Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Container”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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

At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

BAM challenges for design testing of waste containers for the final repository KONRAD

Up to the end of this decade, corresponding to the planned date of starting operation in the final disposal KONRAD for non-heat generating waste in Germany, a lot of efforts are needed to condition and package the radioactive waste in containers certified by BfS (Federal Institute for Radiation Protection). This waste is produced by public sector and industry as well as nuclear energy facilities, which result in more than a half of the actually declared quantity growing especially after the phase out decision of nuclear power production in Germany and the subsequent decommissioning of nuclear power plants. BAM (Federal Institute for Materials Research and Testing) acts as responsible authority on behalf of BfS for design testing under consideration of the KONRAD requirements. Within the assessment procedure of containers, BAM has to verify the application documents, including material qualification, container geometry, corrosion protection, leakage rate and operational and accidental loading, and to evaluate quality assurance measures. Besides the previous completed approvals for various containers such as steel sheet and cast iron box-shaped containers and concrete cylindrical ones, BAM is actually carrying assessments for other types such as cast iron cylindrical containers and 'old' steel sheet box shaped ones. The so called 'old' containers present already loaded containers without any KONRAD certification, currently stored at licensed interim storages. In the benefit of the container assessment, BAM operates design testing facilities for drop and fire tests which are also used for research objectives to improve and expand the evaluation methods such as research project ConDrop. The latter deals with numerical simulations and analyses tools for further precise predictions about unfavorable drop test scenarios, safety margins, and design sensitivities for steel sheet containers. Furthermore, during the assessment procedure, several specifications about the containers, the inventory to be disposed and the required safety level have been intensely debated by authorities and applicants. Based on its experience in qualifying containers, BAM has been commissioned to identify insufficiently specified aspects in the waste acceptance criteria and to propose clearer definitions with regard to the secondary regulations in the planning approval notice. This paper focuses on aspects of the contribution of BAM in certifying KONRAD waste containers. This will include the current state of design testing procedures, evaluation methods that are available or are being developed and the main topics within the KONRAD requirements being actually specified by BAM. (authors)

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

Effects of Laser Etching on the Corrosion Susceptibility of SAVY 4000 and Hagan Containers

Since the late 1990’s, the Hagan container was used as the primary container for packaging of plutonium-bearing materials. The Hagan design consisted of a threaded closure, a Viton® ORing, a carbon-carbon filter, and a 304L stainless steel (SS) body. Over the years, Hagans have shown vulnerability in their design [1]. In 2008, The Department of Energy (DOE) issued DOE M 441.1-1, Nuclear Material Packaging Manual, which detailed an approach to obtain highconfidence in containers by including specific design requirements, material contents and an approach to determine life span from said contents, and surveillance techniques [2]. In response to both the vulnerability issues with the Hagan and DOE M 441.1-1, the SAVY 4000 container with its twist style closure, Viton® O-Ring, Fiberfrax-Gortex filter, and annealed 316L SS body, was designed as the replacement for Hagan containers, but only for a short term lifespan of 5 years [1]. However, both the Hagan and SAVY 4000 are being pushed to maintain a lifespan of 40 years. Therefore, proper confidence must be placed on each component of each container to last a minimum of 40 years. So far, the biggest concern found during surveillance of these containers is corrosion and the potential for failure by corrosion. One concern is that the containers fail due to stress corrosion cracking (SCC), especially around the weld between the collar and the body as welds leave residual stresses. One advantage the SAVY 4000 has is that the body is annealed, but its weld is still susceptible as it was welded after annealing [3, 4]. Moreover, 316L SS is known to have a higher pitting resistance (pits are a precursor to SCC and can also lead to extensive failure of the material), than 304L SS [4]. During recent surveillance activities, two SAVY 4000’s containing Solution Assay Instrument (SAI) solutions were opened. The SAI SAVY 4000’s contained plutonium (Pu) in 3M HCl solution in plastic volumetric flasks placed inside of polyethylene bags. Historically, a SAVY 4000 containing an SAI solution is packaged for 3 weeks, however, these particular containers were not reopened until 14 months later. When opened, brown-red corrosion product was found all over the inside of the container as shown in Figure 1.

