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At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Development of an Activation Analysis Methodology to Support the Disposal of the High Flux Isotope Reactor Original Reflector Container

The original beryllium reflector container was installed prior to the first ever High Flux Isotope Reactor (HFIR) cycle and was removed at the end of cycle 382. The original beryllium reflector was part of every HFIR irradiation cycle from September of 1966 to October of 2000. After its service lifetime, the beryllium reflector container has been interim stored at the fuel pool storage, which is located adjacent to the reactor. Due to space limitations in the pool fuel storage, it is necessary to remove the beryllium reflector container and send it to a location that can be stored permanently. Prior to shipping the beryllium reflector container, the radiological activity must be known. The challenge in determining the overall activity as well as the isotopic inventory and volumetric material fractions lays on the fact that some of the original reflector container stainless steel quantities are not known. Therefore, a methodology to determine a conservative activation scenario was developed.

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

Software Quality Assurance for High Performance Computing Containers

Software containers are a key channel for delivering portable and reproducible scientific software in high performance computing (HPC) environments. HPC environments are different from other types of computing environments primarily due to usage of the message passing interface (MPI) and drivers for specialized hard- ware to enable distributed computing capabilities. This distinction directly impacts how software containers are built for HPC applications and can complicate software quality assurance efforts including portability and performance. This work introduces a strategy for building containers for HPC applications that adopts layering as a mechanism for software quality assurance. The strategy is demonstrated across three different HPC systems, two of them petaflops scale with entirely different interconnect technologies and/or processor chipsets but running the same container. Performance consequences of the containerization strategy are found to be less than 5-14% while still achieving portable and reproducible containers for HPC systems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Development and Application of an Oversize Reusable DOT 7A Type A Overpack Container at the Y-12 National Security Complex - 13150

Waste Management personnel at the Y-12 National Security Complex (Y-12) are concluding a multi-year effort to dispose of a large backlog of low-level waste. Six containers presented a particularly difficult technical challenge in that they each contained large robust equipment (mostly salt baths) with elevated levels of highly enriched uranium (exceeding U.S. Department of Transportation (DOT) fissile-excepted quantities). The equipment was larger than the standard 1.2 m x 1.2 m x 1.8 m (4 ft x 4 ft x 6 ft) DOT Specification 7A Type A box and would have been very difficult to size-reduce because of several inches of steel plate (along with insulating block and concrete) in the equipment design. A critical breakthrough for the success of the project involved procuring and developing two oversize reusable DOT Specification 7A Type A (fissile tested) containers (referred to as the CTI Model 7AF-690-SC) that could be used as overpacks for the original boxes of equipment. The 7A Type A overpack containers are approximately 3.5 m long x 2.7 m wide x 2.8 m high (11.7 ft x 8.9 ft x 9.2 ft) with a maximum gross weight of 10,660 kg (23,500 lb) and a payload capacity of 6,804 kg (15,000 lbs). The boxes were designed and fabricated using a split cavity design that allowed the gasketed and bolted closure to lie along the horizontal centerline of the box. The central closure location in this design allows for strengthening of box corners that tend to be points of weakness or failure in 49CFR173.465 drop tests. By combining the split cavity design with large diameter tubing and diagonal cross bracing, drop test requirements of 49CFR173.465(1) and (2) were met and demonstrated through finite element analysis modeling. The development and use of this new container dramatically reduced the need for down-sizing the equipment and allowed the project to meet objectives within cost and schedule targets. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Legacy Waste Containers for Low and Intermediate Level Radioactive Waste in Germany - Actual Situation and Challenges - 17070

Low and intermediate level radioactive waste has been produced in Germany for several decades. While a part of the waste had been deposited in the Morleben repository and the Asse mine which have been operated for just a limited period of time, the major amount is conditioned and packaged in temporarily stored waste containers. Those containers are designated for final disposal in the KONRAD repository being currently under construction and expected to start operation in 2022. As a consequence, an assessment of the waste container documentation and safety is strongly needed to receive an approval for the Konrad repository. This requalification is challenging because those old containers have to meet the same safety requirements as new containers. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Installation and operation of GeoMelt{sup R} in container vitrification in NNL Central Lab active rig hall at Sellafield - 15328

