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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Water Ingress Testing of the Turbula Jar and U-233 Lead Pig Containers

Understanding the water ingress behavior of containers used at the TA-55 Plutonium Facility has significant implications for criticality safety. The purpose of this report is to document the water ingress behavior of the Turbula Jar with Bakelite lid and Viton gaskets (Turbula Jar) used in oxide blending operations and the U-233 lead pig container used to store and transport U-233 material. The technical basis for water resistant containers at TA-55 is described in LA-UR-15-22781, “Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program.” Testing of the water ingress behavior of various containers is described in LA-CP-13-00695, “Water Penetration Tests on the Filters of Hagan and SAVY Containers,” LA-UR-15-23121, “Water Ingress into Crimped Convenience Containers under Flooding Conditions,” and in LA-UR- 16-2411, “Water Ingress Testing for TA-55 Containers.” Water ingress criteria are defined in TA55-AP-522 “TA-55 Criticality Safety Program”, and in PA-RD-01009 “TA55 Criticality Safety Requirements.” The water ingress criteria for submersion is no more than 50 ml of water ingress at a 6” water column height for a period of 2 hours.

42 ENGINEERING↗
DOE OSTI2024

He Leak Testing of 3013 Inner Container Lids

The 3013-container package is used for the long-term storage of plutonium bearing materials at Department of Energy (DOE) sites. The package consists of three nested stainless-steel containers: outer, inner and convenience, which are made of stainless steel. The requirements of the DOE 3013 Standarda, also entail the development of a surveillance program, tasked to ensure structural integrity of the outer container for at least 50 years. The Materials Identification and Surveillance Working Group (MIS-WG) is specified to be responsible for selecting the containers for the surveillance as well as analyzing results. The MIS-WG determined that corrosion is a potential mechanism for container degradation with the most credible type of failure to be stress corrosion cracking (SCC) through wall breach of the inner container closure weld region (ICCWR). A breach of the inner container has the potential of allowing corrosive gases to be released into the safety class outer container.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

COMPUTATION FLUID DYNAMICS ANALYSIS FOR GENERIC SMALL MODULAR REACTOR CONTAINMENT SEPARATE EFFECTS TEST

It is desirable for fourth-generation Small Modular Reactors to be passively cooled in standard and accident operations. Passive Containment Cooling Systems can reject heat from the containment structure, without using pumps or blowers. The targeted design containment structure is a large, domed, stainless steel, cylindrical vessel. In a postulated Design Basis Accident, steam will flash inside containment. Steam condensation occurs on the inner containment wall and transfers heat through the steel containment into a large body of water known as the annular reservoir (AR) surrounding the vessel serving as the ultimate heat sink. Natural circulation drives the flow in the AR and heat will be released to the environment by evaporation of water. Unique containment geometry requires a separate effects test (SET) facility for the verification and validation of the computer code and evaluation model development and assessment for reactor licensing efforts. In this study, STAR-CCM+, a computational fluid dynamics (CFD) code was used to inform the decision-making process on the design of the SET. The CFD simulation modeled, a two-phase turbulent flow with fluid film development and heat transfer for different containment geometries. The Reactor Excursion and Leak Analysis Program will also be used in a code-to-code verification against the CFD results.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2017

Enhancing reproducibility in scientific computing: Metrics and registry for Singularity containers

Here we present Singularity Hub, a framework to build and deploy Singularity containers for mobility of compute, and the singularity-python software with novel metrics for assessing reproducibility of such containers. Singularity containers make it possible for scientists and developers to package reproducible software, and Singularity Hub adds automation to this workflow by building, capturing metadata for, visualizing, and serving containers programmatically. Our novel metrics, based on custom filters of content hashes of container contents, allow for comparison of an entire container, including operating system, custom software, and metadata. First we will review Singularity Hub’s primary use cases and how the infrastructure has been designed to support modern, common workflows. Next, we conduct three analyses to demonstrate build consistency, reproducibility metric and performance and interpretability, and potential for discovery. This is the first effort to demonstrate a rigorous assessment of measurable similarity between containers and operating systems. We provide these capabilities within Singularity Hub, as well as the source software singularity-python that provides the underlying functionality. Singularity Hub is available at https://singularity-hub.org, and we are excited to provide it as an openly available platform for building, and deploying scientific containers.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Residual Stresses in SAVY 4000 and Hagan Container Bodies

