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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Development of disposal container designs for disposal of high level waste and legacy spent fuel in the UK - 15471

Radioactive Waste Management Limited (RWM), a fully owned subsidiary of the Nuclear Decommissioning Authority (NDA), is responsible for the implementation of geological disposal in the UK. Three illustrative concepts of a geological disposal facility (GDF), corresponding to three generic geological environments - higher strength rock, lower strength sedimentary rock and evaporite rock - have been developed to demonstrate the viability of geological disposal of intermediate level waste, high level waste and spent fuel. It is envisaged that HLW which has been encapsulated in Waste Vitrification Plant (WVP) canisters, advanced gas-cooled reactor (AGR) SF which has been stripped of their spacers and packaged into slotted fuel cans, and PWR SF will be packaged into disposal containers for disposal. A set of schematic designs of these disposal containers was developed in the late 2000's. The design concept for the higher strength rock environment was based on the joint Posiva/SKB disposal container design, and the design concept for the lower strength sedimentary rock and evaporate rock illustrative environments was based on the disposal container concept from Switzerland's National Cooperative for the Disposal of Radioactive Waste (Nagra). The size, mass and dimensions of the disposal containers in the scheme designs vary significantly. The construction method and the lid closing method also vary between the disposal containers for the different waste contents. A design study has indicated that optimization of the number of HLW WVP canisters and the number of AGR slotted fuel cans in the disposal containers, standardizing construction method and closure arrangement across the containers for each geology type, providing handling features at both ends of the disposal container, and balancing of shielding around the containers, could bring significant advantage to the overall GDF programme by reducing the number of disposal containers, reducing GDF footprint, and reducing the extent of excavations, spoil generated and materials that are required for construction. A project to develop the scheme designs to conceptual level was carried in 2011 to 2013. In this project, the above changes were taken into account, technical underpinning of the designs was carried out and manufacturing method was considered. Two disposal container variants for different host rock environments have been developed: - Variant 1, for compatibility with the higher strength rock illustrative disposal system design. This is constructed from copper, with a cast iron insert and is based on the joint Posiva/SKB disposal container design. - Variant 2, for compatibility with the lower strength sedimentary rock and evaporite rock illustrative disposal system designs. This is a steel container based on the NAGRA disposal container design. Each variant has three internal configurations to accommodate 3 HLW vitrified canisters, 16 consolidated cans of AGR SF and 4 PWR spent fuel assemblies. These designs have been shown to bring significant benefit to the GDF designs, bringing a significant reduction to the operating programme and GDF programme cost, including reducing the number of disposal containers by 33% for HLW, 50% for AGR SF and reducing the size of the GDF footprint by 25%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

3013 DE Inner Container Closure Weld Corrosion Evaluation

Destructive evaluation (DE) of 3013 containers is one part of the U. S. Department of Energy Integrated Surveillance Program. During standard DE of 3013 containers, visual examinations for pitting and stress corrosion cracking (SCC) are performed on the accessible surfaces of the outer, inner, and convenience containers, which make up the 3013 container. As a result of 3013 DE additional analysis, the area near the inner container closure weld has been identified as being a region of increased corrosion susceptibility, which may provide a pathway for corrosive gases to the outer container. This area has a higher residual stress, an altered microstructure, and less corrosion resistant weld oxides as a result of the welding process as well as a lower temperature than other areas of the container, which may increase the absorption of moisture on the surface. The deposition of moisture in this stressed region could lead to pitting and stress corrosion cracking. During FY2013, the inner container closure weld area was more closely evaluated on several archived samples from DE containers. These containers included FY09 DE2, FY12 DE4, FY12 DE6 and FY12 DE7 and the Hanford High Moisture Container. The additional examinations included visual observations with a stereomicroscope, scanning electron microscopy along with energy dispersive spectroscopy for chemical analysis, and serial metallography of the sidewall and lid that are part of the inner container closure weld region. Pitting was observed in all the samples taken from the closure weld regions of the examined inner containers. This pitting was generally less 20 µm with most less than 5µm. These pits were similar in depth to those observed in the vapor exposed surfaces of teardrops in the shelf life corrosion testing. Cracking was not observed on either the vapor-exposed surfaces of the teardrop coupons or the inner container closure weld region. Further testing is necessary to determine if the conditions in the welded inner container could support SCC during the 50 year life time for the 3013 container.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Container Approval for the Disposal of Radioactive Waste with Negligible Heat Generation in the German Konrad Repository - 12148

