1. West facade of Plutonium Concentration Facility (Building 233S), ReductionOxidation ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

1. West facade of Plutonium Concentration Facility (Building 233-S), Reduction-Oxidation Building (REDOX-202-S) to the right. Looking east. - Reduction-Oxidation Complex, Plutonium Concentration Facility, 200 West Area, Richland, Benton County, WA

8. VIEW OF FOUNDRY INDUCTION FURNACES, MODULE J. THE FOUNDRY ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

8. VIEW OF FOUNDRY INDUCTION FURNACES, MODULE J. THE FOUNDRY CASTING PROCESS WAS CONDUCTED IN A VACUUM. PLUTONIUM METAL WAS MELTED IN ONE OF FOUR ELECTRIC INDUCTION FURNACES TO FORM INGOTS. - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

Project management plan, Waste Receiving and Processing Facility, Module 1, Project W-026

DOE Office of Scientific and Technical Information (OSTI.GOV)

Starkey, J.G.

1993-05-01

The Hanford Waste Receiving and Processing Facility Module 1 Project (WRAP 1) has been established to support the retrieval and final disposal of approximately 400K grams of plutonium and quantities of hazardous components currently stored in drums at the Hanford Site.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlton, W.H.; Evans, A.G.; Geary, L.A.

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 nuclearmore » 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.« less

Melter Technologies Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perez, J.M. Jr.; Schumacher, R.F.; Forsberg, C.W.

1996-05-01

The problem of controlling and disposing of surplus fissile material, in particular plutonium, is being addressed by the US Department of Energy (DOE). Immobilization of plutonium by vitrification has been identified as a promising solution. The Melter Evaluation Activity of DOE`s Plutonium Immobilization Task is responsible for evaluating and selecting the preferred melter technologies for vitrification for each of three immobilization options: Greenfield Facility, Adjunct Melter Facility, and Can-In-Canister. A significant number of melter technologies are available for evaluation as a result of vitrification research and development throughout the international communities for over 20 years. This paper describes an evaluationmore » process which will establish the specific requirements of performance against which candidate melter technologies can be carefully evaluated. Melter technologies that have been identified are also described.« less

LANL robotics site overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beugelsdijk, T.J.

1990-11-01

This paper reports on robotics applications at the Los Alamos National Laboratory. The topics of the paper include the ROBOCAL project to assay all nuclear materials entering and leaving the process floor at the Los Alamos Plutonium Facility, the isotope detector fabrication project, a plutonium dissolution robotic system, a safeguards waste automated measurement instrument, and DNA filter array construction. This report consists of overheads only.

4. AERIAL VIEW, LOOKING SOUTHSOUTHWEST, OF BUILDING 371 GROUND FLOOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

4. AERIAL VIEW, LOOKING SOUTH-SOUTHWEST, OF BUILDING 371 GROUND FLOOR UNDER CONSTRUCTION. THE GROUND FLOOR, WHICH CONTAINS THE MAJORITY OF THE PLUTONIUM RECOVERY PROCESSING EQUIPMENT, IS DIVIDED INTO COMPARTMENTS BY FIREWALLS, AIRLOCKS, AND USE OF NEGATIVE AIR PRESSURE. (1/7/75) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

Chemical interaction matrix between reagents in a Purex based process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brahman, R.K.; Hennessy, W.P.; Paviet-Hartmann, P.

2008-07-01

The United States Department of Energy (DOE) is the responsible entity for the disposal of the United States excess weapons grade plutonium. DOE selected a PUREX-based process to convert plutonium to low-enriched mixed oxide fuel for use in commercial nuclear power plants. To initiate this process in the United States, a Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) is under construction and will be operated by Shaw AREVA MOX Services at the Savannah River Site. This facility will be licensed and regulated by the U.S. Nuclear Regulatory Commission (NRC). A PUREX process, similar to the one used at La Hague,more » France, will purify plutonium feedstock through solvent extraction. MFFF employs two major process operations to manufacture MOX fuel assemblies: (1) the Aqueous Polishing (AP) process to remove gallium and other impurities from plutonium feedstock and (2) the MOX fuel fabrication process (MP), which processes the oxides into pellets and manufactures the MOX fuel assemblies. The AP process consists of three major steps, dissolution, purification, and conversion, and is the center of the primary chemical processing. A study of process hazards controls has been initiated that will provide knowledge and protection against the chemical risks associated from mixing of reagents over the life time of the process. This paper presents a comprehensive chemical interaction matrix evaluation for the reagents used in the PUREX-based process. Chemical interaction matrix supplements the process conditions by providing a checklist of any potential inadvertent chemical reactions that may take place. It also identifies the chemical compatibility/incompatibility of the reagents if mixed by failure of operations or equipment within the process itself or mixed inadvertently by a technician in the laboratories. (aut0010ho.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

CHARBONEAU, S.L.

The Plutonium Finishing Plant (PFP) consists of a number of process and support buildings for handling plutonium. Building construction began in the late 1940's to meet national priorities and became operational in 1950 producing refined plutonium salts and metal for the United States nuclear weapons program. The primary mission of the PFP was to provide plutonium used as special nuclear material for fabrication into a nuclear device for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race. PFP has nowmore » completed its mission and is fully engaged in deactivation, decontamination and decommissioning (D&D). At this time the PFP buildings are planned to be reduced to ground level (slab-on-grade) and the site remediated to satisfy national, Department of Energy (DOE) and Washington state requirements. The D&D of a highly contaminated plutonium processing facility presents a plethora of challenges. PFP personnel approached the D&D mission with a can-do attitude. They went into D&D knowing they were facing a lot of challenges and unknowns. There were concerns about the configuration control associated with drawings of these old process facilities. There were unknowns regarding the location of electrical lines and process piping containing chemical residues such as strong acids and caustics. The gloveboxes were highly contaminated with plutonium and chemical residues. Most of the glovebox windows were opaque with splashed process chemicals that coated the windows or etched them, reducing visibility to near zero. Visibility into the glovebox was a serious worker concern. Additionally, all the gloves in the gloveboxes were degraded and unusable. Replacing gloves in gloveboxes was necessary to even begin glovebox cleanout. The sheer volume of breathing air needed was also an issue. These and other challenges and PFP's approach to overcome these challengers are described. Many of the challenges to the D&D work at PFP were met with innovative approaches based on new science and/or technology and many were also based on the creativity and motivation of the work force personnel.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Charboneau, S.; Klos, B.; Heineman, R.

The Plutonium Finishing Plant (PFP) consists of a number of process and support buildings for handling plutonium. Building construction began in the late 1940's to meet national priorities and became operational in 1950 producing refined plutonium salts and metal for the United States nuclear weapons program The primary mission of the PFP was to provide plutonium used as special nuclear material for fabrication into a nuclear device for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race. PFP has nowmore » completed its mission and is fully engaged in deactivation, decontamination and decommissioning (D and D). At this time the PFP buildings are planned to be reduced to ground level (slab-on-grade) and the site remediated to satisfy national, Department of Energy (DOE) and Washington state requirements. The D and D of a highly contaminated plutonium processing facility presents a plethora of challenges. PFP personnel approached the D and D mission with a can-do attitude. They went into D and D knowing they were facing a lot of challenges and unknowns. There were concerns about the configuration control associated with drawings of these old process facilities. There were unknowns regarding the location of electrical lines and the condition and contents of process piping containing chemical residues such as strong acids and caustics. The gloveboxes were highly contaminated with plutonium and chemical residues. Most of the glovebox windows were opaque with splashed process chemicals that coated the windows or etched them, reducing visibility to near zero. Visibility into the glovebox was a serious worker concern. Additionally, all the gloves in the gloveboxes were degraded and unusable. Replacing gloves in gloveboxes was necessary to even begin glovebox clean-out. The sheer volume of breathing air needed was also an issue. These and other challenges and PFP's approach to overcome these challengers are described. Many of the challenges to the D and D work at PFP were met with innovative approaches based on new science and/or technology and many were also based on the creativity and motivation of the work force personnel. (authors)« less

Feasibility study of a gamma camera for monitoring nuclear materials in the PRIDE facility

NASA Astrophysics Data System (ADS)

Jo, Woo Jin; Kim, Hyun-Il; An, Su Jung; Lee, Chae Young; Song, Han-Kyeol; Chung, Yong Hyun; Shin, Hee-Sung; Ahn, Seong-Kyu; Park, Se-Hwan

2014-05-01

The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocessing technology, in which actinides are recovered together with plutonium. There is no pure plutonium stream in the process, so it has an advantage of proliferation resistance. Tracking and monitoring of nuclear materials through the pyroprocess can significantly improve the transparency of the operation and safeguards. An inactive engineering-scale integrated pyroprocess facility, which is the PyRoprocess Integrated inactive DEmonstration (PRIDE) facility, was constructed to demonstrate engineering-scale processes and the integration of each unit process. the PRIDE facility may be a good test bed to investigate the feasibility of a nuclear material monitoring system. In this study, we designed a gamma camera system for nuclear material monitoring in the PRIDE facility by using a Monte Carlo simulation, and we validated the feasibility of this system. Two scenarios, according to locations of the gamma camera, were simulated using GATE (GEANT4 Application for Tomographic Emission) version 6. A prototype gamma camera with a diverging-slat collimator was developed, and the simulated and experimented results agreed well with each other. These results indicate that a gamma camera to monitor the nuclear material in the PRIDE facility can be developed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reilly, Sean Douglas; Smith, Paul Herrick; Jarvinen, Gordon D.

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that themore » following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U 3O 8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl 3, and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate buffer would significantly reduce the solubility of PuCl 3 by the precipitation of PuPO 4.« less

18. Process area room. Incinerator to the left. Filter boxes ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

18. Process area room. Incinerator to the left. Filter boxes on the right. Looking north towards change room. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

Literature review for oxalate oxidation processes and plutonium oxalate solubility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nash, C. A.

2015-10-01

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.more » 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.« less

7. Process areas room. Incinerator and glove boxes (hoods) to ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

7. Process areas room. Incinerator and glove boxes (hoods) to the right. Filter boxes to the left. Looking south. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

Dissolution of aerosol particles collected from nuclear facility plutonium production process

DOE PAGES

Xu, Ning; Martinez, Alexander; Schappert, Michael Francis; ...

2015-08-14

Here, 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 °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.

Nuclear Matters. A Practical Guide

DTIC Science & Technology

2008-01-01

plutonium science and engineering. Figure 4.6 depicts LANL workers in Technical Area (TA)-55, the Los Alamos plutonium facility. LANL oversees...facility at Los Alamos to produce plutonium pits in a laboratory environment, with a capacity to produce a small number of pits per year . At that...Office of Secure Transportation (OST). Technical Advisors represent the following organizations: Los Alamos National Chair ATSD(NCB) Vice-Chair

Determination of filter pore size for use in HB line phase II production of plutonium oxide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shehee, T.; Crowder, M.; Rudisill, T.

2014-08-01

H-Canyon and HB-Line are tasked with the production of plutonium oxide (PuO 2) from a feed of plutonium (Pu) metal. The PuO 2 will provide feed material for the Mixed Oxide (MOX) Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, plans are to transfer the solution to HB-Line for purification by anion exchange. Anion exchange will be followed by plutonium(IV) oxalate precipitation, filtration, and calcination to form PuO 2. The filtrate solutions, remaining after precipitation, contain low levels of Pu ions, oxalate ions, and may include solids. These solutions are transferred to H-Canyon for disposition. To mitigatemore » the criticality concern of Pu solids in a Canyon tank, past processes have used oxalate destruction or have pre-filled the Canyon tank with a neutron poison. The installation of a filter on the process lines from the HB-Line filtrate tanks to H-Canyon Tank 9.6 is proposed to remove plutonium oxalate solids. This report describes SRNL’s efforts to determine the appropriate pore size for the filters needed to perform this function. Information provided in this report aids in developing the control strategies for solids in the process.« less

Plutonium storage criteria

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, D.; Ascanio, X.

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 thanmore » 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.« less

Balanced program plan. Analysis for biomedical and environmental research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-06-01

Major issues associated with the use of nuclear power are health hazards of exposure to radioactive materials; sources of radiation exposure; reactor accidents; sabotage of nuclear facilities; diversion of fissile material and its use for extortion; and the presence of plutonium in the environment. Fission fuel cycle technology is discussed with regard to milling, UF/sub 6/ production, uranium enrichment, plutonium fuel fabrication, power production, fuel processing, waste management, and fuel and waste transportation. The following problem areas of fuel cycle technology are briefly discussed: characterization, measurement, and monitoring; transport processes; health effects; ecological processes and effects; and integrated assessment. Estimatedmore » program unit costs are summarized by King-Muir Category. (HLW)« less

Certification of Plutonium Standards for KAMS Neutron Multiplicity Counter

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salaymeh, S.R.

2002-05-31

As part of the implementation of the PEIS record of decision in January of 1997, DOE will pursue two technologies to disposition fifty metric tons of its stockpile of plutonium. As a result of this and in order to expedite the closure of Rocky Flats Environmental Technology Site in Colorado, DOE decided to use existing facilities at the Savannah River Site (SRS) for storing all material containing plutonium at KAMS. A neutron multiplicity counter was designed and built to carry out receipt verification measurement at the facility. Since the material covers a wide range and different levels of impurities, itmore » is essential that we obtain a set of working standards. An agreement was drafted to select the first drums to be these standards. A plan was developed for the certification of these standards using Rocky Flat's existing nondestructive assay equipment. This paper will discuss the types of materials to be shipped to SRS, number of standards to certify for each type of material, and the certification plan. It will also discuss the activities necessary to determine the nuclear content of these working standards to be used at SRS facilities in support of shipment and receipt of the Pu containing materials. Definition of instrument qualifications, measurement control processes, measurement methodologies, and calculations necessary to report the gram quantities and their uncertainties for plutonium, americium-241, uranium-235 (if present) and neptunium-237 (if present) will also be presented.« less

High-Precision Plutonium Isotopic Compositions Measured on Los Alamos National Laboratory’s General’s Tanks Samples: Bearing on Model Ages, Reactor Modelling, and Sources of Material. Further Discussion of Chronometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Khalil J.; Rim, Jung Ho; Porterfield, Donivan R.

2015-06-29

In this study, we re-analyzed late-1940’s, Manhattan Project era Plutonium-rich sludge samples recovered from the ''General’s Tanks'' located within the nation’s oldest Plutonium processing facility, Technical Area 21. These samples were initially characterized by lower accuracy, and lower precision mass spectrometric techniques. We report here information that was previously not discernable: the two tanks contain isotopically distinct Pu not only for the major (i.e., 240Pu, 239Pu) but trace ( 238Pu , 241Pu, 242Pu) isotopes. Revised isotopics slightly changed the calculated 241Am- 241Pu model ages and interpretations.

Guide of good practices for occupational radiological protection in plutonium facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1998-06-01

This Technical Standard (TS) does not contain any new requirements. Its purpose is to provide guides to good practice, update existing reference material, and discuss practical lessons learned relevant to the safe handling of plutonium. the technical rationale is given to allow US Department of Energy (DOE) health physicists to adapt the recommendations to similar situations throughout the DOE complex. Generally, DOE contractor health physicists will be responsible to implement radiation protection activities at DOE facilities and DOE health physicists will be responsible for oversight of those activities. This guidance is meant to be useful for both efforts. This TSmore » replaces PNL-6534, Health Physics Manual of Good Practices for Plutonium Facilities, by providing more complete and current information and by emphasizing the situations that are typical of DOE`s current plutonium operations; safe storage, decontamination, and decommissioning (environmental restoration); and weapons disassembly.« less

10 CFR 830.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

.... Critical assembly means special nuclear devices designed and used to sustain nuclear reactions, which may... reaction becomes self-sustaining. Design features means the design features of a nuclear facility specified... reaction (e.g., uranium-233, uranium-235, plutonium-238, plutonium-239, plutonium-241, neptunium-237...

Evaluation of continuous air monitor placement in a plutonium facility.

