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The use of performance assessments in Yucca Mountain repository waste package design activities

Description: The Yucca Mountain Project is developing performance assessment approaches as part of the evaluations of the suitability of Yucca Mountain as a repository site. Lawrence Livermore National Laboratory is developing design concepts and the scientific performance assessment methodologies and techniques used for the waste package and engineered barrier system components. This paper presents an overview of the approach under development for postclosure performance assessments that will guide the conceptual design activities and assist in the site suitability evaluations. This approach includes establishing and modeling for the long time periods required by regulations: near-field environment characteristics surrounding the emplaced wastes; container materials performance responses; and waste form properties. All technical work is being done under a fully qualified quality assurance program.
Date: January 1, 1990
Creator: Jardine, L.J.
Partner: UNT Libraries Government Documents Department

Development of a Comprehensive Plan for Scientific Research, Exploration, and Design: Creation of an Undergroung Radioactive Waste Isloation Facility at the Nizhnekansky Rock Massif

Description: ISTC Partner Project No.2377, ''Development of a General Research and Survey Plan to Create an Underground RW Isolation Facility in Nizhnekansky Massif'', funded a group of key Russian experts in geologic disposal, primarily at Federal State Unitary Enterprise All-Russian Design and Research Institute of Engineering Production (VNIPIPT) and Mining Chemical Combine Krasnoyarsk-26 (MCC K-26) (Reference 1). The activities under the ISTC Partner Project were targeted to the creation of an underground research laboratory which was to justify the acceptability of the geologic conditions for ultimate isolation of high-level waste in Russia. In parallel to this project work was also under way with Minatom's financial support to characterize alternative sections of the Nizhnekansky granitoid rock massif near the MCC K-26 site to justify the possibility of creating an underground facility for long-term or ultimate isolation of radioactive waste (RW) and spent nuclear fuel (SNF). (Reference 2) The result was a synergistic, integrated set of activities several years that advanced the geologic repository site characterization and development of a proposed underground research laboratory better than could have been expected with only the limited funds from ISTC Partner Project No.2377 funded by the U.S. DOE-RW. There were four objectives of this ISTC Partner Project 2377 geologic disposal work: (1) Generalize and analyze all research work done previously at the Nizhnekansky granitoid massif by various organizations; (2) Prepare and issue a declaration of intent (DOI) for proceeding with an underground research laboratory in a granite massif near the MCC K-26 site. (The DOI is similar to a Record of Decision in U.S. terminology). (3) Proceeding from the data obtained as a result of scientific research and exploration and design activities, prepare a justification of investment (JOI) for an underground research laboratory in as much detail as the available site characterization data allow. Consider the ...
Date: June 15, 2005
Creator: Jardine, L J
Partner: UNT Libraries Government Documents Department

Proceedings of the 6th Annual Meeting for Excess Weapons Plutonium Disposition: Plutonium Packaging, Storage and Transportation and WasteTreatment, Storage and Disposal Activities

Description: The sixth annual Excess Weapons Plutonium Disposition meeting organized by Lawrence Livermore National Laboratory (LLNL) was held November 15-17, 2004, at the State Education Center (SEC), 4 Aerodromnya Drive, St. Petersburg, Russia. The meeting discussed Excess Weapons Plutonium Disposition topics for which LLNL has the US Technical Lead Organization responsibilities. The technical areas discussed included Radioactive Waste Treatment, Storage, and Disposal, and Plutonium Oxide and Plutonium Metal Packaging, Storage and Transportation and Spent Fuel Packaging, Storage and Transportation. The meeting was conducted with a conference format using technical presentations of papers with simultaneous translation into English and Russian. There were 55 Russian attendees from 16 different Russian organizations and four non-Russian attendees from the US. Forty technical presentations were made. The meeting agenda is given in Appendix B and the attendance list is in Appendix C. The 16 different Russian design, industrial sites, and scientific organizations in attendance included staff from Rosatom/Minatom, Federal Nuclear and Radiation Safety Authority of Russia (GOSATOMNADZOR, NIERA/GAN), All Russian Designing & Scientific Research Institute of Complex Power Technology (VNIPIET), Khlopin Radium Institute (KRI), A. A. Bochvar All Russian Scientific Research Institute of Inorganic Materials (VNIINM), All Russian & Design Institute of Production Engineering (VNIPIPT), Ministry of Atomic Energy of Russian Federation Specialized State Designing Institute (GSPI), State Scientific Center Research Institute of Atomic Reactors (RIAR), Siberian Chemical Combine Tomsk (SCC), Mayak PO, Mining Chemical Combine (MCC K-26), Institute of Biophysics (IBPh), Sverdlosk Scientific Research Institute of Chemical Machine Building (SNIIChM), Kurchatov Institute (KI), Institute of Physical Chemistry Russian Academy of Science (IPCh RAS) and Radon PO-Moscow. The four non-Russian attendees included one representative from DOE NNSA, and LLNL, and two from Duratek, The meeting was organized into three major sessions: (1) Waste Treatment, Storage and Disposal; (2) Plutonium Packaging, Storage and Transportation; (3) Spent Fuel ...
Date: June 30, 2005
Creator: Jardine, L J
Partner: UNT Libraries Government Documents Department