Cleanup and subsequent immobilization of wastes on the UK's NDA estate brings major challenges. With the exception of HLW where vitrification is utilized for immobilization, cement is typically used in the management and immobilization of non-heat generating wastes. However some legacy feeds such as those containing reactive metals can be incompatible with cementation technologies, and the application of engineered cement encapsulating product to raw wastes increases wasteform volume. Cementation has also been developed to be compatible with the repository concept, which provides it with a level of protection from the environment. Potential delays in repository implementation and as yet to be determined repository waste form acceptance criteria for some wastes may require the duration of above ground storage of immobilized wastes to be extended with potential effect on the integrity of the wasteform. Vitrification is a treatment option which destroys organics and hazardous inorganics such as asbestos, reduces wasteform volume, and produces a robust leach resistant wasteform. Kurion's GeoMelt{sup R} In-container Vitrification (ICV){sup TM} is a vitrification process wherein the treatment vessel can serve as an interim-storage, transport, and final long-term storage vessel - thus negating the need to pour glass into a separate container, effectively leading to a much simpler system than traditional waste melters. The inherent volume reduction from vitrification results in a dense product with self-shielding properties, and the GeoMelt treatment process has been demonstrated to homogenize elemental metals and radionuclide contaminants of concern, eliminating hot spots and mitigating criticality concerns. All of which opens up alternatives strategies for storage and disposal. GeoMelt has been proven in the commercial and US DOE waste treatment environment with over 26,000 tonnes of hazardous waste treated to date (half of this radioactive). Kurion and the UK's National Nuclear Laboratory (NNL) plan to demonstrate its application to Sellafield radioactive wastes. A small (engineering) scale demonstrator is being installed in the active rig hall at NNL's Central Laboratory at Sellafield UK with a twofold mission; 1) demonstration of the vitrification of ILW/LLW, and 2) processing of smaller volume orphan wastes that currently have no treatment path. The NNL facility will allow the handling of uranium containing materials which form a substantial component of ILW currently stored on the Sellafield site. Accurate DFs across the melter and off gas system are often difficult to obtain using inactive materials at the low concentrations present in radioactive wastes. Installation in the NNL facility will allow trace quantities of active species such as Cs-137 and Tc-99 to be processed as part of a simulant ILW stream and monitoring will allow accurate decontamination factors (DFs) to be established to support future vitrification facility design. In addition to underpinning melter operation and obtaining DFs across the system, the facility will allow the processing of radioactive waste arisings of low activity which may either be typical of larger waste volumes or may present as orphans without a disposition route. In either case the melter facility will be able to vitrify these wastes into a form that will be suitable for interim storage and disposal. Among the feeds envisaged for processing will be asbestos from power stations, contaminated soils, feeds containing uranium metals, ponds solids, sludge, and miscellaneous active laboratory wastes. The NNL Central Lab capability will be used to analyze the active product to support disposal requirements whether through letters of compliance for the Geological Disposal Facility (GDF) or waste acceptance criteria for the Low Level Waste Repository (LLWR). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Flammability Limits of H2 Based Mixtures in Conditions Relevant to Waste Management in Containment - 18032

The aim of the present work is to assess the risk of explosion in closed containments used for the transportation of nuclear materials or nuclear waste. Indeed, it is very well-known that hydrogen can be produced due to (i) the radiolysis of different materials within the containment, (ii) the thermal decomposition of mainly the organic part in the containment. Since hydrogen has a very low ignition energy and a very wide flammability domain, it is important to determine the risk of ignition of the subsequent mixture produced by the aforementioned mechanisms. The quantity of the hydrogen that can be produced can vary depending on the containment type and on the state of the material/waste being transported. It is then mandatory to have a very good knowledge not only of the flammability domain of hydrogen in air diluted with water vapor. To do so, an experimental work on the flammability domain of multiple ternary mixtures containing H{sub 2}/air/H{sub 2}O{sub vap} is being conducted at the CNRS-ICARE laboratory within collaboration with the CEA. The present work will focus on the experimental determination of the lower flammability limit of H{sub 2}/air/H{sub 2}O{sub vap} using the spherical method. These experiments have been carried out for an initial temperature of 200 deg. C and at three different initial pressures: 0.3, 1 and 2 bar. The monitoring of the flammability conditions has been conducted using fast flame imaging and pressure measurements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Fukushima Containment Dose Assessment