36 MATERIALS SCIENCE↗
DOE OSTI2024

ILAW Container Cost Reduction and Optimization Study

This study, performed on behalf of Washington River Protection Solutions (WRPS), examines the potential for making cost savings for the Immobilized Low Active Waste (ILAW) containers over the mission life of the Waste Treatment and Immobilization Plant (WTP). it builds on earlier study assessments 24590-LAW-RPT-M-01-001 Rev 0 ILAW Product Container Specifications Optimization Study and DOE/ORP-2013-02 Rev 0 Decision Analysis of Low-Active Waste Container Finishing Handling System Alternatives. The WTP design and installed equipment has evolved over an extended period ~20 years. Consequently, some of the existing requirements relating the ILAW containers may no longer be applicable under current mission plans where the glass filled containers are placed into the Integrated Disposal Facility (IDF) rather than stacked in an engineered purpose-built ventilated store as originally envisioned. The study team reviewed available WTP project reference documents to establish the functions the container must perform and assign requirements against each of these functions. With the assumptions that some of the existing WTP contract restrictions would not necessarily apply to the subsequent facility operations contract the container requirements were reviewed to identify those which were ‘hard constraints‘, imposed by regulatory requirements. Initially each of these hard constraints were reviewed, evaluated and documented to see if they precluded the use of an alternate less expensive material of construction, carbon steel, to the existing 304 L stainless steel in the current container specification. That review did not generate any fatal flaws for a change in material of construction to carbon steel allowing the study team to examine the impacts of material cost savings, potential changes to the design and requirements and impacts on interfaces of the containers with the existing WTP process equipment from container receipt, transfers, glass filling, cooling and export. In addition, a change in container target fill level was evaluated and qualitative, and quantitative where feasible, impacts to the container fabrication process assessed. Lastly, additional follow on development or testing activities to support the use of carbon steel were assessed and a recommended path forward with a ROM cost and timescale developed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2010

Methods of Using a Magnetic Field Response Sensor Within Closed, Electrically Conductive Containers

Magnetic field response sensors are a class of sensors that are powered via oscillating magnetic fields, and when electrically active, respond with their own magnetic fields with attributes dependent upon the magnitude of the physical quantity being measured. A magnetic field response recorder powers and interrogates the magnetic sensors [see Magnetic-Field-Response Measurement- Acquisition System, NASA Tech Briefs Vol. 30, No, 6 (June 2006, page 28)]. Electrically conductive containers have low transmissivity for radio frequency (RF) energy and thus present problems for magnetic field response sensors. It is necessary in some applications to have a magnetic field response sensor s capacitor placed in these containers. Proximity to conductive surfaces alters the inductance and capacitance of the sensors. As the sensor gets closer to a conductive surface, the electric field and magnetic field energy of the sensor is reduced due to eddy currents being induced in the conductive surface. Therefore, the capacitors and inductors cannot be affixed to a conductive surface or embedded in a conductive material. It is necessary to have a fixed separation away from the conductive material. The minimum distance for separation is determined by the desired sensor response signal to noise ratio. Although the inductance is less than what it would be if it were not in proximity to the conductive surface, the inductance is fixed. As long as the inductance is fixed, all variations of the magnetic field response are due to capacitance changes. Numerous variations of inductor mounting can be utilized, such as providing a housing that provides separation from the conductive material as well as protection from impact damage. The sensor can be on the same flexible substrate with a narrow throat portion of the sensor between the inductor and the capacitor, Figure 1. The throat is of sufficient length to allow the capacitor to be appropriately placed within the container and the inductor placed outside the container. The throat is fed through the orifice in the container wall (e.g., fuel tank opening) and connects to the inductor and capacitor via electrical leads to form a closed circuit, Figure 2. Another embodiment is to have the inductor and capacitor fabricated as separate units. In this embodiment, the inductor is mounted external to the container, and the capacitor is mounted internal to the container, Figure 1. Electrical leads are fed through the orifice to connect the inductor and capacitor, Figure 2. When a container holding multiple sensors is made of a conductive material, an antenna can be placed internal to the container. An internal antenna allows all components of the sensors to reside inside the container. The antenna must be separated from the container wall s conductive surface. Additionally, the inductors must be maintained in a fixed position relative to and separated from the container

