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DOE OSTI2021

High temperature mechanical properties of fluorite crystal structured materials (CeO 2 , ThO 2 , and UO 2 ) and advanced accident tolerant fuels (U 3 Si 2 , UN, and UB 2 )

The mechanical interaction between the fuel and cladding that occurs during operation of a nuclear reactor is important to understand as it can lead to cladding failures and release of radioactive material into the coolant. Additionally, in order to develop better models of the pellet-clad mechanical interactions, the mechanical properties of the fuel at relevant operating temperatures, like the elastic moduli, are needed for current and advanced accident tolerant fuels (ATFs). In this work, elevated temperature nanoindentation and resonant ultrasound spectroscopy were used to measure the moduli and hardness of several fluorite materials (CeO 2 , ThO 2 , UO 2 ) and several ATF candidates (ATF) (U 3 Si 2 , UN, UB 2 ). In addition, a comparison of the two techniques was performed in this study to independently validate the mechanical properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

MPF X-Ray Module Setup for 2018

The Mobile Plutonium Facility (MPF) is a portable module design facility that can be deployed anywhere in the world for the safe recovery, packaging, and shipment of a nations legacy, or excess plutonium stockpile. All of the modules of MPF have specific purposes for operating in a mission if one were to ever come up. For my past four internships here at SRNL I have been working on the support of the X-Ray module for MPF. This year I installed new equipment I had purchased last year for the X-Ray Module and made sure the module was mission ready for the chance that the team deploys in the future for plutonium recovery. Using digital radiography (DR) and computed tomography (CT) provides the MPF team the ability to verify the contents of containers that would be holding material. This also ensures the safety of the team members in the glovebox facilities by verifying that there are no sharps, liquids, or other items that may harm the workers or glovebox operations when working with radioactive material. X-Ray imaging can also pick up low density objects like plastic bags and tape. Note how the DR image hides the puncture in the inner can while the CT image shows the damage. New DR and CT computers with upgraded software, New imaging panel, NOVO Portable X-Ray System. Because of the new upgrades to the X-Ray module system containers can be scanned and evaluated quicker than before. With the last system the 3D model reconstruction alone would take anywhere from 10-12 minutes, now with the new system that time has been brought down to just shy of 1 minute 30 seconds. The new portable system also can 1,2 or 4 images of an object and stitch them together into a mosaic for scanning of larger containers.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

CONSTRAINTS ON THE ORIGIN OF THE FIRST LIGHT FROM SN 2014J

We study the very early light curve of supernova 2014J (SN 2014J) using the high-cadence broad-band imaging data obtained by the Kilodegree Extremely Little Telescope, which fortuitously observed M 82 around the time of the explosion, starting more than 2 months prior to detection, with up to 20 observations per night. These observations are complemented by observations in two narrow-band filters used in an Hα survey of nearby galaxies by the intermediate Palomar Transient Factory that also captured the first days of the brightening of the supernova. The evolution of the light curves is consistent with the expected signal from the cooling of shock heated material of large scale dimensions, ≳1R {sub ☉}. This could be due to heated material of the progenitor, a companion star or pre-existing circumstellar environment, e.g., in the form of an accretion disk. Structure seen in the light curves during the first days after explosion could also originate from radioactive material in the outer parts of an exploding white dwarf, as suggested from the early detection of gamma-rays. The model degeneracy translates into a systematic uncertainty of ±0.3 days on the estimate of the first light from SN 2014J.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2018

Vitrification of Hanford Tank Waste 241-AP-105 in a Continuous Laboratory-Scale Melter