Dose calculations for a BWR Mark I containment are conducted to complement source term research on the meltdowns at Fukushima Daiichi. These efforts are a primary component of Phase II of the OECD/NEA's Benchmark Study on the Accident at Fukushima Daiichi Nuclear Power Station (BSAF). Phase II of BSAF aims to evaluate radionuclide source terms to the containment and to the environment, radionuclide transport behavior, and dose rates inside and near the plant. The radionuclide-focused goals of Phase II make use of thermal-hydraulic and severe accident insights from Phase I of BSAF, which concentrated on severe accident thermal-hydraulics, core degradation and fuel relocation, reactor pressure vessel (RPV) lower head failure, and containment response. The details and assumptions involved in the severe accident progression have first-order impacts on radionuclide transport and consequent dose rates. An MCNP6 model is created to support the BSAF Phase II investigations. The MCNP6 input represents a boiling water reactor (BWR) Mark I containment and pertinent surrounding structures such the reactor building. This paper examines photon dose rates inside and adjacent to the containment. The spatial distribution and energy spectrum of the source, as well as the overall source magnitude, are informed by MELCOR radionuclide transport calculations in conjunction with ORIGEN-S decay calculations. Initial mesh and ring tally calculations are discussed in relation to plant data. Despite the simplified geometry model and uncertainties of the severe accident source, the preliminary calculations yield impressive agreement with the measured data. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Facilitated Strontium Transport by Remobilization of Strontium-Containing Secondary Precipitates in Hanford Site Subsurface

Significantly enhanced immobilization of radionuclides (such as 90Sr and 137Cs) due to adsorption and coprecipitation with neo-formed colloid-sized secondary precipitates has been reported at the U.S. Department of Energy’s Hanford Site. However, the stability of these secondary precipitates containing radionuclides in the subsurface under changeable field conditions is not clear. Here, the authors tested the remobilization possibility of Sr containing secondary precipitates (nitrate-cancrinite) in the subsurface using saturated column experiments under different geochemical and flow conditions. The columns were packed with quartz sand that contained secondary precipitates (nitrate-cancrinite containing Sr), and leached using colloid-free solutions under different flow rates, varying pH, and ionic strength conditions. The results indicate remobilization of the neo-formed secondary precipitates could be possible given a change of background conditions. The remobility of the neo formed precipitates increased with the rise in the leaching solution flow rate and pH (in a range of pH 4 to 11), as well as with decreasing solution ionic strength. The increased mobility of Sr-containing secondary precipitates with changing background conditions can be a potential source for additional radionuclide transport in Hanford Site subsurface environments.

colloid↗
DOE OSTI2017

Microstructural characteristics of σ phase and P phase in Ru-containing single crystal superalloys

Microstructural instability caused by topologically close-packed (TCP) phase precipitation restricts the useful compositional range of advanced Ni-base single crystal superalloys in industrial applications. Limited systematic investigations of TCP formers (Cr and Mo) additions on microstructural evolution of both the σ phase and the P phase in Ru-containing single crystal superalloys have been reported. In this study, the microstructural characteristics of σ phase and P phase were investigated in three Ru-containing superalloys with different levels of Cr and Mo additions at 950 °C and 1100 °C by using phase extraction, X-ray diffraction, scanning electron microscope and high resolution transmission electron microscopy. The experimental results indicated that the high level additions of Cr and Mo promoted the formation of σ phase and P phase, respectively. The amount of σ phase was much higher than that of P phase after long term exposure at 950 °C and 1100 °C. The sheet-like σ phase existed in the alloy with higher Cr addition after thermal exposure at 950 °C and 1100 °C for 1000 h, while the needle-like P phase precipitated in high Mo content alloy after thermal exposure at 1100 °C for 1000 h and the intergrowth of σ phase and P phase was observed after thermal exposure at 950 °C for 500 h. Both the σ phase and P phase were enriched in Re, W, Cr and Mo, but the σ phase contained more Re and Cr while the P phase contained more Mo and Ni, and Ru was found in both phases. The nucleation of σ phase was much easier than P phase due to the more ledge steps in the interfacial structure between σ phase and matrix, as well as the higher partitioning ratios of Re, Cr and Mo. This study is helpful to understand the microstructural evolution of σ phase and P phase, and to optimize the alloy design in Ru-containing superalloys. - Highlights: •Microstructures of σ phase and P phase were characterized in detail. •Cr and Mo influenced the precipitation of σ phase and P phase, respectively. •Partitioning ratios and interfacial relationship decided precipitation behaviors.

