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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Idaho National Laboratory Classified Mockup Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Idaho National Laboratory (INL) Classified Mockup Low-Level Waste (LLW) (INEL208599QR0, Revision 0 [INL 2020]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Classified Mockup LLW meets all performance objectives of U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The INL Classified Mockup LLW stream is recommended for acceptance without conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Idaho National Laboratory Classified Mockup Low-Level Waste, Revision 1, at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Idaho National Laboratory (INL) Classified Mockup Low-Level Waste (LLW) (INEL208599QR0, Revision 1 [INL 2021]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Classified Mockup LLW meets all performance objectives of U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The INL Classified Mockup LLW stream is recommended for acceptance without conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Distributed fiber optic strain sensing of bending deformation of a well mockup in the laboratory

Well integrity is critical to the safety and success of subsurface energy exploration and management, as leakage of fluids from subsurface reservoirs is often induced by compromised wells. This study investigates bending deformation of a mockup of an oil/gas well that is subjected to loads expected in deviated wells under reservoir compaction and assesses the feasibility of utilizing distributed fiber optic strain sensing to monitor the deformation. Here, a well mockup, which consists of outer and inner steel pipes with a cemented annulus, is tested under three-point bending loading and its strain and curvature development is monitored by Brillouin optical time domain reflectometry/analysis (BOTDR/A) as well as optical frequency domain reflectometry (OFDR). The primary objective of this research is to assess the strain sensing performance of newly fabricated fiber optic cables and to identify key cable characteristics which could improve the quality of distributed strain measurements with BOTDR/A. Results show that the tight-buffered cable is best suited for strain sensing as its maximum errors in the strain measurement were -36% and -24% against conventional sensors at the maximum elastic and plastic bending loads, respectively, whereas those of the non-tight-buffered cable were -45% and -71%, respectively. Similar trends were obtained in the bending curvature measurement. The detailed design of the tight-buffered cable is presented to elucidate key characteristics of such a cable, which will facilitate accurate distributed strain sensing in oil and gas wells.

42 ENGINEERING↗
DOE OSTI2026

PIP-II LB650 Cryomodule Mockup

Fermilab’s PIP-II project requires accurate cable routing in the LB650 cryomodule, a confined and difficult-to-access system. We designed a full-scale mockup using a 15-ft, 7-ft 8020 frame and twenty-three PLA interfaces. CAD modeling and lifting analysis has been verified for safe assembly With all components ordered, the mockup will allow technicians to rehearse routing and identify issues before working on the real cryomodule.

Oribello, Francis [Northern Illinois U.]↗
DOE OSTI2022

Cell G 7606A Mockup Measurement Analysis Methods and Results

The production of research isotopes at the Radiochemical Engineering Development Center (REDC) invariably leads to the accumulation of radioactive waste. Storage of this waste onsite at ORNL serves as an interim solution prior to shipment to long term storage facilities, however onsite capacity is limited. Proper characterization of the activity of waste products is essential for determining the appropriate waste stream and, ultimately, mitigating the cost of disposal. This is typically done via gamma spectrometry and use of the In-Situ Object Counting System (ISOCS), a software package from Mirion that serves as an accepted community standard. However, variability in the contents, density, and activity distribution in waste containers can introduce large errors in ISOCS quantification. A measurement campaign using a mockup of the proposed setup in Cell G seeks to quantify the magnitude and source of these systematic errors. Analysis of this data will guide the creation of ISOCS geometry templates and measurement methods designed to minimize overall uncertainty in reported activities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Coupling of CTF and TRANSFORM using the Functional Mockup Interface

An in-memory coupling between the sub-channel thermal hydraulics code COBRA-TF (CTF), which is included in the Virtual Environment for Reactor Applications (VERA), and the systems code Transient Simulation Framework of Reconfigurable Models (TRANSFORM) was developed in this work. Data exchange is accomplished by using the Functional Mock-Up Interface (FMI), an open standard for coupling dynamic systems models together. The FMI-based coupling necessitated the development of a novel FORTRAN wrapper for communicating with Functional Mock-Up Units. This wrapper will facilitate future FMI-based code couplings with other FORTRAN-based programs. CTF-TRANSFORM is exercised on a simplified Molten Salt Reactor Experiment (MSRE) model in steady-state and transient configurations. The coupled CTF-TRANSFORM model is shown to predict core temperature deltas similar to those available in historical MSRE operational data and is shown to be robust to fast power transients.The coupling paves the way to performing sensitivity and design studies impossible with VERA/CTF alone, in which secondary and tertiary loop operating conditions and design parameters may be perturbed and their impact on the core operating conditions studied.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Augmented Monitoring and Condition Assessment Program (AMCAP) - Proof-of-Principle (POP) Mockup for Non-Aluminum Spent Nuclear Fuel Container In-Situ Examinations

