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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Modeling of Radiolytic Production of Oxidants Relevant to Corrosion of a Used Fuel Container in Deep Geologic Repository Environments - 17641

The Canadian used nuclear fuel disposal concept involves a deep geological repository using a multi-barrier system. A key barrier is the used fuel container (UFC). It consists of an inner vessel of carbon steel, which provides the structural strength to withstand repository loads, and an outer coating of copper acting as an external corrosion barrier. Relative to other design concepts, coated containers have thinner walls that provide less shielding for γ-radiation, raising the possibility of radiolysis-induced corrosion on the exterior of the container. Also, any trapped aerated water inside the container or in the bentonite clay may condense on the container surface, be exposed to γ-radiation and influence the container corrosion. Understanding the effect of radiation on container corrosion is one key factor to consider in the assessment of the integrity and longevity of the UFC. Humid air radiolysis produces HNO{sub 3} as the dominant oxidizing species. A conservative hemispherical water droplet model was developed to predict the accumulation of HNO{sub 3} and its impact on corrosion. Radiolysis of groundwater in the presence of Cl{sup -} mainly produces H{sub 2}O{sub 2} and HOCl. The concentration of these species is <10 mM and their sum remains almost constant. Radiation exposure tests showed radiation-induced corrosion on carbon steel and Cu. A systematic study of the Cu corrosion and the corrosion rate as a function of [H{sub 2}O{sub 2}], [HOCl], pH and γ-radiation is on-going. (authors)

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

Experience by BAM in the Transport Package Safety Assessment of German Waste Containers - 18469

Transport and storage containers for low and mid-level radioactive waste are getting more and more of relevance due to the nuclear phase out decision in 2011. For higher activities Type B(U) approved waste containers will be needed for the shut down and dismantling phases of NPPs. It is expected that large quantities are required in the near future. German waste containers are generally approved for transport and interim storage and are also intended for the final disposal in the Konrad repository. BAM is involved in the authority licensing of transport package designs in Germany. In this context, BAM is responsible for the assessment of safe containment, mechanical, thermal and quality management issues. BAM also operates test facilities and performs drop as well as thermal tests during package licensing procedures. This paper summarizes our experience in the transport package design assessment of Type B(U) waste containers. A general overview of the approval process, the requirements and approaches BAM applies are described. Some examples are used to illustrate different aspects and technical issues we are addressing during the package assessment. In the first part of the paper the specific design aspects of German waste containers are described. Here, a general overview is given e.g. about closure systems, impact limiter designs, and the handling concepts. Furthermore, the wide range of radioactive content and their physical behavior including the impact on the packages assessment are described. The second part is focused on questions about the licensing and assessment process of German waste containers from the BAM authority point of view. The general approaches for the strategy of demonstration are outlined on the basis of the test conditions according to IAEA Regulations SSR-6. Furthermore, particular issues of the mechanical and thermal assessment with respect to the specific test conditions are discussed. For accident conditions of transport, aspects to be mentioned are e.g. the assessment of the lid bolts, the axial gap applied between content and lid, and the thermal behavior of the wood filled impact limiter after the fire test. However, issues of the assessment for routine and normal conditions of transport are addressed in this paper, too. The intention of this paper is introduce recent approval procedure experience in Germany, describing technical evaluation issues and so reduce rounds of questions during applications. (authors)

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

On the Durability of Tin‐Containing Perovskite Solar Cells

Abstract Tin (Sn)‐containing perovskite solar cells (PSCs) have gained significant attention in the field of perovskite optoelectronics due to lower toxicity than their lead‐based counterparts and their potential for tandem applications. However, the lack of stability is a major concern that hampers their development. To achieve the long‐term stability of Sn‐containing PSCs, it is crucial to have a clear and comprehensive understanding of the degradation mechanisms of Sn‐containing perovskites and develop mitigation strategies. This review provides a compendious overview of degradation pathways observed in Sn‐containing perovskites, attributing to intrinsic factors related to the materials themselves and environmental factors such as light, heat, moisture, oxygen, and their combined effects. The impact of interface and electrode materials on the stability of Sn‐containing PSCs is also discussed. Additionally, various strategies to mitigate the instability issue of Sn‐containing PSCs are summarized. Lastly, the challenges and prospects for achieving durable Sn‐containing PSCs are presented.