Chloride-induced stress corrosion cracking (SCC) has been investigated as a potential failure mechanism for the SAVY 4000 and the Hagan containers used to store plutonium-bearing materials at Los Alamos National Laboratory. This report discusses the regions of the container bodies most susceptible to SCC and the magnitude of the residual stresses in those regions. Boiling MgCl2 testing indicated that for both containers the region near the top weld was most susceptible to SCC. The Hagan showed through wall cracking after 22-24 hours of exposure both parallel (axial stresses) and perpendicular (hoop stresses) to the weld. The SAVY 4000 container showed significant cracking above and below the weld after 47 hours of exposure but there was no visual evidence of a through wall crack and the cracks did not leak water. Two through wall holes formed in the bottom of the SAVY 4000 container after 44-46 hours of exposure. For both containers, average “through wall” residual stresses were determined from hole drilling data 4 mm below the weld. In the Hagan body, average tensile hoop stresses were 194 MPa and average compressive axial stresses were -120 MPa. In the SAVY 4000 body, average compressive hoop stresses were 11 MPa and average tensile axial stresses were 25 MPa. Results suggest that because the Hagan container exhibited through wall cracking in a shorter time in boiling MgCl2 and had the higher average tensile stress, 194 MPa hoop stress, it is more susceptible to SCC than the SAVY 4000 container.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Surveillance Report on SAVY-4000 and Hagan Nuclear Material Storage Containers for FY 2017

In accordance with the SAVY-4000 Surveillance Plan [1] and DOE M441.1-1 requirements, storage container surveillance continued through fiscal year 2017 at Los Alamos National Laboratory. Surveillance items for the year consisted of 8 SAVY-4000 storage containers, 8 Hagan containers, and 39 SAVY-4000 transfer containers. The SAVY-4000 surveillance items ranged in age from 1 year to 5.6 years and the Hagan containers ranged in age from 6.3 years to 17.6 years. The surveillance containers for this year were selected primarily to better understand the extent of corrosion of the stainless steel components of the containers. Accelerated aging studies indicate that the O-ring and filter components of the SAVY-4000 will last at least 40 years under LANL storage conditions. However, the observation of corrosion on the inside of SAVY-4000 and Hagan surveillance containers has shifted the emphasis to understanding both the nature and the extent of corrosion on the stainless steel body. The restriction on handling soluble residues greater than 500 grams continued this year, delaying the surveillance of some items that was scheduled in earlier surveillance plans.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2018

