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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Visual Examination of Aluminum Containers for Extended Wet Storage of Non-Aluminum-Clad Spent Nuclear Fuel (FY 2021)

The goal of the Augmented Monitoring and Condition Assessment Program (AMCAP) program is to provide a condition assessment of the storage containers for non-aluminum-clad spent nuclear fuel (NASNF) in L Basin and demonstrate continued safe storage of all NASNF pending retrieval for ultimate disposition. The storage configuration of NASNF is vulnerable to inside-out corrosion attack. The approach being used in AMCAP is to develop and deploy examination methods for remote, in-situ inspection of the bundle storage in VTS and for OSC storage in the OSC racks. A method for Visual Examination (VE) using video-recording cameras was readied as part of the full-scale mockup development, and was deployed in L Basin for the initial condition assessment work. This report describes the development and the deployment of the VE for inspection of selected bundles in L Basin. The first deployment of the VE system was for visually examining five spent fuel bundles. Bundles were chosen based on their vulnerability in terms of inner fuel contents, history, storage duration, storage configuration, and possibility of inside out corrosion due to galvanic coupling of fuel cladding with the aluminum bundle wall.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Distribution of U, Pu and Np in a Mo-99 Recovery Process using Sachtopore NP (titania based) Absorbent Material

A laboratory scale mockup of a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (U-238), plutonium (94% Pu-239), neptunium (Np-237), and non-radioisotopes of common fission products (including Mo). The goal of this work was to determine the amounts of U, Pu, and Np in the liquid waste streams created by the wash streams for the production process. The key question was how the initial acid wash volume will be handled, as this solution contains residual raffinate which amounted to ~ 3% ofthe mass ofthe feed solution in this experiment. If the initial 1.9 bed volumes (BV) of acid wash were recycled as in this experiment then the combined wash solutions can be expected to contain 0.4 g U/L, 2.7 pg Pu/L and < 14 pg Np/L. If, on the other extreme, the entire acid wash solution was discarded as waste, then the combined wash solutions can be expected to contain 11 g U/L, 3.8 pg Pu/L and ~ 80 pg Np/L. The Sachtopore NP absorbent (for Mo-99 recovery) retains a minuscule fraction ofthe U (2.3 mg U per g titania or 0.03% ofthe U in the feed). The uncertainties are too large to determine how close the absorbent is to saturation with U. Less mass but a higher fraction ofthe total Np was retained by the absorbent (0.7 pg Np per g titania or ~2% of the Np in the feed). As has been reported elsewhere, a large fraction ofthe Pu fed to the titania bed was retained by the absorbent (92% of Pu in the feed or 20 pg per g titania).

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

NRC Multiphysics Analysis Capability Deployment (FY2021--Part 1)

This report details progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” The tasks completed for this report are as follows: First, Task 1d: The net radiation transfer method was implemented into MOOSE for modeling reactor cavity cooling system geometries. RCCS models for two experiments were created: (1) Natural Convection Shutdown Heat Removal Test Facility (NSTF) experiment R022, and (2) HTTR VCS mockup. For validation, computed temperature distributions were compared to measured temperatures. Next, Task 4c: An algorithm for computing the pebble bed reactor equilibrium core isotopic com-position was developed and an initial version is implemented into the reactor multi-physics code Griffin. Initial results for a simplified axisymmetric pebble bed reactor are presented. Finally, Task 7: generation of a reference plant model for molten salt cooled pebble bed reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

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

This report is the comprehensive description of the completed experiment to evaluate the corrosion behavior of a large section of welded plate material harvested from the Sandia National Laboratories’ (SNL) full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The plate was subjected to conditions aggressive to chloride-induced stress corrosion cracking for a total time of 2 years. The details of the establishment of this experiment, and the corrosion results are reported. This work is intended to inform aging management programs for SNF canisters under conditions of extended dry storage.

