and long-term stability of this thermocouple type was achieved as part of the AGR-5/6/7 thermometry development program. Additionally, long-term testing (7000+ h) at 1250°C of Type N thermocouples utilizing a customized sheath developed at the University of Cambridge has been completed with excellent low-drift results. The results of this testing as well as testing of the improved high-temperature irradiation-resistant design are reported herein. Both the improved high-temperature irradiation-resistant and the Cambridge Type N thermocouple types were incorporated into the AGR-5/6/7 test, which began irradiation in February 2018 and was completed in July 2020. Finally, a summary of the performance of the thermocouples incorporated into the AGR-5/6/7 test is included herein.

In-pile instrumentation↗
DOE OSTI2020

Oak Ridge Response to Versatile Test Reactor Environmental Impact Statement Data Request

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor being developed in the United States under the direction of the US Department of Energy Office of Nuclear Energy (DOE-NE). The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by the US National Laboratories in collaboration with General Electric–Hitachi and Bechtel National, Inc. In support of the VTR project, DOE issued a Notice of Intent (NOI) in the Federal Register on August 5, 2019, announcing the intent to prepare an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act (NEPA) and its implementing regulations. The EIS will evaluate alternatives for a versatile reactor–based fast-neutron source facility and associated facilities for the preparation, irradiation, and post-irradiation examination (PIE) of test/experimental fuels and materials. Specifically, the NOI identified two siting alternatives for the VTR reactor facility: Idaho National Laboratory (INL) or Oak Ridge National Laboratory (ORNL). In addition, the NOI also specified two siting alternatives for VTR fuel fabrication: INL and the Savannah River Site (SRS). This report provides information in response to data requests made to ORNL to fill in site-specific knowledge gaps to develop a high-quality EIS. The responses provided are not required to provide full details in every aspect; instead, they adequately bound possible environmental impacts or provide sufficient information to adequately assess likely environmental impacts. This work is being performed under a subcontract from INL to ORNL using DOE-NE funds and is directed by DOE-NE and DOE-ID. Leidos has been contracted by DOE-NE to write the VTR EIS, so most data requests have come from Leidos but were often routed through INL or DOE-ID. DOE-ID is overseeing the NEPA and EIS processes for the VTR project. Leidos will use the information provided in this report to inform the VTR EIS and will also cite this document to establish a clear, publicly available source of the information. Section 2 of this report briefly describes the proposed ORNL VTR Alternative and illustrates the location of the proposed site for the ORNL VTR Alternative. Sections 3 through 7 provide direct responses to data requests received by ORNL. These sections use a tabular format in which data requests are divided into separate items to be addressed; the items are numbered, the data requests are restated with more topical information included, and then the responses are provided. Initial data requests and follow-on requests for additional information (RAIs) are combined under the original data request fields. Finally, Section 8 presents summarized conclusions and describes future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Closure Report for Corrective Action Unit 116: Area 25 Test Cell C Facility, Nevada National Security Site, Nevada with ROTC-1

CR loaded to this OSTI record. Just adding new file which includes CR plus new ROTC 1 and update the metadata to the following: This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 116, Area 25 Test Cell C Facility. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 [as amended March 2010]). CAU 116 consists of the following two Corrective Action Sites (CASs), located in Area 25 of the Nevada National Security Site: (1) CAS 25-23-20, Nuclear Furnace Piping and (2) CAS 25-41-05, Test Cell C Facility. CAS 25-41-05 consisted of Building 3210 and the attached concrete shield wall. CAS 25-23-20 consisted of the nuclear furnace piping and tanks. Closure activities began in January 2007 and were completed in August 2011. Activities were conducted according to Revision 1 of the Streamlined Approach for Environmental Restoration Plan for CAU 116 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2008). This CR provides documentation supporting the completed corrective actions and provides data confirming that closure objectives for CAU 116 were met. Site characterization data and process knowledge indicated that surface areas were radiologically contaminated above release limits and that regulated and/or hazardous wastes were present in the facility. The Record of Technical Change 1 updated the use restriction information.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Contributions for ENDF/B-VIII.1 Paper

It was discussed at WANDA2020 that some issues were found by the community and users in ENDF/B-VIII.0 covariances after their release. It was mentioned that one issue in testing ENDF/B-VIII.0 covariances was that they were provided by evaluators too late in the β-release phases that insufficient time for thorough testing was available. Therefore, the executive committee requested that first nuclear data covariances should already be delivered for ENDF/B-VIII.1β 1 in February 2023—about half a year earlier than was the case for ENDF/B-VIII.0. This additional time allowed for more stringent testing with LANL’s ENDFtk and CovVal codes, and ORNL’s AMPX code during the release phases. The LLNL covariance testing employed a combination of NJOY, empy tools from EMPIRE, and x4i.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Mounted Accelerometer Frequency Response of Adhesive Products and Aluminum Frame Quick Mounts

An accelerometer mounting technique has large implications on the frequency range and accuracy of the measurement, with stiffness and the mass relative to the monitored structure as the primary concerns. The International Organization for Standardization (ISO) gives an extensive list in 5348:2021, detailing mounting methods, and provides recommendations for testing mounts that are not specifically defined. In the nuclear industry on the laboratory scale, there is a need for vibration measurements for predictive maintenance and process monitoring that are nondestructive and capable of working in high-temperature environments. Commercial adhesive products with easy application and removal were tested as nondestructive methods, while quick mounts to a commonly used aluminum frame were tested as nondestructive and have potential applicability in high-temperature environments. The sinusoidal excitation method was used, measuring frequencies from 50 Hz to 10 kHz in one-third octave band intervals, utilizing three accelerometers and comparing the results to those obtained with the stud-mounting method. Using the lowest ±3 dB threshold across each accelerometer, foam dots and poster strips were not successful, and foam tapes were accurate up to 2000 Hz, hose clamps and zip ties up to 800 Hz, and a custom 3D printed mount up to 1000 Hz. Knowing the limitations of each mounting technique allows for accurate measurements within the appropriate range.

47 - OTHER INSTRUMENTATION↗