A review of plutonium oxalate decomposition reactions and effects of decomposition temperature on the surface area of the plutonium dioxide product

Energy Technology Data Exchange (ETDEWEB)

Orr, R.M.; Sims, H.E.; Taylor, R.J., E-mail: robin.j.taylor@nnl.co.uk

2015-10-15

Plutonium (IV) and (III) ions in nitric acid solution readily form insoluble precipitates with oxalic acid. The plutonium oxalates are then easily thermally decomposed to form plutonium dioxide powder. This simple process forms the basis of current industrial conversion or ‘finishing’ processes that are used in commercial scale reprocessing plants. It is also widely used in analytical or laboratory scale operations and for waste residues treatment. However, the mechanisms of the thermal decompositions in both air and inert atmospheres have been the subject of various studies over several decades. The nature of intermediate phases is of fundamental interest whilst understanding the evolution of gases at different temperatures is relevant to process control. The thermal decomposition is also used to control a number of powder properties of the PuO{sub 2} product that are important to either long term storage or mixed oxide fuel manufacturing. These properties are the surface area, residual carbon impurities and adsorbed volatile species whereas the morphology and particle size distribution are functions of the precipitation process. Available data and experience regarding the thermal and radiation-induced decompositions of plutonium oxalate to oxide are reviewed. The mechanisms of the thermal decompositions are considered with a particular focus on the likely redox chemistry involved. Also, whilst it is well known that the surface area is dependent on calcination temperature, there is a wide variation in the published data and so new correlations have been derived. Better understanding of plutonium (III) and (IV) oxalate decompositions will assist the development of more proliferation resistant actinide co-conversion processes that are needed for advanced reprocessing in future closed nuclear fuel cycles. - Highlights: • Critical review of plutonium oxalate decomposition reactions. • New analysis of relationship between SSA and calcination temperature. â

Particle Generation by Laser Ablation in Support of Chemical Analysis of High Level Mixed Waste from Plutonium Production Operations

International Nuclear Information System (INIS)

Dickinson, J. Thomas; Alexander, Michael L.

2001-01-01

Investigate particles produced by laser irradiation and their analysis by Laser Ablation Inductively Coupled Plasma Mass Spectroscopy (LA/ICP-MS), with a view towards optimizing particle production for analysis of high level waste materials and waste glass. LA/ICP-MS has considerable potential to increase the safety and speed of analysis required for the remediation of high level wastes from cold war plutonium production operations. In some sample types, notably the sodium nitrate-based wastes at Hanford and elsewhere, chemical analysis using typical laser conditions depends strongly on the details of sample history composition in a complex fashion, rendering the results of analysis uncertain. Conversely, waste glass materials appear to be better behaved and require different strategies to optimize analysis

Estimating plutonium production fron long-lived radionuclides in permanent structural components of production reactor cores

International Nuclear Information System (INIS)

Fetter, S.

1992-01-01

This paper reports that the United States and the Soviet Union face critical decisions about the future of plutonium production for nuclear weapons. Both countries could eliminate the economic burden of rebuilding their production complexes by agreeing to ban the production of plutonium for weapons. Such an agreement could also provide important national-security benefits by reinforcing the Non-Proliferation Treaty and by diminishing the ability of both nations to break out of nuclear-arms reduction agreements at a later time - especially if the plutonium in the warheads eliminated by the arms reduction agreements is put under safeguards. A production cutoff would be verifiable

Baseline process description for simulating plutonium oxide production for precalc project

Energy Technology Data Exchange (ETDEWEB)

Pike, J. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-10-26

Savannah River National Laboratory (SRNL) started a multi-year project, the PreCalc Project, to develop a computational simulation of a plutonium oxide (PuO₂) production facility with the objective to study the fundamental relationships between morphological and physicochemical properties. This report provides a detailed baseline process description to be used by SRNL personnel and collaborators to facilitate the initial design and construction of the simulation. The PreCalc Project team selected the HB-Line Plutonium Finishing Facility as the basis for a nominal baseline process since the facility is operational and significant model validation data can be obtained. The process boundary as well as process and facility design details necessary for multi-scale, multi-physics models are provided.

Design safety features of containments used for handling plutonium in Reprocessing Plants

International Nuclear Information System (INIS)

Aherwal, P.; Achuthan, P.V.

2016-01-01

The plutonium present in spent fuel is separated from the associated uranium and fission products using solvent extraction cycles in process cells. Product plutonium nitrate solution containing trace concentrations of uranium and fission products is treated in the reconversion facility through a precipitation-calcination route and converted to sinterable grade plutonium oxide (PuO 2 ). All chemical operations involving materials with high plutonium content, both in solid and solution forms are carried out in glove boxes. Glove box provides an effective isolation from radioactive materials handled and acts as a barrier between the operator and the source of radiation. These glove boxes are interconnected for sequential operations and the interconnected glove box trains are installed within secondary enclosures called double skin which provides double barrier protection to operators

Utilization of plutonium in HTGR and its actinide production

International Nuclear Information System (INIS)

Karin, S.; Brogli, R.; Lefler, W.; Nordheim, L.

1976-01-01

The HTGR is a potential plutonium consumer. In this function it would burn plutonium, produce electricity and the valuable fissile isotope U-233. The advantages of this concept are discussed but particular attention is given to the production and the destruction of the higher actinides due to the high burnup achievable in such a system. The presence of the strong resonances in the plutonium isotopes demanded an extension of the methods for evaluation of self-shielding factors, a different structure for broad groups, and the adaptation of the reactor codes to these changes. Specifications for coated plutonium particles were developed. Also procedures were determined to evaluate the alpha ray and neutron emission rates of the actinide nuclides. First cycle calculations were carried out to establish in detail the characteristics of the plutonium reactors and their results are given

How much plutonium does North Korea really have?

International Nuclear Information System (INIS)

Dreicer, J.S.

1997-01-01

In a previous study, as part of the Global Nuclear Material Control Model effort, the author estimated the maximum quantity of plutonium that could be produced in thermal research reactors in the potential nuclear weapon states (including North Korea), based on their declared power level. D. Albright has estimated the amount of plutonium the North Koreans may have produced since 1986 in the 5-megawatt-electric power reactor at Yongbon. Albright provided an upper-bound estimate of 53 kilograms of weapon-grade plutonium produced cumulatively if the gas-graphite (magnox) reactor had achieved a load factor of 0.80. This cumulative estimate of 53 kilograms ignores the potential plutonium production in the 8-megawatt-thermal research reactor, IRT-DPRK. To better quantify the cumulative North Korean production, the author conducted a study to estimate the amount of plutonium that could have been produced in the IRT-DPRK research reactor operating at the declared power level during the entire period it has operated, including a period it was not safeguarded. The author estimates that, at most, an additional 6 to 33 kilograms of plutonium could have been produced cumulatively in the research reactor operating at the declared power level during the entire period it has operated, including a 12-year period it was not safeguarded, resulting in a total of 13 to 47 kilograms of plutonium possibly produced in both the research and power reactors

Use of plutonium for power production

International Nuclear Information System (INIS)

1965-01-01

The panel reviewed available information on various aspects of plutonium utilization, such as physics of plutonium, technology of plutonium fuels in thermal and fast reactors, behaviour of plutonium fuel under reactor irradiation, technological and economic aspects of plutonium fuel cycle. Refs, figs and tabs

Radiological safety aspects of handling plutonium

International Nuclear Information System (INIS)

Sundararajan, A.R.

2016-01-01

Department of Atomic Energy in its scheme of harnessing the nuclear energy for electrical power generation and strategic applications has given a huge role to utilization of plutonium. In the power production programme, fast reactors with plutonium as fuel are expected to play a major role. This would require establishing fuel reprocessing plants to handle both thermal and fast reactor fuels. So in the nuclear fuel cycle facilities variety of chemical, metallurgical, mechanical operations have to be carried out involving significant inventories of "2"3"9 Pu and associated radionuclides. Plutonium is the most radiotoxic radionuclide and therefore any facility handling it has to be designed and operated with utmost care. Two problems of major concern in the protection of persons working in plutonium handling facilities are the internal exposure to the operating personnel from uptake of plutonium and transplutonic nuclides as they are highly radiotoxic and the radiation exposure of hands and eye lens during fuel fabrication operations especially while handling recycled high burn up plutonium. In view of the fact that annual limit for intake is very small for "2"3"9Pu and its radiation emission characteristics are such that it is a huge challenge for the health physicists to detect Pu in air and in workers. This paper discusses the principles and practices followed in providing radiological surveillance to workers in plutonium handling areas. The challenges in protecting the workers from receiving exposures to hands and eye lens in handling high burn up plutonium are also discussed. The sites having Pu fuel cycle facilities should have trained medical staff to handle cases involving excessive intake of plutonium. (author)

Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

1999-01-01

The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

Review of operating experience at the Los Alamos Plutonium Electrorefining Facility, 1963-1977

International Nuclear Information System (INIS)

Mullins, L.J.; Morgan, A.N.

1981-12-01

This report reviews the operation of the Los Alamos Plutonium Electrorefining Plant at Technical Area 21 for the period 1964 through 1977. During that period, approximately 1568 kg of plutonium metal, > 99.95% pure, was produced in 653 runs from 1930 kg of metal fabrication scrap, 99% pure. General considerations of the electrorefining process and facility operation and recommendations for further improvement of the process are discussed

A probabilistic risk assessment of the LLNL Plutonium Facility's evaluation basis fire operational accident. Revision 1

International Nuclear Information System (INIS)

Brumburgh, G.P.

1995-01-01

The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous programmatic activities involving plutonium to include device fabrication, development of improved and/or unique fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed in July 1994 to address operational safety and acceptable risk to employees, the public, government property, and the environmental. This paper outlines the PRA analysis of the Evaluation Basis Fire (EBF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility

SEPARATION OF PLUTONIUM

Science.gov (United States)

Maddock, A.G.; Smith, F.

1959-08-25

A method is described for separating plutonium from uranium and fission products by treating a nitrate solution of fission products, uranium, and hexavalent plutonium with a relatively water-insoluble fluoride to adsorb fission products on the fluoride, treating the residual solution with a reducing agent for plutonium to reduce its valence to four and less, treating the reduced plutonium solution with a relatively insoluble fluoride to adsorb the plutonium on the fluoride, removing the solution, and subsequently treating the fluoride with its adsorbed plutonium with a concentrated aqueous solution of at least one of a group consisting of aluminum nitrate, ferric nitrate, and manganous nitrate to remove the plutonium from the fluoride.

Materials measurement and accounting in an operating plutonium conversion and purification process. Phase I. Process modeling and simulation

International Nuclear Information System (INIS)

Thomas, C.C. Jr.; Ostenak, C.A.; Gutmacher, R.G.; Dayem, H.A.; Kern, E.A.

1981-04-01

A model of an operating conversion and purification process for the production of reactor-grade plutonium dioxide was developed as the first component in the design and evaluation of a nuclear materials measurement and accountability system. The model accurately simulates process operation and can be used to identify process problems and to predict the effect of process modifications

A probabilistic risk assessment of the LLNL Plutonium facility's evaluation basis fire operational accident

International Nuclear Information System (INIS)

Brumburgh, G.

1994-01-01

The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous involving plutonium to include device fabrication, development of fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed rational safety and acceptable risk to employees, the public, government property, and the environment. This paper outlines the PRA analysis of the Evaluation Basis Fire (EDF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility

Plutonium production in a remote automated cell

International Nuclear Information System (INIS)

Maddux, E.P.; Purcell, J.A.

1987-01-01

Production of 239 Pu metal has traditionally been carried out in glove box enclosures to protect operating personnel and the outside environment from the hazards of radiation. With the advent of more stringent federal regulations for protecting personnel, the environment, and security of special nuclear materials, it is proposed that plutonium be processed in a totally remote environment using automated work stations and robotic materials transfer and equipment maintenance. Investigators envision a facility where operations are carried out in two remotely operated cells. One cell will contain aqueous processing operations while the second will contain powder operations. Various processing steps will be carried out on modular frames of a standard configuration. The processing equipment on each frame will be completely automated and operated via computer. Nonaqueous materials will be transferred between frames by robots using closed-circuit television and computer control. Robots will be able to perform limited in-place maintenance of equipment on each frame for high-maintenance replaceable components. When major repair is required, the processing frame will be decoupled by the robot and moved to a decontamination and maintenance area by a conventional overhead bridge crane. The new design concept will provide for long-term goals, which include exposure reduction, near-real time accountability, safeguards, and waste reduction

Plutonium active operation of the Winfrith modular containment system

International Nuclear Information System (INIS)

Sanders, M.J.; Pengelly, M.G.A.; McSherry, K.

1985-01-01

Three gloveboxes contaminated with plutonium have been dismantled inside the Winfrith Modular Containment System. This system is a portable, demountable pressurised suit area with its own filters and shower entry tunnel. Details of the operation are given. (U.K.)

Solvent anode for plutonium purification

International Nuclear Information System (INIS)

Bowersox, D.F.; Fife, K.W.; Christensen, D.C.

1986-01-01

The purpose of this study is to develop a technique to allow complete oxidation of plutonium from the anode during plutonium electrorefining. This will eliminate the generation of a ''spent'' anode heel which requires further treatment for recovery. Our approach is to employ a solvent metal in the anode to provide a liquid anode pool throughout electrorefining. We use molten salts and metals in ceramic crucibles at 700 0 C. Our goal is to produce plutonium metal at 99.9% purity with oxidation and transfer of more than 98% of the impure plutonium feed metal from the anode into the salt and product phases. We have met these criteria in experiments on the 100 to 1000 g scale. We plan to scale our operations to 4 kg of feed plutonium and to optimize the process parameters

Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

International Nuclear Information System (INIS)

Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

2000-01-01

The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

Preparation of fused chloride salts for use in pyrochemical plutonium recovery operations at Los Alamos

Energy Technology Data Exchange (ETDEWEB)

Fife, K.W.; Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

1986-07-01

The Plutonium Metal Technology Group at Los Alamos routinely uses pyrochemical processes to produce and purify plutonium from impure sources. The basic processes (metal production, metal purification, and residue treatment) involve controlling oxidation and reduction reactions between plutonium and its compounds in molten salts. Current production methods are described, as well as traditional approaches and recent developments in the preparation of solvent salts for electrorefining, molten salt extraction, lean metal (pyroredox) purification, and direct oxide reduction.

Preparation of fused chloride salts for use in pyrochemical plutonium recovery operations at Los Alamos

International Nuclear Information System (INIS)

Fife, K.W.; Bowersox, D.F.; Christensen, D.C.; Williams, J.D.

1986-07-01

The Plutonium Metal Technology Group at Los Alamos routinely uses pyrochemical processes to produce and purify plutonium from impure sources. The basic processes (metal production, metal purification, and residue treatment) involve controlling oxidation and reduction reactions between plutonium and its compounds in molten salts. Current production methods are described, as well as traditional approaches and recent developments in the preparation of solvent salts for electrorefining, molten salt extraction, lean metal (pyroredox) purification, and direct oxide reduction

Design of an integrated non-destructive plutonium assay facility

International Nuclear Information System (INIS)

Moore, C.B.

1984-01-01

The Department of Energy requires improved technology for nuclear materials accounting as an essential part of new plutonium processing facilities. New facilities are being constructed at the Savannah River Plant by the Du Pont Company, Operating Contractor, to recover plutonium from scrap and waste material generated at SRP and other DOE contract processing facilities. This paper covers design concepts and planning required to incorporate state-of-the-art plutonium assay instruments developed at several national laboratories into an integrated, at-line nuclear material accounting facility operating in the production area. 3 figures

A review of plutonium oxalate decomposition reactions and effects of decomposition temperature on the surface area of the plutonium dioxide product

Science.gov (United States)

Orr, R. M.; Sims, H. E.; Taylor, R. J.

2015-10-01

Plutonium (IV) and (III) ions in nitric acid solution readily form insoluble precipitates with oxalic acid. The plutonium oxalates are then easily thermally decomposed to form plutonium dioxide powder. This simple process forms the basis of current industrial conversion or 'finishing' processes that are used in commercial scale reprocessing plants. It is also widely used in analytical or laboratory scale operations and for waste residues treatment. However, the mechanisms of the thermal decompositions in both air and inert atmospheres have been the subject of various studies over several decades. The nature of intermediate phases is of fundamental interest whilst understanding the evolution of gases at different temperatures is relevant to process control. The thermal decomposition is also used to control a number of powder properties of the PuO2 product that are important to either long term storage or mixed oxide fuel manufacturing. These properties are the surface area, residual carbon impurities and adsorbed volatile species whereas the morphology and particle size distribution are functions of the precipitation process. Available data and experience regarding the thermal and radiation-induced decompositions of plutonium oxalate to oxide are reviewed. The mechanisms of the thermal decompositions are considered with a particular focus on the likely redox chemistry involved. Also, whilst it is well known that the surface area is dependent on calcination temperature, there is a wide variation in the published data and so new correlations have been derived. Better understanding of plutonium (III) and (IV) oxalate decompositions will assist the development of more proliferation resistant actinide co-conversion processes that are needed for advanced reprocessing in future closed nuclear fuel cycles.

Method of immobilizing weapons plutonium to provide a durable, disposable waste product

Science.gov (United States)

Ewing, Rodney C.; Lutze, Werner; Weber, William J.

1996-01-01

A method of atomic scale fixation and immobilization of plutonium to provide a durable waste product. Plutonium is provided in the form of either PuO.sub.2 or Pu(NO.sub.3).sub.4 and is mixed with and SiO.sub.2. The resulting mixture is cold pressed and then heated under pressure to form (Zr,Pu)SiO.sub.4 as the waste product.

Plutonium vulnerability issues at Hanford's Plutonium Finishing Plant

International Nuclear Information System (INIS)

Feldt, E.; Templeton, D.W.; Tholen, E.

1995-01-01

The Plutonium Finishing Plant (PFP) at the Hanford, Washington Site was operated to produce plutonium (Pu) metal and oxide for national defense purposes. Due to the production requirements and methods utilized to meet national needs and the abrupt shutdown of the plant in the late 1980s, the plant was left in a condition that poses a risk of radiation exposure to plant workers, of accidental radioactive material release to the environment, and of radiation exposure to the public. In early 1994, an Environmental Impact Statement (EIS) to determine the best methods for cleaning out and stabilizing Pu materials in the PFP was started. While the EIS is being prepared, a number of immediate actions have been completed or are underway to significantly reduce the greatest hazards in the PFP. Recently, increased attention his been paid to Pu risks at Department of Energy (DOE) facilities resulting in the Department-wide Plutonium Vulnerability Assessment and a recommendation by the Defense Nuclear Facilities Safety Board (DNFSB) for DOE to develop integrated plans for managing its nuclear materials

The separation of plutonium from uranium and fission products on zirconium phosphate columns

Energy Technology Data Exchange (ETDEWEB)

Gal, I; Ruvarac, A [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

In recent years special attention has been given to the ion-exchange properties of zirconium phosphate and similar compounds in aqueous solutions. These inorganic cation exchangers are stable in oxidizing media and at elevated temperatures. Their resistance to ionizing radiation makes them particularly suitable for work with radioactive solutions. On account of this we considered ir worthwhile to investigate the separation of plutonium from uranium and fission products on zirconium phosphate columns. We were interested in nitric and solutions containing macro-amounts of uranium (a few grams per litre), and micro-amounts of plutonium and long-lived fission products. To obtain a better insight into the ion-exchange behaviour of the different ionic species towards zirconium phosphate, we first determined the dependence of the distribution coefficients of uranium, plutonium and fission product cations on the aqueous nitric acid concentration. Then, taking the distribution data as a guide, we separated plutonium on small glass columns filled with zirconium phosphate and calculated the decontamination factors (author)

Plutonium metal preparation and purification at Los Alamos, 1984

International Nuclear Information System (INIS)

Christensen, D.C.; Williams, J.D.; McNeese, J.A.; Fife, K.W.

1984-01-01

Plutonium metal preparation and purification are well established at Los Alamos. Metal is prepared by calcothermic reduction of both PuF 4 and PuO 2 . Metal is purified by halide slagging, casting, and electrorefining. The product from the production sequence is ultrapure plutonium metal. All of the processes involve high temperature operation and all but casting involve molten salt media. Development efforts are fourfold: (1) recover plutonium values from residues; (2) reduce residue generation through process improvements and changes; (3) recycle of reagents, and (4) optimize and integrate all processes into a close-loop system. Plutonium residues are comprised of oxides, chlorides, colloidal metal suspensions, and impure metal heels. Pyrochemical recovery techniques are under development to address each residue. In addition, we are looking back at each residue generation step and are making process changes to reduce plutonium content in each residue. Reagent salt is the principle media used in pyrochemical processing. The regeneration and recycle of these reagents will both reduce our waste handling and operating expense. The fourth area, process optimization, involves both existing processes and new process developments. A status of efforts in all four of these areas will be summarized

Smart unattended systems for plutonium safeguards

International Nuclear Information System (INIS)

Menlove, H.O.; Abhold, M.; Eccleston, G.; Puckett, J.M.

1996-01-01

Large automated facilities for fabricating plutonium fuel present both difficulties and challenges for improved accounting of nuclear materials. The traditional methods of sample measurements, requiring the transfer of the sample from the production line to the assay measurement station, are not possible in automated facilities. The robotics used for automation require special containers for nuclear material that cannot be easily removed from the production line. Safety and radiation protection considerations also require that the assay instrumentation be installed in the fuel production lines because, in general, personnel cannot be in the fuel-handling area with nuclear material during operations. Such operational constraints are common in many of the modern facilities that have been designed for fabricating and processing plutonium fuel. A bilateral safeguards agreement between the US Department of Energy (DOE) and Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan was signed to develop and implement nondestructive assay (NDA) systems to provide continuous safeguards measurements for material accountancy in the robot-automated Plutonium Fuel Fabrication Facility (PFFF). The PFFF assay systems were required to operate in unattended mode with a size and fuel mass capability to match the robotics fuel manipulators. Unattended assay systems reduce the requirement for inspector's oversight of measurement operations, reduce the inspector's workload, and improve inspection efficiencies. In addition, unattended measurements become essential when facility constraints limit the access of inspectors to the operations area during material processing. Authentication techniques were incorporated into the NDA systems so that data obtained from unattended assays could be used by independent inspectors such as the IAEA

Reclamation of plutonium from pyrochemical processing residues

International Nuclear Information System (INIS)

Gray, L.W.; Gray, J.H.; Holcomb, H.P.; Chostner, D.F.

1987-04-01

Savannah River Laboratory (SRL), Savannah River Plant (SRP), and Rocky Flats Plant (RFP) have jointly developed a process to recover plutonium from molten salt extraction residues. These NaCl, KCL, and MgCl 2 residues, which are generated in the pyrochemical extraction of 241 Am from aged plutonium metal, contain up to 25 wt % dissolved plutonium and up to 2 wt % americium. The overall objective was to develop a process to convert these residues to a pure plutonium metal product and discardable waste. To meet this objective a combination of pyrochemical and aqueous unit operations was used. The first step was to scrub the salt residue with a molten metal (aluminum and magnesium) to form a heterogeneous ''scrub alloy'' containing nominally 25 wt % plutonium. This unit operation, performed at RFP, effectively separated the actinides from the bulk of the chloride salts. After packaging in aluminum cans, the ''scrub alloy'' was then dissolved in a nitric acid - hydrofluoric acid - mercuric nitrate solution at SRP. Residual chloride was separated from the dissolver solution by precipitation with Hg 2 (NO 3 ) 2 followed by centrifuging. Plutonium was then separated from the aluminum, americium and magnesium using the Purex solvent extraction system. The 241 Am was diverted to the waste tank farm, but could be recovered if desired

Transuranic (Tru) waste volume reduction operations at a plutonium facility

Energy Technology Data Exchange (ETDEWEB)

Cournoyer, Michael E [Los Alamos National Laboratory; Nixon, Archie E [Los Alamos National Laboratory; Dodge, Robert L [Los Alamos National Laboratory; Fife, Keith W [Los Alamos National Laboratory; Sandoval, Arnold M [Los Alamos National Laboratory; Garcia, Vincent E [Los Alamos National Laboratory

2010-01-01

Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA 55) involve working with various amounts of plutonium and other highly toxic, alpha-emitting materials. The spread of radiological contamination on surfaces, airborne contamination, and excursions of contaminants into the operator's breathing zone are prevented through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been discontinued and the room is being modified to support a new customer. The Actin ide Processing Group at TA-55 uses one-meter-long glass columns to process plutonium. Disposal of used columns is a challenge, since they must be size-reduced to get them out of the glovebox. The task is a high-risk operation because the glass shards that are generated can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker's skin when completing the task. One of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into a container for removal from the glove box as non-compactable transuranic (TRU) waste. This size-reduction operation reduces solid TRU waste generation by almost 2% times. Replacing one-time-use bag-out bags with multiple-use glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos

Transuranic (Tru) waste volume reduction operations at a plutonium facility

International Nuclear Information System (INIS)

Cournoyer, Michael E.; Nixon, Archie E.; Dodge, Robert L.; Fife, Keith W.; Sandoval, Arnold M.; Garcia, Vincent E.

2010-01-01

Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA 55) involve working with various amounts of plutonium and other highly toxic, alpha-emitting materials. The spread of radiological contamination on surfaces, airborne contamination, and excursions of contaminants into the operator's breathing zone are prevented through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been discontinued and the room is being modified to support a new customer. The Actin ide Processing Group at TA-55 uses one-meter-long glass columns to process plutonium. Disposal of used columns is a challenge, since they must be size-reduced to get them out of the glovebox. The task is a high-risk operation because the glass shards that are generated can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker's skin when completing the task. One of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into a container for removal from the glove box as non-compactable transuranic (TRU) waste. This size-reduction operation reduces solid TRU waste generation by almost 2% times. Replacing one-time-use bag-out bags with multiple-use glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos National

Transuranic (TRU) waste volume reduction operations at a plutonium facility

International Nuclear Information System (INIS)

Cournoyer, Michael E.; Nixon, Archie E.; Fife, Keith W.; Sandoval, Arnold M.; Garcia, Vincent E.; Dodge, Robert L.

2011-01-01

Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA-55) involve working with various amounts of plutonium and other highly toxic, alpha-emitting materials. The spread of radiological contamination on surfaces, airborne contamination, and excursions of contaminants into the operator's breathing zone are prevented through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been discontinued and the room is being modified to support a new customer. The Actinide Processing Group at TA-55 uses one-meter or longer glass columns to process plutonium. Disposal of used columns is a challenge, since they must be size-reduced to get them out of the glovebox. The task is a high-risk operation because the glass shards that are generated can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker's skin when completing the task. One of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into a container for removal from the glovebox as non-compactable transuranic (TRU) waste. This size-reduction operation reduces solid TRU waste volume generation by almost 2½ times. Replacing one-time-use bag-out bags with multiple-use glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos

Plutonium recycling and the problem of nuclear proliferation

International Nuclear Information System (INIS)

Albright, D.; Feiveson, H.S.

1988-01-01

A typical 1-gigawatt light water reactor (LWR), the dominant commercial power reactor type today, operating at 70% capacity factor, generates approximately 250 kilograms of plutonium annually. This plutonium, which is produced in the reactor through neutron capture by uranium-238, is then discharged from the reactor along with the other constituents of the spent fuel. About 70% of the plutonium, or 175 kilograms, consists of fissile (odd-numbered) plutonium isotopes. As long as the plutonium discharged from the reactor is left intermixed with the highly radioactive fission products also contained in the spent fuel, it cannot readily be used for power or for weapons. However, upon chemical separation from the radioactive fission products and other components of the spent reactor fuel, the plutonium produced each year in a gigawatt reactor could be used, either in recycled fuel (to replace about 175 kilograms of U-235 in a power reactor) or to provide the fissile material for more than 25 nuclear warheads. Commercial separation of plutonium and the introduction of nuclear fuel cycles using recycled plutonium, which are now impending in several countries, force one to balance the probable increased risks of nuclear proliferation due to these activities against various economic and other motives that have been forwarded in their defense. The authors undertake an assessment of this balancing in this article

Plutonium Finishing Plant (PFP) Final Safety Analysis Report (FSAR) [SEC 1 THRU 11

Energy Technology Data Exchange (ETDEWEB)

ULLAH, M K

2001-02-26

The Plutonium Finishing Plant (PFP) is located on the US Department of Energy (DOE) Hanford Site in south central Washington State. The DOE Richland Operations (DOE-RL) Project Hanford Management Contract (PHMC) is with Fluor Hanford Inc. (FH). Westinghouse Safety Management Systems (WSMS) provides management support to the PFP facility. Since 1991, the mission of the PFP has changed from plutonium material processing to preparation for decontamination and decommissioning (D and D). The PFP is in transition between its previous mission and the proposed D and D mission. The objective of the transition is to place the facility into a stable state for long-term storage of plutonium materials before final disposition of the facility. Accordingly, this update of the Final Safety Analysis Report (FSAR) reflects the current status of the buildings, equipment, and operations during this transition. The primary product of the PFP was plutonium metal in the form of 2.2-kg, cylindrical ingots called buttoms. Plutonium nitrate was one of several chemical compounds containing plutonium that were produced as an intermediate processing product. Plutonium recovery was performed at the Plutonium Reclamation Facility (PRF) and plutonium conversion (from a nitrate form to a metal form) was performed at the Remote Mechanical C (RMC) Line as the primary processes. Plutonium oxide was also produced at the Remote Mechanical A (RMA) Line. Plutonium processed at the PFP contained both weapons-grade and fuels-grade plutonium materials. The capability existed to process both weapons-grade and fuels-grade material through the PRF and only weapons-grade material through the RMC Line although fuels-grade material was processed through the line before 1984. Amounts of these materials exist in storage throughout the facility in various residual forms left from previous years of operations.

Plutonium in uranium deposits

International Nuclear Information System (INIS)

Curtis, D.; Fabryka-Martin, J.; Aguilar, R.; Attrep, M. Jr.; Roensch, F.

1992-01-01

Plutonium-239 (t 1/2 , 24,100 yr) is one of the most persistent radioactive constituents of high-level wastes from nuclear fission power reactors. Effective containment of such a long-lived constituent will rely heavily upon its containment by the geologic environment of a repository. Uranium ore deposits offer a means to evaluate the geochemical properties of plutonium under natural conditions. In this paper, analyses of natural plutonium in several ores are compared to calculated plutonium production rates in order to evaluate the degree of retention of plutonium by the ore. The authors find that current methods for estimating production rates are neither sufficiently accurate nor precise to provide unambiguous measures of plutonium retention. However, alternative methods for evaluating plutonium mobility are being investigated, including its measurement in natural ground waters. Preliminary results are reported and establish the foundation for a comprehensive characterization of plutonium geochemistry in other natural environments

Plan for the Initiation of HA-211 Furnace Operations at the Plutonium Finishing Plan (PFP)

International Nuclear Information System (INIS)

WILLIS, H.T.

2000-01-01

This plan provides a phased approach authorizing the use of three additional muffle furnaces for thermal stabilization. Achievement of Thermal Stabilization mission elements require the installation and startup of three additional muffle furnaces for the thermal stabilization of plutonium and plutonium bearing materials at the Plutonium Finishing Plant (PFP). The release to operate these additional furnaces will require an Activity Based Startup Review. The conduct of the Activity Based Startup Review (ABSR) was approved by Fluor Daniel Hanford on October 15, 1999. This plan has been developed with the objective of identifying those activities needed to guide the controlled startup of five furnaces from authorization to unrestricted operations by adding the HA-211 furnaces in an orderly and safe manner after the approval to Startup has been given

Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

International Nuclear Information System (INIS)

Permana, Sidik; Suzuki, Mitsutoshi; Saito, Masaki; Novitrian,; Waris, Abdul; Suud, Zaki

2013-01-01

Highlights: • The paper analyzes the plutonium production of recycling nuclear fuel option. • To evaluate material attractiveness based on intrinsic feature of material barrier. • Evaluation based on isotopic plutonium composition of spent fuel LWR and FBR. • Even mass number of plutonium gives a significant contribution to material barrier, in particular Pu-238 and Pu-240. • Doping MA in FBR blanket is effective to increase material barrier from weapon grade plutonium to more than MOX fuel grade. - Abstract: Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no

Direct oxide reduction (DOR) solvent salt recycle in pyrochemical plutonium recovery operations

International Nuclear Information System (INIS)

Fife, K.W.; Bowersox, D.F.; Davis, C.C.; McCormick, E.D.

1987-02-01

One method used at Los Alamos for producing plutonium metal is to reduce the oxide with calcium metal in molten CaCl 2 at 850 0 C. The solvent CaCl 2 from this reduction step is currently discarded as low-level radioactive waste because it is saturated with the reaction by-product, CaO. We have developed and demonstrated a molten salt technique for rechlorinating the CaO, thereby regenerating the CaCl 2 and incorporating solvent recycle into the batch PuO 2 reduction process. We discuss results from the process development experiments and present our plans for incorporating the technique into an advanced design for semicontinuous plutonium metal production

PFP vault operations containers for Plutonium Handling and Storage Critical Characteristics

International Nuclear Information System (INIS)

BONADIE, E.P.

2000-01-01

This document specifies the critical characteristics for containers procured for Plutonium Finishing Plant's (PFP's) Vault Operations system as required by HNF-PRO-268 and HNF-PRO-1819. These are the minimum specifications that the equipment must meet in order to perform its safety function

Improving Efficiency with 3-D Imaging: Technology Essential in Removing Plutonium Processing Equipment from Plutonium Finishing Plant Gloveboxes

International Nuclear Information System (INIS)

Crow, Stephen H.; Kyle, Richard N.; Minette, Michael J.

2008-01-01

The Plutonium Finishing Plant at Hanford, Washington began operations in 1949 to process plutonium and plutonium products. Its primary mission was to produce plutonium metal, fabricate weapons parts, and stabilize reactive materials. These operations, and subsequent activities, were performed in remote production lines, consisting primarily of hundreds of gloveboxes. Over the years these gloveboxes and processes have been continuously modified. The plant is currently inactive and Fluor Hanford has been tasked to clean out contaminated equipment and gloveboxes from the facility so it can be demolished in the near future. Approximately 100 gloveboxes at PFP have been cleaned out in the past four years and about 90 gloveboxes remain to be cleaned out. Because specific commitment dates for this work have been established with the State of Washington and other entities, it is important to adopt work practices that increase the safety and speed of this effort. The most recent work practice to be adopted by Fluor Hanford D and D workers is the use of 3-D models to improve the efficiency of cleaning out radioactive gloveboxes at the plant. The use of 3-D models has significantly improved the work planning process by providing workers with a clear image of glovebox construction and composition, which is then used to determine cleanout methods and work sequences. The 3-D visual products enhance safety by enabling workers to more easily identify hazards and implement controls. In addition, the ability to identify and target the removal of radiological materials early in the D and D process provides substantial dose reduction for the workers

A solvent proceed for the extraction of the irradiate uranium and plutonium in the reactor core; Un procede par solvant pour l'extraction du plutonium de l'uranium irradie dans les piles

Energy Technology Data Exchange (ETDEWEB)

Goldschmidt, B; Regnaut, P; Prevot, I [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

Description of the conditions of plutonium, fission products and of uranium separation by selective extraction of the nitrates by organic solvent, containing a simultaneous extraction of plutonium and uranium, followed by a plutonium re-extraction after reduction, and an uranium re-extraction. The rates of decontamination being insufficient in this first stage, we also describes the processes of decontamination permitting separately to get the rates wanted for uranium and plutonium. Finally, we describes the beginning of the operation that consists in a nitric dissolution of the active uranium while capturing the products of gaseous fission, as well as the final concentration of the products of fission in a concentrated solution. (authors) [French] Description des conditions de separation du plutonium, des produits de fission et de l'uranium au moyen d'une extraction selective des nitrates par solvant organique, comprenant une extraction simultanee du plutonium et de l'uranium, suivie d'une reextraction du plutonium apres reduction, et d'une reextraction de l'uranium. Les taux de decontamination etant insuffisants dans ce premier stade, on decrit egalement les processus de decontamination permettant separement d'obtenir les taux desires pour l'uranium et le plutonium. Enfin, on decrit aussi le debut de l'operation qui consiste en une dissolution nitrique de l'uranium actif en captant les produits de fission gazeux, ainsi que la concentration finale des produits de fission sous forme de solution concentree. (auteurs)

Plutonium contamination in soils and sediments at Mayak PA, Russia.

Science.gov (United States)

Skipperud, Lindis; Salbu, Brit; Oughton, Deborah H; Drozcho, Eugeny; Mokrov, Yuri; Strand, Per

2005-09-01

The Mayak Production Association (Mayak PA) was established in the late 1940's to produce plutonium for the Soviet Nuclear Weapons Programme. In total, seven reactors and two reprocessing plants have been in operation. Today, the area comprises both military and civilian reactors as well as reprocessing and metallurgical plants. Authorized and accidental releases of radioactive waste have caused severe contamination to the surrounding areas. In the present study, [alpha]-spectrometry and inductively coupled plasma-mass spectrometry (ICP-MS) have been used to determine plutonium activities and isotope ratios in soil and sediment samples collected from reservoirs of the Techa River at the Mayak area and downstream Techa River. The objective of the study was to determine the total inventory of plutonium in the reservoirs and to identify the different sources contributing to the plutonium contamination. Results based on [alpha]-spectrometry and ICP-MS measurements show the presence of different sources and confirmed recent reports of civilian reprocessing at Mayak. Determination of activity levels and isotope ratios in soil and sediment samples from the Techa River support the hypothesis that most of the plutonium, like other radionuclides in the Techa River, originated from the very early waste discharges to the Techa River between 1949 and 1951. Analysis of reservoir sediment samples suggest that about 75% of the plutonium isotopes could have been released to Reservoir 10 during the early weapons production operation of the plant, and that the majority of plutonium in Reservoir 10 originates from discharges from power production or reprocessing. Enhanced 240Pu/239Pu atom ratios in river sediment upper layers (0-2 cm) between 50 and 250 km downstream from the plant indicate a contribution from other, non-fallout sources.

Characteriztion of particulate plutonium released in fuel cycle operations

International Nuclear Information System (INIS)

Seefeldt, W.B.; Mecham, W.J.; Steindler, M.J.

1976-05-01

An estimate of the plutonium source terms is made for the fuel cycles of three reactor types on the basis of currently applied, currently available, and estimated future technology. The three reactors are LWR-U, LWR-Pu, and LMFBR. The source terms are characterized as to quantity, form, and particle size distribution. Historical operating data for existing plants and the state of the art of the technology of air cleaning are reviewed

Plutonium-236 traces determination in plutonium-238 by α spectrometry

International Nuclear Information System (INIS)

Acena, M.L.; Pottier, R.; Berger, R.

1969-01-01

Two methods are described in this report for the determination of plutonium-236 traces in plutonium-238 by a spectrometry using semi-conductor detectors. The first method involves a direct comparison of the areas under the peaks of the α spectra of plutonium-236 and plutonium-238. The electrolytic preparation of the sources is carried out after preliminary purification of the plutonium. The second method makes it possible to determine the 236 Pu/ 238 Pu ratio by comparing the areas of the α peaks of uranium-232 and uranium-234, which are the decay products of the two plutonium isotopes respectively. The uranium in the source, also deposited by electrolysis, is separated from a 1 mg amount of plutonium either by a T.L.A. extraction, or by the use of ion-exchange resins. The report ends with a discussion of the results obtained with plutonium of two different origins. (authors) [fr

Trans-Uranium Doping Utilization for Increasing Protected Plutonium Proliferation of Small Long Life Reactor

Energy Technology Data Exchange (ETDEWEB)

Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suud, Zaki [Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suzuki, Mitsutoshi [Japan Atomic Energy Agency, Nuclear Non-proliferation Science and Technology Center, 2-4 Shirane Shirakata, Tokai-mura, Ibaraki, 319-1195 (Japan)

2009-06-15

for different power output and core operation time. The two dimensional core optimization analysis (SRAC-CITATION) has been used with nuclear data library (JENDL-33). Higher plutonium production rate of Pu-238 can be achieved by Neptunium (NP)-237 doping compared with Minor Actinide (MA) doping. However, higher plutonium production rate of Pu-240 can be obtained by MA doping. Pu-238 production decreases and Pu-240 production increases with increasing core operation time for both MA and NP-237 doping materials. In case of different doping rate of MA, it shows Pu-238 and Pu-240 productions are strongly depends on the content of MA doping and different core operation time. Finally, some plutonium characterizations will be shown based on several basic criteria which have been adopted in this study. (authors)

The plutonium society

International Nuclear Information System (INIS)

Mez, L.; Richter, M.

1981-01-01

The lectures of an institute are reported on, which took place between 25th and 27th January 1980 in Berlin. The subsequent public panel discussion with representations from the political parties is then documentated in a few press-reports. The themes of the 8 lectures are: views and facts on plutonium, plutonium as an energy resource, military aspects of the production of plutonium, economic aspects of the plutonium economy, the position of the trade unions on the industrial reconversion, the alleged inevitability of a plutonium society and the socio-political alternatives and perspectives of nuclear waste disposal. (UA) [de

Light water breeder reactor using a uranium-plutonium cycle

International Nuclear Information System (INIS)

Radkowsky, A.; Chen, R.

1990-01-01

This patent describes a light water receptor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: a prebreeder section having plutonium fuel containing a Pu-241 component, the prebreeder section being operable to produce enriched plutonium having an increased Pu-241 component; and a breeder section for receiving the enriched plutonium from the prebreeder section, the breeder section being operable for breeding fissile material from the enriched plutonium fuel. This patent describes a method of operating a light water nuclear reactor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: operating the prebreeder to produce enriched plutonium fuel having an increased Pu-241 component; fueling a breeder section with the enriched plutonium fuel to breed the fissile material

SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

Science.gov (United States)

Spence, R.; Lister, M.W.

1958-12-16

Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

Plutonium Finishing Plant

Data.gov (United States)

Federal Laboratory Consortium — The Plutonium Finishing Plant, also known as PFP, represented the end of the line (the final procedure) associated with plutonium production at Hanford.PFP was also...

Systems work for Plutonium Fuel Production Facility (PFPF) near-real-time accounting

International Nuclear Information System (INIS)

Picard, R.R.; Hafer, J.F.; Pillay, K.K.S.; Takahashi, S.; Ohtani, T.; Eguchi, K.; Seya, M.

1990-01-01

A joint effort by the Los Alamos National Laboratory and the Power Reactor and Nuclear Fuel Development Corporation of Japan examines materials accounting for the Plutonium Fuel Production Facility. A unique feature of the systems work is a sophisticated data generator. This software follows individual items throughout the process, creating detailed data files for variance propagation. The data generator deals with user-specified process operations and handles related accounting problems, such as the tracking of individual measurements through numerous blending and splitting procedure, frequent decay correction (important for large inventories), scrap recovery, and automated determination of static inventory. There is no need to rely on simplified assumptions regarding process operation and material measurement. Also, the joint study applies recent theoretical work on stratified inspection of nonhomogeneous inventories and sequential analysis of MUF -- D. 4 refs

Expert system for estimating LWR plutonium production

International Nuclear Information System (INIS)

Sandquist, G.M.

1988-01-01

An Artificial Intelligence-Expert System called APES (Analysis of Proliferation by Expert System) has been developed and tested to permit a non proliferation expert to evaluate the capability and capacity of a specified LWR reactor and PUREX reprocessing system for producing and separating plutonium even when system information may be limited and uncertain. APES employs an expert system coded in LISP and based upon an HP-RL (Hewlett Packard-Representational Language) Expert System Shell. The user I/O interface communicates with a blackboard and the knowledge base which contains the quantitative models required to describe the reactor, selected fission product production and radioactive decay processes, Purex reprocessing and ancillary knowledge

Plutonium

International Nuclear Information System (INIS)

Anon.

1995-01-01

Plutonium, which was obtained and identified for the first time in 1941 by chemist Glenn Seaborg - through neutron irradiation of uranium 238 - is closely related to the history of nuclear energy. From the very beginning, because of the high radiotoxicity of plutonium, a tremendous amount of research work has been devoted to the study of the biological effects and the consequences on the environment. It can be said that plutonium is presently one of the elements, whose nuclear and physico-chemical characteristics are the best known. The first part of this issue is a survey of the knowledge acquired on the subject, which emphasizes the sanitary effects and transfer into the environment. Then the properties of plutonium related to energy generation are dealt with. Fissionable, like uranium 235, plutonium has proved a high-performance nuclear fuel. Originally used in breeder reactors, it is now being more and more widely recycled in light water reactors, in MOX fuel. Reprocessing, recycling and manufacturing of these new types of fuel, bound of become more and more widespread, are now part of a self-consistent series of operations, whose technical, economical, industrial and strategical aspects are reviewed. (author)

Inherent protection of plutonium by doping minor actinide in thermal neutron spectra

International Nuclear Information System (INIS)

Peryoga, Yoga; Sagara, Hiroshi; Saito, Masaki; Ezoubtchenko, Alexey

2005-01-01

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% 235 U and 20% 235 U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, 238 Pu, 240 Pu and 242 Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% 235 U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties. (author)

The production control laboratories of the plutonium extraction Plant at Marcoule. Six years operating experience: 1957 - 1963

International Nuclear Information System (INIS)

Fontaine, A.