Integrated approach to nuclear materials safety management in the U.S. and Russia

Description: The United States and Russia are dismantling nuclear weapons and generating hundreds of tons of excess plutonium and high enriched uranium fissile nuclear materials that require disposition. The U.S. Department of Energy and the Ministry of the Russian Federation for Atomic Energy (Minatom) organizations are planning and implementing safe, secure storage and disposition operations for these materials in numerous facilities. This provides a new opportunity for technical exchanges between Russian and Western scientists that can establish an integrated and improved common safety culture for handling these materials. The development and use of personal relationships and joint projects among Russian and Western participants involved in fissile nuclear materials safety management contributes to improving nuclear materials nonproliferation and to making a safer world. Technical exchanges and workshops are being used to systematically identify opportunities in the nuclear fissile materials facilities to improve and ensure the safety of workers, the public, and the environment.
Date: June 1, 1997
Creator: Jardine, L.J.
Partner: UNT Libraries Government Documents Department

Disposition of excess weapons plutonium from dismantled weapons

Description: With the end of the Cold War and the implementation of various nuclear arms reduction agreements, US and Russia have been actively dismantling tens of thousands of nuclear weapons. As a result,large quantities of fissile materials, including more than 100 (tonnes?) of weapons-grade Pu, have become excess to both countries` military needs. To meet nonproliferation goals and to ensure the irreversibility of nuclear arms reductions, this excess weapons Pu must be placed in secure storage and then, in timely manner, either used in nuclear reactors as fuel or discarded in geologic repositories as solid waste. This disposition in US and Russia must be accomplished in a safe, secure manner and as quickly as practical. Storage of this Pu is a prerequisite to any disposition process, but the length of storage time is unknown. Whether by use as fuel or discard as solid waste, disposition of that amount of Pu will require decades--and perhaps longer, if disposition operations encounter delays. Neither US nor Russia believes that long-term secure storage is a substitute for timely disposition of excess Pu, but long-term, safe, secure storage is a critical element of all excess Pu disposition activities.
Date: January 1, 1997
Creator: Jardine, L.J.
Partner: UNT Libraries Government Documents Department

Russian-American strategy for stabilization and immobilization of excess Russian weapons origin plutonium

Description: In the US, impure Pu-containing materials such as residues and scrapes are in storage, in known quantities, and in materials of various compositions with known Pu contents. However, in Russia, there are no substantial quantities of accumulated impure Pu-containing materials awaiting processing either for disposition or for transuranic (TRU) geologic disposal as there are in the Us. during the Cold War, the Russian approach to Pu processing for weapons production was different from that of the US. All impure Pu- containing materials were routinely reprocessed, and the residual Pu was recovered and purified for reuse until residual Pu levels of less than 200 mg/kg (less than 200 ppm) in any discharged solid process waste streams were reached. Wastes containing less than 200 ppm Pu were routinely discharged for burial in cement waste forms. Russia is studying changing from this practice of recovery of impure Pu for reuse to immobilizing future impure Pu-containing materials into solids at higher concentrations of Pu than 200 ppm for eventual geologic disposal.
Date: February 9, 1998
Creator: Jardine, L.J.
Partner: UNT Libraries Government Documents Department

Preclosure safety analysis for a prospective Yucca Mountain conceptual design repository