Woodward, Stanley E.↗
DOE OSTI2025

Measuring 3D Profilometry of SAVY-4000 Nuclear Material Storage Containers: Novacam TubeInspect Capabilities Report

The SAVY-4000 container series is a general-purpose interim storage container for nuclear materials, developed and maintained by Los Alamos National Laboratory (LANL). It is the first vented, general-use nuclear material container to be demonstrated as meeting the requirements outlined in DOE M 441.1-1, the Nuclear Material Packaging Manual. Due to the challenging radiation, thermal, and corrosive storage conditions that the SAVY containers must endure, continuous surveillance techniques are employed to ensure the containers meet all safety standards and specifications. These inspections are typically performed by human operators, who check for issues such as corrosion, O-ring deterioration, corrosion, filter integrity, and potential manufacturing defects. However, human inspections alone are not sufficient, and automated inspection technologies, such as the ATIS system, as well as other automated systems are also utilized. The MicroCam TubeInspect, developed by Novacam Technologies Inc., is designed to address the challenges of understanding how manufacturing variations in the SAVY-4000 container series may affect performance. It is a 3D profilometry measurement system that enables detailed analysis of surface features, including defects, surface roughness, and manufacturing variations. This advanced tool significantly enhances rapid surveillance techniques for both pristine and used containers. In this study, container properties such as surface roughness, thickness, and geometric attributes like circularity are measured for SAVY-4000 containers. Artificially corroded or dented containers are examined to demonstrate the MicroCam's ability to quantify defects. A sensitivity analysis is also conducted, comparing the MicroCam results to those obtained using more precise instruments such as confocal microscopy. This comparison aims to provide valuable insights into container quality, durability, and potential improvements in manufacturing processes.

42 ENGINEERING↗
DOE OSTI2013

Analysis of dose rates to O-ring and membrane for SAVY-4000 container

New generation standard nuclear material containers (NG SNMC aka SAVY 4000) have been designed for storage of various actinide materials whose specific activity is less than or equal to that of heat source plutonium (13.4 Ci/g). The containers also have a limiting decay heat of 25 W. The objective of this analysis is to determine the maximum absorbed dose rate delivered to the Viton o-ring and to the Teflon filter membrane located in the container’s lid under these activity and decay heat constraints. Previous analysis conducted in 2011 showed that the dose rate to the o-ring is higher for the 1-Quart container as compared to the 3-Quart container for the maximum material loading at the density of 3 g cm -3 for all material compositions studied given the size constraints of each container. Ratios of dose rates to the o-ring for the 1-Quart container as compared to the 3-Quart container are summarized in Table 1. In addition, the dose rate to the Teflon filter membrane is 3.4 times higher for the 1-Quart container with MT52+Am as compared to the 3-Quart container with MT52+Am. Thus only the 1-Quart container is analyzed in this report.

61 RADIATION PROTECTION AND DOSIMETRY↗
NASA NTRS2021

Sample Return Container Opening Process Concept for Potential Mars Sample Return

Proposed Mars Sample Return (MSR) missions would require a pristine and secure opening method to ensure sample Contamination Control and backward Planetary Protection. In order to reach the samples, a series of five containers would need to be opened inside of a Mars Returned Sample Handling (MRSH) facility. The process would include opening the Containment Assurance Module (CAM) aboard the Earth Entry Vehicle (EEV), the Secondary Containment Vessel (SCV), Primary Containment Vessel (PCV), and the Orbiting Sample (OS) container, and finally removing and opening the Returned Sample Tube Assemblies (RSTA). Some of these containers can be opened by reversing the mechanical assembly process while others would require destructive means. Several containers are still in a conceptual design phase while others are en route to Mars aboard the M2020 rover. In order to inform the design of the containers that are still in the conceptual stages, a study was performed at each level of the containment assembly. The studies were used to determine the least destructive and most robust methods to disassemble and open the containers while minimizing risk to the samples or the personnel. In addition to the envisioned disassembly and breaching method, contingency plans were made for completing the disassembly if damage to the vehicle and containment system occurred.