PubMed

Whicker, J J; Rodgers, J C; Fairchild, C I; Scripsick, R C; Lopez, R C

1997-05-01

Department of Energy appraisers found continuous air monitors at Department of Energy plutonium facilities alarmed less than 30% of the time when integrated room plutonium air concentrations exceeded 500 DAC-hours. Without other interventions, this alarm percentage suggests the possibility that workers could be exposed to high airborne concentrations without continuous air monitor alarms. Past research has shown that placement of continuous air monitors is a critical component in rapid and reliable detection of airborne releases. At Los Alamos National Laboratory and many other Department of Energy plutonium facilities, continuous air monitors have been primarily placed at ventilation exhaust points. The purpose of this study was to evaluate and compare the effectiveness of exhaust register placement of workplace continuous air monitors with other sampling locations. Polydisperse oil aerosols were released from multiple locations in two plutonium laboratories at Los Alamos National Laboratory. An array of laser particle counters positioned in the rooms measured time-resolved aerosol dispersion. Results showed alternative placement of air samplers generally resulted in aerosol detection that was faster, often more sensitive, and equally reliable compared with samplers at exhaust registers.

SRP engineering and design history, Vol III, 200 F and H Areas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Banick, C.J.

2000-04-17

This volume combines the record of events relating to the development of design for both the 200-F and H Areas. Chronologically, the definition of plant facilities was first established for the 200-F Area. The second area, 200-H, was projected initially to be a supplementary plutonium separations facility. This history explains the differences in character and capacity of the manufacturing facilities in both areas as production requirements and experience with separations processes advanced.

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2013 CFR

2013-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2012 CFR

2012-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

10 CFR 140.91 - Appendix A-Form of nuclear energy liability policy for facilities.

Code of Federal Regulations, 2011 CFR

2011-01-01

... designed or used for (a) separating the isotopes of uranium or plutonium, (b) processing or utilizing spent... processing, fabricating or alloying of special nuclear material if at any time the total amount of such... operations conducted thereat; Nuclear reactor means any apparatus designed or used to sustain nuclear fission...

In search of plutonium: A nonproliferation journey

NASA Astrophysics Data System (ADS)

Hecker, Siegfried

2010-02-01

In February 1992, I landed in the formerly secret city of Sarov, the Russian Los Alamos, followed a few days later by a visit to Snezhinsk, their Livermore. The briefings we received of the Russian nuclear weapons program and tours of their plutonium, reactor, explosives, and laser facilities were mind boggling considering the Soviet Union was dissolved only two months earlier. This visit began a 17-year, 41 journey relationship with the Russian nuclear complex dedicated to working with them in partnership to protect and safeguard their weapons and fissile materials, while addressing the plight of their scientists and engineers. In the process, we solved a forty-year disagreement about the plutonium-gallium phase diagram and began a series of fundamental plutonium science workshops that are now in their tenth year. At the Yonbyon reprocessing facility in January 2004, my North Korean hosts had hoped to convince me that they have a nuclear deterrent. When I expressed skepticism, they asked if I wanted to see their ``product.'' I asked if they meant the plutonium; they replied, ``Well, yes.'' Thus, I wound up holding 200 grams of North Korean plutonium (in a sealed glass jar) to make sure it was heavy and warm. So began the first of my six journeys to North Korea to provide technical input to the continuing North Korean nuclear puzzle. In Trombay and Kalpakkam a few years later I visited the Indian nuclear research centers to try to understand how India's ambitious plans for nuclear power expansion can be accomplished safely and securely. I will describe these and other attempts to deal with the nonproliferation legacy of the cold war and the new challenges ahead. )

25. Plutonium Recovery From Contaminated Materials, Architectural Plans & Details, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

25. Plutonium Recovery From Contaminated Materials, Architectural Plans & Details, Building 232-Z, U.S. Atomic Energy Commission, Hanford Atomic Products Operation, General Electric Company, Dwg. No. H-2-23105, 1959. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

24. Plutonium Recovery From Contaminated Materials, Architectural Details, Building 232z, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

24. Plutonium Recovery From Contaminated Materials, Architectural Details, Building 232-z, U.S. Atomic Energy Commission, Hanford Atomic Products Operation, General Electric Company, Dwg. No. H-2-23106, 1959. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

26. Plutonium Recovery From Contaminated Materials, Architectural Elevations, Sections & ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

26. Plutonium Recovery From Contaminated Materials, Architectural Elevations, Sections & Dets., Building 232-Z, U.S. Atomic Energy Commission, Hanford Atomic Products Operation, General Electric Company, Dwg. No. H-2-23106, 1959. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

NNSA B-Roll: MOX Facility

ScienceCinema

None

2017-12-09

In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

NNSA B-Roll: MOX Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2010-05-21

In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

1. VIEW LOOKING SOUTH AT BUILDING 771 UNDER CONSTRUCTION. BUILDING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

1. VIEW LOOKING SOUTH AT BUILDING 771 UNDER CONSTRUCTION. BUILDING 771 WAS ONE OF THE FIRST FOUR MAJOR BUILDINGS AT THE ROCKY FLATS PLANT, BUILDING 771 WAS ORIGINALLY THE PRIMARY FACILITY FOR PLUTONIUM OPERATIONS. (5/29/52) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

A continuous plutonium aerosol monitor for use in high radon environments.

PubMed

Li, HuiBin; Jia, MingYan; Li, GuoShen; Wang, YinDong

2012-01-01

Radon concentration is very high in underground basements and other facilities. Radon concentration in a nuclear facility locates in the granite tunnel can be as high as 10(4) Bq m(-3) in summer. Monitoring plutonium aerosol in this circumstance is seriously interfered by radon daughters. In order to solve this problem, a new continuous aerosol monitor that can monitor very low plutonium aerosol concentration in high radon background was developed. Several techniques were used to reduce interference of radon daughters, and the minimum detectable concentrations in various radon concentrations were measured.

9. VIEW, LOOKING WEST, OF GLOVE BOXES ASSOCIATED WITH THE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

9. VIEW, LOOKING WEST, OF GLOVE BOXES ASSOCIATED WITH THE ANION EXCHANGE PROCESS IN ROOM 149. THE GLOVE BOXES ON THE LEFT CONTAIN MIXER STIRRERS THAT AID IN THE DISSOLUTION PROCESS THAT OCCURRED PRIOR TO ANION EXCHANGE. (6/20/60) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

15. VIEW OF MODULE H, THE HIGH PRESSURE ASSEMBLY AREA. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

15. VIEW OF MODULE H, THE HIGH PRESSURE ASSEMBLY AREA. PROCESSES IN THIS MODULE OCCURRED UNDER HIGH PRESSURES AND TEMPERATURES. (5/70) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

18. VIEW OF THE CEILING, THE PIPING TRANSPORTED CHEMICALS FROM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

18. VIEW OF THE CEILING, THE PIPING TRANSPORTED CHEMICALS FROM A CHEMICAL PREPARATION ROOM ON THE SECOND FLOOR TO THE FIRST FLOOR PROCESS AREAS. (6/12/73) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

20. VIEW OF WASTE TREATMENT CONTROL ROOM IN BUILDING 374. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

20. VIEW OF WASTE TREATMENT CONTROL ROOM IN BUILDING 374. THE BUILDING 371/374 COMPLEX WAS DESIGNED TO EMPHASIZE AUTOMATICALLY CONTROLLED, REMOTELY OPERATED PROCESSES. (1/80) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

11. SIDE VIEW OF INSTALLATION OF A CONTINUOUS ROTARYTUBE HYDROFLUORINATOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

11. SIDE VIEW OF INSTALLATION OF A CONTINUOUS ROTARY-TUBE HYDROFLUORINATOR LOCATED IN ROOM 146. THE HYDROFLUORINATOR IS BEING INSTALLED INSIDE A GLOVE BOX. HYDROFLUORINATION CONVERTED PLUTONIUM OXIDE TO PLUTONIUM TETRAFLUORIDE. (1/11/62) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

14. END VIEW OF THE PLUTONIUM STORAGE VAULT FROM THE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

14. END VIEW OF THE PLUTONIUM STORAGE VAULT FROM THE REMOTE CONTROL STATION. THE STACKER-RETRIEVER, A REMOTELY-OPERATED, MECHANIZED TRANSPORT SYSTEM, RETRIEVES CONTAINERS OF PLUTONIUM FROM SAFE GEOMETRY PALLETS STORED ALONG THE LENGTH OF THE VAULT. THE STACKER-RETRIEVER RUNS ALONG THE AISLE BETWEEN THE PALLETS OF THE STORAGE CHAMBER. (3/2/86) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

Authorization basis supporting documentation for plutonium finishing plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, J.P., Fluor Daniel Hanford

1997-03-05

The identification and definition of the authorization basis for the Plutonium Finishing Plant (PFP) facility and operations are essential for compliance to DOE Order 5480.21, Unreviewed Safety Questions. The authorization basis, as defined in the Order, consists of those aspects of the facility design basis, i.e., the structures, systems and components (SSCS) and the operational requirements that are considered to be important to the safety of operations and are relied upon by DOE to authorize operation of the facility. These facility design features and their function in various accident scenarios are described in WHC-SD-CP-SAR-021, Plutonium Finishing Plant Final Safety Analysismore » Report (FSAR), Chapter 9, `Accident Analysis.` Figure 1 depicts the relationship of the Authorization Basis to its components and other information contained in safety documentation supporting the Authorization Basis. The PFP SSCs that are important to safety, collectively referred to as the `Safety Envelope` are discussed in various chapters of the FSAR and in WHC-SD-CP-OSR-010, Plutonium Finishing Plant Operational Safety Requirements. Other documents such as Criticality Safety Evaluation Reports (CSERS) address and support some portions of the Authorization Basis and Safety Envelope.« less

All About MOX

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2009-07-29

In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

All About MOX

ScienceCinema

None

2018-01-16

In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

13. VIEW OF THE HYDROFLUORINATOR AFTER INSTALLATION. SUPPLEMENTAL SHIELDING, WHICH ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

13. VIEW OF THE HYDROFLUORINATOR AFTER INSTALLATION. SUPPLEMENTAL SHIELDING, WHICH MOVES ALONG TRACKS IN THE FLOOR AND CEILING PROTECTS WORKERS FROM NEUTRON RADIATION EMISSIONS ASSOCIATED WITH THE PROCESS. (4/29/65) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

Estimation and characterization of decontamination and decommissioning solid waste expected from the Plutonium Finishing Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Millar, J.S.; Pottmeyer, J.A.; Stratton, T.J.

1995-01-01

Purpose of the study was to estimate the amounts of equipment and other materials that are candidates for removal and subsequent processing in a solid waste facility when the Hanford Plutonium Finishing Plant is decontaminated and decommissioned. (Building structure and soil are not covered.) Results indicate that {approximately}5,500 m{sup 3} of solid waste is expected to result from the decontamination and decommissioning of the Pu Finishing Plant. The breakdown of the volumes and percentages of waste by category is 1% dangerous solid waste, 71% low-level waste, 21% transuranic waste, 7% transuranic mixed waste.

2. VIEW LOOKING NORTHEAST AT BUILDING 444 UNDER CONSTRUCTION. BUILDING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

2. VIEW LOOKING NORTHEAST AT BUILDING 444 UNDER CONSTRUCTION. BUILDING 444 WAS THE PRIMARY NON-PLUTONIUM MANUFACTURING FACILITY AT THE ROCKY FLATS PLANT. MANUFACTURING PROCESSES COMPLETED IN THIS BUILDING WERE USED TO FABRICATE WEAPONS COMPONENTS AND ASSEMBLIES FOR A VARIETY OF MATERIALS, INCLUDING DEPLETED URANIUM, BERYLLIUM, STAINLESS STEEL, ALUMINUM, AND VANADIUM. (4/25/52) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

VIEW OF THE INTERIOR OF BUILDING 774, THE ORIGINAL LIQUID ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

VIEW OF THE INTERIOR OF BUILDING 774, THE ORIGINAL LIQUID PROCESS WASTEWATER TREATMENT FACILITY. THE PHOTOGRAPH SHOWS STORAGE TANKS AND ASSOCIATED PLUTONIUM-CONTAMINATED SOLUTIONS. THE GLOVE BOX IS USED BY OPERATORS TO MANUALLY OPERATE PUMPS AND VALVES THAT REQUIRE PERIODIC ADJUSTMENT. OTHER VALVES IN THE ROOM WERE INFREQUENTLY ADJUSTED, AND ARE SEALED IN PLASTIC WRAP - Rocky Flats Plant, Waste Treatment Facility, Adjacent to bldg 771C, in northern portion of protected area, Golden, Jefferson County, CO

Pyrochemical process for extracting plutonium from an electrolyte salt

DOEpatents

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

1982-09-20

A pyrochemical process for extracting plutonium from a plutonium-bearing salt is disclosed. The process is particularly useful in the recovery of plutonium for electrolyte salts which are left over from the electrorefining of plutonium. In accordance with the process, the plutonium-bearing salt is melted and mixed with metallic calcium. The calcium reduces ionized plutonium in the salt to plutonium metal, and also causes metallic plutonium in the salt, which is typically present as finely dispersed metallic shot, to coalesce. The reduced and coalesced plutonium separates out on the bottom of the reaction vessel as a separate metallic phase which is readily separable from the overlying salt upon cooling of the mixture. Yields of plutonium are typically on the order of 95%. The stripped salt is virtually free of plutonium and may be discarded to low-level waste storage.

Pyrochemical process for extracting plutonium from an electrolyte salt

DOEpatents

Mullins, Lawrence J.; Christensen, Dana C.

1984-01-01

A pyrochemical process for extracting plutonium from a plutonium-bearing salt is disclosed. The process is particularly useful in the recovery of plutonium from electrolyte salts which are left over from the electrorefining of plutonium. In accordance with the process, the plutonium-bearing salt is melted and mixed with metallic calcium. The calcium reduces ionized plutonium in the salt to plutonium metal, and also causes metallic plutonium in the salt, which is typically present as finely dispersed metallic shot, to coalesce. The reduced and coalesced plutonium separates out on the bottom of the reaction vessel as a separate metallic phase which is readily separable from the overlying salt upon cooling of the mixture. Yields of plutonium are typically on the order of 95%. The stripped salt is virtually free of plutonium and may be discarded to low-level waste storage.

Introduction to Pits and Weapons Systems (U)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kautz, D.

2012-07-02

A Nuclear Explosive Package includes the Primary, Secondary, Radiation Case and related components. This is the part of the weapon that produces nuclear yield and it converts mechanical energy into nuclear energy. The pit is composed of materials that allow mechanical energy to be converted to electromagnetic energy. Fabrication processes used are typical of any metal fabrication facility: casting, forming, machining and welding. Some of the materials used in pits include: Plutonium, Uranium, Stainless Steel, Beryllium, Titanium, and Aluminum. Gloveboxes are used for three reasons: (1) Protect workers and public from easily transported, finely divided plutonium oxides - (a) Plutoniummore » is very reactive and produces very fine particulate oxides, (b) While not the 'Most dangerous material in the world' of Manhattan Project lore, plutonium is hazardous to health of workers if not properly controlled; (2) Protect plutonium from reactive materials - (a) Plutonium is extremely reactive at ambient conditions with several components found in air: oxygen, water, hydrogen, (b) As with most reactive metals, reactions with these materials may be violent and difficult to control, (c) As with most fabricated metal products, corrosion may significantly affect the mechanical, chemical, and physical properties of the product; and (3) Provide shielding from radioactive decay products: {alpha}, {gamma}, and {eta} are commonly associated with plutonium decay, as well as highly radioactive materials such as {sup 241}Am and {sup 238}Pu.« less

Conceptual designs of NDA instruments for the NRTA system at the Rokkasho Reprocessing Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, T.K.; Klosterbuer, S.F.; Menlove, H.O.

