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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Planning for Receipt of Mk-18A Recovered Nuclear Material at Oak Ridge National Laboratory

Oak Ridge National Laboratory (ORNL) ships and receives a wide variety of radioisotopes. These shipments will include the Mark-18A high-burnup targets, which are being processed at Savannah River National Laboratory to recover valuable isotopes that will then be shipped to ORNL for programmatic use. The Mark-18A targets will be processed to recover two products: a plutonium oxide stream and an oxide material that will contain the americium, curium, and lanthanides. A Type A shipping package has been designed, fabricated, and certified to transport the americium/curium/lanthanide oxide product in special form capsules. The package was designed with both gamma and neutron shielding for contents reading up to 5,000 R/hour. The plutonium stream will be shipped to ORNL in a Model 9975 Type B transport cask. A new content envelope for the Type B cask is being developed. This paper discusses the preparation underway at ORNL to receive the packages and transfer the material into hot cell processing and storage facilities for subsequent use in research and development applications and as feedstock for heavy actinide production.

Patton, Bradley↗
DOE OSTI2019

Improvements for LCLS-II Cryomodule Transportation

The Linear Coherent Light Source (LCLS-II) is currently being constructed at the SLAC National Laboratory. A total of 35 cryomodules will be fabricated at Jefferson Lab (JLab) in Virginia and Fermi National Laboratory (FNAL) in Illinois and transported via road to SLAC. A shipping frame with an inner bed isolated by springs was designed to protect the CMs from shocks and vibrations during shipments. Successful road testing of the JLab prototype paved the way for production CM shipments. The initial production shipments lead to several catastrophic failures in beamline vacuum in the cryomodules. The failures were determined to be due to fatigue in Fundamental Power Coupler (FPC) bellows due to excessive motion during shipment. A series of instrumented CM shipping tests and component tests were undertaken to develop a solution. A modified spring layout was tested and implemented, which reduced shocks on the CMs. FPC coupler bellows restraints were tested on a shaker table and on a CM during shipping; they were able to reduce bellows motion by a factor of three. The updated shipping system is currently in use and has successfully delivered six cryomodules to SLAC from JLab and FNAL.

Huque, N. A.↗
DOE OSTI2023

Mitigating Extremist Maritime Threats to Radiological Material Transport Vessels

The 21st century has experienced a rise in the threat of global radicalism and extremist organizations, and there has been oft-reported interest from these groups in obtaining or exploiting radiological materials. The threat of extremists acquiring radiological materials while in transit or sabotaging a transport vessel carrying materials at sea is one that requires increased attention. Transport is already one of the most vulnerable times for any cargo but including the difficulties in securing a large ship on the open ocean makes oceanic transport of radiological materials a space of elevated vulnerability for sabotage, theft, or other attacks. The risks of violent extremist attacks on trade ships and trading routes are serious, as shipping routes sustain the global economy, but an attack on a radiological transport vessel could be a magnitude worse, impacting national and international security. While radicalists have largely remained on land in the past due to minimal maritime expertise or inability to project power out of their immediate vicinity, the world has witnessed past attacks on ships such as the USS Cole and the MV Limburg and it is likely only a matter of time before groups extend further into the sea and engage in more maritime campaigns against material transport vessels. This paper suggests methods for increasing safety and security on oceanic transport voyages through behavioral science principles and insider threat recognition training, as well as a better understanding of adversarial and extremist psychology. By providing transport personnel with the knowledge and support to identify and respond to unique maritime threats by extremists, these principles and training suggestions should advance and increase security on the high seas.