Low-activity waste (LAW) stored in underground tanks on the Hanford site in Washington State is planned to be filtered for solids removal and processed through ion exchange columns for cesium removal at the Low-Activity Waste Pretreatment System facility. These pretreatment steps will allow the waste to be transferred to the Hanford Tank Waste Treatment and Immobilization Plant LAW facility for immobilization into glass. The liquid waste will be combined with glass-forming chemicals (GFCs) to form a waste feed slurry that can be fed to electric melters for vitrification. The process of continuously converting the aqueous feed slurry into a melt is dynamic and includes multiple reactions, degassing, and dissolution processes that depend on heat from the melt below. In this conversion process, waste components are partitioned into one of two streams: glass and off-gas. Processing information and chemical information associated with these waste product streams has been requested by Washington River Protection Solutions for actual tank waste from tank 241-AP-105 (referred to as AP-105). To acquire this type of information, a small-scale melter system was desired that would not require high volumes of input waste or the large resource commitment of a full-scale melter system, while also providing dynamic information that would be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM), has been designed to operate with a continuous feeding process, while periodically pouring glass product and collecting off-gas. The CLSM vessel has been sized to collect the relevant process and chemical information from obtainable volumes of AP-105 waste samples. Two CLSMs were constructed: one in a non-radioactive environment for processing waste simulants and another in a fume hood capable of handling radioactive material for processing actual Hanford tank waste. This dual-system setup allowed for comparison between simulated and real tank waste with essentially identical test equipment. AP-105 waste simulant was first tested in the non-radioactive CLSM to determine the processability and desired operating window for the actual tank waste. Rhenium (Re) was added to the simulant to act as a surrogate for technetium-99 ( 99 Tc), a key waste component found in the actual tank waste. The partitioning of Re into the glass and off-gas products during the continuous feeding of simulant was determined, which allowed the Re retention, the mass flow rate of Re out of the CLSM via the glass product with respect to the mass flow rate of Re into the CLSM via the feed slurry, to be calculated. Key processing results and the average Re retention, from the CLSM test run with AP-105 simulant feed slurry are shown in Table ES.1.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Management of Documents and Information on Radioactive Waste - 19455

The assurance of the safety of a repository for low- and intermediate-level radioactive waste, during the 300 years of institutional control after closure, depends on assuming that the responsible institution will be capable to control the documents with information regarding the facility, the disposed of waste and the established measures for counteracting events that could breach the isolation of the radioactive material. Management of documents or information does not represent a major challenge in itself in view of consolidated practices and international standards, such as ISO 15489. However, considering the time scale involved in the disposal of the radioactive wastes, efforts should be made to ensure the preservation and proper access to documents and information. In this matter, the technological aspect of document production can be a complicating factor. In fact, it is common to be unable to access electronic documents produced 20 or 30 years ago, due either to problems related to the media where the information is stored, or to a lack of compatible systems for reading the files. This is just one example of complication when thinking about production, organization, representation, preservation and retrieval of recorded information that is to last for centuries. In this way, this work seeks to present the situation of the management of documents and information on radioactive waste in the scope of the Institute of Energy and Nuclear Research (IPEN) and with national regulation in Brazil. (authors)

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DOE OSTI2019

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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DOE OSTI2022

Recommendations for Secure Transport: Material Conveyance Physical Protection Technology

Nuclear material is at higher risk of theft and sabotage during transport than during any other phase of the nuclear fuel lifecycle. Nuclear materials are transported in the public domain where adversaries have an upper hand by taking advantage of the time and location of the theft or sabotage attempt. As such, even modest threat profiles for transport of nuclear and radioactive material can require substantial detection and delay measures to support timely response. Conveyance tracking augmented with technology that improves on-the-scene situational awareness at a remote monitoring center has been adopted as a de-facto standard approach for transportation security. At present, the extended tracking and situational awareness capabilities needed for a nuclear material shipment, as is provided by the purpose designed, Transportation – Security, Tracking and Reporting (T-STAR) System, do not exist in a single commercial off-the-shelf (COTS) solution. Typically, the COTS systems that excel in one area are deficient in other areas, presenting challenges to designing well-rounded, robust systems. Still, COTS solutions can offer the basic set of tracking and situational awareness capabilities by indicating last update time, current location, and route taken. By incorporating vehicle and driver performance measures via the vehicle’s controller area network (CAN bus) and video alongside other data streams allows telemetry and other shipment information to be assessed in novel ways.This paper describes the operation, capabilities and features of various conveyance protection systems and approaches, assesses emerging technologies and how they could be used through a unified interface, and enumerates additional ways to provide early detection using vehicle, onboard technologies, effective delay technologies and approaches and simple equipment to improve protection during transport.