The authors are studying conceptual designs of selected nondestructive assay (NDA) instruments for the near-real-time accounting system at the rokkasho Reprocessing Plant (RRP) of Japan Nuclear Fuel Limited (JNFL). The JNFL RRP is a large-scale commercial reprocessing facility for spent fuel from boiling-water and pressurized-water reactors. The facility comprises two major components: the main process area to separate and produce purified plutonium nitrate and uranyl nitrate from irradiated reactor spent fuels, and the co-denitration process area to combine and convert the plutonium nitrate and uranyl nitrate into mixed oxide (MOX). The selected NDA instruments for conceptual design studies are themore » MOX-product canister counter, holdup measurement systems for calcination and reduction furnaces and for blenders in the co-denitration process, the isotope dilution gamma-ray spectrometer for the spent fuel dissolver solution, and unattended verification systems. For more effective and practical safeguards and material control and accounting at RRP, the authors are also studying the conceptual design for the UO{sub 3} large-barrel counter. This paper discusses the state-of-the-art NDA conceptual design and research and development activities for the above instruments.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hurd, J.R.

The active-passive shuffler installed and certified a few years ago in Los Alamos National Laboratory`s plutonium facility has now been calibrated for different matrices to measure Waste Isolation Pilot Plant (WIPP)-destined transuranic (TRU)-waste. Little or no data presently exist for these types of measurements in plant environments where there may be sudden large changes in the neutron background radiation which causes distortions in the results. Measurements and analyses of twenty-two 55-gallon drums, consisting of mixtures of varying quantities of uranium and plutonium, have been recently completed at the plutonium facility. The calibration and measurement techniques, including the method used tomore » separate out the plutonium component, will be presented and discussed. Particular attention will be directed to those problems identified as arising from the plant environment. The results of studies to quantify the distortion effects in the data will be presented. Various solution scenarios will be indicated, along with those adopted here.« less

APPLICATION OF VACUUM SALT DISTILLATION TECHNOLOGY FOR THE REMOVAL OF FLUORIDE AND CHLORIDE FROM LEGACY FISSILE MATERIALS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pierce, R.; Peters, T.

2011-11-01

Between September 2009 and January 2011, the Savannah River National Laboratory (SRNL) and the Savannah River Site (SRS) HB-Line Facility designed, developed, tested, and successfully deployed a production-scale system for the distillation of sodium chloride (NaCl) and potassium chloride (KCl) from plutonium oxide (PuO{sub 2}). Subsequent efforts adapted the vacuum salt distillation (VSD) technology for the removal of chloride and fluoride from less-volatile halide salts at the same process temperature and vacuum. Calcium chloride (CaCl{sub 2}), calcium fluoride (CaF{sub 2}), and plutonium fluoride (PuF{sub 3}) were of particular concern. To enable the use of the same operating conditions for themore » distillation process, SRNL employed in situ exchange reactions to convert the less-volatile halide salts to compounds that facilitated the distillation of halide without removal of plutonium. SRNL demonstrated the removal of halide from CaCl{sub 2}, CaF{sub 2} and PuF{sub 3} below 1000 C using VSD technology.« less

1. VIEW OF THE CONTROL ROOM FOR THE XY RETRIEVER. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

1. VIEW OF THE CONTROL ROOM FOR THE X-Y RETRIEVER. USING THE X-Y RETRIEVER, OPERATORS RETRIEVED PLUTONIUM METAL FROM THE PLUTONIUM STORAGE VAULTS (IN MODULE K) AND CONVEYED IT TO THE X-Y SHUTTLE AREA WHERE IT WAS CUT AND WEIGHED. FROM THE SHUTTLE AREA THE PLUTONIUM WAS CONVEYED TO MODULES A, J OR K FOR CASTING, OR MODULE B FOR ROLLING AND FORMING. (5/17/71) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

Integrating the stabilization of nuclear materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dalton, H.F.

1996-05-01

In response to Recommendation 94-1 of the Defense Nuclear Facilities Safety Board, the Department of Energy committed to stabilizing specific nuclear materials within 3 and 8 years. These efforts are underway. The Department has already repackaged the plutonium at Rocky Flats and metal turnings at Savannah River that had been in contact with plastic. As this effort proceeds, we begin to look at activities beyond stabilization and prepare for the final disposition of these materials. To describe the plutonium materials being stabilize, Figure 1 illustrates the quantities of plutonium in various forms that will be stabilized. Plutonium as metal comprisesmore » 8.5 metric tons. Plutonium oxide contains 5.5 metric tons of plutonium. Plutonium residues and solutions, together, contain 7 metric tons of plutonium. Figure 2 shows the quantity of plutonium-bearing material in these four categories. In this depiction, 200 metric tons of plutonium residues and 400 metric tons of solutions containing plutonium constitute most of the material in the stabilization program. So, it is not surprising that much of the work in stabilization is directed toward the residues and solutions, even though they contain less of the plutonium.« less

Environmental consequences of postulate plutonium releases from Atomics International's Nuclear Materials Development Facility (NMDF), Santa Susana, California, as a result of severe natural phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jamison, J.D.; Watson, E.C.

1982-02-01

Potential environmental consequences in terms of radiation dose to people are presented for postulated plutonium releases caused by severe natural phenomena at the Atomics International's Nuclear Materials Development Facility (NMDF), in the Santa Susana site, California. The severe natural phenomena considered are earthquakes, tornadoes, and high straight-line winds. Plutonium deposition values are given for significant locations around the site. All important potential exposure pathways are examined. The most likely 50-year committed dose equivalents are given for the maximum-exposed individual and the population within a 50-mile radius of the plant. The maximum plutonium deposition values likely to occur offsite are alsomore » given. The most likely calculated 50-year collective committed dose equivalents are all much lower than the collective dose equivalent expected from 50 years of exposure to natural background radiation and medical x-rays. The most likely maximum residual plutonium contamination estimated to be deposited offsite following the earthquake, and the 150-mph and 170-mph tornadoes are above the Environmental Protection Agency's (EPA) proposed guideline for plutonium in the general environment of 0.2 ..mu..Ci/m/sup 2/. The deposition values following the 110-mph and the 130-mph tornadoes are below the EPA proposed guideline.« less

OPEN AIR DEMOLITION OF FACILITIES HIGHLY CONTAMINATED WITH PLUTONIUM

DOE Office of Scientific and Technical Information (OSTI.GOV)

LLOYD, E.R.

2007-05-31

The demolition of highly contaminated plutonium buildings usually is a long and expensive process that involves decontaminating the building to near free- release standards and then using conventional methods to remove the structure. It doesn't, however, have to be that way. Fluor has torn down buildings highly contaminated with plutonium without excessive decontamination. By removing the select source term and fixing the remaining contamination on the walls, ceilings, floors, and equipment surfaces; open-air demolition is not only feasible, but it can be done cheaper, better (safer), and faster. Open-air demolition techniques were used to demolish two highly contaminated buildings tomore » slab-on-grade. These facilities on the Department of Energy's Hanford Site were located in, or very near, compounds of operating nuclear facilities that housed hundreds of people working on a daily basis. To keep the facilities operating and the personnel safe, the projects had to be creative in demolishing the structures. Several key techniques were used to control contamination and keep it within the confines of the demolition area: spraying fixatives before demolition; applying fixative and misting with a fine spray of water as the buildings were being taken down; and demolishing the buildings in a controlled and methodical manner. In addition, detailed air-dispersion modeling was done to establish necessary building and meteorological conditions and to confirm the adequacy of the proposed methods. Both demolition projects were accomplished without any spread of contamination outside the modest buffer areas established for contamination control. Furthermore, personnel exposure to radiological and physical hazards was significantly reduced by using heavy equipment rather than ''hands on'' techniques.« less

Why SRS Matters - K Area

ScienceCinema

Hunt, Paul; Lawson, Janice

2018-06-22

A video series presenting an overview of the Savannah River Site (SRS) mission and operations. Each episode features a specific area/operation and how it contributes to help make the world safer. This episode features K Area's mission and operations. K area is the former production reactor that's been re-purposed to serve as a plutonium processing and storage facility.

Los Alamos National Laboratory and Lawrence Livermore National Laboratory Plutonium Sustainment Monthly Program Report September 2012

DOE Office of Scientific and Technical Information (OSTI.GOV)

McLaughlin, Anastasia Dawn; Storey, Bradford G.; Bowidowicz, Martin

In March of 2012 the Plutonium Sustainment program at LANL completed or addressed the following high-level activities: (1) Delivered Revision 2 of the Plutonium Sustainment Manufacturing Study, which incorporated changes needed due to the release of the FY2013 President's Budget and the delay in the Chemistry and Metallurgy Research Replacement Nuclear Facility (CMRRNF). (2) W87 pit type development activities completed a detailed process capability review for the flowsheet in preparation for the Engineering Development Unit Build. (3) Completed revising the Laser Beam Welding schedule to address scope and resource changes. (4) Completed machining and inspecting the first set of high-fidelitymore » cold parts on Precitech 2 for Gemini. (5) The Power Supply Assembly Area started floor cutting with a concrete saw and continued legacy equipment decommissioning. There are currently no major issues associated with achieving MRT L2 Milestones 4195-4198 or the relevant PBIs associated with Plutonium Sustainment. There are no budget issues associated with FY12 final budget guidance. Table 1 identifies all Baseline Change Requests (BCRs) that were initiated, in process, or completed during the month. The earned value metrics overall for LANL are within acceptable thresholds, so no high-level recovery plan is required. Each of the 5 major LANL WBS elements is discussed in detail.« less

15. VIEW OF LABORATORY EQUIPMENT IN THE BUILDING 771 ANALYTICAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

15. VIEW OF LABORATORY EQUIPMENT IN THE BUILDING 771 ANALYTICAL LABORATORY. THE LAB ANALYZED SAMPLES FOR PLUTONIUM, AMERICIUM, URANIUM, NEPTUNIUM, AND OTHER RADIOACTIVE ISOTOPES. (9/25/62) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

Safeguardability of the vitrification option for disposal of plutonium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pillay, K.K.S.

1996-05-01

Safeguardability of the vitrification option for plutonium disposition is rather complex and there is no experience base in either domestic or international safeguards for this approach. In the present treaty regime between the US and the states of the former Soviet Union, bilaterial verifications are considered more likely with potential for a third-party verification of safeguards. There are serious technological limitations to applying conventional bulk handling facility safeguards techniques to achieve independent verification of plutonium in borosilicate glass. If vitrification is the final disposition option chosen, maintaining continuity of knowledge of plutonium in glass matrices, especially those containing boron andmore » those spike with high-level wastes or {sup 137}Cs, is beyond the capability of present-day safeguards technologies and nondestructive assay techniques. The alternative to quantitative measurement of fissile content is to maintain continuity of knowledge through a combination of containment and surveillance, which is not the international norm for bulk handling facilities.« less

PROCESS OF FORMING PLUOTONIUM SALTS FROM PLUTONIUM EXALATES

DOEpatents

Garner, C.S.

1959-02-24

A process is presented for converting plutonium oxalate to other plutonium compounds by a dry conversion method. According to the process, lower valence plutonium oxalate is heated in the presence of a vapor of a volatile non- oxygenated monobasic acid, such as HCl or HF. For example, in order to produce plutonium chloride, the pure plutonium oxalate is heated to about 700 deg C in a slow stream of hydrogen plus HCl. By the proper selection of an oxidizing or reducing atmosphere, the plutonium halide product can be obtained in either the plus 3 or plus 4 valence state.

Radioisotope contaminations from releases of the Tomsk-Seversk nuclear facility (Siberia, Russia).

PubMed

Gauthier-Lafaye, F; Pourcelot, L; Eikenberg, J; Beer, H; Le Roux, G; Rhikvanov, L P; Stille, P; Renaud, Ph; Mezhibor, A

2008-04-01

Soils have been sampled in the vicinity of the Tomsk-Seversk facility (Siberia, Russia) that allows us to measure radioactive contaminations due to atmospheric and aquatic releases. Indeed soils exhibit large inventories of man-made fission products including 137Cs (ranging from 33,000 to 68,500 Bq m(-2)) and actinides such as plutonium (i.e. 239+240Pu from 420 to 5900 Bq m(-2)) or 241Am (160-1220 Bq m(-2)). Among all sampling sites, the bank of the Romashka channel exhibits the highest radioisotope concentrations. At this site, some short half-life gamma emitters were detected as well indicating recent aquatic discharge in the channel. In comparison, soils that underwent atmospheric depositions like peat and forest soils exhibit lower activities of actinides and 137Cs. Soil activities are too high to be related solely to global fallout and thus the source of plutonium must be discharges from the Siberian Chemical Combine (SCC) plant. This is confirmed by plutonium isotopic ratios measured by ICP-MS; the low 241Pu/239Pu and 240Pu/239Pu atomic ratios with respect to global fallout ratio or civil nuclear fuel are consistent with weapons grade signatures. Up to now, the influence of Tomsk-Seversk plutonium discharges was speculated in the Ob River and its estuary. Isotopic data from the present study show that plutonium measured in SCC probably constitutes a significant source of plutonium in the aquatic environment, together with plutonium from global fallout and other contaminated sites including Tomsk, Mayak (Russia) and Semipalatinsk (Republic of Kazakhstan). It is estimated that the proportion of plutonium from SCC source can reach 45% for 239Pu and 60% for 241Pu in the sediments.

Nonproliferation and Threat Reduction Assistance: U.S, Programs in the Former Soviet Union

DTIC Science & Technology

2008-03-26

reconfigure its large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious diseases. For FY2004, the Bush...program to eliminate its plutonium, opting instead for the construction of fast breeder reactors that could burn plutonium directly for energy production...The United States might not fund this effort, as many in the United States argue that breeder reactors , which produce more plutonium than they

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerber, Mark A.; Schmidt, Andrew J.; Delegard, Calvin H.

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 quantifymore » the effects of key processing variables on processing time (primarily neutralization and filtration time) and calcined product properties.« less

13. SIDE VIEW OF THE STACKERRETRIEVER CRANE FROM THE TRANSFER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

13. SIDE VIEW OF THE STACKER-RETRIEVER CRANE FROM THE TRANSFER BAY. THE STACKER-RETRIEVER IS A REMOTELY-OPERATED, MECHANIZED TRANSPORT SYSTEM FOR RETRIEVING PLUTONIUM CONTAINERS FROM THE STORAGE VAULT. (1/80) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iseki, Tadahiro; Inaba, Makoto; Takahashi, Naoki

During the second and third steps of Active Test at Rokkasho Reprocessing Plant (RRP), the performances of the Separation Facility have been checked; (A) diluent washing efficiency, (B) plutonium stripping efficiency, (C) decontamination factor of fission products and (D) plutonium and uranium leakage into raffinate and spent solvent. Test results were equivalent to or better than expected. (authors)

Integrated approaches for reducing sample size for measurements of trace elemental impurities in plutonium by ICP-OES and ICP-MS

DOE PAGES

Xu, Ning; Chamberlin, Rebecca M.; Thompson, Pam; ...

2017-10-07

This study has demonstrated that bulk plutonium chemical analysis can be performed at small scales (\\50 mg material) through three case studies. Analytical methods were developed for ICP-OES and ICP-MS instruments to measure trace impurities and gallium content in plutonium metals with comparable or improved detection limits, measurement accuracy and precision. In two case studies, the sample size has been reduced by 109, and in the third case study, by as much as 50009, so that the plutonium chemical analysis can be performed in a facility rated for lower-hazard and lower-security operations.

Integrated approaches for reducing sample size for measurements of trace elemental impurities in plutonium by ICP-OES and ICP-MS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Ning; Chamberlin, Rebecca M.; Thompson, Pam

This study has demonstrated that bulk plutonium chemical analysis can be performed at small scales (\\50 mg material) through three case studies. Analytical methods were developed for ICP-OES and ICP-MS instruments to measure trace impurities and gallium content in plutonium metals with comparable or improved detection limits, measurement accuracy and precision. In two case studies, the sample size has been reduced by 109, and in the third case study, by as much as 50009, so that the plutonium chemical analysis can be performed in a facility rated for lower-hazard and lower-security operations.