14 SOLAR ENERGY↗
DOE OSTI2025

An approach for spent nuclear fuel containment integrity verification using gas tagging

Verification of containment integrity is required for spent nuclear fuel (SNF) managed by the commercial nuclear industry and U.S. Department of Energy (DOE), especially after extended storage. Certain SNF storage systems, such as the DOE road-ready dry storage system, hold several packaged containments within a welded over-canister. These packaged containments are called Department of Energy Standard Canisters (DOESCs). DOESC leakage identification is challenging because their containment boundary cannot be accessed for testing and their contents (i.e., SNF and fill gas) are often similar. There are concerns that this could result in costly characterization and repackaging operations of DOE road-ready dry storage systems if compromised DOESCs are suspected. Here, to address these concerns, this paper presents an approach for applying a gas tagging process using xenon to uniquely identify compromised inaccessible containments following extended storage. The containments considered for this application are seven DOESCs, each packaged within a single over-canister. Two different SNF loading configurations from the Advanced Test Reactor and Fort Saint Vrain nuclear power plant are considered. These configurations are used to represent research reactor aluminum-clad spent nuclear fuel (ASNF) and TRi-structural ISOtropic (TRISO) SNF types. Results for this application show that for ASNF and TRISO type fuels for which the selected fuels are representative, the volume of taggant required at loading is determined primarily by the lower detection limit and leak rate of taggant from a compromised DOESC, rather than the amount of fission-generated xenon in the loaded fuel. While the application presented is suited for larger leaks, smaller leaks could be detected by modifying certain design parameters. This gas tagging approach can also be applied to other DOE containments and advanced reactor SNF storage systems.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2014

SU-E-T-273: Radiation Shielding for a Fixed Horizontal-Beam Linac in a Shipping Container and a Conventional Treatment Vault

Purpose: A fixed horizontal-beam linac, where the patient is treated in a seated position, could lower the overall costs of the treatment unit and room shielding substantially. This design also allows the treatment room and control area to be contained within a reduced space, such as a shipping container. The main application is the introduction of low-cost, high-quality radiation therapy to low- and middle-income regions. Here we consider shielding for upright treatments with a fixed-6MV-beam linac in a shipping container and a conventional treatment vault. Methods: Shielding calculations were done for two treatment room layouts using calculation methods in NCRP Report 151: (1) a shipping container (6m × 2.4m with the remaining space occupied by the console area), and (2) the treatment vault in NCRP 151 (7.8m by 5.4m by 3.4m). The shipping container has a fixed gantry that points in one direction at all times. For the treatment vault, various beam directions were evaluated. Results: The shipping container requires a primary barrier of 168cm concrete (4.5 TVL), surrounded by a secondary barrier of 3.6 TVL. The other walls require between 2.8–3.3 TVL. Multiple shielding calculations were done along the side wall. The results show that patient scatter increases in the forward direction and decreases dramatically in the backward direction. Leakage scatter also varies along the wall, depending largely on the distance between the gantry and the wall. For the treatment room, fixed-beam requires a slightly thicker primary barrier than the conventional linac (0.6 TVL), although this barrier is only needed in the center of one wall. The secondary barrier is different only by 0–0.2 TVL. Conclusion: This work shows that (1) the shipping container option is achievable, using indigenous materials for shielding and (2) upright treatments can be performed in a conventional treatment room with minimal additional shielding. Varian Medical Systems.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Characterization Data Package for Containerized Sludge Samples Collected from Engineered Container SCS-CON-210