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

IER-539 CED-1: Preliminary Design of a New Horizontal Split Table

This report presents the preliminary design (CED-1) of IER-539, focused on a new Horizontal Split Table (HST) critical assembly machine. The Department of Energy (DOE) National Nuclear Criticality Safety Program (NCSP) funds critical experiment R&D and its Mission and Vision document identifies a long-standing technical gap of a general purpose HST. HSTs have historically been used for experiments that are impractical to assemble on a vertical lift machine (VLM), usually due to their large physical size or weight. While many HSTs were historically employed in the US (including at LLNL), there are currently no surviving HSTs in the US. While there are two critical experiments facilities remaining in the US, the National Critical Experiments Research Center (NCERC) operated by LANL and the Critical Experiments Facility (CX) operated by SNL, these facilities are not equipped with a general purpose HST. There are a number of experiment designs that require an HST, including needs for Defense Programs and Nuclear Energy. Lack of an HST constraints the national ability to perform large and heavy experiments. Large footprints are needed to successfully create mockup experiments for advanced reactors (mock-up cold critical reactor cores, which is a urgent need for new reactor designs using novel materials and longer cycle lengths), to create solution experiments in lattices, or to test nuclear data of heavy metals. A new HST is a vital piece of criticality experiment infrastructure that would allow for the conduct of integral critical experiments to address needs in the nuclear criticality safety, nuclear data, and other DOE stakeholder communities.

42 ENGINEERING↗
DOE OSTI2022

Niowave Flow Testing of the High Power Converter - FY 21

Design work for a 100 kW converter to produce a neutron flux via electron impact on a flowing stream of liquid metal lead-bismuth (LBE) was conducted in FY20. In FY21, the focus shifted to a 20 kW converter which was designed in collaboration with Niowave engineers. A flow visualization model is needed to validate model results for the 20 kW converter and ensure adequate coolant flow over containment structures. This work highlights the mockup that was made with quartz side plates by LANL.

42 ENGINEERING↗
DOE OSTI2022

Operation Plumbbob

On June 24, 1957, a thirty-seven kiloton nuclear device, codenamed Priscilla, exploded 700 feet over the Nevada Test Site’s Frenchman dry lake. Below, on the lakebed, was a bank vault designed by the Mosler Safe Company for one of several Federal Civil Defense Administration experiments. The vault was constructed “to evaluate the effectiveness of a reinforced-concrete vault and steel vault door in providing protection against the effects of a nuclear detonation.” Blast damage to the exterior was largely confined to the side walls, whose concrete was stripped off; its steel reinforcing rods bent and twisted. The vault door was essentially undamaged and was opened with minimal difficulty. The vault contents - sandbags, movie camera, and thermometers - were only slightly disturbed. The camera operated as planned, running 40 ft of film. The film, however, was completely fogged. Two clinical type thermometers registered 88°F, and a 24-hr stylus American temperature recorder registered a constant 88°F. Priscilla was the fifth of twenty-nine tests conducted in 1957 as part of Operation Plumbbob, whose stated purpose was to conduct tests “aimed to attain new knowledge important to the defense of the United States and the Free World.” Individual events included “proof tests of air defense and antisubmarine warheads to be stockpiled, development tests of bomb components and mockups, and design tests of smaller and lighter warheads.” One test, John, was that of an air-to-air missile detonated 18,500 feet over the test site.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Summary of Chloride-induced Stress Corrosion Cracking Experiments with a Spent Fuel Storage Canister Large Plate

This report provides a comprehensive summary of the crack growth rate testing and the experiments to evaluate the corrosion behavior of welded plate material taken from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The information includes that previously reported, and new work on detailed surface characterization of the corrosion attack from the large plate experiment. The work and the results inform aging management programs for the SNF canisters.