Radiochemical Processing Laboratory (RPL) at PNNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peurrung, Tony; Clark, Sue; Bryan, Sam

2017-03-23

Nuclear research is one of the core components of PNNL's mission. The centerpiece of PNNL's nuclear research is the Radiochemical Processing Laboratory (RPL), a Category 2 nuclear facility with state-of-the-art instrumentation, scientific expertise, and specialized capabilities that enable research with significant quantities of fissionable materials and other radionuclides—from tritium to plutonium. High impact radiological research has been conducted in the RPL since the 1950's, when nuclear weapons and energy production at Hanford were at the forefront of national defense. Since then, significant investments have been made in the RPL to maintain it as a premier nuclear science research facility supportingmore » multiple programs. Most recently, PNNL is developing a world-class analytical electron microscopy facility dedicated to the characterization of radiological materials.« less

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

DOEpatents

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.

METHOD FOR OBTAINING PLUTONIUM METAL AND ALLOYS OF PLUTONIUM FROM PLUTONIUM TRICHLORIDE

DOEpatents

Reavis, J.G.; Leary, J.A.; Maraman, W.J.

1962-11-13

A process is given for both reducing plutonium trichloride to plutonium metal using cerium as the reductant and simultaneously alloying such plutonium metal with an excess of cerium or cerium and cobalt sufficient to yield the desired nuclear reactor fuel composition. The process is conducted at a temperature from about 550 to 775 deg C, at atmospheric pressure, without the use of booster reactants, and a substantial decontamination is effected in the product alloy of any rare earths which may be associated with the source of the plutonium. (AEC)

PROCESS OF ELIMINATING HYDROGEN PEROXIDE IN SOLUTIONS CONTAINING PLUTONIUM VALUES

DOEpatents

Barrick, J.G.; Fries, B.A.

1960-09-27

A procedure is given for peroxide precipitation processes for separating and recovering plutonium values contained in an aqueous solution. When plutonium peroxide is precipitated from an aqueous solution, the supernatant contains appreciable quantities of plutonium and peroxide. It is desirable to process this solution further to recover plutonium contained therein, but the presence of the peroxide introduces difficulties; residual hydrogen peroxide contained in the supernatant solution is eliminated by adding a nitrite or a sulfite to this solution.

Plutonium inventories for stabilization and stabilized materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, A.K.

1996-05-01

The objective of the breakout session was to identify characteristics of materials containing plutonium, the need to stabilize these materials for storage, and plans to accomplish the stabilization activities. All current stabilization activities are driven by the Defense Nuclear Facilities Safety Board Recommendation 94-1 (May 26, 1994) and by the recently completed Plutonium ES&H Vulnerability Assessment (DOE-EH-0415). The Implementation Plan for accomplishing stabilization of plutonium-bearing residues in response to the Recommendation and the Assessment was published by DOE on February 28, 1995. This Implementation Plan (IP) commits to stabilizing problem materials within 3 years, and stabilizing all other materials withinmore » 8 years. The IP identifies approximately 20 metric tons of plutonium requiring stabilization and/or repackaging. A further breakdown shows this material to consist of 8.5 metric tons of plutonium metal and alloys, 5.5 metric tons of plutonium as oxide, and 6 metric tons of plutonium as residues. Stabilization of the metal and oxide categories containing greater than 50 weight percent plutonium is covered by DOE Standard {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides{close_quotes} December, 1994 (DOE-STD-3013-94). This standard establishes criteria for safe storage of stabilized plutonium metals and oxides for up to 50 years. Each of the DOE sites and contractors with large plutonium inventories has either started or is preparing to start stabilization activities to meet these criteria.« less

Mortality among workers with chronic radiation sickness

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shilnikova, N.S.; Koshurnikova, N.A.; Bolotnikova, M.G.

1996-07-01

This study is based on a registry containing medical and dosimetric data of the employees who began working at different plants of the Mayak nuclear complex between 1948 and 1958 who developed chronic radiation sickness. Mayak is the first nuclear weapons plutonium production enterprise built in Russia and includes nuclear reactors, a radiochemical plant for plutonium separation, and a plutonium production enterprise built in Russia and includes nuclear reactors, a radiochemical plant for plutonium separation, and a plutonium production plant.Workers whose employment began between 1948 and 1958 exhibited a 6-28% incidence of chronic radiation sickness at the different facilities. Theremore » were no cases of chronic radiation sickness among those who began working after 1958. Data on doses of external whole-body gamma-irradiation and mortality in workers with chronic radiation sickness are presented. 6 refs., 5 tabs.« less

GrayQb TM Single-Faced Version 2 (SF2) Hanford Plutonium Reclamation Facility (PRF) deployment report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Plummer, J. R.; Immel, D. M.; Serrato, M. G.

2015-11-18

The Savannah River National Laboratory (SRNL) in partnership with CH2M Plateau Remediation Company (CHPRC) deployed the GrayQb TM SF2 radiation imaging device at the Hanford Plutonium Reclamation Facility (PRF) to assist in the radiological characterization of the canyon. The deployment goal was to locate radiological contamination hot spots in the PRF canyon, where pencil tanks were removed and decontamination/debris removal operations are on-going, to support the CHPRC facility decontamination and decommissioning (D&D) effort. The PRF canyon D&D effort supports completion of the CHPRC Plutonium Finishing Plant Decommissioning Project. The GrayQb TM SF2 (Single Faced Version 2) is a non-destructive examinationmore » device developed by SRNL to generate radiation contour maps showing source locations and relative radiological levels present in the area under examination. The Hanford PRF GrayQbTM Deployment was sponsored by CH2M Plateau Remediation Company (CHPRC) through the DOE Richland Operations Office, Inter-Entity Work Order (IEWO), DOE-RL IEWO- M0SR900210.« less

History of 232-F, tritium extraction processing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blackburn, G.W.

1994-08-01

In 1950 the Atomic Energy Commission authorized the Savannah River Project principally for the production of tritium and plutonium-239 for use in thermonuclear weapons. 232-F was built as an interim facility in 1953--1954, at a cost of $3.9M. Tritium extraction operations began in October, 1955, after the reactor and separations startups. In July, 1957 a larger tritium facility began operation in 232-H. In 1958 the capacity of 232-H was doubled. Also, in 1957 a new task was assigned to Savannah River, the loading of tritium into reservoirs that would be actual components of thermonuclear weapons. This report describes the historymore » of 232-F, the process for tritium extraction, and the lessons learned over the years that were eventually incorporated into the new Replacement Tritium Facility.« less

A Report to Congress on Long-Term Stewardship. Volume II, Site Summaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

2001-01-01

During World War II and the Cold War, the Federal government developed and operated a vast network of industrial facilities for the research, production, and testing of nuclear weapons, as well as for other scientific and engineering research. These processes left a legacy of radioactive and chemical waste, environmental contamination, and hazardous facilities and materials at well over a 100 sites in 30 States and one U.S. Territory. Hundreds of thousand of acres of residually contaminated soils, contaminated groundwater, surface water and sediment contamination, and contaminated buildings are present at many sites across the country. These sites range in sizemore » from less than one acre, containing only a single facility, to large sites spanning over 100,000 acres with huge uranium enrichment plants and plutonium processing canyons. Since 1989, the U.S. Department of Energy’s (DOE) Environmental Management (EM) program has made significant progress in addressing this environmental legacy. Millions of cubic meters of waste have been removed, stabilized, or disposed of, resulting in significant risk and cost reduction. In addition, DOE began disposing of transuranic (i.e., plutonium-contaminated) waste in the nation’s first deep geologic repository – the Waste Isolation Pilot Plant in New Mexico. DOE is now carrying out its long-term stewardship obligations at dozens of sites, including smaller sites where DOE has completed cleanup work for the entire site and many larger sites where DOE has remediated portions of the site.« less

2. VIEW OF THE GLOVE BOX WHERE, ON SEPTEMBER 11, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

2. VIEW OF THE GLOVE BOX WHERE, ON SEPTEMBER 11, 1957, A FIRE STARTED. THE FIRE SPREAD TO THE REST OF THE BUILDING, RESULTING IN THE TRANSFER OF PLUTONIUM FOUNDRY, FABRICATION, AND ASSEMBLY OPERATIONS TO BUILDING 776/777. (9/16/57) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

DOEpatents

Boyd, G.E.; Adamson, A.W.; Schubert, J.; Russell, E.R.

1958-10-01

A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This process provides a convenient and efficient means for isolating plutonium.

PROCESS FOR THE RECOVERY OF PLUTONIUM

DOEpatents

Ritter, D.M.

1959-01-13

An improvement is presented in the process for recovery and decontamination of plutonium. The carrier precipitate containing plutonium is dissolved and treated with an oxidizing agent to place the plutonium in a hexavalent oxidation state. A lanthanum fluoride precipitate is then formed in and removed from the solution to carry undesired fission products. The fluoride ions in the reniaining solution are complexed by addition of a borate sueh as boric acid, sodium metaborate or the like. The plutonium is then reduced and carried from the solution by the formation of a bismuth phosphate precipitate. This process effects a better separation from unwanted flssion products along with conccntration of the plutonium by using a smaller amount of carrier.

Facility effluent monitoring plan for the plutonium uranium extraction facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wiegand, D.L.

A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438-01. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan shall ensure long-range integrity of themore » effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated at a minimum of every three years.« less

PRECIPITATION OF PLUTONOUS PEROXIDE

DOEpatents

Barrick, J.G.; Manion, J.P.

1961-08-15

A precipitation process for recovering plutonium values contained in an aqueous solution is described. In the process for precipitating plutonium as plutonous peroxide, hydroxylamine or hydrazine is added to the plutoniumcontaining solution prior to the addition of peroxide to precipitate plutonium. The addition of hydroxylamine or hydrazine increases the amount of plutonium precipitated as plutonous peroxide. (AEC)

METHOD OF SEPARATING PLUTONIUM

DOEpatents

Brown, H.S.; Hill, O.F.

1958-02-01

Plutonium hexafluoride is a satisfactory fluorinating agent and may be reacted with various materials capable of forming fluorides, such as copper, iron, zinc, etc., with consequent formation of the metal fluoride and reduction of the plutonium to the form of a lower fluoride. In accordance with the present invention, it has been found that the reactivity of plutonium hexafluoride with other fluoridizable materials is so great that the process may be used as a method of separating plutonium from mixures containing plutonium hexafluoride and other vaporized fluorides even though the plutonium is present in but minute quantities. This process may be carried out by treating a mixture of fluoride vapors comprising plutonium hexafluoride and fluoride of uranium to selectively reduce the plutonium hexafluoride and convert it to a less volatile fluoride, and then recovering said less volatile fluoride from the vapor by condensation.

PROCESS OF PRODUCING SHAPED PLUTONIUM

DOEpatents

Anicetti, R.J.

1959-08-11

A process is presented for producing and casting high purity plutonium metal in one step from plutonium tetrafluoride. The process comprises heating a mixture of the plutonium tetrafluoride with calcium while the mixture is in contact with and defined as to shape by a material obtained by firing a mixture consisting of calcium oxide and from 2 to 10% by its weight of calcium fluoride at from 1260 to 1370 deg C.

Removal of the Plutonium Recycle Test Reactor - 13031

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herzog, C. Brad; Guercia, Rudolph; LaCome, Matt

2013-07-01

The 309 Facility housed the Plutonium Recycle Test Reactor (PRTR), an operating test reactor in the 300 Area at Hanford, Washington. The reactor first went critical in 1960 and was originally used for experiments under the Hanford Site Plutonium Fuels Utilization Program. The facility was decontaminated and decommissioned in 1988-1989, and the facility was deactivated in 1994. The 309 facility was added to Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) response actions as established in an Interim Record of Decision (IROD) and Action Memorandum (AM). The IROD directs a remedial action for the 309 facility, associated waste sites, associatedmore » underground piping and contaminated soils resulting from past unplanned releases. The AM directs a removal action through physical demolition of the facility, including removal of the reactor. Both CERCLA actions are implemented in accordance with U.S. EPA approved Remedial Action Work Plan, and the Remedial Design Report / Remedial Action Report associated with the Hanford 300-FF-2 Operable Unit. The selected method for remedy was to conventionally demolish above grade structures including the easily distinguished containment vessel dome, remove the PRTR and a minimum of 300 mm (12 in) of shielding as a single 560 Ton unit, and conventionally demolish the below grade structure. Initial sample core drilling in the Bio-Shield for radiological surveys showed evidence that the Bio-Shield was of sound structure. Core drills for the separation process of the PRTR from the 309 structure began at the deck level and revealed substantial thermal degradation of at least the top 1.2 m (4LF) of Bio-Shield structure. The degraded structure combined with the original materials used in the Bio-Shield would not allow for a stable structure to be extracted. The water used in the core drilling process proved to erode the sand mixture of the Bio-Shield leaving the steel aggregate to act as ball bearings against the core drill bit. A redesign is being completed to extract the 309 PRTR and entire Bio-Shield structure together as one monolith weighing 1100 Ton by cutting structural concrete supports. In addition, the PRTR has hundreds of contaminated process tubes and pipes that have to be severed to allow for a uniformly flush fit with a lower lifting frame. Thirty-two 50 mm (2 in) core drills must be connected with thirty-two wire saw cuts to allow for lifting columns to be inserted. Then eight primary saw cuts must be completed to severe the PRTR from the 309 Facility. Once the weight of the PRTR is transferred to the lifting frame, then the PRTR may be lifted out of the facility. The critical lift will be executed using four 450 Ton strand jacks mounted on a 9 m (30 LF) tall mobile lifting frame that will allow the PRTR to be transported by eight 600 mm (24 in) Slide Shoes. The PRTR will then be placed on a twenty-four line, double wide, self powered Goldhofer for transfer to the onsite CERCLA Disposal Cell (ERDF Facility), approximately 33 km (20 miles) away. (authors)« less

CONVERSION OF PLUTONIUM TRIFLUORIDE TO PLUTONIUM TETRAFLUORIDE

DOEpatents

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.

PROCESSES FOR SEPARATING AND RECOVERING CONSTITUENTS OF NEUTRON IRRADIATED URANIUM

DOEpatents

Connick, R.E.; Gofman, J.W.; Pimentel, G.C.

1959-11-10

Processes are described for preparing plutonium, particularly processes of separating plutonium from uranium and fission products in neutron-irradiated uraniumcontaining matter. Specifically, plutonium solutions containing uranium, fission products and other impurities are contacted with reducing agents such as sulfur dioxide, uranous ion, hydroxyl ammonium chloride, hydrogen peroxide, and ferrous ion whereby the plutoninm is reduced to its fluoride-insoluble state. The reduced plutonium is then carried out of solution by precipitating niobic oxide therein. Uranium and certain fission products remain behind in the solution. Certain other fission products precipitate along with the plutonium. Subsequently, the plutonium and fission product precipitates are redissolved, and the solution is oxidized with oxidizing agents such as chlorine, peroxydisulfate ion in the presence of silver ion, permanganate ion, dichromate ion, ceric ion, and a bromate ion, whereby plutonium is oxidized to the fluoride-soluble state. The oxidized solution is once again treated with niobic oxide, thus precipitating the contamirant fission products along with the niobic oxide while the oxidized plutonium remains in solution. Plutonium is then recovered from the decontaminated solution.

5. VIEW OF A GLOVE BOX FIREWALL DETAIL. THE FIREWALL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

5. VIEW OF A GLOVE BOX FIREWALL DETAIL. THE FIREWALL WAS A SAFETY FEATURE TO PREVENT THE SPREAD OF FIRE BETWEEN INTERCONNECTED GLOVE BOXES. PLUTONIUM IS PYROPHORIC, AND MAY IGNITE IN THE PRESENCE OF OXYGEN. (5/8/70) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

PROCESS OF SEPARATING PLUTONIUM FROM URANIUM

DOEpatents

Brown, H.S.; Hill, O.F.