This data package contains the K Basin sludge characterization results obtained by Pacific Northwest National Laboratory during processing and analysis of four sludge core samples collected from Engineered Container SCS-CON-210 in 2010 as requested by CH2M Hill Plateau Remediation Company. Sample processing requirements, analytes of interest, detection limits, and quality control sample requirements are defined in the KBC-33786, Rev. 2. The core processing scope included reconstitution of a sludge core sample distributed among four to six 4-L polypropylene bottles into a single container. The reconstituted core sample was then mixed and subsampled to support a variety of characterization activities. Additional core sludge subsamples were combined to prepare a container composite. The container composite was fractionated by wet sieving through a 2,000 micron mesh and a 500-micron mesh sieve. Each sieve fraction was sampled to support a suite of analyses. The core composite analysis scope included density determination, radioisotope analysis, and metals analysis, including the Waste Isolation Pilot Plant Hazardous Waste Facility Permit metals (with the exception of mercury). The container composite analysis included most of the core composite analysis scope plus particle size distribution, particle density, rheology, and crystalline phase identification. A summary of the received samples, core sample reconstitution and subsampling activities, container composite preparation and subsampling activities, physical properties, and analytical results are presented. Supporting data and documentation are provided in the appendices. There were no cases of sample or data loss and all of the available samples and data are reported as required by the Quality Assurance Project Plan/Sampling and Analysis Plan.

core sampling↗
DOE OSTI2018

An Assessment of MELCOR 2.1: Containment Thermal-Hydraulic Tests in the Heissdampfreaktor (HDR) Facility

MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. It provides a capability for independently auditing analyses submitted by reactor manufacturers and utilities. In order to assess the adequacy of containment thermal - hydraulic modeling incorporated in the MELCOR code, a key containment test facility was analyzed. This report documents MELCOR code calculations for simulating steam - water blowdown tests performed in the Heissdampfreaktor ( HDR) de-commissioned containment facility located near Frankfurt , Germany . These tests are a series of blowdown experiments in a large scaled test facility ; including some tests with the addition of hydrogen release which are intended to simulate a variety of postulated break s inside large containment buildings. The key objectives of this MELCOR assessment are to study: (1) the expansion and transport of high energy steam - water releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment gas mixing and stratification. Moreover, MELCOR results are compared to the CONTAIN code for the same tests.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells

A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers'’ for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field's Metal eutectic alloy with a melting point of 62°C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions. It is recommended that glass vessels of various volumes ranging from 125 mL to 1 gallon with modified tops containing the specified Field's Metal eutectic alloy which are appropriately designed to avoid flashing spray material releases be added to the DSA container list. These vessels will be utilized in the Shielded Cells for the storage of radioactive liquid samples of sludge slurries. Use of these vessels will also be advantageous for other liquid samples containing accountable amounts of nuclear material throughout other SRNL laboratories. Similarly constructed paraffin vessel lids could also potentially be added to the approved list, though additional testing would be required.

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

Simulations of Criticality Control Overpack Container Compaction at the Waste Isolation Pilot Plant

Criticality Control Overpack (CCO) containers are being considered for the disposal of defense-related nuclear waste at the Waste Isolation Pilot Plant (WIPP). At WIPP, these containers would be placed in underground disposal rooms, which will naturally close and compact the containers closer to one another over several centuries. This report details simulations to predict the final container configuration as an input to nuclear criticality assessments. Each container was discretely modeled, including the plywood and stainless steel pipe inside the 55-gallon drum, in order to capture its complex mechanical behavior. Although these high-fidelity simulations were computationally intensive, several different material models were considered in an attempt to reasonably bound the horizontal and vertical compaction percentages. When exceptionally strong materials were used for the containers, the horizontal and vertical closure respectively stabilized at 43:9 % and 93:7 %. At the other extreme, when the containers completely degraded and the clay seams between the salt layers were glued, the horizontal and vertical closure reached respective final values of 48:6 % and 100 %.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1973

Analysis of temperature and pressure distribution of containers for nuclear waste material disposal in space

A computer program was adapted from a previous generation program to analyze the temperature and internal pressure response of a radioactive nuclear waste material disposal container following impact on the earth. This program considers component melting, LiH dissociation, temperature dependent properties and pressure and container stress response. Analyses were performed for 21 cases with variations in radioactive power level, container geometry, degree of deformation of the container, degree of burial and soil properties. Results indicated that the integrity of SS-316 containers could be maintained with partial burials of either underformed or deformed containers. Results indicated that completely buried waste containers, with power levels above 5 kW, experienced creep stress rupture failures in 4 to 12 days.