Shannon, Michael↗

DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

This study focuses on requirements and challenges pertaining to advanced transportable nuclear energy systems, which are currently in development and feature varying degrees of mobility. These reactors mostly use high-assay low-enriched uranium (HALEU) fuel, and only limited refueling is required during their lifecycles. Due to the smaller source terms, higher burnup fuel, and potentially higher thermal efficiency involved, the waste inventory will be less than that generated by conventional light-water reactors. However, the relevant regulatory requirements must be met when transporting HALEU fuel (whether fresh or used) inside the reactor vessel—mainly the fuel material transportation, siting, and physical protection requirements outlined in US Nuclear Regulatory Commission (NRC) regulations 10 CFR 71 and 73. Spent nuclear fuel (SNF), as well as irradiated fuel (used during any interval of reactor operation), contains highly radioactive materials that generate heat and necessitate special precautions during transportation. This study addresses the challenges that this incurs, and outlines potential solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

A Novel and Cost Effective Approach to the Decommissioning and Decontamination of Legacy Glove Boxes - Minimizing TRU Waste and Maximizing LLW Waste - 13634

This paper describes the process of decommissioning two gloveboxes at the Argonne National Laboratory (ANL) that were employed for work with plutonium and other radioactive materials. The decommissioning process involved an initial phase of clearing tools and materials from the glove boxes and disconnecting them from the laboratory infrastructure. The removed materials, assessed as Transuranic (TRU) waste, were packaged into 55 gallon (200 litre) drums and prepared for ultimate disposal at the Waste Isolation Pilot Plant (WIPP) at Carlsbad New Mexico. The boxes were then sampled to determine the radioactive contents by means of smears that were counted with alpha and beta detectors to determine the residual surface contamination, especially in terms of alpha particle emitters that are an indicator of TRU activity. Paint chip samples were also collected and sent for laboratory analysis in order to ascertain the radioactive contamination contributing to the TRU activity as a fixed contamination. The investigations predicted that it may be feasible to reduce the residual surface contamination and render the glovebox structure low level waste (LLW) for disposal. In order to reduce the TRU activity a comprehensive decontamination process was initiated using chemical compounds that are particularly effective for lifting and dissolving radionuclides that adhere to the inner surfaces of the gloveboxes. The result of the decontamination process was a reduction in the TRU surface activity on the inner surfaces of the gloveboxes by four orders of magnitude in terms of disintegrations per unit area (DPA). The next phase of the process involved a comprehensive assay of the gloveboxes using a combination of passive neutron and gamma ray scintillation detectors and a shielded and collimated high purity Germanium (HPGe) gamma ray detector. The HPGe detector was used to obtain gamma ray spectra for a variety of measurement positions within the glovebox. The spectra were used to determine the TRU content of the boxes by assessing the activity of Am-241 (59 keV) and Pu-241 (414 keV). Using the data generated it was possible for qualified subject matter experts (SME) to assess that the gloveboxes could be consigned for disposition as LLW and not as TRU. Once this determination was assessed and accepted the gloveboxes were prepared for final disposition to the Nevada National Security Site (NNSS) - formerly the Nevada Test Site (NTS). This preparation involved fixing any remaining radioactive contamination within the gloveboxes by filling them with a foam compound, prior to transportation. Once the remaining contamination was fixed the gloveboxes were removed from the laboratory and prepared for transported by road to NNSS. This successful glovebox decontamination and decommissioning process illustrates the means by which TRU waste generation has been minimized, LLW generation has been maximized, and risk has been effectively managed. The process minimizes the volume of TRU waste and reduced the decommissioning time with significant cost savings as the result. (authors)

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DOE OSTI2012

Data Quality Objectives Supporting Radiological Air Emissions Monitoring for the PNNL Site

Pacific Northwest National Laboratory (PNNL) is in the process of developing a radiological air monitoring program for the PNNL Site that is distinct from that of the nearby Hanford Site. The original DQO (PNNL-19427) considered radiological emissions at the PNNL Site from Physical Sciences Facility (PSF) major emissions units. This first revision considers PNNL Site changes subsequent to the implementation of the original DQO. A team was established to determine how the PNNL Site changes would continue to meet federal regulations and address guidelines developed to monitor air emissions and estimate offsite impacts of radioactive material operations.