A disciplined engineering approach is being followed to develop an engineered system of tooling and sensors, characterization techniques, and deployment subsystems for in-situ inspection of the several container types used for the storage of non-aluminum spent nuclear fuel (NASNF) in L Basin under the Augmented Monitoring and Condition Assessment Program (AMCAP). Inspecting the containers to provide information on their structural condition helps ensure the safe handling and storage of the NASNF containers pending final disposition. This report describes the work performed in the initial two phases of this developmental work, namely the bench scale and proof-of-principle (POP) scale, which focus on sensor selection and the tooling design and fabrication for two remote non-destructive examination (NDE) methods. These methods include visual testing (VT) for a visual examination of the container surfaces and ultrasonic testing (UT) for a n examination to characterize container wall material thickness and flaws.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Progress Report on Mockup Irradiation Capsule Fuel Measurements at LANSCE

Irradiation tests are a key component of nuclear fuel development and identifying typical and atypical regions in the irradiated fuel volume relies on very few characterization techniques. The goal of the effort reported here is to provide complementary measurements adding to the available parameter space for post irradiation examination as well as to inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Pulsed neutrons, enabling diffraction as well as energy-resolved neutron imaging and neutron absorption resonance spectroscopy, offer unique capabilities for this purpose. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel specimen with spatial resolution of 1 mm 3 to 1 cm 3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm 3 was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions, lattice strains (indicative of residual stresses or chemistry variations) and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotope densities of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Uncertainty Quantification and Sensitivity Analysis of Non-Nuclear Advanced Controls Testbed Reactor Mockup

The research presented in this report describes our progress in applying stochastic methods and uncertainty quantification, parametric study, and variance-based sensitivity analysis (also known as Sobol sensitivity analysis) to a full-core model of a nuclear thermal propulsion (NTP) system simulated with Griffin, with the goal of developing a reduced order (surrogate) model which can be rapidly sampled while perturbing multiple input parameters. In this NTP system, reactivity and power feedback affect the rotation of control drums, which are controlled by a hybrid proportional, integral and derivative (PID) controller, actuated by the power demand and reactivity feedback from the numerical model. This model uses reactor kinetic feedback (mean generation time and $\beta$ from a transient Griffin simulation executed with the improved quasi-static method to provide the kinetic parameters) as inputs to functions which control the CD rotation angle. Using a number of stochastic method approaches, we developed a dual purpose training-surrogate model of the NTP system using polynomial regression. The trained model can be rapidly sampled while simultaneously perturbing various input parameters of the model, such as coefficients on the PID control, or temperature (directly affect the neutron cross section). The surrogate model delivers accurate results orders-of-magnitude faster (minutes, not days) than the base model. Once the base model has been trained, distributions of the uncertain parameters can be changed at will to investigate the effects of perturbing multiple inputs and their effect on the output. For example, coefficients used in the PID control system may vary due to some physical interference, or there may be uncertainty in the temperature of the neutron cross sections in various regions of the reactor. A distribution can be placed on these parameters and operational boundaries can be determined. The goal of this work is to support development of an advanced control system to operate CDs in a functioning NTP system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Update on Full-Scale Retrieval and Transfer System Testing for Pneumatic Conveyance of High-Level Calcine Waste - 20241

The Idaho National Laboratory processed spent nuclear fuel from 1953 to 1994. The process used solvent extraction to separate uranium, resulting in a liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive solid powder, referred to as calcine. The calcine is stored in stainless steel bins contained within concrete vaults called the Calcined Solids Storage Facilities. There are a total of seven CSSFs. The Calcine Retrieval Project is tasked with retrieving the calcine from the bins of CSSF 1 and pneumatically transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. Once the calcine is removed from CSSF1, that facility may be closed, reducing the footprint of waste storage at the INL. The same process for removing Calcine from CSSF1 may then be utilized to remove the calcine from the remaining CSSFs for final treatment and transport to a long-term storage facility. To overcome these inherent challenges with accessing CSSF 1 and extracting the calcine in a timely manner, a full-scale integrated mockup has been constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. All essential elements of the transport design are represented in the Mockup, including motive and makeup air supply (air compressor), a replica of the CSSF6 cyclone, pre-filter, HEPA filter, vacuum pump, and retrieval control devices. Construction on the Mockup was completed in 2018 and testing began in February 2019. It has already proven very beneficial to the CRP. Experience in running the Mockup have aided in the development of retrieval and transport methodologies and the selection of process parameter values. It has also uncovered some unforeseen technological challenges and will continue to be valuable to the testing team as these hurdles are overcome and the process is proven effective. Operations personnel have also benefitted from its use as a training tool and will continue to use the Mockup as a simulator to reproduce problems encountered during extraction and transfer of Calcine from CSSF 1. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Pretest Modeling A Spent Nuclear Fuel Seismic Shake Test