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

Volume 10: Flow Test Facilities

Adding additional flow testing facilities has been considered as part of the High Flux Isotope Reactor (HFIR) Sustaining and Enhancing Neutron Science (SENSe) Initiative at Oak Ridge National Laboratory (ORNL) to support HFIR operations and experiments. This prospect has prompted many ideas and discussions regarding potential features, configurations, locations, and applications for the facilities. A working group of ORNL staff members was formed in fiscal year 2022 to recommend one or more configurations to best support future HFIR operations and scientific capacities and to develop order-of magnitude cost estimates and timing. The ideas discussed in this report include options ranging from upgrading existing small-scale testing facilities to building a full-scale HFIR mockup for detailed thermohydraulic testing and fuel assessment.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

CFD-Chemical Model of One-Third Scale Demonstration of DOE Sealed Canister with ATR Fuel

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. A demonstration case for these DOE sealed canisters will be eventually performed with a one-third scale mockup that has been instrumented with thermocouples and gas concentration probes. This study seeks to model the planned canister for validation of the previously developed model, such that confidence in its long-term prediction can be increased. The deployment of the instrumented lid for online monitoring in intended for a period of 10+ years based on previous monitoring of commercial fuel storage; however, the model is still run for the previously used 50-year storage periods. This case is modeled with a G-value using a bi-linear function such that it decreases at higher dose rates to be consistent with experiments. Validation efforts of the model would likely revolve around the results for the 1st year, so the results of this timeframe are also highlighted. For dried fuel of the nominal decay heat (18W), the predicted hydrogen concentration is 0.19% after 1 year, 0.91% after 10 years, and 2.8% after 50 years, with a maximum pressure of 1.26 atm. Consistent with previous modeling, the decay heat of the fuel is the main factor that influences the results. For undried fuel in pure helium, the hydrogen concentration ranges from 0.23 to 1.2% after 1 year, 1.25-6.2% after 10 years and 8.8 to 18.84% after 50 years. For dried fuel in pure helium the hydrogen concentration ranges from 0.06% to 0.34% after 1 year, 0.35-2.11% after 10 years and 0.95 to 6.6% after 50 years. While long-term results in the presence of residual air are mostly the same, early reactions with O 2 can delay significant production of H 2 until it is consumed to form more water vapor, lowering the range to 0.14 to 1.0% after 1 year. The maximum absolute pressure that is reached across any scenario is 2.06 atm. In the event of residual air, the presence of nitric acid is possible in the range of 18-131 ppm after 1 year, 174-1180 ppm after 10 years, and 586-3500 ppm after 50 years. As long as the fuel is sufficiently dried, or of nominal decay heat, a 4% lower flammability limit of hydrogen will not be reached within a 10-year monitoring period.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Divertor Component Testing

Power exhaust is an immense challenge in tokamaks. Commonwealth Fusion Systems (CFS), in collaboration with MIT and others, is working on the design of a compact (R 0 = 1.85 m), high-field (B 0 = 12 T) tokamak for the demonstration of net fusion energy, called SPARC. Empirical scaling laws indicate that the unmitigated heat flux in the boundary plasma will exceed 10 GW/m 2 . Such high heat fluxes will be a formidable challenge. The scope of this work is to utilize the HHF testing expertise maintained by ORNL, used in support of NSTX-U and to demonstrate the thermal performance of tungsten-coated graphitic foam targets, to help CFS and its collaborators to assess the performance of (1) base materials (tungsten and its alloys) and (2) divertor mockups under SPARC-like divertor heat flux conditions. These activities will be performed by ORNL personnel using electron beam exposure facilities at the Pennsylvania State University’s Applied Research Laboratory (ARL). ARL is a Department of Defense University Affiliated Research Center where ORNL has successfully executed NSTX-U and graphitic foam high heat flux testing using electron beam (e-beam) facilities, which is a nonstandard service. For the assessment of base materials, CFS will do an initial thermal analysis, in consultation with ORNL, to understand under what scenarios the materials are most likely to fail. CFS will supply to ORNL various tungsten-based material samples to test with testing taking 3 weeks. ORNL will thermally stress the samples under SPARC divertor-like conditions to either failure or to a prescribed number of pulses. ORNL will record experimental data, including IR surface temperature and embedded thermocouple readings, and report results to CFS