Vanbibber, L. E.↗
NASA NTRS1982

Results of tests of MTA-2 TPS on the SRB hold-down bolt blast container

The four solid rocket booster (SRB) hold-down posts are fastened to the mobile launch platform (MLP) with four large nuts. At liftoff the nuts are split with explosive changes to release the SRB/Shuttle. A blast container is placed over the nuts to protect the vehicle from flying debris. The blast container is a reusable part and has to be protected from aerodynamic heating during flight. The thermal protection system (TPS) used to protect these blast containers is cork. Fitting the flat cork sheet to this hemispherical shaped blast container is both time consuming and expensive. Another problem is removing the charred cork and epoxy glue from the blast containers. Replacements of this cork with another TPS material such as MTA-2 was examined. Heating rates along the centerline of the forward facing areas of the blast container were determined. The feasibility of using 1/2 in. MTA-2 on the SRB blast containers for protection from ascent, plume impingement and reentry heating is demonstrated.

Dean, W. G.↗
NASA NTRS2002

The Evaluation and Implementation of a Water Containment System to Support Aerospace Flywheel Testing

High-energy flywheel systems for aerospace power storage and attitude control applications are being developed because of the potential for increasing the energy density and reducing operational costs. A significant challenge facing the development of the test hardware is containment of the rotating elements in the event of a failure during the development and qualification stages of testing. This containment is critical in order to ensure the safety of the test personnel and the facility. A containment system utilizing water as the containment media is presented. Water containment was found to be a low cost, flexible, and highly effective containment system. Ballistic test results and analytical results are discussed along with a description of a flywheel test facility that was designed and built utilizing the water containment system at the NASA Glenn Research Center at Lewis Field in Cleveland, Ohio.

Trase, Larry M.↗
NASA NTRS1984

Method of forming a relatively stable slip of silicon metal particles and yttrium containing particles

The method concerns forming a relatively stable slip of silicon metal particles and yttrium containing particles. In one embodiment, a casting slip of silicon metal particles is formed in water. Particles of a yttrium containing sintering aid are added to the casting slip. The yttrium containing sintering aid is a compound which has at least some solubility in water to form Y.sup.+3 ions which have a high potential for totally flocculating the silicon metal particles into a semiporous solid. A small amount of a fluoride salt is added to the casting slip which contains the yttrium containing sintering aid. The fluoride salt is one which will produce fluoride anions when dissolved in water. The small amount of the fluoride anions produced are effective to suppress the flocculation of the silicon metal particles by the Y.sup.+3 ions so that all particles remain in suspension in the casting slip and the casting slip has both an increased shelf life and can be used to cast articles having a relatively thick cross-section. The pH of the casting slip is maintained in a range from 7.5 to 9. Preferably, the fluoride salt used is one which is based on a monovalent cation such as sodium or ammonia. The steps of adding the yttrium containing sintering aid and the fluoride salt may be interchanged if desired, and the salt may be added to a solution containing the sintering aid prior to addition of the silicon metal particles.