Description: A preliminary probabilistic risk assessment was performed for the prospective Yucca Mountain conceptual design repository. A new methodology to quantify radioactive source terms was developed and applied in the analysis. The study identified 42 event trees comprising 278 accident scenarios. The maximum offsite dose evaluated in this study is about 1000 mrem. For the majority of the accident scenarios, either the offsite dose is less than 100 mrem or the probability of occurrence is less than 1 {times} 10{sup {minus}9}/yr. Only 11 accident scenarios with a dose larger than 100 mrem and an associated probability greater than 1 {times} 10{sup {minus}9}/yr were identified. A more detailed follow-on analysis for seismic events of various severity was also performed, and similar results were obtained. Therefore, based on the results of this analysis, no significant risk to the general public was identified during the preclosure period for the conceptual repository design. 13 refs., 4 figs., 2 tabs.
Date: December 1, 1989
Creator: Ma, C.W. & Jardine, L.J.
Partner: UNT Libraries Government Documents Department

Perspective of metal encapsulation of waste. [Evaluation of solid waste encapsulation in lead alloys]

Description: A conceptual flow sheet is presented for encapsulating solid, stabilized calcine (e.g., supercalcine) in a solid lead alloy, using existing or developing technologies. Unresolved and potential problem areas of the flow sheet are outlined and suggestions are made as how metal encapsulation might be applied to other solid wastes from the fuel cycle. It is concluded that metal encapsulation is a technique applicable to many forms of solid wastes and is likely to meet future waste isolation criteria and regulations.
Date: January 1, 1978
Creator: Jardine, L.J. & Steindler, M.J.
Partner: UNT Libraries Government Documents Department

Metal matrix encapsulation of waste

Description: The ultimate disposition of nuclear wastes is frequently mentioned by opponents of nuclear power as an unresolved issue, and it is true that adequate demonstrations of nuclear waste disposal have not been performed. It has been suggested, however, that technology is either available or readily amenable for full-scale radioactive demonstrations once necessary Federal policy and criteria decisions are made. Public acceptance of nuclear power would be enhanced if the uncertainty of nuclear waste disposal is dispelled by successful waste disposal demonstrations under the full scrutiny of the public. It is our opinion that only full-scale radioactive demonstrations of waste disposal would qualify as an adequate demonstration and thereby reduce the antagonism which retards development of nuclear power. Thus, now is the time to initiate such demonstrations and we suggest that the concept of metal encapsulation of solidified high level waste forms be used as the method that can be acceptable to both the public and industry. This paper will briefly introduce the metal encapsulation concept by presenting a process flow sheet for encapsulation of wastes that would be produced by a 5 Mg/day reprocessing plant. Some probable attributes of metal-encapsulated waste forms and glass monoliths and of the fabrication processes for these waste forms will be discussed in order to illustrate the bases for the recommendation that metal encapsulation be the preferred route to achieving successful early demonstrations of nuclear waste disposal.
Date: January 1, 1978
Creator: Jardine, L.J. & Steindler, M.J.
Partner: UNT Libraries Government Documents Department

Review of metal-matrix encapsulation of solidified radioactive high-level waste

Description: Literature describing previous and current work on the encapsulation of solidified high-level waste forms in a metal matrix was reviewed. Encapsulation of either stabilized calcine pellets or glass beads in alloys by casting techniques was concluded to be the most developed and direct approach to fabricating solid metal-matrix waste forms. Further characterizations of the physical and chemical properties of metal-matrix waste forms are still needed to assess the net attributes of metal-encapsulation alternatives. Steady-state heat transfer properties of waste canisters in air and water environments were calculated for four reference waste forms: (1) calcine, (2) glass monoliths, (3) metal-encapsulated calcine, and (4) metal-encapsulated glass beads. A set of criteria for the maximum allowable canister centerline and surface temperatures and heat generation rates per canister at the time of shipment to a Federal repository was assumed, and comparisons were made between canisters of these reference waste forms of the shortest time after reactor discharge that canisters could be filled and the subsequent ''interim'' storage times prior to shipment to a Federal repository for various canister diameters and waste ages. A reference conceptual flowsheet based on existing or developing technology for encapsulation of stabilized calcine pellets is discussed. Conclusions and recommendations are presented.
Date: May 1, 1978
Creator: Jardine, L.J. & Steindler, M.J.
Partner: UNT Libraries Government Documents Department