1964-01-01

In this paper, the author attempts to sum up the conditions prevailing, after six years of operation, in the Laboratories of the Plutonium Extraction Plant. The origins and objectives are briefly reviewed, the technology and staff recruitment policy are examined, and progress made is shown. The methods used as well as the scope of application and limits imposed at the present state are considered. Past achievements and further possibilities in the next future are examined. An attempt has been made to bring out the outlooks for the more distant future and to investigate the conditions required for the successful carrying out of the program. (author) [fr

Management of Russian military plutonium

International Nuclear Information System (INIS)

Zaleski, C.P.

1996-01-01

The objective of this paper is to propose and discuss a solution which enables storing as quickly as possible all weapons-grade plutonium from Russian military program in a way which would prevent diversion. Two main conditions apply to this solution. First, it should be achieved in a manner acceptable to Russian government, notably by preserving plutonium for possible future energy production, and second, the economics of the total system should be good enough to ensure no charge or limited charge for the storage of plutonium. A proposal is made to store plutonium in a specially designed fast reactor or specially designed reactor core. This solution could be favorable in comparison to other solutions applying the above mentioned goal and conditions. Additionally the proposed solution would have the following side advantages: utilizing available personnel and installations of the Russian nuclear complex; providing possible basis for decommissioning of older and less safe Russian reactors; giving experience of construction and operation of a series of sodium-cooled fast reactors. The major problem however is the need for large capital investment with the risk of getting no adequate return on investment due to difficult political and economic situation in Russia

The production and certification of a plutonium equal-atom reference material: NBL CRM 128

International Nuclear Information System (INIS)

Crawford, D.W.

1990-07-01

This report describes the design, production, and certification of the New Brunswick Laboratory plutonium equal-atom certified reference material (CRM), NBL CRM 128. The primary use of this CRM is for the determination of bias corrections encountered in the operation of a mass spectrometer. This reference material is available to the US Department of Energy contractor-operated and government-operated laboratories, as well as to the international nuclear safeguards community. The absolute, or unbiased, certified value for the CRM's Pu-242/Pu-239 ratio is 1.00063 ± 0.00026 (95% confidence interval) as of October 1, 1984. This value was obtained through the quantitative blending of high-purity, chemically and isotopically characterized separated isotopes, as well as through intercomparisons of CRM samples with calibration mixtures using thermal ionization mass spectrometry. 32 tabs

Plutonium metal burning facility

International Nuclear Information System (INIS)

Hausburg, D.E.; Leebl, R.G.

1977-01-01

A glove-box facility was designed to convert plutonium skull metal or unburned oxide to an oxide acceptable for plutonium recovery and purification. A discussion of the operation, safety aspects, and electrical schematics are included

Ventilation design modifications at Los Alamos Scientific Laboratory major plutonium operational areas

International Nuclear Information System (INIS)

Stafford, R.G.; Gallimore, J.C. Jr.; Mitchell, R.N.; Maraman, W.J.; McNeese, W.D.

1975-01-01

Major ventilation design modifications in plutonium operational areas at Los Alamos have occurred during the past two years. An additional stage of HEPA filters has been added to DP West glove-box process exhaust resulting in significant effluent reductions. The additional stage of HEPA filters is unique in that each filter may be individually DOP tested. Radiological filter efficiencies of each process exhaust stage is presented. DP West room air ventilation systems have been modified to incorporate a single stage of HEPA filters in contrast to a previous American Air Filter PL-24 filtration system. Plutonium effluent reductions of 10 2 to 10 3 have resulted in these new systems. Modified DOP testing procedures for room air filtration systems are discussed. Major plutonium areas of the CMR Building utilizing Aerosolve 95 process exhaust filtration systems have been upgraded with two stages of HEPA filters. Significant reductions in effluent are evident. A unique method of DOP testing each bank of HEPA filters is discussed. Radiological efficiencies of both single and two-stage filters are discussed. (U.S.)

Procedure for hazards analysis of plutonium gloveboxes used in analytical chemistry operations

International Nuclear Information System (INIS)

Delvin, W.L.

1977-06-01

A procedure is presented to identify and assess hazards associated with gloveboxes used for analytical chemistry operations involving plutonium. This procedure is based upon analytic tree methodology and it has been adapted from the US Energy Research and Development Administration's safety program, the Management Oversight and Risk Tree

The plutonium challenge for the future

International Nuclear Information System (INIS)

Gray, L.W.

2000-01-01

In this paper author deal with the weapons-usable plutonium and with the possibilities of their managing. Russia has not disclosed the amount of plutonium produced, but various estimates indicate that the production was about 130 tonnes. Production has been curtailed in Russia; three dual-purpose reactors still produce weapons-grade plutonium - two at Tomsk-7 (renamed Seversk) and one at Krasnoyarsk-26 (renamed Zheleznogorsk Mining and Chemical Combine). In a 1994 United States-Russian agreement that has yet to enter into force, Russia agreed to close the remaining operating reactors by the year 2000. Treaties between the United States and Russia have already cut the number of nuclear warheads from more than 10,000 to about 6,000 under START 1, which has been ratified, and to about 3,500 under START 2, which still awaits approval. If Russia and the United States conclude START 3, that number could drop to between 2,000 and 2,500. On September 2, 1998, the Presidents of the United States and Russia signed the 'Joint statement of principles for Management and Disposition of Plutonium, Designated as No Longer Required for Defense Purposes.' In this joint statement the Presidents affirm the intention of each country to remove by stages approximately 50 metric tons of plutonium and to convert the nuclear weapons programs, and to convert this material so that it can never be used in nuclear weapons. These 100 tonne of plutonium must be managed in proper way such that it becomes neither a proliferation for an environmental risk. The United States has proposed that it manage it's 50 tonnes by a dual approach-once through MOX burning of a portion of the plutonium and immobilization in a ceramic matrix followed by en- casement in high level waste glass. Russia has proposed that it manage its full 50 tonnes by burning in a reactor. The MOX program in the United States would bum the cleaner plutonium metal and residues. Weapons components would be converted to plutonium oxide

The use of plutonium

International Nuclear Information System (INIS)

Marshall, W.

1980-01-01

The use of plutonium as a vital energy source producing maximum economic benefit with minimum proliferation risks is discussed. Having considered the production of plutonium, several possible plutonium fuel cycle options are identified and the economic value to be attached to plutonium for each examined. It is shown how the use of plutonium in fast reactors gives an opportunity for a non-proliferation policy not available when plutonium is used only in thermal reactors. From the technical considerations reviewed concerning plutonium and fast reactors it is shown that an economic regime involving international trade in spent thermal reactor fuel is possible which benefits equally those countries with fast reactors and those without and also assists in avoiding the proliferation of nuclear weapons. (U.K.)

Plutonium solution storage in plastic bottles: Operational experience and safety issues

International Nuclear Information System (INIS)

Conner, W.V.

1995-01-01

Computer spread sheet models were developed to gain a better understanding of the factors that lead to pressurization and failure of plastic bottles containing plutonium solutions. These models were developed using data obtained from the literature on gas generation rates for plutonium solutions. Leak rates from sealed plastic bottles were obtained from bottle leak tests conducted at Rocky Flats. Results from these bottle leak tests showed that narrow mouth four liter bottles will seal much better than wide mouth four liter bottles. The gas generation rate and leak rate data were used to develop models for predicting the rate of pressurization and maximum pressures expected in sealed bottles of plutonium solution containing various plutonium and acid concentrations. The computer models were used to develop proposed time limits for storing or transporting plutonium solutions in sealed plastic bottles. For plutonium solutions containing 1.5 g/l plutonium, storage in sealed bottles should not be allowed. However, transportation of higher concentration plutonium solution in sealed bottles is required, and safe transportation times of 1 shift to 6 days are proposed

Determination of uranium and plutonium in metal conversion products from electrolytic reduction process

International Nuclear Information System (INIS)

Lee, Chang Heon; Suh, Moo Yul; Joe, Kih Soo; Sohn, Se Chul; Jee, Kwang Young; Kim, Won Ho

2005-01-01

Chemical characterization of process materials is required for the optimization of an electrolytic reduction process in which uranium dioxide, a matrix of spent PWR fuels, is electrolytically reduced to uranium metal in a medium of LiCl-Li 2 O molten at 650 .deg. C. A study on the determination of fissile materials in the uranium metal products containing corrosion products, fission products and residual process materials has been performed by controlled-potential coulometric titration which is well known in the field of nuclear science and technology. Interference of Fe, Ni, Cr and Mg (corrosion products), Nd (fission product) and LiCl molten salt (residual process material) on the determination of uranium and plutonium, and the necessity of plutonium separation prior to the titration are discussed in detail. Under the analytical condition established already, their recovery yields are evaluated along with analytical reliability

Simulation of uranium and plutonium oxides compounds obtained in plasma

Science.gov (United States)

Novoselov, Ivan Yu.; Karengin, Alexander G.; Babaev, Renat G.

2018-03-01

The aim of this paper is to carry out thermodynamic simulation of mixed plutonium and uranium oxides compounds obtained after plasma treatment of plutonium and uranium nitrates and to determine optimal water-salt-organic mixture composition as well as conditions for their plasma treatment (temperature, air mass fraction). Authors conclude that it needs to complete the treatment of nitric solutions in form of water-salt-organic mixtures to guarantee energy saving obtainment of oxide compounds for mixed-oxide fuel and explain the choice of chemical composition of water-salt-organic mixture. It has been confirmed that temperature of 1200 °C is optimal to practice the process. Authors have demonstrated that condensed products after plasma treatment of water-salt-organic mixture contains targeted products (uranium and plutonium oxides) and gaseous products are environmental friendly. In conclusion basic operational modes for practicing the process are showed.

Plutonium storage phenomenology

International Nuclear Information System (INIS)

Szempruch, R.

1995-12-01

Plutonium has been produced, handled, and stored at Department of Energy (DOE) facilities since the 1940s. Many changes have occurred during the last 40 years in the sources, production demands, and end uses of plutonium. These have resulted in corresponding changes in the isotopic composition as well as the chemical and physical forms of the processed and stored plutonium. Thousands of ordinary food pack tin cans have been used successfully for many years to handle and store plutonium. Other containers have been used with equal success. This paper addressees the exceptions to this satisfactory experience. To aid in understanding the challenges of handling plutonium for storage or immobilization the lessons learned from past storage experience and the necessary countermeasures to improve storage performance are discussed

Properties of plutonium

International Nuclear Information System (INIS)

Ahn, Jin Su; Yoon, Hwan Ki; Min, Kyung Sik; Kim, Hyun Tae; Ahn, Jong Sung; Kwag, Eon Ho; Ryu, Keon Joong

1996-03-01

Plutonium has unique chemical and physical properties. Its uniqueness in use has led to rare publications, in Korea. This report covers physical aspects of phase change of metal plutonium, mechanical properties, thermal conductivity, etc, chemical aspects of corrosion, oxidation, how to produce plutonium from spent fuels by describing various chemical treatment methods, which are currently used and were used in the past. It also contains characteristics of the purex reprocessing process which is the most widely used nowadays. And show processes to purify and metalize from recovered plutonium solution. Detection and analysis methods are introduced with key pints for handling, critical safety, toxicity, and effects on peoples. This report gives not only a general idea on what plutonium is, rather than deep technical description, but also basic knowledge on plutonium production and safeguards diversion from the view point of nonproliferation. 18 refs. (Author) .new

Properties of plutonium

Energy Technology Data Exchange (ETDEWEB)

Ahn, Jin Su; Yoon, Hwan Ki; Min, Kyung Sik; Kim, Hyun Tae; Ahn, Jong Sung; Kwag, Eon Ho; Ryu, Keon Joong [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of)

1996-03-01

Plutonium has unique chemical and physical properties. Its uniqueness in use has led to rare publications, in Korea. This report covers physical aspects of phase change of metal plutonium, mechanical properties, thermal conductivity, etc, chemical aspects of corrosion, oxidation, how to produce plutonium from spent fuels by describing various chemical treatment methods, which are currently used and were used in the past. It also contains characteristics of the purex reprocessing process which is the most widely used nowadays. And show processes to purify and metalize from recovered plutonium solution. Detection and analysis methods are introduced with key pints for handling, critical safety, toxicity, and effects on peoples. This report gives not only a general idea on what plutonium is, rather than deep technical description, but also basic knowledge on plutonium production and safeguards diversion from the view point of nonproliferation. 18 refs. (Author) .new.

The plutonium fuel cycles

International Nuclear Information System (INIS)

Pigford, T.H.; Ang, K.P.

1975-01-01

The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

Decontamination of Battelle-Columbus' Plutonium Facility. Final report

International Nuclear Information System (INIS)

Rudolph, A.; Kirsch, G.; Toy, H.L.

1984-01-01

The Plutonium Laboratory, owned and operated by Battelle Memorial Institute's Columbus Division, was located in Battelle's Nuclear Sciences area near West Jefferson, Ohio, approximately 17 miles west of Columbus, Ohio. Originally built in 1960 for plutonium research and processing, the Plutonium Laboratory was enlarged in 1964 and again in 1967. With the termination of the Advanced Fuel Program in March, 1977, the decision was made to decommission the Plutonium Laboratory and to decontaminate the building for unrestricted use. Decontamination procedures began in January, 1978. All items which had come into contact with radioactivity from the plutonium operations were cleaned or disposed of through prescribed channels, maintaining procedures to ensure that D and D operations would pose no risk to the public, the environment, or the workers. The entire program was conducted under the cognizance of DOE's Chicago Operations Office. The building which housed the Plutonium Laboratory has now been decontaminated to levels allowing it to house ordinary laboratory and office operations. A ''Finding of No Significant Impact'' (FNSI) was issued in May, 1980

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity [O1] - Fundamentals of P{sup 3} Mechanism and Methodology Development for Plutonium Categorization

Energy Technology Data Exchange (ETDEWEB)

Saito, Masaki [Research Laboratory for Nuclear Reactor, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro, Tokyo, 1528550 (Japan)

2009-06-15

'Protected Plutonium Production (P{sup 3})' has been proposed to enhance the proliferation resistance of plutonium by the transmutation of Minor Actinides (MAs). Doping the small amount of MAs such as {sup 237}Np or {sup 241}Am with large neutron capture cross-section into the uranium fuel to enhance the production of {sup 238}Pu or {sup 242}Pu, which have high spontaneous fission neutron source or also high decay heat to makes the process of the nuclear weapon manufacture and maintenance technologically difficult, can be effective for improving the isotopic barrier of proliferation resistance of the plutonium in thermal reactors. Super weapon grade plutonium could be produced in the blanket of a conventional FBR. However, by increasing the {sup 238}Pu or {sup 242}Pu ratio in the total plutonium by MAs doping into the fresh blanket, the protected plutonium with high proliferation-resistance can be bred. A new evaluation function, 'attractiveness', defined as a ratio of potential of fission yield to the technological difficulties of nuclear explosive device, has been proposed to evaluate the proliferation resistance of Pu based on the nuclear material property for Plutonium Categorization. In the conference, the fundamentals of P{sup 3} mechanism by transmutation of MA, and the comparison of the 'attractiveness' of the Pu produced in advanced reactors based on P{sup 3} mechanism and in the conventional reactors will be presented. Instead of the geological disposal or just their burning of MAs by the fission reaction, they should be treated as valuable fertile materials to enhance the proliferation resistance of plutonium produced in the thermal and fast breeder reactors for peace and sustainable prosperity in future. Acknowledgement: Some parts of this work have been supported by the Ministry of Education, Culture, Sports, Science and Technology in Japan. (authors)

The use of plutonium rapides surregenerateurs aspects techniques et economiques

International Nuclear Information System (INIS)

Guillet, H.; Delayre, R.; Mougniot, J.C.; Ferrari, A.

1977-01-01

Nuclear energy production utilizing U 235 and U 238 inevitably results in the formation of plutonium. Some of this is directly used by the reactor in power production. Some reactors, e.g. the Candu type, burn most of their plutonium ''in situ''. However the surplus quantity of plutonium produced is increasing, and by 1990 the world stock of plutonium is predicted to be about 1000 tons (300 tons in Europe and 400 in the USA). This represents approximately 0.1 Q of potential power, where Q=10 21 joules. Proposals for dealing with this plutonium include its storage, either as irradiated fuel or as a refined substance (plutonium nitrate or oxide); its use in thermal reactors as a substitute for U 235 ; its use in fast breeder reactors which can act as consumers of plutonium as well as producers and which can therefore regulate the world's plutonium stocks. Leaving aside the question of storage, certain technical conditions must be fulfilled. It is essential that reprocessing plants should be operational and available in adequate numbers. An industry able to make use of plutonium should be developed with large capacity units. There seem to be no problems with plutonium use in reactors: conclusive experiments have proved the accuracy of calculations regarding plutonium recycling in thermal reactors and a number of fast breeders have proved the possibility of using plutonium in them. Experience acquired during the past ten years permits us to deal with the safety and safeguards problem of using plutonium. Many economic questions remain, however, including questions affecting the utilization of plutonium, such as the cost of regenerating irradiated fuel and of making fuel, and questions affecting the strategy of using plutonium in different ways, such as the price at which it should be sold. Also the cost of producing electricity using plutonium is not yet clear. It is unavoidable that plutonium will be used in nuclear power development. The technical and economic problems

Prospects for the Use of Plutonium in Reactors; Prospective d'Utilisation du Plutonium dans les Reacteurs

Energy Technology Data Exchange (ETDEWEB)

Fossoul, E.; Haubert, P. [BELGONUCLEAIRE (Belgium); Hirschberg, D.; Morlet, E. [International Business Machines of Belgium, Bruxelles (Belgium)

1967-09-15

The introduction, at an increasing rate, of power reactors using slightly enriched uranium will inevitably lead to the production of considerable quantities of plutonium over the next decade. Fast reactors will not be capable of absorbing this material before 1980. The question thus arises of whether one should store the plutonium far future use in fast reactors, recycle it in existing thermal reactors, or try to sell it. The problem has been studied for an electric power generating system that does not foresee selling the plutonium produced by its reactors and does not buy plutonium outside, which enables a good approximation to be made and eliminates the major unknown quantity represented by the future market price of plutonium. Assuming within this system a programme that provides for the construction of power reactors of a given type and capacity at specific dates, the utilization of the plutonium produced can be optimized by linear programming techniques so as to minimize the discounted total cost of the power generated over a given period. A later stage consists in optimizing, by various techniques, not only the utilization but also the production of plutonium by appropriate selection of the power reactor types to be constructed. (author) [French] L'implantation, a un rythme croissant, de centrales nucleaires a uranium legerement enrichi entrainera la production ineluctable d'une quantite importante de plutonium au cours de la prochaine decennie. Les reacteurs a neutrons rapides ne seront capables d'absorber cette production qu'apres 1980. La question se pose donc de savoir s'il est preferable de stocker le plutonium en vue de son utilisation ulterieure dans les reacteurs a neutrons rapides plutot que de le recycler dans les reacteurs actuels a neutrons thermiques ou d'essayer de le vendre. Ce probleme a ete etudie dans le cadre d'un systeme de production d'energie electrique qui ne prevoirait pas la vente du plutonium produit par ses reacteurs nucleaires ni

History and stabilization of the Plutonium Finishing Plant (PFP) complex, Hanford Site

Energy Technology Data Exchange (ETDEWEB)

Gerber, M.S., Fluor Daniel Hanford

1997-02-18

The 231-Z Isolation Building or Plutonium Metallurgy Building is located in the Hanford Site`s 200 West Area, approximately 300 yards north of the Plutonium Finishing Plant (PFP) (234-5 Building). When the Hanford Engineer Works (HEW) built it in 1944 to contain the final step for processing plutonium, it was called the Isolation Building. At that time, HEW used a bismuth phosphate radiochemical separations process to make `AT solution,` which was then dried and shipped to Los Alamos, New Mexico. (AT solution is a code name used during World War II for the final HEW product.) The process was carried out first in T Plant and the 224-T Bulk Reduction Building and B Plant and the 224-B Bulk Reduction Building. The 224-T and -B processes produced a concentrated plutonium nitrate stream, which then was sent in 8-gallon batches to the 231-Z Building for final purification. In the 231-Z Building, the plutonium nitrate solution underwent peroxide `strikes` (additions of hydrogen peroxide to further separate the plutonium from its carrier solutions), to form the AT solution. The AT solution was dried and shipped to the Los Alamos Site, where it was made into metallic plutonium and then into weapons hemispheres.` The 231-Z Building began `hot` operations (operations using radioactive materials) with regular runs of plutonium nitrate on January 16, 1945.

Analytic determination of plutonium in the environment

International Nuclear Information System (INIS)

Ballada, J.

1967-01-01

The work described in this report was undertaken with a view to determining the plutonium content in the fall-out from nuclear explosions. In the first part are described in turn the importance of the problems due to the plutonium, the physico-chemical properties of the radioelement and the biological dangers which it presents. A detailed and critical analysis is made of the radio-toxicological determination of the plutonium as reported in the literature prior to this report. The second part consists in the presentation of a judicious choice of techniques making it possible to determine plutonium in air, rain-water, soils and ash. After a detailed description of the measurement equipment and the operational techniques which have been developed, a justification of these techniques is given with particular reference to their sensitivity and specificity. After a brief conclusion concerning the preceding chapters, the results are presented. These are then discussed in the ease of each element in which the plutonium has been determined. This discussion is concluded by a consideration of the importance of the occurrence of fall-out plutonium on problems relating to public health. From a consideration of 200 analyses carried out, it is concluded that the contribution of plutonium to the exposure of populations is still very small compared to that of natural radiation and that due to such fission products as strontium 90. The report includes 63 literature references, 26 figures and 11 tables. (author) [fr

A performance indicator for reduction in vulnerability through stabilization of plutonium

International Nuclear Information System (INIS)

Marchese, A.R.; Neogy, P.; Azarm, M.A.

1997-01-01

The US Department of Energy (DOE) is currently storing several metric tons of plutonium in various forms in a variety of facilities throughout the DOE complex. Since the cessation of weapons production in 1990, many of these facilities with plutonium in storage have not operated. Since the shutdown was regarded as temporary, little attempt was made at that time to empty the process lines of plutonium, or to place the plutonium in containers or packages that would provide safe storage for extended periods of time. As a result, the packages and containers providing interim storage are vulnerable to failure through leakage, rupture and other modes, and pose potential hazards to facility workers, the public and the environment. Here, an approach to measuring and tracking the reduction in vulnerabilities resulting from stabilizing and repackaging plutonium is developed and presented. The approach utilizes results obtained by the DOE Working Group on the vulnerabilities associated with plutonium storage

Provision of NDA instrumentation for the control of operations on plutonium finishing and waste plants at the Sellafield nuclear fuel reprocessing facility

International Nuclear Information System (INIS)

Whitehouse, K.R.; Orr, C.H.

1995-01-01

On BNFL's Sellafield site a significant number of major plants are involved in the handling, processing and storage of plutonium in various forms including nitrate, oxide and mixed oxide (MOX). Other plants in operation or under construction treat and prepare for storage, plutonium bearing wastes in the form of plutonium contaminated materials -- PCM (transuranic waste -- TRU) or low level waste. Concurrently, a number of old plutonium handling plants are being decommissioned. The safety and cost effectiveness of these widely varying operations has been ensured by the development and installation of a wide range of special radiometric instrumentation. These systems based on a range of neutron counting and high resolution gamma spectrometric techniques -- singly or in combination -- enable BNFL to maintain a detailed and comprehensive picture of the disposition of plutonium within each plant and across the site. This paper describes an overview of the range of plant and paper prove waste measurement systems in this context, highlighting the specific roles of the Plutonium Inventory Measurement System (PIMS) for real time accountancy and the Decommissioning In-Situ Plutonium Inventory Monitor (DISPIM) for material control during decommissioning

Conceptual Design for the Pilot-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

Energy Technology Data Exchange (ETDEWEB)

Lumetta, Gregg J.; Meier, David E.; Tingey, Joel M.; Casella, Amanda J.; Delegard, Calvin H.; Edwards, Matthew K.; Jones, Susan A.; Rapko, Brian M.

2014-08-05

This report describes a conceptual design for a pilot-scale capability to produce plutonium oxide for use as exercise and reference materials, and for use in identifying and validating nuclear forensics signatures associated with plutonium production. This capability is referred to as the Pilot-scale Plutonium oxide Processing Unit (P3U), and it will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including plutonium dioxide (PuO2) dissolution, purification of the Pu by ion exchange, precipitation, and conversion to oxide by calcination.

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

International Nuclear Information System (INIS)

Chodak, P. III

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO 2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239 Pu and ≥90% total Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

Energy Technology Data Exchange (ETDEWEB)

Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

1996-05-01

This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO₂ assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the ²³⁹Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

Some plutonium IV polymers properties in Purex process

International Nuclear Information System (INIS)

Scoazec, H.; Pasquiou, J.Y.; Germain, M.

1990-01-01

The metabolism of plutonium polymers in fuel reprocessing using the Purex process with tributylphosphate as solvent, and its practical consequence in real operation conditions are examined. Precipitation with dibutylphosphoric acid, a solvent degradation product, occurs both in extraction and stripping units when polymers are present. (author)

IMPROVED PROCESS OF PLUTONIUM CARRIER PRECIPITATION

Science.gov (United States)

Faris, B.F.

1959-06-30

This patent relates to an improvement in the bismuth phosphate process for separating and recovering plutonium from neutron irradiated uranium, resulting in improved decontamination even without the use of scavenging precipitates in the by-product precipitation step and subsequently more complete recovery of the plutonium in the product precipitation step. This improvement is achieved by addition of fluomolybdic acid, or a water soluble fluomolybdate, such as the ammonium, sodium, or potassium salt thereof, to the aqueous nitric acid solution containing tetravalent plutonium ions and contaminating fission products, so as to establish a fluomolybdate ion concentration of about 0.05 M. The solution is then treated to form the bismuth phosphate plutonium carrying precipitate.

Plutonium Plant, Trombay

International Nuclear Information System (INIS)

Yadav, J.S.; Agarwal, K.

2017-01-01

The journey of Indian nuclear fuel reprocessing started with the commissioning of Plutonium Plant (PP) at Trombay on 22"n"d January, 1965 with an aim to reprocess the spent fuel from research reactor CIRUS. The basic process chosen for the plant was Plutonium Uranium Reduction EXtraction (PUREX) process. In seventies, the plant was subjected to major design modifications and replacement of hardware, which later met the additional demand from research reactor DHRUVA. The augmented plutonium plant has been operating since 1983. Experience gained from this plant was very much helpful to design future reprocessing plant in the country

What is plutonium stabilization, and what is safe storage of plutonium?

International Nuclear Information System (INIS)

Forsberg, C.W.

1995-01-01

The end of the cold war has resulted in the shutdown of nuclear weapons production and the start of dismantlement of significant numbers of nuclear weapons. This, in turn, is creating an inventory of plutonium requiring interim and long-term storage. A key question is, ''What is required for safe, multidecade, plutonium storage?'' The requirements for storage, in turn, define what is needed to stabilize the plutonium from its current condition into a form acceptable for interim and long-term storage. Storage requirements determine if research is required to (1) define required technical conditions for interim and long-term storage and (2) develop or improve current stabilization technologies. Storage requirements depend upon technical, policy, and economic factors. The technical issues are complicated by several factors. Plutonium in aerosol form is highly hazardous. Plutonium in water is hazardous. The plutonium inventory is in multiple chemical forms--some of which are chemically reactive. Also, some of the existing storage forms are clearly unsuitable for storage periods over a few years. Gas generation by plutonium compounds complicates storage: (1) all plutonium slowly decays creating gaseous helium and (2) the radiation from plutonium decay can initiate many chemical reactions-some of which generate significant quantities of gases. Gas generation can pressurize sealed storage packages. Last nuclear criticality must be avoided

Plutonium

Energy Technology Data Exchange (ETDEWEB)

Miner, William N

1964-01-01

This pamphlet discusses plutonium from discovery to its production, separation, properties, fabrication, handling, and uses, including use as a reactor fuel and use in isotope power generators and neutron sources.

Wastes from plutonium conversion and scrap recovery operations

Energy Technology Data Exchange (ETDEWEB)

Christensen, D.C.; Bowersox, D.F.; McKerley, B.J.; Nance, R.L.

1988-03-01

This report deals with the handling of defense-related wastes associated with plutonium processing. It first defines the different waste categories along with the techniques used to assess waste content. It then discusses the various treatment approaches used in recovering plutonium from scrap. Next, it addresses the various waste management approaches necessary to handle all wastes. Finally, there is a discussion of some future areas for processing with emphasis on waste reduction. 91 refs., 25 figs., 4 tabs.

Wastes from plutonium conversion and scrap recovery operations

International Nuclear Information System (INIS)

Christensen, D.C.; Bowersox, D.F.; McKerley, B.J.; Nance, R.L.

1988-03-01

This report deals with the handling of defense-related wastes associated with plutonium processing. It first defines the different waste categories along with the techniques used to assess waste content. It then discusses the various treatment approaches used in recovering plutonium from scrap. Next, it addresses the various waste management approaches necessary to handle all wastes. Finally, there is a discussion of some future areas for processing with emphasis on waste reduction. 91 refs., 25 figs., 4 tabs

Double shell tanks plutonium inventory assessment

International Nuclear Information System (INIS)

Tusler, L.A.

1995-01-01

This report provides an evaluation that establishes plutonium inventory estimates for all DSTs based on known tank history information, the DST plutonium inventory tracking system, tank characterization measurements, tank transfer records, and estimated average concentration values for the various types of waste. These estimates use data through December 31, 1994, and give plutonium estimates as of January 1, 1995. The plutonium inventory values for the DSTs are given in Section 31. The plutonium inventory estimate is 224 kg for the DSTs and 854 kg for the SSTs for a total of 1078 kg. This value compares favorably with the total plutonium inventory value of 981 kg obtained from the total plutonium production minus plutonium recovery analysis estimates

The differential radiological impact of plutonium recycle in the light-water reactor fuel cycle: effluent discharges during normal operation

International Nuclear Information System (INIS)

Bouville, A.; Guetat, P.; Jones, J.A.; Kelly, G.N.; Legrand, J.; White, I.F.

1980-01-01

The radiological impact of a light-water reactor fuel cycle utilizing enriched uranium fuel may be altered by the recycle of plutonium. Differences in impact may arise during various operations in the fuel cycle: those which arise from effluents discharged during normal operation of the various installations comprising the fuel cycle are evaluated in this study. The differential radiological impact on the population of the European Communities (EC) of effluents discharged during the recycling of 10 tonnes of fissile plutonium metal is evaluated. The contributions from each stage of the fuel cycle, i.e. fuel fabrication, reactor operation and fuel reprocessing and conversion, are identified. Separate consideration is given to airborne and liquid effluents and account is taken of a wide range of environmental conditions, representative of the EC, in estimating the radiological impact. The recycle of plutonium is estimated to result in a reduction in the radiological impact from effluents of about 30% of that when using enriched uranium fuel

HB-Line Plutonium Oxide Data Collection Strategy

Energy Technology Data Exchange (ETDEWEB)

Watkins, R. [Savannah River Nuclear Solutions; Varble, J. [Savannah River Nuclear Solutions; Jordan, J. [Savannah River Nuclear Solutions

2015-05-26

HB-Line and H-Canyon will handle and process plutonium material to produce plutonium oxide for feed to the Mixed Oxide Fuel Fabrication Facility (MFFF). However, the plutonium oxide product will not be transferred to the MFFF directly from HB-Line until it is packaged into a qualified DOE-STD-3013-2012 container. In the interim, HB-Line will load plutonium oxide into an inner, filtered can. The inner can will be placed in a filtered bag, which will be loaded into a filtered outer can. The outer can will be loaded into a certified 9975 with getter assembly in compliance with onsite transportation requirement, for subsequent storage and transfer to the K-Area Complex (KAC). After DOE-STD-3013-2012 container packaging capabilities are established, the product will be returned to HB-Line to be packaged into a qualified DOE-STD-3013-2012 container. To support the transfer of plutonium oxide to KAC and then eventually to MFFF, various material and packaging data will have to be collected and retained. In addition, data from initial HB-Line processing operations will be needed to support future DOE-STD-3013-2012 qualification as amended by the HB-Line DOE Standard equivalency. As production increases, the volume of data to collect will increase. The HB-Line data collected will be in the form of paper copies and electronic media. Paper copy data will, at a minimum, consist of facility procedures, nonconformance reports (NCRs), and DCS print outs. Electronic data will be in the form of Adobe portable document formats (PDFs). Collecting all the required data for each plutonium oxide can will be no small effort for HB-Line, and will become more challenging once the maximum annual oxide production throughput is achieved due to the sheer volume of data to be collected. The majority of the data collected will be in the form of facility procedures, DCS print outs, and laboratory results. To facilitate complete collection of this data, a traveler form will be developed which

Plutonium fires; Incendies de plutonium

Energy Technology Data Exchange (ETDEWEB)

Mestre, E.

1959-06-23

The author reports an information survey on accidents which occurred when handling plutonium. He first addresses accidents reported in documents. He indicates the circumstances and consequences of these accidents (explosion in glove boxes, fires of plutonium chips, plutonium fire followed by filter destruction, explosion during plutonium chip dissolution followed by chip fire). He describes hazards associated with plutonium fires: atmosphere and surface contamination, criticality. The author gives some advices to avoid plutonium fires. These advices concern electric installations, the use of flammable solvents, general cautions associated with plutonium handling, venting and filtration. He finally describes how to fight plutonium fires, and measures to be taken after the fire (staff contamination control, atmosphere control)

French en engineering and operation rules for plutonium facilities

International Nuclear Information System (INIS)

Bertolotti, G.; Drain, F.; Dubois, G.; Monnatte, J.; Mathieu, P.

1998-01-01

COGEMA is operating large size purifying and conditioning plutonium facilities at LA HAGUE and MOX fuels fabrication plant at Marcoule. A high safety standard is recognised for these facilities. It is mainly based on : - prevention of spreading of radioactive materials to workers and environment by physical barriers ensuring static containment and by a cascade of pressure differentials ensuring dynamic containment, - radiation shielding and remote controlled processes ensuring very low dose to workers, - prevention of criticality accident by criticality control methods and double contingency principle, - prevention of fire risks by control of ignition sources, adequate management of combustible materials, physical separation between zones where there is a risk of fire and the remainder of the facility. The facilities are operated while respecting safety requirements as described in the General Operating Rules. The equipment involved in safety functions are monitored and periodically checked. Continuous improvement by incorporation of feed back of safety experience results in: - effective decrease of exposure to operating staff; - reduction of solid waste, liquid and gaseous effluents; - no significant incident recorded. (author)

Fused salt processing of impure plutonium dioxide to high-purity plutonium metal

International Nuclear Information System (INIS)

Mullins, L.J.; Christensen, D.C.; Babcock, B.R.

1982-01-01

A process for converting impure plutonium dioxide (approx. 96% pure) to high-purity plutonium metal (>99.9%) was developed. The process consists of reducing the oxide to an impure plutonium metal intermediate with calcium metal in molten calcium chloride. The impure intermediate metal is cast into an anode and electrorefined to produce high-purity plutonium metal. The oxide reduction step is being done now on a 0.6-kg scale with the resulting yield being >99.5%. The electrorefining is being done on a 4.0-kg scale with the resulting yield being 80 to 85%. The purity of the product, which averages 99.98%, is essentially insensitive to the purity of the feed metal. The yield, however, is directly dependent on the chemical composition of the feed. To date, approximately 250 kg of impure oxide has been converted to pure metal by this processing sequence. The availability of impure plutonium dioxide, together with the need for pure plutonium metal, makes this sequence a valuable plutonium processing tool

Use of a moving-bed ion-exchange column for plutonium purification; Utilisation d'une colonne echangeuse d'ions a lit mobile pour la purification du plutonium

Energy Technology Data Exchange (ETDEWEB)

Sabatier, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1965-07-01

When large amounts of fissile matter have to be purified on ion exchange resins, it is difficult to use a fixed bed because of its limiting maximum size. With a moving bed it is possible to ensure a continuous production which can easily be integrated into a purification line on account of its large production capacity. The installation described in this report is derived from an American prototype designed for uranium separation. As a result of many modifications, it is suitable for the purification of plutonium several such columns will shortly be operating in various French centres. The moving bed column, which has a diameter of 25 mm, was first tried with the uranium-thorium mixture; then, after modifications on the plutonium-uranium mixture. The production capacity will depend on the plutonium concentration which can be tolerated in the effluents. It is possible to treat 150 gm/day of plutonium alone; the effluents obtained have a concentration of around of 1 mg/l. The plutonium-uranium separation is improved by a 5 N acidic rinsing as well as by a temperature increase. The decontamination factor increased from 14 in 7 N nitric acid solution to 115 in 5 N nitric acid solution. A temperature increase of about 20 C leads to a decontamination factor of over 500. This result is sufficient encouraging for the possibility of future installations operating in optimum temperature conditions, i.e. 60-65 C, to be considered. (author) [French] Des que l'on desire purifier sur resine echangeuse d'ions des quantites importantes de matieres fissiles, le lit fixe devient difficilement exploitable par suite des dimensions maximum possibles. Le lit mobile permet une production continue pouvant s'integrer facilement par sa capacite de traitement dans une chaine de purification. L'installation decrite dans ce rapport est derivee d'un prototype americain destine a la separation de l'uranium. De nombreuses modifications en font un ensemble utilisable pour la purification du

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

Energy Technology Data Exchange (ETDEWEB)

Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

2002-02-26

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment

Stop plutonium; Stop plutonium

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-02-01

This press document aims to inform the public on the hazards bound to the plutonium exploitation in France and especially the plutonium transport. The first part is a technical presentation of the plutonium and the MOX (Mixed Oxide Fuel). The second part presents the installation of the plutonium industry in France. The third part is devoted to the plutonium convoys safety. The highlight is done on the problem of the leak of ''secret'' of such transports. (A.L.B.)

HANFORD PLUTONIUM FINISHG PLAN (PFP) COMPLETES PLUTONIUM STABILIZATION KEY SAFETY ISSUES CLOSED

International Nuclear Information System (INIS)

GERBER, M.S.

2004-01-01

A long and intense effort to stabilize and repackage nearly 18 metric tons (MT) of plutonium-bearing leftovers from defense production and nuclear experiments concluded successfully in February, bringing universal congratulations to the Department of Energy's Hanford Site in southeast Washington State. The victorious stabilization and packaging endeavor at the Plutonium Finishing Plant (PFP), managed and operated by prime contractor Fluor Hanford, Inc., finished ahead of all milestones in Hanford's cleanup agreement with regulators, and before deadlines set by the Defense Nuclear Facilities Safety Board (DNFSB), a part of the federal Executive Branch that oversees special nuclear materials. The PFP stabilization and packaging project also completed under budget for its four-year tenure, and has been nominated for a DOE Secretarial Award. It won the Project of the Year Award in the local chapter competition of the Project Management Institute, and is being considered for awards at the regional and national level

A World made of Plutonium?

International Nuclear Information System (INIS)

Broda, E.

1976-01-01

This lecture by Engelbert Broda was written for the 26th Pugwash Conference in Mühlhausen, Germany, 26 – 31 August 1976: Public doubts about nuclear energy are generally directed at the problems of routine emissions of radionuclides, of catastrophic accidents, and of terminal waste disposal. Curiously, the most important problem is not being given sufficient attention: The use of plutonium from civilian reactors fpr weapons production. According to current ideas about a nuclear future, 5000 tons (order of magnitude) of plutonium are to be made annually by year 2000, and about 10 000 tons will all the time be in circulation (transport, reprocessing, reproduction of fuel elements, etc.). It is a misconception that plutonium from power reactors is unsuitable as a nuclear explosive. 5000 tons are enough for several hundred thousand (!) of bombs, Nagasaki type. By the year 2000 maybe 40 – 50 countries will have home-made plutonium. Plutonium production and proliferation are the most serious problems in a nuclear world. (author)

Selecting a plutonium vitrification process

Energy Technology Data Exchange (ETDEWEB)

Jouan, A. [Centre d`Etudes de la Vallee du Rhone, Bagnols sur Ceze (France)

1996-05-01

Vitrification of plutonium is one means of mitigating its potential danger. This option is technically feasible, even if it is not the solution advocated in France. Two situations are possible, depending on whether or not the glass matrix also contains fission products; concentrations of up to 15% should be achievable for plutonium alone, whereas the upper limit is 3% in the presence of fission products. The French continuous vitrification process appears to be particularly suitable for plutonium vitrification: its capacity is compatible with the required throughout, and the compact dimensions of the process equipment prevent a criticality hazard. Preprocessing of plutonium metal, to convert it to PuO{sub 2} or to a nitric acid solution, may prove advantageous or even necessary depending on whether a dry or wet process is adopted. The process may involve a single step (vitrification of Pu or PuO{sub 2} mixed with glass frit) or may include a prior calcination step - notably if the plutonium is to be incorporated into a fission product glass. It is important to weigh the advantages and drawbacks of all the possible options in terms of feasibility, safety and cost-effectiveness.

Operating experience of centrifugal contactors used in a third plutonium purification cycle at the Marcoule reprocessing plant

International Nuclear Information System (INIS)

Coste, J.A.; Breschet, C.A.; Delafontaine, G.L.

1991-01-01

Multistage centrifugal contactors of the type SGN-ROBATEL LX 208 NSC are used in a third plutonium cycle at the Marcoule Reprocessing Plant, they have been smooth-running since the commissioning in 1984. The four centrifugal contactors, totalling 32 stages, are used for Extraction. Scrub and Acidic Strip, and a bank of three flat mixer-settlers is used for plutonium blocking, and soda washing of the solvent. The plutonium product, the residual activity of which is less than one microcurie per gram, is diluent washed in a bank of three flat mixer-settlers, before been concentrated in a zirconium evaporator to 200 g. 1 -1

Analytic determination of plutonium in the environment; Determination analytique du plutonium dans l'environnement

Energy Technology Data Exchange (ETDEWEB)

Ballada, J [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-07-01

The work described in this report was undertaken with a view to determining the plutonium content in the fall-out from nuclear explosions. In the first part are described in turn the importance of the problems due to the plutonium, the physico-chemical properties of the radioelement and the biological dangers which it presents. A detailed and critical analysis is made of the radio-toxicological determination of the plutonium as reported in the literature prior to this report. The second part consists in the presentation of a judicious choice of techniques making it possible to determine plutonium in air, rain-water, soils and ash. After a detailed description of the measurement equipment and the operational techniques which have been developed, a justification of these techniques is given with particular reference to their sensitivity and specificity. After a brief conclusion concerning the preceding chapters, the results are presented. These are then discussed in the ease of each element in which the plutonium has been determined. This discussion is concluded by a consideration of the importance of the occurrence of fall-out plutonium on problems relating to public health. From a consideration of 200 analyses carried out, it is concluded that the contribution of plutonium to the exposure of populations is still very small compared to that of natural radiation and that due to such fission products as strontium 90. The report includes 63 literature references, 26 figures and 11 tables. (author) [French] Les travaux decrits dans ce memoire ont ete entrepris et eflectues dans le but de mettre en evidence le plutonium des retombees radioactives consecutives aux explosions nucleaires. Dans la premiere partie nous etudions successivement l'importance des problemes poses par le plutonium puis les proprietes physicochimiques du radioelement et les dangers qu'il presente du point de vue biologique. Nous effectuons une analyse detaillee et critique des techniques

Interference from radon-thoron daughters in plutonium channel of a continuous plutonium-in-air monitor

International Nuclear Information System (INIS)

Pendharkar, K.A.; Krishnamony, S.