Gas composition observed by destructive examination of 3013 containers

The Department of Energy’s Integrated Surveillance and Monitoring Program implements requirements to measure gas pressure and gas composition during destructive examination of 3013 containers as required by DOE’s 3013 Standard. Gas pressures and gas compositions measured by the Program during destructive examination of 3013 containers from 2007 to 2017 are reported here. The maximum gas pressure within the population of containers that have undergone destructive examination was 299 kPa (43.4 psia) and this container had 0.52 wt% moisture which is higher than the Standard allows. The majority of containers have essentially no gas pressurization. Hydrogen was quantifiable above the measurement uncertainty in less than half (47 out of 114) of the containers that have undergone destructive examination. The amount of hydrogen that is generated within the sealed 3013 containers is always less than the amount predicted by the 3013 Standard’s pressure equation with the maximum observed hydrogen fraction (observed moles of hydrogen/predicted moles of hydrogen) of 0.18. It is recommended that for the purposes of calculating a maximum hydrogen pressure using the 3013 equation, a correction to the amount of hydrogen produced from decomposition of the water be incorporated. The pressure (or moles) of H 2 should be reduced by multiplying by a hydrogen fraction as described in the text. Helium and nitrogen are the other major constituents. Oxygen which is initially present in nearly all of the containers undergoing destructive examination was consumed to low values in all cases. The highest observed oxygen concentration was 1.7%, and in this case the H 2 concentration was essentially zero. Other minor gases observed in some cases included CO 2 with maximum content of 3.5%, CO with maximum content of 0.8%, N 2 O with maximum content of 1.4%, and CH 4 with maximum content of 0.4%.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Simulations of Pipe Overpack Container Compaction at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is an operating geologic repository in southeastern New Mexico for transuranic (TRU) waste from nuclear defense activities. Past nuclear criticality concerns have generally been low at the WIPP due to the low initial concentration of fissile material and the natural tendency of fissile solute to disperse during fluid transport in porous media (Rechard et al. 2000). On the other hand, the list of acceptable WIPP waste types has expanded over the years to include Criticality Control Overpack (CCO) containers and Pipe Overpack (POP) containers. Containers bound for WIPP are bundled together in hexagon shaped 7-packs (six containers surround one container in the center). Two 7-packs are often combined into a TRUPACT-II package for a total of 14 containers. Most TRUPACT-II packages are restricted to a maximum fissile mass equivalent to plutonium (FMEP) between 0.1 and 0.38 kg, but a CCO TRUPACT-II package and a POP TRUPACT-II package are respectively permitted to have 5.32 kg and 2.80 kg FMEP (see Section 3 of US DOE (2013)). Consequently, CCO container criticality after emplacement at the WIPP was evaluated in Saylor and Scaglione (2018), and Oak Ridge National Laboratories is currently at work on POP container criticality analyses.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

SMR Containment Cable and Electrical Penetration Assembly System

GAIN Project CRADA Number NFE-21-08839, entitled “SMR Containment Cable and EPA System” was initiated by the partnership of Oak Ridge National Laboratory (ORNL) and Engineered Solutions Group (ESG) to test an ESG-designed Electrical Penetration Assembly and Containment Cabling System Qualified for not only SMR technologies, but also Advanced Reactors design. An Electrical Penetration Assembly (EPA) is a component used to allow electrical power and signal conductors as well as optical fiber through the nuclear reactor’s containment structure while maintaining a pressure barrier. The EPA ensures the containment's integrity both during the normal operation and also accident conditions by providing a sealed passage for power and signals. This project was undertaken to fill the equipment gap of EPA and Cabling Systems that require much more severe environmental requirements than legacy plant applications present due to their smaller containment volumes that result in high energy densities compared to legacy designs. This high energy density results in severe accident environments and also more severe normal operating conditions.This document describes cost effective approaches to qualify a unique EPA for Small Modular Reactor (SMR) technologies as legacy LWR EPA technologies will likely have inherent material performance insufficiencies. The DBA profiles for SMRs (and some Advanced Reactor Technologies) are more severe than the legacy qualification requirements making the design of the qualification testing system challenging as the temperatures and pressures can approach the limits permitted by ASME Pressure Vessel Code, as well as accident temperatures exceeding the capability of polymeric gaskets and dielectrics. Small Modular Reactors are designed to have, as the name suggests, modularity which implies sized for factory fabrication and subsequent assembly at the power plant site. SMR containment is much smaller than a legacy Light Water Reactor containment. The containment walls are likely to be comprised of stainless steel. Qualifying Electrical Penetration Assemblies (EPAs) through these steel vessels are of interest in this work. Methods are presented here that discuss critical safety and cost effectiveness for these SMR EPA’s test systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Method to produce water-soluble sugars from biomass using solvents containing lactones

A process to produce an aqueous solution of carbohydrates that contains C6-sugar-containing oligomers, C6 sugar monomers, C5-sugar-containing oligomers, C5 sugar monomers, or any combination thereof is presented. The process includes the steps of reacting biomass or a biomass-derived reactant with a solvent system including a lactone and water, and an acid catalyst. The reaction yields a product mixture containing water-soluble C6-sugar-containing oligomers, C6-sugar monomers, C5-sugar-containing oligomers, C5-sugar monomers, or any combination thereof. A solute is added to the product mixture to cause partitioning of the product mixture into an aqueous layer containing the carbohydrates and a substantially immiscible organic layer containing the lactone.