Dickie, Ray A.↗
NASA NTRS2013

Ultrasonic Low-Friction Containment Plate for Thermal and Ultrasonic Stir Weld Processes

The thermal stir welding (TSW) process is finding applications in fabrication of space vehicles. In this process, workpieces to be joined by TSW are drawn, by heavy forces, between "containment plates," past the TSW tool that then causes joining of the separate plates. It is believed that the TSW process would be significantly improved by reducing the draw force, and that this could be achieved by reducing the friction forces between the workpieces and containment plates. Based on use of high-power ultrasonics in metal forming processes, where friction reduction in drawing dies has been achieved, it is believed that ultrasonic vibrations of the containment plates could achieve similar friction reduction in the TSW process. By applying ultrasonic vibrations to the containment plates in a longitudinal vibration mode, as well as by mounting and holding the containment plates in a specific manner such as to permit the plates to acoustically float, friction between the metal parts and the containment plates is greatly reduced, and so is the drawing force. The concept was to bring in the ultrasonics from the sides of the plates, permitting the ultrasonic hardware to be placed to the side, away from the equipment that contains the thermal stir tooling and that applies clamping forces to the plates. Tests demonstrated that one of the major objectives of applying ultrasonics to the thermal stir system, that of reducing draw force friction, should be achievable on a scaled-up system.

Graff, Karl↗
DOE OSTI2017

Passive filtration of air egressing from nuclear containment

A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into an inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Addendum 2 to the Closure Report for CAU 577 Area 5 Chromium Containing Waste Disposal Cells, NNSS, Nevada

Corrective Action Unit (CAU) 577, “Area 5 Chromium Containing Waste Disposal Cells,” includes five low-level waste cells at the Nevada National Security Site Area 5 Radioactive Waste Management Site where buried waste received from Nuclear Fuel Services, Inc., was subsequently determined to contain chromium that exceeded the toxicity characteristic leaching procedure regulatory limit, which would require the waste to carry hazardous waste code D007. CAU 577 was created to satisfy the requirements of the Settlement Agreement (SA) executed between the Nevada Division of Environmental Protection (NDEP) and the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) on April 25, 2019 (NDEP 2019). The SA required that the chromium-containing waste received from NFS would be addressed following the closure process laid out in the Federal Facility Agreement and Consent Order (FFACO). The FFACO process ensures proper closure of the chromium-containing waste and documentation of that closure through FFACO-type documents. The following three Corrective Action Sites (CASs) were closed, and their closure was documented in the Closure Report for Corrective Action Unit 577: Area 5 Chromium Containing Waste Disposal Cells, Nevada National Security Site, Nevada, DOE/EMNV--0030, dated September 2021 (U.S. Department of Energy [DOE] Environmental Management [EM] Nevada Program 2021a): • CAS 05-21-02, Waste Disposal Cell 12 • CAS 05-21-03, Waste Disposal Cell 15 • CAS 05-21-04, Waste Disposal Cell 17 Closure of the following CAS was previously documented in the Addendum to the Closure Report for Corrective Action Unit 577: Area 5 Chromium Containing Waste Disposal Cells, Nevada National Security Site, Nevada, DOE/EMNV--0030-ADD, dated September 2022 (DOE EM Nevada Program 2022): • CAS 05-21-05, Waste Disposal Cell 20 This second addendum to the Closure Report documents the closure activities that have occurred for CAS 05-21-06, Waste Disposal Cell 21. This is the last CAS in CAU 577. The final waste shipment was placed in the waste disposal cell on November 21, 2022. Following this, closure activities began on November 28, 2022, and were conducted according to the Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for CAU 577 (DOE EM Nevada Program 2021b). The following closure activities were performed: • Constructing an engineered evapotranspiration cover • Installing two subsidence monuments and vadose zone monitoring equipment • Seeding the cover with a mixture of native plant species • Installing four concrete monuments on the corners of the cover and placing two use restriction (UR) warning signs on each monument These activities fulfill applicable federal and state regulations for closure of CAS 05-21-06 and minimize potential future exposure pathways to buried waste. Completed closure activities are also consistent with closure of the nine historical Resource Conservation and Recovery Act (RCRA) units included in Section 10.2.2 of the RCRA Permit that governs hazardous waste management activities at the Nevada National Security Site (Permit NEV HW0101) (NDEP 2023). UR documentation for this CAS is included in Appendix B of this report. The post-closure plan is presented in detail in the CADD/CAP for CAU 577 (DOE EM Nevada Program 2021b), and the requirements are summarized in Section 5.2 of this document. In accordance with paragraph 5D of the SA, a request to incorporate the requirements for post-closure monitoring of CAU 577 was included with the permit application for RCRA Permit NEV HW0101 that was submitted in January 2022 (NNSA/NFO 2022). The request included the post-closure requirements for the three CASs that had been closed at the time of submittal of the application as well as requirements that would be implemented upon future approval of closure of the remaining two CASs. All CAU 577 post-closure monitoring requirements have been captured in the April 4, 2023, Revision 7 of the RCRA Permit (NDEP 2023). As the RCRA Permit NEV HW0101 has been revised since the submittal of the original CAU 577 Closure Report (DOE EM Nevada Program 2021a) and Addendum 1 (DOE EM Nevada Program 2022), the post-closure requirements in this Addendum 2 report do not align with the previous documents. Specific changes resulting from the issuance of Revision 7 of the RCRA Permit are discussed in Section 5.2 of this report. The requirements in this report are consistent with the current permit (NDEP 2023) and supersede all requirements listed in the CAU 577 Closure Report and Addendum 1. All CAU 577 post-closure requirements should be conducted in accordance with the version of the RCRA Permit that is current at the time of the activities being performed. The DOE EM Nevada Program is requesting a Notice of Completion from NDEP for closure of CAU 577. Although CAU 577 is not a legacy site, the FFACO process is being followed to ensure proper closure of the chromium-containing waste. Therefore, transfer of CAU 577 from Appendix III of the FFACO to Appendix IV, Closed Corrective Action Units, is requested, as all closure activities have been completed.