Liu, Yang↗
DOE OSTI2021

Towards Improving Container Security by Preventing Runtime Escapes

Container escapes enable the adversary to execute code on the host from inside an isolated container. Notably, these high severity escape vulnerabilities originate from three sources: (1) container profile misconfigurations, (2) Linux kernel bugs, and (3) container runtime vulnerabilities. While the first two cases have been studied in the literature, no works have investigated the impact of container runtime vulnerabilities. In this paper, to fill this gap, we study 59 CVEs for 11 different container runtimes. As a result of our study, we found that five of the 11 runtimes had nine publicly available PoC container escape exploits covering 13 CVEs. Our further analysis revealed all nine exploits are the result of a host component leaked into the container. Here, we apply a user namespace container defense to prevent the adversary from leveraging leaked host components and demonstrate that the defense stops seven of the nine container escape exploits.

42 ENGINEERING↗
DOE OSTI2015

Development of a multi-purpose container system for the storage, transport and disposal of PWR spent fuel in the UK - 15466

Radioactive Waste Management Limited (RWM) is a wholly owned subsidiary of UK's Nuclear Decommissioning Authority, and is responsible for implementing UK Government policy on geological disposal of higher activity radioactive waste. An aspect of the generic work programme that underpins the current generic disposal concepts includes the development of the associated disposal containers. To date RWM has focused on the development of a range of disposal container designs that fulfil the requirements for transport and disposal in a Geological Disposal Facility (GDF) for legacy high heat generating wastes, one of which includes legacy Pressurised Water Reactor (PWR) spent nuclear fuel (SF). In line with a number of European countries exploring deep geological disposal, the conceptual disposal container solution developed for this waste type has a capacity of 4 PWR fuel assemblies. Since 2013, the feasibility of developing an alternative higher capacity disposal container design based around the Multi-Purpose Container (MPC) concept has been explored. In general terms, MPCs are containers that are designed to meet requirements for safe containment of radioactive waste during storage, transport and disposal. The fundamental design philosophy of the MPC is that once the waste is packaged in an MPC, it would not need to be handled again. That is, after a period of initial cooling in a fuel pond, the SF could be packaged for long term dry storage and would not require direct handling for its eventual transport to, and disposal in, a GDF. The novel feature of the MPC system design considered is the inclusion of an additional overpack specifically for the final deep geological disposal phase. The conceptual MPC and its associated storage, transfer, transport and thick walled carbon steel disposal overpack have been developed to a conceptual level of design detail to demonstrate concept feasibility, with the final design solution confirming a MPC system with a capacity of 12 PWR SF assembles. This represents the highest capacity PWR MPC system that complies with the mass and rail gauge limits of the UK W6a rail gauge. Such a capacity realises a contents loading three times that of the standard PWR disposal container and presented technical challenges from both a structural performance and thermal perspective; necessitating the need for the inclusion of features to enhance heat transfer from the central fuel assemblies within the MPC and novel lid features on the transport container overpack to maintain structural integrity and containment during the hypothetical accident conditions of transport tests. In developing the MPC system conceptual design it is recognised that, depending upon the safety functions of the MPC disposal overpack, the choice of carbon steel material may not be suitable for all geological environments. Consequently, the MPC system design and its materials of construction will require further evaluation once detailed knowledge of the disposal concept to be implemented, its safety functions and geological environment are known. (authors)