09 BIOMASS FUELS↗
DOE OSTI2017

Method to produce water-soluble sugars from biomass using solvents containing lactones

A process to produce an aqueous solution of carbohydrates that contains C6-sugar-containing oligomers, C6 sugar monomers, C5-sugar-containing oligomers, C5 sugar monomers, or any combination thereof is presented. The process includes the steps of reacting biomass or a biomass-derived reactant with a solvent system including a lactone and water, and an acid catalyst. The reaction yields a product mixture containing water-soluble C6-sugar-containing oligomers, C6-sugar monomers, C5-sugar-containing oligomers, C5-sugar monomers, or any combination thereof. A solute is added to the product mixture to cause partitioning of the product mixture into an aqueous layer containing the carbohydrates and a substantially immiscible organic layer containing the lactone.

09 BIOMASS FUELS↗
DOE OSTI2018

Conformal core cooling and containment structure

A nuclear reactor includes a pressure vessel and a nuclear reactor core comprising fissile material disposed inside the pressure vessel at the bottom of the pressure vessel. A secondary core containment structure includes a containment basket comprising insulation with a maximum stable temperature of at least 2200K cladded by steel. The bottom of the pressure vessel and the nuclear reactor core are disposed inside the containment basket with the containment basket spaced apart from the bottom of the pressure vessel by a clearance gap. The containment basket may comprise zirconia insulation cladded by steel. In some embodiments the clearance gap between the containment basket and the bottom of the pressure vessel is no larger than one meter. The secondary core containment structure may further comprise conduits arranged to inject water into the clearance gap between the containment basket and the bottom of the pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Conformal core cooling and containment structure

A nuclear reactor includes a pressure vessel and a nuclear reactor core comprising fissile material disposed inside the pressure vessel at the bottom of the pressure vessel. A secondary core containment structure includes a containment basket comprising insulation with a maximum stable temperature of at least 2200K cladded by steel. The bottom of the pressure vessel and the nuclear reactor core are disposed inside the containment basket with the containment basket spaced apart from the bottom of the pressure vessel by a clearance gap. The containment basket may comprise zirconia insulation cladded by steel. In some embodiments the clearance gap between the containment basket and the bottom of the pressure vessel is no larger than one meter. The secondary core containment structure may further comprise conduits arranged to inject water into the clearance gap between the containment basket and the bottom of the pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

The SAVY 4000 Container Storage Program at Los Alamos National Laboratory - 17313

Incidents involving release of nuclear materials stored in containers of convenience such as food pack cans, slip lid taped cans, paint cans, etc. has resulted in the U.S. Defense Nuclear Safety Board concerns over the lack of prescriptive performance requirements for interim storage of nuclear materials. Los Alamos National Laboratory (LANL) has shared in these incidents and in response proactively moved into developing a performance based storage container design, the SAVY-4000{sup R}. The SAVY-4000{sup R} is the first vented general use nuclear material container demonstrated to meet the requirements of DOE M 441.1-1, Nuclear Material Packaging Manual. The SAVY-4000{sup R} is an innovative and creative design demonstrated by the fact that it can be opened and closed in a few seconds without torque wrenches or other tools; has a built-in, fire-rated filter that prevents the build-up of hydrogen gas, yet retains 99.97% of plutonium particulates, and prevents release of material even in a 12 foot drop. Finally, it has been fire then drop tested and will reduce the risk to the public in the event of an earthquake/fire scenario. This will allow major nuclear facilities to credit the container towards source term Material at Risk (MAR) reduction. The container was approved for nuclear material storage in the TA-55 Plutonium Facility on March 15, 2011, and there are currently over 2000 SAVY-4000{sup R} containers available for use within the facility. A Safety Analysis Report was developed to demonstrate Manual 441.1-1 compliance and was approved April, 2014. The first four SAVY-4000{sup R} containers were packaged with plutonium on August 2, 2011. Key aspects of the SAVY-4000{sup R} vented storage container design will be discussed which include design description, qualification and testing, surveillance strategy, and the design life extension program as enhanced by surveillance activities with the intent to extend well beyond the current five year design life. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of the B-25 Low Level Radioactive Waste Container in Taiwan - 1900x