Kinney, Justin↗
DOE OSTI2016

GNF fuel technology update

In 2015, Global Nuclear Fuel (GNF) launched the GNF3 fuel design by inserting lead-use assemblies into two boiling water reactors (BWRs) in the U.S. A lead use assembly is a nuclear fuel assembly which is inserted in a reactor core to confirm its performance. A third set of GNF3 lead-use assemblies are scheduled to be inserted into another reactor in the spring of 2017. GNF3 is a robust next-generation 10*10 fuel design that will improve both fuel cycle and fuel reliability. In addition, GNF3 will be delivered in a fully assembled condition - forgoing the need to insert fuel channels over the bundles at the site. The process of shipping BWR fuel in a channeled configuration is complex and requires an understanding of the potential high-cycle and low-cycle fatigue conditions that can occur during shipping, as characterized by accelerometer data from ship tests. The key component in evaluating channeled fuel shipping is the spacer, and the GNF3 spacer fatigue performance was found to be approximately twice that of GNF2. While fuel reliability has traditionally been related to the avoidance of fuel leakers, for the past 15 years channel - control blade interference has been a close second for concern. To eliminate this problem, GNF3 will come standard with NSF (a zirconium alloy) channels that are effectively resistant to fluence gradient-induced bow and shadow corrosion-induced bow. In addition to the development of GNF3, GNF is evaluating the use of advancing 3D metal printing or additive manufacturing for manufacturing fuel bundle components. Additive manufacture offers the hope of more flexibility in design and shorter manufacturing development cycles. However, since there is little experience with this technology in the nuclear industry, GNF has been evaluating the use of additive manufacture to fabricate small scale components that are easily retrieved from the reactor for testing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of Type A Transport Packagings - 17073a

ORNL ships a wide variety of radioisotopes. Many of these are associated with the DOE Isotope Program and Office of Nuclear Materials Integration (ONMI) and can be shipped as DOT Type A quantities of material. The DOT Specification 6M Type B packaging previously used for many shipments at ORNL is no longer authorized. Most materials could be shipped in Type B packagings, but many of them can meet the requirements for shipping as Type A. As a result, ORNL is working with a packaging manufacturer to develop, test, and certify several Type A packagings. These include a 38 L (10 gal) Type A packaging primarily to be used for transport of Cf-252 sources, a 208 L (55 gal) Type A packaging primarily to be used for transport of americium/curium/lanthanide oxide materials recovered from high-burnup targets, and a Type A packaging for transport of irradiated actinium targets. These are being developed with consideration for both gamma and neutron shielding for contents reading up to 5000 R/h. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

What's in WIPP? Packaging TRU Waste to Enhance WIPP's Capacity - 16515

Beginning in 1999, and continuing for 15 years, the US DOE Waste Isolation Pilot Plant (WIPP) repository successfully disposed of defense Transuranic (TRU) waste. Six of the originally planned ten disposal panels were filled before two unrelated accidents temporarily halted disposal operations in February 2014. Based on counting the gross volume of all containers emplaced in WIPP to date, a little more than half of the legislated volume capacity has been subscribed (91,000 cubic meters). This paper describes the many different packaging configurations that have been used to ship TRU waste to WIPP, and makes observations that may enhance the repository capacity when shipments resume. The contents of shipments to WIPP are limited by many requirements of the repository waste acceptance criteria. These criteria derive from three primary sources; 1) the Certificate of Compliance (CoC), issued by the NRC for Type B shipping packaging used by WIPP; 2) The Hazardous Waste Facility Permit, issued by the New Mexico Environment Department; and 3) the WIPP facility's own Documented Safety Analysis (DSA), developed under DOE's regulation 10CFR830. To meet these criteria, various TRU waste forms have been packaged in different ways, with varying packaging efficiency. Of course, there is an overall upper limit on WIPP's volume capacity that was legislated as part of the WIPP Land Withdrawal Act. The volume counted against this limit has been calculated by simply summing the volumes of individual payload containers shipped to and emplaced at WIPP. This method counts significant non-waste volumes inside the containers due to over-packing inner containers and partially filled containers. Without some change to how disposal volumes are calculated, or how waste is packaged, the amount of TRU waste yet to be shipped to WIPP could exceed the legislated capacity. Concepts to enhance the packaging efficiency of future waste shipments to WIPP are presented. Approximate estimates of the waste versus non-waste volumes that have been emplaced to date are discussed. This paper only considers contact handled TRU waste (i.e., contact dose rates < 2 mSv/hr). (authors)

