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At least 253 records · Page 14

All sources NASA NTRS DOE OSTI

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.

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DOE OSTI2018

Evaluating Corrosion Effects on the Stainless Steel Components of the SAVY-4000/Hagan Nuclear Material Storage Containers

Surveillance of vented, stainless-steel Hagan and SAVY-4000 storage containers with some plutonium materials has 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-4000 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 describes the potential cause of corrosion, the extent of corrosion, and types of corrosion observed, along with the results of accelerated aging experiments. The combined results of the surveillance and lifetime extension studies establish a technical basis for extending the design life of the stainless steel components of the SAVY-4000 container from 5 years to 15 years.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗

DOE OSTI2021

Rules of Use Report Engineering evaluations of nuclear material storage containers against the packaging requirements at TA-55

A set of obligations in the yearly storage container surveillance program, ensures that a retrospective approach is applied with loaded containers in inventory to confirm compliance with all “users” to ensure nuclear material storage container requirements are being followed. Validation that packaged containers are properly used, within a specific set of container types prescribed “bounding conditions of use,” is based on meeting requirements as identified in operating procedures associated with TA55-DOP-091, TA-55 Nuclear Material Packaging, and PA-RD-01022, Nuclear Material Packaging Requirements. The surveillance plan obligation requires the application of local area nuclear material accountability software or (LANMAS) to produce queries. The data queries are used to assess attributes of containers in storage these investigations are conducted on a bi-annual basis. This rules of use (ROU) compliance process ensures proper usage of storage containers as containment systems and therefore provides effective worker protection. It is structured as a supporting effort by implementing aspects of PA-AP-01207, Nuclear Material Container Safety Management at TA-55.

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DOE OSTI2022

Engineering evaluations of nuclear material storage containers against the packaging requirements at TA-55 (FY22 Rules of Use Report)

36 MATERIALS SCIENCE↗

DOE OSTI2022

A Demonstration of Intelligent Container Surveillance Using Stationary and Mobile Camera Platforms

Surveillance of nuclear material storage containers is required to ensure that the container safety boundary is maintained during the service life of the container. As many of these container types are the first and only containment barrier to release protecting the worker, public, and environment, it is imperative to develop robust surveillance tools to characterize container degradation. Machine learning (ML) techniques have matured quickly in the past decade and are slowly becoming a routine application in data analysis. While more popular architectures have been developed with different applications in mind, these can be easily translated to container surveillance requirements. In this study, we developed a ML model based on the Detectron2 framework to identify slip lid containers and common exterior container defects, primarily dents. Commercial-off-the-shelf cameras are combined into a stationary camera array or are mounted to a robotic arm to provide more mobility for positioning needs. Results from the model evaluations on the images indicate that the stationary camera arrays outperform the accuracy of the mobile camera systems for both individual camera and composite image detections. Overall, however, the detection accuracies of the systems fell short of 50% and were less than satisfactory. Future efforts to improve the system involve focusing on a single camera deployable solution that combines a controllable light source that eliminates one of the biggest environmental factors that influence ML model detection performance. Additional training set collection is planned to build a more robust model that enables accurate detection from a wider range of imaging conditions.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2024

Lifetime extension drop-test of real-world corroded 5 Quart Hagan nuclear material storage container

A 5Qt Hagan container with a 20-year history of nuclear material storage was challenged with three successive drop tests at a height of 3.7 meters. The total mass of the test package was 12.1 kg. The 1st and 2nd drop tests (center of gravity over the container bottom corner but 180 degrees apart on the container bottom face) passed the pre- and post- impact helium leak criterion at less than 1.00E-6 atm-cc/sec (ambient cubic centimeters per second). The 3rd and final test (center of gravity over top corner) failed with a post-impact gross leak of 1.1E-1 atmcc/sec. The RRFMC (Respirable Release Fraction Measurement Chamber) is a drop tower test system that is critical for the sustainability of the SAVY-4000™ series and Hagan-type (NFT Inc. Golden CO) nuclear material storage containers. These are the primary in-use nuclear material storage container types at the Los Alamos National Laboratory TA-55 facility. Results are presented to expand the technical knowledge basis for container lifetime, regarding actual exposure to corrosive gas species on the container inner surfaces. The primary source of general corrosion throughout the container is gaseous hydrogen chloride (HCl). This gas is generated by the degradation of the polyvinylchloride (PVC) bag-out bag. Additionally, in most cases, the nuclear material itself also releases HCl gas (due to residual chemical components associated with the material formation). The RRFMC drop tower gives the end-user the ability record and analyze high-speed video and photography and if needed aerosol mass release measurements. In this report the principal issue is the physical deformation of the 5Qt Hagan container. There were no mass release experiments of test aerosol mass in the present study.