1958-09-01

A process is presented for recovering plutonium values from aqueous solutions. It comprises forming a uranous hydroxide precipitate in such a plutonium bearing solution, at a pH of at least 5. The plutonium values are precipitated with and carried by the uranium hydroxide. The carrier precipitate is then redissolved in acid solution and the pH is adjusted to about 2.5, causing precipitation of the uranous hydroxide but leaving the still soluble plutonium values in solution.

COLUMBIC OXIDE ADSORPTION PROCESS FOR SEPARATING URANIUM AND PLUTONIUM IONS

DOEpatents

Beaton, R.H.

1959-07-14

A process is described for separating plutonium ions from a solution of neutron irradiated uranium in which columbic oxide is used as an adsorbert. According to the invention the plutonium ion is selectively adsorbed by Passing a solution containing the plutonium in a valence state not higher than 4 through a porous bed or column of granules of hydrated columbic oxide. The adsorbed plutonium is then desorbed by elution with 3 N nitric acid.

PROCESS OF SEPARATING PLUTONIUM VALUES BY ELECTRODEPOSITION

DOEpatents

Whal, A.C.

1958-04-15

A process is described of separating plutonium values from an aqueous solution by electrodeposition. The process consists of subjecting an aqueous 0.1 to 1.0 N nitric acid solution containing plutonium ions to electrolysis between inert metallic electrodes. A current density of one milliampere io one ampere per square centimeter of cathode surface and a temperature between 10 and 60 d C are maintained. Plutonium is electrodeposited on the cathode surface and recovered.

Experience gained with the Synroc demonstration plant at ANSTO and its relevance to plutonium immobilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jostsons, A.; Ridal, A.; Mercer, D.J.

1996-05-01

The Synroc Demonstration Plant (SDP) was designed and constructed at Lucas Heights to demonstrate the feasibility of Synroc production on a commercial scale (10 kg/hr) with simulated Purex liquid HLW. Since commissioning of the SDP in 1987, over 6000 kg of Synroc has been fabricated with a range of feeds and waste loadings. The SDP utilises uniaxial hot-pressing to consolidate Synroc. Pressureless sintering and hot-isostatic pressing have also been studied at smaller scales. The results of this extensive process development have been incorporated in a conceptual design for a radioactive plant to condition HLW from a reprocessing plant with amore » capacity to treat 800 tpa of spent LWR fuel. Synroic containing TRU, including Pu, and fission products has been fabricated and characterised in a glove-box facility and hot cells, respectively. The extensive experience in processing of Synroc over the past 15 years is summarised and its relevance to immobilization of surplus plutonium is discussed.« less

Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Plutonium Metals, Oxides, and Solutions on the High Performance Computing Platform Moonlight

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chapman, Bryan Scott; Gough, Sean T.

This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.

Characterization of Representative Materials in Support of Safe, Long Term Storage of Surplus Plutonium in DOE-STD-3013 Containers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Narlesky, Joshua E.; Stroud, Mary Ann; Smith, Paul Herrick

2013-02-15

The Surveillance and Monitoring Program is a joint Los Alamos National Laboratory/Savannah River Site effort funded by the Department of Energy-Environmental Management to provide the technical basis for the safe, long-term storage (up to 50 years) of over 6 metric tons of plutonium stored in over 5,000 DOE-STD-3013 containers at various facilities around the DOE complex. The majority of this material is plutonium that is surplus to the nuclear weapons program, and much of it is destined for conversion to mixed oxide fuel for use in US nuclear power plants. The form of the plutonium ranges from relatively pure metalmore » and oxide to very impure oxide. The performance of the 3013 containers has been shown to depend on moisture content and on the levels, types and chemical forms of the impurities. The oxide materials that present the greatest challenge to the storage container are those that contain chloride salts. Other common impurities include oxides and other compounds of calcium, magnesium, iron, and nickel. Over the past 15 years the program has collected a large body of experimental data on 54 samples of plutonium, with 53 chosen to represent the broader population of materials in storage. This paper summarizes the characterization data, moisture analysis, particle size, surface area, density, wattage, actinide composition, trace element impurity analysis, and shelf life surveillance data and includes origin and process history information. Limited characterization data on fourteen nonrepresentative samples is also presented.« less

Characterization of representative materials in support of safe, long term storage of surplus plutonium in DOE-STD-3013 containers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Paul H; Narlesky, Joshua E; Worl, Laura A

2010-01-01

The Surveillance and Monitoring Program (SMP) is a joint LANL/SRS effort funded by DOE/EM to provide the technical basis for the safe, long-term storage (up to 50 years) of over 6 metric tons of plutonium stored in over 5000 DOE-STD-3013 containers at various facilities around the DOE complex. The majority of this material is plutonium that is surplus to the nuclear weapons program, and much of it is destined for conversion to mixed oxide fuel for use in US nuclear power plants. The form of the plutonium ranges from relatively pure metal and oxide to very impure oxide. The performancemore » of the 3013 containers has been shown to depend on moisture content and on the levels, types and chemical forms of the impurities. The oxide materials that present the greatest challenge to the storage container are those that contain chloride salts. The chlorides (NaCl, KCl, CaCl{sub 2}, and MgCl{sub 2}) range from less than half of the impurities present to nearly all the impurities. Other common impurities include oxides and other compounds of calcium, magnesium, iron, and nickel. Over the past 15 years the program has collected a large body of experimental data on over 60 samples of plutonium chosen to represent the broader population of materials in storage. This paper will summarize the characterization data, including the origin and process history, particle size, surface area, density, calorimetry, chemical analysis, moisture analysis, prompt gamma, gas generation and corrosion behavior.« less

ADSORPTION-BISMUTH PHOSPHATE METHOD FOR SEPARATING PLUTONIUM

DOEpatents

Russell, E.R.; Adamson, A.W.; Boyd, G.E.

1960-06-28

A process is given for separating plutonium from uranium and fission products. Plutonium and uranium are adsorbed by a cation exchange resin, plutonium is eluted from the adsorbent, and then, after oxidation to the hexavalent state, the plutonium is contacted with a bismuth phosphate carrier precipitate.

PLUTONIUM-HYDROGEN REACTION PRODUCT, METHOD OF PREPARING SAME AND PLUTONIUM POWDER THEREFROM

DOEpatents

Fried, S.; Baumbach, H.L.

1959-12-01

A process is described for forming plutonlum hydride powder by reacting hydrogen with massive plutonium metal at room temperature and the product obtained. The plutonium hydride powder can be converted to plutonium powder by heating to above 200 deg C.

PRODUCTION OF PLUTONIUM METAL

DOEpatents

Lyon, W.L.; Moore, R.H.

1961-01-17

A process is given for producing plutonium metal by the reduction of plutonium chloride, dissolved in alkali metal chloride plus or minus aluminum chloride, with magnesium or a magnesium-aluminum alloy at between 700 and 800 deg C and separating the plutonium or plutonium-aluminum alloy formed from the salt.

Author Contribution to the Pu Handbook II: Chapter 37 LLNL Integrated Sample Preparation Glovebox (TEM) Section

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wall, Mark A.

The development of our Integrated Actinide Sample Preparation Laboratory (IASPL) commenced in 1998 driven by the need to perform transmission electron microscopy studies on naturally aged plutonium and its alloys looking for the microstructural effects of the radiological decay process (1). Remodeling and construction of a laboratory within the Chemistry and Materials Science Directorate facilities at LLNL was required to turn a standard radiological laboratory into a Radiological Materials Area (RMA) and Radiological Buffer Area (RBA) containing type I, II and III workplaces. Two inert atmosphere dry-train glove boxes with antechambers and entry/exit fumehoods (Figure 1), having a baseline atmospheremore » of 1 ppm oxygen and 1 ppm water vapor, a utility fumehood and a portable, and a third double-walled enclosure have been installed and commissioned. These capabilities, along with highly trained technical staff, facilitate the safe operation of sample preparation processes and instrumentation, and sample handling while minimizing oxidation or corrosion of the plutonium. In addition, we are currently developing the capability to safely transfer small metallographically prepared samples to a mini-SEM for microstructural imaging and chemical analysis. The gloveboxes continue to be the most crucial element of the laboratory allowing nearly oxide-free sample preparation for a wide variety of LLNL-based characterization experiments, which includes transmission electron microscopy, electron energy loss spectroscopy, optical microscopy, electrical resistivity, ion implantation, X-ray diffraction and absorption, magnetometry, metrological surface measurements, high-pressure diamond anvil cell equation-of-state, phonon dispersion measurements, X-ray absorption and emission spectroscopy, and differential scanning calorimetry. The sample preparation and materials processing capabilities in the IASPL have also facilitated experimentation at world-class facilities such as the Advanced Photon Source at Argonne National Laboratory, the European Synchrotron Radiation Facility in Grenoble, France, the Stanford Synchrotron Radiation Facility, the National Synchrotron Light Source at Brookhaven National Laboratory, the Advanced Light Source at Lawrence Berkeley National Laboratory, and the Triumph Accelerator in Canada.« less

ION EXCHANGE ADSORPTION PROCESS FOR PLUTONIUM SEPARATION

DOEpatents

Boyd, G.E.; Russell, E.R.; Taylor, M.D.

1961-07-11

Ion exchange processes for the separation of plutonium from fission products are described. In accordance with these processes an aqueous solution containing plutonium and fission products is contacted with a cation exchange resin under conditions favoring adsorption of plutonium and fission products on the resin. A portion of the fission product is then eluted with a solution containing 0.05 to 1% by weight of a carboxylic acid. Plutonium is next eluted with a solution containing 2 to 8 per cent by weight of the same carboxylic acid, and the remaining fission products on the resin are eluted with an aqueous solution containing over 10 per cent by weight of sodium bisulfate.

The Fireball integrated code package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dobranich, D.; Powers, D.A.; Harper, F.T.

1997-07-01

Many deep-space satellites contain a plutonium heat source. An explosion, during launch, of a rocket carrying such a satellite offers the potential for the release of some of the plutonium. The fireball following such an explosion exposes any released plutonium to a high-temperature chemically-reactive environment. Vaporization, condensation, and agglomeration processes can alter the distribution of plutonium-bearing particles. The Fireball code package simulates the integrated response of the physical and chemical processes occurring in a fireball and the effect these processes have on the plutonium-bearing particle distribution. This integrated treatment of multiple phenomena represents a significant improvement in the state ofmore » the art for fireball simulations. Preliminary simulations of launch-second scenarios indicate: (1) most plutonium vaporization occurs within the first second of the fireball; (2) large non-aerosol-sized particles contribute very little to plutonium vapor production; (3) vaporization and both homogeneous and heterogeneous condensation occur simultaneously; (4) homogeneous condensation transports plutonium down to the smallest-particle sizes; (5) heterogeneous condensation precludes homogeneous condensation if sufficient condensation sites are available; and (6) agglomeration produces larger-sized particles but slows rapidly as the fireball grows.« less

Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 andmore » Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.« less

SEPARATION OF PLUTONIUM FROM FISSION PRODUCTS BY A COLLOID REMOVAL PROCESS

DOEpatents

Schubert, J.

1960-05-24

A method is given for separating plutonium from uranium fission products. An acidic aqueous solution containing plutonium and uranium fission products is subjected to a process for separating ionic values from colloidal matter suspended therein while the pH of the solution is maintained between 0 and 4. Certain of the fission products, and in particular, zirconium, niobium, lanthanum, and barium are in a colloidal state within this pH range, while plutonium remains in an ionic form, Dialysis, ultracontrifugation, and ultrafiltration are suitable methods of separating plutonium ions from the colloids.

Chemical Disposition of Plutonium in Hanford Site Tank Wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Delegard, Calvin H.; Jones, Susan A.

2015-05-07

This report examines the chemical disposition of plutonium (Pu) in Hanford Site tank wastes, by itself and in its observed and potential interactions with the neutron absorbers aluminum (Al), cadmium (Cd), chromium (Cr), iron (Fe), manganese (Mn), nickel (Ni), and sodium (Na). Consideration also is given to the interactions of plutonium with uranium (U). No consideration of the disposition of uranium itself as an element with fissile isotopes is considered except tangentially with respect to its interaction as an absorber for plutonium. The report begins with a brief review of Hanford Site plutonium processes, examining the various means used tomore » recover plutonium from irradiated fuel and from scrap, and also examines the intermediate processing of plutonium to prepare useful chemical forms. The paper provides an overview of Hanford tank defined-waste–type compositions and some calculations of the ratios of plutonium to absorber elements in these waste types and in individual waste analyses. These assessments are based on Hanford tank waste inventory data derived from separately published, expert assessments of tank disposal records, process flowsheets, and chemical/radiochemical analyses. This work also investigates the distribution and expected speciation of plutonium in tank waste solution and solid phases. For the solid phases, both pure plutonium compounds and plutonium interactions with absorber elements are considered. These assessments of plutonium chemistry are based largely on analyses of idealized or simulated tank waste or strongly alkaline systems. The very limited information available on plutonium behavior, disposition, and speciation in genuine tank waste also is discussed. The assessments show that plutonium coprecipitates strongly with chromium, iron, manganese and uranium absorbers. Plutonium’s chemical interactions with aluminum, nickel, and sodium are minimal to non-existent. Credit for neutronic interaction of plutonium with these absorbers occurs only if they are physically proximal in solution or the plutonium present in the solid phase is intimately mixed with compounds or solutions of these absorbers. No information on the potential chemical interaction of plutonium with cadmium was found in the technical literature. Definitive evidence of sorption or adsorption of plutonium onto various solid phases from strongly alkaline media is less clear-cut, perhaps owing to fewer studies and to some well-attributed tests run under conditions exceeding the very low solubility of plutonium. The several studies that are well-founded show that only about half of the plutonium is adsorbed from waste solutions onto sludge solid phases. The organic complexants found in many Hanford tank waste solutions seem to decrease plutonium uptake onto solids. A number of studies show plutonium sorbs effectively onto sodium titanate. Finally, this report presents findings describing the behavior of plutonium vis-à-vis other elements during sludge dissolution in nitric acid based on Hanford tank waste experience gained by lab-scale tests, chemical and radiochemical sample characterization, and full-scale processing in preparation for strontium-90 recovery from PUREX sludges.« less

PLUTONIUM-CUPFERRON COMPLEX AND METHOD OF REMOVING PLUTONIUM FROM SOLUTION

DOEpatents

Potratz, H.A.

1959-01-13

A method is presented for separating plutonium from fission products present in solutions of neutronirradiated uranium. The process consists in treating such acidic solutions with cupferron so that the cupferron reacts with the plutonium present to form an insoluble complex. This plutonium cupferride precipitates and may then be separated from the solution.

EXAFS/XANES studies of plutonium-loaded sodalite/glass waste forms

NASA Astrophysics Data System (ADS)

Richmann, Michael K.; Reed, Donald T.; Kropf, A. Jeremy; Aase, Scott B.; Lewis, Michele A.

2001-09-01

A sodalite/glass ceramic waste form is being developed to immobilize highly radioactive nuclear wastes in chloride form, as part of an electrochemical cleanup process. Two types of simulated waste forms were studied: where the plutonium was alone in an LiCl/KCl matrix and where simulated fission-product elements were added representative of the electrometallurgical treatment process used to recover uranium from spent nuclear fuel also containing plutonium and a variety of fission products. Extended X-ray absorption fine structure spectroscopy (EXAFS) and X-ray absorption near-edge spectroscopy (XANES) studies were performed to determine the location, oxidation state, and particle size of the plutonium within these waste form samples. Plutonium was found to segregate as plutonium(IV) oxide with a crystallite size of at least 4.8 nm in the non-fission-element case and 1.3 nm with fission elements present. No plutonium was observed within the sodalite in the waste form made from the plutonium-loaded LiCl/KCl eutectic salt. Up to 35% of the plutonium in the waste form made from the plutonium-loaded simulated fission-product salt may be segregated with a heavy-element nearest neighbor other than plutonium or occluded internally within the sodalite lattice.