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

Idaho-Developed Technologies, Processes Offer Unique Solutions to Waste Management Challenges - 19281

INL Site cleanup contractor Fluor Idaho, LLC (Fluor Idaho) and its predecessors have developed a host of state-of-the-art technologies and specialized processes for nuclear waste handling, treatment, storage, and shipping. The cleanup program has also repurposed and upgraded previously idled Cold War-era buildings to treat and repackage both on- and off-site generated radioactive and hazardous wastes. And lastly, Fluor Idaho is using one-of-a-kind equipment to protect waste management personnel as they undertake their work to meet cleanup commitments that the Department of Energy (DOE) made to the state of Idaho. Such applications, although developed for Idaho's wastes, have the potential to benefit other facilities within the DOE Complex. Engineers and fabricators have teamed up to develop truly unique devices made of specialized metals, like stainless steel alloys, and more readily available materials such as Lexan. All waste types that Fluor Idaho manages daily - spent nuclear fuel, both high- and low-level wastes, and both remote-handled (RH) and CH-TRU - benefit from these innovations. Without question, the greatest number of specialized equipment, processes, and worker-protection technologies developed are for the retrieving, characterizing, repackaging, and shipping of TRU wastes, which makes sense considering the INL Site stores the most TRU waste in the DOE Complex. On the CH-TRU retrieval front, crews are using specially modified Gradall excavators to protect workers while they exhume Rocky Flats Plant wastes that were sent to Idaho in the 1950's and 1960's and buried in a shallow landfill. The excavators are supplied with extra Lexan windows to protect operators from the potential for flying debris during exhumation. On-board supplied air, pumped into the cab, and individual air supplies protect workers from inhalation hazards. Also on-board infrared cameras allow operators to detect heat generating from reactive metals at the dig face. At the Advanced Mixed Waste Treatment Facility (AMWTP), Brokk robotic arm operators use a 'super clam' claw, designed and built at the INL Site, on the remotely operated heavy equipment to size-reduce waste boxes, drums, concrete, and metal containing TRU constituents. The super clam is so powerful and dexterous that operators can crush a 55-gallon drum into the size of a basketball, but also delicately turn the pages of a magazine. Boxes too large to be sized at the AMWTP's box-lines are sent to the Accelerated Retrieval Project VII facility to be sized by Waste Management personnel. Such waste boxes, which sometimes contain 'flighty' plutonium-238 material, have forced the project to employ a heavily ventilated and filtered interior tent. Not all technologies developed for the Waste Management program are high-technology; in fact, one piece of equipment is so simple it's literally ingenious. Operators who inspect and repackage RH-TRU inside a hot cell at the Idaho Nuclear Technology and Engineering Center (INTEC) needed a backup plan in case the hot cell port couldn't be closed in the event of a power outage. Engineers and fabricators collaborated to develop a port cover actuator mechanism, which allows operators to open and close the cell's hatch cover with nothing more than a cordless drill. Also at INTEC, Maintenance Shop fabricators developed a Lexan container or 'pig' suitable for shielding neutron-emitting TRU waste sources. Lexan was chosen as the preferred shielded container material for this special type of waste because neutrons are best shielded with materials containing lots of hydrogen, such as plastics, which are composed of carbon and hydrogen molecules. Once approved for shipment, the high-activity waste will be loaded into specialty casks and shipped to WIPP for permanent disposal. Although all these technologies and processes were initially developed for resolving challenges at the INL Site, their applications could be far-reaching. Like the INL Site, many other facilities in the DOE Complex manage, treat, store, and ship TRU, low-and high-level wastes, and spent nuclear fuel. The technologies and processes developed in Idaho protect Waste Management personnel as they undertake their work, but these benefits could be utilized by other facilities and personnel doing similar work. (authors)