Alternate strategy for commercial high-level radioactive-waste management

Description: A current strategy of geologic disposal of immobilized commercial, high-level, nuclear wastes provides long-term storage (hundreds of thousands of years) for a wide spectrum of wastes from the Purex process which would be immobilized in a borosilicate glass. When implaced in a repository, the temperature increases and peaks within the geologic formations housing the waste repository during the first several hundred years after burial and then declines towards the initial temperature. During this thermal and radiolysis pulse period, the geologic formation and waste packages could be significantly perturbed unless the effects are controlled by some engineered approach. Many of the proposed solutions introduce new economic penalties and/or have serious impacts on how the volume of waste must be handled in production, transportation and final interment in the repository. It is noted that the majority of the thermal energy (has high as 98% after 30 years) in commercial waste aged between 3 and 150 years is due to only two radioactive isotopes, /sup 90/Sr and /sup 137/Cs and their decay chains, which constitute < 10 wt. % of the total elements in HLW. Thus removal of cesium and strontium from all the other HLW components greatly reduces the geologic/waste package perturbations caused by the thermal/radiolysis pulse.
Date: January 1, 1982
Creator: Northrup, C.J.; Jardine, L.J. & Steindler, M.J.
Partner: UNT Libraries Government Documents Department

Disposition of excess plutonium using ``off-spec`` MOX pellets as a sintered ceramic waste form

Description: The authors describe a potential strategy for the disposition of excess weapons plutonium in a way that minimizes (1) technological risks, (2) implementation costs and completion schedules, and (3) requirements for constructing and operating new or duplicative Pu disposition facilities. This is accomplished by an optimized combination of (1) using existing nuclear power reactors to ``burn`` relatively pure excess Pu inventories as mixed oxide (MOX) fuel and (2) using the same MOX fuel fabrication facilities to fabricate contaminated or impure excess Pu inventories into an ``off-spec`` MOX solid ceramic waste form for geologic disposition. Diversion protection for the SCWF to meet the ``spent fuel standard`` introduced by the National Academy of Sciences can be achieved in at least three ways. (1) One can utilize the radiation field from defense high-level nuclear waste by first packaging the SCWF pellets in 2- to 4-L cans that are subsequently encapsulated in radioactive glass in the Defense Waste Processing Facility (DWPF) glass canisters (a ``can-in-canister`` approach). (2) One can add {sup 137}Cs (recovered from defense wastes at Hanford and currently stored as CsCl in capsules) to an encapsulating matrix such as cement for the SCWF pellets in a small hot-cell facility and thus fabricate large monolithic forms. (3) The SCWF can be fabricated into reactor fuel-like pellets and placed in tubes similar to fuel assemblies, which can then be mixed in sealed repository containers with irradiated spent nuclear fuel for geologic disposition.
Date: February 1, 1996
Creator: Armantrout, G.A. & Jardine, L.J.
Partner: UNT Libraries Government Documents Department

Secondary wastes and high explosive residues generated during production of main high explosive charges for nuclear weapons. Revision 1

Description: This study identifies the sources of high-explosive (HE) residues and hazardous and nonhazardous wastes generated during the production of the main HE charges for nuclear weapons, and estimates their quantities and characteristics. The results can be used as a basis for design of future handling and treatment systems for solid and liquid HE residues and wastes at any proposed new HE production facilities. This paper outlines a general methodology for documenting and estimating the volumes and characteristics of the solid and liquid HE residues and hazardous and nonhazardous wastes. To facilitate the estimating, we separated the HE main-charge production process into ten discrete unit operations and four support operations, and identified the corresponding solid and liquid HE residues and waste quantities. Four different annual HE main-charge production rates of 100, 500, 1000, and 2000 HE units/yr were assumed to develop the volume estimates and to establish the sensitivity of the estimates to HE production rates. The total solids (HE residues and hazardous and nonhazardous wastes) estimated range from 800 to 2800 ft{sup 3}/yr and vary uniformly with the assumed HE production rate. The total liquids estimated range from 73,000 to 1,448.000 gal/yr and also vary uniformly with the assumed production rate. Of the estimated solids, the hazardous wastes (e.g., electrical vehicle batteries and light tubes) were about 2% of the total volumes. The generation of solid HE residues varied uniformly with the HE production rates and ranged from about 20% of the total solids volume for the 100 HE units/yr case to about 60% for the 2000 units/yr case. The HE machining operations generated 60 to 80% of the total solid HE residues, depending on the assumed production rate, and were also the sources of the most concentrated HE residues.
Date: January 1, 1995
Creator: Jardine, L.J. & McGee, J.T.
Partner: UNT Libraries Government Documents Department