TA-55 and Sigma Overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spearing, Dane Robert

These are slides from a facility overview presentation for visiting agencies to Los Alamos National Laboratory (LANL). The TA-55 Plutonium Facility (PF-4) is discussed in detail. PF-4 is a unique resource for US plutonium programs. The basic design is flexible and has adapted to changing national needs. It is a robust facility with strong safety and security implementation. It supports a variety of national programs. It will continue for many years into the future. Sigma is then discussed in detail, which handles everything from hydrogen to uranium. It has been in long term service to the Nation (nearly 60 years).more » It has a flexible authorization basis to handle almost the entire periodic table. It has a wide breadth of prototyping and characterization capabilities. It has integrated program and line management.« less

CONCENTRATION PROCESS FOR PLUTONIUM IONS, IN AN OXIDATION STATE NOT GREATER THAN +4, IN AQUEOUS ACID SOLUTION

DOEpatents

Seaborg, G.T.; Thompson, S.G.

1960-06-14

A process for concentrating plutonium is given in which plutonium is first precipitated with bismuth phosphate and then, after redissolution, precipitated with a different carrier such as lanthanum fluoride, uranium acetate, bismuth hydroxide, or niobic oxide.

METHOD FOR RECOVERING PLUTONIUM VALUES FROM SOLUTION USING A BISMUTH HYDROXIDE CARRIER PRECIPITATE

DOEpatents

Faris, B.F.

1961-04-25

Carrier precipitation processes for separating plutonium values from aqueous solutions are described. In accordance with the invention a bismuth hydroxide precipitate is formed in the plutonium-containing solution, thereby carrying plutonium values from the solution.

PROCESS OF TREATING OR FORMING AN INSOLUBLE PLUTONIUM PRECIPITATE IN THE PRESENCE OF AN ORGANIC ACTIVE AGENT

DOEpatents

Balthis, J.H.

1961-07-18

Carrier precipitation processes for the separation of plutonium from fission products are described. In a process in which an insoluble precipitate is formed in a solution containing plutonium and fission products under conditions whereby plutonium is carried by the precipitate, and the precipitate is then separated from the remaining solution, an organic surface active agent is added to the mixture of precipitate and solution prior to separation of the precipitate from the supernatant solution, thereby improving the degree of separation of the precipitate from the solution.

PROCESS USING BISMUTH PHOSPHATE AS A CARRIER PRECIPITATE FOR FISSION PRODUCTS AND PLUTONIUM VALUES

DOEpatents

Finzel, T.G.

1959-03-10

A process is described for separating plutonium from fission products carried therewith when plutonium in the reduced oxidation state is removed from a nitric acid solution of irradiated uranium by means of bismuth phosphate as a carrier precipitate. The bismuth phosphate carrier precipitate is dissolved by treatment with nitric acid and the plutonium therein is oxidized to the hexavalent oxidation state by means of potassium dichromate. Separation of the plutonium from the fission products is accomplished by again precipitating bismuth phosphate and removing the precipitate which now carries the fission products and a small percentage of the plutonium present. The amount of plutonium carried in this last step may be minimized by addition of sodium fluoride, so as to make the solution 0.03N in NaF, prior to the oxidation and prccipitation step.

PROCESS OF TREATING URANIUM HEXAFLUORIDE AND PLUTONIUM HEXAFLUORIDE MIXTURES WITH SULFUR TETRAFLUORIDE TO SEPARATE SAME

DOEpatents

Steindler, M.J.

1962-07-24

A process was developed for separating uranium hexafluoride from plutonium hexafluoride by the selective reduction of the plutonium hexafluoride to the tetrafluoride with sulfur tetrafluoride at 50 to 120 deg C, cooling the mixture to --60 to -100 deg C, and volatilizing nonreacted sulfur tetrafluoride and sulfur hexafluoride formed at that temperature. The uranium hexafluoride is volatilized at room temperature away from the solid plutonium tetrafluoride. (AEC)

PROCESS OF REMOVING PLUTONIUM VALUES FROM SOLUTION WITH GROUP IVB METAL PHOSPHO-SILICATE COMPOSITIONS

DOEpatents

Russell, E.R.; Adamson, A.W.; Schubert, J.; Boyd, G.E.

1957-10-29

A process for separating plutonium values from aqueous solutions which contain the plutonium in minute concentrations is described. These values can be removed from an aqueous solution by taking an aqueous solution containing a salt of zirconium, titanium, hafnium or thorium, adding an aqueous solution of silicate and phosphoric acid anions to the metal salt solution, and separating, washing and drying the precipitate which forms when the two solutions are mixed. The aqueous plutonium containing solution is then acidified and passed over the above described precipi-tate causing the plutonium values to be adsorbed by the precipitate.

APPLICATION OF VACUUM SALT DISTILLATION TECHNOLOGY FOR THE REMOVAL OF FLUORIDE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pierce, R.; Pak, D.

2011-08-10

Vacuum distillation of chloride salts from plutonium oxide (PuO{sub 2}) and simulant PuO{sub 2} has been previously demonstrated at Department of Energy (DOE) sites using kilogram quantities of chloride salt. The apparatus for vacuum distillation contains a zone heated using a furnace and a zone actively cooled using either recirculated water or compressed air. During a vacuum distillation operation, a sample boat containing the feed material is placed into the apparatus while it is cool, and the system is sealed. The system is evacuated using a vacuum pump. Once a sufficient vacuum is attained, heating begins. Volatile salts distill frommore » the heated zone to the cooled zone where they condense, leaving behind the non-volatile materials in the feed boat. The application of vacuum salt distillation (VSD) is of interest to the HB-Line Facility and the MOX Fuel Fabrication Facility (MFFF) at the Savannah River Site (SRS). Both facilities are involved in efforts to disposition excess fissile materials. Many of these materials contain chloride and fluoride salt concentrations which make them unsuitable for dissolution without prior removal of the chloride and fluoride salts. Between September 2009 and January 2011, the Savannah River National Laboratory (SRNL) and HB-Line designed, developed, tested, and successfully deployed a system for the distillation of chloride salts. Subsequent efforts are attempting to adapt the technology for the removal of fluoride. Fluoride salts of interest are less-volatile than the corresponding chloride salts. Consequently, an alternate approach is required for the removal of fluoride without significantly increasing the operating temperature. HB-Line Engineering requested SRNL to evaluate and demonstrate the feasibility of an alternate approach using both non-radioactive simulants and plutonium-bearing materials. Whereas the earlier developments targeted the removal of sodium chloride (NaCl) and potassium chloride (KCl), the current activities are concerned with the removal of the halide ions associated with plutonium trifluoride (PuF{sub 3}), plutonium tetrafluoride (PuF{sub 4}), calcium fluoride (CaF{sub 2}), and calcium chloride (CaCl{sub 2}). This report discusses non-radioactive testing of small-scale and pilot-scale systems and radioactive testing of a small-scale system. Experiments focused on demonstrating the chemistry for halide removal and addressing the primary engineering questions associated with a change in the process chemistry.« less

PLUTONIUM CLEANING PROCESS

DOEpatents

Kolodney, M.

1959-12-01

A method is described for rapidly removing iron, nickel, and zinc coatings from plutonium objects while simultaneously rendering the plutonium object passive. The method consists of immersing the coated plutonium object in an aqueous acid solution containing a substantial concentration of nitrate ions, such as fuming nitric acid.

Safeguards and retrievability from waste forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Danker, W.

1996-05-01

This report describes issues discussed at a session from the PLutonium Stabilization and Immobilization Workshop related to safeguards and retrievability from waste forms. Throughout the discussion, the group probed the goals of disposition efforts, particularly an understanding of the {open_quotes}spent fuel standard{close_quotes}, since the disposition material form derives from these goals. The group felt strongly that not only the disposition goals but safeguards to meet these goals could affect the material form. Accordingly, the Department was encouraged to explore and apply safeguards as early in the implementation process as possible. It was emphasized that this was particularly true for anymore » planned use of existing facilities. It is much easier to build safeguards approaches into the development of new facilities, than to backfit existing facilities. Accordingly, special safeguards challenges are likely to be encountered, given the cost and schedule advantages offered by use of existing facilities.« less

STRIPPING PROCESS FOR PLUTONIUM

DOEpatents

Kolodney, M.

1959-10-01

A method for removing silver, nickel, cadmium, zinc, and indium coatings from plutonium objects while simultaneously rendering the plutonium object passive is described. The coated plutonium object is immersed as the anode in an electrolyte in which the plutonium is passive and the coating metal is not passive, using as a cathode a metal which does not dissolve rapidly in the electrolyte. and passing an electrical current through the electrolyte until the coating metal is removed from the plutonium body.

20. VIEW OF THE INTERIOR OF THE ADVANCED SIZE REDUCTION ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

20. VIEW OF THE INTERIOR OF THE ADVANCED SIZE REDUCTION FACILITY USED TO CUT PLUTONIUM CONTAMINATED GLOVE BOXES AND MISCELLANEOUS LARGE EQUIPMENT DOWN TO AN EASILY PACKAGED SIZE FOR DISPOSAL. ROUTINE OPERATIONS WERE PERFORMED REMOTELY, USING HOISTS, MANIPULATOR ARMS, AND GLOVE PORTS TO REDUCE BOTH INTENSITY AND TIME OF RADIATION EXPOSURE TO THE OPERATOR. (11/6/86) - Rocky Flats Plant, Plutonium Fabrication, Central section of Plant, Golden, Jefferson County, CO

Nonproliferation and Threat Reduction Assistance: U.S. Programs in the Former Soviet Union

DTIC Science & Technology

2009-07-31

seeks to help Russia reconfigure its large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious...opting instead for the construction of fast breeder reactors that could burn plutonium directly for energy production. The United States might not fund...this effort, as many in the United States argue that breeder reactors , which produce more plutonium than they consume, would undermine

Nonproliferation and Threat Reduction Assistance: U.S. Programs in the Former Soviet Union

DTIC Science & Technology

2011-04-26

large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious diseases. The Global Threat Reduction...indicated that it may not pursue the MOX program to eliminate its plutonium, opting instead for the construction of fast breeder reactors that could...burn plutonium directly for energy production. The United States might not fund this effort, as many in the United States argue that breeder reactors

Improvement of INVS Measurement Uncertainty for Pu and U-Pu Nitrate Solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn Thomas; Menlove, Howard Olsen; Marlow, Johnna Boulds

2017-04-27

In the Tokai Reprocessing Plant (TRP) and the Plutonium Conversion Development Facility (PCDF), a large amount of plutonium nitrate solution which is recovered from light water reactor (LWR) and advanced thermal reactor (ATR), FUGEN are being stored. Since the solution is designated as a direct use material, the periodical inventory verification and flow verification are being conducted by Japan Safeguard Government Office (JSGO) and International Atomic Agency (IAEA).

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

Verification study of an emerging fire suppression system

DOE PAGES

Cournoyer, Michael E.; Waked, R. Ryan; Granzow, Howard N.; ...

2016-01-01

Self-contained fire extinguishers are a robust, reliable and minimally invasive means of fire suppression for gloveboxes. Moreover, plutonium gloveboxes present harsh environmental conditions for polymer materials; these include radiation damage and chemical exposure, both of which tend to degrade the lifetime of engineered polymer components. Several studies have been conducted to determine the robustness of selfcontained fire extinguishers in plutonium gloveboxes in a nuclear facility, verification tests must be performed. These tests include activation and mass loss calorimeter tests. In addition, compatibility issues with chemical components of the self-contained fire extinguishers need to be addressed. Our study presents activation andmore » mass loss calorimeter test results. After extensive studies, no critical areas of concern have been identified for the plutonium glovebox application of Fire Foe™, except for glovebox operations that use large quantities of bulk plutonium or uranium metal such as metal casting and pyro-chemistry operations.« less

Verification study of an emerging fire suppression system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cournoyer, Michael E.; Waked, R. Ryan; Granzow, Howard N.

Self-contained fire extinguishers are a robust, reliable and minimally invasive means of fire suppression for gloveboxes. Moreover, plutonium gloveboxes present harsh environmental conditions for polymer materials; these include radiation damage and chemical exposure, both of which tend to degrade the lifetime of engineered polymer components. Several studies have been conducted to determine the robustness of selfcontained fire extinguishers in plutonium gloveboxes in a nuclear facility, verification tests must be performed. These tests include activation and mass loss calorimeter tests. In addition, compatibility issues with chemical components of the self-contained fire extinguishers need to be addressed. Our study presents activation andmore » mass loss calorimeter test results. After extensive studies, no critical areas of concern have been identified for the plutonium glovebox application of Fire Foe™, except for glovebox operations that use large quantities of bulk plutonium or uranium metal such as metal casting and pyro-chemistry operations.« less

PROCESS OF SECURING PLUTONIUM IN NITRIC ACID SOLUTIONS IN ITS TRIVALENT OXIDATION STATE

DOEpatents

Thomas, J.R.

1958-08-26

>Various processes for the recovery of plutonium require that the plutonium be obtalned and maintained in the reduced or trivalent state in solution. Ferrous ions are commonly used as the reducing agent for this purpose, but it is difficult to maintain the plutonium in a reduced state in nitric acid solutions due to the oxidizing effects of the acid. It has been found that the addition of a stabilizing or holding reductant to such solution prevents reoxidation of the plutonium. Sulfamate ions have been found to be ideally suitable as such a stabilizer even in the presence of nitric acid.

CONCENTRATION OF Pu USING AN IODATE PRECIPITATE

DOEpatents

Fries, B.A.

1960-02-23

A method is given for separating plutonium from lanthanum in a lanthanum fluoride carrier precipitation process for the recovery of plutonium values from an aqueous solution. The carrier precipitation process includes the steps of forming a lanthanum fluoride precipi- . tate, thereby carrying plutonium out of solution, metathesizing the fluoride precipitate to a hydroxide precipitate, and then dissolving the hydroxide precipitate in nitric acid. In accordance with the invention, the nitric acid solution, which contains plutonium and lanthanum, is made 0.05 to 0.15 molar in potassium iodate. thereby precipitating plutonium as plutonous iodate and the plutonous iodate is separated from the lanthanum- containing supernatant solution.

IMPROVED PROCESS OF PLUTONIUM CARRIER PRECIPITATION

DOEpatents

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 Oxide Containment and the Potential for Water-Borne Transport as a Consequence of ARIES Oxide Processing Operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wayne, David Matthew; Rowland, Joel C.

2015-02-01

The question of oxide containment during processing and storage has become a primary concern when considering the continued operability of the Plutonium Facility (PF-4) at Los Alamos National Laboratory (LANL). An Evaluation of the Safety of the Situation (ESS), “Potential for Criticality in a Glovebox Due to a Fire” (TA55-ESS-14-002-R2, since revised to R3) first issued in May, 2014 summarizes these concerns: “The safety issue of fire water potentially entering a glovebox is: the potential for the water to accumulate in the bottom of a glovebox and result in an inadvertent criticality due to the presence of fissionable materials inmore » the glovebox locations and the increased reflection and moderation of neutrons from the fire water accumulation.” As a result, the existing documented safety analysis (DSA) was judged inadequate and, while it explicitly considered the potential for criticality resulting from water intrusion into gloveboxes, criticality safety evaluation documents (CSEDs) for the affected locations did not evaluate the potential for fire water intrusion into a glovebox.« less

Volatile fluoride process for separating plutonium from other materials

DOEpatents

Spedding, F. H.; Newton, A. S.

1959-04-14

The separation of plutonium from uranium and/or fission products by formation of the higher fluorides off uranium and/or plutonium is described. Neutronirradiated uranium metal is first converted to the hydride. This hydrided product is then treated with fluorine at about 315 deg C to form and volatilize UF/sub 6/ leaving plutonium behind. Thc plutonium may then be separated by reacting the residue with fluorine at about 5004DEC and collecting the volatile plutonium fluoride thus formed.

VOLATILE FLUORIDE PROCESS FOR SEPARATING PLUTONIUM FROM OTHER MATERIALS

DOEpatents

Spedding, F.H.; Newton, A.S.