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

The SD-20K, a large Type B(U)F Package for Waste Containers - 18181

The type B(U)F package SD-20K, currently being developed by DAHER Nuclear Technologies GmbH (DNT), will facilitate the transport of legacy waste from interim storage facilities to the German final repository Konrad. The package SD-20K could also be used in other cases where legacy packages are not fit for a final repository and/or for transport but could be easily conditioned into a suitable package. Since the 1990's German utilities, reactor operators and service providers conditioned waste according to the requirements of the preliminary acceptance criteria for the final repository Konrad. Some of this waste was conditioned in containers which were at that time licensed as type B packages but for which the license expired due to various reasons. Other waste was reclassified so that the used package type does no longer comply with the content. In recent times Konrad containers of type IP-2 are loaded with e. g. core components in compliance with the Konrad acceptance criteria but overloaded with respect to the transport regulations. DAHER Nuclear Technologies GmbH (DNT) in cooperation with Siemens AG investigated possible solutions in a general approach. The result of the evaluation was that the best solution for the transport problem was to develop a versatile type B(U)F package large enough to accommodate all containers and packages accepted for the final repository Konrad. Possible solutions like repacking (high costs and high dose exposure) or requalification (high costs and risk of failure to achieve the goal) were not considered to be generally applicable. The SD-20K packaging is currently designed to accommodate Konrad Containers type III, IV and VI; however, the dimensions of its cavity are sufficient to accommodate any container and package that meets the acceptance criteria for the final repository Konrad. The SD-20K packaging consists of an inner container which constitutes the containment, surrounded by a rather large shock absorbing structure which also contains the thermal protection system. The outer shape of the SD-20K resembles a 20' ISO container. A special feature of the SD-20K is the automatic unloading system to minimize dose exposure of the personnel in the final repository. After arrival in the repository the shock absorbing structure as well as the door of the containment system are opened automatically and the content is moved out of the SD-20K by a remote-controlled carrier system. (authors)

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

Destructive Examination Protocol for 3013-Container-Package Storing Plutonium-Bearing Materials

The 3013-container-package consists of a convenience, inner and outer container and is used for long-term storage of plutonium-bearing materials. A destructive examination (DE) protocol has been developed to examine the container package visually and with microscopic aid to find any corrosion conditions that could result in the loss of the integrity of the container package over its lifetime. The DE protocol contains three main steps: initial container examination, helium (He) leak testing, and detailed imaging and analysis of the inner container closure weld region (ICCWR). The ICCWR has been determined to be bounding, defined as exhibiting worst case conditions for stress corrosion cracking (SCC) of the inner container. To assess SCC in the ICCWR, the inner container lid is cut into quarters and the weld and He-leak testing gasket are removed. Then a citric acid wash is performed to remove adherent chlorides from the ICCWR. The wash is then sent for analysis to determine the concentration of chlorides in the ICCWR. While the analysis for chloride concentration is being performed, the quarter sections are further sectioned into 1/8th subsections by cutting each section in half. These subsections are washed using nitric acid to remove corrosion products. Then each subsection is imaged using a Wide Angle 3-D Measurement System (WAMS). After analysis of microscope images for potential SCC, additional imaging can be performed, including subsurface imaging. After review, a determination is made of whether the container integrity may potentially affect the safe storage of the material.

container↗
DOE OSTI2019

Characterization of Mechanical and Thermal Factors Related to Glovebox Container Water Ingress (Final Report)

This report summarizes tests conducted to quantify water ingress into nuclear material containers when damaged. The damage inflicted is meant to replicate realistic worst-case seismic/fire scenario at the Los Alamos National Laboratory (LANL). Four types (SAVY, Hagan, K-flange, and ROCK) of three-quart containers were subject to drop and then furnace testing. At each stage (as-delivered, post-drop, and post-furnace), water ingress potential was tested by subjecting the containers to a static water column. Key results of these experiments include: • LANL provided two water ingress testing criteria. The first, and most stringent, was a 24 inAq column for a period of one week. These parameters were used for baseline testing on as-delivered containers. The second was a 6 inAq column for a period of two hours. These parameters were used to test containers after impact and furnace testing. • Under the more severe criteria for as-delivered containers, the measured water ingress was: SAVY 7–14 g; Hagan 168–493 g; K-flange 4–5 g; and ROCK 24–280 g of water ingress. • Under the less severe criteria and after drop testing, all containers permitted less than 50 g of water ingress with one exception a ROCK container that had one of three latches release during impact. • Finally, after furnace testing (500°C, 45 minutes), which destroyed the o-ring/gasket materials, and under the same 6 inAq, 2 hr column test, the measured water ingress exceeded 50 g for all but one K-flange and one ROCK container.