36 MATERIALS SCIENCE↗
DOE OSTI2023

Evaluation of Flaw Detection with SLDV Methods Before/After Mechanical Patch

This report provides a summary of the testing and the experiments to evaluate the utility of laser ultrasonic testing (UT) in detecting and characterizing defects on the surface of steel plates before and after the application of an engineered composite patch for crack repair. Included are the detection and characterization of defects from a section of welded plate material taken from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister which had been pre-flawed and exposed to deliquescent sea salt conditions for 2 years. The information includes a review of previously reported metallurgical characterization, and new work on detailed surface characterization of the corrosion attack using laser UT.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Full Scale HEPA Filter Encapsulation in Ultra-High-Performance Grout - Proof of Concept

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) currently being constructed to treat radioactive waste, includes vitrification facilities for both the high-level waste (HLW) and low activity waste (LAW) fractions. Operation of the WTP will produce contaminated high-efficiency particulate air filters (HEPA), as part of the solid secondary waste (SSW) stream. The HEPA filters receive off-gas from the vessel vent header and primary off-gas treatment system in LAW Facility and remove particulate contaminants including 99 Tc and 129 I salts. The current disposal method for HEPA filters is encapsulation in metal containers using cementitious material (CM) and disposal in the Integrated Disposal Facility (IDF). Results from the 2017 IDF Performance Assessment (PA) WRPS (2018) demonstrated that while compliance is maintained for the 1000 year compliance period mandated by DOE O 435.1 and its accompanying manual, release of constituents from the HEPA filters result in exceedance of the performance objective imposed as the groundwater regulatory limit at later times. For example, at about 1500 years post-closure, solid secondary waste (SSW), including HEPA filters is predicted to become a dominant contributor to 99 Tc release and over the 10,000-year sensitivity analysis period, SSW is the dominant contributor of 129 I release to the groundwater. The estimated release could potentially be reduced if the HEPA filters are not compacted and waste containers could be distributed throughout a large space, thereby diluting the contaminant release. Additionally, a better cementitious material could be used to encapsulate the HEPA filters. One alternative method for disposal of contaminated HEPA filters is encapsulation of the filters in ultra high-performance grout (UHPG). UHPG is a variation of ultra-high-performance concrete (UHPC) is commonly used in the prestressed concrete industry for large structural members. Recent studies of UHPG show it has excellent properties for containing radionuclides such as 99 Tc and 129 I Nichols and Kaplan (2021). This report presents the results of the first attempt to encapsulate a clean, full-size HEPA filter in UHPG and evaluate the effectiveness of the immobilization process and final waste form. A full-scale proof-of-concept simulated waste form was prepared by encapsulating a HEPA filter in a 110-gallon stainless steel (SS) drum using UHPG. A change from Type I/II PC to Type 1L PLC was made after American Rock Products informed the team that they would no longer be using Type I/II by the end of 2024 and the northwest was phasing out Type I/II PC overall. Type I/II PC used in previous studies of UHPG for encapsulation (Nichols and Kaplan 2021). After the UHPG was cured both the scaled mockup and the full-scale simulated waste forms were sectioned for visual examination. UHPG completely encapsulated the filters and bonded to the external surfaces of materials comprising the filters. No cracks were observed in the sectioned waste forms.

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

Development of a Half-Meter Scale Traveling-Wave (TW) SRF Cavity

While a demonstration of TW resonance excitation in the 3-cell structure in 2K liquid helium had been prepared and carried out at Fermilab in collaboration with Euclid Techlabs, the RF design process of 0.5~1 meter scale TW cavity was begun at Fermilab as the next step of TW development towards an accelerator-scale one. Considering the physical dimensions of existing SRF facilities (for fabrication, processing, and cryogenic testing), Fermilab has proposed a half-meter scale TW RF design consisting of a 7-cell structure and a power feedback waveguide (WG) loop. The WG loop design includes the new RF configurations for TW resonance control during a high-power operation. 1-year US-Japan collaboration program focused on EBW optimization for the TW shape iris joint within the narrow gap was awarded and the efforts has been made by Fermilab, Jlab, and KEK. 1-year LDRD program of Fermilab is awarded recently to fabricate a low-cost mockup of the WG loop with new RF configurations and validate them. Here we will present a preliminary 7-cell TW RF design and will report the progress and challenges through the programs awarded.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