36 MATERIALS SCIENCE↗
DOE OSTI2018

High-CO 2 Requirement as a Mechanism for the Containment of Genetically Modified Cyanobacteria

As researchers engineer cyanobacteria for biotechnological applications, we must consider potential environmental release of these organisms. Previous theoretical work has considered cyanobacterial containment through elimination of the CO 2 -concentrating mechanism (CCM) to impose a high-CO 2 requirement (HCR), which could be provided in the cultivation environment but not in the surroundings. In this work, we experimentally implemented an HCR containment mechanism in Synechococcus sp. strain PCC7002 (PCC7002) through deletion of carboxysome shell proteins and showed that this mechanism contained cyanobacteria in a 5% CO 2 environment. We considered escape through horizontal gene transfer (HGT) and reduced the risk of HGT escape by deleting competence genes. We showed that the HCR containment mechanism did not negatively impact the performance of a strain of PCC7002 engineered for L-lactate production. In conclusion, we showed through coculture experiments of HCR strains with ccm-containing strains that this HCR mechanism reduced the frequency of escape below the NIH recommended limit for recombinant organisms of one escape event in 10 8 CFU.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2023

Preliminary Nuclear Containment Vessel Modeling for Multi-Hazard Probabilistic Risk Assessment under Seismic Hazards and Concrete Degradation

The current practice for natural phenomena hazards (NPH) risk assessment of nuclear facilities is to compute the risk for each hazard independently and then compound the total risk as a combination of single hazard risks. This state of practice does not consider correlations between hazards and the cascading impacts to structures, systems, and components (SSCs), and could thus underestimate the NPH risk or overestimate the nuclear facility safety. Events such as the Fukushima Daiichi accident have highlighted the importance of multi-hazard risk considerations to nuclear power plants (NPPs) that quantify the cascading damage effects to SSCs in the risk models. Moreover, the current fleet of NPPs in the United States is aging; these NPPs are now expected to operate well beyond their initially planned design life. Aging-related deterioration can potentially decrease the capacity of critical structures such as containment vessels to withstand NPH. Such aging considerations may not be adequately accounted for by the current NPH risk assessment guidelines. This paper presents a preliminary modeling and simulation of a representative reinforced concrete containment vessel subjected to seismic mainshock and aftershock considering concrete degradation due to alkali silica reaction. The broader aim is to develop multi-hazard time-dependent fragility functions that could be subsequently used in the probabilistic risk assessment (PRA) model. The multi-hazard component comes into play due to the consideration of damage to the containment vessel under seismic loads and concrete degradation. Consideration of concrete degradation also brings into play the time-dependent nature of the containment vessel response. The response of the containment vessel under varying degrees of concrete degradation to seismic loads is investigated. The results presented are simulated using the Multi-hazard Analysis for STOchastic time-DOmaiN phenomena (MASTODON) software for seismic analysis and the Blackbear software for concrete degradation and damage modeling. Both software are open source and developed within the Multiphysics Object-Oriented Simulation Environment (MOOSE).

42 ENGINEERING↗
DOE OSTI2014

Similarity Analysis: Comparison of a Standard 60mL With a 50mL Container for MicroCT

A 50mL container with a one inch outer diameter has been proposed for holding specimen materials scanned at LLNL and Tyndall Air Force Research Laboratory (TAFRL). In order to determine whether or not this new container would significantly affect MicroCT results, a similarity analysis procedure developed at LLNL was used. This procedure compares multiple first and second order statistics obtained from CT reconstructed images generated by the MicroCT analysis process, and prescribes similarity constraints as guidance. For this study, water was scanned on the MicroCT system at LLNL, twice in the standard 60mL container and twice in the proposed 50mL container. Data were analyzed following the Standard Operating Procedure (SOP) for MicroCT analysis. Comparison of analytical results shows that the new container does not satisfy all similarity analysis criteria (certain second order statistics are outside of similarity analysis constraints).