1983-01-01

This paper summarises the results of a study conducted to define the extent of interference from the daughter products of radon/thoron to the plutonium channel of a continuous plutonium-in-air monitor. The effect on the detection limits of the instrument due to chemical form (transportable or non-transportable) and isotopic composition of plutonium aerosol are briefly discussed. (author)

GLASS FABRICATION AND PRODUCT CONSISTENCY TESTING OF LANTHANIDE BOROSILICATE FRIT X COMPOSITION FOR PLUTONIUM DISPOSITION

Energy Technology Data Exchange (ETDEWEB)

Marra, J

2006-11-15

The Department of Energy Office of Environmental Management (DOE/EM) plans to conduct the Plutonium Disposition Project at the Savannah River Site (SRS) to disposition excess weapons-usable plutonium. A plutonium glass waste form is the preferred option for immobilization of the plutonium for subsequent disposition in a geologic repository. A reference glass composition (Lanthanide Borosilicate (LaBS) Frit B) was developed during the Plutonium Immobilization Program (PIP) to immobilize plutonium in the late 1990's. A limited amount of performance testing was performed on this baseline composition before efforts to further pursue Pu disposition via a glass waste form ceased. Recent FY05 studies have further investigated the LaBS Frit B formulation as well as development of a newer LaBS formulation denoted as LaBS Frit X. The objectives of this present task were to fabricate plutonium loaded LaBS Frit X glass and perform corrosion testing to provide near-term data that will increase confidence that LaBS glass product is suitable for disposal in the Yucca Mountain Repository. Specifically, testing was conducted in an effort to provide data to Yucca Mountain Project (YMP) personnel for use in performance assessment calculations. Plutonium containing LaBS glass with the Frit X composition with a 9.5 wt% PuO{sub 2} loading was prepared for testing. Glass was prepared to support Product Consistency Testing (PCT) at Savannah River National Laboratory (SRNL). The glass was thoroughly characterized using x-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS) prior to performance testing. A series of PCTs were conducted at SRNL using quenched Pu Frit X glass with varying exposed surface areas. Effects of isothermal and can-in-canister heat treatments on the Pu Frit X glass were also investigated. Another series of PCTs were performed on these different heat-treated Pu Frit X glasses. Leachates from all these PCTs

Plutonium Disposition by Immobilization

International Nuclear Information System (INIS)

Gould, T.; DiSabatino, A.; Mitchell, M.

2000-01-01

The ultimate goal of the Department of Energy (DOE) Immobilization Project is to develop, construct, and operate facilities that will immobilize between 17 to 50 tonnes (MT) of U.S. surplus weapons-usable plutonium materials in waste forms that meet the ''spent fuel'' standard and are acceptable for disposal in a geologic repository. Using the ceramic can-in-canister technology selected for immobilization, surplus plutonium materials will be chemically combined into ceramic forms which will be encapsulated within large canisters of high level waste (HLW) glass. Deployment of the immobilization capability should occur by 2008 and be completed within 10 years. In support of this goal, the DOE Office of Fissile Materials Disposition (MD) is conducting development and testing (D and T) activities at four DOE laboratories under the technical leadership of Lawrence Livermore National Laboratory (LLNL). The Savannah River Site has been selected as the site for the planned Plutonium Immobilization Plant (PIP). The D and T effort, now in its third year, will establish the technical bases for the design, construction, and operation of the U. S. capability to immobilize surplus plutonium in a suitable and cost-effective manner. Based on the D and T effort and on the development of a conceptual design of the PIP, automation is expected to play a key role in the design and operation of the Immobilization Plant. Automation and remote handling are needed to achieve required dose reduction and to enhance operational efficiency

Plutonium working group report on environmental, safety and health vulnerabilities associated with the Department's plutonium storage. Volume I: Summary

International Nuclear Information System (INIS)

1994-11-01

At the conclusion of the Cold War, the Department of Energy (DOE) stopped plutonium processing for nuclear weapons production. Facilities used for that purpose now hold significant quantities of plutonium in various forms. Unless properly stored and handled, plutonium can present environment, safety and health (ES ampersand H) hazards. Improperly stored plutonium poses a variety of hazards. When containers or packaging fail to fully protect plutonium metal from exposure to air, oxidation can occur and cause packaging failures and personnel contamination. Contamination can also result when plutonium solutions leak from bottles, tanks or piping. Plutonium in the form of scrap or residues generated by weapons production are often very corrosive, chemically reactive and difficult to contain. Buildings and equipment that are aging, poorly maintained or of obsolete design contribute to the overall problem. Inadvertent accumulations of plutonium of any form in sufficient quantities within facilities can result in nuclear criticality events that could emit large amounts of radiation locally. Contamination events and precursors of criticality events are causing safety and health concerns for workers at the Department's plutonium facilities. Contamination events also potentially threaten the public and the surrounding environment

Studies on resin degradation products encountered during purification of plutonium by anion exchange

International Nuclear Information System (INIS)

Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Dhumwad, R.K.

1991-01-01

Among the methods available for the purification of plutonium in Purex process, anion exchange method offers several advantages. However, on repeated use, the resin gets degraded due to thermal, radiolytic and chemical attacks resulting in chemical as well as physical damage. Frequently, plutonium product eluted from such resin contains significant quantities of white precipitates. A few anion exchange resins were leached with 8 M HNO 3 at 60-80degC and the resin degradation products (RDP) in the leach-extract were found to give similar precipitates with tetravalent metal ions like Pu(IV), Th(IV) etc. Tetra propyl ammonium hydroxide in 8 M HNO 3 (TPAN) also gave a white precipitate with plutonium similar to the one found in the elution streams. The results indicate that delinked quaternary ammonium functional groups might be responsible for the formation of precipitate. The characteristics of precipitates Th-RDP, Th-TPAN and that isolated from elution stream have been investigated. In a separate study a tentative formula for Th-RDP compound is proposed. The influence of RDP on the extraction of plutonium and other components in Purex process was studied and it was found that RDP complexes metal ions thus marginally affecting the kd values. A spectrophotometric method has been standardised to monitor the extent of degradation of anion exchange resins which is based on the ability of RDP to reduce the colour intensity of Th-thoron complex. This technique can be used to study the stability of the anion exchange resins. (author). 8 refs., 8 tabs., 5 figs.,

Plutonium story

International Nuclear Information System (INIS)

Seaborg, G.T.

1981-09-01

The first nuclear synthesis and identification (i.e., the discovery) of the synthetic transuranium element plutonium (isotope 238 Pu) and the demonstration of its fissionability with slow neutrons (isotope 239 Pu) took place at the University of California, Berkeley, through the use of the 60-inch and 37-inch cyclotrons, in late 1940 and early 1941. This led to the development of industrial scale methods in secret work centered at the University of Chicago's Metallurgical Laboratory and the application of these methods to industrial scale production, at manufacturing plants in Tennessee and Washington, during the World War II years 1942 to 1945. The chemical properties of plutonium, needed to devise the procedures for its industrial scale production, were studied by tracer and ultramicrochemical methods during this period on an extraordinarily urgent basis. This work, and subsequent investigations on a worldwide basis, have made the properties of plutonium very well known. Its well studied electronic structure and chemical properties give it a very interesting position in the actinide series of inner transition elements

Six-kilogram-scale electrorefining of plutonium metal

International Nuclear Information System (INIS)

Mullins, L.J.; Morgan, A.N.; Apgar, S.A. III; Christensen, D.C.

1982-09-01

The electrorefining of metallic plutonium scrap to produce high purity metal has been an established procedure at Los Alamos since 1964. This is a batch process and was limited to 4-kg plutonium because of criticality safety considerations. Improvements in critical mass measurements have permitted us to develop a process for 6-kg plutonium. The 6-kg process is now operational. The increased size of the process, together with other improvements which have been made, makes plutonium electrorefining the principal industrial tool for processing and purifying metallic plutonium scrap

Some investigations into the behaviour of plutonium in the marine environment

International Nuclear Information System (INIS)

Hetherington, J.A.; Jefferies, D.F.; Lovett, M.B.

1975-01-01

Plutonium and other transuranic elements are produced during the operation of magnox reactors, although they do not usually present any problems of waste management until the fuel reaches the reprocessing stage. At this stage one of the chief aims is to separate and recover the plutonium from the depleted uranium and the fission products, the incentive being the value of plutonium itself as a nuclear fuel particularly for the fast breeder reactors. For this reason the amounts of plutonium appearing in the low activity waste streams for discharge to the environment are usually small, and their radiological impact insignificant compared with that of the more abundant fission products. However, with the prospect of use of large quantities of plutonium in the fast reactors, considerable interest surrounds all aspects of the properties of plutonium, and in an attempt to give a fuller understanding of the behaviour of plutonium in the marine environment use has been made of the small amounts which have been discharged in recent years to the north-east Irish Sea from the fuel reprocessing plant at Windscale in Cumbria. Details of the amounts of 239 Pu and 240 Pu discharged to sea are given and the resultant measured concentrations of the isotopes in the water throughout the Irish Sea are shown. The distribution of plutonium in sea water is shown to follow the same basic pattern as the fission products, details of which have been reported previously. By comparing the concentration gradients with distance from the point of discharge for plutonium and caesium, a value for the rate of loss of plutonium from the water to the other important compartment, namely the sediment, has been deduced. Measurements have been made of 239 Pu and 240 Pu in seabed and estuarine sediments and by combining these with the water data values of the concentration factor for the nuclide in sediment have been calculated. The results of a number of core samples have been used to calculate an

MIS High-Purity Plutonium Oxide Metal Oxidation Product TS707001 (SSR123): Final Report

Energy Technology Data Exchange (ETDEWEB)

Veirs, Douglas Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stroud, Mary Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Berg, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Narlesky, Joshua Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Max A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carillo, Alex [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-08-09

A high-purity plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. Sample TS707001 represents process plutonium oxides from several metal oxidation operations as well as impure and scrap plutonium from Hanford that are currently stored in 3013 containers. After calcination to 950°C, the material contained 86.98% plutonium with no major impurities. This study followed over time, the gas pressure of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. Gas compositions were measured periodically over a six year period. The maximum observed gas pressure was 138 kPa. The increase over the initial pressure of 80 kPa was primarily due to generation of nitrogen and carbon dioxide gas in the first six months. Hydrogen and oxygen were minor components of the headspace gas. At the completion of the study, the internal components of the sealed container showed signs of corrosion, including pitting.

The plutonium danger

International Nuclear Information System (INIS)

Ruiter, W. de

1983-01-01

Nobody can ignore the fact that plutonium is potentially very dangerous and the greatest danger concerning it lies in the spreading of nuclear weapons via nuclear energy programmes. The following seven different attitudes towards this problem are presented and discussed: 1) There is no connection between peaceful and military applications; 2) The problem cannot be prevented; 3) A technical solution must be found; 4) plutonium must be totally inaccessible to countries involved in acquiring nuclear weapons; 5) The use of plutonium for energy production should only occur in one multinational centre; 6) Dogmas in the nuclear industry must be enfeebled; 7) All developments in this area should stop. (C.F.)

Smart unattended systems for plutonium safeguards

International Nuclear Information System (INIS)

Menlove, H.O.; Abhold, M.; Eccleston, G.; Puckett, J.M.

1996-01-01

During the past decade, IAEA inspectors, national inspectors, and facility operators have used neutron coincidence counters and gamma-ray isotopics measurements extensively to measure the plutonium content of various forms of nuclear materials in the fuel cycle. Large automated facilities for fabricating plutonium fuel present both difficulties and challenges for improved accounting of nuclear materials. The traditional methods of sample measurements, requiring the transfer of the sample from the production line to the assay measurement station, are not possible in automated facilities. A bilateral safeguards agreement between the US Department of Energy (DOE) and Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan was signed to develop and implement nondestructive assay (NDA) systems to provide continuous safeguards measurements for material accountancy in the robot-automated Plutonium Fuel Fabrication Facility (PFFF). The PFFF assay systems were required to operate in unattended mode with a size and fuel mass capability to match the robotics fuel manipulators. Unattended assay systems reduce the requirement for inspector''s oversight of measurement operations, reduce the inspector''s workload, and improve inspection efficiencies. In addition, unattended measurements become essential when facility constraints limit the access of inspectors to the operations area during material processing. Authentication techniques were incorporated into the NDA systems so that data obtained form unattended assays could be used by independent inspectors such as the IAEA. The standardized containers and robot-controlled fuel movements in automated facilities enable more accurate nondestructive assay (NDA) measurements than are possible in conventional nonautomated facilities. The NDA instrumentation can be custom designed and optimized for the particular measurement goal in the automated facility

Method of processing plutonium and uranium solution

International Nuclear Information System (INIS)

Otsuka, Katsuyuki; Kondo, Isao; Suzuki, Toru.

1989-01-01

Solutions of plutonium nitrate solutions and uranyl nitrate recovered in the solvent extraction step in reprocessing plants and nuclear fuel production plants are applied with low temperature treatment by means of freeze-drying under vacuum into residues containing nitrates, which are denitrated under heating and calcined under reduction into powders. That is, since complicate processes of heating, concentration and dinitration conducted so far for the plutonium solution and uranyl solution are replaced with one step of freeze-drying under vacuum, the process can be simplified significantly. In addition, since the treatment is applied at low temperature, occurrence of corrosion for the material of evaporation, etc. can be prevented. Further, the number of operators can be saved by dividing the operations into recovery of solidification products, supply and sintering of the solutions and vacuum sublimation. Further, since nitrates processed at a low temperature are powderized by heating dinitration, the powderization step can be simplified. The specific surface area and the grain size distribution of the powder is made appropriate and it is possible to obtain oxide powders of physical property easily to be prepared into pellets. (N.H.)

Plutonium again (smuggling and movements)

International Nuclear Information System (INIS)

Anon.

1994-01-01

A link is discounted between nuclear proliferation and the recently discovered smuggled plutonium from the former Soviet Union at Munich airport and other places in Germany. It is argued that governments wishing to obtain nuclear materials to develop a weapons programme would not arrange to have it smuggled in a suitcase. Instead, it is speculated that a link exists between the plutonium smuggling incidents and the desire to promote the production of mixed oxide (MOX) fuel. Such incidents, by further raising public anxiety, may be intended to turn public opinion in favour of MOX fuel production as a sensible way of getting rid of surplus plutonium. (UK)

Analytic determination of plutonium in the environment; Determination analytique du plutonium dans l'environnement

Energy Technology Data Exchange (ETDEWEB)

Ballada, J. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-07-01

The work described in this report was undertaken with a view to determining the plutonium content in the fall-out from nuclear explosions. In the first part are described in turn the importance of the problems due to the plutonium, the physico-chemical properties of the radioelement and the biological dangers which it presents. A detailed and critical analysis is made of the radio-toxicological determination of the plutonium as reported in the literature prior to this report. The second part consists in the presentation of a judicious choice of techniques making it possible to determine plutonium in air, rain-water, soils and ash. After a detailed description of the measurement equipment and the operational techniques which have been developed, a justification of these techniques is given with particular reference to their sensitivity and specificity. After a brief conclusion concerning the preceding chapters, the results are presented. These are then discussed in the ease of each element in which the plutonium has been determined. This discussion is concluded by a consideration of the importance of the occurrence of fall-out plutonium on problems relating to public health. From a consideration of 200 analyses carried out, it is concluded that the contribution of plutonium to the exposure of populations is still very small compared to that of natural radiation and that due to such fission products as strontium 90. The report includes 63 literature references, 26 figures and 11 tables. (author) [French] Les travaux decrits dans ce memoire ont ete entrepris et eflectues dans le but de mettre en evidence le plutonium des retombees radioactives consecutives aux explosions nucleaires. Dans la premiere partie nous etudions successivement l'importance des problemes poses par le plutonium puis les proprietes physicochimiques du radioelement et les dangers qu'il presente du point de vue biologique. Nous effectuons une analyse detaillee et critique des techniques

Plutonium recovery from carbonate wash solutions

International Nuclear Information System (INIS)

Gray, J.H.; Reif, D.J.; Chostner, D.F.; Holcomb, H.P.

1991-01-01

540Periodically higher than expected levels of plutonium are found in carbonate solutions used to wash second plutonium cycle solvent. The recent accumulation of plutonium in carbonate wash solutions has led to studies to determine the cause of that plutonium accumulation, to evaluate the quality of all canyon solvents, and to develop additional criteria needed to establish when solvent quality is acceptable. Solvent from three canyon solvent extraction cycles was used to evaluate technology required to measure tributyl phosphate (TBP) degradation products and was used to evaluate solvent quality criteria during the development of plutonium recovery processes. 1 fig

Plan for the Startup of HA-21I Furnace Operations at the Plutonium Finishing Plant (PFP)

International Nuclear Information System (INIS)

WILLIS, H.T.

2000-01-01

Achievement of Thermal Stabilization mission elements require the installation and startup of three additional muffle furnaces for the thermal stabilization of plutonium and plutonium bearing materials at the Plutonium Finishing Plant (PFP). The release to operate these additional furnaces will require an Activity Based Startup Review. The conduct of the Activity Based Startup Review (ABSR) was approved by Fluor Daniel Hanford on October 15, 1999. This plan has been developed with the objective of identifying those activities needed to guide the controlled startup of five furnaces from authorization to unrestricted operations by adding the HA-211 furnaces in an orderly and safe manner after the approval to Startup has been given. The Startup Plan provides a phased approach that bridges the activities between the completion of the Activity Based Startup Review authorizing the use of the three additional furnaces and the unrestricted operation of the five thermal stabilization muffle furnaces. The four phases are: (1) the initiation of five furnace operations using three empty (simulated full) boat charges from HA-211 and two full charges from HC-21C; (2) three furnace operations (one full charge from HA-211 and two full charges from HC-21C); (3) four furnace operations (two full charges from HA-211 and two full charges from HC-21C); and (4) integrated five furnace operations and unrestricted operations. Phase 1 of the Plan will be considered as the cold runs. This Plan also provides management oversight and administrative controls that are to be implemented until unrestricted operations are authorized. It also provides a formal review process for ensuring that all preparations needed for full five furnace operations are completed and formally reviewed prior to proceeding to the increased activity levels associated with five furnace operations. Specific objectives include: (1) To ensure that activities are conducted in a safe manner. (2) To provide supplemental

Long-term plutonium storage: Design concepts

International Nuclear Information System (INIS)

Wilkey, D.D.; Wood, W.T.; Guenther, C.D.

1994-01-01

An important part of the Department of Energy (DOE) Weapons Complex Reconfiguration (WCR) Program is the development of facilities for long-term storage of plutonium. The WCR design goals are to provide storage for metals, oxides, pits, and fuel-grade plutonium, including material being held as part of the Strategic Reserve and excess material. Major activities associated with plutonium storage are sorting the plutonium inventory, material handling and storage support, shipping and receiving, and surveillance of material in storage for both safety evaluations and safeguards and security. A variety of methods for plutonium storage have been used, both within the DOE weapons complex and by external organizations. This paper discusses the advantages and disadvantages of proposed storage concepts based upon functional criteria. The concepts discussed include floor wells, vertical and horizontal sleeves, warehouse storage on vertical racks, and modular storage units. Issues/factors considered in determining a preferred design include operational efficiency, maintenance and repair, environmental impact, radiation and criticality safety, safeguards and security, heat removal, waste minimization, international inspection requirements, and construction and operational costs

Plutonium assemblies in reload 1 of the Dodewaard Reactor

International Nuclear Information System (INIS)

Bairiot, H.; Deramaix, P.; Vandenberg, C.; Leenders, L.; Mostert, P.

1977-01-01

Since 1963, Belgonucleaire has been developing the design of plutonium assemblies of the island type (i.e., plutonium rods inserted in the control zone of the assembly and enriched uranium rods at the periphery) for light water reactors. The application to boiling water reactors (BWRs) led to the introduction, in April 1971, of two prototype plutonium island assemblies in the Dodewaard BWR (The Netherlands): Those assemblies incorporating plutonium in 42 percent of the rods are interchangeable with standard uranium assemblies of the same reload. Their design, which had to meet these criteria, was performed using the routine order in use at Belgonucleaire; experimental checks included a mock-up configuration simulated in the VENUS critical facility at Mol and open-vessel cold critical experiments performed in the Dodewaard core. The pelleted plutonium rods were fabricated and controlled by Belgonucleaire following the manufacturing procedures developed at the production plant. In one of the assemblies, three vibrated plutonium fuel rods with a lower fuel density were introduced in the three most highly rated positions to reduce the power rating. Those plutonium assemblies experienced peak pellet ratings up to 535 W/cm and were discharged in April 1974 after having reached a mean burnup of approximately 21,000 MWd/MT. In-core instrumentation during operation, visual examinations, and reactivity substitution experiments during reactor shutdown did not indicate any special feature for those assemblies compared to the standard uranium assemblies, thereby demonstrating their interchangeability

Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

International Nuclear Information System (INIS)

Brownson, D.A.; Hanson, D.J.; Blackman, H.S.

1993-06-01

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition

Weapons-grade plutonium dispositioning. Volume 2: Comparison of plutonium disposition options

Energy Technology Data Exchange (ETDEWEB)

Brownson, D.A.; Hanson, D.J.; Blackman, H.S. [and others

1993-06-01

The Secretary of Energy requested the National Academy of Sciences (NAS) Committee on International Security and Arms Control to evaluate disposition options for weapons-grade plutonium. The Idaho National Engineering Laboratory (INEL) offered to assist the NAS in this evaluation by investigating the technical aspects of the disposition options and their capability for achieving plutonium annihilation levels greater than 90%. This report was prepared for the NAS to document the gathered information and results from the requested option evaluations. Evaluations were performed for 12 plutonium disposition options involving five reactor and one accelerator-based systems. Each option was evaluated in four technical areas: (1) fuel status, (2) reactor or accelerator-based system status, (3) waste-processing status, and (4) waste disposal status. Based on these evaluations, each concept was rated on its operational capability and time to deployment. A third rating category of option costs could not be performed because of the unavailability of adequate information from the concept sponsors. The four options achieving the highest rating, in alphabetical order, are the Advanced Light Water Reactor with plutonium-based ternary fuel, the Advanced Liquid Metal Reactor with plutonium-based fuel, the Advanced Liquid Metal Reactor with uranium-plutonium-based fuel, and the Modular High Temperature Gas-Cooled Reactor with plutonium-based fuel. Of these four options, the Advanced Light Water Reactor and the Modular High Temperature Gas-Cooled Reactor do not propose reprocessing of their irradiated fuel. Time constraints and lack of detailed information did not allow for any further ratings among these four options. The INEL recommends these four options be investigated further to determine the optimum reactor design for plutonium disposition.

Recovery studies for plutonium machining oil coolant

International Nuclear Information System (INIS)

Navratil, J.D.; Baldwin, C.E.

1977-01-01

Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

Swiss R and D on uranium-free LWR fuels for plutonium incineration

International Nuclear Information System (INIS)

Stanculescu, A.; Chawla, R.; Degueldre, C.; Kasemeyer, U.; Ledergerber, G.; Paratte, J.M.

1999-01-01

The most efficient way to enhance the plutonium consumption in LWRs is to eliminate plutonium production altogether. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. The inert matrix material studied at PSI is zirconium oxide. For reactivity control reasons, adding a burnable poison to this fuel proves to be necessary. The studies performed at PSI have identified erbium oxide as the most suitable candidate for this purpose. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of the chosen single phase zirconium/ erbium/plutonium oxide material stabilised as a cubic solution by yttrium. The results to-date, obtained for inert matrix samples containing thorium or cerium as plutonium substitute, confirm the robustness and stability of this material. With regard to reactor physics aspects, our studies indicate the feasibility of uranium-free, plutonium-fuelled cores having operational characteristics quite similar to those of conventional UO 2 -fuelled ones, and much higher plutonium consumption rates, as compared to 100% MOX loadings. The safety features of such cores, based on results obtained from static neutronics calculations, show no cliff edges. However, the need for further detailed transient analyses is clearly recognised. Summarising, PSI's studies indicate the feasibility of a uranium-free plutonium fuel to be considered in 'maximum plutonium consumption LWRs' operating in a 'once-through' mode. With regard to reactor physics, future efforts will concentrate on strengthening the safety case of uranium-free cores, as well as on improving the integral data base for validation of the neutronics calculations. Material technology studies will be continued to investigate the physico-chemical properties of the inert matrix fuel containing plutonium and will focus on the planning and evaluation of

Los Alamos DP West Plutonium Facility decontamination project

International Nuclear Information System (INIS)

Garde, R.; Cox, E.J.; Valentine, A.M.

1982-01-01

The DP West Plutonium Facility operated by the Los Alamos National Laboratory, Los Alamos, New Mexico, was decontaminated between April 1978 and April 1981. The facility was constructed in 1944 to 1945 to produce plutonium metal and fabricate parts for nuclear weapons. It was continually used as a plutonium processing and research facility until mid-1978. Decontamination operations included dismantling and removing gloveboxes and conveyor tunnels; removing process systems, utilities, and exhaust ducts; and decontaminating all remaining surfaces. This report describes glovebox and conveyor tunnel separations, decontamination techniques, health and safety considerations, waste management procedures, and costs of the operation

Plutonium Oxide Process Capability Work Plan

Energy Technology Data Exchange (ETDEWEB)

Meier, David E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2014-02-28

Pacific Northwest National Laboratory (PNNL) has been tasked to develop a Pilot-scale Plutonium-oxide Processing Unit (P3U) providing a flexible capability to produce 200g (Pu basis) samples of plutonium oxide using different chemical processes for use in identifying and validating nuclear forensics signatures associated with plutonium production. Materials produced can also be used as exercise and reference materials.

ZPR-9 airborne plutonium monitoring system

International Nuclear Information System (INIS)

Rusch, G.K.; McDowell, W.P.; Knapp, W.G.

1975-01-01

An airborne plutonium monitoring system which is installed in the ZPR-9 (Zero Power Reactor No. 9) facility at Argonne National Laboratory is described. The design and operational experience are discussed. This monitoring system utilizes particle size and density discrimination, alpha particle energy discrimination, and a background-subtraction techique operating in cascade to separate airborne-plutonium activity from other, naturally occurring, airborne activity. Relatively high sensitivity and reliability are achieved

R.4. Innovative concept for plutonium finishing facility

International Nuclear Information System (INIS)

Bertolotti, G.; Laguerie, I.V. de; Richter, R.; Gillet, B.

1998-01-01

After complete shutdown of the units of the previous UP2 plant, the new R4 facility will ensure the purification of Plutonium of the UP2-800 plant for the whole range of fuel to be reprocessed in the years to come. This facility features four main units: - Purification of plutonium nitrate; - Conversion into plutonium oxide; - PuO 2 conditioning into cans; Acid recovery. An extensive R and D program resulted in significant innovations. From a technological aspect the centrifugal extractor and the sub-critical tube bundle tank contributed to the reduction of the building dimensions. The extensive use of on-line analyses enables a more efficient follow-up of the process while minimizing the effluent production. On the other hand, the organization of the building which consists in grouping the rooms presenting the same risk of dispersal of nuclear materials also contributed to reduce the active zone volume. This facility, as any other facilities on the LA HAGUE site, will be remotely operated. (author)

Automation of plutonium spectrophotometry

International Nuclear Information System (INIS)

Perez, J.J.; Boisde, G.; Goujon de Beauvivier, M.; Chevalier, G.; Isaac, M.

1980-01-01

Instrumentation was designed and constructed for automatic control of plutonium by molecular absorption spectrophotometry, on behalf of the reprocessing facilities, to meet two objectives: on-line measurement, of the valency state of plutonium, on by-pass, with the measured concentration covering the process concentration range up to a few mg.l -1 ; laboratory measurement of plutonium adjusted to valency VI, with operation carried out using a preparative system meeting the required containment specifications. For this two objectives, the photometer, optical cell connections are made by optical fibers resistant to β, γ radiation. Except this characteristic the devices are different according to the quality required for result [fr

Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

International Nuclear Information System (INIS)

Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

2001-01-01

In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

CONVERSION OF PLUTONIUM TRIFLUORIDE TO PLUTONIUM TETRAFLUORIDE

Science.gov (United States)

Fried, S.; Davidson, N.R.

1957-09-10

A large proportion of the trifluoride of plutonium can be converted, in the absence of hydrogen fluoride, to the tetrafiuoride of plutonium. This is done by heating plutonium trifluoride with oxygen at temperatures between 250 and 900 deg C. The trifiuoride of plutonium reacts with oxygen to form plutonium tetrafluoride and plutonium oxide, in a ratio of about 3 to 1. In the presence of moisture, plutonium tetrafluoride tends to hydrolyze at elevated temperatures and therefore it is desirable to have the process take place under anhydrous conditions.

Search for plutonium salt deposits in the plutonium extraction batteries of the Marcoule plant (1963)

International Nuclear Information System (INIS)

Bouzigues, H.; Reneaud, J.M.

1963-01-01

This report describes a method and a special apparatus making it possible to detach the insoluble plutonium salt deposits in the extraction chain of an irradiated fuel treatment plant. The process chosen allows the detection, in the extraction batteries or in the highly active chemical engineering equipment, of plutonium quantities of a few grains. After four years operation it has been impossible to detect measurable quantities of plutonium in any part of the extraction chain. The results have been confirmed by visual examinations carried out with a specially constructed endoscope. (authors) [fr

Plutonium Vulnerability Management Plan

International Nuclear Information System (INIS)

1995-03-01

This Plutonium Vulnerability Management Plan describes the Department of Energy's response to the vulnerabilities identified in the Plutonium Working Group Report which are a result of the cessation of nuclear weapons production. The responses contained in this document are only part of an overall, coordinated approach designed to enable the Department to accelerate conversion of all nuclear materials, including plutonium, to forms suitable for safe, interim storage. The overall actions being taken are discussed in detail in the Department's Implementation Plan in response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-1. This is included as Attachment B

Salt stripping: a pyrochemical approach to the recovery of plutonium electrorefining salt residues

International Nuclear Information System (INIS)

Christensen, D.C.; Mullins, L.J.

1982-10-01

A pyrochemical process has been developed to take the salt residue from the plutonium electrorefining process and strip the plutonium from it. The process, called salt stripping, uses calcium as a reducing/coalescing agent. In a one-day operation, greater than 95% of the plutonium can be recovered as a metallic button. As much as 88% of the residue is either reused as metal or discarded as a clean salt. A thin layer of black salts, which makes up the bulk of the unrecovered Pu, is a by-product of the initial reductions. A number of black salts can be collected together and re-reduced in a second step. Greater than 88% of this plutonium can be successfully recovered in this second stage with the resulting residues being discardable. The processing time, number of processor hours, and the volume of secondary residues are greatly reduced over the classical aqueous recovery methods. In addition, the product metal is of sufficient quality to be fed directly to the electrorefining process for purification. 8 figures, 7 tables

Furnace System Testing to Support Lower-Temperature Stabilization of High Chloride Plutonium Oxide Items at the Hanford Plutonium Finishing Plant

International Nuclear Information System (INIS)

Schmidt, Andrew J.; Gerber, Mark A.; Fischer, Christopher M.; Elmore, Monte R.

2003-01-01

High chloride content plutonium (HCP) oxides are impure plutonium oxide scrap which contains NaCl, KCl, MgCl2 and/or CaCl2 salts at potentially high concentrations and must be stabilized at 950 C per the DOE Standard, DOE-STD-3013-2000. The chlorides pose challenges to stabilization because volatile chloride salts and decomposition products can corrode furnace heating elements and downstream ventilation components. Thermal stabilization of HCP items at 750 C (without water washing) is being investigated as an alternative method for meeting the intent of DOE STD 3013-2000. This report presents the results from a series of furnace tests conducted to develop material balance and system operability data for supporting the evaluation of lower-temperature thermal stabilization

Remote handling in the Plutonium Immobilization Project: Plutonium conversion and first stage immobilization

International Nuclear Information System (INIS)

Brault, J.R.

2000-01-01

Since the break up of the Soviet Union at the end of the Cold War, the United States and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

The use of calorimetry for plutonium assay

International Nuclear Information System (INIS)

Mason, J.A.

1982-12-01

Calorimetry is a technique for measuring the thermal power of heat-producing substances. The technique may be applied to the measurement of plutonium-bearing materials which evolve heat as a result of alpha and beta decay. A calorimetric measurement of the thermal power of a plutonium sample, combined with a knowledge or measurement of the plutonium isotopic mass ratios of the sample provides a convenient and accurate, non-destructive measure of the total plutonium mass of the sample. The present report provides a description, and an assessment of the calorimetry technique applied to the assay of plutonium-bearing materials. Types and characteristics of plutonium calorimeters are considered, as well as calibration and operating procedures. The instrumentation used with plutonium calorimeters is described and the use of computer control for calorimeter automation is discussed. A critical review and assessment of plutonium calorimetry literature since 1970 is presented. Both fuel element and plutonium-bearing material calorimeters are considered. The different types of plutonium calorimeters are evaluated and their relative merits are discussed. A combined calorimeter and gamma-ray measurement assay system is considered. The design principles of plutonium assay calorimeters are considered. An automatic, computer-based calorimeter control system is proposed in conjunction with a general plutonium assay calorimeter design. (author)

Fast Thorium Molten Salt Reactors Started with Plutonium

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

2006-01-01

One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

Chloride-catalyzed corrosion of plutonium in glovebox atmospheres

International Nuclear Information System (INIS)

Burgess, M.; Haschke, J.M.; Allen, T.H.; Morales, L.A.; Jarboe, D.M.; Puglisi, C.V.

1998-04-01

Characterization of glovebox atmospheres and the black reaction product formed on plutonium surfaces shows that the abnormally rapid corrosion of components in the fabrication line is consistent with a complex salt-catalyzed reaction involving gaseous hydrogen chloride (HCl) and water. Analytical data verify that chlorocarbon and HCl vapors are presented in stagnant glovebox atmospheres. Hydrogen chloride concentrations approach 7 ppm at some locations in the glovebox line. The black corrosion product is identified as plutonium monoxide monohydride (PuOH), a product formed by hydrolysis of plutonium in liquid water and salt solutions at room temperature. Plutonium trichloride (PuCl 3 ) produced by reaction of HCl at the metal surface is deliquescent and apparently forms a highly concentrated salt solution by absorbing moisture from the glovebox atmosphere. Rapid corrosion is attributed to the ensuing salt-catalyzed reaction between plutonium and water. Experimental results are discussed, possible involvement of hydrogen fluoride (HF) is examined, and methods of corrective action are presented in this report

An MS-DOS-based program for analyzing plutonium gamma-ray spectra

International Nuclear Information System (INIS)

Ruhter, W.D.; Buckley, W.M.

1989-01-01

A plutonium gamma-ray analysis system that operates on MS-DOS-based computers has been developed for the International Atomic Energy Agency (IAEA) to perform in-field analysis of plutonium gamma-ray spectra for plutonium isotopics. The program titled IAEAPU consists of three separate applications: (1) a data-transfer application for transferring spectral data from a CICERO multichannel analyzer to a binary data file, (2) a data-analysis application to analyze plutonium gamma-ray spectra for plutonium isotopic ratios and weight percents of total plutonium, (3) and a data-quality assurance application to check spectral data for proper data-acquisition setup and performance. Volume 2 describes the operations of these applications and the installation and maintenance of the software

Plutonium separation by reduction stripping. Application to processing of mixed oxide (U,Pu)O2 fuel fabrication wastes

International Nuclear Information System (INIS)

Arnal, Thierry; Cousinou, Gerard; Ganivet, Michel.

1978-11-01

A procedure is described for separating plutonium from a uranium VI and plutonium IV mixture contained in an organic phase (tributyl phosphate diluted in dodecane). This separation is obtained by extracting the plutonium III using two organic reducers: hydrazine and paraminophenol. Paraminophenol has excellent reducing qualities, similar to those of ferrous sulphamate, but has the added advantage of not contaminating extracted plutonium. This procedure is currently used in processing production wastes from mixed oxide (U,Pu)O 2 fuels; the installation using this procedure is described in detail in this paper. Operating results show the remarkable efficiency of this procedure: the separated plutonium and uranium mass flows have been increased to 185 and 350 g.h -1 respectively; the uranium contains less than 0.1 ppm of plutonium on completion of the purification cycle [fr

Examination of the potential for diversion or clandestine dual use of a pebble-bed reactor to produce plutonium

International Nuclear Information System (INIS)

Ougouag, A.M.; Terry, W.K.; Gougar, H.D.

2002-01-01

This paper explores the susceptibility of Pebble-Bed Reactors (PBRs) to be used overtly or covertly for the production of plutonium for nuclear weapons. The basic assumption made for the consideration of overt production is that a country would purchase a PBR with the ostensible motive of producing electric power; then, after the power plant was built, the country would divert the facility entirely to the production of weapons material. It is assumed that the country would then have to manufacture production pebbles from natural uranium. The basic assumption made for covert production is that the country would obtain and use a PBR for power production, but that it would clandestinely feed plutonium production pebbles through the reactor in such small numbers that the perturbation on power plant operation would be very difficult to detect. This paper shows the potential rate of plutonium production under such constraints. It is demonstrated that the PBR is a very poor choice for either form of proliferation-intent use. (author)

The first metallurgical tests on plutonium; Premiers essais metallurgiques sur le plutonium

Energy Technology Data Exchange (ETDEWEB)

Grison, E; Abramson, R; Anselin, F; Monti, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

Metallic plutonium was first prepared in France in January 1956, as soon as we had access to quantities of the order of several grams of plutonium, which had been extracted from the rods of the pile EL2 at Saclay. Since up to the present this reactor, of thermal power 2 000 kW, has been our only source of plutonium, we have so far only worked on experimental quantities sufficient for the basic tests but not for tests on a scale of possible applications. It is this work, carried out during this phase of preliminary research, which is described below. With the starting up of the plutonium extraction plant at Marcoule, where the reactor G1 has been operating at power for more than a year, we shall go on next to a another order of magnitude which will allow the manufacture and experimentation of prototype fuel elements. (author) [French] La premiere elaboration de plutonium metallique en France fut faite en janvier 1956, des que nous pumes disposer de quantites de plutonium de l'ordre de quelques grammes, qui avaient ete retires des barreaux de la pile EL2 de Saclay. Ce reacteur, d'une puissance thermique de 2 000 kW, ayant ete jusqu'a present notre seule source de plutonium, nous n'avons encore travaille que sur des quantites experimentales suffisantes pour les essais de base, mais non pour des essais a l'echelle d'applications possibles. Ce sont les travaux effectues pendant cette phase de recherches preliminaires qui seront evoques ci-dessous. Avec la mise eu route de l'usine d'extraction de plutonium de Marcoule, ou le reacteur G1 fonctionne en puissance depuis plus d'un an, nous allons passer prochainement a un autre ordre de grandeur, qui nous permettra la fabrication et l'experimentation d'elements combustibles prototypes. (auteur)

Plutonium-239 production rate study using a typical fusion reactor

International Nuclear Information System (INIS)

Faghihi, F.; Havasi, H.; Amin-Mozafari, M.

2008-01-01

The purpose of the present paper is to compute fissile 239 Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m 2 ) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate 239 Pu production rate. Produced 239 Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, 239 Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type

Enzymatic degradation of plutonium-contaminated cellulose products

International Nuclear Information System (INIS)

Heintz, C.E.; Rainwater, K.A.; Swift, L.M.; Barnes, D.L.; Worl, L.; Avens, L.

1999-01-01

Enzyme solutions produced for commercial purposes unrelated to waste management have the potential for reducing the volume of wastes in streams containing cellulose, lipid and protein materials. For example, the authors have shown that cellulases used in denim production and in detergent formulations are able to digest cellulose-containing sorbents and other cellulose-based wastes contaminated either with crude oil or with radionuclides. This presentation describes the use of one such enzyme preparation (Rapidase trademark) for the degradation of cotton sorbents intentionally contaminated with low levels of plutonium. This is part of a feasibility study to determine if such treatments have a role in reducing the volume of low level and transuranic wastes to minimize the amount of radionuclide-contaminated waste that must be disposed of in secured storage areas

Enzymatic degradation of plutonium-contaminated cellulose products

International Nuclear Information System (INIS)

Heintz, C.E.; Rainwater, K.A.; Swift, L.M.; Barnes, D.L.; Worl, L.A.

1999-01-01

Enzyme solutions produced for commercial purposes unrelated to waste management have the potential for reducing the volume of wastes in streams containing cellulose, lipid and protein materials. For example, the authors have shown previously that cellulases used in denim production and in detergent formulations are able to digest cellulose-containing sorbents and other cellulose-based wastes contaminated either with crude oil or with uranium. This presentation describes the use of one such enzyme preparation (Rapidase trademark, manufactured by Genencor, Rochester, NY) for the degradation of cotton sorbents intentionally contaminated with low levels of plutonium. This is part of a feasibility study to determine if such treatments have a role in reducing the volume of low level and transuranic wastes to minimize the amount of radionuclide-contaminated waste destined for costly disposal options

Remote material handling in the Plutonium Immobilization Project. Revision 1

International Nuclear Information System (INIS)

Brault, J.R.

2000-01-01

With the downsizing of the US and Russian nuclear stockpiles, large quantities of weapons-usable plutonium in the US are being declared excess and will be disposed of by the Department of Energy Fissile Materials Disposition Program. To implement this program, DOE has selected the Savannah River Site (SRS) for the construction and operation of three new facilities: pit disassembly and conversion; mixed oxide fuel fabrication; and plutonium immobilization. The Plutonium Immobilization Project (PIP) will immobilize a portion of the excess plutonium in a hybrid ceramic and glass form containing high level waste for eventual disposal in a geologic repository. The PIP is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. Processing technology for the PIP is being developed jointly by the Lawrence Livermore National Laboratory and Westinghouse Savannah River Company. This paper will discuss development of the automated unpacking and sorting operations in the conversion area, and the automated puck and tray handling operations in the first stage immobilization area. Due to the high radiation levels and toxicity of the materials to be disposed of, the PIP will utilize automated equipment in a contained (glovebox) facility. Most operations involving plutonium-bearing materials will be performed remotely, separating personnel from the radiation source. Source term materials will be removed from the operations during maintenance. Maintenance will then be performed hands on within the containment using glove ports

GLASS FABRICATION AND PRODUCT CONSISTENCY TESTING OF LANTHANIDE BOROSHILICATE FRIT X COMPOSITION FOR PLUTONIUM DISPOSITION

Energy Technology Data Exchange (ETDEWEB)

Marra, J

2006-11-21

The Department of Energy Office of Environmental Management (DOE/EM) plans to conduct the Plutonium Disposition Project at the Savannah River Site (SRS) to disposition excess weapons-usable plutonium. A plutonium glass waste form is the preferred option for immobilization of the plutonium for subsequent disposition in a geologic repository. A reference glass composition (Lanthanide Borosilicate (LaBS) Frit B) was developed during the Plutonium Immobilization Program (PIP) to immobilize plutonium in the late 1990's. A limited amount of performance testing was performed on this baseline composition before efforts to further pursue Pu disposition via a glass waste form ceased. Recent FY05 studies have further investigated the LaBS Frit B formulation as well as development of a newer LaBS formulation denoted as LaBS Frit X. The objectives of this present task were to fabricate plutonium loaded LaBS Frit X glass and perform corrosion testing to provide near-term data that will increase confidence that LaBS glass product is suitable for disposal in the Yucca Mountain Repository. Specifically, testing was conducted in an effort to provide data to Yucca Mountain Project (YMP) personnel for use in performance assessment calculations. Plutonium containing LaBS glass with the Frit X composition with a 9.5 wt% PuO{sub 2} loading was prepared for testing. Glass was prepared to support Product Consistency Testing (PCT) at Savannah River National Laboratory (SRNL). The glass was thoroughly characterized using x-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS) prior to performance testing. A series of PCTs were conducted at SRNL using quenched Pu Frit X glass with varying exposed surface areas. Effects of isothermal and can-in-canister heat treatments on the Pu Frit X glass were also investigated. Another series of PCTs were performed on these different heat-treated Pu Frit X glasses. Leachates from all these PCTs

Assessment of plutonium in the Savannah River Site environment. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Carlton, W.H.; Evans, A.G.; Geary, L.A.; Murphy, C.E. Jr.; Pinder, J.E.; Strom, R.N.

1992-12-31

Plutonium in the Savannah River Site Environment is published as a part of the Radiological Assessment Program (RAP). It is the fifth in a series of eight documents on individual radioisotopes released to the environment as a result of Savannah River Site (SRS) operations. These are living documents, each to be revised and updated on a two-year schedule. This document describes the sources of plutonium in the environment, its release from SRS, environmental transport and ecological concentration of plutonium, and the radiological impact of SRS releases to the environment. Plutonium exists in the environment as a result of above-ground nuclear weapons tests, the Chernobyl accident, the destruction of satellite SNAP 9-A, plane crashes involving nuclear weapons, and small releases from reactors and reprocessing plants. Plutonium has been produced at SRS during the operation of five production reactors and released in small quantities during the processing of fuel and targets in chemical separations facilities. Approximately 0.6 Ci of plutonium was released into streams and about 12 Ci was released to seepage basins, where it was tightly bound by clay in the soil. A smaller quantity, about 3.8 Ci, was released to the atmosphere. Virtually all releases have occurred in F- and H-Area separation facilities. Plutonium concentration and transport mechanisms for the atmosphere, surface water, and ground water releases have been extensively studied by Savannah River Technology Center (SRTC) and ecological mechanisms have been studied by Savannah River Ecology Laboratory (SREL). The overall radiological impact of SRS releases to the offsite maximum individual can be characterized by a total dose of 15 mrem (atmospheric) and 0.18 mrem (liquid), compared with the dose of 12,960 mrem from non-SRS sources during the same period of time (1954--1989). Plutonium releases from SRS facilities have resulted in a negligible impact to the environment and the population it supports.

Plutonium Protection System (PPS). Volume 2. Hardware description. Final report

International Nuclear Information System (INIS)

Miyoshi, D.S.