U.S.-Russian experts NATO collaborative research grant exchange visit meeting on excess Pu ceramics formulations and characterizations

Description: This document contains the agenda and meeting notes. Topics of discussion included US Pu disposition ceramics activities, Russian experience and proposals in Pu ceramics, and development of possible Russian ceramic proposals or collaborations.
Date: November 24, 1998
Creator: Jardine, L.J., LLNL
Partner: UNT Libraries Government Documents Department

Yucca Mountain Site Characterization Project Waste Package Plan

Description: The goal of the US Department of Energy`s (DOE) Yucca Mountain Site Characterization Project (YMP) waste package program is to develop, confirm the effectiveness of, and document a design for a waste package and associated engineered barrier system (EBS) for spent nuclear fuel and solidified high-level nuclear waste (HLW) that meets the applicable regulatory requirements for a geologic repository. The Waste Package Plan describes the waste package program and establishes the technical approach against which overall progress can be measured. It provides guidance for execution and describes the essential elements of the program, including the objectives, technical plan, and management approach. The plan covers the time period up to the submission of a repository license application to the US Nuclear Regulatory Commission (NRC). 1 fig.
Date: February 1991
Creator: Harrison-Giesler, D. J.; Morissette, R. P. & Jardine, L. J.
Partner: UNT Libraries Government Documents Department

Using a systems engineering process to develop engineered barrier system design concepts

Description: The methodology used to develop conceptual designs of the engineered barrier system and waste packages for a geologic repository is based on an iterative systems engineering process. The process establishes a set of general mission requirements and then conducts detailed requirements analyses using functional analyses, system concept syntheses, and trade studies identifications to develop preliminary system concept descriptions. The feasible concept descriptions are ranked based on selection factors and criteria and a set of preferred concept descriptions is then selected for further development. For each of the selected concept descriptions, a specific set of requirements, including constraints, is written to provide design guidance for the next and more detailed phase of design. The process documents all relevant waste management system requirements so that the basis and source for the specific design requirements are traceable and clearly established. Successive iterations performed during design development help to insure that workable concepts are generated to satisfy the requirements. 4 refs., 2 figs.
Date: May 1, 1991
Creator: Jardine, L.J. & Short, D.W.
Partner: UNT Libraries Government Documents Department

Comparison of costs for solidification of high-level radioactive waste solutions: glass monoliths vs metal matrices

Description: A comparative economic analysis was made of four solidification processes for liquid high-level radioactive waste. Two processes produced borosilicate glass monoliths and two others produced metal matrix composites of lead and borosilicate glass beads and lead and supercalcine pellets. Within the uncertainties of the cost (1979 dollars) estimates, the cost of the four processes was about the same, with the major cost component being the cost of the primary building structure. Equipment costs and operating and maintenance costs formed only a small portion of the building structure costs for all processes.
Date: May 1, 1981
Creator: Jardine, L.J.; Carlton, R.E. & Steindler, M.J.
Partner: UNT Libraries Government Documents Department

Measurement of Leaching From Simulated Nuclear-Waste Glass Using Radiotracers

Description: The use of radiotracer spiking as a method of measuring the leaching from simulated nuclear-waste glass is shown to give results comparable with other analytical detection methods. The leaching behavior of /sup 85/Sr, /sup 106/Ru, /sup 133/Ba, /sup 137/Cs, /sup 141/Ce, /sup 152/Eu, and other isotopes is measured for several defense waste glasses. These tests show that radiotracer spiking is a sensitive, multielement technique that can provide leaching data, for actual waste elements, that are difficult to obtain by other methods. Additionally, a detailed procedure is described that allows spiked glass to be prepared with a suitable distribution of radionuclides.
Date: September 1982
Creator: Bates, J. K.; Jardine, L. J. & Steindler, M. J.
Partner: UNT Libraries Government Documents Department