Compliance↗

DOE OSTI2025

Microscopic Measurement: Need for Nuclear Material Sample Specimen Characterization Capability

The Surveillance, Fabrication, and Off-Site Operations team at Lawrence Livermore National Laboratory is beginning an update to the nuclear material machining lab in Superblock. Plutonium and other nuclear materials tend to accelerate the aging process for equipment and machines that would otherwise be maintainable for long lives. They have their lifespans shortened from deteriorated seals, embrittled polymer components, accelerated corrosion, and damage to electronics from exposure to radioactive materials. The cost of installation of equipment into a glove box (GB) tends to outstrip the cost of the equipment itself. Due to the magnification of cost, the historical approach to equipment installs has been to purchase and install the highest-quality and most robust equipment available to maximize time between failures and updates. To make room for the installation of a new lathe, the inspection GB is being moved to a different location within the machining lab. As part of the overall effort, the equipment within the inspection GB is being revisited. Current measurement capabilities are antiquated and starting to become problematic due to age. Advances in metrology are sought for more comprehensive characterization of scientific samples. This document aims to identify several options for the replacement of current measurement equipment and identify a system for purchase and eventual installation. The intention is not to select the cheapest system, but rather to identify a system that best meets the needs of both the operators and principal investigators (PI). Following the convention of the SYSM-5620 Design Thinking and System Engineering, it was determined that a laser confocal measurement system from Keyence is the most suitable option for our programmatic need. Form factor, capability, and maintainability were considered. Along with stakeholder input, to ensure the best decision for the Nuclear Material Technology Program.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Fracture Toughness Evaluation for Spent Nuclear Fuel Clad Systems Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials. These materials are often held in long-term storage as an interim stage within their lifecycle, including reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations and follow-on SNF transportations to its final repository sites are very important, due to a great concentration of fission products, actinides and activation products. Fracture mechanics approach in applying to SNF system reliability investigation, especially for the high burn up (HBU) SNF, during SNF long term dry storage or SNF transportation is necessary due to the inherited flaws and inhomogeneity structures existed in a SNF system from the reactor operation. For instance, such as hydride and oxide formation, surface flaw induced by SNF assembly contact interactions, or internal flaws induced by pellet-clad mechanical interaction (PCMI). However, none of the existing fracture toughness data deal with fuel cladding specific geometry or SNF material conditions, such as cladding structure with the segment fuel pellets configuration and the PCMI mechanism. Consequently, the application of existing fracture toughness data to SNF system reliability investigation can be error prone; thus, the development of an industry-wide consensus fracture mechanics approach is needed. The objective of this research is to develop an in-situ fracture mechanics testing protocol, including the associated analytical procedure, that is suitable for evaluating the SNF fracture toughness. Fracture toughness data was obtained under quasi-static fracture loading using Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support SNF reliability investigation during long-tern SNF dry storage stage or in the follow-on SNF system transport to the final repository sites. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre cracks are needed for brittle materials specimens. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient pre-crack length to increase the sample localized constraint at crack front. Fatigue pre-crack growth techniques with the associated compliance function estimated was developed for estimating the crack penetration depth to monitor the fatigue crack growth evolution. In order to extend SNTT approach to a thin shell cladding structure material, a new testing protocol and the associated analytical procedure for estimating the SNF fracture toughness was developed. The detailed SNTT approach and its estimated fracture toughness for the Zr-4 cladding structure with segment alumina inserts are presented in this report. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to Zr-4 cladding material. For the medium and the short crack length, the estimate J Q upon fracture for the baseline Zr-4 cladding is at 285.7 lb/in with 2-sigma uncertainty of 18.59 lb/in., and the associated K Q is at 61.4 Ksi√in. For a long crack length, the crack initialization orientation is apparently deviated from the principle tensile stress contour – 45° spiral crack front (i.e., Mode I + Mode III failure mechanism); and the estimated J Q is at 108 lb/in., the associated K Q is at 37.7 Ksi√in. The estimate Mode I J Q (along the 45° pitch principle tensile stress contour) is at 200 lb/in.; which indicate a significant reduction in the fracture toughness in a tubing structure under mixed mode loading condition, Mode I + Mode III, compared to that of Mode I (tensile stress dominated) alone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