1979-05-01

The Plutonium Protection System (PPS) is an integrated safeguards system developed by Sandia Laboratories for the Department of Energy, Office of Safeguards and Security. The system is designed to demonstrate and test concepts for the improved safeguarding of plutonium. Volume 2 of the PPS final report describes the hardware elements of the system. The major areas containing hardware elements are the vault, where plutonium is stored, the packaging room, where plutonium is packaged into Container Modules, the Security Operations Center, which controls movement of personnel, the Material Accountability Center, which maintains the system data base, and the Material Operations Center, which monitors the operating procedures in the system. References are made to documents in which details of the hardware items can be found

Qualitative chemical analysis of plutonium by Alpha spectroscopy

International Nuclear Information System (INIS)

Ramirez G, J Qumica.J.

1994-01-01

In this work the separation and purification of plutonium from irradiated uranium was done. The plutonium, produced by the irradiation of uranium in a nuclear reactor and the β decay of 239 Np, was stabilized to Pu +4 with sodium nitrite. Plutonium was separated from the fission products and uranium by ion exchange using the resin Ag 1 X 8. It was electrodeposited on stainless steel discs and the alpha radioactivity of plutonium was measured in a surface barrier detector. The results showed that plutonium was separated with a radiochemical purity higher than 99 %. (Author)

How not to reduce plutonium stocks. The danger of MOX-fuelled nuclear reactors

International Nuclear Information System (INIS)

1999-01-01

Plutonium is a radioactive by-product of nuclear reactor operation and one of the most toxic substances known. The world would be a safer place if the governments of countries with stocks of it, including Britain, would adopt effective policies for reducing and managing them. Two recent authoritative reports recommend that the British government take urgent action to reduce its 'civil' plutonium stock - currently one quarter of the world's total and set to rise to about two-thirds by the year 2010. The March 1999 House of Lords report, Management of Nuclear Waste, concludes that British government policy on plutonium 'should be the maintenance of the minimum strategic stock, and the declaration of the remainder as waste'. A report from the Royal Society, Britain's main learned society, meanwhile states that: 'In addition to disposing of some of the plutonium already in the stockpile, steps should be taken to reduce the amount added to it each year, primarily by reducing the amount of reprocessing carried out'. The government's reply to the House of Lords is expected to be followed by a public consultation before changes in legislation are proposed. But, at the same time, the government is considering an application from British Nuclear Fuels Limited (BNFL), the government-owned company which separates plutonium from spent nuclear fuel rods, for a licence to operate a new plant at Sellafield in Cumbria to produce mixed-oxide (MOX) nuclear fuel from its plutonium stockpile. The nuclear industry justifies the Sellafield MOX plant as one way of reducing plutonium stocks. But critics point out that this is not a rational way to manage plutonium. This briefing aims to contribute to an informed debate during the current flurry of British government nuclear policymaking by explaining why. (author)

Search for plutonium salt deposits in the plutonium extraction batteries of the Marcoule plant (1963); Recherche de depots de sels de plutonium dans les batteries d'extraction du plutonium de l'usine de Marcoule (1963)

Energy Technology Data Exchange (ETDEWEB)

Bouzigues, H; Reneaud, J M [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1963-07-01

This report describes a method and a special apparatus making it possible to detach the insoluble plutonium salt deposits in the extraction chain of an irradiated fuel treatment plant. The process chosen allows the detection, in the extraction batteries or in the highly active chemical engineering equipment, of plutonium quantities of a few grains. After four years operation it has been impossible to detect measurable quantities of plutonium in any part of the extraction chain. The results have been confirmed by visual examinations carried out with a specially constructed endoscope. (authors) [French] Ce rapport decrit une methode et un montage special permettant de detacher les accumulations de sels de plutonium insolubles dans les chaines d'extraction d'une usine de traitement de combustible irradie. Le procede retenu permet de reperer, dans des batteries d'extraction ou dans l'appareillage de genie chimique fortement actif, des masses de plutonium de quelques grammes. Apres quatre annees de fonctionnement, il n'a pas ete possible de deceler des quantites ponderables de plutonium en aucun endroit de la chaine d'extraction. Ces resultats ont ete confirmes par les examens visuels effectues a l'aide d'un endoscope concu specialement pour cet usage. (auteurs)

Burning weapons-grade plutonium in reactors

International Nuclear Information System (INIS)

Newman, D.F.

1993-06-01

As a result of massive reductions in deployed nuclear warheads, and their subsequent dismantlement, large quantities of surplus weapons- grade plutonium will be stored until its ultimate disposition is achieved in both the US and Russia. Ultimate disposition has the following minimum requirements: (1) preclude return of plutonium to the US and Russian stockpiles, (2) prevent environmental damage by precluding release of plutonium contamination, and (3) prevent proliferation by precluding plutonium diversion to sub-national groups or nonweapons states. The most efficient and effective way to dispose of surplus weapons-grade plutonium is to fabricate it into fuel and use it for generation of electrical energy in commercial nuclear power plants. Weapons-grade plutonium can be used as fuel in existing commercial nuclear power plants, such as those in the US and Russia. This recovers energy and economic value from weapons-grade plutonium, which otherwise represents a large cost liability to maintain in safeguarded and secure storage. The plutonium remaining in spent MOX fuel is reactor-grade, essentially the same as that being discharged in spent UO 2 fuels. MOX fuels are well developed and are currently used in a number of LWRs in Europe. Plutonium-bearing fuels without uranium (non-fertile fuels) would require some development. However, such non-fertile fuels are attractive from a nonproliferation perspective because they avoid the insitu production of additional plutonium and enhance the annihilation of the plutonium inventory on a once-through fuel cycle

Plutonium speciation affected by environmental bacteria

International Nuclear Information System (INIS)

Neu, M.P.; Icopini, G.A.; Boukhalfa, H.

2005-01-01

Plutonium has no known biological utility, yet it has the potential to interact with bacterial cellular and extracellular structures that contain metal-binding groups, to interfere with the uptake and utilization of essential elements, and to alter cell metabolism. These interactions can transform plutonium from its most common forms, solid, mineral-adsorbed, or colloidal Pu(IV), to a variety of biogeochemical species that have much different physico-chemical properties. Organic acids that are extruded products of cell metabolism can solubilize plutonium and then enhance its environmental mobility, or in some cases facilitate plutonium transfer into cells. Phosphate- and carboxylate-rich polymers associated with cell walls can bind plutonium to form mobile biocolloids or Pu-laden biofilm/mineral solids. Bacterial membranes, proteins or redox agents can produce strongly reducing electrochemical zones and generate molecular Pu(III/IV) species or oxide particles. Alternatively, they can oxidize plutonium to form soluble Pu(V) or Pu(VI) complexes. This paper reviews research on plutonium-bacteria interactions and closely related studies on the biotransformation of uranium and other metals. (orig.)

Transplutonium elements production program: extraction chromatographic process for plutonium irradiated targets

International Nuclear Information System (INIS)

Bourges, J.; Madic, C.; Koehly, G.

1980-01-01

The treatment of irradiated plutonium targets by extraction chromatography allowed the purification of the isotopes 243 Am and 244 Cm on the scale of few tens of grams. This process proved to be extremely simple and flexible, and yielded results which are reproducible in time. The chief advantage of the TBP process over the HDEHP process in high and medium activity conditions lies in the rapid absorption/desorption kinetics of the elements to be purified and in the separation of americium from curium, which largely offsets its lower selectivity for lanthanide elements. it is certainly possible to improve the performance of this process by: a) optimization of the characteristics of the stationary phase, b) improvement in the filling technique and in hydraulic operation of the columns, c) on-line analysis of americium (the key element in actinide/lanthanide separation) in the eluate. The application of extraction chromatography with HD(DiBM)P to the purification of 243 Am of the end of treatment makes the process more consistent, eliminates the delicate stages implemented in hot cell, and considerably improves final product quality

Volume reduction and plutonium recovery in alpha wastes by cryogenic crushing and lixiviation

International Nuclear Information System (INIS)

Arnal, T.; Pajot, J.

1986-06-01

The industry of plutonium generates solid alpha wastes of medium activity called ''technological wastes''. They are mainly produced during the fabrication and reprocessing of nuclear reactor fuels and they are of a wide variety i.e: vinyl bags, gloves, glass, steel materials used in glove box operation, etc... These wastes contain relevant residual quantities of uranium and plutonium in the form of oxides or nitrates, reaching up to several dozen grams per cubic meter. Up to the beginning of the eighties, they were conditionned without any treatment and stored as such on the production site. However, for an economic and safe storage, recovering of the plutonium contained in these waste streams and reduction of their volume is of obvious importance. At the plutonium ''Complexe de Fabrication des Combustibles de Cadarache'' was developed a new technical solution of this problem that combines cryogenic crushing of the solid waste and plutonium recovery from the crushed material by chemical lixiviation. The first results obtained in applying this system on the industrial scale are reported briefly

Plutonium in nature; Le plutonium dans la nature

Energy Technology Data Exchange (ETDEWEB)

Madic, C.

1994-12-31

Plutonium in nature comes from natural sources and anthropogenic ones. Plutonium at the earth surface comes principally from anthropogenic sources. It is easily detectable in environment. The plutonium behaviour in environment is complex. It seems necessary for the future to reduce releases in environment, to improve predictive models of plutonium behaviour in geosphere, to precise biological impact of anthropogenic plutonium releases.

Japan's spent fuel and plutonium management challenge

International Nuclear Information System (INIS)

Katsuta, Tadahiro; Suzuki, Tatsujiro

2011-01-01

Japan's commitment to plutonium recycling has been explicitly stated in its long-term program since 1956. Despite the clear cost disadvantage compared with direct disposal or storage of spent fuel, the Rokkasho reprocessing plant started active testing in 2006. Japan's cumulative consumption of plutonium has been only 5 tons to date and its future consumption rate is still uncertain. But once the Rokkasho reprocessing plant starts its full operation, Japan will separate about 8 tons of plutonium annually. Our analysis shows that, with optimum use of available at-reactor and away-from-reactor storage capacity, there would be no need for reprocessing until the mid-2020s. With an additional 30,000 tons of away-from-reactor (AFR) spent-fuel storage capacity reprocessing could be avoided until 2050. Deferring operation of the Rokkasho plant, at least until the plutonium stockpile had been worked down to the minimum required level, would also minimize international concern about Japan's plutonium stockpile. The authors are happy to acknowledge Frank von Hippel, Harold Feiveson, Jungming Kang, Zia Mian, M.V. Ramana, and other IPFM members, as well as the generous grant from the MacArthur Foundation for helping make this research possible.

Production and industrial applications of plutonium

International Nuclear Information System (INIS)

Lafontaine, I.

1975-01-01

In the present document, an evaluation is made of the quantities of plutonium which will be produced by all nuclear power stations until 1980 and of the plants which are actually able to treat burned nuclear fuels with a view to recover this material. As soon as the plutonium, in the form of dioxide, becomes available, it is transported towards fuel rod and fuel assembly fabrication plants, in containers especially commissioned by the competent authorities; these containers have to resist succesfully to very severe tests specified by the International Atomic Energy Agency in Vienna. Additional protections are foreseen during transportation, namely to prevent an attempt of nuclear materials' diversion. The plant for fabrication of nuclear fuels is designed on the basis of extensive safety studies. Indeed, various probabilities of accidents and associated risks have been evaluated and have given rise namely to safe working rules and the provision of elements of protection [fr

Technical report for generic site add-on facility for plutonium polishing

International Nuclear Information System (INIS)

1998-06-01

The purpose of this report is to provide environmental data and reference process information associated with incorporating plutonium polishing steps (dissolution, impurity removal, and conversion to oxide powder) into the genetic-site Mixed-Oxide Fuel Fabrication Facility (MOXFF). The incorporation of the plutonium polishing steps will enable the removal of undesirable impurities, such as gallium and americium, known to be associated with the plutonium. Moreover, unanticipated impurities can be removed, including those that may be contained in (1) poorly characterized feed materials, (2) corrosion products added from processing equipment, and (3) miscellaneous materials contained in scrap recycle streams. These impurities will be removed to the extent necessary to meet plutonium product purity specifications for MOX fuels. Incorporation of the plutonium polishing steps will mean that the Pit Disassembly and Conversion Facility (PDCF) will need to produce a plutonium product that can be dissolved at the MOXFF in nitric acid at a suitable rate (sufficient to meet overall production requirements) with the minimal usage of hydrofluoric acid, and its complexing agent, aluminum nitrate. This function will require that if the PDCF product is plutonium oxide powder, that powder must be produced, stored, and shipped without exceeding a temperature of 600 C

Calculating the plutonium in spent fuel elements

International Nuclear Information System (INIS)

Barnham, Keith

1992-01-01

Many members of the public are concerned about plutonium. They are worried about its environmental, health and proliferation risks. Fundamental to all such considerations are two related questions: how much plutonium do nuclear reactors produce ? and how accurately do the relevant authorities know these production figures ? These two questions have been studied with particular reference to the UK civil Magnox reactors. In 1990 these were still the only UK civil reactors whose spent fuel had been reprocessed to extract plutonium in routine production. It has not been possible to conclude that the relevant government industry and safeguard authorities are aware of how much plutonium these reactors produce and that the figures are known to the highest achievable accuracy. To understand why, this chapter will outline some of the history of the attempts to get answers to these two questions. (author)

Los Alamos DP West Plutonium Facility decontamination project, 1978-1981

International Nuclear Information System (INIS)

Garde, R.; Cox, E.J.; Valentine, A.M.

1982-09-01

The DP West Plutonium Facility operated by the Los Alamos National Laboratory, Los Alamos, New Mexico was decontaminated between April 1978 and April 1981. The facility was constructed in 1944 to 1945 to produce plutonium metal and fabricate parts for nuclear weapons. It was continually used as a plutonium processing and research facility until mid-1978. Decontamination operations included dismantling and removing gloveboxes and conveyor tunnels; removing process systems, utilities, and exhaust ducts; and decontaminating all remaining surfaces. This report describes glovebox and conveyor tunnel separations, decontamination techniques, health and safety considerations, waste management procedures, and costs of the operation

The medical treatment of the plutonium-contaminated wound

International Nuclear Information System (INIS)

Liu Yanling

1988-01-01

Some experiences in medical treatment of plutonium-contaminated wound gained through the animal experiments and clinical practices were described. For the treatment of plutonium-contaminated wound, much attention should be devoted to the early emergency measure at accident site. In some case when the surgical interference is needed, radioactivity in local wound and in regional lympho-nodes of wound, firstly, must be determined and the contaminated area must be demonstrated before operation. Selecting a proper approach for anaesthesia is one of the critical factors for successfulness of operation. The operation should be performed under the coordination of monitoring workers. During operation, the rules for decontamination should be followed to avoid recontamination. In addition to conventional administration of chelating agents, the local application of such agents during and after operation is a better supplementary therapeutic procedure for some cases when the residual amount of plutonium in wound is less and any other surgical procedures will not be performed further

Decontaminaion of metals containing plutonium and americium

International Nuclear Information System (INIS)

Seitz, M.G.; Gerding, T.J.; Steindler, M.J.

1979-06-01

Melt-slagging (melt-refining) techniques were evaluated as a decontamination and consolidation step for metals contaminated with oxides of plutonium and americium. Experiments were performed in which mild steel, stainless steel, and nickel contaminated with oxides of plutonium and americium were melted in the presence of silicate slags of various compositions. The metal products were low in contamination, with the plutonium and americium strongly fractionated to the slags. Partition coefficients (plutonium in slag/plutonium in steel) of 7 x 10 6 were measured with boro-silicate slag and of 3 x 10 6 with calcium, magnesium silicate slag. Decontamination of metals containing as much as 14,000 ppM plutonium appears to be as efficient as for metals with plutonium levels of 400 ppM. Staged extraction, that is, a remelting of processed metal with clean slag, results in further decontamination of the metal. The second extraction is effective with either resistance-furnace melting or electric-arc melting. Slag adhering to the metal ingots and in defects within the ingots is in the important contributors to plutonium retained in processed metals. If these sources of plutonium are controlled, the melt-refining process can be used on a large scale to convert highly contaminated metals to homogeneous and compact forms with very low concentrations of plutonium and americium. A conceptual design of a melt-refining process to decontaminate plutonium- and americium-contaminated metals is described. The process includes single-stage refining of contaminated metals to produce a metal product which would have less than 10 nCi/g of TRU-element contamination. Two plant sizes were considered. The smaller conceptual plant processes 77 kg of metal per 8-h period and may be portable.The larger one processes 140 kg of metal per 8-h period, is stationary, and may be near te maximum size that is practical for a metal decontamination process

Recent improvements in plutonium gamma-ray analysis using MGA

International Nuclear Information System (INIS)

Ruhter, W.D.; Gunnink, R.

1992-06-01

MGA is a gamma-ray spectrum analysis program for determining relative plutonium isotopic abundances. It can determine plutonium isotopic abundances better than 1% using a high-resolution, low-energy, planar germanium detector and measurement times ten minutes or less. We have modified MGA to allow determination of absolute plutonium isotopic abundances in solutions. With calibration of a detector using a known solution concentration in a well-defined sample geometry, plutonium solution concentrations can be determined. MGA can include analysis of a second spectrum of the high-energy spectrum to include determination of fission product abundances relative to total plutonium. For the high-energy gamma-ray measurements we have devised a new hardware configuration, so that both the low- and high-energy gamma-ray detectors are mounted in a single cryostat thereby reducing weight and volume of the detector systems. We describe the detector configuration, and the performance of the MGA program for determining plutonium concentrations in solutions and fission product abundances

Expected behavior of plutonium in the IFR fuel cycle

International Nuclear Information System (INIS)

Steunenberg, R.K.; Johnson, I.

1985-01-01

The Integral Fast Reactor (IFR) is a metal-fueled, sodium-cooled reactor that will consist initially of a U-Zr alloy core in which the enriched uranium will be replaced gradually by plutonium bred in a uranium blanket. The plutonium is concentrated to the required level by extraction from the molten blanket material with a CaCl 2 -BaCl 2 salt containing MgCl 2 as an oxidant (halide slagging). The CaCl 2 -BaCl 2 salt containing dissolved PuCl 3 and UCl 3 is added to the core process where fission products are removed by electrorefining, using a liquid cadmium anode, a metal cathode, and a LiCl-NaCl-CaCl 2 -BaCl 2 molten salt electrolyte. The product is recovered as a metallic deposit on the cathode. The halide slagging step is operated at about 1250 0 and the electrorefining step at about 450 0 C. These processes are expected to give low fission-product decontamination factors of the order of 100

Disposition of plutonium-239 via production of fission molybdenum-99

Energy Technology Data Exchange (ETDEWEB)

Mushtaq, A., E-mail: muahtaq_a1953@hotmail.co [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

2011-04-15

A heritage of physical consequences of the U.S.-Soviet arms race has accumulated, the weapons-grade plutonium (WPu), which will become excess as a result of the dismantlement of the nuclear weapons under the arms reduction agreements. Disposition of Pu has been proposed by mixing WPu with high-level radioactive waste with subsequent vitrification into large, highly radioactive glass logs or fabrication into mixed oxide fuel with subsequent irradiation in existing light water reactors. A potential option may be the production of medical isotope molybdenum-99 by using Pu-239 targets.

Continuous precipitation process of plutonium salts

International Nuclear Information System (INIS)

Richard, P.

1967-03-01

This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [fr

Nondestructive analysis of plutonium contaminated soil

International Nuclear Information System (INIS)

Smith, H.E.; Taylor, L.H.

1977-01-01

Plutonium contaminated soil is currently being removed from a covered liquid waste disposal trench near the Pu Processing facility on the Hanford Project. This soil with the plutonium is being mined using remote techniques and equipment. The mined soil is being packaged for placement into retrievable storage, pending possible recovery. To meet the requirements of criticality safety and materials accountability, a nondestructive analysis program has been developed to determine the quantity of plutonium in each packing-storage container. This paper describes the total measurement program: equipment systems, calibration techniques, matrix assumption, instrument control program and a review of laboratory operating experience

Continuous precipitation process of plutonium salts; Procede continu de precipitation des sels de plutonium

Energy Technology Data Exchange (ETDEWEB)

Richard, P [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-03-01

This work concerns the continuous precipitation process of plutonium oxalate. Investigations about the solubility of different valence states in nitric-oxalic and in nitric-sulfuric-oxalic medium lead to select the precipitation process of tetravalent plutonium oxalate. Settling velocity and granulometry of tetravalent oxalate plutonium have been studied with variation of several precipitation parameters such as: temperature, acidity, excess of oxalic acid and aging time. Then are given test results of some laboratory continuous apparatus. Conditions of operation with adopted tubular apparatus are defined in conclusion. A flow-sheet is given for a process at industrial scale. (author) [French] Cette etude porte sur la precipitation continue de l'oxalate de plutonium. L'etude de la solubilite des differentes valences du plutonium dans des milieux acides nitrique-oxalique, puis nitrique-sulfurique-oxalique conduit a choisir la precipitation de l'oxalate de plutonium tetravalent. L'etude porte ensuite sur la sedimentation et la granulometrie de l'oxalate de Pu{sup 4+} obtenue en faisant varier differents parametres de la precipitation : la temperature, l'acidite, l'exces oxalique et le temps de murissement. La derniere partie traite des resultats obtenus avec plusieurs types d'appareils continus essayes au laboratoire. En conclusion sont donnees les conditions de marche de l'appareil tubulaire adopte, ainsi qu'une extrapolation a l'echelle industrielle sous forme d'un flow-sheet. (auteur)

Continuous plutonium(IV) oxalate precipitation, filtration, and calcination process. [From product streams from Redox, Purex, or Recuplex solvent extraction plants

Energy Technology Data Exchange (ETDEWEB)

Beede, R L

1956-09-27

A continuous plutonium (IV) oxalate precipitation, filtration, and calcination process has been developed. Continuous and batch decomposition of the oxalate in the filtrates has been demonstrated. The processes have been demonstrated in prototype equipment. Plutonium (IV) oxalate was precipitated continuously at room temperature by the concurrent addition of plutonium (IV) nitrate feed and oxalic acid into the pan of a modified rotary drum filter. The plutonium (IV) oxalate was calcined to plutonium dioxide, which could be readily hydrofluorinated. Continuous decomposition of the oxalate in synthetic plutonium (IV) oxalate filtrates containing plutonium (IV) oxalate solids was demonstrated using co-current flow in a U-shaped reactor. Feeds containing from 10 to 100 g/1 Pu, as plutonium (IV) nitrate, and 1.0 to 6.5 M HNO/sub 3/, respectively, can be processed. One molar oxalic acid is used as the precipitant. Temperatures of 20 to 35/sup 0/C for the precipitation and filtration are satisfactory. Plutonium (IV) oxalate can be calcined at 300 to 400/sup 0/C in a screw-type drier-calciner to plutonium dioxide and hydrofluorinated at 450 to 550/sup 0/C. Plutonium dioxide exceeding purity requirements has been produced in the prototype equipment. Advantages of continuous precipitation and filtration are: uniform plutonium (IV) oxalate, improved filtration characteristics, elimination of heating and cooling facilities, and higher capacities through a single unit. Advantages of the screw-type drier-calciner are the continuous production of an oxide satisfactory for feed for the proposed plant vibrating tube hydrofluorinator, and ease of coupling continuous precipitation and filtration to this proposed hydrofluorinator. Continuous decomposition of oxalate in filtrates offers advantages in decreasing filtrate storage requirements when coupled to a filtrate concentrator. (JGB)

Plutonium uniqueness

International Nuclear Information System (INIS)

Silver, G.L.

1984-01-01

A standard is suggested against which the putative uniqueness of plutonium may be tested. It is common folklore that plutonium is unique among the chemical elements because its four common oxidation states can coexist in the same solution. Whether this putative uniqueness appears only during transit to equilibrium, or only at equilibrium, or all of the time, is not generally made clear. But while the folklore may contain some truth, it cannot be put to test until some measure of 'uniqueness' is agreed upon so that quantitative comparisons are possible. One way of measuring uniqueness is as the magnitude of the product of the mole fractions of the element at equilibrium. A 'coexistence index' is defined and discussed. (author)

On the Burning of Plutonium Originating from Light Water Reactor Use in a Fast Molten Salt Reactor—A Neutron Physical Study

Directory of Open Access Journals (Sweden)

Bruno Merk

2015-11-01

Full Text Available An efficient burning of the plutonium produced during light water reactor (LWR operation has the potential to significantly improve the sustainability indices of LWR operations. The work offers a comparison of the efficiency of Pu burning in different reactor configurations—a molten salt fast reactor, a LWR with mixed oxide (MOX fuel, and a sodium cooled fast reactor. The calculations are performed using the HELIOS 2 code. All results are evaluated against the plutonium burning efficiency determined in the Consommation Accrue de Plutonium dans les Réacteurs à Neutrons RApides (CAPRA project. The results are discussed with special view on the increased sustainability of LWR use in the case of successful avoidance of an accumulation of Pu which otherwise would have to be forwarded to a final disposal. A strategic discussion is given about the unavoidable plutonium production, the possibility to burn the plutonium to avoid a burden for the future generations which would have to be controlled.

Recovery of americium-241 from aged plutonium metal

International Nuclear Information System (INIS)

Gray, L.W.; Burney, G.A.; Reilly, T.A.; Wilson, T.W.; McKibben, J.M.

1980-12-01

About 5 kg of ingrown 241 Am was recovered from 850 kg of aged plutonium using a process developed specifically for Savannah River Plant application. The aged plutonium metal was first dissolved in sulfamic acid. Sodium nitrite was added to oxidize the plutonium to Pu(IV) and the residual sulfamate ion was oxidized to nitrogen gas and sulfate. The plutonium and americium were separated by one cycle of solvent extraction. The recovered products were subsequently purified by cation exchange chromatography, precipitated as oxalates, and calcined to the oxides. Plutonium processng was routine. Before cation exchange purification, the aqueous americium solution from solvent extraction was concentrated and stripped of nitric acid. More than 98% of the 241 Am was then recovered from the cation exchange column where it was effectively decontaminated from all major impurities except nickel and chromium. This partially purified product solution was concentrated further by evaporation and then denitrated by reaction with formic acid. Individual batches of americium oxalate were then precipitated, filtered, washed, and calcined. About 98.5% of the americium was recovered. The final product purity averaged 98% 241 AmO 2 ; residual impurities were primarily lead and nickel

Sample preparation automation for dosing plutonium in urine

International Nuclear Information System (INIS)

Jeanmaire, Lucien; Ballada, Jean; Ridelle Berger, Ariane

1969-06-01

After having indicated that dosing urinary plutonium by using the Henry technique can be divided into three stages (plutonium concentration by precipitation, passing the solution on an anionic resin column and plutonium elution, and eluate evaporation to obtain a source of which the radioactivity is measured), and recalled that the automation of the second stage has been reported in another document, this document describes the automation of the first stage, i.e. obtaining from urine a residue containing the plutonium, and sufficiently mineralized to be analyzed by means of ion exchanging resins. Two techniques are proposed, leading to slightly different devices. The different operations to be performed are indicated. The different components of the apparatus are described: beakers, hot plate stirrers, reagent circuits, a system for supernatant suction, and a control-command circuit. The operation and use are then described, and results are given

Plutonium-239 production rate study using a typical fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Faghihi, F. [Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Havasi, H.; Amin-Mozafari, M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of)

2008-05-15

The purpose of the present paper is to compute fissile {sup 239}Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m{sup 2}) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate {sup 239}Pu production rate. Produced {sup 239}Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, {sup 239}Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type.

Plutonium Proliferation: The Achilles Heel of Disarmament

International Nuclear Information System (INIS)

Leventhal, Paul

2001-01-01

Plutonium is a byproduct of nuclear fission, and it is produced at the rate of about 70 metric tons a year in the world's nuclear power reactors. Concerns about civilian plutonium ran high in the 1970s and prompted enactment of the Nuclear Non-Proliferation Act of 1978 to give the United States a veto over separating plutonium from U.S.-supplied uranium fuel. Over the years, however, so-called reactor-grade plutonium has become the orphan issue of nuclear non-proliferation, largely as a consequence of pressures from plutonium-separating countries. The demise of the fast breeder reactor and the reluctance of utilities to introduce plutonium fuel in light-water reactors have resulted in large surpluses of civilian, weapons-usable plutonium, which now approach in size the 250 tons of military plutonium in the world. Yet reprocessing of spent fuel for recovery and use of plutonium proceeds apace outside the United States and threatens to overwhelm safeguards and security measures for keeping this material out of the hands of nations and terrorists for weapons. A number of historical and current developments are reviewed to demonstrate that plutonium commerce is undercutting efforts both to stop the spread of nuclear weapons and to work toward eliminating existing nuclear arsenals. These developments include the breakdown of U.S. anti-plutonium policy, the production of nuclear weapons by India with Atoms-for-Peace plutonium, the U.S.-Russian plan to introduce excess military plutonium as fuel in civilian power reactors, the failure to include civilian plutonium and bomb-grade uranium in the proposed Fissile Material Cutoff Treaty, and the perception of emerging proliferation threats as the rationale for development of a ballistic missile defense system. Finally, immobilization of separated plutonium in high-level waste is explored as a proliferation-resistant and disarmament-friendly solution for eliminating excess stocks of civilian and military plutonium.

The uranium-plutonium breeder reactor fuel cycle

International Nuclear Information System (INIS)

Salmon, A.; Allardice, R.H.

1979-01-01

All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

Preparation and application of potassium and sodium titanate for removal of plutonium from basic solution

International Nuclear Information System (INIS)

Patil, Prashant; Pathak, Sachin S.; Pius, I.C.; Mukerjee, S.K.

2014-01-01

In PUREX process, after extraction and stripping of uranium and plutonium, the extractant, tributyl phosphate is usually washed with sodium carbonate solution before reuse for the removal of radiolytic/hydrolytic degradation products of TBP and small amounts of HNO 3 , uranium and plutonium goes into aqueous phase during carbonate washings. Partial neutralization of carbonate by the acid converts it to bicarbonate. Removal of plutonium from such sodium carbonate/bicarbonate streams facilitates their disposal. In the present work, studies were carried out to prepare inorganic ion-exchangers such as potassium and sodium titanates for their application as ion-exchange material. It is essential to prepare these materials in granular form to obtain good liquid flow property for ion exchange column operations, however, it is also important that the final product is having good surface area and porosity so that they may exhibit good ion exchange capacity

Design of the Laboratory-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

Energy Technology Data Exchange (ETDEWEB)

Lumetta, Gregg J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meier, David E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tingey, Joel M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Orton, Robert D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rapko, Brian M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smart, John E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-05-01

This report describes a design for a laboratory-scale capability to produce plutonium oxide (PuO₂) for use in identifying and validating nuclear forensics signatures associated with plutonium production, as well as for use as exercise and reference materials. This capability will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including PuO₂ dissolution, purification of the Pu by ion exchange, precipitation, and re-conversion to PuO₂ by calcination.

Plutonium Disposition Now exclamation point

International Nuclear Information System (INIS)

Buckner, M.R.

1995-01-01

A means for use of existing processing facilities and reactors for plutonium disposition is described which requires a minimum capital investment and allows rapid implementation. The scenario includes interim storage and processing under IAEA control, and fabrication into MOX fuel in existing or planned facilities in Europe for use in operating reactors in the two home countries. Conceptual studies indicate that existing Westinghouse four-loop designs can safety dispose of 0.94 MT of plutonium per calendar year. Thus, it would be possible to consume the expected US excess stockpile of about 50 MT in two to three units of this type, and it is highly likely that a comparable amount of the FSU excess plutonium could be deposed of in a few VVER-1000's. The only major capital project for this mode of plutonium disposition would be the weapons-grade plutonium processing which could be done in a dedicated international facility or using existing facilities in the US and FSU under IAEA control. This option offers the potential for quick implementation at a very low cost to the governments of the two countries

Plutonium

International Nuclear Information System (INIS)

Watson, G.M.

1976-01-01

Discovery of the neutron made it easy to create elements which do not exist in nature. One of these is plutonium, and its isotope with mass number 239 has nuclear properties which make it both a good fuel for nuclear power reactors and a good explosive for nuclear weapons. Since it was discovered during a war the latter characteristic was put to use, but it is now evident that use of plutonium in a particular kind of nuclear reactor, the fast breeder reactor, will allow the world's resources of uranium to last for millennia as a major source of energy. Plutonium is very radiotoxic, resembling radium in this respect. Therefore the widespread introduction of fast breeder reactors to meet energy demands can be contemplated only after assurances on two points; that adequate control of the radiological hazard resulting from the handling of very large amounts of plutonium can be guaranteed, and that diversion of plutonium to illicit use can be prevented. The problems exist to a lesser degree already, since all types of nuclear reactor produce some plutonium. Some plutonium has already been dispersed in the environment, the bulk of it from atmospheric tests of nuclear weapons. (author)

Improved plutonium consumption in a pressurised water reactor

International Nuclear Information System (INIS)

Puill, A.; Bergeron, J.

1995-01-01

Our goal is to improve plutonium consumption in a dedicated PWR while limiting the production of minor actinides. For lack of proving the system's reliability, we stay in reasonable configurations in which power capacity is maintained. Three ways are investigated in determining the fuel assembly: (a) standard geometry with mixed oxide in enriched uranium base; (b) standard geometry with plutonium oxide included in an inert matrix; (c) new geometry with special all-plutonium consumption varies from 50 kg/TWeh (a) up to 140 kg/TWeh (b) (upper point). The new geometry with special all plutonium rods mixed with standard uranium rods appears promising with a burning rate of 92 kg/TWeh for a production of minor actinides of 10 kg/TWeh. (authors). 3 refs., 3 figs., 4 tabs

Disposition of plutonium from dismantled warheads: Belgonucleaire's proposal

International Nuclear Information System (INIS)

Haas, D.; Vanderborck, Y.; Vandenberg, C.; Vliet, J. van

1996-01-01

Set up in 1957 by Union Miniere, Belgonucleaire (BN) has been working since its origin in the field of plutonium and has accumulated extensive experience in the design and fabrication of mixed-oxide (MOX) fuel, as well as in the fuel management and licensing of MOX cores for fast breeder reactors (FBRs), pressurized water reactors (PWRs), and boiling water reactors (BWRs). Although BN's MOX plant first went into operation in 1973 (it was used for FBRs and light water reactor (LWR) demonstration fuel fabrication), industrial production started in 1986. The MOX fuel produced (280 t heavy metal (HM) until the end of 1995) has been loaded in PWRs and BWRs in four countries in Europe: France, Germany, Switzerland, and Belgium. They propose the development of MOX plants as the means for disposal of plutonium from warhead disassembly

International collaborations about fuel studies for reactor recycling of military quality plutonium

International Nuclear Information System (INIS)

Bernard, H.; Chaudat, J.P.

1997-01-01

In November 1992, an agreement was signed between the French and Russian governments to use in Russia and for pacific purposes the plutonium recovered from the Russian nuclear weapons dismantling. This plutonium will be transformed into mixed oxide fuels (MOX) for nuclear power production. The French Direction of Military Applications (DAM) of the CEA is the operator of the French-Russian AIDA program. The CEA Direction of Fuel Cycle (DCC) and Direction of Nuclear Reactors (DRN) are involved in the transformation of metallic plutonium into sinterable oxide powder for MOX fuel manufacturing. The Russian TOMOX (Treatment of MOX powder Metallic Objects) and DEMOX (MOX Demonstration) plants will produce the MOX fuel assemblies for the 4 VVER 1000 reactors of Balakovo and the fast BN 600 reactor. The second part of the program will involve the German Siemens and GRS companies for the safety studies of the reactors and fuel cycle plants. The paper gives also a brief analysis of the US policy concerning the military plutonium recycling. (J.S.)

Plutonium recovery from spent reactor fuel by uranium displacement

Science.gov (United States)

Ackerman, J.P.

1992-03-17

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Plutonium recovery from spent reactor fuel by uranium displacement

International Nuclear Information System (INIS)

Ackerman, J.P.

1992-01-01

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished

Probability of production of mobile plutonium in environments of soil and sediment

International Nuclear Information System (INIS)

Mahara, Y.; Kudo, A.

1998-01-01

Mobile plutonium was found in the bottom sediment in the Nishiyama reservoir in Nagasaki after more than 40 years from deposition of local fallout released in the explosion of the A-bomb in 1945. Less than 10% of total deposited plutonium had turned into a mobile form in the bottom environment of the reservoir. The environmental conditions at bottom sediment is expected to be rich organic materials and high bacterial population under anaerobic conditions. Anaerobic bacteria have a high ability to uptake plutonium into cell during their growth. The K d of plutonium to living bacteria is 20 times greater than the dead bacteria under anaerobic conditions. The results of field observations combined with empirical laboratory tests indicate that mobile plutonium in soil and sediment may be affected not only by binding with dissolved natural organic materials but also by the number of living anaerobic bacteria. (orig.)

Plutonium Immobilization Can Loading Equipment Review

International Nuclear Information System (INIS)

Kriikku, E.; Ward, C.; Stokes, M.; Randall, B.; Steed, J.; Jones, R.; Hamilton, L.

1998-05-01

This report lists the operations required to complete the Can Loading steps on the Pu Immobilization Plant Flow Sheets and evaluates the equipment options to complete each operation. This report recommends the most appropriate equipment to support Plutonium Immobilization Can Loading operations

Weapons-grade plutonium dispositioning. Volume 3: A new reactor concept without uranium or thorium for burning weapons-grade plutonium

International Nuclear Information System (INIS)

Ryskamp, J.M.; Schnitzler, B.G.; Fletcher, C.D.

1993-06-01

The National Academy of Sciences (NAS) requested that the Idaho National Engineering Laboratory (INEL) examine concepts that focus only on the destruction of 50,000 kg of weapons-grade plutonium. A concept has been developed by the INEL for a low-temperature, low-pressure, low-power density, low-coolant-flow-rate light water reactor that destroys plutonium quickly without using uranium or thorium. This concept is very safe and could be designed, constructed, and operated in a reasonable time frame. This concept does not produce electricity. Not considering other missions frees the design from the paradigms and constraints used by proponents of other dispositioning concepts. The plutonium destruction design goal is most easily achievable with a large, moderate power reactor that operates at a significantly lower thermal power density than is appropriate for reactors with multiple design goals. This volume presents the assumptions and requirements, a reactor concept overview, and a list of recommendations. The appendices contain detailed discussions on plutonium dispositioning, self-protection, fuel types, neutronics, thermal hydraulics, off-site radiation releases, and economics

Chemical species of plutonium in Hanford radioactive tank waste

International Nuclear Information System (INIS)

Barney, G.S.

1997-01-01

Large quantities of radioactive wastes have been generated at the Hanford Site over its operating life. The wastes with the highest activities are stored underground in 177 large (mostly one million gallon volume) concrete tanks with steel liners. The wastes contain processing chemicals, cladding chemicals, fission products, and actinides that were neutralized to a basic pH before addition to the tanks to prevent corrosion of the steel liners. Because the mission of the Hanford Site was to provide plutonium for defense purposes, the amount of plutonium lost to the wastes was relatively small. The best estimate of the amount of plutonium lost to all the waste tanks is about 500 kg. Given uncertainties in the measurements, some estimates are as high as 1,000 kg (Roetman et al. 1994). The wastes generally consist of (1) a sludge layer generated by precipitation of dissolved metals from aqueous wastes solutions during neutralization with sodium hydroxide, (2) a salt cake layer formed by crystallization of salts after evaporation of the supernate solution, and (3) an aqueous supernate solution that exists as a separate layer or as liquid contained in cavities between sludge or salt cake particles. The identity of chemical species of plutonium in these wastes will allow a better understanding of the behavior of the plutonium during storage in tanks, retrieval of the wastes, and processing of the wastes. Plutonium chemistry in the wastes is important to criticality and environmental concerns, and in processing the wastes for final disposal. Plutonium has been found to exist mainly in the sludge layers of the tanks along with other precipitated metal hydrous oxides. This is expected due to its low solubility in basic aqueous solutions. Tank supernate solutions do not contain high concentrations of plutonium even though some tanks contain high concentrations of complexing agents. The solutions also contain significant concentrations of hydroxide which competes with other

Plutonium Finishing Plant. Interim plutonium stabilization engineering study

Energy Technology Data Exchange (ETDEWEB)

Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J. [Pacific Northwest Lab., Richland, WA (United States); Nass, R. [Nuclear Fuel Services, Inc. (United States)

1995-08-01

This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage.

Plutonium Finishing Plant. Interim plutonium stabilization engineering study

International Nuclear Information System (INIS)

Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J.; Nass, R.

1995-08-01

This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage

In situ remediation of plutonium from glovebox exhaust ducts at the Department of Energy's Rocky Flats Plant

International Nuclear Information System (INIS)

Dugdale, J.S.; Humiston, T.J.; Omer, G.E.

1993-01-01

Plutonium and other miscellaneous hold-up materials have been accumulating in the glovebox exhaust ducts at the Rocky Flats Plant over the 40 years of weapons production at the site. The Duct Remediation Project was undertaken to assess the safety impacts of this material, and to remove it from the ductwork. The project necessitated the development of specialized tools, equipment and methods to remediate the material from continuously operating ventilation systems. Special engineered access locations were also required to provide access to the ductwork, and to ensure that safety and system operability were not degraded as a result of the remediation efforts. Operations personnel underwent significant training and development, and became an important asset to the success of the project. In total, the project succeeded in removing over 40 kilograms of plutonium-bearing material from one of the major weapons production buildings at the plant

Effects Influencing Plutonium-Absorber Interactions and Distributions in Routine and Upset Waste Treatment Plant Operations

Energy Technology Data Exchange (ETDEWEB)

Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sinkov, Sergey I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-05-01

This report is the third in a series of analyses written in support of a plan to revise the Hanford Waste Treatment and Immobilization Plant (WTP) Preliminary Criticality Safety Evaluation Report (CSER) that is being implemented at the request of the U.S. Department of Energy (DOE) Criticality Safety Group. A report on the chemical disposition of plutonium in Hanford tank wastes was prepared as Phase 1 of this plan (Delegard and Jones 2015). Phase 2 is the provision of a chemistry report to describe the potential impacts on criticality safety of waste processing operations within the WTP (Freer 2014). In accordance with the request from the Environmental and Nuclear Safety Department of the WTP (Miles and Losey 2012), the Phase 2 report assessed the potential for WTP process conditions within and outside the range of normal control parameters to change the ratio of fissile material to neutron-absorbing material in the waste as it is processed with an eye towards potential implications for criticality safety. The Phase 2 study also considered the implications should WTP processes take place within the credible range of chemistry upset conditions. In the present Phase 3 report, the 28 phenomena described in the Phase 2 report were considered with respect to the disposition of plutonium and various absorber elements. The phenomena identified in the Phase 2 report are evaluated in light of the Phase 1 report and other resources to determine the impacts these phenomena might have to alter the plutonium/absorber dispositions and ratios. The outcomes of the Phase 3 evaluations then can be used to inform subsequent engineering decisions and provide reasonable paths forward to mitigate or overcome real or potential criticality concern in plant operations.

Stop plutonium

International Nuclear Information System (INIS)

2003-02-01

This press document aims to inform the public on the hazards bound to the plutonium exploitation in France and especially the plutonium transport. The first part is a technical presentation of the plutonium and the MOX (Mixed Oxide Fuel). The second part presents the installation of the plutonium industry in France. The third part is devoted to the plutonium convoys safety. The highlight is done on the problem of the leak of ''secret'' of such transports. (A.L.B.)

An MCNP model of glove boxes in a plutonium processing facility

International Nuclear Information System (INIS)

Dooley, D.E.; Kornreich, D.E.

1998-01-01

Nuclear material processing usually occurs simultaneously in several glove boxes whose primary purpose is to contain radioactive materials and prevent inhalation or ingestion of radioactive materials by workers. A room in the plutonium facility at Los Alamos National Laboratory has been slated for installation of a glove box for storing plutonium metal in various shapes during processing. This storage glove box will be located in a room containing other glove boxes used daily by workers processing plutonium parts. An MCNP model of the room and glove boxes has been constructed to estimate the neutron flux at various locations in the room for two different locations of the storage glove box and to determine the effect of placing polyethylene shielding around the storage glove box. A neutron dose survey of the room with sources dispersed as during normal production operations was used as a benchmark to compare the neutron dose equivalent rates calculated by the MCNP model

Surplus plutonium disposition draft environmental impact statement. Summary

International Nuclear Information System (INIS)

1998-07-01

On May 22, 1997, DOE published a Notice of Intent (NOI) in the Federal Register (62 Federal Register 28009) announcing its decision to prepare an environmental impact statement (EIS) that would tier from the analysis and decisions reached in connection with the Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic EIS (Storage and Disposition PEIS). DOE's disposition strategy allows for both the immobilization of surplus plutonium and its use as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of the immobilized plutonium and MOX fuel (as spent nuclear fuel) in a geologic repository. The Surplus Plutonium Disposition Environmental Impact Statement analyzes alternatives that would use the immobilization approach (for some of the surplus plutonium) and the MOX fuel approach (for some of the surplus plutonium); alternatives that would immobilize all of the surplus plutonium; and the No Action Alternative. The alternatives include three disposition facilities that would be designed so that they could collectively accomplish disposition of up to 50 metric tons (55 tons) of surplus plutonium over their operating lives: (1) the pit disassembly and conversion facility would disassemble pits (a weapons component) and convert the recovered plutonium, as well as plutonium metal from other sources, into plutonium dioxide suitable for disposition; (2) the immobilization facility would include a collocated capability for converting nonpit plutonium materials into plutonium dioxide suitable for immobilization and would be located at either Hanford or SRS. DOE has identified SRS as the preferred site for an immobilization facility; (3) the MOX fuel fabrication facility would fabricate plutonium dioxide into MOX fuel

Recovery of plutonium from incinerator ash at Rocky Flats

International Nuclear Information System (INIS)

Johnson, T.C.