42 ENGINEERING↗
DOE OSTI2014

Similarity Analysis: Comparison of a Standard 60mL With an Anti-Static Container for MicroCT

Pursuant to safety concerns, an anti-static container has been proposed for holding specimen materials scanned by an International Partner (IP) on a MicroCT system. In order to determine whether or not this new container would significantly affect MicroCT results, a similarity analysis procedure developed at LLNL was used. This procedure compares multiple first and second order statistics obtained from CT reconstructed images generated by the MicroCT analysis process, and prescribes similarity constraints as guidance. For this study, water was scanned on a MicroCT system in both a standard container and the proposed anti-static container under Test Plan 83. Data were analyzed following the Standard Operating Procedure (SOP) for MicroCT analysis. Comparison of analytical results shows that the anti-static container satisfies all similarity analysis criteria.

42 ENGINEERING↗
DOE OSTI2016

Feasibility Study of Non-Destructive Techniques to Measure Corrosion in SAVY Containers

Stainless Steel SAVY containers are used to transport and store nuclear material. They are prone to interior corrosion in the presence of certain chemicals and a low-oxygen environment. SAVY containers also have relatively thin walls to reduce their weight, making their structural integrity more vulnerable to the effects of corrosion. A nondestructive evaluation system that finds and monitors corrosion within containers in use would improve safety conditions and preclude hazards. Non-destructive testing can determine whether oxidation or corrosion is occurring inside the SAVY containers, and there are a variety of non-destructive testing methods that may be viable. The feasibility study described will objectively decide which method best fits the requirements of the facility and the problem. To improve efficiency, the containers cannot be opened during the non-destructive examination. The chosen technique should also be user-friendly and relatively quick to apply. It must also meet facility requirements regarding wireless technology and maintenance. A feasibility study is an objective search for a new technology or product to solve a particular problem. First, the design, technical, and facility feasibility requirements are chosen and ranked in order of importance. Then each technology considered is given a score based upon a standard ranking system. The technology with the highest total score is deemed the best fit for a certain application.

42 ENGINEERING↗
DOE OSTI2018

Microscopic Examination of the Corroded Hagan Container Used to Store Molten Salt Extraction Residue XBPS333

A microscopic examination was performed on a Hagan container after corrosion and a white powder were found on the outside surface of the lid and filter cover while in storage in the vault. The inside surfaces of the container were covered by a wipeable brown coating. The coating was removed and portions of the container were sectioned and imaged with optical microscopy. General corrosion was found throughout the inside of the container, and pitting corrosion was observed in several locations on the lid and sidewall. This report describes the type and extent of the corrosion observed and its impact on the storage lifetime of the container.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Evaluation of the Performance of the SAVY Container Subjected to Various Impact Events using Finite Element Methods

The SAVY container is a Safety Class Design Feature utilized both inside engineered features and outside engineered features (e.g. gloveboxes) at TA-55 Plutonium Facility Building 4. These containers were certified as low damage ratio containers by demonstrating their ability to retain material after exposure to impacts and elevated temperatures resulting from a design basis seismic event. Original testing of the SAVY did not evaluate the effects of various facility specific insults to the SAVY while within a DOT Type A Container, caused either by design basis accident or an operational mishap. These insults include: 1. Impact of a structural I-beam (drum upright & drum on side); 2. Impact of a section of Positive Pressure Chilled Water Pipe (PPCWS); 3. Forklift fork tine impact; 4. Fuel truck impact. Abaqus Explicit Dynamics Finite Element Analysis was performed on a nested SAVY/DOT Type 7A container configuration to assess the structural integrity of the SAVY under the accident conditions listed.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Flanged Tritium Waste Containers: Background and Current Path to Disposal

The MLLW Flanged Tritium Waste Container (FTWC) Project has been operational since 2007 when, during an audit of the FTWC waste stream, it was determined that four FTWC containers contained lead in amounts greater than LDR treatment standards. From June to August of 2007, four FTWC containers were shipped to TA-54 for disposal as LLW from WETF. The contents of the FTWCs were characterized as high activity LLW and were thought to be in compliance with LANL's Waste Acceptance Criteria (WAC). During the packaging of these FTWC's, 36 intact squib valves (aka actuators), containing detonated squibs were included with the waste stream After being transported to TA-54 Area G, it was determined that the explosive used in some of the squibs was Normal Lead Styphnate (NLS). A recharacterization of the entire contents in the FTWC's was performed with the conclusion that the squib valves exhibited the RCRA toxic characteristic for Lead (D008), thus resulting in the contents of the FTWCs being reclassified as a Mixed LLW.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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↗