H-Canyon Jumper Gasket Sealing Evaluation

H-Canyon Documented Safety Analysis approved December 2019 credits the leak integrity of H-Canyon (HCA) process tanks, piping, and jumpers in preventing the release of radioactive material in a post seismic environment. Demonstration of leak tightness capability for this equipment was achieved through substantial analytical effort. However, one potential weak point that affects leak integrity exists at the gasketed joint between Hanford Connectors (HC) and nozzles. Review of the HCA Crane Log over a three-year period indicates the HCs periodically leak. Savannah River National Laboratory (SRNL) was requested to perform gasket compression tests that evaluate gasket relaxation characteristics, i.e., creep, and provide improvements to the HC installation process that can reduce or eliminate the frequency of leaks at HCs. A HC torque value of 312 ft-lbf is the target value used because it is the minimum torque value, with 95% confidence, when an unlubricated HC is impacted for standard sealing period of 9 seconds. In the HCA pipe-jumper connections are installed remotely by using HCs. HCs inherently add to the uncertainty of a gasketed joint because a calibrated torque value cannot be applied. The gasket material is a special material made of asbestos-fiber impregnated with Teflon. This material has been used for decades and generally performs well but does leak intermittently as gaskets creep and preloaded torque is lost. Gasket types are either snap-ring (SR) or lampshade (LS). Both gasket types cover the same seating area. The SR gasket is designed to remain with the jumper when the jumper is removed leaving the nozzle faces clean of any debris. SR gaskets are initially installed when the jumper is fabricated. Changing gaskets is time consuming and the radioactive environment presents risks to personnel. When jumpers are removed LS gaskets could be used to better seal the connection to reduce down time and ALARA concerns. That is, an LS gasket would be remotely positioned over the nozzle and the HC reinstalled. Testing also included two situations not typical of plant operation. One situation was the use of the non-standard double-layered gaskets (DG), and the other situation was the use of a LS without removal of the installed SR gasket. The DG was used when the supply of LS gaskets ended, so that testing could continue. It was of interest to demonstrate this gasket’s creep response because the DG, which was two gaskets glued together, was thicker than the LS gasket and the thicker material was expected to elicit a greater creep characteristics. The second situation of using two gaskets together by installing an LS gasket without removing the installed gasket is not allowed by the ASME Codes. However, by demonstrating the creep history of such a gasket arrangement it could be considered as an option that could possibly seal a connection while significantly reducing operational down time and personnel risk. Forty tests were conducted, and the overall results demonstrate the initial applied force and how the gasket changes with time. The change in sealing force, in terms of torque, was then measured to obtain a force history as the gaskets creep. This report documents the results of those tests as well as the method used to perform the tests.

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DOE OSTI2023

Update of Receptor Locations for INL National Emissions Standard for Hazardous Air Pollutants (NESHAP) Assessments

This report documents the selection of updated public receptor locations for use in conducting regulatory assessments of radioactive air emissions for Idaho National Laboratory (INL) facilities. The basis for this receptor location analysis is found in 40 CFR 61 Subpart H, “National Emissions Standards for Emissions of Radionuclide Other than Radon from Department of Energy Facilities,” where the requirements for identifying National Emission Standards for Hazardous Air Pollutants (NESHAP) receptor locations are outlined. Receptor locations (residences, schools, businesses, or farm operations) were identified using high-resolution satellite imagery. This assessment identified 31 unique receptor locations for all INL Site facilities, and 30 receptor locations for INL in-town facilities that have the potential to emit radioactive materials. Identification of receptors for each in-town facility is new to this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Chemistry Envelope for Pitting and Stress Corrosion Cracking Mitigation