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DOE OSTI2025

Examination of 3013 Containers Baseline Surface Features

The Surveillance and Monitoring Program at Los Alamos National Laboratory (LANL) was tasked with evaluating the baseline features of 3013 containers. This baseline is to be used as a basis for comparison for 3013 containers that had been packaged with corrosive plutonium materials. The LANL team evaluated an unwelded container, an unused welded container, and a welded container that had held plutonium metal without any corrosive impurities. These three containers had features with depths no larger than 6 µm and had similar depth distributions. The features observed in the baseline containers were all shallower than those seen in containers packaged with corrosive plutonium materials. This work establishes workflows for feature identification and measurement, as well as establishment of baseline data for future comparison with corroded containers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

A Vertex-Aligned Model for Packing 4-Hexagonal Clusters in a Regular Hexagonal Container

This paper deals with a problem the packing polyhex clusters in a regular hexagonal container. It is a common problem in many applications with various cluster shapes used, but symmetric polyhex is the most useful in engineering due to its geometrical properties. Hence, we concentrate on mathematical modeling in such an application, where using the “bee” tetrahex is chosen for the new Compact Muon Solenoid (CMS) design upgrade, which is one of four detectors used in Large Hadron Collider (LHC) experiment at European Laboratory for Particle Physics (CERN). We start from the existing hexagonal containers with hexagonal cells packed inside, and uniform clustering applied. We compare the center-aligned (CA) and vertex-aligned (VA) models, analyzing cluster rotations providing the increased packing efficiency. We formally describe the geometrical properties of clustering approaches and show that cluster sharing is inevitable at the container border with uniform clustering. In addition, we propose a new vertex-aligned model decreasing the number of shared clusters in the uniform scenario, but with a smaller number of clusters contained inside the container. Also, we describe a non-uniform tetrahex cluster packing scheme in the proposed container model. With the proposed cluster packing solution, it is accomplished that all clusters are contained inside the container region. Since cluster-sharing is completely avoided at the container border, the maximal packing efficiency is obtained compared to the existing models.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2023

Aluminum-fiber composites containing intermetallic phase at the matrix-fiber interface

A solid aluminum-fiber composite comprising: (i) an aluminum-containing matrix comprising elemental aluminum; (ii) coated or uncoated fibers embedded within said aluminum-containing matrix, wherein said fibers have a different composition than said aluminum-containing matrix and impart additional strength to said aluminum-containing matrix as compared to said aluminum-containing matrix in the absence of said fibers embedded therein; and (iii) an intermetallic layer present as an interface between each of said fibers and the aluminum-containing matrix, wherein said intermetallic layer has a composition different from said aluminum-containing matrix and said fibers, and said intermetallic layer contains at least one element that is also present in the aluminum-containing matrix and at least one element present in the fibers, whether from the coated or interior portion of the fibers. Methods of producing the above-described composite are also described.

Rios, Orlando↗

DOE OSTI2015

Load planning in the dark: RH-TRU waste container load planning at Sandia National Laboratories/New Mexico - 15116

Sandia National Laboratories (Sandia) performed several sets of experiments during the 1970's and 1980's with fresh and irradiated mixed oxide (MOX) and uranium oxide (UO{sub 2}) materials designed to simulate various fuel responses under varying reactor conditions. These research materials were from light water and breeder mixed oxide reactors. The experimental vessels and extraneous piece parts resulting from the experiments were identified as eligible for disposal at the Waste Isolation Pilot Plant (WIPP) as remote-handled (RH) transuranic (TRU) waste. Once Sandia received the approval to dispose at WIPP, discussions began with the Carlsbad Field Office (CBFO) and Nuclear Waste Partnerships, LLC (NWP) to provide Central Characterization Program (CCP) support for certified acceptable knowledge, visual examination, and dose-to-curie for the repackaging effort. In addition to the WIPP support, Sandia was required to prepare a load plan describing the configuration of the parent containers and how they could be repackaged to meet WIPP and other regulatory requirements. However, all of these containers were packaged as material twenty (20) or thirty (30) years ago and the records of the packaging process and configuration were not documented as waste. Therefore, it seemed that the load plans were prepared with limited information regarding parent container configuration, hence in the dark. The design of the Auxiliary Hot Cell Facility (AHCF) in technical area (TA) V, where the repackaging was to take place, along with the shipping requirements for the RH 72-B cask, limited the options for repackaging. This configuration required that 0.114 m{sup 3} (30-gallon) containers be packaged in the hot cell, removed and placed into 0.208 m{sup 3} (55-gallon) containers to mitigate contamination issues. The material had been tracked as accountable nuclear material and Sandia had verified Material Control and Accountability (MC and A) data on each parent container. Based on the MC and A data, scaling factors could be determined and sampling and analysis was not required. However, it meant that parent containers could not be split into two daughter containers, though two or more parent containers could be combined into one daughter. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Drop Testing of a Container for the Storage, Transport and Disposal of Intermediate Level Waste - 18078

Impact tests were performed, as part of a corresponding container's substantiation, during design development of a shielded container proposed for use at Sellafield for waste retrieval from the First Generation Magnox Storage Pond (FGMSP) in the UK. The mechanical test program comprised a 9 m free drop test onto an unyielding target in a container long lid edge down orientation at ambient conditions. Further, a 0,5 m free drop test onto a punch target was performed. Here, the container was orientated with the lid downwards in a way that the punch, a mild steel bar, impacts a filter lid. The test specimen was instrumented with strain gauges and accelerometers for the drop tests. Transient strains at selected points of the inner and outer container walls, at the shielding lid, as well as at the lid bolts were measured during the container's impact. Furthermore, decelerations of the container body, container lid, and the skip were measured. The complex geometrical changes of the container due to impact were determined by optical 3d- deformation measures using the projected fringes method in combination with multi-image photogrammetry. This paper summarizes the performance of the drop tests and various drop test results in context with the design development as well as aspects regarding the associated Finite Element (FE) analyses and post-test evaluation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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↗