INMM

Packaging, Transportation & Disposition Abstracts

Abstract #113
Abstract #169
Abstract #180
Abstract #181
Abstract #201
Abstract #217
Abstract #345
Abstract #384


Tuesday, July 24

 

Concurrent Session III

Abstract #217 (2:40 p.m.)


Decision Methodology for Final Disposal of Radioactive Sources

Randall L. Beatty1, Jose Miguel Roncero-Martin2
1Oak Ridge National Laboratory, Oak RIdge, TN, USA, 2International Atomic Energy Agency, Vienna, Austria

Radioactive sources are used throughout the world, and their beneficial uses are many. Sources are at the core of many applications in different branches of industry, medicine, agriculture and research. Sources are present in a very wide range of equipment that is used for, inter alia, cancer treatment, killing bacteria in food, sterilizing medical supplies, measuring instruments such as gauges used to measure soil moisture and soil density and many other components, irradiating seeds for enhancing food production, protecting buildings from lightning strikes, mapping underground sources of water, prospecting for oil and gas reserves, measuring density of soil for construction projects, or even detecting smoke. At their “cradle” and when in use, radioactive sources are usually properly managed and controlled. It is when a source has reached the end of its useful life when they are at a higher risk of being under inadequate control, poorly managed, or even becoming orphan. Sources that have reached the end of their useful life must be carefully and safely managed. The “grave” of disused sources needs a proper disposal process to prevent them from posing a potential threat to people or the environment. To protect the public and the environment from the potential hazards of ionizing radiation, and to prevent disused sources from becoming orphan, a “cradle-to-grave” control of radioactive sources is essential, as promoted by the Code of Conduct on the Safety and Security of Radioactive Sources. This approach requires a national policy and strategy, an adequate legal and regulatory framework, and adequate resources and infrastructure. Many IAEAMember States have expressed the need for effective solutions to this challenge. There is also a gap related to the life cycle costs of potential disposal practices. This paper proposes a multi-attribute decision making methodology to compare disposal options using both economic and non-economic criteria leading to a preferred option for disposal and a strategy for final disposal of disused radioactive sources that is specific and optimized for each IAEA Member State.

Abstract #169 (3:00 p.m.)


Design and Structural Analysis of a Compact Type B Packaging for Disused Radiological Sources

Zenghu Han1Yung Y. Liu2, James Shuler3
1Argonne National Laboratory, Argonne, IL, USA, 2Argonne National Laboratory, Lemont, IL, USA, 3Department of Energy, Washington D.C., DC, USA

Radiological sources are common in many countries because of beneficial uses in medical and industrial applications. Some of the radioisotopes used in these applications have relatively short half-lives—e.g. 73.8 days for Ir-192 and 5.27 years for Co-60—while others have much longer half-lives—e.g. 30.17 and 28.79 years, respectively, for Cs-137 and Sr-90. These radioisotopes are high-energy β-ɣ emitters, and the lack of a disposition pathway for the disused radiological sources poses a significant risk in terms of inadvertent or deliberate misuse of the material and other problems. The U.S. Department of Energy has planned since the mid-1980s to dispose of all high-level radioactive waste (HLW) and spent nuclear fuel (SNF), regardless of commercial, defense, or research origin, in a common mined geologic repository. A separate mined repository was proposed in 2015 for DOE-managed SNF and HLW, as well as an option for deep borehole disposal of “small” waste forms, such as the Cs and Sr capsules currently stored in pool cells at Hanford’s Waste Encapsulation Storage Facility. This paper presents the design and structural analysis of a compact Type B packaging that can be used for transportation and storage of the disused radiological sources, as well as for direct disposal at a mined geological repository, or a deep borehole, without repackaging. Structural analysis was performed by using finite element code ABAQUS. The results showed that the structural performance of the packaging meets the regulatory requirements under both normal conditions of transport (NCT) and hypothetical accident conditions (HAC), as prescribed in the U.S. federal regulations 10 CFR 71 Packaging and Transportation of Radioactive Material. The all stainless-steel structure materials (i.e., 304/304L stainless steels for the packaging components) and the packaging design also provide adequate heat dissipation and radiation shielding for the disused radiological sources, as well as excellent long-term performance against general corrosion and stress corrosion cracking during extended dry storage. Moreover, the packaging design is suitable for subsequent transportation and direct disposal, without repackaging. Finally, the compact Type B packaging design enables optional use of the ARG-US remote monitoring system, which enhances safety and security during extended storage, transportation, and disposal.