1959-04-14

The separation of plutonium from uranium and/or tission products by formation of the higher fluorides of uranium and/or plutonium is discussed. Neutronirradiated uranium metal is first convcrted to the hydride. This hydrided product is then treatced with fluorine at about 315 deg C to form and volatilize UF/sup 6/ leaving plutonium behind. The plutonium may then be separated by reacting the residue with fluorine at about 500 deg C and collecting the volatile plutonium fluoride thus formed.

Plutonium dissolution process

DOEpatents

Vest, Michael A.; Fink, Samuel D.; Karraker, David G.; Moore, Edwin N.; Holcomb, H. Perry

1996-01-01

A two-step process for dissolving plutonium metal, which two steps can be carried out sequentially or simultaneously. Plutonium metal is exposed to a first mixture containing approximately 1.0M-1.67M sulfamic acid and 0.0025M-0.1M fluoride, the mixture having been heated to a temperature between 45.degree. C. and 70.degree. C. The mixture will dissolve a first portion of the plutonium metal but leave a portion of the plutonium in an oxide residue. Then, a mineral acid and additional fluoride are added to dissolve the residue. Alteratively, nitric acid in a concentration between approximately 0.05M and 0.067M is added to the first mixture to dissolve the residue as it is produced. Hydrogen released during the dissolution process is diluted with nitrogen.

HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kyser, E. A.; King, W. D.

2012-07-31

Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Usemore » of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after ~4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.« less

HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kyser, E.; King, W.

2012-04-25

Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Usemore » of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after {approx}4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.« less

Selecting a plutonium vitrification process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jouan, A.

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 ofmore » 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.« less

The cement solidification systems at LANL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Veazey, G.W.

1990-01-01

There are two major cement solidification systems at Los Alamos National Laboratory. Both are focused primarily around treating waste from the evaporator at TA-55, the Plutonium Processing Facility. The evaporator receives the liquid waste stream from TA-55's nitric acid-based, aqueous-processing operations and concentrates the majority of the radionuclides in the evaporator bottoms solution. This is sent to the TA-55 cementation system. The evaporator distillate is sent to the TA-50 facility, where the radionuclides are precipitated and then cemented. Both systems treat TRU-level waste, and so are operated according to the criteria for WIPP-destined waste, but they differ in both cementmore » type and mixing method. The TA-55 systems uses Envirostone, a gypsum-based cement and in-drum prop mixing; the TA-50 systems uses Portland cement and drum tumbling for mixing.« less

REMOVAL OF LEGACY PLUTONIUM MATERIALS FROM SWEDEN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, Kerry A.; Bellamy, J. Steve; Chandler, Greg T.

2013-08-18

U.S. Department of Energy’s National Nuclear Security Administration (NNSA) Office of Global Threat Reduction (GTRI) recently removed legacy plutonium materials from Sweden in collaboration with AB SVAFO, Sweden. This paper details the activities undertaken through the U.S. receiving site (Savannah River Site (SRS)) to support the characterization, stabilization, packaging and removal of legacy plutonium materials from Sweden in 2012. This effort was undertaken as part of GTRI’s Gap Materials Program and culminated with the successful removal of plutonium from Sweden as announced at the 2012 Nuclear Security Summit. The removal and shipment of plutonium materials to the United States wasmore » the first of its kind under NNSA’s Global Threat Reduction Initiative. The Environmental Assessment for the U.S. receipt of gap plutonium material was approved in May 2010. Since then, the multi-year process yielded many first time accomplishments associated with plutonium packaging and transport activities including the application of the of DOE-STD-3013 stabilization requirements to treat plutonium materials outside the U.S., the development of an acceptance criteria for receipt of plutonium from a foreign country, the development and application of a versatile process flow sheet for the packaging of legacy plutonium materials, the identification of a plutonium container configuration, the first international certificate validation of the 9975 shipping package and the first intercontinental shipment using the 9975 shipping package. This paper will detail the technical considerations in developing the packaging process flow sheet, defining the key elements of the flow sheet and its implementation, determining the criteria used in the selection of the transport package, developing the technical basis for the package certificate amendment and the reviews with multiple licensing authorities and most importantly integrating the technical activities with the Swedish partners.« less

Plutonium

NASA Astrophysics Data System (ADS)

Clark, David L.; Hecker, Siegfried S.; Jarvinen, Gordon D.; Neu, Mary P.

The element plutonium occupies a unique place in the history of chemistry, physics, technology, and international relations. After the initial discovery based on submicrogram amounts, it is now generated by transmutation of uranium in nuclear reactors on a large scale, and has been separated in ton quantities in large industrial facilities. The intense interest in plutonium resulted fromthe dual-use scenario of domestic power production and nuclear weapons - drawing energy from an atomic nucleus that can produce a factor of millions in energy output relative to chemical energy sources. Indeed, within 5 years of its original synthesis, the primary use of plutonium was for the release of nuclear energy in weapons of unprecedented power, and it seemed that the new element might lead the human race to the brink of self-annihilation. Instead, it has forced the human race to govern itself without resorting to nuclear war over the past 60 years. Plutonium evokes the entire gamut of human emotions, from good to evil, from hope to despair, from the salvation of humanity to its utter destruction. There is no other element in the periodic table that has had such a profound impact on the consciousness of mankind.

PREPARATION OF PLUTONIUM TRIFLUORIDE

DOEpatents

Burger, L.L.; Roake, W.E.

1961-07-11

A process of producing plutonium trifluoride by reacting dry plutonium(IV) oxalate with chlorofluorinated methane or ethane at 400 to 450 deg C and cooling the product in the absence of oxygen is described.

3S (Safeguards, Security, Safety) based pyroprocessing facility safety evaluation plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ku, J.H.; Choung, W.M.; You, G.S.

The big advantage of pyroprocessing for the management of spent fuels against the conventional reprocessing technologies lies in its proliferation resistance since the pure plutonium cannot be separated from the spent fuel. The extracted materials can be directly used as metal fuel in a fast reactor, and pyroprocessing reduces drastically the volume and heat load of the spent fuel. KAERI has implemented the SBD (Safeguards-By-Design) concept in nuclear fuel cycle facilities. The goal of SBD is to integrate international safeguards into the entire facility design process since the very beginning of the design phase. This paper presents a safety evaluationmore » plan using a conceptual design of a reference pyroprocessing facility, in which 3S (Safeguards, Security, Safety)-By-Design (3SBD) concept is integrated from early conceptual design phase. The purpose of this paper is to establish an advanced pyroprocessing hot cell facility design concept based on 3SBD for the successful realization of pyroprocessing technology with enhanced safety and proliferation resistance.« less

Experimental and Numerical Investigations on Colloid-facilitated Plutonium Reactive Transport in Fractured Tuffaceous Rocks

NASA Astrophysics Data System (ADS)

Dai, Z.; Wolfsberg, A. V.; Zhu, L.; Reimus, P. W.

2017-12-01

Colloids have the potential to enhance mobility of strongly sorbing radionuclide contaminants in fractured rocks at underground nuclear test sites. This study presents an experimental and numerical investigation of colloid-facilitated plutonium reactive transport in fractured porous media for identifying plutonium sorption/filtration processes. The transport parameters for dispersion, diffusion, sorption, and filtration are estimated with inverse modeling for minimizing the least squares objective function of multicomponent concentration data from multiple transport experiments with the Shuffled Complex Evolution Metropolis (SCEM). Capitalizing on an unplanned experimental artifact that led to colloid formation and migration, we adopt a stepwise strategy to first interpret the data from each experiment separately and then to incorporate multiple experiments simultaneously to identify a suite of plutonium-colloid transport processes. Nonequilibrium or kinetic attachment and detachment of plutonium-colloid in fractures was clearly demonstrated and captured in the inverted modeling parameters along with estimates of the source plutonium fraction that formed plutonium-colloids. The results from this study provide valuable insights for understanding the transport mechanisms and environmental impacts of plutonium in fractured formations and groundwater aquifers.

PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM CONTAMINATING ELEMENTS

DOEpatents

Sutton, J.B.

1958-02-18

This patent relates to an improved method for the decontamination of plutonium. The process consists broadly in an improvement in a method for recovering plutonium from radioactive uranium fission products in aqueous solutions by decontamination steps including byproduct carrier precipitation comprising the step of introducing a preformed aqueous slurry of a hydroxide of a metal of group IV B into any aqueous acidic solution which contains the plutonium in the hexavalent state, radioactive uranium fission products contaminant and a by-product carrier precipitate and separating the metal hydroxide and by-product precipitate from the solution. The process of this invention is especially useful in the separation of plutonium from radioactive zirconium and columbium fission products.

Excess Weapons Plutonium Immobilization in Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jardine, L.; Borisov, G.B.

2000-04-15

The joint goal of the Russian work is to establish a full-scale plutonium immobilization facility at a Russian industrial site by 2005. To achieve this requires that the necessary engineering and technical basis be developed in these Russian projects and the needed Russian approvals be obtained to conduct industrial-scale immobilization of plutonium-containing materials at a Russian industrial site by the 2005 date. This meeting and future work will provide the basis for joint decisions. Supporting R&D projects are being carried out at Russian Institutes that directly support the technical needs of Russian industrial sites to immobilize plutonium-containing materials. Special R&Dmore » on plutonium materials is also being carried out to support excess weapons disposition in Russia and the US, including nonproliferation studies of plutonium recovery from immobilization forms and accelerated radiation damage studies of the US-specified plutonium ceramic for immobilizing plutonium. This intriguing and extraordinary cooperation on certain aspects of the weapons plutonium problem is now progressing well and much work with plutonium has been completed in the past two years. Because much excellent and unique scientific and engineering technical work has now been completed in Russia in many aspects of plutonium immobilization, this meeting in St. Petersburg was both timely and necessary to summarize, review, and discuss these efforts among those who performed the actual work. The results of this meeting will help the US and Russia jointly define the future direction of the Russian plutonium immobilization program, and make it an even stronger and more integrated Russian program. The two objectives for the meeting were to: (1) Bring together the Russian organizations, experts, and managers performing the work into one place for four days to review and discuss their work with each other; and (2) Publish a meeting summary and a proceedings to compile reports of all the excellent Russian plutonium immobilization contract work. This proceedings document presents the wide extent of Russian immobilization activities, provides a reference for their work, and makes it available to others.« less

PROCESS FOR SEPARATING PLUTONIUM FROM IMPURITIES

DOEpatents

Wahl, A.C.

1957-11-12

A method is described for separating plutonium from aqueous solutions containing uranium. It has been found that if the plutonium is reduced to its 3+ valence state, and the uranium present is left in its higher valence state, then the differences in solubility between certain salts (e.g., oxalates) of the trivalent plutonium and the hexavalent uranium can be used to separate the metals. This selective reduction of plutonium is accomplished by adding iodide ion to the solution, since iodide possesses an oxidation potential sufficient to reduce plutonium but not sufficient to reduce uranium.

PRODUCTION OF PLUTONIUM FLUORIDE FROM BISMUTH PHOSPHATE PRECIPITATE CONTAINING PLUTONIUM VALUES

DOEpatents

Brown, H.S.; Bohlmann, E.G.

1961-05-01

A process is given for separating plutonium from fission products present on a bismuth phosphate carrier. The dried carrier is first treated with hydrogen fluoride at between 500 and 600 deg C whereby some fission product fluorides volatilize away from plutonium tetrafluoride, and nonvolatile fission product fluorides are formed then with anhydrous fluorine at between 400 and 500 deg C. Bismuth and plutonium distill in the form of volatile fluorides away from the nonvolatile fission product fluorides. The bismuth and plutonium fluorides are condensed at below 290 deg C.

PLUTONIUM COMPOUNDS AND PROCESS FOR THEIR PREPARATION

DOEpatents

Wolter, F.J.; Diehl, H.C. Jr.

1958-01-01

This patent relates to certain new compounds of plutonium, and to the utilization of these compounds to effect purification or separation of the plutonium. The compounds are organic chelate compounds consisting of tetravalent plutonium together with a di(salicylal) alkylenediimine. These chelates are soluble in various organic solvents, but not in water. Use is made of this property in extracting the plutonium by contacting an aqueous solution thereof with an organic solution of the diimine. The plutonium is chelated, extracted and effectively separated from any impurities accompaying it in the aqueous phase.

METHOD OF MAKING PLUTONIUM DIOXIDE

DOEpatents

Garner, C.S.

1959-01-13

A process is presented For converting both trivalent and tetravalent plutonium oxalate to substantially pure plutonium dioxide. The plutonium oxalate is carefully dried in the temperature range of 130 to300DEC by raising the temperature gnadually throughout this range. The temperature is then raised to 600 C in the period of about 0.3 of an hour and held at this level for about the same length of time to obtain the plutonium dioxide.

METHOD OF PRODUCING PLUTONIUM TETRAFLUORIDE

DOEpatents

Tolley, W.B.; Smith, R.C.

1959-12-15

A process is presented for preparing plutonium tetrafluoride from plutonium(IV) oxalate. The oxalate is dried and decomposed at about 300 deg C to the dioxide, mixed with ammonium bifluoride, and the mixture is heated to between 50 and 150 deg C whereby ammonium plutonium fluoride is formed. The ammonium plutonium fluoride is then heated to about 300 deg C for volatilization of ammonium fluoride. Both heating steps are preferably carried out in an inert atmosphere.

Lithium metal reduction of plutonium oxide to produce plutonium metal

DOEpatents

Coops, Melvin S.

1992-01-01

A method is described for the chemical reduction of plutonium oxides to plutonium metal by the use of pure lithium metal. Lithium metal is used to reduce plutonium oxide to alpha plutonium metal (alpha-Pu). The lithium oxide by-product is reclaimed by sublimation and converted to the chloride salt, and after electrolysis, is removed as lithium metal. Zinc may be used as a solvent metal to improve thermodynamics of the reduction reaction at lower temperatures. Lithium metal reduction enables plutonium oxide reduction without the production of huge quantities of CaO--CaCl.sub.2 residues normally produced in conventional direct oxide reduction processes.

Developing a physiologically based approach for modeling plutonium decorporation therapy with DTPA.

PubMed

Kastl, Manuel; Giussani, Augusto; Blanchardon, Eric; Breustedt, Bastian; Fritsch, Paul; Hoeschen, Christoph; Lopez, Maria Antonia

2014-11-01

To develop a physiologically based compartmental approach for modeling plutonium decorporation therapy with the chelating agent Diethylenetriaminepentaacetic acid (Ca-DTPA/Zn-DTPA). Model calculations were performed using the software package SAAM II (©The Epsilon Group, Charlottesville, Virginia, USA). The Luciani/Polig compartmental model with age-dependent description of the bone recycling processes was used for the biokinetics of plutonium. The Luciani/Polig model was slightly modified in order to account for the speciation of plutonium in blood and for the different affinities for DTPA of the present chemical species. The introduction of two separate blood compartments, describing low-molecular-weight complexes of plutonium (Pu-LW) and transferrin-bound plutonium (Pu-Tf), respectively, and one additional compartment describing plutonium in the interstitial fluids was performed successfully. The next step of the work is the modeling of the chelation process, coupling the physiologically modified structure with the biokinetic model for DTPA. RESULTS of animal studies performed under controlled conditions will enable to better understand the principles of the involved mechanisms.

NUCLEAR CLEANUP: Progress Made at Rocky Flats, but Closure by 2006 Is Unlikely, and Costs May Increase

DTIC Science & Technology

2001-02-01

liquids or residues from process pipes and tanks. The contractor also dismantled plutonium - processing furnaces, stripped out contaminated process...Soil Cleanup Levels on the Scope and Cost of the 903 Pad Cleanup 30 Figures Figure 1: Workers in Protective Clothing Handling Plutonium - Contaminated ...activities—shipping nuclear materials such as plutonium - contaminated metals and powders—is expected to be completed in 2002. Another activity

PROCESS OF MAKING A NEUTRONIC REACTOR FUEL ELEMENT COMPOSITION

DOEpatents

Alter, H.W.; Davidson, J.K.; Miller, R.S.; Mewherter, J.L.