42 ENGINEERING↗
DOE OSTI2017

Thermal Analysis of a Pipe Over-pack Container with a Combustible Waste Matrix for Los Alamos National Laboratory - 17307

Los Alamos National Laboratory (LANL) Technical Area 55 (TA 55) utilizes several different container types to store and transfer special nuclear material and waste for numerous programs. The Nuclear Process Infrastructure - Infrastructure Operations Group (NPI-2) analyzes and certifies containers to meet the specifications for distinct processes and storage. Engineering analysis performed on these containers include but are not limited to: leak testing, filter efficiency testing, pressure drop test, drop testing, water penetration testing, polymer (o-ring) hardness testing, and thermal steady state testing. Of all of these tests, thermal steady-state testing is most crucial to ensure critical components for specific containers do not fail under an internal heat payload. This testing provides data for setting heat payload limits for different containers. The Pipe Over-pack Container (POC) is a vented carbon steel container with a removable lid designed for storing or transferring nuclear waste. Additionally, TA 55 has been tasked to manage and store all of its Transuranic (TRU) waste for the foreseeable future. Being able to place more material into a single container conserves the physical storage space available at TA 55. The Pipe Over pack Handling and Operations Manual requires each user to ensure that the maximum temperature of each component in the POC is not exceeded. Conducting a thermal steady-state test with a combustible waste matrix on the POC can benefit stakeholders on increasing the material limits of the container. The results of these measurements will establish new wattage limits to meet LANL's transportation requirements. The current LANL Transportation Safety Document (TSD) limits the amount of heat source plutonium to 10 grams per POC, corresponding to approximately 5 Watts of heat. The result of this test was provided to the Packaging and Transportation (OS-PT) group to evaluate new wattage limits that can satisfy the needs of TA 55, while maintaining a safe packaging configuration. (authors)

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

Lifetime Extension of the Stainless-Steel Components of the SAVY-4000 Storage Container - 18572

Surveillance of vented, stainless-steel Hagan and SAVY-4000 storage containers with some plutonium materials have revealed evidence of corrosion inside the container. There are ∼3000 items stored in Hagan containers and ∼600 items currently stored in SAVY-4000 containers at TA-55. The Hagan is made of 304L steel while the SAVY is made of 316L and should be less susceptible to corrosion. Given the relatively thin body wall of these containers (nominally 0.762 millimeters), it is essential to understand the effects of corrosion on the lifetime of the containers to ensure worker safety. Studies are being conducted on both container types to determine the relative susceptibility of the two containers to corrosion. This paper will describe the potential cause of the corrosion, the extent of corrosion, the types of corrosion observed, along with the results of accelerating aging experiments. The combined results of the surveillance and lifetime extension studies will be used to predict the container lifetime. (authors)

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

Development of the 3 m{sup 3} Low Level Radioactive Waste Container in Taiwan - 20037