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

Environmentally improved CdTe photovoltaic recycling through novel technologies and facility location strategies

Abstract Photovoltaic waste is projected to reach up to 78 million tons across globally dispersed locations through 2050. Current recycling infrastructure is inadequate to process these waste volumes responsibly. This necessitates commercializing novel, environmentally advantageous photovoltaic recycling technologies that improve upon incumbent industrial operations. CdTe photovoltaic recycling is a promising candidate for improvements as 25,000 tons of spent modules are recycled worldwide annually. This paper evaluates the operational performance and compares six novel technologies, the incumbent technology used in industry and one technology which extends the incumbent process across ten environmental impact categories. The tradeoff between incurring a transportation burden (ship or road) to recycle in a large‐scale centralized facility with a higher operational efficiency and avoiding transportation by recycling in a small‐scale decentralized facility with a lower operational efficiency is evaluated. Thermal delamination to eliminate the ethylene vinyl acetate and separate the photovoltaic glass panels is preferable to the incumbent mechanical process across nine environmental impact categories and decreases the climate‐change impact of CdTe photovoltaic recycling by 23%. Bath and probe sonication are ineffective for delamination, and the use of organic solvents is more environmentally burdensome than the incumbent mechanical process. Centralized recycling with shipping is environmentally preferable than with road‐based transportation. For every 100‐km increase in road transportation from the decentralized to the centralized facility, the inventory requirement in the centralized facility should be 6% lower than the decentralized facility for centralized recycling to have a lower climate‐change impact than decentralized recycling. The corresponding value for shipping is 0.4%.

Ravikumar, Dwarakanath↗
DOE OSTI2018

Improved Reanalysis and Prediction of Atmospheric Fields Over the Southern Ocean Using Campaign‐Based Radiosonde Observations

Abstract This study investigated the impact of radiosonde observations from the Southern Ocean obtained by the Australian R/V Aurora Australis on the ALERA2 experimental ensemble reanalysis data set and ensemble forecast experiment. An observing system experiment (OSE) that included additional ship‐launched radiosonde data captured the atmospheric structure over the Southern Ocean. ALERA2 without additional radiosondes had positive temperature biases exceeding 7 °C in the upper troposphere when low‐pressure cyclonic systems passed over the ship. The spread in the upper level was reduced by 15% in the OSE, which propagated downstream from the ship's position because of the sparse observing network over southern high latitudes. Comparison of two 63‐member ensemble forecast experiments initialized by ALERA2 and the OSE revealed that prediction of midlatitude cyclone tracks was improved by the realistic representation of upper‐level troughs in the OSE forecast. This confirms that additional radiosondes over the Southern Ocean reduce uncertainty and error in midlatitude cyclone forecasts.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Impact of Anthropogenic Emission Injection Height Uncertainty on Global Sulfur Dioxide and Aerosol Distribution

Anthropogenic sulfur compounds play an important role in acid deposition, aerosol particle formation, and subsequent radiative forcing and human fine particulate exposure. There are substantial uncertainties in processes influencing sulfate and precursor distributions, however, that have not yet been resolved through comparisons with observations. We find here an underappreciated factor that has a large impact on model results: uncertain emission height. Global aerosol-climate model simulations indicate that the assumed effective anthropogenic emission height is very important to SO 2 near-surface concentrations and vertical profile. The global range of near-surface SO 2 concentration over land (ocean) due to uncertainty in industrial (international shipping) emission injection height is 81% (76%), relative to the average concentration. This sensitivity is much larger than the uncertainty of SO 2 emission rates. Black carbon and primary organic matter concentration and profiles are also sensitive to emission heights (53% over land and 28% over oceans). The impact of emission height uncertainty is larger in winter for land-based emissions, but larger in summer over the Northern Hemisphere ocean for shipping emissions. The variation in aerosol optical depth related to shipping emission injection heights is 11% over oceans, revealing the potential importance of injection height on aerosol forcing and climatic effects. The large impact on SO 2 concentrations can confound attempts to use surface, aircraft, and satellite observations to constrain the importance of other processes that govern sulfur compound distributions in the atmosphere. In conclusion, the influence of emission height on vertical SO 2 column also will impact the accuracy of satellite retrievals.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Comparison of a Full-Scale and a 1:10 Scale Low-Speed Two-Stroke Marine Engine Using Computational Fluid Dynamics