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

Evaluation of Pad 18 Spent Mercury Gold Trap Stainless Steel Container Failure

Failure of the Pad 18 spent mercury gold trap stainless steel waste container is principally attributed to corrosion induced by degradation of plasticized polyvinyl chloride (pPVC) waste packaging material. Dehydrochlorination of pPVC polymer by thermal and/or radiolytic degradation is well-known to evolve HCl gas, which is highly corrosive to stainless steel and other metals in the presence of moisture. Degradation of the pPVC packaging material was likely caused by radiolysis in the presence of tritium gas within the waste container, though other degradation mechanisms (aging, thermo-oxidation, plasticizer migration) over 30 years storage may have contributed. Corrosion was also likely enhanced by the crevice in the container weld design, and may have been enhanced by the presence of tritiated water. Similar non-failed spent mercury gold trap waste containers did not show radiographic evidence of plastic packaging or trapped free liquid within the container. Therefore, those containers are not expected to exhibit similar failures. Halogenated polymers such as pPVC subject to degradation can evolve halide gases such as HCl, which is corrosive in the presence of moisture and can generate pressure in sealed systems.

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

Model for Collaboration among Carriers to Reduce Empty Container Truck Trips

In recent years, intermodal transport has become an increasingly attractive alternative to freight shippers. However, the current intermodal freight transport is not as efficient as it could be. Oftentimes an empty container needs to be transported from the empty container depot to the shipper, and conversely, an empty container needs to be transported from the receiver to the empty container depot. These empty container movements decrease the freight carrier’s profit, as well as increase traffic congestion, decrease roadway safety, and add unnecessary emissions to the environment. To this end, our study evaluates a potential collaboration strategy to be used by carriers for domestic intermodal freight transport based on an optimization approach to reduce the number of empty container trips. A binary integer-linear programming model is developed to determine each freight carrier’s optimal schedule while minimizing its operating cost. The model ensures that the cost for each carrier with collaboration is less than or equal to its cost without collaboration. It also ensures that average savings from the collaboration are shared equally among all participating carriers. Additionally, two stochastic models are provided to account for uncertainty in truck travel times. The proposed collaboration strategy is tested using empirical data and is demonstrated to be effective in meeting all of the shipment constraints.

42 ENGINEERING↗