1976-01-01

Incineration of combustible materials highly contaminated with plutonium produces a residue of incinerator ash. Recovery of plutonium from incinerator ash residues at Rocky Flats is accomplished by a continuous leaching operation with nitric acid containing fluoride ion. Special equipment used in the leaching operation consists of a screw feeder, air-lift dissolvers, filters, solids dryer, and vapor collection system. Each equipment item is described in detail. The average dissolution efficiency of plutonium experienced with the process was 68% on the first pass, 74% on the second pass, and 64% on each subsequent pass. Total-solids dissolution efficiencies averaged 47% on the first pass and about 25% on each subsequent pass

Preconceptual design for separation of plutonium and gallium by ion exchange

International Nuclear Information System (INIS)

DeMuth, S.F.

1997-01-01

The disposition of plutonium from decommissioned nuclear weapons, by incorporation into commercial UO 2 -based nuclear reactor fuel, is a viable means to reduce the potential for theft of excess plutonium. This fuel, which would be a combination of plutonium oxide and uranium oxide, is referred to as a mixed oxide (MOX). Following power generation in commercial reactors with this fuel, the remaining plutonium would become mixed with highly radioactive fission products in a spent fuel assembly. The radioactivity, complex chemical composition, and large size of this spent fuel assembly, would make theft difficult with elaborate chemical processing required for plutonium recovery. In fabricating the MOX fuel, it is important to maintain current commercial fuel purity specifications. While impurities from the weapons plutonium may or may not have a detrimental affect on the fuel fabrication or fuel/cladding performance, certifying the effect as insignificant could be more costly than purification. Two primary concerns have been raised with regard to the gallium impurity: (1) gallium vaporization during fuel sintering may adversely affect the MOX fuel fabrication process, and (2) gallium vaporization during reactor operation may adversely affect the fuel cladding performance. Consequently, processes for the separation of plutonium from gallium are currently being developed and/or designed. In particular, two separation processes are being considered: (1) a developmental, potentially lower cost and lower waste, thermal vaporization process following PuO 2 powder preparation, and (2) an off-the-shelf, potentially higher cost and higher waste, aqueous-based ion exchange (IX) process. While it is planned to use the thermal vaporization process should its development prove successful, IX has been recommended as a backup process. This report presents a preconceptual design with material balances for separation of plutonium from gallium by IX

Radiological impact of plutonium recycle in the fuel cycle of LWR type reactors: professional exposure during mormal operation

International Nuclear Information System (INIS)

White, I.F.; Kelly, G.N.

1983-01-01

The radiological impact of the fuel cycle of light water type reactors using enriched uranium may be changed by plutonium recycle. The impact on human population and on the persons professionally exposed may be different according to the different steps of the fuel cycle. This report analyses the differential radiological impact on the different types of personnel involed in the fuel cycle. Each step of the fuel cycle is separately studied (fuel fabrication, reactor operation, fuel reprocessing), as also the transport of the radioactive materials between the different steps. For the whole fuel cycle, one estimates that, with regard to the fuel cycle using enriched uranium, the plutonium recycle involves a small increase of the professional exposure

Determination of plutonium in soils by mass spectrometry

International Nuclear Information System (INIS)

Storms, H.A.; Carlson, D.C.; Hunter, F.F.

1974-01-01

A procedure is described in which mass spectrometry is utilized for the determination of plutonium in soils. Using this procedure we have measured plutonium isotopic compositions at concentrations as low as 2 x 10 -14 grams Pu per gram soil. A thermal ionization source with canoe-shaped rhenium filament, is utilized in the mass spectrometer. The plutonium, when loaded onto the filament, is contained in a single Dowex-1 resin bead which is about 350 micrometers in diameter. Concentrating the plutonium within this single bead is a key step in the procedure and produces a relatively clean plutonium fraction. The resin bead also serves as an effective diffusion barrier such that the plutonium is prevented from being removed with the lower boiling impurities. The Pu remains in the bead until the temperature is sufficiently high for efficient production of Pu + ions. Plutonium ionization efficiencies as high as 2.5 percent have been measured

Radiological safety aspects in the fabrication of mixed oxide fuel elements. [Derived working limits in air and water for plutonium, enriched uranium and their mixture

Energy Technology Data Exchange (ETDEWEB)

Krishnamurthi, T.N.; Janardhanan, S.; Soman, S.D. (Bhabha Atomic Research Centre, Bombay (India). Health Physics Div.)

The problems of radiological safety in the fabrication of (U, Pu)O/sub 2/ fuel assemblies for fast reactors utilising high exposure plutonium are discussed. Derived working limits for plutonium as a function of the burn-up of RAPS (Rajasthan Atomic Power Station) fuel, external gamma and neutron exposures from feed product batches, finished fuel pins and assemblies are presented. Shielding requirements for the various glove box operations are also indicated. In general, high exposure plutonium handling calls for remote fabrication and automation at various stages would play a key role in minimising exposures to personnel in a large production plant.

The first metallurgical tests on plutonium

International Nuclear Information System (INIS)

Grison, E.; Abramson, R.; Anselin, F.; Monti, H.

1958-01-01

Metallic plutonium was first prepared in France in January 1956, as soon as we had access to quantities of the order of several grams of plutonium, which had been extracted from the rods of the pile EL2 at Saclay. Since up to the present this reactor, of thermal power 2 000 kW, has been our only source of plutonium, we have so far only worked on experimental quantities sufficient for the basic tests but not for tests on a scale of possible applications. It is this work, carried out during this phase of preliminary research, which is described below. With the starting up of the plutonium extraction plant at Marcoule, where the reactor G1 has been operating at power for more than a year, we shall go on next to a another order of magnitude which will allow the manufacture and experimentation of prototype fuel elements. (author) [fr

Constitutional problems in the handling of plutonium

International Nuclear Information System (INIS)

Witt, S. de.

1989-01-01

Reprocessing and final storage involve two different systems of nuclear energy utilization: with or without the use of plutonium. There is a choice available between these two systems. The paper discusss the constitutional implications of this choice. The permission of the use of plutonium as nuclear fuel by the Atomic Energy Law is irreconcilable with the Basic Law, i.e. the Constitution. If the corresponding provisions of the Atomic Energy Law are repealed, then only the plutonium-related branch will be revoked and not the legal permission of nuclear energy as a whole. The fact is not ignored that the Atomic Energy law does not permit the construction and operation of a plant or the handling of plutonium if this were to violate a basic right. However, the plutonium-related branch of nculear energy utilization inevitably results in such basic right violations; hence the Atomic Energy law is unconstitutional in this respect. (orig./HSCH) [de

Accelerator-driven assembly for plutonium transformation (ADAPT)

Science.gov (United States)

Tuyle, Greorgy J. Van; Todosow, Michael; Powell, James; Schweitzer, Donald

1995-01-01

A particle accelerator-driven spallation target and corresponding blanket region are proposed for the ultimate disposition of weapons-grade plutonium being retired from excess nuclear weapons in the U.S. and Russia. The highly fissle plutonium is contained within .25 to .5 cm diameter silicon-carbide coated graphite beads, which are cooled by helium, within the slightly subcritical blanket region. Major advantages include very high one-pass burnup (over 90%), a high integrity waste form (the coated beads), and operation in a subcritical mode, thereby minimizing the vulnerability to the positive reativity feedbacks often associated with plutonium fuel.

Plutonium and textbooks

International Nuclear Information System (INIS)

Silver, G.L.

2008-01-01

Chemical equilibrium is not characterized by equilibrium constants alone. At least one conservation principle is necessary. Textbook descriptions of plutonium chemistry that are based on two-reaction-product disproportionation equations, or do not recognize the conservation principles, are incomplete and potentially misleading. (author)

The future of plutonium - an overview

International Nuclear Information System (INIS)

Larson, C.E.

1975-01-01

Plutonium is the underpinning of the nuclear industry. Without it it is estimated that the fuel will run out not long after the turn of the century. With plutonium in fast breeders nuclear reactors can be operated for tens of thousands of years and the depleted uranium now available can be utilized The fuel cycle contemplated is similar to that of the light water reactor with some important differences at least partially related to the greater radioactivity of the resulting mixture of plutonium isotopes. The regulatory program does recognize the problems, including those of toxicity, safeguards and transportation. The concept of an integrated fuel cycle facility at a single location must be seriously considered. (author)

Work at Aldermaston on plutonium binding with biological systems

International Nuclear Information System (INIS)

Popplewell, D.S.

1976-01-01

Over a number of years, the Chemistry Division of AWRE, Aldermaston, studied the nature of plutonium in various biological systems. The object of the work was to identify those natural products of the body which play a part in the transport and deposition of accidental intakes of plutonium. These results should be of value in formulating more effective therapeutic measures for plutonium removal from the body. The paper reviews the work at AWRE on the interaction of plutonium with serum and rat liver. Experiments on the uptake of plutonium into cell cultures are described. The aim of these experiments was to see whether a model system could be set up for testing the efficacy by which chelating agents could remove plutonium from within cells. A simple hypothesis is presented for the mode of transport of plutonium within animals. (author)

An MS-DOS-based program for analyzing plutonium gamma-ray spectra

International Nuclear Information System (INIS)

Ruhter, W.D.; Buckley, W.M.

1989-01-01

A plutonium gamma-ray analysis system that operates on MS-DOS-based computers has been developed for the International Atomic Energy Agency (IAEA) to perform in-field analysis of plutonium gamma-ray spectra for plutonium isotopics. The program titled IAEAPU consists of three separate applications: a data-transfer application for transferring spectral data from a CICERO multichannel analyzer to a binary data file, a data-analysis application to analyze plutonium gamma-ray spectra, for plutonium isotopic ratios and weight percents of total plutonium, and a data-quality assurance application to check spectral data for proper data-acquisition setup and performance. Volume 3 contains the software listings for these applications

Plutonium finishing plant safety systems and equipment list

International Nuclear Information System (INIS)

Bergquist, G.G.

1995-01-01

The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex

Determination of plutonium in pure plutonium nitrate solutions - Gravimetric method

International Nuclear Information System (INIS)

1987-01-01

This International Standard specifies a precise and accurate gravimetric method for determining the concentration of plutonium in pure plutonium nitrate solutions and reference solutions, containing between 100 and 300 g of plutonium per litre, in a nitric acid medium. The weighed portion of the plutonium nitrate is treated with sulfuric acid and evaporated to dryness. The plutonium sulfate is decomposed and formed to oxide by heating in air. The oxide is ignited in air at 1200 to 1250 deg. C and weighed as stoichiometric plutonium dioxide, which is stable and non-hygroscopic

Design-Only Conceptual Design Report: Plutonium Immobilization Plant

International Nuclear Information System (INIS)

DiSabatino, A.; Loftus, D.

1999-01-01

This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant

Effect of compositional variation in plutonium on process shielding design

Energy Technology Data Exchange (ETDEWEB)

Brown, T.H.

1997-11-01

Radiation dose rate from plutonium with high {sup 239}Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations.

Effect of compositional variation in plutonium on process shielding design

International Nuclear Information System (INIS)

Brown, T.H.

1997-11-01

Radiation dose rate from plutonium with high 239 Pu content varies with initial nuclidic content, radioactive decay time, and impurity elemental content. The two idealized states of old plutonium and clean plutonium, whose initial compositions are given, provide approximate upper and lower bounds on dose rate variation. Whole-body dose rates were calculated for the two composition states, using unshielded and shielded plutonium spheres of varying density. The dose rates from these variable density spheres are similar to those from expanded plutonium configurations encountered during processing. The dose location of 40 cm from the sphere center is representative of operator standoff for direct handling of plutonium inside a glove box. The results have shielding implications for glove boxes with only structurally inherent shielding, especially for processing of old plutonium in an expanded configuration. Further reduction in total dose rate by using lead to reduce photon dose rate is shown for two density cases representing compact and expanded plutonium configurations

Accelerator-based conversion (ABC) of reactor and weapons plutonium

Energy Technology Data Exchange (ETDEWEB)

Jensen, R.J.; Trapp, T.J.; Arthur, E.D.; Bowman, C.D.; Davidson, J.W.; Linford, R.K.

1993-06-01

An accelerator-based conversion (ABC) system is presented that is capable of rapidly burning plutonium in a low-inventory sub-critical system. The system also returns fission power to the grid and transmutes troublesome long-lived fission products to short lived or stable products. Higher actinides are totally fissioned. The system is suited not only to controlled, rapid burning of excess weapons plutonium, but to the long range application of eliminating or drastically reducing the world total inventory of plutonium. Deployment of the system will require the successful resolution of a broad range of technical issues introduced in the paper.

Accelerator-based conversion (ABC) of reactor and weapons plutonium

International Nuclear Information System (INIS)

Jensen, R.J.; Trapp, T.J.; Arthur, E.D.; Bowman, C.D.; Davidson, J.W.; Linford, R.K.

1993-01-01

An accelerator-based conversion (ABC) system is presented that is capable of rapidly burning plutonium in a low-inventory sub-critical system. The system also returns fission power to the grid and transmutes troublesome long-lived fission products to short lived or stable products. Higher actinides are totally fissioned. The system is suited not only to controlled, rapid burning of excess weapons plutonium, but to the long range application of eliminating or drastically reducing the world total inventory of plutonium. Deployment of the system will require the successful resolution of a broad range of technical issues introduced in the paper

Separation of Plutonium from Irradiated Fuels and Targets

Energy Technology Data Exchange (ETDEWEB)

Gray, Leonard W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Holliday, Kiel S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murray, Alice [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thompson, Major [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thorp, Donald T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yarbro, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Venetz, Theodore J. [Hanford Site, Benton County, WA (United States)

2015-09-30

Spent nuclear fuel from power production reactors contains moderate amounts of transuranium (TRU) actinides and fission products in addition to the still slightly enriched uranium. Originally, nuclear technology was developed to chemically separate and recover fissionable plutonium from irradiated nuclear fuel for military purposes. Military plutonium separations had essentially ceased by the mid-1990s. Reprocessing, however, can serve multiple purposes, and the relative importance has changed over time. In the 1960’s the vision of the introduction of plutonium-fueled fast-neutron breeder reactors drove the civilian separation of plutonium. More recently, reprocessing has been regarded as a means to facilitate the disposal of high-level nuclear waste, and thus requires development of radically different technical approaches. In the last decade or so, the principal reason for reprocessing has shifted to spent power reactor fuel being reprocessed (1) so that unused uranium and plutonium being recycled reduce the volume, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security and (2) to reduce the volume and radioactivity of the waste by recovering all long-lived actinides and fission products followed by recycling them in fast reactors where they are transmuted to short-lived fission products; this reduces the volume to about 20%, reduces the long-term radioactivity level in the high-level waste, and complicates the possibility of the plutonium being diverted from civil use – thereby increasing the proliferation resistance of the fuel cycle. In general, reprocessing schemes can be divided into two large categories: aqueous/hydrometallurgical systems, and pyrochemical/pyrometallurgical systems. Worldwide processing schemes are dominated by the aqueous (hydrometallurgical) systems. This document provides a historical review of both categories of reprocessing.

Plutonium use - Present status and prospects

International Nuclear Information System (INIS)

Dievoet, J. van; Fossoul, E.; Jonckheere, E.; Bemden, E. van den

1977-01-01

The use of plutonium in thermal and fast reactors is a demonstrated, if not proven, technology. Moreover, plutonium is being produced in increasing quantities. Evaluation of this production on a world scale shows that it would be theoretically possible to construct numerous breeders and thus to make the best use of plutonium, while considerably reducing uranium consumption. This source of plutonium is nevertheless dependent on the reprocessing of irradiated fuel. Long delays in installing and adequate world reprocessing capacity are weakening the prospects for introducing breeders. Furthermore, the critical situation regarding reprocessing may delay the development of complementary reprocessing methods for fuels with a high plutonium content and high burnup. The recycling of plutonium is now a well-known technique and any objections to it hardly bear analysis. Utilization of plutonium offers an appreciable saving in terms of uranium and separative work units; and it can also be shown that immediate reprocessing of the recycling fuel is not essential for the economics of the concept. Temporary storage of recycled fuel is a particularly safe form of concentrating plutonium, namely in irradiated plutonium-bearing fuel assemblies. Finally, recycling offers such flexibility that it represents no obstacle to fuel management at power plants with light-water reactors. These strategic considerations imply that the technology of using plutonium for fabricating thermal or fast reactor fuels is both technically reliable and economically viable. The methods used in industrial facilities are fully reassuring in this respect. Although various unsolved problems exist, none seems likely to impede current developments, while the industrial experience gained has enabled the economics and reliability of the methods to be improved appreciably. Apart from the techno-economic aspects, the plutonium industry must face extremely important problems in connection with the safety of personnel

''FIXBOX'' - a new technique for the reliable conditioning of plutonium waste solutions

International Nuclear Information System (INIS)

Bruchertseifer, H.; Sommer, E.; Steinemann, M.; Bart, G.

1994-01-01

''FIXBOX'' - A new technique and facility for the conditioning of plutonium waste solutions has been developed and brought into operation in the Hot-laboratory at PSI, for the solidification of the waste from the research programmes. The facility is situated in glove-boxes for handling alpha activity and gamma-shielded for conditioning of fission product-containing waste. This report gives a brief description of the FIXBOX facility, the procedure and the first results of the cementation of plutonium waste solutions. As a result of this solidification, the actinide waste is homogeneous and strongly bound in the cement. The presence of gluconic acid and other complexing agents in the waste solution will not disturb this process. (author) figs., tabs., refs

A technique for the assessment of the masses of residual plutonium in gloveboxes using thermoluminescent dosimeters

International Nuclear Information System (INIS)

Day, B.; Godward, D.F.

1979-01-01

A means of measuring the mass of residual plutonium in gloveboxes in the size range 1 to 10 m 3 has been developed using multiple thermoluminescent detectors. By optimising the location and the number of detectors, and by using suitable filtration, the mean response from them has been made insensitive to the distribution and the composition of the plutonium. It is possible to detect 10 g of plutonium in the largest glovebox considered. The measurement and mass estimation processes have been reduced to simple operations which can be carried out by skilled industrial staff. The routine application of the technique has been arranged to minimise disturbance to be production work going on in the gloveboxes by making unattended measurements during silent hours

Behaviour of plutonium and other actinides in the environment

International Nuclear Information System (INIS)

Grauby, Andre

1976-01-01

The presence of plutonium in the environment is essentially due to nuclear experiences, as well as to some specific accidents and also controlled discharges. Having studied the cycle of this substance in water, the sediments and the living aquatic medium, plutonium behaviour in the earth medium is analysed: soils and vegetable production. It appears in particular from all the observations presented that the exchangeable fraction of plutonium in the main compartments of the natural medium is extremely low, i.e. about 0.1% of the plutonium available [fr

Recovery of plutonium from pyrolysis and incineration residues

International Nuclear Information System (INIS)

Isaacs, J.W.; McDonald, L.A.; Roberts, W.G.; Sutcliffe, P.W.; Wilkins, J.D.

1981-01-01

The effect of ashes prepared from typical plutonium-handling, glove box, combustible wastes on the dissolution of PuO 2 is described. Synthetic ashes have been prepared by doping inactively-prepared ashes with various plutonium-containing compounds, followed by heating at temperatures in the range 550-1200 0 C. The resulting ashes have been leach-tested in order to provide information on the relationship between leachability, the nature of the ashes, the type of plutonium contamination and temperature of thermal treatment. Optimum temperatures for the recovery of plutonium and for the production of inert ''slag'' -type residues have been identified. A furnace for producing model incinerator ashes and pyrolysis chars under carefully controlled conditions is described. Preliminary results on the leaching of these plutonium-active ashes and chars are discussed. (author)

Plutonium controversy

International Nuclear Information System (INIS)

Gofman, J.W.

1976-01-01

If the world chooses to seek a solution to the energy dilemma through nuclear energy, the element plutonium will become an article of commerce to be handled in quantities of thousands of tonnes annually. Plutonium is a uniquely potent inhalation carcinogen, the potential induction of lung cancer dwarfing other possible toxic effects. For reasons to be presented here, it is the author's opinion that plutonium's carcinogenicity has been very seriously underestimated. If one couples the corrected carcinogenicity with the probable degree of industrial containment of the plutonium, it appears that the commercialization of a plutonium-based energy economy is not an acceptable option for society. Sagan's statement that ''the experience of 30 years supports the contention that plutonium can be used safely'' is manifestly indefensible. No meaningful epidemiological study of plutonium-exposed workers for that 30-year period has ever been done. Since thousands of those possibly exposed have left the industry and are not even available to follow-up, it is doubtful that any meaningful study of ''the experience of 30 years'' will ever be accomplished

Health physics manual of good practices for plutonium facilities. [Contains glossary

Energy Technology Data Exchange (ETDEWEB)

Brackenbush, L.W.; Heid, K.R.; Herrington, W.N.; Kenoyer, J.L.; Munson, L.F.; Munson, L.H.; Selby, J.M.; Soldat, K.L.; Stoetzel, G.A.; Traub, R.J.

1988-05-01

This manual consists of six sections: Properties of Plutonium, Siting of Plutonium Facilities, Facility Design, Radiation Protection, Emergency Preparedness, and Decontamination and Decommissioning. While not the final authority, the manual is an assemblage of information, rules of thumb, regulations, and good practices to assist those who are intimately involved in plutonium operations. An in-depth understanding of the nuclear, physical, chemical, and biological properties of plutonium is important in establishing a viable radiation protection and control program at a plutonium facility. These properties of plutonium provide the basis and perspective necessary for appreciating the quality of control needed in handling and processing the material. Guidance in selecting the location of a new plutonium facility may not be directly useful to most readers. However, it provides a perspective for the development and implementation of the environmental surveillance program and the in-plant controls required to ensure that the facility is and remains a good neighbor. The criteria, guidance, and good practices for the design of a plutonium facility are also applicable to the operation and modification of existing facilities. The design activity provides many opportunities for implementation of features to promote more effective protection and control. The application of ''as low as reasonably achievable'' (ALARA) principles and optimization analyses are generally most cost-effective during the design phase. 335 refs., 8 figs., 20 tabs.

Preparation of hexavalent plutonium and its determination in the presence of tetravalent plutonium; Preparation de plutonium hexavalent et dosage en presence de plutonium tetravalent

Energy Technology Data Exchange (ETDEWEB)

Corpel, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Corpel, J [Institut du Radium, 75 - Paris (France)

1958-07-01

In order to study the eventual reduction of plutonium from the VI-valent state to the IV-valent state, in sulphuric medium, under the influence of its own {alpha} radiation or of the {gamma}-rays from a cobalt-60 source, we have developed a method for preparing pure hexavalent plutonium and two methods for determining solutions containing tetravalent and hexavalent plutonium simultaneously. Hexavalent plutonium was prepared by anodic oxidation at a platinum electrode. Study of the oxidation yield as a function of various factors has made it possible to define experimental conditions giving complete oxidation. For concentrations in total plutonium greater than 1.5 x 10{sup -3} M, determination of the two valencies IV and VI was carried out by spectrophotometry at two wavelengths. For lower concentrations, the determination was done by counting, after separation of the tetravalent plutonium in the form of fluoride in the presence of a carrier. (author) [French] Afin d'etudier l'eventuelle reduction du plutonium de l'etat de valence VI a l'etat de valence IV, en milieu sulfurique sous l'influence de son propre rayonnement {alpha} ou des rayons {gamma} d'une source de cobalt-60, nous avons mis au point une methode de preparation de plutonium hexavalent pur et deux methodes de dosage des solutions contenant simultanement du plutonium tetravalent et du plutonium hexavalent. Nous avons prepare le plutonium hexavalent par oxydation anodique au contact d'une electrode de platine. L'etude de rendement de l'oxydation en fonction des divers facteurs nous a permis de definir des conditions experimentales donnant une oxydation complete. Pour des concentrations en plutonium total superieures a 1,5.10{sup -3} M, le dosage des deux valences IV et VI a ete realise par spectrophotometrie a deux longueurs d'onde. Pour des concentrations inferieures, le dosage a ete effectue par comptage apres separation du plutonium tetravalent sous la forme du fluorure en presence d'un entraineur

International plutonium policy

International Nuclear Information System (INIS)

1979-02-01

As any other fissile material, the plutonium raises several utilization problems, particularly as far as diversion by sub-national groups or proliferation are concerned. The purpose of this paper is to show that these problems can be given reasonable solutions avoiding over penalties on energy production

Recovery of plutonium from lean organic in presence of Ru - activity

International Nuclear Information System (INIS)

Govindan, P.; Vijayan, K.S.; Dhamodharan, K.; Sharma, P.K.; Desigan, N.; Mohan, S.V.; Subba Rao, R.V.

2006-01-01

Recovery of plutonium from the scrubbed organic by uranous stabilized with hydrazine is carried out. It is observed that the quantitative stripping of plutonium is obtained by a single contact leaving organic with 25.0 mg/L of plutonium containing 0.468mCi/L Ru activity. The retention of plutonium in the lean organic may be attributed to the presence of degradation products and ruthenium activity may be due to the stable complex of ruthenium with TBP. The final strip product almost free from ruthenium is subjected to oxalate precipitation after pre concentration.The proposed method is applied to actual plant stream during reprocessing of high burn up fuel

Plutonium in nature

International Nuclear Information System (INIS)

Madic, C.

1994-01-01

Plutonium in nature comes from natural sources and anthropogenic ones. Plutonium at the earth surface comes principally from anthropogenic sources. It is easily detectable in environment. The plutonium behaviour in environment is complex. It seems necessary for the future to reduce releases in environment, to improve predictive models of plutonium behaviour in geosphere, to precise biological impact of anthropogenic plutonium releases

Robotic sample preparation for radiochemical plutonium and americium analyses

International Nuclear Information System (INIS)

Stalnaker, N.; Beugelsdijk, T.; Thurston, A.; Quintana, J.

1985-01-01

A Zymate robotic system has been assembled and programmed to prepare samples for plutonium and americium analyses by radioactivity counting. The system performs two procedures: a simple dilution procedure and a TTA (xylene) extraction of plutonium. To perform the procedures, the robotic system executes 11 unit operations such as weighing, pipetting, mixing, etc. Approximately 150 programs, which require 64 kilobytes of memory, control the system. The system is now being tested with high-purity plutonium metal and plutonium oxide samples. Our studies indicate that the system can give results that agree within 5% at the 95% confidence level with determinations performed manually. 1 ref., 1 fig., 1 tab

Pulmonary carcinogenesis from plutonium-containing particles

International Nuclear Information System (INIS)

Thomas, R.G.; Smith, D.M.; Anderson, E.C.

1980-01-01

Induction of lung tumors by various types of radiation is of paramount concern to the nuclear industry. The data presented were obtained by exposing the pulmonary system of Syrian hamsters to particles of zirconium oxide containing various amounts of either plutonium-238 or -239 as the alpha radiation source. These particles were injected intravenously and lodged permanently in the capillary bed of the lung. When less than 20% of the lung tissue was irradiated, simulating the ''hot particle'' mode, tumors were not evident with lung burdens up to 500 nCi plutonium. More diffuse irradiation significantly increased the tumor incidence, with lung burdens of 50 to 150 nCi. When plutonium-laden microspheres were administered intratracheally, tumor production was considerably increased and the addition of 3 mg of iron oxide intratracheally further increased the incidence. Using the zirconium oxide matrix for the carrier of plutonium in aerosol particles produced tumor incidences of up to 50% in Syrian hamsters exposed by inhalation. Initial pulmonary (alveolar) burdens reached 100 nCi of plutonium. Similar inhalation studies using plutonium dioxide alone (no matrix) failed to produce any increase in lung tumorigenesis. The results are discussed in terms of possible mechanisms necessary for lung carcinogenesis. (H.K.)

Extraction and purification of plutonium by a tertiary amine; Extraction et purification du plutonium par une amine tertiaire

Energy Technology Data Exchange (ETDEWEB)

Trentinian, M de; Chesne, A [Commissariat a l' Energie Atomique, Fontenay aux Roses, Section de Chimie des Actimides (France).Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1960-07-01

Trilaurylamine diluted with a paraffinic solvent (dodecane) was studied as part of the research dealing with the separation and purification of plutonium. The physical properties (solubility of nitrates in the amine as a function of temperature) and the resistance to radiations of this substance were examined. The extraction characteristics of nitric solutions of plutonium, uranium and certain fission products are given as a function of the following factors: concentration of the various ions in solution, valency states. A method of plutonium purification based on these results is presented. (author) [French] La trilaurylamine diluee par un solvant paraffinique (dodecane) a ete etudiee dans le cadre des recherches concernant la separation et la purification du plutonium. Une etude des caracteres physiques (solubilite des nitrates dans l'amine en fonction de la temperature) s'ajoute a celle de la tenue aux radiations de ce corps. Les caracteristiques d'extraction de solutions nitriques de plutonium, uranium, et certains produits de fission, sont donnes en fonction des facteurs suivants: concentration des differents ions en solution, etats de valence. On presente une methode de purification du plutonium basee sur ces resultats. (auteur)

Plutonium in the arctic marine environment--a short review.

Science.gov (United States)

Skipperud, Lindis

2004-06-18

Anthropogenic plutonium has been introduced into the environment over the past 50 years as the result of the detonation of nuclear weapons and operational releases from the nuclear industry. In the Arctic environment, the main source of plutonium is from atmospheric weapons testing, which has resulted in a relatively uniform, underlying global distribution of plutonium. Previous studies of plutonium in the Kara Sea have shown that, at certain sites, other releases have given rise to enhanced local concentrations. Since different plutonium sources are characterised by distinctive plutonium-isotope ratios, evidence of a localised influence can be supported by clear perturbations in the plutonium-isotope ratio fingerprints as compared to the known ratio in global fallout. In Kara Sea sites, such perturbations have been observed as a result of underwater weapons tests at Chernaya Bay, dumped radioactive waste in Novaya Zemlya, and terrestrial runoff from the Ob and Yenisey Rivers. Measurement of the plutonium-isotope ratios offers both a means of identifying the origin of radionuclide contamination and the influence of the various nuclear installations on inputs to the Arctic, as well as a potential method for following the movement of water and sediment loads in the rivers.

Political influences in plutonium recycling

International Nuclear Information System (INIS)

Patak, H.N.

1982-01-01

The history of plutonium safeguards is one of political error and misunderstandings, as well as a lack of technical knowledge. Although there was widespread support for preventing the proliferation of nuclear explosives, with over 100 nations signing the Nonproliferation Treaty of 1969, India's 1974 nuclear test brought renewed political activity to prevent another such occurrence. Opposition has been directed only at how to pursue this goal, but the status of four major experiments aimed at minimizing weapons proliferation is one of failure, intensified by a weakening of the International Atomic Energy Agency (IAEA). If the link between plutonium power and weapons production can be broken through on-site reprocessing, the situation could improve. One course would be for the nuclear power industry to adopt its own system for safe guarding plutonium

Guide to good practices at plutonium facilities

International Nuclear Information System (INIS)

Faust, L.G.; Brackenbush, L.W.; Carter, L.A.; Endres, G.W.R.; Glenn, R.D.; Jech, J.J.; Selby, J.M.; Smith, R.C.; Waite, D.A.; Walsh, W.P.

1977-09-01

This manual establishes guidelines and principles for use in setting up a sound radiation protection program for work with plutonium. The guidance presented is based on the experiences of Energy Research and Development Administration (ERDA) contractors and those portions of private industry concerned with the operation of plutonium facilities, specifically with the fabrication of mixed oxide reactor fuel. The manual is directed primarily to those facilities which have as their sole purpose the handling of large quantities of plutonium for military or industrial uses. It is not intended for use by facilities engaged in reactor or chemical separation operations nor for partial or occasional use by analytical laboratories; while these facilities would find the manual beneficial, it would be incomplete for their needs. The manual addresses good practices that should be observed by management, staff and designers, since the benefits of a good radiation protection program are the result of their joint efforts. Methods for the diagnostic evaluation of internally deposited Pu are included

Technical considerations and policy requirements for plutonium management

International Nuclear Information System (INIS)

Christensen, D.C.; Dinehart, S.M.; Yarbro, S.L.

1995-01-01

The goals for plutonium management have changed dramatically over the past few years. Today, the challenge is focused on isolating plutonium from the environment and preparing it for permanent disposition. In parallel, the requirements for managing plutonium are rapidly changing. For example, there is a significant increase in public awareness on how facilities operate, increased attention to environmental safety and health (ES and H) concerns, greater interest in minimizing waste, more emphasis on protecting material from theft, providing materials for international inspection, and a resurgence of interest in using plutonium as an energy source. Of highest concern, in the immediate future, is protecting plutonium from theft or diversion, while the national policy on disposition is debated. These expanded requirements are causing a broadening of responsibilities within the Department of Energy (DOE) to include at least seven organizations. An unavoidable consequence is the divergence in approach and short-term goals for managing similar materials within each organization. The technology base does exist, properly, safely, and cost effectively to extract plutonium from excess weapons, residues, waste, and contaminated equipment and facilities, and to properly stabilize it. Extracting the plutonium enables it to be easily inventoried, packaged, and managed to minimize the risk of theft and diversion. Discarding excess plutonium does not sufficiently reduce the risk of diversion, and as a result, long-term containment of plutonium from the environment may not be able to be proven to the satisfaction of the public

Technical considerations and policy requirements for plutonium management

International Nuclear Information System (INIS)

Christensen, D.C.; Dinehart, S.M.; Yarbro, S.L.

1996-01-01

The goals for plutonium management have changed dramatically over the past few years. Today, the challenge is focused on isolating plutonium from the environment and preparing it for permanent disposition. In parallel, the requirements for managing plutonium are rapidly changing. For example, there is a significant increase in public awareness on how facilities operate, increased attention to environmental safety and health (ES and H) concerns, greater interest in minimizing waste, more emphasis on protecting material from theft, providing materials for international inspection, and a resurgence of interest in using plutonium as an energy source. Of highest concern, in the immediate future, is protecting plutonium from theft or diversion, while the national policy on disposition is debated. These expanded requirements are causing a broadening of responsibilities within the Department of Energy (DOE) to include at least seven organizations. An unavoidable consequence is the divergence in approach and short-term goals for managing similar materials within each organization. The technology base does exist, properly, safely, and cost effectively to extract plutonium from excess weapons, residues, waste, and contaminated equipment and facilities, and to properly stabilize it. Extracting the plutonium enables it to be easily inventoried, packaged, and managed to minimize the risk of theft and diversion. Discarding excess plutonium does not sufficient reduce the risk of diversion, and as a result, long-term containment of plutonium from the environment may not be able to be proven to the satisfaction of the public

Review of major plutonium pyrochemical technology

International Nuclear Information System (INIS)

Moser, W.S.; Navratil, J.D.

1983-01-01

The past twenty years have seen significant growth in the development and application of pyrochemical technology for processing of plutonium. For particular feedstocks and specific applications, non-aqueous high-temperature processes offer key advantages over conventional hydrometallurgical systems. Major processes in use today include: (1) direct oxide reduction for conversion of PuO 2 to metal, (2) molten salt extraction for americium removal from plutonium, (3) molten salt electrorefining for Pu purification, and (4) hydriding to remove plutonium from host substrates. This paper reviews current major pyrochemical processes from the classical calcination-hydrofluorination-bomb reduction sequence through new techniques under development. Each process is presented and brief descriptions of production equipment are given. 47 references, 5 figures

CANDU - a versatile reactor for plutonium disposition or actinide burning

International Nuclear Information System (INIS)

Chan, P.S.W.; Gagnon, M.J.N.; Boczar, P.G.; Ellis, R.J.; Verrall, R.A.

1997-10-01

High neutron economy, on-line refuelling, and a simple fuel-bundle design result in a high degree of versatility in the use of the CANDU reactor for the disposition of weapons-derived plutonium and for the annihilation of long-lived radioactive actinides, such as plutonium, neptunium, and americium isotopes, created in civilian nuclear power reactors. Inherent safety features are incorporated into the design of the bundles carrying the plutonium and actinide fuels. This approach enables existing CANDU reactors to operate with various plutonium-based fuel cycles without requiring major changes to the current reactor design. (author)

Japan`s civil use of foreign military plutonium

Energy Technology Data Exchange (ETDEWEB)

Suzuki, A. [Tokyo Univ. (Japan). Dept. of Quantum Engineering and Systems Sciences

1995-12-31

This paper is intended to propose one of the MOX options of international cooperation for safer and more secured management of excess military plutonium. The proposal was made with special reference to the Japanese public`s view. Owing to the domestic plutonium shortage anticipated soon after the 200 in Japan, some specific reactors will be available to get rid of foreign excess weapon plutonium. According to the Japan AEC`s new long-term programme, the shortage will be approximately 0.5 tonne of plutonium per annum, which is a sort of the least amount that Japan can buy from a certain external source. With international requests for a more positive Japanese contribution, however, the amount of Japanese purchase would be increased. It follows from the preliminary estimate shown in this paper that roughly 2 tonnes of plutonium can be burned annually in the reactors without any major modifications concerning safe reactor operation. (author) 10 refs.

Experience with thermal recycle of plutonium and uranium

International Nuclear Information System (INIS)

Beer, O.; Schlosser, G.; Spielvogel, F.

1985-01-01

The Federal Republic of Germany (FRG) decided to close the fuel cycle by erecting the reprocessing plant WA350 at Wackersdorf. As long as the plutonium supply from reprocessing plants exceeds the plutonium demand of fast breeder reactors, recycling of plutonium in LWR's is a convenient solution by which a significant advanced uranium utilization is achieved. The demonstration of plutonium recycling performed to date in the FRG in BWR's and PWR's shows that thermal plutonium recycling on an industrial scale is feasible and that the usual levels of reliability and safety can be achieved in reactor operation. The recycling of reprocessed uranium is presently demonstrated in the FRG, too. As regards fuel cycle economy thermal recycling allows savings in natural uranium and separative work. Already under present cost conditions the fuel cycle costs for mixed oxide or enriched reprocessed uranium fuel assemblies are equal or even lower than for usual uranium fuel assemblies

Plutonium dissolution from Rocky Flats Plant incinerator ash

International Nuclear Information System (INIS)

Delegard, C.H.

1985-06-01

Rockwell Hanford Operations (Rockwell) soon will commence recovery of plutonium from Rocky Flats Plant incinerator ash. In preparation for this processing, Rockwell undertook literature and laboratory studies to identify, select and optimize plutonium dissolution methods for treating the ash. Ash reburning, followed by dissolution in nitric acid containing calcium fluoride, was selected as the processing method for the ash. Recommended values of process parameters were identified. Using the selected process, 99.5% plutonium recovery was achieved, leaving about 12.7 wt % heel residue for an equal weight composite of the three ashes tested. 15 refs., 26 figs

Economic considerations of plutonium utilization in the nuclear power strategy of Finland

International Nuclear Information System (INIS)

Silvennoinen, P.; Tusa, E.; Routti, J.T.

1977-01-01

Based on the current and predicted share of nuclear power in the national energy supply strategy, an optimal programme is developed for the exploitation of plutonium in both light-water and fast reactor systems. Assuming cost trends beyond the year 2000 for uranium, plutonium, uranium enrichment, fuel fabrication and assessing the availability of plutonium from domestic power plants and from abroad, the nuclear construction programme is optimized economically in view of the estimated development in the investment costs of various plant types. Given the expected nuclear share of the energy procurement this sector is covered by the alternative production schemes, i.e. light-water reactors with and without plutonium recycle, and fast reactors. Defining the objective function in terms of minimized revenue requirement in plant amortization and operation the generated scenarios are screened off and they finally converge to the optimal policy of nuclear power construction up to the year 2000. The established technology is associated with a larger share of the domestic manufacturing and the introduction of a new fuel or reactor type is taken to correspond to a reduced domestic investment share. In the investment costs the domestic fraction is regarded competitive up to a certain marginal excess. Plutonium recycle is seen to be competitive from 1985 or as soon as the required amount of fuel has been reprocessed. The domestic accumulation of plutonium will be able to support the introduction of the LMFBR in 1997. Owing to the uncertainties prevailing in the forecasts, sensitivity studies are performed as functions of the major economic parameters and their temporal development. (author)

Progress on immobilisation of plutonium residues and shredded plutonium contaminated materials in cement

International Nuclear Information System (INIS)

Landles, A.J.; Awmack, A.F.; Baxter, W.

1987-03-01

Laboratory scale experiments have been carried out to study the feasibility of encapsulating plutonium contaminated materials in cement. A proposed grout of a 3:1 PFA/OPC mixture has been tested and some product evaluation carried out. (author)

Facility Operations 1993 fiscal year work plan: WBS 1.3.1

Energy Technology Data Exchange (ETDEWEB)

1992-11-01

The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO{sub 3}) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP).

Facility Operations 1993 fiscal year work plan: WBS 1.3.1

International Nuclear Information System (INIS)

1992-11-01

The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO 3 ) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP)

Plutonium use - present status and perspectives

International Nuclear Information System (INIS)

Dievoet, J. van; Fossoul, E.; Jonckheere, E.; Bemden, E. van den

1977-01-01

Plutonium is being produced in increasing quantities in the so-called proven reactors, which are mostly of the light-water type. Evaluation of this production on a world scale shows that it would be theoretically possible to construct a large number of breeders and thus to make the best use of the intrinsic qualities of plutonium as a fissionable material, while considerably reducing the consumption of uranium. This source of plutonium is nevertheless dependent on an essential stage of the fuel cycle, namely reprocessing of irradiated fuel. The long delays in installing an adequate world reprocessing capacity are substantially weakening the prospects for the introduction of breeders. Furthermore, the critical situation as regards reprocessing may delay the development of complementary reprocessing methods for fuels with a high plutonium content and high burn-up. When it is recalled that fast reactors themselves may suffer some delay in their technological development, if only because of the intention to build power plants of very high unit capacity immediately, it must be concluded that another use will have to be considered for the plutonium available in future -use in thermal reactors, i.e. recycling. The recycling of plutonium is a well-known technique today and the objections which could be raised against it hardly stand up to analysis. Utilization of plutonium offers an appreciable saving in terms of uranium and separative work units, the consumption being of a low order of magnitude in comparison with the total amount of plutonium needed for the eventual fabrication of the first fast reactor cores. It can also be shown that immediate reprocessing of the recycling fuel is not essential for the economics of the concept. Temporary storage of recycled fuel has the advantage of concentrating plutonium in a particularly safe form, namely in irradiated plutonium-bearing fuel assemblies. Lastly, recycling offers such flexibility that it does not in practice represent

Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

Energy Technology Data Exchange (ETDEWEB)

Hoover, Andrew S [Los Alamos National Laboratory; Rudy, Cliff R [Los Alamos National Laboratory; Tobin, Steve J [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Stafford, A [TEXAS A& M; Strohmeyer, D [TEXAS A& M; Saavadra, S [ORNL

2009-01-01

Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

International Nuclear Information System (INIS)

Rubini, L.A.

1978-02-01

An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

Learning more about plutonium

International Nuclear Information System (INIS)

2005-01-01

This document offers chemical, metallurgical and economical information on the plutonium, a hard white radioelement. It deals also on the plutonium formation in the earth, the plutonium use in the nuclear industry, the plutonium in the environment and the plutonium toxicity. (A.L.B.)

Design of plutonium processing facilities

International Nuclear Information System (INIS)

Derbyshire, W.; Sills, R.J.

1982-01-01

Five considerations for the design of plutonium processing facilities are identified. These are: Toxicity, Radiation, Criticality, Containment and Remote Operation. They are examined with reference to reprocessing spent nuclear fuel and application is detailed both for liquid and dry processes. (author)

On-line monitoring of low-level plutonium concentrations

International Nuclear Information System (INIS)

Hofstetter, K.J.; Huff, G.A.; Rebagay, T.V.