Abstract #113 (3:20 p.m.)



Removal and Decommissioning of Disused Gamma Irradiator in Minsk, Belarus

Pavel Mikhalevich1, Stephen Mladineo2, Vladimir Bogdanov1, Aliaksandr Talai1, Gary Stubblefield3, Edward Godfrey4
1CJSC "Isotope Technologies", Minsk, Belarus, 2PNNL, Richland, WA, USA, 3Vantage Systems, Inc., Missoula, MT, USA, 4Mission Support and Test Services, LLC, Las Vegas, NV, USA

Under the scope of The Department of Energy’s Office of Radiological Security (ORS) Program the Belarusian company, “Isotope Technologies” (IT), in concert with assistance of “Radii” (a company based in Moscow, Russia) dismantled a disused gamma irradiator RKhM-γ-20. This project required innovation, flexibility, and prior experience. The team successfully removed the radioactive sources from the site and moved them to the local radioactive waste facility Ekores. The gamma unit had been installed at the Minsk, Belarus Research Institute of Physical Organic Chemistry in 1971. This research center is a part of the National Academy of Sciences of Belarus. There were 48 highly radioactive Cobalt 60 sources inside the unit. During the dismantling process it was discovered that the source retaining drum was broken. Therefore, the removal proved to be much more complicated than originally envisioned. IT and Radii were unable to remove a number of sources from the drum by means of standard removal equipment and finally needed to remove the entire unit itself along with the remaining sources. The problem was further complicated by a very short timeframe allowed for work implementation. This was because the unit was located near an elementary school requiring the team to complete the project prior to the end of summer school holidays. Moreover, the unit itself was not easily accessible. The project team developed a technical solution and accomplished the removal.

Abstract #384 (3:40 p.m.)



A Methodology for Comparing and Evaluating Electronic Continuous Tracking, Monitoring, and Locating Technologies for Mobile Radioactive Sources in Transport

Esther M. Bryan
US Department of Energy National Nuclear Security Administration Office of Radiological Security, Washington, DC, USA

Numerous electronic tracking, monitoring and locating technologies are available commercially, but only a few may be appropriate for use in the transport of mobile radioactive sources. If applied effectively to radioactive source shipments, such technologies could add an extra layer of security during the most vulnerable phase when materials are being shipped from one location to another. Given the dynamic nature of tracking, monitoring and locating technologies, users need more systematic methods to comparatively analyze these technologies in the selection of an optimal one for a given application. This paper provides a methodology to help users, developers, and decision makers compare and evaluate electronic continuous tracking, monitoring and locating technologies for mobile radioactive source transport. The methodology proposes 21 general criteria defined and organized into the following seven categories: (1) communications; (2) adaptable physical attributes; (3) operational attributes; (4) flexibility; (5) system performance; (6) durability and longevity; and (7) practicality. Ultimately, the optimal technology for any given application must be further derived from user defined requirements, and this methodology provides a general framework to inform a given technology’s limitations and advantages.

Abstract #180 (4:00 p.m.)



Crediting the Inner O-Ring of a Chalfant‑Style Containment Vessel

Glenn Abramczyk, Bradley M. Loftin, Charles A. McKeel, John S. Bellamy, Donald J. Trapp
Savannah River National Laboratory, Aiken, SC, USA

Type B Radioactive Material Packaging are required by federal regulation to provide containment of the radioactive materials being transported with an allowable release rate based upon the conditions and the material. Type B Packaging design incorporate a Containment Vessel (CV) which performs this function. CV designs typically have two separate seals, one that performs the content containment function and a second that is used to verify the installation and functionality of the containment seal. The Savannah River National Laboratory has designed the “Chalfant‑style” Containment Vessel, used on a family of Type B radioactive material packages, which utilizes a conical plug with dual O‑Rings which is held in contact by a threaded nut to a conical sealing surface on the vessel body. The “outer” O‑Ring is credited in the safety basis documents for providing containment. This paper discusses the challenges and benefits of crediting the “inner” O‑Ring of a Chalfant‑Style CV.

Abstract #201 (4:40 p.m.)



Fabrication of the Mk-18A Transfer Cask

Bradley M. Loftin, Mark Bowers, Glenn Abramczyk
Savannah River National Laboratory, Aiken, SC, USA

The Mk-18A transfer cask was designed for transfer of legacy Mk-18A target assemblies from the L-Basin Storage Area at the Savannah River Site (SRS) to the Savannah River National Laboratory (SRNL). SRNL needed to design a site-specific cask as all existing designs required extensive facility modifications to perform the necessary onsite shielding functions; in the L-Basin, during transfer, and at the SRNL. Once the design was approved, SRNL procurement requested quotes for its fabrication and awarded the contract to Petersen, Inc. in Ogden, Utah. This paper will provide a detailed look at how the Mk-18A will be used to accomplish the mission of recovering the Mk-18A target assemblies, it will briefly describe the process used at SRNL for procurement of radioactive material packagings, it will document and describe challenges in the fabrication of the cask, and it will explore requested design changes made by the fabricator and any impacts those changes have on the cask’s ability to meet its performance requirements.

Abstract #181 (5:00 p.m.)



Evaluating a Type B Radioactive Material Packaging to Air Transport Conditions

Glenn Abramczyk, Bradley M. Loftin, Charles A. McKeel, John S. Bellamy
Savannah River National Laboratory, Aiken, SC, USA

Type B Radioactive Material Packagings (RAM) are designed and built to performance standards under normal conditions of transport (NCT) and hypothetical accident conditions (HAC) as defined by Title 10 Part 71 of the Code of Federal Regulations. Recently, Type B RAM Packages were transported by air. This event raised the question as to the expected performance and resultant condition of these packagings following this experience. This paper discusses the similarities and differences in the regulatory requirements for Type B and air transport packagings and evaluates the performance, both hypothetical and by testing, of the packages that were transported by air.

Abstract #345 (5:20 p.m.)



SAVY-4000 Storage Container and Program Update

Kirk Reeves, Robin Cunningham, Timothy Stone
Los Alamos National Laboratory, Los Alamos, NM, USA

The SAVY-4000™ is used at Los Alamos National Laboratory as a DOE Manual M441.1-1 compliant container. The SAVY-4000™ has been approved for use since 2013 and 3912 containers have been manufactured to date. The SAVY-4000™ is an innovative and creative design demonstrated by the fact that it can be opened and closed in a few seconds without torque wrenches or other tools; has a built-in, fire-rated filter that prevents the build-up of hydrogen gas, yet retains 99.97% of plutonium particulates, and prevents release of material even in a 12 foot drop. Finally, it has been fire then drop tested and will reduce the risk to the public in the event of an earthquake/fire scenario. This will allow nuclear facilities to credit the container towards source term Material at Risk (MAR) reduction. Over the course of the use of the container, improvements, new sizes, and life-time extension efforts have continued. The life-time extension efforts initially revolved around the polymer components and a 40-year design life is established. The surveillance activities performed as part of the M441.1-1 requirements led to the investigation of the effect of corrosion on the stainless steel components of the container. The investigation is on-going. The membrane used to keep the container water-tight is a PTFE material. The PTFE material does not withstand heat or radiation well, and as the SAVY-4000™ has been used in ways that were not intended in the initial design scope, a new water resistant filter was developed. The filter consists of a PFOTS treated ceramic fiber filter. This filter has been seen to remain hydrophobic up to 450 degrees Celsius and an improvement to gamma and alpha radiation resistance. A new container size was also developed to optimize the storage in certain locations. A new bag-out bag material is being proposed as well as a new in-glovebox container to the overall storage package.