Since the license for the Konrad repository was finally confirmed by legal decision in 2007, the Federal Institute for Radiation Protection (BfS) has been performing further planning and preparation work to prepare the repository for operation. Waste conditioning and packaging has been continued by different waste producers as the nuclear industry and federal research institutes on the basis of the official disposal requirements. The necessary prerequisites for this are approved containers as well as certified waste conditioning and packaging procedures. The Federal Institute for Materials Research and Testing (BAM) is responsible for container design testing and evaluation of quality assurance measures on behalf of BfS under consideration of the Konrad disposal requirements. Besides assessing the container handling stability (stacking tests, handling loads), design testing procedures are performed that include fire tests (800 deg. C, 1 hour) and drop tests from different heights and drop orientations. This paper presents the current state of BAM design testing experiences about relevant container types (box shaped, cylindrical) made of steel sheets, ductile cast iron or concrete. It explains usual testing and evaluation methods which range from experimental testing to analytical and numerical calculations. Another focus has been laid on already existing containers and packages. The question arises as to how they can be evaluated properly especially with respect to lack of completeness of safety assessment and fabrication documentation. At present BAM works on numerous applications for container design testing for the Konrad repository. Some licensing procedures were successfully finished in the past and BfS certified several container types like steel sheet, concrete until cast iron containers which are now available for waste packaging for final disposal. However, large quantities of radioactive wastes had been placed into interim storage using containers which are not already licensed for the Konrad repository. Safety assessment of these so-called 'old' containers is a big challenge for all parties because documentation sheets about container design testing and fabrication often contain gaps or have not yet been completed. Appropriate solution strategies are currently under development and discussion. Furthermore, BAM has successfully initiated and established an information forum, called 'ERFA QM Konrad Containers', which facilitates discussions on various issues of common interest with respect to Konrad container licensing procedures as well as the interpretation of disposal requirements under consideration of operational needs. Thus, it provides additional, valuable supports for container licensing procedures. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2011

Processes to Open the Container and the Sample Catcher of the Hayabusa Returned Capsule in the Planetary Material Sample Curation Facility of JAXA

Japanese spacecraft Hayabusa, which returned from near-Earth-asteroid Itokawa, successfully returned its reentry capsule to the Earth, the Woomera Prohibited Area in Australia in Jun 13th, 2010, as detailed in another paper [1]. The capsule introduced into the Planetary Material Sample Curation Facility in the Sagamihara campus of JAXA in the early morning of June 18th. Hereafter, we describe a series of processes for the returned capsule and the container to recover gas and materials in there. A transportation box of the recovered capsule was cleaned up on its outer surface beforehand and introduced into the class 10,000 clean room of the facility. Then, the capsule was extracted from the box and its plastic bag was opened and checked and photographed the outer surface of the capsule. The capsule was composed of the container, a backside ablator, a side ablator, an electronic box and a supporting frame. The container consists of an outer lid, an inner lid, a frame for latches, a container and a sample catcher, which is composed of room A and B and a rotational cylinder. After the first check, the capsule was packed in a plastic bag with N2 again, and transferred to the Chofu campus in JAXA, where the X-ray CT instrument is situated. The first X-ray CT analysis was performed on the whole returned capsule for confirming the conditions of latches and O-ring seal of the container. The analysis showed that the latches of the container should have worked normally, and that the double Orings of the container seemed to be sealed its sample catcher with no problem. After the first X-ray CT, the capsule was sent back to Sagamihara and introduced in the clean room to exclude the electronic box and the side ablator from the container by hand tools. Then the container with the backside ablator was set firmly to special jigs to fix the lid of container tightly to the container and set to a milling machine. The backside ablator was drilled by the machine to expose heads of bolts, which combined the ablator to the outer lid of the container, and after the drilling had been finished, all the bolts were unscrewed and the backside ablator was removed from the container. Then, the container was sent to the Chofu X-ray facility again to examine in detail by a micro X-ray CT instrument in order to reconfirm that the condition of the latches of the lid of container was normal and that its double O-ring seemed to have been sealed after the last X-ray CT analysis.

Fujimura, A.↗
DOE OSTI2017

Qualification of Pre-existing Containers According to the Acceptance Criteria of the Repository Konrad - 17204

As well as working in the field of waste management, Kerntechnische Entsorgung Karlsruhe GmbH (KBE) is, for historical reasons, one of Germany's three constructors of repository containers for the repository Konrad. The focus of this container production is merely on the side of requalification of the existing legacy containers, which are partly already loaded and grouted. To a smaller extend, KTE also produces new containers for its own use because the containers exactly match the waste properties present on the Karlsruhe site. There are three major container types in the process of requalification: - The Type-IV-Drum-Steel-Container FSC-IV-D was successfully licensed in 2014 and regularly produced thereafter. At the same time, the pre-existing containers of this type are re-qualified on the basis of this license by proof of the compliance with this license. - The Type-IV-Product-Container PSC-I-D is identical to the FSC-VI-D in terms of steel construction but also contains a liner of concrete for shielding and protection of the steel construction. This container is designed for large components and demolition waste. After a successful drop test, it is currently in the advanced process of licensing. - Type I cylindrical concrete containers are simple single drum shieldings which were produced in large numbers in the 1970's and 1980's. KTE owns 7,000 of those containers, thus making requalification and licensing cheaper than retreatment of the waste after demolition of the containers. (authors)