The present waste chemistry controls for the Hanford double-shell tanks (DSTs) were established in the 1980s in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits were coupled with corrosion testing results from Pacific Northwest Laboratory (PNL) and adopted for corrosion control of the DST waste at the Hanford Site. Since that time, the temperatures of many DST wastes have decreased to less than 50 °C as the radioactive material has decayed, the testing program in Argentina demonstrated that chloride ion in Hanford Site waste could cause pitting, and other work showed that dilute solutions could cause pitting at the liquid air interface (LAI). Moreover, future transfers from the single shell tanks and the waste treatment plant and process changes may shift the DST waste chemistries to a broader range of compositions with higher aggressive anion concentrations than the wastes that are presently stored. The decreasing temperatures led to the formulation of new chemistry controls for the minimization of SCC risks for operations below 50 °C in 2010, and these controls were partially adopted. The Tank Integrity Expert Panel (TIEP) Corrosion Subgroup (CSG) recommended that additional testing be performed to assess the potential for pitting and LAI corrosion. Since 2015, DNV-GL, SRNL, and the Hanford Site laboratory have investigated pitting and LAI corrosion at current and anticipated DST chemistries and Savannah River National Laboratory (SRNL) undertook a statistically based investigation of the role of nitrate and halide ion induced pitting corrosion. The objective was to develop a comprehensive waste chemistry envelope for the simultaneous minimization of the pitting and SCC risks caused by halide and nitrate ions. This report summarizes the technical basis for the new waste chemistry envelope to minimize the potential for both pitting and cracking corrosion induced by nitrate and halide ions, and the strategies used for the validation of the new controls.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Data Analysis and Detector Troubleshooting for the Silicon Muon Scanner

Muon tomography is the calculation of the density distribution of a material from the probable angles of muon Coulomb scattering in a sampled object. This allows the probing of internal object structures that cannot otherwise be observed. It has many important applications for both national security and other fields, such as archaeology and geology. Its uses range from detecting radioactive material intended for nuclear weaponry to searching for hidden chambers in the Egyptian pyramids. In this project we intend to demonstrate the capability of a tomographic system which uses cosmic rays as the muon source and tracks each muon's initial and final trajectories to ascertain information about their scattering pattern. This method can be used to determine the structure or composition of an object, such as testing to see if a nuclear fuel cask is empty or full. We intend this as an early R&D exploration into the practicality of implementing silicon microstrip sensors in border scans for radioactive contraband, as an alternative to existing drift tube technology.

43 PARTICLE ACCELERATORS↗

DOE OSTI2016

Developments that are making the Versa-Pac Indeed More Versatile (and More Useful to Industry) - 16267

Daher-TLI acquired the assets and employees of Century Industries, a radioactive material packaging manufacturing facility, in 2013. Included in that acquisition was the Versa-Pac Shipping Package. The Versa-Pac is an NRC licensed Type AF packaging (Certificate of Compliance no. USA/9342/AF-96) and was originally designed as a replacement for standard 7A drum-type transportation packaging. It can be used directly or in conjunction with pails, drums or inserts that can be manufactured to accommodate a variety of smaller containers and vessels. The Versa-Pac comes in two sizes, the VP-55 (i.e., 55-gallon drum version) and VP-110 (i.e., 110-gallon drum version). It features a patented design concept in combination with the drum exterior to provide enhanced structural protection to payloads under Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC), per 10CFR71.71 and 10CFR71.73 respectively. Since its introduction in 2010, the Versa-Pac has been utilized mainly as a one-time use package for disposing of radioactive waste. The package is also designed as a reusable package for transporting front-end materials. Discussions with current and prospective clients have resulted in numerous improvements, accomplished and planned, to the package's payload capability through additional engineering analyses and actual design modifications. Revision 10 to the NRC Certificate of Compliance, dated August 6, 2015, introduces an enrichment-loading table that allows shipping larger quantities of U-235 for enrichments less than 100 wt.%. It also authorizes shipment of natural thorium. Other planned and proposed improvements include provisions to transport UF{sub 6} in 1S, 2S, and 5B cylinders, design changes that further increases the amount of U-235 at enrichments less than 100 wt.%, increasing the capacity of the VP-110, developing a lighter 30-gallon version, the VP-30, and introducing content limits specific to uranium-oxides in multiple forms. This paper traces the engineering and design efforts involved in expanding the Versa-Pac's versatility. (authors)

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