1979-10-01

An on-line monitor has been developed to assay plutonium in nitric acid solutions. The performance of the monitor has been assessed by a laboratory experimentation program using solutions with plutonium concentrations from 0.1 to 10 g/l. These conditions are typical of the plutonium solutions in an input stream to a plutonium-purification cycle in a reprocessing plant following uranium/plutonium partitioning. The monitoring system can be fully automated and shows great promise for detecting and quantifying plutonium in situ, thus minimizing the reliance on traditional sampling and laboratory-analysis techniques. The total concentration and isotopic abundance of plutonium are determined by measuring the absolute intensities of the low-energy gamma rays characteristics of 238 Pu, 239 Pu, and 240 Pu nuclides by direct gamma-ray spectroscopy and computer analysis of the spectral data. The addition of a monitoring system of this type to the input stream of a plutonium-purification cycle along with other suitable monitors on the waste streams and on the product stream provides the basis for a near real-time materials control and inventory system. Results of the laboratory-evaluation program employing plutonium in solutions with isotopic compositions typical of those involved in processing light water reactor fuels are presented. The detailed design of a monitoring cell and detection system is given. The precision and accuracy of the results relative to those measured by mass spectrometry and controlled potential coulometry are also summarized

Prototype fast neutron counter for the assay of impure plutonium

International Nuclear Information System (INIS)

Wachter, J.R.; Adams, E.L.; Ensslin, N.

1987-01-01

A fast coincident neutron counter using liquid scintillators and gamma-ray/neutron pulse-shape discrimination has been constructed for the analysis of plutonium samples with unknown self-multiplication and (α,n) production. The counter was used to measure plutonium-bearing materials that cover a range of masses and (α,n) reaction rates of importance to the safeguards community. Measured values of the 240 Pu effective mass differed, on average, from their declared values by 0.4% for plutonium oxides and by -2.2% for metal and MgO-loaded samples. Poorer results were obtained for materials with large (α,n) reaction rates and low self-multiplication such as plutonium ash and plutonium fluoride

Safeguarding the Plutonium Fuel Cycle

International Nuclear Information System (INIS)

Johnson, S.J.; Lockwood, D.

2013-01-01

In developing a Safeguards Approach for a plutonium process facility, two general diversion and misuse scenarios must be addressed: 1) Unreported batches of undeclared nuclear material being processed through the plant and bypassing the accountancy measurement points, and 2) The operator removing plutonium at a rate that cannot be detected with confidence due to measurement uncertainties. This paper will look at the implementation of international safeguards at plutonium fuel cycle facilities in light of past lessons learned and current safeguards approaches. It will then discuss technical areas which are currently being addressed as future tools to improve on the efficiency of safeguards implementation, while maintaining its effectiveness. The discussion of new improvements will include: safeguards by design (SBD), process monitoring (PM), measurement and monitoring equipment, and data management. The paper is illustrated with the implementation of international safeguards at the Rokkasho Reprocessing Plant in Japan and its accountancy structure is detailed. The paper is followed by the slides of the presentation

Multiple recycling of plutonium in advanced PWRs

International Nuclear Information System (INIS)

Kloosterman, J.L.

1998-04-01

The influence of the moderator-to-fuel ratio in MOX fueled PWRs on the moderator void coefficient, the fuel temperature coefficient, the moderator temperature coefficient, the boron reactivity worth, the critical boron concentration, the mean neutron generation time and the effective delayed neutron fraction has been assessed. Increasing the moderator-to-fuel ratio to values larger than three, gives a moderator void coefficient sufficiently large to recycle the plutonium at least four times. Scenario studies show that four times recycling of plutonium in PWRs reduces the plutonium mass produced with a factor of three compared with a reference once-through reactor park, but that the americium and curium production triple. If the minor actinides and the remaining plutonium after four times recycling are disposed of, the reduction of the radiotoxicity reaches only a factor of two. This factor increases to five at the maximum when the plutonium is further recycled. Recycling of americium and curium is needed to further reduce the radiotoxicity of the spent fuel. 4 refs

Surplus plutonium disposition draft environmental impact statement. Volume 2

International Nuclear Information System (INIS)

1998-07-01

On May 22, 1997, DOE published a Notice of Intent (NOI) in the Federal Register (62 Federal Register 28009) announcing its decision to prepare an environmental impact statement (EIS) that would tier from the analysis and decisions reached in connection with the Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic EIS (Storage and Disposition PEIS). DOE's disposition strategy allows for both the immobilization of surplus plutonium and its use as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of the immobilized plutonium and MOX fuel (as spent nuclear fuel) in a geologic repository. The Surplus Plutonium Disposition Environmental Impact Statement analyzes alternatives that would use the immobilization approach (for some of the surplus plutonium) and the MOX fuel approach (for some of the surplus plutonium); alternatives that would immobilize all of the surplus plutonium; and the No Action Alternative. The alternatives include three disposition facilities that would be designed so that they could collectively accomplish disposition of up to 50 metric tons (55 tons) of surplus plutonium over their operating lives: (1) the pit disassembly and conversion facility would disassemble pits (a weapons component) and convert the recovered plutonium, as well as plutonium metal from other sources, into plutonium dioxide suitable for disposition; (2) the immobilization facility would include a collocated capability for converting nonpit plutonium materials into plutonium dioxide suitable for immobilization and would be located at either Hanford or SRS. DOE has identified SRS as the preferred site for an immobilization facility; (3) the MOX fuel fabrication facility would fabricate plutonium dioxide into MOX fuel. Volume 2 contains the appendices to the report and describe the following: Federal Register notices; contractor nondisclosure statement; adjunct melter

The use of plutonium in Swedish reactors

International Nuclear Information System (INIS)

Forsstroem, H.

1982-09-01

The report deals with the utilization of plutonium in Swedish nuclear power plants. The plutonium content of the mixed oxide fuel will normally be 3-7 per cent. The processing of spent nuclear fuel will produce about 6 ton plutonium. The use of mixed oxide fuel in Forsmark 3 and Oskarshamn 3 is discussed. The fuel cycle will start with the manufacturing of the fuel elements abroad and proceeds with transport and utilization, storing of spent fuel about 40 years in Sweden followed by direct disposal. The manufacture and use of mixed oxide (MOX) fuel is based on well-known techniques. Approximately 20 000 MOX fuel rods have been irradiated and the fuel is essentially equivalent to uranium oxide fuel. 30-50 per cent of the core may be composed of MOX-fuel without any effect on the operation and safety of the reactor which has been originally designed for uranium fuel. The evaluation of international fuel cycle (INFCE) states that the proliferation risks are very small. The recycling of plutonium will reduce demand for enriched uranium and the calculations show that 6.3 ton plutonium will replace the enrichment of 600 ton natural uranium. (G.B.)

Microwave calcination for plutonium immobilization and residue stabilization

Energy Technology Data Exchange (ETDEWEB)

Harris, M.J.; Rising, T.L.; Roushey, W.J.; Sprenger, G.S. [Kaiser-Hill Co., Golden, CO (United States)

1995-12-01

In the late 1980`s development was begun on a process using microwave energy to vitrify low level mixed waste sludge and transuranic mixed waste sludge generated in Building 374 at Rocky Flats. This process was shown to produce a dense, highly durable waste form. With the cessation of weapons production at Rocky Flats, the emphasis has changed from treatment of low level and TRU wastes to stabilizaiton of plutonium oxide and residues. This equipment is versatile and can be used as a heat source to calcine, react or vitrify many types of residues and oxides. It has natural economies in that it heats only the material to be treated, significantly reducing cycle times over conventional furnaces. It is inexpensive to operate in that most of the working components remain outside of any necessary contamination enclosure and therefore can easily be maintained. Limited testing has been successfully performed on cerium oxide (as a surrogate for plutonium oxide), surrogate electrorefining salts, surrogate residue sludge and residue ash. Future plans also include tests on ion exchange resins. In an attempt to further the usefullness of this technology, a mobile, self-contained microwave melting system is currently under development and expected to be operational at Rocky Flats Enviromental Technology Site by the 4th quarter of FY96.

Microwave calcination for plutonium immobilization and residue stabilization

International Nuclear Information System (INIS)

Harris, M.J.; Rising, T.L.; Roushey, W.J.; Sprenger, G.S.

1995-01-01

In the late 1980's development was begun on a process using microwave energy to vitrify low level mixed waste sludge and transuranic mixed waste sludge generated in Building 374 at Rocky Flats. This process was shown to produce a dense, highly durable waste form. With the cessation of weapons production at Rocky Flats, the emphasis has changed from treatment of low level and TRU wastes to stabilizaiton of plutonium oxide and residues. This equipment is versatile and can be used as a heat source to calcine, react or vitrify many types of residues and oxides. It has natural economies in that it heats only the material to be treated, significantly reducing cycle times over conventional furnaces. It is inexpensive to operate in that most of the working components remain outside of any necessary contamination enclosure and therefore can easily be maintained. Limited testing has been successfully performed on cerium oxide (as a surrogate for plutonium oxide), surrogate electrorefining salts, surrogate residue sludge and residue ash. Future plans also include tests on ion exchange resins. In an attempt to further the usefullness of this technology, a mobile, self-contained microwave melting system is currently under development and expected to be operational at Rocky Flats Enviromental Technology Site by the 4th quarter of FY96

LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

Energy Technology Data Exchange (ETDEWEB)

Bronson, M.C.

1997-10-01

The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

International Nuclear Information System (INIS)

Bronson, M.C.

1997-01-01

The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium

An improved, computer-based, on-line gamma monitor for plutonium anion exchange process control

International Nuclear Information System (INIS)

Pope, N.G.; Marsh, S.F.

1987-06-01

An improved, low-cost, computer-based system has replaced a previously developed on-line gamma monitor. Both instruments continuously profile uranium, plutonium, and americium in the nitrate anion exchange process used to recover and purify plutonium at the Los Alamos Plutonium Facility. The latest system incorporates a personal computer that provides full-feature multichannel analyzer (MCA) capabilities by means of a single-slot, plug-in integrated circuit board. In addition to controlling all MCA functions, the computer program continuously corrects for gain shift and performs all other data processing functions. This Plutonium Recovery Operations Gamma Ray Energy Spectrometer System (PROGRESS) provides on-line process operational data essential for efficient operation. By identifying abnormal conditions in real time, it allows operators to take corrective actions promptly. The decision-making capability of the computer will be of increasing value as we implement automated process-control functions in the future. 4 refs., 6 figs

Reactor-grade plutonium inventory taking in the RT-1 storage production association Mayak site

International Nuclear Information System (INIS)

Zatorskij, Yu.M.

1999-01-01

General characterization of plutonium physical inventory taking (PIT) system at the RT-1 plant is given in the paper. The proposed PIT procedure is based on measuring indirect parameters of inventory plutonium, the so called attributes of items - containers with plutonium. The method how to assess quality of item attribute measurements is proposed. If any defect is detected, the container is sent back to the place where it been filed and packed in order to perform direct measurements of plutonium mass. Taking into account the above-mentioned techniques and the availability of several access control means it is proposed the period between two PIT procedures be equal to 12 months [ru

Civil plutonium management

International Nuclear Information System (INIS)

Sicard, B.; Zaetta, A.

2004-01-01

During 1960 and 1970 the researches on the plutonium recycling in fast neutrons reactors were stimulated by the fear of uranium reserves diminishing. At the beginning of 1980, the plutonium mono-recycling for water cooled reactors is implementing. After 1990 the public opinion concerning the radioactive wastes management and the consequences of the disarmament agreements between Russia and United States, modified the context. This paper presents the today situation and technology associated to the different options and strategical solutions of the plutonium management: the plutonium use in the world, the neutronic characteristics, the plutonium effect on the reactors characteristics, the MOX behavior in the reactors, the MOX fabrication and treatment, the possible improvements to the plutonium use, the concepts performance in a nuclear park. (A.L.B.)

The production control laboratories of the plutonium extraction Plant at Marcoule. Six years operating experience: 1957 - 1963; Les laboratoires de fabrication de l'usine d'extraction du plutonium de Marcoule. Experience des 6 premieres annees de fonctionnement: 1957 - 1963

Energy Technology Data Exchange (ETDEWEB)

Fontaine, A [CEA Marcoule, Centre de Production de Plutonium, 30 (France)

1964-07-01

In this paper, the author attempts to sum up the conditions prevailing, after six years of operation, in the Laboratories of the Plutonium Extraction Plant. The origins and objectives are briefly reviewed, the technology and staff recruitment policy are examined, and progress made is shown. The methods used as well as the scope of application and limits imposed at the present state are considered. Past achievements and further possibilities in the next future are examined. An attempt has been made to bring out the outlooks for the more distant future and to investigate the conditions required for the successful carrying out of the program. (author) [French] Le present rapport tente, apres 6 ans de vie, de faire le point de la situation des Laboratoires du Service Extraction du Plutonium. Apres un rapide retour sur les origines et les objectifs, apres quelques considerations sur le recrutement et la technologie situant le contexte de leurs progres, nous etudierons les methodes, leurs domaines d'application, leurs limites actuelles. Nous ferons un bilan des realisations et des possibilites pour un avenir proche. Nous tenterons de degager quelques vues plus lointaines et les conditions pour les mener a bonne fin. (auteur)

Civil plutonium in the world: an estimate by the code REACTOR

International Nuclear Information System (INIS)

Braet, J.; Carchon, R.; Van der Meer, K.

1996-11-01

The computer code REACTOR that was developed by the Belgian Nuclear Research Centre SCK/CEN to study the built-up of plutonium stockpiles in the world is described. The code consists of a central database, containing general information about most commercial civil nuclear facilities. Using this code, an overview is given of the evolution of the nuclear energy production in the world, in the past and the medium term future. The nuclear energy production results in the accumulation of spent fuel stocks, containing vast amounts of energy enclosed in the plutonium. The presence and built-up of large stockpiles of spent fuel and separated plutonium originating from the civil fuel cycle is estimated. In this report several possible scenarios are considered for the use of that plutonium, with the aim of minimizing those stocks. According to the different national policies, scenarios such as open fuel cycle, thermal reactors or fast reactor cycle with the burning of plutonium in fast reactors are envisaged

Plutonium waste incineration using pyrohydrolysis

International Nuclear Information System (INIS)

Meyer, M.L.

1991-01-01

Waste generated by Savannah River Site (SRS) plutonium operations includes a contaminated organic waste stream. A conventional method for disposing of the organic waste stream and recovering the nuclear material is by incineration. When the organic material is burned, the plutonium remains in the incinerator ash. Plutonium recovery from incinerator ash is highly dependent on the maximum temperature to which the oxide is exposed. Recovery via acid leaching is reduced for a high fired ash (>800 degree C), while plutonium oxides fired at lower decomposition temperatures (400--800 degrees C) are more soluble at any given acid concentration. To determine the feasibility of using a lower temperature process, tests were conducted using an electrically heated, controlled-air incinerator. Nine nonradioactive, solid, waste materials were batch-fed and processed in a top-heated cylindrical furnace. Waste material processing was completed using a 19-liter batch over a nominal 8-hour cycle. A processing cycle consisted of 1 hour for heating, 4 hours for reacting, and 3 hours for chamber cooling. The water gas shift reaction was used to hydrolyze waste materials in an atmosphere of 336% steam and 4.4% oxygen. Throughput ranged from 0.14 to 0.27 kg/hr depending on the variability in the waste material composition and density

Monitoring of atmospheric contamination by plutonium in laboratories

International Nuclear Information System (INIS)

Pomarola, J.; Risselin, A.; Feliers, P.

1965-02-01

Immediate detection of atmospheric contamination by plutonium is necessary for warning of operators. A precise estimate of the level of this contamination is also necessary in order to give the Medical Section proper information. Experiments have been carried out at CEN-FAR using atmospheric contamination by plutonium monitors. This paper deals successively with: important problems of monitoring, the carrying out of the experiments and the results yielded. (authors) [fr

Development of isotope dilution gamma-ray spectrometry for plutonium analysis

Energy Technology Data Exchange (ETDEWEB)

Li, T.K.; Parker, J.L. (Los Alamos National Lab., NM (United States)); Kuno, Y.; Sato, S.; Kurosawa, A.; Akiyama, T. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))

1991-01-01

We are studying the feasibility of determining the plutonium concentration and isotopic distribution of highly radioactive, spent-fuel dissolver solutions by employing high-resolution gamma-ray spectrometry. The study involves gamma-ray plutonium isotopic analysis for both dissolver and spiked dissolver solution samples, after plutonium is eluted through an ion-exchange column and absorbed in a small resin bead bag. The spike is well characterized, dry plutonium containing {approximately}98% of {sup 239}Pu. By using measured isotopic information, the concentration of elemental plutonium in the dissolver solution can be determined. Both the plutonium concentration and the isotopic composition of the dissolver solution obtained from this study agree well with values obtained by traditional isotope dilution mass spectrometry (IDMS). Because it is rapid, easy to operate and maintain, and costs less, this new technique could be an alternative method to IDMS for input accountability and verification measurements in reprocessing plants. 7 refs., 4 figs., 4 tabs.

Cost-benefit analysis of unfired PuO2 pellets as an alternative plutonium shipping form

International Nuclear Information System (INIS)

Mishima, J.; Brackenbush, L.W.; Libby, R.A.; Soldat, K.L.; White, G.D.

1983-10-01

A limited cost-benefit evaluation was performed concerning use of unfired plutonium dioxide pellets as a shipping form. Two specific processing operations are required for this use, one to form the pellet (pelletizing) and a second to reconstitute an acceptable powder upon receipt (reconstitution). The direct costs for the pelletizing operation are approximately $208,000 for equipment and its installation and $122 per kg of plutonium processed (based upon a 20-kg plutonium/day facility). The direct costs for reconstitution are approximately $90,000 for equipment and its installation and $81 per kg of plutonium processed. The indirect cost considered was personnel exposure from these operations. Whole body exposures ranged from 0.04 man-rem per 100 kg of low-exposure plutonium reconstituted to 0.9 man-rem per 100 kg of average-exposure plutonium pelletized. Hand exposures were much higher - 17 man-rem power 100 kg of low-exposure plutonium reconstituted to 67 man-rem per 100 kg of average plutonium pelletized. The principal benefit is a potential twentyfold reduction of airborne release in the event of an accident. An experimental plan is outlined to fill the data gaps uncovered during this study in the areas of pelletizing and reconstitution process parameters and pellet response behavior to accident-generated stresses. A study to enhance the containment potential of the inner packaging used during shipment is also outlined

Study on the process variables in the anion exchange plutonium separation process

Energy Technology Data Exchange (ETDEWEB)

Nishimura, D T

1957-11-15

This report discusses the study of the process variables in the Anion Exchange Process Pilot Plant for the separation of plutonium from irradiated uranium. Variables associated with the feed, wash and elution cycles were studied with the aim of improving the quality of the final plutonium product, reduce cycling time and reagent requirements, and also to obtain data for prediction of resin column behaviour under various feed conditions. A cation resin column and a silica gel column were installed in the system and these were studied for plutonium recovery and product quality. The product obtained from the plant was acceptable in all the impurities except the associated gamma activity which was too high for easy product handling. (author)

Plutonium peroxide precipitation: review and current research

International Nuclear Information System (INIS)

Hagan, P.G.; Miner, F.J.

1980-01-01

Increasing the HNO 3 concentration decreases the filtration time but increases the plutonium concentration in the filtrate. A compromise was therefore necessary. If a minimum plutonium concentration is required in the filtrate, the acidity could be lowered to 1.9M with an approximate doubling in the filtration time. The H 2 O 2 concentration has little effect on filtration time. However, the higher the H 2 O 2 concentration, the less plutonium lost to the filtrate. Concentrations higher than the 22 moles/mole Pu recommended (at least up to 30 molar which was the highest investigated) would be beneficial if reagent costs are not excessive and production capacity exists for destroying the excess H 2 O 2 in the filtrate. Although the effect is not large, filtration time is shorter and the plutonium concentration in the filtrate is lower if metallic impurities are present. The slowest rate of H 2 O 2 addition investigated gives a plutonium peroxide precipitate with the fastest filtration time. The rate of addition has very little effect on the plutonium concentration in the filtrate. The temperature has little effect on the filtration time. 14 0 C is recommended since decomposition of H 2 O 2 would be slower at 14 0 C than at 22 0 C (min. Pu content in the filtrate). The effect of digestion time on both the filtration time and the plutonium content in the filtrate is minor, so the shortest digestion time investigated is recommended

The particulate material formed during the combusion of polonium and plutonium

Energy Technology Data Exchange (ETDEWEB)

Stewart, K.

1963-11-15

Over the last decade the production and use of uranium, plutonium and other radioactive materials has developed steadily. The work described in this paper is part of a reasonably comprehensive study of the extent and seriousness of the radiological hazards which could occur as the result of selected types of accidents. Additionally, these studies demonstrate the types of material and operations which are the more likely to produce a serious situation should a mishap occur. The emphasis in the paper, therefore, is on the differences between an element, polonium, of high specific activity which is volatile, and a reactive material, plutonium, of lower specific activity which is non-volatile but of somewhat greater radiotoxicity. The modes of formation of oxide aerosols from these two elements are, to a very considerable degree, representative of two important and contrasting processes.

Fuel cycles using adulterated plutonium

International Nuclear Information System (INIS)

Brooksbank, R.E.; Bigelow, J.E.; Campbell, D.O.; Kitts, F.G.; Lindauer, R.B.

1978-01-01

Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with 238 U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial

Non-fertile fuels for burning weapons plutonium in thermal fission reactors

International Nuclear Information System (INIS)

Lombardi, C.; Mazzola, A.; Vettraino, F.

1996-01-01

In the last few years, the excess plutonium disposition has become ever more a topical and critical issue. As a matter of fact, more than 200 MT of plutonium coming from spent fuel reprocessing have been already stockpiled and over the next decade, under the already ratified agreements, another about 200 MT of weapon-grade plutonium are expected to be available from nuclear weapons dismantlement. On this basis, an ever growing plutonium production is no longer the goal and the already stored quantities should be burnt in power reactors by taking care that no new plutonium is generated under irradiation. This new outlook in considering plutonium has led many designers to reassess the Fast Breeder Reactors (FBR) role and shifting from breeder to burner machines perspective. Several solutions for burning plutonium have been so far proposed and discussed from the safeguards, proliferation resistance, environmental safety, technological background, economy and time schedule standpoint. A proposal for plutonium burning in commercial Pressurized Water Reactors (PWR) by using a non-fertile oxide-type fuel consisting of PuO 2 diluted in an inert matrix is reported hereafter. This solution appears to receive an ever growing interest in the nuclear community. In order not to produce new plutonium during irradiation an innovative U-free fuel is being researched, based on an inert matrix which will consist in a mixed compound of inert oxides, such as ZrO 2 , Al2O 3 , MgO, CeO 2 where the plutonium oxide is dispersed in. The matrix will fulfill the following requirements: good chemical compatibility, acceptable thermal conductivity, good nuclear properties, good stability under irradiation, good dissolution resistance. The plutonium relative content will be comparable to that used in MOX fuel. The fuel is expected to be characterized by a high chemical stability (rock-like fuel), so that after discharge from reactor and adequate cooling time, it can be considered a High Level

Simulation and control synthesis for a pulse column separation system for plutonium--uranium recovery

International Nuclear Information System (INIS)

McCutcheon, E.B.

1975-05-01

Control of a plutonium-uranium partitioning column was studied using a mathematical model developed to simulate the dynamic response and to test postulated separation mechanisms. The column is part of a plutonium recycle flowsheet developed for the recovery of plutonium and uranium from metallurgical scrap. In the first step of the process, decontamination from impurities is achieved by coextracting plutonium and uranium in their higher oxidation states. In the second step, reduction of the plutonium to a lower oxidation state allows partitioning of the plutonium and uranium. The use of hydroxylamine for the plutonium reduction in this partitioning column is a unique feature of the process. The extraction operations are carried out in pulse columns. (U.S.)

Plutonium solubilities

International Nuclear Information System (INIS)

Puigdomnech, I.; Bruno, J.

1991-02-01

Thermochemical data has been selected for plutonium oxide, hydroxide, carbonate and phosphate equilibria. Equilibrium constants have been evaluated in the temperature range 0 to 300 degrees C at a pressure of 1 bar to T≤100 degrees C and at the steam saturated pressure at higher temperatures. Measured solubilities of plutonium that are reported in the literature for laboratory experiments have been collected. Solubility data on oxides, hydroxides, carbonates and phosphates have been selected. No solubility data were found at temperatures higher than 60 degrees C. The literature solubility data have been compared with plutonium solubilities calculated with the EQ3/6 geochemical modelling programs, using the selected thermodynamic data for plutonium. (authors)

The US plutonium materials conversion program in Russia

International Nuclear Information System (INIS)

Zygmunt, S.J.; Mason, C.F.V.; Hahn, W.K.

2000-01-01

Progress has been made in Russia towards the conversion of weapons-grade plutonium (w-Pu) into plutonium oxide (PuO 2 ) suitable for further manufacture into mixed oxide (MOX) fuels. This program was started in 1998 in response to US proliferation concerns and the acknowledged international need to decrease the available weapons-grade Pu. A similar agenda is being followed in the US to address disposition of US surplus weapons-grade Pu. In Russia a conversion process has been selected and a site proposed. This paper discusses the present state of the program in support of this future operating facility that will process up to 5 metric tons of plutonium a year. (authors)

Plutonium in the Arctic Marine Environment — A Short Review

Directory of Open Access Journals (Sweden)

Lindis Skipperud

2004-01-01

Full Text Available Anthropogenic plutonium has been introduced into the environment over the past 50 years as the result of the detonation of nuclear weapons and operational releases from the nuclear industry. In the Arctic environment, the main source of plutonium is from atmospheric weapons testing, which has resulted in a relatively uniform, underlying global distribution of plutonium. Previous studies of plutonium in the Kara Sea have shown that, at certain sites, other releases have given rise to enhanced local concentrations. Since different plutonium sources are characterised by distinctive plutonium-isotope ratios, evidence of a localised influence can be supported by clear perturbations in the plutonium-isotope ratio fingerprints as compared to the known ratio in global fallout. In Kara Sea sites, such perturbations have been observed as a result of underwater weapons tests at Chernaya Bay, dumped radioactive waste in Novaya Zemlya, and terrestrial runoff from the Ob and Yenisey Rivers. Measurement of the plutonium-isotope ratios offers both a means of identifying the origin of radionuclide contamination and the influence of the various nuclear installations on inputs to the Arctic, as well as a potential method for following the movement of water and sediment loads in the rivers.

Recovery and purification of uranium-234 from aged plutonium-238

International Nuclear Information System (INIS)

Keister, P.L.; Figgins, P.W.; Watrous, R.M.

1978-01-01

The current production methods used to recover and purify uranium-234 from aged plutonium-238 at Mound Laboratory are presented. The three chemical separation steps are described in detail. In the initial separation step, the bulk of the plutonium is precipitated as the oxalate. Successively lower levels of plutonium are achieved by anion exchange in nitrate media and by anion exchange in chloride media. The procedures used to characterize and analyze the final U 3 O 8 are given

Safety problems relating to plutonium recycling in light water reactors

International Nuclear Information System (INIS)

Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P.

1991-01-01

This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies

Safety problems relating to plutonium recycling in light water reactors

International Nuclear Information System (INIS)

Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P

1991-01-01

This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies. (author) [fr

The problem of utilization of the military uranium and plutonium

International Nuclear Information System (INIS)

Feoktistov, L.P.

1995-01-01

The problem on military uranium and plutonium is considered from the viewpoint of their utilization as a source of fissionable materials for NPPs. The solution consists in combining spherical geometry of critical mass with enriched center and the uranium burnup expansion recess. It is necessary thereby to obtain the minimum plutonium consumption in order to draw in unlimited quaintness of uranium-238 in the burnup process. It is estimated that hundred reactors with the total capacity of several hundred gigawatt may be involved into operation with the help of military plutonium. Refs. 2

Conversion of metal plutonium to plutonium dioxide by pyrochemical method

Energy Technology Data Exchange (ETDEWEB)

Panov, A.V.; Subbotin, V.G. [Russian Federal Nuclear Center, ALL-Russian Science and Research Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

2000-07-01

Report contains experimental results on metal plutonium of weapon origin samples conversion to plutonium dioxide by pyrochemical method. Circuits of processes are described. Their advantages and shortcomings are shown. Parameters of plutonium dioxide powders (phase and fraction compositions, poured density) manufactured by pyrochemical method in RFNC-VNIITF are shown as well. (authors)

Aqueous recovery of plutonium from pyrochemical processing residues

International Nuclear Information System (INIS)

Gray, L.W.; Gray, J.H.

1984-01-01

Pyrochemical processes provide rapid methods to reclaim plutonium from scrap residues. Frequently, however, these processes yield an impure plutonium product and waste residues that are contaminated with actinides and are therefore nondiscardable. The Savannah River Laboratory and Plant and the Rocky Flats Plant are jointly developing new processes using both pyrochemistry and aqueous chemistry to generate pure product and discardable waste. An example of residue being treated is that from the molten salt extraction (MSE), a mixture of NaCl, KCl, MgCl 2 , PuCl 3 , AmCl 3 , PuO 2 , and Pu 0 . This mixture is scrubbed with molten aluminum containing a small amount of magnesium to produce a nonhomogeneous Al-Pu-Am-Mg alloy. This process, which rejects most of the NaCl-KCl-MgCl 2 salts, results in a product easily dissolved in 6M HNO 3 -0.1M HF. Any residual chloride in the product is removed by precipitation with Hg(I) followed by centrifuging. Plutonium and americium are then separated by the standard Purex process. The americium, initially diverted to the solvent extraction waste stream, can either be recovered or sent to waste

PROCESS USING POTASSIUM LANTHANUM SULFATE FOR FORMING A CARRIER PRECIPITATE FOR PLUTONIUM VALUES

Science.gov (United States)

Angerman, A.A.

1958-10-21

A process is presented for recovering plutonium values in an oxidation state not greater than +4 from fluoride-soluble fission products. The process consists of adding to an aqueous acidic solution of such plutonium values a crystalline potassium lanthanum sulfate precipitate which carries the plutonium values from the solution.

Plutonium-238 and plutonium-239 metabolism in dairy cows following ingestion of mixed oxides

International Nuclear Information System (INIS)

Patzer, R.G.; Mullen, A.A.; Sutton, W.W.; Potter, G.D.; Mosley, R.E.; Efurd, D.W.; Stalnaker, N.D.

1985-01-01

Dairy cows were given oral dosage of plutonium-238 and plutonium-239 dioxide particles in a study to determine the relative gastrointestinal absorption and tissue distribution of the nuclides. Two cows were given particles in which the two isotopes were homogeneously mixed within the particles. A third cow was given two batches of particles which contained either plutonium-238 or plutonium-239. Results indicate that, when the two isotopes of plutonium are homogeneous within the particles, there is no difference between plutonium-238 and plutonium-239 in the relative gastrointestinal absorption and tissue distribution

High-temperature enthalpies of plutonium monocarbide and plutonium sesquicarbide

International Nuclear Information System (INIS)

Oetting, F.L.

1979-01-01

The high-temperature enthalpies of plutonium monocarbide and plutonium sesquicarbide have been determined with a copper-block calorimeter of the isoperibol type. The experimental enthalpy data, which was measured relative to 298 K, covered the temperature range from 400 to 1500 K. The calculation of the temperature rise of the calorimeter takes into account the added heat evolution from the radioactive decay of the plutonium samples. These enthalpy results, combined with the heat capacity and entropy of the respective carbide at 298 K available from the literature, has made it possible to generate tables of thermodynamic functions for the plutonium carbides. The behavior of the heat capacity of both of the plutonium carbides, i.e., a relatively steep increase in the heat capacity as the temperature increases, may be attributed to a premelting effect with the formation of vacancies within the crystal lattice although a theoretical treatment of this phenomenon is not given

A vision for environmentally conscious plutonium processing

International Nuclear Information System (INIS)

Avens, L.R.; Eller, P.G.; Christensen, D.C.; Miller, W.L.

1998-01-01

Regardless of individual technical and political opinions about the uses of plutonium, it is virtually certain that plutonium processing will continue on a significant global scale for many decades for the purposes of national defense, nuclear power and remediation. An unavoidable aspect of plutonium processing is that radioactive contaminated gas, liquid, and solid streams are generated. These streams need to be handled in a manner that is not only in full compliance with today's laws,but also will be considered environmentally and economically responsible now and in the future. In this regard, it is indeed ironic that the multibillion dollar and multidecade radioactive cleanup mortgage that the US Department of Energy (and its Russian counterpart) now owns resulted from waste management practices that were at the time in full legal compliance. The theme of this paper is that recent dramatic advances in actinide science and technology now make it possible to drastically minimize or even eliminate the problematic waste streams of traditional plutonium processing operations. Advanced technology thereby provides the means to avoid passing on to our children and grandchildren significant environmental and economic legacies that traditional processing inevitably produces. This paper will describe such a vision for plutonium processing that could be implemented fully within five years at a facility such as the Los Alamos Plutonium Facility (TA55). As a significant bonus, even on this short time scale, the initial technology investment is handsomely returned in avoided waste management costs

International and institutional aspects of reprocessing and plutonium management

International Nuclear Information System (INIS)

1978-09-01

Various institutional alternatives applicable to reprocessing, plutonium management and recycle are considered, not as a definitive analysis but rather as a basis for identifying the institutional approaches and measures which the Working Group might wish to examine more thoroughly. Seven alternatives arrangements for reprocessing are presented. These range from suspending the operation of existing reprocessing plants through placing national facilities under safeguards to limiting reprocessing to a few large facilities subject to plutonium management, multinational or international control. Finally, the comprehensive alternative of an International Nuclear Fuel Authority with worldwide responsibility for reprocessing and plutonium management is considered. Plutonium management alternatives to complement the reprocessing options, are then outlined. These include national discretion on the separation and disposition of plutonium under safeguards, an agreed Code of Practice for plutonium management at national facilities and the international storage of plutonium. The advantages and disadvantages of the alternative are discussed tentatively. It is recognised that the alternatives are presented in a simplified form and that their elements can be combined or separated in many ways. Although strengthening the institutions relating to the peaceful uses of nuclear energy is imperative and can contribute to non-proliferation, such arrangements might open other proliferation risks through the spread of sensitive materials, facilities and technology. While there are risks with any fuel cycle, where plutonium in quantity is separated these risks are of a high order. Although these can be mitigated, they will have to be set against the energy and economic case for reprocessing and alternatives other than plutonium considered

Plutonium: An introduction

International Nuclear Information System (INIS)

Condit, R.H.

1993-10-01

This report is a summary of the history and properties of plutonium. It presents information on the atoms, comparing chemical and nuclear properties. It looks at the history of the atom, including its discovery and production methods. It summarizes the metallurgy and chemistry of the element. It also describes means of detecting and measuring the presence and quantity of the element

The solubility of solid fission products in carbides and nitrides of uranium and plutonium. Part I: literature review on experimental results

International Nuclear Information System (INIS)

Benedict, U.

1977-01-01

This review compiles the available data on the solubility of the most important non-volatile fission products in the carbides, nitrides, and carbonitrides of uranium and plutonium. It includes some elements which are not fission products, but belong to a group of the Periodic Table which contains one or more fission products elements

MICROBIAL TRANSFORMATIONS OF PLUTONIUM AND IMPLICATIONS FOR ITS MOBILITY.

Energy Technology Data Exchange (ETDEWEB)

FRANCIS, A.J.

2000-09-30

The current state of knowledge of the effect of plutonium on microorganisms and microbial activity is reviewed, and also the microbial processes affecting its mobilization and immobilization. The dissolution of plutonium is predominantly due to their production of extracellular metabolic products, organic acids, such as citric acid, and sequestering agents, such as siderophores. Plutonium may be immobilized by the indirect actions of microorganisms resulting in changes in Eh and its reduction from a higher to lower oxidation state, with the precipitation of Pu, its bioaccumulation by biomass, and bioprecipitation reactions. In addition, the abundance of microorganisms in Pu-contaminated soils, wastes, natural analog sites, and backfill materials that will be used for isolating the waste and role of microbes as biocolloids in the transport of Pu is discussed.

Plutonium Immobilization Bagless Transfer Can Size Evaluation

International Nuclear Information System (INIS)

Kriikku, E.; Stokes, M.; Rogers, L.; Ward, C.

1998-02-01

This report identifies and documents the most appropriate bagless transfer can size to support Plutonium Immobilization Can Loading operations. Also, this report considers can diameter, can wall thickness, and can length

LITERATURE REVIEW FOR OXALATE OXIDATION PROCESSES AND PLUTONIUM OXALATE SOLUBILITY

Energy Technology Data Exchange (ETDEWEB)

Nash, C.

2012-02-03

A literature review of oxalate oxidation processes finds that manganese(II)-catalyzed nitric acid oxidation of oxalate in precipitate filtrate is a viable and well-documented process. The process has been operated on the large scale at Savannah River in the past, including oxidation of 20 tons of oxalic acid in F-Canyon. Research data under a variety of conditions show the process to be robust. This process is recommended for oxalate destruction in H-Canyon in the upcoming program to produce feed for the MOX facility. Prevention of plutonium oxalate precipitation in filtrate can be achieved by concentrated nitric acid/ferric nitrate sequestration of oxalate. Organic complexants do not appear practical to sequester plutonium. Testing is proposed to confirm the literature and calculation findings of this review at projected operating conditions for the upcoming campaign. H Canyon plans to commence conversion of plutonium metal to low-fired plutonium oxide in 2012 for eventual use in the Mixed Oxide Fuel (MOX) Facility. The flowsheet includes sequential operations of metal dissolution, ion exchange, elution, oxalate precipitation, filtration, and calcination. All processes beyond dissolution will occur in HB-Line. The filtration step produces an aqueous filtrate that may have as much as 4 M nitric acid and 0.15 M oxalate. The oxalate needs to be removed from the stream to prevent possible downstream precipitation of residual plutonium when the solution is processed in H Canyon. In addition, sending the oxalate to the waste tank farm is undesirable. This report addresses the processing options for destroying the oxalate in existing H Canyon equipment.

Dynamic process model of a plutonium oxalate precipitator. Final report

Energy Technology Data Exchange (ETDEWEB)

Miller, C.L.; Hammelman, J.E.; Borgonovi, G.M.

1977-11-01

In support of LLL material safeguards program, a dynamic process model was developed which simulates the performance of a plutonium (IV) oxalate precipitator. The plutonium oxalate precipitator is a component in the plutonium oxalate process for making plutonium oxide powder from plutonium nitrate. The model is based on state-of-the-art crystallization descriptive equations, the parameters of which are quantified through the use of batch experimental data. The dynamic model predicts performance very similar to general Hanford oxalate process experience. The utilization of such a process model in an actual plant operation could promote both process control and material safeguards control by serving as a baseline predictor which could give early warning of process upsets or material diversion. The model has been incorporated into a FORTRAN computer program and is also compatible with the DYNSYS 2 computer code which is being used at LLL for process modeling efforts.

Dynamic process model of a plutonium oxalate precipitator. Final report

International Nuclear Information System (INIS)

Miller, C.L.; Hammelman, J.E.; Borgonovi, G.M.

1977-11-01

In support of LLL material safeguards program, a dynamic process model was developed which simulates the performance of a plutonium (IV) oxalate precipitator. The plutonium oxalate precipitator is a component in the plutonium oxalate process for making plutonium oxide powder from plutonium nitrate. The model is based on state-of-the-art crystallization descriptive equations, the parameters of which are quantified through the use of batch experimental data. The dynamic model predicts performance very similar to general Hanford oxalate process experience. The utilization of such a process model in an actual plant operation could promote both process control and material safeguards control by serving as a baseline predictor which could give early warning of process upsets or material diversion. The model has been incorporated into a FORTRAN computer program and is also compatible with the DYNSYS 2 computer code which is being used at LLL for process modeling efforts

Preparation of plutonium hexafluoride. Recovery of plutonium from waste dross (1962)

International Nuclear Information System (INIS)

Gendre, R.

1962-01-01

The object of this work is to study the influence of various physical factors on the rate of fluorination of solid plutonium tetrafluoride by fluorine. In a horizontal oven with a circulation for pure fluorine at atmospheric pressure and 520 deg. C, at a fluorine rate of 9 litres/hour, it is possible to transform 3 g of tetrafluoride to hexafluoride with about 100 per cent transformation and a recovery yield of over 90 per cent, in 4 to 5 hours. The fluorination rate is a function of the temperature, of the fluorine flow-rate, of the crucible surface, of the depth of the tetrafluoride layer and of the reaction time. It does not depend on the diffusion of the fluorine into the solid but is determined by the reaction at the gas-solid interface and obeys the kinetic law (1 - T T ) 1/3 = kt + 1. The existence of intermediate fluorides, in particular Pu 4 F 17 , is confirmed by a break in the Arrhenius plot at about 370 deg. C, by differences in the fluorination rates inside the tetrafluoride layer, and by reversible colour changes. The transformation to hexafluoride occurs with a purification with respect of the foreign elements present in the initial plutonium. Recovery of plutonium from waste dross: The study is based on the transformation of occluded plutonium particles to gaseous hexafluoride which is then decomposed thermally to the tetrafluoride which can be reintroduced directly in the production circuit. Under the conditions considered this process is not applicable industrially. After milling, it is possible to separate the dross into enriched (75 per cent Pu in 2.6 per cent by weight of dross) and depleted portions. By prolonged fluorination (16 hours) of the various fractions it is possible to recover about 80 per cent of the plutonium. A treatment plant using fluidization, as described at the end of this study, should make it possible to substantially improve the yield. (author) [fr

Plutonium

International Nuclear Information System (INIS)

Koelzer, W.

1989-03-01

This report contains with regard to 'plutonium' statements on chemistry, occurrence and reactions in the environment, handling procedures in the nuclear fuel cycle, radiation protection methods, biokinetics, toxicology and medical treatment to make available reliable data for the public discussion on plutonium especially its use in nuclear power plants and its radiological assessment. (orig.) [de

Surplus plutonium disposition draft environmental impact statement. Volume 1, Part A

International Nuclear Information System (INIS)

1998-07-01

On May 22, 1997, DOE published a Notice of Intent (NOI) in the Federal Register (62 Federal Register 28009) announcing its decision to prepare an environmental impact statement (EIS) that would tier from the analysis and decisions reached in connection with the Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic EIS (Storage and Disposition PEIS). DOE's disposition strategy allows for both the immobilization of surplus plutonium and its use as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of the immobilized plutonium and MOX fuel (as spent nuclear fuel) in a geologic repository. The Surplus Plutonium Disposition Environmental Impact Statement analyzes alternatives that would use the immobilization approach (for some of the surplus plutonium) and the MOX fuel approach (for some of the surplus plutonium); alternatives that would immobilize all of the surplus plutonium; and the No Action Alternative. The alternatives include three disposition facilities that would be designed so that they could collectively accomplish disposition of up to 50 metric tons (55 tons) of surplus plutonium over their operating lives: (1) the pit disassembly and conversion facility would disassemble pits (a weapons component) and convert the recovered plutonium, as well as plutonium metal from other sources, into plutonium dioxide suitable for disposition; (2) the immobilization facility would include a collocated capability for converting nonpit plutonium materials into plutonium dioxide suitable for immobilization and would be located at either Hanford or SRS. DOE has identified SRS as the preferred site for an immobilization facility; (3) the MOX fuel fabrication facility would fabricate plutonium dioxide into MOX fuel. This volume includes background information; purpose of and need for the proposed action; alternatives for disposition of surplus weapons useable plutonium; and

Plutonium

International Nuclear Information System (INIS)

Mueller-Christiansen, K.; Wollesen, M.

1979-01-01

As emotions and fear of plutonium are neither useful for the non-professionals nor for the political decision makers and the advantages and disadvantages of plutonium can only put against each other under difficulties, the paper wants to present the most essential scientific data of plutonium in a generally understandable way. Each of the individual sections is concluded and they try to give an answer to the most discussed questions. In order to make understanding easier, the scientific facts are only brought at points where it cannot be done without for the correctness of the presentation. Many details were left out knowingly. On the other hand, important details are dealt with several times if it seems necessary for making the presentation correct. The graphical presentations and the figures in many cases contain more than said in the text. They give the interested reader hints to scientific-technical coherences. The total material is to enable the reader to form his own opinion on plutonium problems which are being discussed in public. (orig./HP) [de

Improvements in the detection of airborne plutonium

International Nuclear Information System (INIS)

Ryden, D.J.

1981-02-01

It is shown how it is possible to compensate individually for each of the background components on the filter paper used to collect samples. Experimentally it has been shown that the resulting compensated background count-rate averages zero with a standard deviation very close to the fundamental limit set by random statistical variations. Considerable improvements in the sensitivity of detecting airborne plutonium have been achieved. Two new plutonium-in-air monitors which use the compensation schemes described in this report are now available. Both have operated successfully in high concentrations of radon daughters. (author)

Development of a plutonium solution-assay instrument with isotopic capability

International Nuclear Information System (INIS)

Hsue, S.T.; Marks, T.