Chinshan Nuclear Power Plant stopped its business operation in July, 2019, and Kuosheng Nuclear Power Plant will also be permanently shut down in 2023. Nuclear Reactor Facilities Regulation Act requires a nuclear facility must be dismantled in 25 years after being permanently ceased operation. According to the approved decommissioning plan of Chinshan Nuclear Power Plant and the proposed one of Kuosheng Nuclear Power Plant, the 3 m{sup 3} low level radioactive waste container is planned to store the low level radioactive wastes with higher radiation produced during the decommissioning. However, 3 m{sup 3} low level radioactive waste containers are not used or developed in Taiwan before. To meet such need, Institute of Nuclear Energy Research develops a new type of 3 m{sup 3} low level radioactive waste container and performs the corresponding analyses and tests based on Guidelines of Applying the Usage License of Low Level Radioactive Waste Containers and Regulations for the Safe Transport of Radioactive Material. In this article, it provides the information of the low level radioactive waste containers that have been approved or is under reviewing in Taiwan, and so do the design specifications of the developed 3 m{sup 3} waste container. Besides, the numerical evaluation and test results of the developed 3 m{sup 3} waste container are also included in this article. Based on the evaluation and test results, the developed 3 m{sup 3} waste container satisfies the requirements of the storage container and industrial package type 2. Institute of Nuclear Energy Research will submit the usage license applications to the authority this year, and the usage licenses would be granted in 2021, expectedly. (authors)

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

Sensitivity Analyses of Natural Convection in the AP1000 Passive Containment Cooling System Following LBLOCA Using CFD

The AP1000 power plant has implemented multiple layers of passive safety systems to enhance reactor safety and ensure system integrity during postulated scenarios, such as loss of coolant accident (LOCA) and main steam line break (MSLB). Among these, the passive containment cooling system (PCCS) is a safety-related system designed to prevent the containment from exceeding the design limits of both pressure and temperature following a postulated design basis accident, by transferring heat from the steel containment vessel to the atmosphere. During LOCA, natural air convection plays a vital role in removing heat from the steel containment vessel to the atmosphere. This is integrated with the condensate film from the Passive Containment Cooling Water Storage Tank (PCCWST). Air natural convection is also the only heat removal mechanism after the dry-out of the PCCWST and during the loss of shutdown decay heat removal (LOSDHR) event. Therefore, it is essential to investigate the thermal hydraulics behavior of the containment under transient conditions to ensure its safety and integrity. This paper presents a simplified CFD/ANSYS FLUENT model developed to analyze the impact of different parameters, such as air relative humidity, air temperature, and steel containment temperature, on the natural convection capability to remove the decay heat following LOCA. The model aims to examine the thermal behavior of the containment and provides a comprehensive understanding of the system's natural convection capability during postulated accidents. The results obtained from the model can be used to improve the safety and integrity of the containment system and provide valuable insights for future research in the field.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Balancing Performance and Portability with Containers in HPC: An OpenSHMEM Example

There is a growing interest in using Linux containers to streamline software development and application deployment. A container enables the user to bundle the salient elements of the software stack from an application’s perspective. In this paper, we discuss initial experiences in using the Open MPI implementation of OpenSHMEM with containers on HPC resources. We provide a brief overview of two container runtimes, Docker & Singularity, highlighting elements that are of interest for HPC users. The Docker platform offers a rich set of services that are widely used in enterprise environments, whereas Singularity is an emerging container runtime that is specifically written for use on HPC systems. We describe our procedure for container assembly and deployment that strives to maintain the portability of the container-based application. We show performance results for the Graph500 benchmark running along the typical continuum of development testbed up to full production supercomputer (ORNL’s Titan). The results show consistent performance between the native and Singularity (container) tests. The results also showed an unexplained drop in performance when using the Cray Gemini network with Open MPI’s OpenSHMEM, which was unrelated to the container usage.

Naughton III, Thomas↗
DOE OSTI2017

Corrosion Testing of 304L SS 3013 Inner Container and Teardrop Samples

The Department of Energy (DOE) 3013 Standard specifies a minimum of two containers to be used for the storage of plutonium-bearing materials containing at least 30 wt.% plutonium and uranium. Three nested containers are typically used, the outer, inner, and convenience containers, shown in Figure 1. Both the outer and inner containers are sealed with a weld while the innermost convenience container must not be sealed. Lifetime of the containers is expected to be fifty years. The containers are fabricated of austenitic stainless steels (SS) due to their high corrosion resistance. Potential failure mechanisms of the storage containers have been examined by Kolman and Lillard et al.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