FY23 Update: Surface Sampling Activities for the Canister Deposition Field Demonstration

This report describes the results of a field demonstration of the proposed surface sampling techniques and plan for the multi-year Canister Deposition Field Demonstration (CDFD). The CDFD will evaluate salt deposition rates on three commercial 32PTH2 NUHOMS welded stainless steel storage canisters in Advanced Horizontal Storage Modules. Exposure testing is planned for up to 10 years and will incorporate periodic surface sampling campaigns. The goal of the planned dust sampling and analysis is to determine important environmental parameters that impact the potential occurrence of stress corrosion cracking on spent nuclear fuel (SNF) dry storage canisters. Specifically, measured dust deposition rates and deposited particle sizes will improve parameterization of dust deposition models employed to predict the potential occurrence and timing of stress corrosion cracks on the stainless steel SNF canisters. Previously, a preliminary sampling plan was developed, identifying possible sampling locations on the canister surfaces and sampling intervals; possible sampling methods were also described. Building from previous work, this report documents hand sampling from a spent nuclear fuel canister on a transfer skid mockup designed by Sandia National Laboratories. The sampling took place from a boom lift and salts were collected from mounted sample plates. The results of these efforts are presented in this report and compared to previous laboratory-controlled tests. The information obtained from the CDFD will be critical for ongoing efforts to develop a detailed understanding of the potential for stress corrosion cracking of SNF dry storage canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Assembly of CMS Endcap MIP Timing Detector Module at FNAL

The High-Luminosity LHC (HL-LHC) will enable a more detailed exploration of new phenomena thanks to an anticipated increase in collisions where pileup is expected to reach approximately 200 simultaneous interactions. Many CMS systems will be significantly upgraded to prepare for this new era, including the MIP Timing Detector (MTD) project. The MTD is designed to mitigate the effect of pileup and is set to provide a timestamp accurate to 30 ~ 40 picoseconds for every event, ensuring sustained detector performance at HL-LHC. The MTD is divided into two sections, Barrel Timing Layer (BTL) and Endcap Timing Layer (ETL) which utilize different sensor and ASIC technologies due to the difference in active surfaces, irradiation conditions, and installation schedules. The ETL, composed of two double-sided disks, employs the Low Gain Avalanche Detector (LGAD) sensor and the Endcap Timing Readout Chip (ETROC). More than 8,000 modules, each consisting of four LGAD sensors and ETROCs are required for the ETL detector. These modules will be assembled using an automated robotic gantry that guarantees precise placement at a level of 10 micrometers. In addition, the full assembly of ETL modules includes film application with the jig, wire-bonding, encapsulation with the automated dispensing robot for protecting the wire-bonding, and film curing with a vacuum oven. This talk reports on the successfully completed throughput test with mockup components using the gantry and the successful assembly of real functional modules for beam tests at CERN and FNAL, including the first official ETL module.

Apresyan, Artur↗
DOE OSTI2024

Benchmark Study Matrix for Microreactor Geometries Relevant to Multiple Developers

A benchmark study was developed that include design, development, manufacturing, and performance measurement of agnostic reactor relevant geometries to support industry’s adoption of advanced manufacturing in a variety of structures. A matrix of five microreactor component geometries, specifically based on the recent feasibility study on a Marvel microreactor liner, was developed and the Pacific Northwest National Laboratory team initiated one material/process combination, namely 316H using laser powder directed energy deposition (DED). Although the benchmark starts initially with simplistic cubical and cylindrical forms, it builds up to a mock-up of a non-proprietary design that can demonstrate a variety of features potentially useful for presenting knowledge to specific designers of microreactors and for the matter also for other reactor type designers. The initial cubical and cylindrical forms are initial steps to obtain surface features and dimensional responses to the identified process parameters to be used in the non-proprietary design mockup.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