1992-01-01

A new generation of solution-assay instrument has been developed to satisfy all the assay requirements of an aqueous plutonium-recovery operation. The assay is based on a transmission-corrected passive assay technique. We have demonstrated that the system can cover a concentration range of 0.5--300 g/ell with simultaneous isotopic determination. The system can be used to assay input and eluate streams of the recovery operation. The system can be modified to measure low-concentration effluent solutions from the recovery operation covering 0.01--40 g/ell. The same system has also been modified to assay plutonium solutions enriched in 242 Pu. 6 refs

Plutonium in the aquatic environment around the Rocky Flats facility

International Nuclear Information System (INIS)

Thompson, M.A.

1975-01-01

The Rocky Flats Plant of the United States Energy Research and Development Administration has been fabricating and chemically recovering plutonium for over 20 years. During that time, small amounts of plutonium have been released with liquid process and sanitary waste discharges. The liquid waste flows through a series of holding ponds from which it is discharged into a creek that is part of a municipal drinking water supply. The water flows for about 1.5 km between the last holding pond and the municipal drinking water reservoir. In addition, liquid wastes containing high levels of chemical contaminants and plutonium concentrations less than allowable drinking water standards have been discharged to large evaporation ponds. The fate of the plutonium in both the surface and subsurface aquatic environment has been extensively monitored and studied. It has been found that plutonium does not move very far or very rapidly through subsurface water. The majority of the plutonium released through surface water has been contained in the sediments of the plant holding ponds. Small amounts of plutonium have also been found in the sediments of the draining creek and in the sediments of the receiving reservoir. Higher than normal amounts of plutonium were released from the waste treatment plants during times when suspended solids were high. Various biological species have been examined and plutonium concentration factors determined. Considerably less than 1% of the 210 mCi of plutonium released has been detected in biological systems including man. After more than 20 years of large scale operations, no health or environmental hazard has been identified due to the release of small amounts of plutonium. (author)

Plutonium Immobilization Project - Robotic canister loading

International Nuclear Information System (INIS)

Hamilton, R.L.

2000-01-01

The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF)

Utilization of Cs137 to generate a radiation barrier for weapons grade plutonium immobilized in borosilicate glass canisters. Revision 1

International Nuclear Information System (INIS)

Jardine, L.J.; Armantrout, G.A.; Collins, L.F.

1995-01-01

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 137 . One hundred thousand curies of Cs 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 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 137 inventories exist in the United States' HLWs from past reprocessing operations, should additional curies be required. Using only one fission product, CS 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

1981 Annual Status Report. Plutonium fuels and actinide programme

International Nuclear Information System (INIS)

1981-01-01

In this 1981 report the work carried out by the European Institute for Transuranium elements is reviewed. Main topics are: operation limits of plutonium fuels: swelling of advanced fuels, oxide fuel transients, equation of state of nuclear materials; actinide cycle safety: formation of actinides (FACT), safe handling of plutonium fuel (SHAPE), aspects of the head-end processing of carbide fuel (RECARB); actinide research: crystal chemistry, solid state studies, applied actinide research

Radial plutonium redistribution in mixed-oxide fuel

International Nuclear Information System (INIS)

Lawrence, L.A.; Schwinkendorf, K.N.; Karnesky, R.A.

1981-10-01

Alpha autoradiographs from all HEDL fuel pin metallography samples are evaluated and catalogued according to different plutonium distribution patterns. The data base is analyzed for effects of fabrication and operating parameters on redistribution

Plutonium scrap processing at the Los Alamos Scientific Laboratory

International Nuclear Information System (INIS)

Nixon, A.E.; McKerley, B.J.; Christensen, E.L.

1980-01-01

The Los Alamos Scientific Laboratory currently has the newest plutonium handling facility in the nation. Los Alamos has been active in the processing of plutonium almost since the discovery of this man-made element in 1941. One of the functions of the new facility is the processing of plutonium scrap generated at LASL and other sites. The feed for the scrap processing program is extremely varied, and a wide variety of contaminants are often encountered. Depending upon the scrap matrix and contaminants present, the majority of material receives a nitric acid/hydrofluoric acid or nitric acid/calcium fluoride leach. The plutonium nitrate solutions are then loaded onto an anion exchange column charged with DOWEX 1 x 4, 50 to 100 mesh, nitrate form resin. The column is eluted with 0.48 M hydroxyl amine nitrate. The Pu(NO 3 ) 3 is then precipitated as plutonium III oxalate which is calcined at 450 to 500 0 C to yield a purified PuO 2 product

Plutonium controversy

International Nuclear Information System (INIS)

Richmond, C.R.

1980-01-01

The toxicity of plutonium is discussed, particularly in relation to controversies surrounding the setting of radiation protection standards. The sources, amounts of, and exposure pathways of plutonium are given and the public risk estimated

Perspective on plutonium

International Nuclear Information System (INIS)

Sun, L.S.

1993-01-01

This paper is intended as a brief overview on the element plutonium. Plutonium is the first primarily man-made element to play a significant role not only in technological development, but also in the economic growth of many countries. The importance of plutonium centers around its enormous energy making it ideal for wide-scale use in reactors, while the nuclear industry continues to work toward improving safety and efficiency of plutonium as a reactor fuel politicians and the public still debate over the safety and benefits of nuclear power. (30 refs.)

Remote handling in the Plutonium Immobilization Project: Puck handling

International Nuclear Information System (INIS)

Brault, J.R.

2000-01-01

Since the break up of the Soviet Union at the end of the Cold War, the US and Russia have been negotiating ways to reduce their nuclear stockpiles. Economics is one of the reasons behind this, but another important reason is safeguarding these materials from unstable organizations and countries. With the downsizing of the nuclear stockpiles, large quantities of plutonium are being declared excess and must be safely disposed of. The Savannah River Site (SRS) has been selected as the site where the immobilization facility will be located. Conceptual design and process development commenced in 1998. SRS will immobilize excess plutonium in a ceramic waste form and encapsulate it in vitrified high level waste in the Defense Waste Processing Facility (DWPF) canister. These canisters will then be interred in the national repository at Yucca Mountain, New Mexico. The facility is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. This paper will discuss the first two operations

Surplus plutonium disposition draft environmental impact statement. Volume 1, Part B

International Nuclear Information System (INIS)

1998-07-01

On May 22, 1997, DOE published a Notice of Intent (NOI) in the Federal Register (62 Federal Register 28009) announcing its decision to prepare an environmental impact statement (EIS) that would tier from the analysis and decisions reached in connection with the Storage and Disposition of Weapons-Usable Fissile Materials Final Programmatic EIS (Storage and Disposition PEIS). DOE's disposition strategy allows for both the immobilization of surplus plutonium and its use as mixed oxide (MOX) fuel in existing domestic, commercial reactors. The disposition of surplus plutonium would also involve disposal of the immobilized plutonium and MOX fuel (as spent nuclear fuel) in a geologic repository. The Surplus Plutonium Disposition Environmental Impact Statement analyzes alternatives that would use the immobilization approach (for some of the surplus plutonium) and the MOX fuel approach (for some of the surplus plutonium); alternatives that would immobilize all of the surplus plutonium; and the No Action Alternative. The alternatives include three disposition facilities that would be designed so that they could collectively accomplish disposition of up to 50 metric tons (55 tons) of surplus plutonium over their operating lives: (1) the pit disassembly and conversion facility would disassemble pits (a weapons component) and convert the recovered plutonium, as well as plutonium metal from other sources, into plutonium dioxide suitable for disposition; (2) the immobilization facility would include a collocated capability for converting nonpit plutonium materials into plutonium dioxide suitable for immobilization and would be located at either Hanford or SRS. DOE has identified SRS as the preferred site for an immobilization facility; (3) the MOX fuel fabrication facility would fabricate plutonium dioxide into MOX fuel. This volume has chapters on environmental consequences; environmental regulations, permits, and consultations; a glossary; list of preparers; distribution list

Neutron Resonance Transmission Analysis (NRTA): Initial Studies of a Method for Assaying Plutonium in Spent Fuel

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester; James W. Sterbentz

2011-05-01

Neutron Resonance Transmission Analysis (NRTA) is an analytical technique that uses neutrons to assay the isotopic content of bulk materials. The technique uses a pulsed accelerator to produce an intense, short pulse of neutrons in a time-of-flight configuration. These neutrons, traveling at different speeds according to their energy, can be used to interrogate a spent fuel (SF) assembly to determine its plutonium content. Neutron transmission through the assembly is monitored as a function of neutron energy (time after the pulse), similar to the way neutron cross-section data is often collected. The transmitted neutron intensity is recorded as a function of time, with faster (higher-energy) neutrons arriving first and slower (lower-energy) neutrons arriving later. The low-energy elastic scattering and absorption resonances of plutonium and other isotopes modulate the transmitted neutron spectrum. Plutonium content in SF can be determined by analyzing this attenuation. Work is currently underway at Idaho National Laboratory, as a part of United States Department of Energy's Next Generation Safeguards Initiative (NGSI), to investigate the NRTA technique and to assess its feasibility for quantifying the plutonium content in SF and for determining the diversion of SF pins from assemblies. Preliminary results indicate that NRTA has great potential for being able to assay intact SF assemblies. Operating in the 1-40 eV range, it can identify four plutonium isotopes (239, 240, 241, & 242Pu), three uranium isotopes (235, 236, & 238U), and six resonant fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm). It can determine the areal density or mass of these isotopes in single- or multiple-pin integral transmission scans. Further, multiple observables exist to allow the detection of material diversion (pin defects) including fast-neutron and x-ray radiography, gross-transmission neutron counting, plutonium resonance absorption analysis, and fission-product

Plutonium controversy

Energy Technology Data Exchange (ETDEWEB)

Richmond, C.R.

1980-01-01

The toxicity of plutonium is discussed, particularly in relation to controversies surrounding the setting of radiation protection standards. The sources, amounts of, and exposure pathways of plutonium are given and the public risk estimated. (ACR)

Plutonium, nuclear fuel; Le plutonium, combustible nucleaire

Energy Technology Data Exchange (ETDEWEB)

Grison, E [Commissariat a l' Energie Atomique, Fontenay aux Roses (France). Centre d' Etudes Nucleaires, Saclay

1960-07-01

A review of the physical properties of metallic plutonium, its preparation, and the alloys which it forms with the main nuclear metals. Appreciation of its future as a nuclear fuel. (author) [French] Apercu sur les proprietes physiques du plutonium metallique, sa preparation, ses alliages avec les principaux metaux nucleaires. Consideration sur son avenir en tant que combustible nucleaire. (auteur)

Measurement of Plutonium Isotopic Composition - MGA

Energy Technology Data Exchange (ETDEWEB)

Vo, Duc Ta [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-08-21

In this module, we will use the Canberra InSpector-2000 Multichannel Analyzer with a high-purity germanium detector (HPGe) and the MGA isotopic anlysis software to assay a variety of plutonium samples. The module provides an understanding of the MGA method, its attributes and limitations. You will assess the system performance by measuring a range of materials similar to those you may assay in your work. During the final verification exercise, the results from MGA will be combined with the 240Pueff results from neutron coincidence or multiplicity counters so that measurements of the plutonium mass can be compared with the operator-declared (certified) values.

Advanced fuels for plutonium management in pressurized water reactors

International Nuclear Information System (INIS)

Vasile, A.; Dufour, Ph.; Golfier, H.; Grouiller, J.P.; Guillet, J.L.; Poinot, Ch.; Youinou, G.; Zaetta, A.

2003-01-01

Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1 . More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

Matrix Characterization of Plutonium Residues by Alpha-Particle Self-Interrogation

International Nuclear Information System (INIS)

Prettyman, T.H.; Foster, L.A.; Staples, P.

1998-01-01

Legacy plutonium residues often have inadequate item descriptions. Nondestructive characterization can help segregate these items for reprocessing or provide information needed for disposal or storage. Alpha particle-induced gamma-ray spectra contain a wealth of information that can be used for matrix characterization. We demonstrate how this information can be used for item identification. Gamma-ray spectra were recorded at the Los Alamos Plutonium Facility from a variety of legacy, plutonium-processing residues and product materials. The comparison and analysis of these spectra are presented

Phenomenology of uranium-plutonium homogenization in nuclear fuels

International Nuclear Information System (INIS)

Marin, J.M.

1988-01-01

The uranium and plutonium cations distribution in mixed oxide fuels (U 1-y Pu y )O 2 with y ≤ 0.1 has been studied in laboratory with industrial fabrication methods. Our experiences has showed a slow cations migration. In the substoichiometry (UPu)O 2-x the diffusion is in connection with the plutonium valence which is an indicator of the oxidoreduction state of the crystal lattice. The plutonium valence is in connection with the oxygen ion deficit in order to compensate the electrical charge. The oxygen ratio of the solid depends of the oxygen partial pressure prevailing at the time of product elaboration but it can be modified by impurities. These impurities permit to increase or decrease the fuel characteristics and performances. An homogeneity analysis methodology is proposed, its objective is to classify the mixed oxide fuels according to the uranium and plutonium ions distribution [fr

A vision for environmentally conscious plutonium processing

International Nuclear Information System (INIS)

Avens, L.R.; Eller, P.G.; Christensen, D.C.; Miller, W.L.

1998-01-01

Regardless of individual technical and political opinions about the uses of plutonium, it is virtually certain that plutonium processing will continue on a significant global scale for many decades for the purposes of national defense, nuclear power, and remediation. An unavoidable aspect of plutonium processing is that radioactively contaminated gas, liquid, and solid waste streams are generated. These streams need to be handled in a manner that not only is in full compliance with today's laws but also will be considered environmentally and economically responsible now and in the future. In this regard, it is indeed ironic that the multibillion dollar and multidecade radioactive cleanup mortgage that the US Department of Energy (and its Russian counterpart) now owns resulted from waste management practices that were at the time in full legal compliance. It is now abundantly evident that in the long run, these practices have proven to be neither environmentally nor economically sound. Recent dramatic advances in actinide science and technology now make it possible to drastically minimize or even eliminate the problematic waste streams of traditional plutonium processing operations. Advanced technology thereby provides the means to avoid passing on to children and grandchildren significant environmental and economic legacies that traditional processing inevitably produces. The authors describe such a vision for plutonium processing that could be implemented fully within 5 yr at a facility such as the Los Alamos National Laboratory Plutonium Facility (TA55). As a significant bonus, even on this short timescale, the initial technology investment is handsomely returned in avoided waste management costs

Preparation of plutonium hexafluoride. Recovery of plutonium from waste dross (1962); Preparation de l'hexafluorure de plutonium. Recuperation du plutonium des scories d'elaboration (1962)

Energy Technology Data Exchange (ETDEWEB)

Gendre, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1962-07-01

The object of this work is to study the influence of various physical factors on the rate of fluorination of solid plutonium tetrafluoride by fluorine. In a horizontal oven with a circulation for pure fluorine at atmospheric pressure and 520 deg. C, at a fluorine rate of 9 litres/hour, it is possible to transform 3 g of tetrafluoride to hexafluoride with about 100 per cent transformation and a recovery yield of over 90 per cent, in 4 to 5 hours. The fluorination rate is a function of the temperature, of the fluorine flow-rate, of the crucible surface, of the depth of the tetrafluoride layer and of the reaction time. It does not depend on the diffusion of the fluorine into the solid but is determined by the reaction at the gas-solid interface and obeys the kinetic law (1 - T{sub T}){sup 1/3} = kt + 1. The existence of intermediate fluorides, in particular Pu{sub 4} F{sub 17}, is confirmed by a break in the Arrhenius plot at about 370 deg. C, by differences in the fluorination rates inside the tetrafluoride layer, and by reversible colour changes. The transformation to hexafluoride occurs with a purification with respect of the foreign elements present in the initial plutonium. Recovery of plutonium from waste dross: The study is based on the transformation of occluded plutonium particles to gaseous hexafluoride which is then decomposed thermally to the tetrafluoride which can be reintroduced directly in the production circuit. Under the conditions considered this process is not applicable industrially. After milling, it is possible to separate the dross into enriched (75 per cent Pu in 2.6 per cent by weight of dross) and depleted portions. By prolonged fluorination (16 hours) of the various fractions it is possible to recover about 80 per cent of the plutonium. A treatment plant using fluidization, as described at the end of this study, should make it possible to substantially improve the yield. (author) [French] L'objet de l'etude est l'influence des differents

Standard test method for plutonium assay by plutonium (III) diode array spectrophotometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

2002-01-01

1.1 This test method describes the determination of total plutonium as plutonium(III) in nitrate and chloride solutions. The technique is applicable to solutions of plutonium dioxide powders and pellets (Test Methods C 697), nuclear grade mixed oxides (Test Methods C 698), plutonium metal (Test Methods C 758), and plutonium nitrate solutions (Test Methods C 759). Solid samples are dissolved using the appropriate dissolution techniques described in Practice C 1168. The use of this technique for other plutonium-bearing materials has been reported (1-5), but final determination of applicability must be made by the user. The applicable concentration range for plutonium sample solutions is 10–200 g Pu/L. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropria...

The plutonium recycle for PWR reactors from brazilian nuclear program

International Nuclear Information System (INIS)

Rubini, L.A.

1978-01-01

The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

The approaching plutonium surplus: a Japanese/European predicament

International Nuclear Information System (INIS)

Berkhout, F.; Suzuki, Tatsujiro; Walker, W.

1990-01-01

The vision of cheap energy independence through civil nuclear power has faded in the late 1980s as the costs of nuclear programmes have come to light. But present nuclear plans in Japan, France and Britain are set to produce quantities of plutonium in the 1990s which, the authors argue, will be surplus to requirements and very difficult to use. The over-production of plutonium will pose political and security problems - particularly for Japan, which owns large stocks of separated plutonium in Europe that it may not be able to bring back home. Nuclear strategy in Japan and Europe needs to be rethought if serious international problems are to be avoided. (author)

Plutonium multi-recycling in increased moderating ratio reactors (IMR)

International Nuclear Information System (INIS)

Barbrault, P.; Larderet, P.

1998-01-01

The large core of the future jointly defined European PWR (EPR), would be compatible with an increased Moderating Ratio (MR) enabling better plutonium burnout. The purpose of current work on the subject is to assess plutonium multi-recycling possibilities in IMR reactors. What additional operating constraints would be involved under normal and accidental conditions and are they acceptable? The conclusion is that Plutonium multi-recycling in a PWR of the type envisaged for the EPR raises no major problems under the following conditions: use of an IMR MOX core, enhancing both plutonium burnout and absorber efficiency; use of enriched boron in both the primary coolant soluble boron and the B4C boron carbide in the control rods. Deeper investigation should be performed concerning the partial or total core drain-out, in view of the high total Pu concentrations involved (13%) and the types of core considered (100% MOX). (author)

Dissolution of aerosol particles collected from nuclear facility plutonium production process

International Nuclear Information System (INIS)

Ning Xu; Martinez, Alex; Schappert, Michael; Montoya, D.P.; Martinez, Patrick; Tandon, Lav

2016-01-01

A simple, robust analytical chemistry method has been developed to dissolve plutonium containing particles in a complex matrix. The aerosol particles collected on Marple cascade impactor substrates were shown to be dissolved completely with an acid mixture of 12 M HNO 3 and 0.1 M HF. A pressurized closed vessel acid digestion technique was utilized to heat the samples at 130 deg C for 16 h to facilitate the digestion. The dissolution efficiency for plutonium particles was 99 %. The resulting particle digestate solution was suitable for trace elemental analysis and isotope composition determination, as well as radiochemistry measurements. (author)

Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

International Nuclear Information System (INIS)

Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

2009-01-01

Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

Plutonium storage criteria

Energy Technology Data Exchange (ETDEWEB)

Chung, D. [Scientech, Inc., Germantown, MD (United States); Ascanio, X. [Dept. of Energy, Germantown, MD (United States)

1996-05-01

The Department of Energy has issued a technical standard for long-term (>50 years) storage and will soon issue a criteria document for interim (<20 years) storage of plutonium materials. The long-term technical standard, {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides,{close_quotes} addresses the requirements for storing metals and oxides with greater than 50 wt % plutonium. It calls for a standardized package that meets both off-site transportation requirements, as well as remote handling requirements from future storage facilities. The interim criteria document, {open_quotes}Criteria for Interim Safe Storage of Plutonium-Bearing Solid Materials{close_quotes}, addresses requirements for storing materials with less than 50 wt% plutonium. The interim criteria document assumes the materials will be stored on existing sites, and existing facilities and equipment will be used for repackaging to improve the margin of safety.

Plutonium finishing plant dangerous waste training plan

International Nuclear Information System (INIS)

ENTROP, G.E.

1999-01-01

This training plan describes general requirements, worker categories, and provides course descriptions for operation of the Plutonium Finish Plant (PFP) waste generation facilities, permitted treatment, storage and disposal (TSD) units, and the 90-Day Accumulation Areas

Toxicology of plutonium

International Nuclear Information System (INIS)

Bair, W.J.

1974-01-01

Data are reviewed from studies on the toxicity of Pu in experimental animals. Of the several plutonium isotopes, only 238 Pu and 239 Pu have been studied well. Sufficient results have been obtained to show that the behavior of 238 Pu in biological systems and the resulting biological effects cannot be precisely predicted from studies of 239 Pu. This probably applies also to other radiologically important plutonium isotopes which have half-lives ranging from 45 days to 10 7 years and decay by β-emission, electron capture, and spontaneous fission, as well as by emission of α-particles. All the biological effects of plutonium described in this review are attributed to alpha-particle radiation emitted by the plutonium. However, since plutonium is a chemically active heavy metal, one cannot ignore the possibility of chemical toxicity of the low-specific-activity isotopes, 239 Pu, 242 Pu, and 244 Pu. The preponderance of our knowledge of plutonium toxicology has come from short-term studies of relatively high dosage levels in several animal species. The consequences of high-level internal exposures can be predicted with confidence in experimental animals and probably also in man. However, considering the care with which plutonium is handled in the nuclear industry, a high-level contamination event is unlikely. Considerably less is known about the long-term effects of low levels of contamination. (250 references) (U.S.)

Plutonium chemistry of the ocean

International Nuclear Information System (INIS)

Folsom, T.R.

1972-01-01

Plutonium is a man-made element whose behavior in the marine environment is inadequately known at present. It has been studied intensively in connection with production of weapons and power sources and has been characterized as an extremely toxic substance. Nevertheless, only a few dozen measurements have been made of concentrations in seawater and in the associated organisms and sediments. The first of these were as recent as 1964. There are reasons to believe its chemical behavior in the ocean is different from what has been observed on land, and that it will be difficult to predict how plutonium will distribute itself in the ocean. The consequences of increased environmental concentrations of Pu are discussed

Plutonium isotopic assay of reprocessing product solutions in the KfK K-edge densitometer

International Nuclear Information System (INIS)

Eberle, H.; Ottmar, H.; Matussek, P.

1985-04-01

The KfK K-edge densiometer, designed for accurate element concentration measurements using the technique of X-ray absorptiometry at the K absorption edge, provides as an additional option the possibility to determine the isotopic composition of freshly separated plutonium from an gamma-spectrometric analysis of its self-radiation. This report describes the underlying methodology and experimental procedures for the isotopic analysis in the K-edge densitometer. The paper also presents and discusses the experimental results so far obtained from routine measurements on reprocessing product solutions. (orig.)

Decommissioning the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I)

International Nuclear Information System (INIS)

Harper, J.R.; Garde, R.

1981-11-01

The Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I) was decommissioned at the Los Alamos National Laboratory, Los Alamos, New Mexico, in 1980. The LAMPRE I was a sodium-cooled reactor built to develop plutonium fuels for fast breeder applications. It was retired in the mid-1960s. This report describes the decommissioning procedures, the health physics programs, the waste management, and the costs for the operation

Production of americium isotopes in France

International Nuclear Information System (INIS)

Koehly, G.; Bourges, J.; Madic, C.; Nguyen, T.H.; Lecomte, M.

1984-12-01

The program of productions of americium 241 and 243 isotopes is based respectively on the retreatment of aged plutonium alloys or plutonium dioxide and on the treatment of plutonium targets irradiated either in CELESTIN reactors for Pu-Al alloys or OSIRIS reactor for plutonium 242 dioxide. All the operations, including americium final purifications, are carried out in hot cells equipped with remote manipulators. The chemical processes are based on the use of extraction chromatography with hydrophobic SiO 2 impregnated with extracting agents. Plutonium targets and aged plutonium alloys are dissolved in nitric acid using conventional techniques while plutonium dioxide dissolutions are performed routine at 300 grams scale with electrogenerated silver II in 4M HNO 3 at room temperature. The separation between plutonium and americium is performed by extraction of Pu(IV) either on TBP/SiO 2 or TOAHNO 3 /SiO 2 column. Americium recovery from waste streams rid of plutonium is realized by chromatographic extraction of Am(III) using mainly TBP and episodically DHDECMP as extractant. The final purification of both americium isotopes uses the selective extraction of Am(VI) on HDDiBMP/SiO 2 column at 60 grams scale. Using the overall process a total amount of 1000 grams of americium 241 and 100 grams of americium 243 has been produced nowadays and the AmO 2 final product indicates a purity better than 98.5%

Cycle downstream: the plutonium question; Aval du cycle la question du plutonium

Energy Technology Data Exchange (ETDEWEB)

Zask, G [Electricite de France, EDF/DAC, 75 - Paris (France); Rome, M [Electricite de France, EDF, Service Etudes et Projets Thermiques et Nucleaires, 92 - Courbevoie (France); Delpech, M [CEA Cadarache, Dept. d' Etudes des Reacteurs/SPRC, 13 - Saint-Paul-lez-Durance (France); and others

1998-06-29

This day, organized by the SFEN, took place at Paris the 4 june 1998. Nine papers were presented. They take stock on the plutonium physics and its utilization as a nuclear fuel. This day tried to bring information to answer the following questions: do people have to keep the plutonium in the UOX fuel or in the MOX fuel in order to use it for future fast reactors? Do people have to continue obstinately the plutonium reprocessing in the MOX for the PWR type reactors? Will it be realized a underground disposal? Can it be technically developed plutonium incinerators and is it economically interesting? The plutonium physics, the experimental programs and the possible solutions are presented. (A.L.B.)

Economic considerations of plutonium utilization in the nuclear power strategy of Finland

International Nuclear Information System (INIS)

Silvennoinen, P.; Tusa, E.; Routti, J.T.

1977-01-01

Based on the current and prospected share of nuclear power in the national energy supply strategy an optimal programme is developed for exploitation of plutonium in both light water and fast reactor systems. Assuming cost trends until and beyond the year 2000 for uranium, plutonium, uranium enrichment, fuel fabricaton and assessing the availability of plutonium from the domestic power plants and from abroad the nuclear construction programme is optimized economically in view of the estimated development in the investment costs of various plant types. Given the expected nuclear share of the energy procurement this sector is covered by the alternative production schemes, i.e. light water reactors with and without plutonium recycle and fast reactors. The plant sizes are allowed to be either 500 MWe or 1000 MWe. The installation dates are fixed manually with a minor flexibility of time but with all the three degrees of freedom in the plant types. Defining the objective function in terms of minimized revenue requirement in plant amortization and operation the generated scenarios are screened off and they finally converge to the optimal policy of nuclear power construction up to the year 2000. Special attention is placed on the constraints which eliminate excessive proliferation of reactor types. This is mainly implemented by the criterion of increasing the domestic share in the investments. The established technology is associated with a larger share of the Finnish manufacturing and the introduction of new fuel or reactor type is taken to correspond to a reduced domestic investment share. The results yield the time schedule and installed capacity of the three different production means. Due to the uncertainties prevailing in the forecasts sensitivity studies are performed as functions of the major economic parameters and their temporal development

The handling of plutonium hexafluoride (1962); Manipulation de l'hexafluorure de plutonium (1962)

Energy Technology Data Exchange (ETDEWEB)

Berard, Ph [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1961-11-15

The major problem posed in this work is the instability of plutonium hexafluoride. The influence of various factors on the decomposition of the fluoride has been studied: physical aspect of the walls, nature of the metal and its pretreatment, influence of the temperature. A means of detecting plutonium-239 in a metallic apparatus by {gamma}-ray counting has been developed; the sensitivity is of the order of half a milligram, but the precision is very low (about 50 per cent). Yields of over 95 per cent have been obtained for the transfer of plutonium during the preparation and sublimation of the hexafluoride. This study confirms the possibility of using plutonium hexafluoride for the extraction of plutonium from irradiated fuel elements by a dry method. (author) [French] Le probleme majeur de cette etude a ete l'instabilite de l'hexafluorure de plutonium. Nous avons etudie l'influence de divers facteurs sur la decomposition de l'hexafluorure: aspect physique des parois, nature du metal et de son pretraitement, influence de la temperature. Nous avons mis au point un mode de detection du plutonium-239 dans un appareillage metallique par comptage du rayonnement {gamma}; la sensibilite est de l'ordre du demi-milligramme, mais la precision est tres faible (50 pour cent environ). Nous avons obtenu des rendements depassant 95 pour cent dans le transfert du plutonium au cours de la fabrication et de la sublimation de l'hexafluorure. Cette etude confirme la possibilite d'utiliser l'hexafluorure de plutonium dans l'extraction du plutonium des combustibles irradies par voie seche. (auteur)

METHOD OF SEPARATING URANIUM VALUES, PLUTONIUM VALUES AND FISSION PRODUCTS BY CHLORINATION

Science.gov (United States)

Brown, H.S.; Seaborg, G.T.

1959-02-24

The separation of plutonium and uranium from each other and from other substances is described. In general, the method comprises the steps of contacting the uranium with chlorine in the presence of a holdback material selected from the group consisting of lanthanum oxide and thorium oxide to form a uranium chloride higher than uranium tetrachloride, and thereafter heating the uranium chloride thus formed to a temperature at which the uranium chloride is volatilized off but below the volatilizalion temperature of plutonium chloride.

Determination of plutonium 241 in solutions of nuclear wastes

International Nuclear Information System (INIS)

Raymond, A.; Bilcot, J.B.; Poletiko, C.

1990-09-01

Determination of plutonium 241 in nuclear wastes is important because of long period and high energy of some daughter products. In this report are presented two quantitative analysis methods using both scintillation techniques: A complete method, in any case, by selective extraction of plutonium on an anionic resin allowing simultaneous determination of Pu 241 and the sum of other plutonium isotopes; a simplified method when alpha activity is higher than beta/gamma activity by liquid extraction with TTA. These methods are applied for analysis of 4 waste types: cement encapsulated wastes, bitumen encapsulated wastes, incineration ashes, leaching of encapsulated incineration ashes. In these 4 examples, Pu 241 activity is equal or higher than the sum of alpha plutonium isotope activity. Separation efficiency, measured from Pu 239 or with Pu 236 as tracer, is between 90 and 99% [fr

Plutonium re-cycle in HTR

Energy Technology Data Exchange (ETDEWEB)

Desoisa, J. A.

1974-03-15

The study of plutonium cycles in HTRs using reprocessed plutonium from Magnox and AGR fuel cycles has shown that full core plutonium/uranium loadings are in general not feasible, burn-up is limited due the need for lower loadings of plutonium to meet reload core reactivity limits, on-line refueling is not practicable due to the need for higher burnable poison loadings, and low conversion rates in the plutonium-uranium cycles cannot be mitigated by axial loading schemes so that fissile make-up is needed if HTR plutonium recycle is desired.

Concerning temporary method for return transport of plutonium

International Nuclear Information System (INIS)

1990-01-01

One of the urgent matters under deliberation in the Working Group for Nuclear Fuel Cycle, Atomic Energy Comission of Japan, is to develop a method for smooth and safe transport of plutonium to be returned from Britain and France, where spent fuel sent from Japan is reprocessed by contract. Return transport of plutonium from these nations will start in 1992. Preparatory work has been conducted in Japan for air transport of plutonium, which is to be used by Power Reactor and Nuclear Fuel Development Corporation. For this, efforts have been made since 1984 to develop containers for air transport of plutonium. Though the corporation will continue research, it will take considerable time until practical containers are developed, so the material have to be transported by sea for the time being. The corporation should play the leading role in conducting the return transport operation in cooperation with other organizations concerned including power companies. To achieve this, the corporation should be active in making preparations including the development of transport plans. (N.K.)

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

PLUTONIUM-ZIRCONIUM ALLOYS

Science.gov (United States)

Schonfeld, F.W.; Waber, J.T.

1960-08-30

A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

Extraction of hexavalent uranium, tetravalent plutonium and fission products by N, N'-tetraalkyldiamides

International Nuclear Information System (INIS)

Charbonnel, M.C.

1988-10-01

This study deals with the extractive properties of N, N'-tetraalkylglutaramides of generic formula R 2 NC(0)(CH 2 ) 3 C(0)NR 2 . These molecules were considered as alternative extractants to tributylphosphate in nuclear fuels reprocessing. They are selective extractants of uranium and plutonium as far as trivalent actinides and lanthanides remain in aqueous nitric solutions. Distribution ratios measurements and F.T. Infra-Red investigations show that HN0 3 extraction takes place via the formation of the following species: 2L.HN0 3 , L.HN0 3 and L.2HN0 3 in the organic phase (L: glutaramide). Distribution ratios of actinide ions followed by UV-visible spectroscopy and Infra-Red investigations agree with formation of the following neutral organometallic complexes in low nitric acidity conditions: L.U0 2 (N0 3 ) 2 and L.Pu(N0 3 ) 4 and the anionic species at higher acidities: L.U0 2 (N0 3 ) 3 H and L.Pu(N0 3 ) 6 H 2 . Interactions occur through neutral complexes and free molecules of diamides which explain the non ideality of the organic phase. Degradation products of these molecules don't seem to alter the extractive properties of these extractants towards uranium and plutonium [fr

Massive subcritical compact arrays of plutonium metal

Energy Technology Data Exchange (ETDEWEB)

Rothe, R.E.

1998-04-01

Two experimental critical-approach programs are reported. Both were performed at the Rocky Flats Plant near Denver, Colorado; and both date back to the late 1960s. Both involve very large arrays of massive plutonium ingots. These ingots had been cast in the foundry at the Rocky Flats Plant as part of their routine production operations; they were not specially prepared for either study. Consequently, considerable variation in ingot mass is encountered. This mass varied between approximately 7 kg and a little more than 10 kg. One program, performed in the spring of 1969, involved stacked arrays of ingots contained within cylindrical, disk-shaped, thin, steel cans. This program studied four arrays defined by the pattern of steel cans in a single layer. The four were: 1 x N, 3 x N, 2 x 2 x N, and 3 x 3 x N. The second was a tightly-packed, triangular-pitched patterns; the last two were square-pitched patterns. The other program, performed about a year earlier, involved similar ingots also contained in similar steel cans, but these canned plutonium ingots were placed in commercial steel drums. This study pertained to one-, two-, and three-layered horizontal arrays of drums. All cases proved to be well subcritical. Most would have remained subcritical had the parameters of the array under study been continued infinitely beyond the reciprocal multiplication safety limit. In one case for the drum arrays, an uncertain extrapolation of the data of the earlier program suggests that criticality might have eventually been attained had several thousand additional kilograms of plutonium been available for use.

Massive subcritical compact arrays of plutonium metal

International Nuclear Information System (INIS)

Rothe, R.E.

1998-01-01

Two experimental critical-approach programs are reported. Both were performed at the Rocky Flats Plant near Denver, Colorado; and both date back to the late 1960s. Both involve very large arrays of massive plutonium ingots. These ingots had been cast in the foundry at the Rocky Flats Plant as part of their routine production operations; they were not specially prepared for either study. Consequently, considerable variation in ingot mass is encountered. This mass varied between approximately 7 kg and a little more than 10 kg. One program, performed in the spring of 1969, involved stacked arrays of ingots contained within cylindrical, disk-shaped, thin, steel cans. This program studied four arrays defined by the pattern of steel cans in a single layer. The four were: 1 x N, 3 x N, 2 x 2 x N, and 3 x 3 x N. The second was a tightly-packed, triangular-pitched patterns; the last two were square-pitched patterns. The other program, performed about a year earlier, involved similar ingots also contained in similar steel cans, but these canned plutonium ingots were placed in commercial steel drums. This study pertained to one-, two-, and three-layered horizontal arrays of drums. All cases proved to be well subcritical. Most would have remained subcritical had the parameters of the array under study been continued infinitely beyond the reciprocal multiplication safety limit. In one case for the drum arrays, an uncertain extrapolation of the data of the earlier program suggests that criticality might have eventually been attained had several thousand additional kilograms of plutonium been available for use

Adaptation of the IBM ECR [electric cantilever robot] robot to plutonium processing applications

International Nuclear Information System (INIS)

Armantrout, G.A.; Pedrotti, L.R.; Halter, E.A.; Crossfield, M.

1990-12-01

The changing regulatory climate in the US is adding increasing incentive to reduce operator dose and TRU waste for DOE plutonium processing operations. To help achieve that goal the authors have begun adapting a small commercial overhead gantry robot, the IBM electric cantilever robot (ECR), to plutonium processing applications. Steps are being taken to harden this robot to withstand the dry, often abrasive, environment within a plutonium glove box and to protect the electronic components against alpha radiation. A mock-up processing system for the reduction of the oxide to a metal was prepared and successfully demonstrated. Design of a working prototype is now underway using the results of this mock-up study. 7 figs., 4 tabs

Optimization and plutonium equilibrium

International Nuclear Information System (INIS)

Silver, G.L.

1976-01-01

The sequential simplex method has been used to estimate the extent of disproportionation of tetravalent plutonium in dilute acid. A method for simulating potentiometric titrations is proposed, and this method suggests that the stoichiometric end point and the inflection point may not always correspond in the potentiometric titration of plutonium. A possible characteristic equation for the nitrite-plutonium reaction is illustrated, and the method of proportional equations is extended to the iron-plutonium reaction

Assesment of Plutonium 238 and Plutonium 239+240 in soils of different agricultural regions of Guatemala

International Nuclear Information System (INIS)

Gutierrez Martinez, E.A.

1998-02-01

In this report an assesment and measurement of PLUTONIUM 238, PLUTONIUM 239, and PLUTONIUM 240 are made. Samples of cultivated soils in 15 provinces of Guatemala were taken. To separate plutonium isotopes a radiochemical method was made using extraction, precipitation and ionic interchange. By electrodeposition the plutonium was measured using an alpha spectroscopy by PIPS method. The radioactivity ranges from 2.84 mBq/Kg to 36.38 mBq/Kg for plutonium 238, and 8.46 mBq/Kg to 26.61 mBq/Kg for plutonium 239+240

Evaluation of chloride-ion-specific electrodes as in situ chemical sensors for monitoring total chloride concentration in aqueous solutions generated during the recovery of plutonium from molten salts used in plutonium electrorefining operations

International Nuclear Information System (INIS)

Smith, W.H.

1992-10-01

Two commercially available chloride-ion-specific electrodes (CLISEs), a solid-state type and a membrane type, were evaluated as potential in situ chemical sensors for determining total chloride ion concentration in mixed sodium chloride/potassium chloride/hydrochloric acid solutions generated during the recovery of plutonium from molten salts used in plutonium electrorefining operations. Because the response of the solid-state CLISE was closer than was the response of the membrane-type CLISE to the theoretical response predicted by the Nernst equation, the solid-state CLISE was selected for further evaluation. A detailed investigation of the characteristics of the chloride system and the corresponding CLISE response to concentration changes suggested four methods by which the CLISE could be used either as a direct, in situ sensor or as an indirect sensor through which an analysis could be performed on-line with a sample extracted from the process solution

233-S Plutonium Concentration Facility data quality objectives

International Nuclear Information System (INIS)

Encke, D.B.

1996-08-01

This document is a summary of the decision-making associated with the Data Quality Objective process that pertains to the characterization activities in the 233-S Plutonium Concentration Facility at the Hanford Site in Richland, Washington. The 233-S Plutonium Concentration Facility is located adjacent to, and north of, the REDOX Plant. The facility was used to concentrate the plutonium nitrate product solution from the REDOX facility. The 233-S Pipe Gallery, Control Room, SWP Change Room, Toilet, Equipment Room and the Electrical Cubicle are currently scheduled for decontamination and cleanout to support future demolition (D and D). Identification of the radiological contamination and presence of hazardous materials is needed to allow for disposal of the D and D debris

Plutonium roundtable discussion

International Nuclear Information System (INIS)

Penneman, R.A.

1982-01-01

The roundtable discussion began with remarks by the chairman who pointed out the complicated nature of plutonium chemistry. Judging from the papers presented at this symposium, he noticed a pattern which indicated to him the result of diminished funding for investigation of basic plutonium chemistry and funding focused on certain problem areas. Dr. G.L. silver pointed to plutonium chemists' erroneous use of a simplified summary equation involving the disproportionation of Pu(EV) and their each of appreciation of alpha coefficients. To his appreciation of alpha coefficients. To his charges, Dr. J.T. Bell spoke in defense of the chemists. This discussion was followed by W.W. Schulz's comments on the need for experimental work to determine solubility data for plutonium in its various oxidation states under geologic repository conditions. Discussion then turned to plutonium pyrachemical process with Dana C. Christensen as the main speaker. This paper presents edited versions of participants' written version

Equipping a glovebox for waste form testing and characterization of plutonium bearing materials

International Nuclear Information System (INIS)

Noy, M.; Johnson, S.G.; Moschetti, T.L.

1997-01-01

The recent decision by the Department of Energy to pursue a hybrid option for the disposition of weapons plutonium has created the need for additional facilities that can examine and characterize waste forms that contain Pu. This hybrid option consists of the placement of plutonium into stable waste forms and also into mixed oxide fuel for commercial reactors. Glass and glass-ceramic waste forms have a long history of being effective hosts for containing radionuclides, including plutonium. The types of tests necessary to characterize the performance of candidate waste forms include: static leaching experiments on both monolithic and crushed waste forms, microscopic examination, and density determination. Frequently, the respective candidate waste forms must first be produced using elevated temperatures and/or high pressures. The desired operations in the glovebox include, but are not limited to the following: (1) production of vitrified/sintered samples, (2) sampling of glass from crucibles or other vessels, (3) preparing samples for microscopic inspection and monolithic and crushed static leach tests, and (4) performing and analyzing leach tests in situ. This paper will describe the essential equipment and modifications that are necessary to successfully accomplish the goal of outfitting a glovebox for these functions

Preliminary safety evaluation for the plutonium stabilization and packaging system

International Nuclear Information System (INIS)

Shapley, J.E.

1997-01-01

This Preliminary Safety Evaluation (PSE) describes and analyzes the installation and operation of the Plutonium Stabilization and Packaging System (SPS) at the Plutonium Finishing Plant (PFP). The SPS is a combination of components required to expedite the safe and timely storage of Plutonium (Pu) oxide. The SPS program will receive site Pu packages, process the Pu for storage, package the Pu into metallic containers, and safely store the containers in a specially modified storage vault. The location of the SPS will be in the 2736- ZB building and the storage vaults will be in the 2736-Z building of the PFP, as shown in Figure 1-1. The SPS will produce storage canisters that are larger than those currently used for Pu storage at the PFP. Therefore, the existing storage areas within the PFP secure vaults will require modification. Other modifications will be performed on the 2736-ZB building complex to facilitate the installation and operation of the SPS

A Plutonium Ceramic Target for MASHA

International Nuclear Information System (INIS)

Wilk, P A; Shaughnessy, D A; Moody, K J; Kenneally, J M; Wild, J F; Stoyer, M A; Patin, J B; Lougheed, R W; Ebbinghaus, B B; Landingham, R L; Oganessian, Y T; Yeremin, A V; Dmitriev, S N

2004-01-01

We are currently developing a plutonium ceramic target for the MASHA mass separator. The MASHA separator will use a thick plutonium ceramic target capable of tolerating temperatures up to 2000 C. Promising candidates for the target include oxides and carbides, although more research into their thermodynamic properties will be required. Reaction products will diffuse out of the target into an ion source, where they will then be transported through the separator to a position-sensitive focal-plane detector array. Experiments on MASHA will allow us to make measurements that will cement our identification of element 114 and provide for future experiments where the chemical properties of the heaviest elements are studied

Investigation of plutonium (4) hydroxoformates

International Nuclear Information System (INIS)

Andryushin, V.G.; Belov, V.A.; Galaktionov, S.V.; Kozhevnikov, P.B.; Matyukha, V.A.; Shmidt, V.S.