1959-01-13

A process is presented for making a ceramic-like material suitable for use as a nuclear fuel. The material consists of a solid solution of plutonium dioxide in uranium dioxide and is produced from a uranyl nitrate -plutonium nitrate solution containing uraniunm and plutonium in the desired ratio. The uranium and plutonium are first precipitated from the solution by addition of NH/ sub 4/OH and the dried precipitate is then calcined at 600 C in a hydrogen atmosphere to yield the desired solid solution of PuO/sub 2/ in UO/sub 2/.

Processes for metal extraction

NASA Technical Reports Server (NTRS)

Bowersox, David F.

1992-01-01

This report describes the processing of plutonium at Los Alamos National Laboratory (LANL), and operation illustrating concepts that may be applicable to the processing of lunar materials. The toxic nature of plutonium requires a highly closed system for processing lunar surface materials.

Methodology for worker neutron exposure evaluation in the PDCF facility design.

PubMed

Scherpelz, R I; Traub, R J; Pryor, K H

2004-01-01

A project headed by Washington Group International is meant to design the Pit Disassembly and Conversion Facility (PDCF) to convert the plutonium pits from excessed nuclear weapons into plutonium oxide for ultimate disposition. Battelle staff are performing the shielding calculations that will determine appropriate shielding so that the facility workers will not exceed target exposure levels. The target exposure levels for workers in the facility are 5 mSv y(-1) for the whole body and 100 mSv y(-1) for the extremity, which presents a significant challenge to the designers of a facility that will process tons of radioactive material. The design effort depended on shielding calculations to determine appropriate thickness and composition for glove box walls, and concrete wall thicknesses for storage vaults. Pacific Northwest National Laboratory (PNNL) staff used ORIGEN-S and SOURCES to generate gamma and neutron source terms, and Monte Carlo (computer code for) neutron photon (transport) (MCNP-4C) to calculate the radiation transport in the facility. The shielding calculations were performed by a team of four scientists, so it was necessary to develop a consistent methodology. There was also a requirement for the study to be cost-effective, so efficient methods of evaluation were required. The calculations were subject to rigorous scrutiny by internal and external reviewers, so acceptability was a major feature of the methodology. Some of the issues addressed in the development of the methodology included selecting appropriate dose factors, developing a method for handling extremity doses, adopting an efficient method for evaluating effective dose equivalent in a non-uniform radiation field, modelling the reinforcing steel in concrete, and modularising the geometry descriptions for efficiency. The relative importance of the neutron dose equivalent compared with the gamma dose equivalent varied substantially depending on the specific shielding conditions and lessons were learned from this effect. This paper addresses these issues and the resulting methodology.

Stabilizing stored PuO2 with addition of metal impurities

NASA Astrophysics Data System (ADS)

Moten, Shafaq; Huda, Muhammad

Plutonium oxides is of widespread significance due its application in nuclear fuels, space missions, as well as the long-termed storage of plutonium from spent fuel and nuclear weapons. The processes to refine and store plutonium bring many other elements in contact with the plutonium metal and thereby affect the chemistry of the plutonium. Pure plutonium metal corrodes to an oxide in air with the most stable form of this oxide is stoichiometric plutonium dioxide, PuO2. Defects such as impurities and vacancies can form in the plutonium dioxide before, during and after the refining processes as well as during storage. An impurity defect manifests itself at the bottom of the conduction band and affects the band gap of the unit cell. Studying the interaction between transition metals and plutonium dioxide is critical for better, more efficient storage plans as well as gaining insights to provide a better response to potential threats of exposure to the environment. Our study explores the interaction of a few metals within the plutonium dioxide structure which have a likelihood of being exposed to the plutonium dioxide powder. Using Density Functional Theory, we calculated a substituted metal impurity in PuO2 supercell. We repeated the calculations with an additional oxygen vacancy. Our results reveal interesting volume contraction of PuO2 supercell when one plutonium atom is substituted with a metal atom. The authors acknowledge the Texas Computing Center (TACC) at The University of Texas at Austin and High Performance Computing (HPC) at The University of Texas at Arlington.

Development of the Direct Fabrication Process for Plutonium Immobilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Congdon, J.W.

2001-07-10

The current baseline process for fabricating pucks for the Plutonium Immobilization Program includes granulation of the milled feed prior to compaction. A direct fabrication process was demonstrated that eliminates the need for granulation.

Development of a Si-PM based alpha camera for plutonium detection in nuclear fuel facilities

NASA Astrophysics Data System (ADS)

Morishita, Yuki; Yamamoto, Seiichi; Izaki, Kenji; Kaneko, Junichi H.; Toi, Kohei; Tsubota, Youichi

2014-05-01

Alpha particles are monitored for detecting nuclear fuel material (i.e., plutonium and uranium) at nuclear fuel facilities. Currently, for monitoring the airborne contamination of nuclear fuel, only energy information measured by Si-semiconductor detectors is used to distinguish nuclear fuel material from radon daughters. In some cases, however, such distinguishing is difficult when the radon concentration is high. In addition, a Si-semiconductor detector is generally sensitive to noise. In this study, we developed a new alpha-particle imaging system by combining a Si-PM array, which is insensitive to noise, with a Ce-doped Gd3Al2Ga3O12(GAGG) scintillator, and evaluated our developed system's fundamental performance. The scintillator was 0.1-mm thick, and the light guide was 3.0 mm thick. An 241Am source was used for all the measurements. We evaluated the spatial resolution by taking an image of a resolution chart. A 1.6 lp/mm slit was clearly resolved, and the spatial resolution was estimated to be less than 0.6-mm FWHM. The energy resolution was 13% FWHM. A slight distortion was observed in the image, and the uniformity near its center was within ±24%. We conclude that our developed alpha-particle imaging system is promising for plutonium detection at nuclear fuel facilities.

State waste discharge permit application: 200 Area Treated Effluent Disposal Facility (Project W-049H)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-08-01

As part of the original Hanford Federal Facility Agreement and Concent Order negotiations, US DOE, US EPA and the Washington State Department of Ecology agreed that liquid effluent discharges to the ground to the Hanford Site are subject to permitting in the State Waste Discharge Permit Program (SWDP). This document constitutes the SWDP Application for the 200 Area TEDF stream which includes the following streams discharged into the area: Plutonium Finishing Plant waste water; 222-S laboratory Complex waste water; T Plant waste water; 284-W Power Plant waste water; PUREX chemical Sewer; B Plant chemical sewer, process condensate, steam condensate; 242-A-81more » Water Services waste water.« less

PROCESS FOR PRODUCTION OF PLUTONIUM FROM ITS OXIDES

DOEpatents

Weissman, S.I.; Perlman, M.L.; Lipkin, D.

1959-10-13

A method is described for obtaining a carbide of plutonium and two methods for obtaining plutonium metal from its oxides. One of the latter involves heating the oxide, in particular PuO/sub 2/, to a temperature of 1200 to 1500 deg C with the stoichiometrical amount of carbon to fornn CO in a hard vacuum (3 to 10 microns Hg), the reduced and vaporized plutonium being collected on a condensing surface above the reaction crucible. When an excess of carbon is used with the PuO/sub 2/, a carbide of plutonium is formed at a crucible temperature of 1400 to 1500 deg C. The process may be halted and the carbide removed, or the reaction temperature can be increased to 1900 to 2100 deg C at the same low pressure to dissociate the carbide, in which case the plutonium is distilled out and collected on the same condensing surface.

Analysis of a nuclear accident: fission and activation product releases from the Fukushima Daiichi nuclear facility as remote indicators of source identification, extent of release, and state of damaged spent nuclear fuel.

PubMed

Schwantes, Jon M; Orton, Christopher R; Clark, Richard A

2012-08-21

Researchers evaluated radionuclide measurements of environmental samples taken from the Fukushima Daiichi nuclear facility and reported on the Tokyo Electric Power Co. Website following the 2011 tsunami-initiated catastrophe. This effort identified Units 1 and 3 as the major source of radioactive contamination to the surface soil near the facility. Radionuclide trends identified in the soils suggested that: (1) chemical volatility driven by temperature and reduction potential within the vented reactors' primary containment vessels dictated the extent of release of radiation; (2) all coolant had likely evaporated by the time of venting; and (3) physical migration through the fuel matrix and across the cladding wall were minimally effective at containing volatile species, suggesting damage to fuel bundles was extensive. Plutonium isotopic ratios and their distance from the source indicated that the damaged reactors were the major contributor of plutonium to surface soil at the source, decreasing rapidly with distance from the facility. Two independent evaluations estimated the fraction of the total plutonium inventory released to the environment relative to cesium from venting Units 1 and 3 to be ∼0.002-0.004%. This study suggests significant volatile radionuclides within the spent fuel at the time of venting, but not as yet observed and reported within environmental samples, as potential analytes of concern for future environmental surveys around the site. The majority of the reactor inventories of isotopes of less volatile elements like Pu, Nb, and Sr were likely contained within the damaged reactors during venting.

Nuclear Warheads: The Reliable Replacement Warhead Program and the Life Extension Program

DTIC Science & Technology

2006-12-13

Defense Nuclear Facilities Safety Board was created by Congress 1988 "as an independent oversight organization within the Executive Branch charged... nuclear facilities ." U.S. Defense Nuclear Facilities Safety Board. “Who We Are,” at [http://www.dnfsb.gov/about/index.html]. involving CHE and plutonium...approach, if successful, would “reduce or eliminate the need for ESD controls.”42 Kent Fortenberry, Technical Director of the Defense Nuclear Facilities

Search for Plutonium Salt Deposits in the Plutonium Extraction Batteries of the Marcoule Plant; RECHERCHE DE DEPOTS DE SELS DE PLUTONIUM DANS LES BATTERIES D'EXTRACTION DU PLUTONIUM DE L'USINE DE MARCOULE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bouzigues, H.; Reneaud, J.-M.

1963-01-01

A method and a special apparatus are described which make 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 grams. 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. (auth)

SEPARATION OF PLUTONIUM HYDROXIDE FROM BISMUTH HYDROXIDE

DOEpatents

Watt, G.W.

1958-08-19

An tmproved method is described for separating plutonium hydroxide from bismuth hydroxide. The end product of the bismuth phosphate processes for the separation amd concentration of plutonium is a inixture of bismuth hydroxide amd plutonium hydroxide. It has been found that these compounds can be advantageously separated by treatment with a reducing agent having a potential sufficient to reduce bismuth hydroxide to metalltc bisinuth but not sufficient to reduce the plutonium present. The resulting mixture of metallic bismuth and plutonium hydroxide can then be separated by treatment with a material which will dissolve plutonium hydroxide but not metallic bismuth. Sodiunn stannite is mentioned as a preferred reducing agent, and dilute nitric acid may be used as the separatory solvent.

SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyd, G.E.; Adamson, A.W.; Schubert, J.

A chromatographic adsorption process is presented for the separation of plutonium from other fission products formed by the irradiation of uranium. The plutonium and the lighter element fission products are adsorbed on a sulfonated phenol-formaldehyde resin bed from a nitric acid solution containing the dissolved uranium. Successive washes of sulfuric, phosphoric, and nitric acids remove the bulk of the fission products, then an eluate of dilute phosphoric and nitric acids removes the remaining plutonium and fission products. The plutonium is selectively removed by passing this solution through zirconium phosphate, from which the plutonium is dissolved with nitric acid. This processmore » provides a convenient and efficient means for isolating plutonium.« less

SEPARATION OF PLUTONIUM FROM URANIUM

DOEpatents

Feder, H.M.; Nuttall, R.L.

1959-12-15

A process is described for extracting plutonium from powdered neutron- irradiated urarium metal by contacting the latter, while maintaining it in the solid form, with molten magnesium which takes up the plutonium and separating the molten magnesium from the solid uranium.

PLUTONIUM PROCESSING OPTIMIZATION IN SUPPORT OF THE MOX FUEL PROGRAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

GRAY, DEVIN W.; COSTA, DAVID A.

2007-02-02

After Los Alamos National Laboratory (LANL) personnel completed polishing 125 Kg of plutonium as highly purified PuO{sub 2} from surplus nuclear weapons, Duke, COGEMA, Stone, and Webster (DCS) required as the next process stage, the validation and optimization of all phases of the plutonium polishing flow sheet. Personnel will develop the optimized parameters for use in the upcoming 330 kg production mission.

Project Overview: LA07-LAB072-PD02

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stanley, Floyd E.

2017-09-28

The goal of this project was to identify and characterize sources of plutonium processing signatures, and understand how fate and transport impact these signatures, with an emphasis on establishing a foundation for the use of aerosolized particle characteristics as indicators of historic and current activities within a facility. Targeted activities included: 1) Pu metal reprocessing via direct oxide reduction, 2) Breakout of α-phase and δ-phase materials, 3) CNC machining of alloyed, δ-phase Pu metal, and 4) Low speed cutting of unalloyed, α-phase metal and alloyed, δ-phase Pu metal.

Plutonium (TRU) transmutation and 233U production by single-fluid type accelerator molten-salt breeder (AMSB)

NASA Astrophysics Data System (ADS)

Furukawa, Kazuo; Kato, Yoshio; Chigrinov, Sergey E.

1995-09-01

For practical/industrial disposition of Pu(TRU) by accelerator facility, not only physical soundness and safety but also the following technological rationality should be required: (1) few R&D items including radiation damage, heat removal and material compatibility: (2) few operation/maintenance/processing works; (3) few reproduction of radioactivity; (4) effective energy production in parallel. This will be achieved by the new modification of Th-fertilizing Single-Fluid type Accelerator Molten-Salt Breeder (AMSB), by which a global nuclear energy strategy for next century might be prepared.

A Plutonium-Contaminated Wound, 1985, USA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doran M. Christensen, DO, REAC /TS Associate Director and Staff Physician Eugene H. Carbaugh, CHP, Staff Scientist, Internal Dosimetry Manager, Pacific Northwest National Laboratory, Richland, Washington

2012-02-02

A hand injury occurred at a U.S. facility in 1985 involving a pointed shaft (similar to a meat thermometer) that a worker was using to remove scrap solid plutonium from a plastic bottle. The worker punctured his right index finger on the palm side at the metacarpal-phalangeal joint. The wound was not through-and- through, although it was deep. The puncture wound resulted in deposition of ~48 kBq of alpha activity from the weapons-grade plutonium mixture with a nominal 12 to 1 Pu-alpha to {sup 241}Am-alpha ratio. This case clearly showed that DTPA was very effective for decorporation of plutonium andmore » americium. The case is a model for management of wounds contaminated with transuranics: (1) a team approach for dealing with all of the issues surrounding the incident, including the psychological, (2) early surgical intervention for foreign-body removal, (3) wound irrigation with DTPA solution, and (4) early and prolonged DTPA administration based upon bioassay and in vivo dosimetry.« less

METHOD OF FORMING PLUTONIUM-BEARING CARRIER PRECIPITATES AND WASHING SAME

DOEpatents

Faris, B.F.

1959-02-24

An improvement of the lanthanum fluoride carrier precipitation process for the recovery of plutonium is presented. In this process the plutonium is first segregated in the LaF/su precipitate and this precipitate is later dissolved and the plutonium reprecipitated as the peroxide. It has been found that the loss of plutonium by its remaining in the supernatant liquid associated with the peroxide precipitate is greatly reduced if, before dissolution, the LaF/ sub 3/ precipitate is subjected to a novel washing step which constitutes the improvement of this patent. The step consists in intimately contactifng the LaF/ sub 3/ precipitate with a 4 to 10 percent solution of sodium hydrogen sulfate at a temperature between 10 and 95 deg C for 1/2 to 3 hours.

A XAS study of the local environments of cations in (U, Ce)O 2

NASA Astrophysics Data System (ADS)

Martin, Philippe; Ripert, Michel; