Author: Publisher:
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Languages : en
Pages :
Book Description
Application of Spent Fuel Treatment Technology to Plutonium Immobilization
Author: An Analysis of Plutonium Immobilization Versus the "spent Fuel" Standard
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Safe Pu management is an important and urgent task with profound environmental, national, and international security implications. Presidential Policy Directive 13 and analyses by scientific, technical, and international policy organizations brought about a focused effort within the Department of Energy (DOE) to identify and implement long-term disposition paths for surplus Pu. The principal goal is to render surplus Pu as inaccessible and unattractive for reuse in nuclear weapons as Pu in spent reactor fuel. In the Programmatic Environmental Impact Statement and Record of Decision for the Storage and Disposition of Weapons- Usable Fissile Materials (1997), DOE announced pursuit of two disposition technologies: (1) irradiation of Pu as MOX fuel in existing reactors and (2) immobilization of Pu into solid forms containing fission products as a radiation barrier. DOE chose an immobilization approach that includes use of the can-in-canister option. . for a portion of the surplus, non-pit Pu material. In the can-in-canister approach, cans of glass or ceramic forms containing Pu are encapsulated within canisters of HLW glass. In support of the selection process, a technical evaluation of retrievability and recoverability of Pu from glass and ceramic forms by a host nation and by rogue nations or subnational groups was completed. The evaluation involved determining processes and flowsheets for Pu recovery, comparing these processes against criteria and metrics established by the Fissile Materials Disposition Program and then comparing the recovery processes against each other and against SNF processes.
The Office of Environmental Management Technical Reports
Author: Publisher:
ISBN:
Category : Environmental management
Languages : en
Pages : 972
Book Description
Integrated Development and Testing Plan for the Plutonium Immobilization Project
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
This integrated plan for the DOE Office of Fissile Materials Disposition describes the technology development and major project activities necessary to support the deployment of the immobilization approach for disposition of surplus weapons-usable plutonium. The plan describes details of the development and testing tasks needed to provide technical data for design and operation of a plutonium immobilization plant based on the ceramic can-in-canister technology. The plan also presents tasks for characterization and performance testing of the immobilization form to support a repository licensing application and to develop the basis for repository acceptance of the plutonium form. Essential elements of the plant project (design, construction, facility activation, etc.) are described, but not developed in detail, to indicate how the test results tie into the overall plant project. Given the importance of repository acceptance, specific activities to be conducted by the Office of Civilian Radioactive Waste Management to incorporate the plutonium form in the repository licensing application are provided in this document, together with a summary of how immobilization activities provide input to the license activity and waste qualification. The ultimate goal of the immobilization project is to develop, construct, and operate facilities that will immobilize from about 18 to 50 tonnes of US surplus plutonium materials in a manner that meets the ''spent fuel'' standard and is acceptable for disposal in a geologic repository. The can-in-canister technology is accomplished by encapsulating the plutonium-containing ceramic forms within large canisters of high level waste glass.
Improving the Scientific Basis for Managing DOE's Excess Nuclear Materials and Spent Nuclear Fuel
Author: National Research CouncilPublisher: National Academies Press
ISBN: 0309087228
Category : Science
Languages : en
Pages : 124
Book Description
The production of nuclear materials for the national defense was an intense, nationwide effort that began with the Manhattan Project and continued throughout the Cold War. Now many of these product materials, by-products, and precursors, such as irradiated nuclear fuels and targets, have been declared as excess by the Department of Energy (DOE). Most of this excess inventory has been, or will be, turned over to DOE's Office of Environmental Management (EM), which is responsible for cleaning up the former production sites. Recognizing the scientific and technical challenges facing EM, Congress in 1995 established the EM Science Program (EMSP) to develop and fund directed, long-term research that could substantially enhance the knowledge base available for new cleanup technologies and decision making. The EMSP has previously asked the National Academies' National Research Council for advice for developing research agendas in subsurface contamination, facility deactivation and decommissioning, high-level waste, and mixed and transuranic waste. For this study the committee was tasked to provide recommendations for a research agenda to improve the scientific basis for DOE's management of its high-cost, high-volume, or high-risk excess nuclear materials and spent nuclear fuels. To address its task, the committee focused its attention on DOE's excess plutonium-239, spent nuclear fuels, cesium-137 and strontium-90 capsules, depleted uranium, and higher actinide isotopes.