International marine shipping is a growing component of international trade; a vast majority of all the world’s goods are being transported on large ocean-going vessels. The International Maritime Organization (IMO) introduced the Energy Efficiency Design Index in 2013, a regulatory framework of associated metrics for reducing emissions of CO 2 per tonne-mile from shipping by approximately 10% each decade. Therefore, decarbonizing the maritime sector requires the development of new fuel sources. Because of the extremely large physical size of the internal combustion engines present in shipping vessels, experimental iterative development of the engine and fuel system is cost-prohibitive. Thus, the ability to perform combustion system development in a scaled platform that can be more easily operated and modeled computationally is of interest. To that end, scaling relationships are needed to translate the results from a smaller engine to a larger counterpart. Scaling studies to date have been restricted to low scaling ratios, four-stroke light-duty engines, and under-resolved computational fluid dynamic simulations that likely do not accurately capture the physics of scaling. In this work, computational models of a 1:10 scale and a full-scale two-stroke crosshead low-speed marine engine were created and validated against experiments obtained in a real 1:10 scale engine installed at Oak Ridge National Laboratory. Further, due to the large size of the full-scale engine, the model required large high-performance computing resources to be evaluated. The availability of high-performance computing resources at the Department of Energy’s Leadership Computing Facilities is an enabler of the current work. The results of the small- and large-scale engine simulations were compared to analyze the effectiveness of the appropriate scaling laws under these extreme scaling ratio conditions.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2015

SU-E-T-675: Remote Dosimetry with a Novel PRESAGE Formulation

Purpose: 3D-gel dosimetry provides high-resolution treatment validation; however, scanners aren’t widely available. In remote dosimetry, dosimeters are shipped out from a central base institution to a remote site for irradiation, then shipped back for scanning and analysis, affording a convenient service for treatment validation to institutions lacking the necessary equipment and resources. Previous works demonstrated the high-resolution performance and temporal stability of PRESAGE. Here the newest formulation is investigated for remote dosimetry use. Methods: A new formulation of PRESAGE was created with the aim of improved color stability post irradiation. Dose sensitivity was determined by irradiating cuvettes on a Varian Linac (6MV) from 0–15Gy and measuring change in optical density at 633nm. Sensitivity readings were tracked over time in a temperature control study to determine long-term stability. A large volume study was performed to evaluate the accuracy for remote dosimetry. A 1kg dosimeter was pre-scanned, irradiated on-site with an 8Gy 4field box treatment, post-scanned and shipped to Princess Margaret Hospital for remote reading on an identical scanner. Results: Dose sensitivities ranged from 0.0194–0.0295 ΔOD/(Gy*cm)—similar to previous formulations. Post-irradiated cuvettes stored at 10°C retained 100% initial sensitivity over 5 days and 98.6% over 10 weeks while cuvettes stored at room temperature fell to 95.8% after 5 days and 37.4% after 10 weeks. The immediate and 5-day scans of the 4field box dosimeter data was reconstructed, registered to the corresponding eclipse dose-distribution, and compared with analytical tools in CERR. Immediate and 5-day scans looked visually similar. Line profiles revealed close agreement aside from a slight elevation in dose at the edge in the 5-day readout. Conclusion: The remote dosimetry formulation exhibits excellent temporal stability in small volumes. While immediate and 5-day readout scans of large volume dosimeters show promising agreement, further development is required to reduce an apparent time dependent edge elevation.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2018

Franklin expedition lead exposure: New insights from high resolution confocal x-ray fluorescence imaging of skeletal microstructure