1982-01-01

Deposition processes of plutonium (4) hydroxoformates in the system Pu(NO 3 ) 4 -HNO 3 -HCoOH-N6 4 OH-H 2 O have been studied in pH range 0.2-10.7 at total plutonium concentration in the system 100 g/l. It is shown that under the conditions plutonium (4) hydrolysis takes place with the formation of hydroxoformates. A local maximum of plutonium (4) hydroxoformate solubility in the range pH=3.8-4.8, which is evidently conditioned by the formation of soluble formate complex of plutonium in the region, is pointed out. The basic plutonium (4) formates of the composition PuOsub(x)(OH)sub(y)(COOH)sub(4-2x-y)xnHsub(2)O, where 1,3 >=x >= 0.7, 1.7 >= y >= 1.0 and n=1.5-7.0, are singled out, their thermal stability being studied. Density of the crystals and plutonium dioxide, formed during their thermal decomposition, is measured. It is established that for plutonium (4) hydroxoformates common regularities of the influence of salt composition (OH - -, CHOO - - and H 2 O-group numbers in the mulecule) on position of temperature decomposition effects and on the density of compounds, which have been previously found during the study of thorium and plutonium hydroxosalts are observed. It is shown that the density of plutonium dioxide decreases with the increase of hydration and hydrolysis degree of the initial plutonium hydroxoformate

Prospects for the establishment of plutonium recycle in thermal reactors in the Foratom countries. Status and assessment

International Nuclear Information System (INIS)

Chamberlain, A.; Melches, C.

1977-01-01

The paper reviews the technical status of plutonium recycle in thermal reactors in the Foratom countries and assesses the prospect for it becoming established in the future with the implicit assumptions that uranium oxide reprocessing capacity will be installed commensurate with the projected programmes for thermal reactor installation and that there will be no insuperable environmental, security or safeguards obstacles to the use of plutonium as a fuel. It is argued that the feasibility of using plutonium as an alternative to 235 U as the fuel for thermal reactors, particularly LWRs, has been extensively demonstrated by a number of Foratom countries and the main problem areas are fuel fabrication and fuel reprocessing. Mixed-oxide fuel fabrication has been well established on the prototype plant scale using low-irradiation plutonium, but it is recognized that the future design of production-scale plants will need to cater for the significantly higher radiation levels from high burnup plutonium and meet stricter environmental requirements on operator dosage and waste arisings. The main constraint on the establishment of recycle up to now has been the lack of available plutonium owing to the absence of significant uranium-oxide fuel reprocessing capacity. An assessment of the plutonium arisings in Europe, based on the projected uranium-oxide reprocessing capacity, shows that by 1990 plutonium, surplus to FBR requirements, should be accumulating by about 10t/a, sufficient to fuel about 8000MW(e) of LWRs. A further constraint would then be the availability and technical problems of mixed-oxide reprocessing, which is one of the areas identified for international collaboration. It is concluded that whilst there is unlikely to be substantial recycle of plutonium in thermal reactors in the Foratom countries before the early 1990s, an incentive could possibly arise about that time. The strength of this incentive will depend on a number of factors including the status of

Diffusion in the uranium - plutonium system and self-diffusion of plutonium in epsilon phase; Diffusion dans le systeme uranium-plutonium et autodiffusion du plutonium epsilon

Energy Technology Data Exchange (ETDEWEB)

Dupuy, M [Commissariat a l' Energie Atomique, Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

1967-07-01

A survey of uranium-plutonium phase diagram leads to confirm anglo-saxon results about the plutonium solubility in {alpha} uranium (15 per cent at 565 C) and the uranium one in {zeta} phase (74 per cent at 565 C). Interdiffusion coefficients, for concentration lower than 15 per cent had been determined in a temperature range from 410 C to 640 C. They vary between 0.2 and 6 10{sup 12} cm{sup 2} s{sup -1}, and the activation energy between 13 and 20 kcal/mole. Grain boundary, diffusion of plutonium in a uranium had been pointed out by micrography, X-ray microanalysis and {alpha} autoradiography. Self-diffusion of plutonium in {epsilon} phase (bcc) obeys Arrhenius law: D = 2. 10{sup -2} exp -(18500)/RT. But this activation energy does not follow empirical laws generally accepted for other metals. It has analogies with 'anomalous' bcc metals ({beta}Zr, {beta}Ti, {beta}Hf, U{sub {gamma}}). (author) [French] Une etude du diagramme d'equilibre uranium-plutonium conduit a confirmer les resultats anglo-saxons relatifs a la solubilite du plutonium dans l'uranium {alpha} (15 pour cent a 565 C) et de l'uranium dans la phase {zeta} (74 pour cent a 565 C). Les coefficients de diffusion chimique, pour des concentrations inferieures a 15 pour cent ont ete determines a des temperatures comprises entre 410 et 640 C. Ils se situent entre 0.2 et 6. 10{sup 12} cm{sup 2} s{sup -1}. L'energie d'activation varie entre 13 et 20 kcal/mole. La diffusion intergranulaire du plutonium dans l'uranium a a ete mise en evidence par micrographie, microanalyse X et autoradiographie {alpha}. L' autodiffusion du plutonium {beta} cubique centree obeit a la loi d'Arrhenius D = 2. 10{sup -2} exp - (18500)/RT. Son energie d'activation n'obeit pas aux lois empiriques generalement admises pour les autres metaux. Elle possede des analogies avec les cubiques centres ''anormaux'' (Zr{beta}, Ti{beta}, Hf{beta}, U{gamma}). (auteur)

Sludge stabilization at the Plutonium Finishing Plant, Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1994-10-01

This Environmental Assessment evaluates the proposed action to operate two laboratory-size muffle furnaces in glovebox HC-21C, located in the Plutonium Finishing Plant (PFP), Hanford Site, Richland, Washington. The muffle furnaces would be used to stabilize chemically reactive sludges that contain approximately 25 kilograms (55 pounds) of plutonium by heating to approximately 500 to 1000 degrees C (900 to 1800 degrees F). The resulting stable powder, mostly plutonium oxide with impurities, would be stored in the PFP vaults. The presence of chemically reactive plutonium-bearing sludges in the process gloveboxes poses a risk to workers from radiation exposure and limits the availability of storage space for future plant cleanup. Therefore, there is a need to stabilize the material into a form suitable for long-term storage. This proposed action would be an interim action, which would take place prior to completion of an Environmental Impact Statement for the PFP which would evaluate stabilization of all plutonium-bearing materials and cleanout of the facility. However, only 10 percent of the total quantity of plutonium in reactive materials is in the sludges, so this action will not limit the choice of reasonable alternatives or prejudice the Record of Decision of the Plutonium Finishing Plant Environmental Impact Statement

Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

International Nuclear Information System (INIS)

1996-09-01

Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU's. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work

Crystalline ceramics: Waste forms for the disposal of weapons plutonium

International Nuclear Information System (INIS)

Ewing, R.C.; Lutze, W.; Weber, W.J.

1995-05-01

At present, there are three seriously considered options for the disposition of excess weapons plutonium: (i) incorporation, partial burn-up and direct disposal of MOX-fuel; (ii) vitrification with defense waste and disposal as glass ''logs''; (iii) deep borehole disposal (National Academy of Sciences Report, 1994). The first two options provide a safeguard due to the high activity of fission products in the irradiated fuel and the defense waste. The latter option has only been examined in a preliminary manner, and the exact form of the plutonium has not been identified. In this paper, we review the potential for the immobilization of plutonium in highly durable crystalline ceramics apatite, pyrochlore, monazite and zircon. Based on available data, we propose zircon as the preferred crystalline ceramic for the permanent disposition of excess weapons plutonium

A method for the gravimetric determination of plutonium in pure plutonium nitrate concentrate solution

International Nuclear Information System (INIS)

Mair, M.A.; Savage, D.J.

1986-12-01

Plutonium nitrate solution is treated with sulphuric acid before being heated and finally ignited. The stoichiometric plutonium dioxide so formed is weighed and hence the plutonium content is calculated. (author)

F. Biological hazards of plutonium

International Nuclear Information System (INIS)

1976-01-01

Plutonium is an unavoidable result of present nuclear power programmes. Its predominant isotopes are extremely long-lived and very toxic if absorbed in the body. In view of the increasing potential for plutonium and man to come into contact, the consequences of any plutonium release into the environment should be scientifically examined. This report is an attempt to place in one document a fully referenced account of the on-going work in many areas. There are three sections. Part 1 deals with the amounts of plutonium available in the fuel cycle, its properties and the probability of routine or accidental release. Part 2 examines the ways in which plutonium can reach man, in particular through food chains or inhalation. Part 3 details the biological effects of plutonium once it is absorbed into the body, assesses the amounts likely to be harmful and discusses the adequacy of present standards for plutonium burdens. There are two appendices. The likely differences between Pu-239, the most studied isotope, and other plutonium isotopes or transuranic nuclides are outlined in Appendix A. Appendix B contains a fuller account of the ways in which the fate of ingested or inhaled plutonium have been determined

Study of the Effect of (U0.8Pu0.2O2 Uranium–Plutonium Mixed Fuel Fission Products on a Living Organism

Directory of Open Access Journals (Sweden)

Ayagoz Baimukhanova

2016-08-01

Full Text Available The article describes the results of experiments conducted on pigs to determine the effect of plutonium, which is the most radiotoxic and highly active element in the range of mixed fuel (U0.8Pu0.2O2 fission products, on living organisms. The results will allow empirical prediction of the emergency plutonium radiation dose for various organs and tissues of humans in case of an accident in a reactor running on mixed fuel (U0.8Pu0.2O2.

Thorium utilization as a Pu-burner: proposal of Plutonium-Thorium Mixed Oxide (PT-MOX) Project

International Nuclear Information System (INIS)

Aizawa, Otohiko

2000-01-01

In this paper, a Pu-Th mixed oxide (PT-MOX) project is proposed for a thorium utilization and a plutonium burning. None of plutonium can be newly produced from PT-MOX fuel, and the plutonium mass of about 1 ton can be consumed with one reactor (total heavy metal assumed: 100 tons) for 1 year. In order to consume plutonium produced from usual Light Water Reactor, it should be better to operate one PT-MOX reactor for three to five Light Water Reactors. (author)

A perspective on the proliferation risks of plutonium mines

Energy Technology Data Exchange (ETDEWEB)

Lyman, E.S. [Nuclear Control Institute, Washington, DC (United States)

1996-05-01

The program of geologic disposal of spent fuel and other plutonium-containing materials is increasingly becoming the target of criticism by individuals who argue that in the future, repositories may become low-cost sources of fissile material for nuclear weapons. This paper attempts to outline a consistent framework for analyzing the proliferation risks of these so-called {open_quotes}plutonium mines{close_quotes} and putting them into perspective. First, it is emphasized that the attractiveness of plutonium in a repository as a source of weapons material depends on its accessibility relative to other sources of fissile material. Then, the notion of a {open_quotes}material production standard{close_quotes} (MPS) is proposed: namely, that the proliferation risks posed by geologic disposal will be acceptable if one can demonstrate, under a number of reasonable scenarios, that the recovery of plutonium from a repository is likely to be as difficult as new production of fissile material. A preliminary analysis suggests that the range of circumstances under which current mined repository concepts would fail to meet this standard is fairly narrow. Nevertheless, a broad application of the MPS may impose severe restrictions on repository design. In this context, the relationship of repository design parameters to easy of recovery is discussed.

A perspective on the proliferation risks of plutonium mines

International Nuclear Information System (INIS)

Lyman, E.S.

1996-01-01

The program of geologic disposal of spent fuel and other plutonium-containing materials is increasingly becoming the target of criticism by individuals who argue that in the future, repositories may become low-cost sources of fissile material for nuclear weapons. This paper attempts to outline a consistent framework for analyzing the proliferation risks of these so-called open-quotes plutonium minesclose quotes and putting them into perspective. First, it is emphasized that the attractiveness of plutonium in a repository as a source of weapons material depends on its accessibility relative to other sources of fissile material. Then, the notion of a open-quotes material production standardclose quotes (MPS) is proposed: namely, that the proliferation risks posed by geologic disposal will be acceptable if one can demonstrate, under a number of reasonable scenarios, that the recovery of plutonium from a repository is likely to be as difficult as new production of fissile material. A preliminary analysis suggests that the range of circumstances under which current mined repository concepts would fail to meet this standard is fairly narrow. Nevertheless, a broad application of the MPS may impose severe restrictions on repository design. In this context, the relationship of repository design parameters to easy of recovery is discussed

Radionuclide characterization of graphite stacks from plutonium production reactors of the Siberian group of chemical enterprises

International Nuclear Information System (INIS)

Bushuev, A.V.; Verzilov, Yu.M.; Zubarev, V.N.

2001-01-01

The residual radionuclide concentrations and distributions in graphite from moderator stack of plutonium production reactors at Tomsk-7 have been investigated. It was found that the dominant activity of graphite is 14 C. To gain information on surface and volume contamination of graphite blocks from the moderator stack, the special sets of samples were collected and assayed. The schemes are proposed for evaluation of individual radionuclide inventories together with results of the evaluations performed. (author)

Some aspects of a technology of processing weapons grade plutonium to nuclear fuel

International Nuclear Information System (INIS)

Bibilashvili, Y.; Glagovsky, E.M.; Zakharkin, B.S.; Orlov, V.K.; Reshetnikov, F.G.; Rogozkin, B.G.; Soloni-N, M.I.

2000-01-01

The concept by Russia to use fissile weapons-grade materials, which are being recovered from nuclear pits in the process of disarmament, is based on an assessment of weapons-grade plutonium as an important energy source intended for use in nuclear power plants. However, in the path of involving plutonium excessive from the purposes of national safety into industrial power engineering there are a lot of problems, from which effectiveness and terms of its disposition are being dependent upon. Those problems have political, economical, financial and environmental character. This report outlines several technology problems of processing weapons-grade metallic plutonium into MOX-fuel for reactors based on thermal and fast neutrons, in particular, the issue of conversion of the metal into dioxide from the viewpoint of fabrication of pelletized MOX-fuel. The processing of metallic weapons-grade plutonium into nuclear fuel is a rather complicated and multi-stage process, every stage of which is its own production. Some of the stages are absent in production of MOX-fuel, for instance the stage of the conversion, i.e. transferring of metallic plutonium into dioxide of the ceramic quality. At this stage of plutonium utilization some tasks must be resolved as follows: I. As a result of the conversion, a material purified from ballast and radiogenic admixtures has to be obtained. This one will be applied to fabricate pelletized MOX-fuel going from morphological, physico-mechanical and technological properties. II. It is well known that metallic gallium, which is used as an alloying addition in weapons-grade plutonium, actively reacts with multiple metals. Therefore, an important issue is to study the effect of gallium on the technology of MOX-fuel production, quality of the pellets, as well as the interaction of gallium oxide with zirconium and steel shells of fuel elements depending upon the content of gallium in the fuel. The rate of the interaction of gallium oxide

World status report: plutonium

International Nuclear Information System (INIS)

Dircks, W.

1992-01-01

In a recent speech in Japan, the Deputy Director General of the International Atomic Energy Agency (IAEA) said that the economic case for reprocessing spent nuclear fuel had been severely eroded. An edited version of the speech is given. The changed prospects for nuclear energy is given as the reason why the demand for plutonium has declined sharply. The oil crisis of the 1970s reduced the demand for electric power and the economic justification for the use of recycled plutonium. The stockpile of isolated plutonium is growing rapidly giving rise to worries about its security. From this point of view, isolated plutonium is best kept in reactor fuel not separated out. In this connection the IAEA has offered to help in the storage of plutonium so that vigorous safety and security requirements are met. In Japan there is a debate about the plutonium which is dependent on the future of the fast breeder reactor programme. (UK)

PLUTONIUM PURIFICATION PROCESS EMPLOYING THORIUM PYROPHOSPHATE CARRIER

Science.gov (United States)

King, E.L.

1959-04-28

The separation and purification of plutonium from the radioactive elements of lower atomic weight is described. The process of this invention comprises forming a 0.5 to 2 M aqueous acidffc solution containing plutonium fons in the tetravalent state and elements with which it is normally contaminated in neutron irradiated uranium, treating the solution with a double thorium compound and a soluble pyrophosphate compound (Na/sub 4/P/sub 2/O/sub 7/) whereby a carrier precipitate of thorium A method is presented of reducing neptunium and - trite is advantageous since it destroys any hydrazine f so that they can be removed from solutions in which they are contained is described. In the carrier precipitation process for the separation of plutonium from uranium and fission products including zirconium and columbium, the precipitated blsmuth phosphate carries some zirconium, columbium, and uranium impurities. According to the invention such impurities can be complexed and removed by dissolving the contaminated carrier precipitate in 10M nitric acid, followed by addition of fluosilicic acid to about 1M, diluting the solution to about 1M in nitric acid, and then adding phosphoric acid to re-precipitate bismuth phosphate carrying plutonium.

Development of the plutonium oxide vitrification system

International Nuclear Information System (INIS)

Marshall, K.M.; Marra, J.C.; Coughlin, J.T.; Calloway, T.B.; Schumacher, R.F.; Zamecnik, J.R.; Pareizs, J.M.

1998-01-01

Repository disposal of plutonium in a suitable, immobilized form is being considered as one option for the disposition of surplus weapons-usable plutonium. Accelerated development efforts were completed in 1997 on two potential immobilization forms to facilitate downselection to one form for continued development. The two forms studied were a crystalline ceramic based on Synroc technology and a lanthanide borosilicate (LaBS) glass. As part of the glass development program, melter design activities and component testing were completed to demonstrate the feasibility of using glass as an immobilization medium. A prototypical melter was designed and built in 1997. The melter vessel and drain tube were constructed of a Pt/Rh alloy. Separate induction systems were used to heat the vessel and drain tube. A Pt/Rh stirrer was incorporated into the design to facilitate homogenization of the melt. Integrated powder feeding and off-gas systems completed the overall design. Concurrent with the design efforts, testing was conducted using a plutonium surrogate LaBS composition in an existing (near-scale) melter to demonstrate the feasibility of processing the LaBS glass on a production scale. Additionally, the drain tube configuration was successfully tested using a plutonium surrogate LaBS glass

Fuel Cycle Impacts of Uranium-Plutonium Co-extraction

International Nuclear Information System (INIS)

Taiwo, Temitope; Szakaly, Frank; Kim, Taek-Kyum; Hill, Robert

2008-01-01

A systematic investigation of the impacts of uranium and plutonium co-extraction during fuel separations on reactor performance and fuel cycle has been performed. Proliferation indicators, critical mass and radiation source levels of the separation products or fabricated fuel, were also evaluated. Using LWR-spent-uranium-based MOX fuel instead of natural-uranium-based fuel in a PWR MOX core requires a higher initial plutonium content (∼1%), and results in higher Np-237 content (factor of 5) in the spent fuel, and less consumption of Pu-238 (20%) and Am-241 (14%), indicating a reduction in the effective repository space utilization. Additionally, minor actinides continue to accumulate in the fuel cycle, and thus a separate solution is required for them. Differences were found to be quite smaller (∼0.4% in initial transuranics) between the equilibrium cycles of advanced fast reactor cores using spent and depleted uranium for make-up, in additional to transuranics. The critical masses of the co-extraction products were found to be higher than for weapons-grade plutonium (WG-Pu) and the decay heat and radiation sources of the materials (products) were also found to be generally higher than for WG-Pu in the transuranics content range of 10% to 100% in the heavy-metal. (authors)

Liquid waste treatment at plutonium fuels fabrication facility, 2

International Nuclear Information System (INIS)

Matsumoto, Ken-ichi; Itoh, Ichiroh; Ohuchi, Jin; Miyo, Hiroaki

1974-01-01

The economics in the management of the radioactive liquid waste from Plutonium Fuels Fabrication Facility with sludge-blanket type flocculators has been evaluated. (1) Cost calculation: The cost of chemicals and electricity to treat 1 cubic meter of liquid waste is about 876 yen, while the total operating cost is 250 thousand yen per cubic meter in the case of 140 m 3 /year treatment. These figures are much higher than those for ordinary wastes, due to the particular operation against plutonium. (2) Proposal of the closed system for liquid waste treatment at PFFF: In the case of a closed system using evaporator, ion exchange column and rotary-kiln calciner, the operating cost is estimated at 40 thousand yen per cubic meter of liquid waste. Final radioactivity of treated liquid is below 10 -8 micro curies/ml. (Mori, K.)

The export of weapons grade plutonium to the USA

International Nuclear Information System (INIS)

Kollerstrom, N.

1986-01-01

Reprocessed spent Magnox fuel from British nuclear power plants has led, it is claimed, to the production of plutonium, some of weapons grade. Some of this has been exported to the USA where, it is assumed, it is used for military purposes. The route and agreements which make this possible and the quantities involved are reported. Inspection by IAEA is insufficient to check the Central Electricity Generating Board's (CEGB) claim that no CEGB plutonium has been used for a military purpose. The CEGB case, presented at the Sizewell Inquiry is discussed. In the United States it is not clear whether plutonium from Britain, at present in a civil stockpile, will be transferred to military use or not. (U.K.)

Fluid bed direct denitration process for plutonium nitrate to oxide conversion

International Nuclear Information System (INIS)

Souply, K.R.; Neal, D.H.

1977-01-01

The fluid bed direct-denitration process appears feasible for reprocessing Light Water Reactor fuel. Considerable experience with the fluid bed process exists in the denitration of uranyl nitrate and it shows promise for use in the denitration of plutonium nitrate. The process will require some development work before it can be used in a production-size facility. This report describes a fluid bed direct-denitration process for converting plutonium nitrate to plutonium oxide, and the information should be used when making comparisons of alternative processes or as a basis for further detailed studies

Cation exchange process for recovery of plutonium from laboratory solutions containing chloride

International Nuclear Information System (INIS)

Gray, L.W.

1978-10-01

A cation exchange technique was developed for the separation of plutonium from laboratory solutions containing either Pu(III) or Pu(III)--Pu(IV) mixtures in acidic solutions containing chloride ions. The procedure consists of adjusting the acid concentration to less than one molar and adjusting the valence of the plutonium ion to the (III) state, if necessary. The adjusted solution is fed to a cation exchange column and washed with distilled water to remove residual chlorides from the column. Plutonium is then eluted from the column with 5M nitric acid containing 0.34M sulfamic acid. This procedure was used to separate plutonium from 1.2M chloride solution on a production-scale column. Typical plutonium recovery was 99.97%, while greater than 96% of the original chloride was rejected

Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.

1977-01-01

The increased cost of uranium and separative work has increased the attractiveness of plutonium use in both uranium and thorium fuel cycles in thermal reactors. A technology, fuel utilization, and economic evaluation is given for uranium and thorium fuel cycles in various reactor types, along with the use of plutonium and 238 U. Reactors considered are LWRs, HWRs, LWBRs, HTGRs, and FBRs. Key technology factors are fuel irradiation performance and associated physical property values. Key economic factors are unit costs for fuel fabrication and reprocessing, and for refabrication of recycle fuels; consistent cost estimates are utilized. In thermal reactors, the irradiation performance of ceramic fuels appears to be satisfactory. At present costs for uranium ore and separative work, recycle of plutonium with thorium rather than uranium is preferable from fuel utilization and economic viewpoints. Further, the unit recovery cost of plutonium is lower from LWR fuels than from natural-uranium HWR fuels; use of LWR product permits plutonium/thorium fueling to compete with uranium cycles. Converting uranium cycles to thorium cycles increases the energy which can be extracted from a given uranium resource. Thus, additional fuel utilization improvement can be obtained by fueling all thermal reactors with thorium, but this requires use of highly enriched uranium; use of 235 U with thorium is most economic in HTGRs followed by HWRs and then LWRs. Marked improvement in long-term fuel utilization can be obtained through high thorium loadings and short fuel cycle irradiations as in the LWBR, but this imposes significant economic penalties. Similar operating modes are possible in HWRs and HTGRs. In fast reactors, use of the plutonium-uranium cycle gives advantageous fuel resource utilization in both LMFBRs and GCFRs; use of the thorium cycle provides more negative core reactivity coefficients and more flexibility relative to use of recycle fuels containing uranium of less than 20

Plutonium ocean shipment safety between Europe and Japan

International Nuclear Information System (INIS)

Pierce, J.D.; Hohnstreiter, G.F.; McClure, J.D.; Smith, J.D.; Dukart, R.J.; Koski, J.A.; Braithwate, J.W.; Sorenson, N.R.; Yamamoto, K.; Kitamura, T.; Shibata, K.; Ouchi, Y.; Ito, T.

2004-01-01

Sandia National Laboratories (SNL) and Japan Nuclear Cycle Development Institute (JNC) have conducted an extensive emergency response planning study of the safety of the sea transport of plutonium for JNC. This study was conducted in response to international concerns about the safety of the marine transport of PuO2 powder that began with the sea transport of plutonium powder from France to Japan in 1992 using a purposebuilt ship. This emergency response planning study addressed four topics to better define the accident environment for long-range sea transport of nuclear materials. The first topic is a probabilistic safety analysis that evaluates the technical issues of transporting plutonium between Europe and Japan. An engine-room fire aboard a purposebuilt ship is evaluated as the second topic to determine the vulnerability and safety margin of radioactive material packaging for plutonium designed to meet International Atomic Energy Agency (IAEA) standards. The third topic is a corrosion study performed for generic plutonium packaging to estimate the time required to breach the containment boundary in the event of submersion in seawater. The final study topic is a worldwide survey of information on high-value cargo salvage capabilities from sunken ships. The primary purpose of this overall emergency response planning study is to describe and analyze the safety of radioactive material transportation operations for the international transportation of radioactive materials by maritime cargo vessels

Plutonium ocean shipment safety between Europe and Japan

Energy Technology Data Exchange (ETDEWEB)

Pierce, J D; Hohnstreiter, G F; McClure, J D; Smith, J D; Dukart, R J; Koski, J A; Braithwate, J W; Sorenson, N R [Sandia National Labs., Albuquerque, NM (United States); Yamamoto, K; Kitamura, T; Shibata, K; Ouchi, Y; Ito, T [Japan Nuclear Cycle Development Inst., Tokai-mura (Japan)

2004-07-01

Sandia National Laboratories (SNL) and Japan Nuclear Cycle Development Institute (JNC) have conducted an extensive emergency response planning study of the safety of the sea transport of plutonium for JNC. This study was conducted in response to international concerns about the safety of the marine transport of PuO2 powder that began with the sea transport of plutonium powder from France to Japan in 1992 using a purposebuilt ship. This emergency response planning study addressed four topics to better define the accident environment for long-range sea transport of nuclear materials. The first topic is a probabilistic safety analysis that evaluates the technical issues of transporting plutonium between Europe and Japan. An engine-room fire aboard a purposebuilt ship is evaluated as the second topic to determine the vulnerability and safety margin of radioactive material packaging for plutonium designed to meet International Atomic Energy Agency (IAEA) standards. The third topic is a corrosion study performed for generic plutonium packaging to estimate the time required to breach the containment boundary in the event of submersion in seawater. The final study topic is a worldwide survey of information on high-value cargo salvage capabilities from sunken ships. The primary purpose of this overall emergency response planning study is to describe and analyze the safety of radioactive material transportation operations for the international transportation of radioactive materials by maritime cargo vessels.

Evaluation of the Magnesium Hydroxide Treatment Process for Stabilizing PFP Plutonium/Nitric Acid Solutions

Energy Technology Data Exchange (ETDEWEB)

Gerber, Mark A.; Schmidt, Andrew J.; Delegard, Calvin H.; Silvers, Kurt L.; Baker, Aaron B.; Gano, Susan R.; Thornton, Brenda M.

2000-09-28

This document summarizes an evaluation of the magnesium hydroxide [Mg(OH)2] process to be used at the Hanford Plutonium Finishing Plant (PFP) for stabilizing plutonium/nitric acid solutions to meet the goal of stabilizing the plutonium in an oxide form suitable for storage under DOE-STD-3013-99. During the treatment process, nitric acid solutions bearing plutonium nitrate are neutralized with Mg(OH)2 in an air sparge reactor. The resulting slurry, containing plutonium hydroxide, is filtered and calcined. The process evaluation included a literature review and extensive laboratory- and bench-scale testing. The testing was conducted using cerium as a surrogate for plutonium to identify and quantify the effects of key processing variables on processing time (primarily neutralization and filtration time) and calcined product properties.

The Plutonium Fuel Laboratory at Studsvik and Its Activities

Energy Technology Data Exchange (ETDEWEB)

Hultgren, A.; Berggren, G.; Brown, A.; Eng, H. U.; Forsyth, R. S. [AB Atomenergi, Studsvik (Sweden)

1967-09-15

The plutonium fuel laboratory at Studsvik is engaged in development work on plutonium-enriched fuel. At present, low enriched fuel for thermal reactors is being studied: work on fuel with a higher plutonium content for fast reactors is foreseen at a later date. So far only the pellet technique is under consideration, and a number of pellet rod specimens will be produced and irradiated in the reactor R2. These specimens include pellets from both co-precipitated uranium-plutonium salts and from physically mixed oxides. Comparison of these two materials will be extended to different density levels and different heat ratings. The methods and techniques used and studied include wet chemical work for powder preparation (continuous precipitation of Pu(IV)-oxalate with oxalic acid, continuous co-precipitation of plutonium and uranium with ammonia, optimization of.precipitation conditions using U(IV) and U(VI) respectively) ; powder preparation (drying, calcination, reduction, mixing, milling, binder addition, granulation); pellet preparation (pressing, debonding, sintering, inspection): encapsulation (charging, welding of end plug, helium filling, end sealing by welding, leak detection, decontamination); metallography (specimen preparation (moulding, polishing), etching, microscopy); structure investigations (thermal analysis (TG, DTA), X-ray diffraction, neutron diffraction, data handling by computer analysis); radiometric methods (direct plutonium determination by gamma spectrometry, non-destructive burn-up analysis by high resolution gamma spectrometry, using a Ge(Li) detector) ; rework of waste (recovery of plutonium from fuel waste by extraction with trilauryl amine and anion exchange). The plutonium fuel laboratory forms part of the Active Central Laboratory. The equipment is contained in four adjacent 10 x 15 m rooms; .for diffraction work and inactive uranium work additional space is available. All the forty glove boxes in operation except two are of AB Atomenergi

Solubility of Plutonium (IV) Oxalate During Americium/Curium Pretreatment

International Nuclear Information System (INIS)

Rudisill, T.S.

1999-01-01

Approximately 15,000 L of solution containing isotopes of americium and curium (Am/Cm) will undergo stabilization by vitrification at the Savannah River Site (SRS). Prior to vitrification, an in-tank pretreatment will be used to remove metal impurities from the solution using an oxalate precipitation process. Material balance calculations for this process, based on solubility data in pure nitric acid, predict approximately 80 percent of the plutonium in the solution will be lost to waste. Due to the uncertainty associated with the plutonium losses during processing, solubility experiments were performed to measure the recovery of plutonium during pretreatment and a subsequent precipitation process to prepare a slurry feed for a batch melter. A good estimate of the plutonium content of the glass is required for planning the shipment of the vitrified Am/Cm product to Oak Ridge National Laboratory (ORNL).The plutonium solubility in the oxalate precipitation supernate during pretreatment was 10 mg/mL at 35 degrees C. In two subsequent washes with a 0.25M oxalic acid/0.5M nitric acid solution, the solubility dropped to less than 5 mg/mL. During the precipitation and washing steps, lanthanide fission products in the solution were mostly insoluble. Uranium, and alkali, alkaline earth, and transition metal impurities were soluble as expected. An elemental material balance for plutonium showed that greater than 94 percent of the plutonium was recovered in the dissolved precipitate. The recovery of the lanthanide elements was generally 94 percent or higher except for the more soluble lanthanum. The recovery of soluble metal impurities from the precipitate slurry ranged from 15 to 22 percent. Theoretically, 16 percent of the soluble oxalates should have been present in the dissolved slurry based on the dilution effects and volumes of supernate and wash solutions removed. A trace level material balance showed greater than 97 percent recovery of americium-241 (from the beta dec

Setting for technological control of vibropacked uranium-plutonium fuel pins

International Nuclear Information System (INIS)

Golushko, V.V.; Semenov, A.L.; Chukhlova, O.P.; Kuznetsov, A.M.; Korchkov, Yu.N.; Kandrashina, T.A.

1991-01-01

Scanning set-up providing for control of fuel pins by quality of fuel distribution in them is described. The gamma absorption method of fuel density measurement and the method of its own radiation registration are applied. Scintillation detection blocks are used in the measuring equipment mainly consisting of standard CAMAC blocks. Automation of measurements is performed on the basis of the computer complex MERA-60. A complex of programs for automation of the procedures under way is developed, when the facility operates within the test production line of vibroracked uranium-plutonium fuel pins. 6 refs.; 4 figs.; 1 tabs

Exploiting the plutonium stockpiles in PWRs by using inert matrix fuel

International Nuclear Information System (INIS)

Lombardi, C.; Mazzola, A.

1996-01-01

The plutonium coming from dismantled warheads and that already stockpiled coming from spent fuel reprocessing have raised many concerns related to proliferation resistance, environmental safety and economy. The option of disposing of plutonium by fission is one of the most widely discussed and many proposals for plutonium burning in a safe and economical manner have been put forward. Due to their diffusion, PWRs appear to be the main candidates for the reduction of the plutonium stockpiles. In order to achieve a high plutonium consumption rate, a uranium-free fuel may be conceived, based on the dilution of PuO 2 within a carrier matrix made of inert oxide. In this paper, a partial loading of inert matrix fuel in a current technology PWR was investigated with 3-D calculations. The results indicated that this solution has good plutonium elimination capabilities: commercial PWRs operating in a once-through cycle scheme can transmute more than 98% of the loaded Pu-239 and 73 or 81% of the overall initially loaded reactor grade or weapons grade plutonium, respectively. The plutonium still let in the spent fuel was of poor quality and then offered a better proliferation resistance. Power peaking problems could be faced with the adoption of burnable absorbers: IFBA seemed to be particularly suitable. In spite of a reduction of the overall plutonium loaded mass by a factor 3.7 or 5.4 depending on its quality, there was no evidence of an increase of the minor actinides radiotoxicity after a time period of about 25 years. (author)

Reduction stripping of plutonium (IV) with hydroxylamine nitrate in the centrifugal contactor

International Nuclear Information System (INIS)

Zhang Zefu; Ding Dachun; Fei Hongcheng

1989-01-01

Reduction stripping of Pu(IV) with hydroxylamine nitrate under conditions of Purex 3B contactor is carried out with 16-stage mixer-centrifugal settler in laboratory scale. Results show that the apparent recovery of Pu is more than 99.9%. But there is a great difference in the concentration of plutonium of every stage in extraction equipment at different temperature. For example, plutonium concentrations in organic phase of several stages near the organic effluent at 45-50 deg C are lower than that at 20-24 deg C. Thus, recovery efficiency of plutonium can be assured only for the former. However, plutonium will probably be lost, if operating conditions are fluctuated. In addition, plutonium accumulated in the extraction equipment increased appreciably for the case of 20-24 deg C. Therefore, in order to ensure recovery efficiency of Pu, reduction stripping of Pu should be made at higher temperature

Determination of plutonium in environment

International Nuclear Information System (INIS)

Sakanoue, Masanobu

1978-01-01

Past and present methods of determining the amount of plutonium in the environment are summarized. Determination of the amount of plutonium in uranium ore began in 1941. Plutonium present in polluted environments due to nuclear explosions, nuclear power stations, etc. was measured in soil and sand in Nagasaki in 1951 and in ash in Bikini in 1954. Analytical methods of measuring the least amount of plutonium in the environment were developed twenty years later. Many studies on and reviews of these methods have been reported all over the world, and a standard analytical procedure has been adopted. A basic analytical method of measurement was drafted in Japan in 1976. The yield, treatment of samples, dissolution, separation, control of measurable ray sources determination by α spectrometry, cross-check determination, and treatment of samples containing hardly soluble plutonium were examined. At present, the amount of plutonium can be determined by all of these methods. The presence of plutonium was studied further, and the usefulness of determination of the plutonium isotope ratio is discussed. (Kumagai, S.)

MIS High-Purity Plutonium Oxide Hydride Product 5501579 (SSR124): Final Report

Energy Technology Data Exchange (ETDEWEB)

Veirs, Douglas Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stroud, Mary Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Berg, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Narlesky, Joshua Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinex, Max A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carillo, Alex [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-02-08

A high-purity plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. Sample 5501579 represents process plutonium oxides from hydride oxide from Rocky Flats that are currently stored in 3013 containers. After calcination to 950°C, the material contained 87.42% plutonium with no major impurities. This study followed over time, the gas pressure of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. Gas compositions were measured periodically over a six year period. The maximum observed gas pressure was 124 kPa. The increase over the initial pressure of 70 kPa was primarily due to generation of nitrogen and carbon dioxide gas. Hydrogen and oxygen were minor components of the headspace gas. At the completion of the study, the internal components of the sealed container showed signs of corrosion.

Multi-generational stewardship of plutonium

International Nuclear Information System (INIS)

Pillay, K.K.S.

1997-01-01

The post-cold war era has greatly enhanced the interest in the long-term stewardship of plutonium. The management of excess plutonium from proposed nuclear weapons dismantlement has been the subject of numerous intellectual discussions during the past several years. In this context, issues relevant to long-term management of all plutonium as a valuable energy resource are also being examined. While there are differing views about the future role of plutonium in the economy, there is a recognition of the environmental and health related problems and proliferation potentials of weapons-grade plutonium. The long-term management of plutonium as an energy resource will require a new strategy to maintain stewardship for many generations to come

Plutonium Immobilization Project Binder Burnout and Sintering Studies (Milestone 6.6a)

International Nuclear Information System (INIS)

Chandler, G.

1999-01-01

The Plutonium Immobilization Team has developed an integrated test program to understand and optimize the controlling variables for the sintering step of the plutonium immobilization process. Sintering is the key process step that controls the product mineralogy. It is expected that the sintering will be the limiting process step that controls the throughput of the production line. The goal of the current sintering test program is to better understand factors that affect the sintering process

Plutonium Immobilization Project Binder Burnout and Sintering Studies (Milestone 6.6a)

Energy Technology Data Exchange (ETDEWEB)

Chandler, G.

1999-10-28

The Plutonium Immobilization Team has developed an integrated test program to understand and optimize the controlling variables for the sintering step of the plutonium immobilization process. Sintering is the key process step that controls the product minerology. It is expected that the sintering will be the limiting process step that controls the throughput of the production line. The goal of the current sintering test program is to better understand factors that affect the sintering process.

Plutonium in the marine environment

International Nuclear Information System (INIS)

Jarvis, N.V.; Linder, P.W.; Wade, P.W.

1994-01-01

The shipping of plutonium from Europe to Japan around the Cape is a contentious issue which has raised public concern that South Africa may be at risk to plutonium exposure should an accident occur. The paper describes the containers in which the plutonium (in the form of plutonium oxide, PuO 2 ) is housed and consequences of the unlikely event of these becoming ruptured. Wind-borne pollution is considered not to be a likely scenario, with the plutonium oxide particles more likely to remain practically insoluble and sediment. Plutonium aqueous and environmental chemistry is briefly discussed. Some computer modelling whereby plutonium oxide is brought into contact with seawater has been performed and the results are presented. The impact on marine organisms is discussed in terms of studies performed at marine dump sites and after the crash of a bomber carrying nuclear warheads in Thule, Greenland in 1968. Various pathways from the sea to land are considered in the light of studies done at Sellafield, a reprocessing plant in the United Kingdom. Some recent debates in the popular scientific press, such as that on the leukemia cluster at Sellafield, are described. Plutonium biochemistry and toxicity are discussed as well as medical histories of workers exposed to plutonium. 35 refs., 2 tabs., 1 fig

Dosage of plutonium by isotopic dilution in irradiated fuels; Dosage du plutonium par dilution isotopique dans les combustibles irradies

Energy Technology Data Exchange (ETDEWEB)

Lucas, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Plutonium determination in irradiated fuels has been carried out for several years by isotopic dilution by Sebaci and SSM in collaboration. SECACI has made available to the SSM the necessary space and equipment in its Fontenay laboratories. This work has shown the importance of the valency cycle which should make it possible to obtain a uniform isotopic distribution in sample tracer mixtures, and also a satisfactory U/Pu separation. Now it has been noticed that the presence of an excess of uranium considerably modifies the oxidation-reduction reaction kinetics of the plutonium. We have therefore been led to change certain parts of the operational technique so as to have an efficient cycle and to thereby improve the U/Pu separation; the stability of the thermionic emission of the plutonium, connected to the quantity of residual uranium, has at the same time been improved and we can now carry out more precise isotopic analyses. We have also tried to eliminate as far as possible the isotopic contaminations by:using a more rational operational method; the equipment used has been the object of a special study. The evaporations are carried out so as to prevent the formation of saturated vapours inside the glove box. The material which cannot be changed after each operation is carefully cleaned every time a new sample is treated. With this technique, a second calibration of the tracer T{sub 2} has been undertaken using a new standard solution. This solution has been prepared very carefully by weighing uranium and plutonium of known chemical purity, and we believe that it can be guaranteed to be a good reference solution. The value of the {sup 233}U/{sup 242}Pu ratio of the tracer has been obtained with a relative accuracy of 0,5 per cent. This modified method is now being applied to the analysis of rods irradiated in G-3. (author) [French] La determination du plutonium par dilution isotopique dans les combustibles irradies est pratiquee depuis plusieurs annees en

Plutonium Training Opportunities

Energy Technology Data Exchange (ETDEWEB)

Balatsky, Galya Ivanovna [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wolkov, Benjamin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-03-26

This report was created to examine the current state of plutonium training in the United States and to discover ways in which to ensure that the next generation of plutonium workers are fully qualified.

Qualitative chemical analysis of plutonium by Alpha spectroscopy.; Determinacion cualitativa de plutonio mediante espectroscopia alfa.

Energy Technology Data Exchange (ETDEWEB)

Ramirez G, J Qumica.J.

1994-12-31

In this work the separation and purification of plutonium from irradiated uranium was done. The plutonium, produced by the irradiation of uranium in a nuclear reactor and the {beta} decay of {sup 239} Np, was stabilized to Pu {sup +4} with sodium nitrite. Plutonium was separated from the fission products and uranium by ion exchange using the resin Ag 1 X 8. It was electrodeposited on stainless steel discs and the alpha radioactivity of plutonium was measured in a surface barrier detector. The results showed that plutonium was separated with a radiochemical purity higher than 99 %. (Author).

Optimizing Plutonium stock management

International Nuclear Information System (INIS)

Niquil, Y.; Guillot, J.

1997-01-01

Plutonium from spent fuel reprocessing is reused in new MOX assemblies. Since plutonium isotopic composition deteriorates with time, it is necessary to optimize plutonium stock management over a long period, to guarantee safe procurement, and contribute to a nuclear fuel cycle policy at the lowest cost. This optimization is provided by the prototype software POMAR

Recovery of uranium and plutonium from Redox off-standard aqueous waste streams

Energy Technology Data Exchange (ETDEWEB)

Holm, C.H.; Matheson, A.R.

1949-12-31

In the operation of countercurrent extraction columns as in the Redox process, it is possible, and probable, that from unexpected behaviour of a column, operator error, colloid formation, etc., there will result from time to time excessive losses of uranium and plutonium in the overall process. These losses will naturally accumulate in the waste streams, particularly in the aqueous waste streams. If the loss is excessively high, and such lost material can be recovered by some additional method, then if economical and within reason, the recovered materials ran be returned to a ISF column for further processing. The objective of this work has been to develop such a method to recover uranium and plutonium from such off-standard waste streams in a form whereby the uranium send plutonium can be returned to the process line and subsequently purified and separated.

Chemical states of fission products in irradiated uranium-plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Kurosaki, Ken; Uno, Masayoshi; Yamanaka, Shinsuke

1999-01-01

The chemical states of fission products (FPs) in irradiated uranium-plutonium mixed oxide (MOX) fuel for the light water reactor (LWR) were estimated by thermodynamic equilibrium calculations on system of fuel and FPs by using ChemSage program. A stoichiometric MOX containing 6.1 wt. percent PuO 2 was taken as a loading fuel. The variation of chemical states of FPs was calculated as a function of oxygen potential. Some pieces of information obtained by the calculation were compared with the results of the post-irradiation examination (PIE) of UO 2 fuel. It was confirmed that the multicomponent and multiphase thermodynamic equilibrium calculation between fuel and FPs system was an effective tool for understanding the behavior of FPs in fuel. (author)

a Plutonium Ceramic Target for Masha

Science.gov (United States)

Wilk, P. A.; Shaughnessy, D. A.; Moody, K. J.; Kenneally, J. M.; Wild, J. F.; Stoyer, M. A.; Patin, J. B.; Lougheed, R. W.; Ebbinghaus, B. B.; Landingham, R. L.; Oganessian, Yu. Ts.; Yeremin, A. V.; Dmitriev, S. N.

2005-09-01

We are currently developing a plutonium ceramic target for the MASHA mass separator. The MASHA separator will use a thick plutonium ceramic target capable of tolerating temperatures up to 2000 °C. Promising candidates for the target include oxides and carbides, although more research into their thermodynamic properties will be required. Reaction products will diffuse out of the target into an ion source, where they will then be transported through the separator to a position-sensitive focal-plane detector array. Experiments on MASHA will allow us to make measurements that will cement our identification of element 114 and provide for future experiments where the chemical properties of the heaviest elements are studied.