Chinshan Nuclear Power Plant will be permanently stopped its operation in the middle of 2019 and will be the first nuclear power plant stepping into the decommissioning phase in Taiwan. According to Nuclear Reactor Facilities Regulation Act, a nuclear facility must be dismantled in 25 years after being permanently ceased operation. In its decommissioning plan, the B -25 low level radioactive waste container was one of the proposed containers used for this decommissioning task. However, the B-25 waste container was not used or developed in Taiwan before. To meet this need, Institute of Nuclear Energy Research developed a new type of B-25 waste container and will apply to the authority for its usage license. This article provided the information of the low level radioactive waste containers have been approved in Taiwan and described the design specifications of the developed B-25 waste container. Besides, the numerical evaluation and test results were also included in this article. Based on the evaluation and test results, the developed B-25 waste container satisfied the requirements of the storage container and industrial package type 2. It is believed that the usage license would be granted soon after applying it to the authority. (authors)

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

A study on computational fluid dynamics modeling of a refrigerated container for COVID-19 vaccine distribution with experimental validation

A key issue with the distribution of vaccines to prevent COVID-19 is the temperature level required during transport, storage, and distribution. Typical refrigerated transport containers can provide a temperature-controlled environment down to -30 °C. However, the Pfizer vaccine must be carefully transported and stored under a lower temperature between -80 °C and -60 °C. One way to provide the required temperature is to pack the vaccine vials into small packages containing dry ice. Dry ice sublimates from a solid to a gas, which limits the allowable transport duration. This can be mitigated by transporting in a -30 °C refrigerated container. Moreover, because the dry ice will sublimate and thereby release CO 2 gas into the transport container, monitoring the CO 2 concentration within the refrigerated container is also essential. In the present work, a 3D computational fluid dynamics model was developed based on a commercially available refrigerated container and validated with experimental data. The airflow, temperature distribution, and CO 2 concentration within the container were obtained from the simulations. The modeling results can provide guidance on preparing experimental setups, thus saving time and lowering cost, and also provide insight into safety precautions needed to avoid hazardous conditions associated with the release of CO 2 during vaccine distribution.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2017

Containment Analysis Capability of UNF-ST&DARDS

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) methodology to perform automated containment analyses for potential transportation packages based on canister loading map information is described in this paper, and its capability is illustrated with example results. The allowable leakage rate is calculated with the procedures provided in ANSI N14.5-2014 and NUREG/CR-6487, which were adapted for containment analysis of a transportation package containing fuel assemblies with different nuclear characteristics (e.g., initial enrichment, burnup, and cooling time) and clad integrity (intact or damaged). UNF-ST&DARDS applies different source term calculation methodologies for low-burnup fuel (LBF) (i.e., <45 GWd/tonne U) assemblies and high-burnup fuel (HBF) (i.e., ≥45 GWd/tonne U) assemblies. The LBF radionuclide activities are based on actual fuel assembly burnup, initial enrichment, and cooling time. Bounding radionuclide activities based on a fuel pellet burnup value of 65 GWd/tonne U and actual fuel assembly cooling time are used for HBF assemblies. The fraction of failed fuel rods and the release fractions for the contributors to releasable source terms recommended in NUREG-1617 are used in the containment analysis regardless of fuel assembly burnup. However, UNF-ST&DARDS supports different parameter values for LBF and HBF assemblies. Finally, the containment analysis methodology for as-loaded transportation packages is presented in detail, and the UNF-ST&DARDS containment analysis capability is illustrated with results for simulated transportation packages containing spent nuclear fuel canisters in dry storage at selected sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program

Criticality safety at TA-55 relies on nuclear material containers that are water resistant to prevent significant amounts of water from coming into contact with fissile material in the event of a fire that causes a breach of glovevbox confinement and subsequent fire water ingress. A “water tight container” is a container that will not allow more than 50ml of water ingress when fully submerged, except when under sufficient pressure to produce structural discontinuity. There are many types of containers, welded containers, hermetically sealed containers, filtered containers, etc.

61 RADIATION PROTECTION AND DOSIMETRY↗