The U.S. Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program is planning to conduct a series of full-scale shake table tests to simulate hypothetical earthquake conditions and record the response of surrogate spent nuclear fuel (SNF) assemblies in a dry canister storage system mockup. The shake table motions will represent a range of hypothetical earthquake conditions at hypothetical locations in the continental U.S. to generally define the range of mechanical loads that SNF can be expected to experience during extended dry storage periods. This paper describes the pretest predictions made with LS-DYNA models of the mockup storage systems. The test will use two dry storage system configurations, a mockup vertical concrete cask (VCC) and a mockup horizontal storage module (HSM). The test will use a production-quality canister and basket. Within the canister will be four instrumented fuel assemblies with fuel rods containing surrogate mass and 28 instrumented dummy assemblies that are intended to match the mass and outer dimension of a fuel assembly. The finite element models include models of the VCC and HSM on the shake table to calculate the system level dynamic responses and separate single fuel assembly models to calculate stress and strain in fuel assembly components. Both types of models include nonlinear behavior like rod-to-rod contact and the ability for VCC’s to rock and slide. This paper presents the expected response of the VCC, HSM, and fuel assemblies to the shake table testing that is planned to start in April of 2024. The earthquake conditions represent seismic hazards in the 2,000-to-20,000-year return period range. The test data is expected to confirm the expectation that fuel rod cladding will remain intact, fuel assembly structural components like guide tubes will remain intact, and no significant VCC sliding or tipping will occur in the range of conditions to be tested with the shake table.

Klymyshyn, Nicholas A.↗
DOE OSTI2020

Virtual and Augmented Realities to Support Development, Training, Deployment, and Operations of Remote Systems - 20516

Throughout the U.S. Department of Energy complex, there is a need to deploy new, unique equipment and systems for cleanup activities, waste processing, infrastructure repair, and decommissioning and demolition in radioactive and hazardous environments. The new equipment often consists of custom systems created by integrating and modifying commercial technologies. In addition, the systems are deployed in unique environments due to limited access, waste properties, and the presence of previously installed equipment or infrastructure that has been left in place. The deployment of custom systems that are 'turnkey' ready for first-time installation and application has its risks. Use of test platforms and mockups for cold (i.e., non-radioactive) testing is one method to mitigate that risk: however, full-scale test mockups for operations such as waste retrieval and tank inspection can be expensive and must be modified to incorporate the uniqueness of each application. An alternative to full-scale mockups is to use three-dimensional (3D) virtual environments and virtual reality (VR) technology to develop interactive virtual test beds. These can be used throughout the down-selection, development, system integration, operational planning, operator training, and mission execution phases of a task. Virtual test beds are readily reconfigured, can be used for multiple tasks, support multiple users in parallel, and can accommodate enhancements and changes not originally planned or conceived in the initial design. Virtual test beds also provide a platform for developing and testing augmented reality (AR) tools for improving operator efficiency. A VR experience has been generated for waste retrieval from a Hanford waste tank to present the potential applications of a virtual test bed. Features of the VR experience are presented, potential applications of VR and AR for remote applications are discussed, and development areas for improving future application are considered. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Large Plate Experiment of Chloride-induced Stress Corrosion Cracking in Spent Nuclear Fuel Storage Canisters

This report provides a description of the testing (in progress at the Savannah River National Laboratory) to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The details of the establishment of this plate test, the interim corrosion results (from visual examination), and the general plan for examination of the plate in CY21 to evaluate crack growth behavior (if any) are provided. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage. A “large plate” (approximately 51 x 46 cm with thickness 16 mm and weight about 30 kg) and containing a circumferential weldment from the mockup is the test specimen. The knowledge of the salt, temperature, humidity conditions for chloride-induced stress corrosion cracking (CISCC) including the threshold stress intensity factor, K iscc , were used to setup the large plate test in which a set of part-through-wall and through-wall electrical discharge machined (EDM) defects were positioned in the weld residual stress field of the welded plate material from the mockup canister. A marine salt solution was applied to this plate specimen with an air-brush spray, then dried, and then exposed to a sustained 71-73% RH condition at an ambient room temperature of 22 °C. The salt loading and environmental condition is expected to be aggressive to cause CISCC

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

In-Sodium Testing of a Prototype Thermoacoustic Power Sensor for Sodium-Cooled Fast Reactors