In the summer of 1845, under the command of Sir John Franklin, 128 officers and men aboard Royal Navy ships HMS Erebus and HMS Terror sailed into Lancaster Sound and entered the waters of Arctic North America. The goal of this expedition was to complete the discovery of a northwest passage by navigating the uncharted area between Barrow Strait and Simpson Strait. Franklin and his crew spent the first winter at Beechey Island, where three crewmen died and were buried. In September 1846, the ships became stranded in ice off the northwest coast of King William Island, where they remained until April 1848. At that time, the crew, reduced to 105, deserted the ships and retreated south along the island’s western and southern shores in a desperate attempt to reach the mainland and via the Back River, to obtain aid at a Hudson’s Bay Company Post. Sadly, not one individual survived. Previous analyses of bone, hair, and soft tissue samples from expedition remains found that crewmembers’ tissues contained elevated lead (Pb) levels, suggesting that Pb poisoning may have contributed to their demise; however, questions remain regarding the timing and degree of exposure and, ultimately, the extent to which the crewmembers may have been impacted. To address this historical question, we investigated three hypotheses. First, if elevated Pb exposure was experienced by the crew during the expedition, we hypothesized that those sailors who survived longer (King William Island vs. Beechey Island) would exhibit more extensive uptake of Pb in their bones and vice versa. Second, we hypothesized that Pb would be elevated in bone microstructural features forming at or near the time of death compared with older tissue. Finally, if Pb exposure played a significant role in the failure of the expedition we hypothesized that bone samples would exhibit evidence of higher and more sustained uptake of Pb than that of a contemporary comparator naval population from the 19 th century. To test these hypotheses, we analyzed bone and dental remains of crew members and compared them against samples derived from the Royal Navy cemetery in Antigua. Synchrotron-based high resolution confocal X-ray fluorescence imaging was employed to visualize Pb distribution within bone and tooth microstructures at the micro scale. The data did not support our first hypothesis as Pb distribution within the samples from the two different sites was similar. Evidence of Pb within skeletal microstructural features formed near the time of death lent support to our second hypothesis but consistent evidence of a marked elevation in Pb levels was lacking. Finally, the comparative analysis with the Antigua samples did not support the hypothesis that the Franklin sailors were exposed to an unusually high level of Pb for the time period. Taken all together our skeletal microstructural results do not support the conclusion that Pb played a pivotal role in the loss of Franklin and his crew.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2014

AGC-2 Disassembly Report

The Next Generation Nuclear Plant (NGNP) Graphite Research and Development (R&D) Program is currently measuring irradiated material properties for predicting the behavior and operating performance of new nuclear graphite grades available for use within the cores of new very high temperature reactor designs. The Advanced Graphite Creep (AGC) experiment, consisting of six irradiation capsules, will generate irradiated graphite performance data for NGNP reactor operating conditions. The AGC experiment is designed to determine the changes to specific material properties such as thermal diffusivity, thermal expansion, elastic modulus, mechanical strength, irradiation induced dimensional change rate, and irradiation creep for a wide variety of nuclear grade graphite types over a range of high temperature, and moderate doses. A series of six capsules containing graphite test specimens will be used to expose graphite test samples to a dose range from 1 to 7 dpa at three different temperatures (600, 900, and 1200°C) as described in the Graphite Technology Development Plan. Since irradiation induced creep within graphite components is considered critical to determining the operational life of the graphite core, some of the samples will also be exposed to an applied load to determine the creep rate for each graphite type under both temperature and neutron flux. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR). AGC-1 and AGC-2 will be irradiated in the south flux trap and AGC-3–AGC-6 will be irradiated in the east flux trap. The change in flux traps is due to NGNP irradiation priorities requiring the AGC experiment to be moved to accommodate Fuel irradiation experiments. After irradiation, all six AGC capsules will be cooled in the ATR Canal, sized for shipment, and shipped to the Materials and Fuels Complex (MFC) where the capsule will be disassembled in the Hot Fuel Examination Facility (HFEF). During disassembly, the metallic capsule will be machined open and the individual samples removed from the interior graphite body containing the samples. Samples removed from the capsule will be loaded in a shipping drum and shipped to the Idaho National Laboratory (INL) Research Center (IRC) for initial post-irradiation examination (PIE) and storage for any future testing at the newly completed Carbon Characterization Laboratory (CCL). All work was performed under an ASME NQA-1-2008;1a-2009 compliant quality assurance program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Used Fuel Testing Transportation Model