The Office of Environmental Management Technical Reports
Author: Publisher:
ISBN:
Category : Environmental management
Languages : en
Pages : 968
Book Description
Investigations of Plutonium Immobilization Into the Vitreous Compositions
Author: Publisher:
ISBN:
Category :
Languages : en
Pages : 6
Book Description
Disposal of radioactive waste is a central problem and among the most important concerns of the nuclear fuel cycle. The Russian concept of nuclear fuel-cycle management is aimed at reprocessing spent fuel with the maximum, economically justified extraction of useful components for their recycling. The technology currently used in Russia for reprocessing spent nuclear fuel gives rise to liquid high- level waste (HLW) with minor concentrations of valuable components such as uranium (U) and plutonium (Pu) [1]. The liquid radioactive wastes formed in the course of reprocessing are converted into the solid forms suitable for the transportation, storage, and burial. Of special importance is management of high-level waste (HLW). Although various technological approaches underlying the processes for the solidification or immobilization of liquid HLW are used at the research institutes of the MINATOM RF [1-5], all these approaches have in common the idea of a strong bonding of radionuclides in the resulting solid matrices. Therefore, development of solidification technologies must include the mandatory stages of investigating the behavior of HLW components during the immobilization process and in the prepared solidified compositions and characterizing their properties under conditions for subsequent transportation, storage, and burial. An important technological area of exploration is study of the behavior of long-lived alpha radionuclides during the course of the vitrification process and the ultimate long-range influence of these radionuclides on the properties of the immobilized forms. For the most part, immobilization of alpha radionuclides, particularly plutonium, in vitreous compositions involves investigations on the properties of final materials and the effect of alpha-decay radiation on the synthesized solid compositions. Another direction of investigation is study on the behavior of plutonium and transplutonium elements upon vitrification of liquid HLW, as applied to the one-stage process for immobilizing HLW by using different types of melters. Such studies were carried out to forecast the behavior of the above radionuclides during long-term operation of the ceramic melter at the vitrification facility of PU 'Mayak.' The results of many investigations on the behavior of plutonium upon immobilization into phosphate and borosilicate vitreous compositions developed in Russia are generalized and summarized in the present work. In the conducted investigations of plutonium immobilization into both phosphate and borosilicate vitreous compositions used for the solidification of high-level liquid wastes upon vitrification in ceramic melters,0272 plutonium exhibited a limited solubility in the studied glass matrices. The solubility of plutonium, using plutonium dioxide powders, in phosphate and borosilicate glasses of specifically studied compositions was limited to 0.2 -0.4 wt %. The degree of incorporation (i.e., solubility) of plutonium, using plutonium in the form of nitrate solutions, in borosilicate glasses was also equal to 0.2-0.4 wt %. The degree of incorporation (i.e., solubility) of plutonium, using plutonium in the form of nitrate solutions, in phosphate glasses depended considerably on the chemical compositions of the solution to be solidified and on the specific glass matrix (i. e., on the composition of final solidified product) and was equal to 0.4-1.0 wt %. Available experimental data also allow one to assume that the use of the cold-crucible- induction melter (CCIM) method for immobilizing plutonium-containing wastes [6-8] provides a means of synthesizing the high-quality final solid-glass products with a plutonium content.
The Spent Fuel Standard - Does the Can-in-canister Concept for Plutonium Immobilization Measure Up?.
Author: Publisher:
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Languages : en
Pages :
Book Description
Critics continue to question whether or not the can-in-canister concept for immobilization and disposal of surplus plutonium meets the ''Spent Fuel Standard.'' Following this standard would make this plutonium roughly as ''inaccessible for weapons use as the much larger and growing quantity of plutonium that exists in spent fuel from commercial reactors.'' These critics take a narrower view of the ''Spent Fuel Standard'' than was intended in the National Academy reports, rather than considering the total effective barrier. This paper directly compares retrieval and recovery of plutonium from a can-in-canister to a spent fuel assembly. The conclusion from this study, as from earlier studies, is that the plutonium in the can-in-canister form is less accessible and less attractive to a potential proliferate than the plutonium that exists in spent fuel from commercial reactors.
Proceedings
Author: Plutonium Disposition and the U.S. Mixed Oxide Fuel Fabrication Facility
Author: United States. Congress. House. Committee on Armed Services. Strategic Forces SubcommitteePublisher:
ISBN:
Category : Science
Languages : en
Pages : 120
Book Description