The ultimate goal of this project is to develop and demonstrate a thermoacoustic power sensor (TAPS) for Sodium-Cooled Fast Reactors (SFRs), with potential application also envisioned to other nuclear technologies such as Lead-Cooled Fast Reactors (LFRs), Molten Salt Reactors (MSRs), in addition to Light Water Reactors (LWRs). The project was led by Westinghouse Electric Company, LLC (Westinghouse) and carried out in collaboration with Argonne National Laboratory (ANL) and the University of Pittsburgh. A TAPS is a passive (self-powered), non-invasive (wireless) sensor envisioned for measuring key parameters, such as local temperature and neutron flux, in a nuclear reactor core. The sensor generates pressure waves (i.e., sound waves) with a frequency and amplitude dependent upon nuclear operating conditions (coolant temperature or power changes). The acoustic waves are able to travel through the core and associated structures, and reach to the sensor network placed outside and/or inside of the reactor vessel. These sensors require a very small amount of power which, during loss of power events, can be provided, for example, by harvesting gamma radiation energy, thus resulting in a monitoring system that can function both during normal operation and during loss of power events. Westinghouse and the University of Pittsburgh designed and fabricated TAPS prototypes for Argonne National Laboratory (ANL) to carry out in-sodium testing to evaluate the effects of sodium on the TAPS and the performance of the TAPS technique in sodium. Argonne received a TAPS prototype from Westinghouse, and the prototype was modified such that it can be installed into a test vessel and function in sodium at elevated temperature without potentially leaking. A water mockup test apparatus was constructed to validate proper working of the prototype. An instrumentation and control (I&C) system, running on the National Instruments (NI) LabView platform, was developed to: 1) operate both the water mockup test and the in-sodium test facility; and 2) process and analyze the received acoustic signals from an array of accelerometers and the Argonne sodium-submersible high-temperature acoustic sensor. The prototype was successfully tested in a water bath at different temperatures. Water mockup tests demonstrated that the TAPS prototype is working properly and its resonance frequency changes linearly with the coolant (water) temperature. A TAPS test apparatus was constructed and integrated with the upgraded Under-Sodium Viewing (USV) sodium test facility. The integrated USV-TAPS sodium test facility has been operational. The TAPS prototype and a high-temperature sodium-submersible acoustic sensor (SSAS) developed by Argonne were both installed inside the TAPS test vessel. Being operated within argon cover gas under ambient conditions, the TAPS prototype demonstrated that it was functioning properly with a resonance frequency at 1407.2 Hz, which was successfully detected by the accelerometers mounted on the external wall of the vessel and the high-temperature SSAS installed inside the vessel. After successfully transferring sodium into the vessel, in-sodium tests of the prototype were conducted. Tests of the TAPS prototype demonstrated that the resonance frequency of the TAPS changes linearly with respect to the temperature difference between the interior of the TAPS and bulk sodium. The early tests showed that the TAPS prototype could not establish a continuous and consistent resonance in sodium. The resonance diminished before the prototype reached its operating temperature. A signal postprocessor was added to the DAQ module latterly to isolate interferences, enhance signal conditioning, improve peak detection, and generate resonance frequency versus temperature plots. After testing in molten sodium and immersion at higher temperature for several weeks, the TAPS prototype was able to establish a continuous resonance. Performance evaluation of the TAPS prototype was then conducted in sodium. The tests included the investigation of 1) the effects of the temperature difference between the TAPS and bulk sodium, 2) the effects of sodium flowrate; and 3) the performance of the different sensor-receiver systems positioned inside or outside the vessel. Results of a test demonstrated that, with limited sodium circulation, a continuous and consistent resonance of the TAPS prototype was established occasionally. The tests also demonstrated that, because of the nature of detection principles and mounting methods, the high temperature SSAS is more affected by acoustic noise, while accelerometers are more affected by vibrations, in the test environment. It is unknown why the TAPS prototype only occasionally established a continuous and consistent resonance when the TAPS temperature reached its operating temperature in molten sodium, and why it ultimately failed to resonate at all. A failure modes assessment was conducted and a few potential causes of failure were identified. Different post in-sodium tests were conducted to obtain additional information potentially relevant to the cause of the failure. Nondestructive evaluation techniques are suggested to examine the internal integrity as well as the gas mixture of the prototype. If they prove inconclusive, the prototype should be cut open to conduct a thorough inspection of its internal integrity and determine the state of the gas mixture.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Evaluation of Advanced Phased-Array Techniques: Interim Results [Slides]

Report: “Evaluation of Advanced Phased-Array Techniques: Interim Results”, August 2023 ML23216A009, PNNL-34622 Results showed that detection rate drops as microstructures/materials become more complex Standard PA had the highest detection rates overall Detection rates in the CASS-CASS mockups were lowest overall FMC had the most accurate length sizing in the fine-grained WSS mockups. All methods struggled with length sizing in coarse-grained mockups SNR vs method depended on the specimen type SNR was consistently high with standard PA SNR was consistently the lowest with PWI FMC performed best on fine-grained WSS A lack of data received from some Round Robin participants resulted in low statistics for many of the analyses.

42 ENGINEERING↗