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

Testing and Versatility of the SAVY 4000 Nuclear Material Container - 15634

The SAVY 4000 container was designed by Nuclear Filter Technology, Inc. to meet DOE M441.1-1 requirements for safe storage of nuclear material, and it is currently being used at Los Alamos National Laboratory, Livermore National Laboratory, Idaho National Laboratory and Nevada National Security Site. The 316L stainless steel container can be manufactured in a range of sizes from 1 quart to 10 gallon capacity, it can be opened and closed easily by hand, it is sealed with a Viton{sup R} O-ring, and the filter retains particulates and resists liquid water entry while allowing hydrogen gas to escape. This innovative design has proven to provide benefits and uses beyond its original intent. For example, the container was tested in a simulated facility design basis accident involving high temperatures, fuel fire, and subsequent drop, and the damage ratio was measured. This has allowed the LANL plutonium facility to take credit for a material-at-risk reduction of approximately two orders of magnitude for material stored in the SAVY 4000 container. In addition, the water resistant filter allows the nuclear facility to credit the container as a safety feature to mitigate the risk of a criticality accident in the event of facility sprinkler system activation. The container has also been approved to transfer small quantities of solid nuclear material by hand carry between Los Alamos facilities, and it has been approved and used as a waste container for transuranic waste destined for the Waste Isolation Pilot Plant. The 1 quart capacity container and the future 2 quart design both fit inside the 9977 shipping container, and this has facilitated the transport of nuclear material from Idaho National Laboratory to Nevada Nuclear Security Site, obviated the need for repackaging the material for storage upon arrival. Finally, Los Alamos has recently contracted with Nuclear Filter Technology, Inc. to perform testing to certify the SAVY 4000 container as a DOT Type A liquid transport container. This paper will provide an overview of the various testing regimes and describe the resulting versatility of the container for nuclear material containment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2012

Segmented Gamma Scanner for Small Containers of Uranium Processing Waste- 12295

The Segmented Gamma Scanner (SGS) is commonly utilized in the assay of 55-gallon drums containing radioactive waste. Successfully deployed calibration methods include measurement of vertical line source standards in representative matrices and mathematical efficiency calibrations. The SGS technique can also be utilized to assay smaller containers, such as those used for criticality safety in uranium processing facilities. For such an application, a Can SGS System is aptly suited for the identification and quantification of radionuclides present in fuel processing wastes. Additionally, since the significant presence of uranium lumping can confound even a simple 'pass/fail' measurement regimen, the high-resolution gamma spectroscopy allows for the use of lump-detection techniques. In this application a lump correction is not required, but the application of a differential peak approach is used to simply identify the presence of U-235 lumps. The Can SGS is similar to current drum SGSs, but differs in the methodology for vertical segmentation. In the current drum SGS, the drum is placed on a rotator at a fixed vertical position while the detector, collimator, and transmission source are moved vertically to effect vertical segmentation. For the Can SGS, segmentation is more efficiently done by raising and lowering the rotator platform upon which the small container is positioned. This also reduces the complexity of the system mechanism. The application of the Can SGS introduces new challenges to traditional calibration and verification approaches. In this paper, we revisit SGS calibration methodology in the context of smaller waste containers, and as applied to fuel processing wastes. Specifically, we discuss solutions to the challenges introduced by requiring source standards to fit within the confines of the small containers and the unavailability of high-enriched uranium source standards. We also discuss the implementation of a previously used technique for identifying the presence of uranium lumping. The SGS technique is a well-accepted NDA technique applicable to containers of almost any size. It assumes a homogenous matrix and activity distribution throughout the entire container; an assumption that is at odds with the detection of lumps within the assay item typical of uranium-processing waste. This fact, in addition to the difficultly in constructing small reference standards of uranium-bearing materials, required the methodology used for performing an efficiency curve calibration to be altered. The solution discussed in this paper is demonstrated to provide good results for both the segment activity and full container activity when measuring heterogeneous source distributions. The application of this approach will need to be based on process knowledge of the assay items, as biases can be introduced if used with homogenous, or nearly homogenous, activity distributions. The bias will need to be quantified for each combination of container geometry and SGS scanning settings. One recommended approach for using the heterogeneous calibration discussed here is to assay each item using a homogenous calibration initially. Review of the segment activities compared to the full container activity will signal the presence of a non-uniform activity distribution as the segment activity will be grossly disproportionate to the full container activity. Upon seeing this result, the assay should either be reanalyzed or repeated using the heterogeneous calibration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

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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.

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