Hydration process of nuclear-waste glass: an interim report

Description: Aging of simulated nuclear waste glass by contact with a controlled-temperature, humid atmosphere results in the formation of a double hydration layer penetrating the glass, as well as the formation of minerals on the glass surface. The hydration process can be described by Arrhenius behavior between 120 and 240/sup 0/C. Results suggest that simulated aging reactions are necessary for demonstrating that nuclear waste forms can meet projected Nuclear Regulatory Commission regulations. 16 figures, 4 tables.
Date: July 1, 1982
Creator: Bates, J.K.; Jardine, L.J. & Steindler, M.J.
Partner: UNT Libraries Government Documents Department

Framing a bilateral US-Russian geologic repository initiative

Description: This document summarizes a framework for the development of a bilateral United States�Russian geologic repository initiative to enable cooperative work on the science and technology of geologic disposal of high-level nuclear wastes and fissile-containing materials. Three different types of integrated technical activities in Russia are employed to focus and organize a Department of Energy (DOE) Office of Civilian Radioactive Waste Management (RW) FY00 initiative. We have specified the items for initial negotiations with the Russians for start-up activities in FY99 and early FY00. These first interactions will generate other activities which, by utilizing Russia�s unique capabilities, may assist us in the development and validation of the US geologic repository program. The current International Science and Technology Center (ISTC) cooperative study of 30years of heat effects on underground hardrock rock media at the closed city of Krasnoyarsk-26 (Zheleznorgorsk) is but one example of such a Russian geologic repository analogue project that may assist the US geologic repository program.
Date: September 8, 1998
Creator: Jardine, L J
Partner: UNT Libraries Government Documents Department

Utilization of Cs{sup 137} to generate a radiation barrier for weapons grade plutonium immobilized in borosilicate glass canisters. Revision 1

Description: One of the ways recommended by a recent National Academy of Sciences study to dispose of excess weapons-grade plutonium is to encapsulate the plutonium in a glass in combination with high-level radioactive wastes (HLW) to generate an intense radiation dose rate field. The objective is to render the plutonium as difficult to access as the plutonium contained in existing US commercial spent light-water reactor (LWR) fuel until it can be disposed of in a permanent geological repository. A radiation dose rate from a sealed canister of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years after fabrication was assumed in this paper to be a radiation dose comparable to spent LWR fuel. This can be achieved by encapsulating the plutonium in a borosilicate glass with an adequate amount of a single fission product in the HLWS, namely radioactive Cs{sup 137}. One hundred thousand curies of Cs{sup 137} will generate a dose rate of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years when imbedded into canisters of the size proposed for the Savannah River Site`s vitrified high-level wastes. The United States has a current inventory of 54 MCi of CS{sup 137} that has been separated from defense HLWs and is in sealed capsules. This single curie inventory is sufficient to spike 50 metric tons of excess weapons-grade plutonium if plutonium can be loaded at 5.5 wt% in glass, or 540 canisters. Additional CS{sup 137} inventories exist in the United States` HLWs from past reprocessing operations, should additional curies be required. Using only one fission product, CS{sup 137}, rather than the multiple chemical elements and compounds in HLWs to generate a high radiation dose rate from a glass canister greatly simplifies the processing engineering retirement for encapsulating plutonium in a borosilicate glass.
Date: January 1, 1995
Creator: Jardine, L.J.; Armantrout, G.A. & Collins, L.F.
Partner: UNT Libraries Government Documents Department

Investigations of plutonium immobilization into the vitreous compositions

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 ...
Date: March 2, 1998
Creator: Matyunin, Y.I. & Jardine, L.J.
Partner: UNT Libraries Government Documents Department

Qualifying radioactive waste forms for geologic disposal

Description: We have developed a phased strategy that defines specific program-management activities and critical documentation for producing radioactive waste forms, from pyrochemical processing of spent nuclear fuel, that will be acceptable for geologic disposal by the US Department of Energy. The documentation of these waste forms begins with the decision to develop the pyroprocessing technology for spent fuel conditioning and ends with production of the last waste form for disposal. The need for this strategy is underscored by the fact that existing written guidance for establishing the acceptability for disposal of radioactive waste is largely limited to borosilicate glass forms generated from the treatment of aqueous reprocessing wastes. The existing guidance documents do not provide specific requirements and criteria for nonstandard waste forms such as those generated from pyrochemical processing operations.
Date: September 1, 1994
Creator: Jardine, L.J.; Laidler, J.J. & McPheeters, C.C.
Partner: UNT Libraries Government Documents Department