This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Temperature Environment for 9975 Packages Stored in KAC

Plutonium materials are stored in the K Area Complex (KAC) in shipping packages, typically the 9975 shipping package. In order to estimate realistic degradation rates for components within the shipping package (i.e. the fiberboard overpack and O-ring seals), it is necessary to understand actual facility temperatures, which can vary daily and seasonally. Relevant facility temperature data available from several periods throughout its operating history have been reviewed. The annual average temperature within the Crane Maintenance Area has ranged from approximately 70 to 74 °F, although there is significant seasonal variation and lesser variation among different locations within the facility. The long-term average degradation rate for 9975 package components is very close to that expected if the component were to remain continually at the annual average temperature. This result remains valid for a wide range of activation energies (which describes the variation in degradation rate as the temperature changes), if the activation energy remains constant over the seasonal range of component temperatures. It is recommended that component degradation analyses and service life estimates incorporate these results. Specifically, it is proposed that future analyses assume an average facility ambient air temperature of 94 °F. This value is bounding for all packages, and includes margin for several factors such as increased temperatures within the storage arrays, the addition of more packages in the future, and future operational changes.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Results of Hg speciation testing on tanks 30, 32, and 37 depth samples

The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team. The twelfth shipment of samples was designated to include 3H evaporator system Tank 30, 32, and 37 depth samples. The Tank 30 depth sample (HTF-30-15-70) was taken at 190 inches from the tank bottom and the Tank 32 depth sample (HTF-32-15-68) was taken at 89 inches from the tank bottom and both were shipped to SRNL on June 29, 2015 in an 80 mL stainless steel dip bottles. The Tank 37 surface sample (HTF-37-15-94) was taken around 253.4 inches from the tank bottom and shipped to SRNL on July 21, 2015 in an 80 mL stainless steel dip bottle. All samples were placed in the SRNL Shielded Cells and left unopened until intermediate dilutions were made on July 24, 2015 using 1.00 mL of sample diluted to 100.00 mL with deionized H 2 O. A 30 mL Teflon® bottle was rinsed twice with the diluted tank sample and then filled leaving as little headspace as possible. It was immediately removed from the Shielded Cells and transferred to refrigerated storage where it remained at 4 °C until final dilutions were made on October 20. A second portion of the cells diluted tank sample was poured into a shielded polyethylene bottle and transferred to Analytical Development for radiochemical analysis data needed for Hazardous Material Transportation calculations.

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

Hazardous Material Packaging and Transportation

This is a student training course. Some course objectives are to: recognize and use standard international and US customary units to describe activities and exposure rates associated with radioactive material; determine whether a quantity of a single radionuclide meets the definition of a class 7 (radioactive) material; determine, for a given single radionuclide, the shipping quantity activity limits per 49 Code of Federal Regulations (CFR) 173.435; determine the appropriate radioactive material hazard class proper shipping name for a given material; determine when a single radionuclide meets the DOT definition of a hazardous substance; determine the appropriate packaging required for a given radioactive material; identify the markings to be placed on a package of radioactive material; determine the label(s) to apply to a given radioactive material package; identify the entry requirements for radioactive material labels; determine the proper placement for radioactive material label(s); identify the shipping paper entry requirements for radioactive material; select the appropriate placards for a given radioactive material shipment or vehicle load; and identify allowable transport limits and unacceptable transport conditions for radioactive material.

07 ISOTOPE AND RADIATION SOURCES↗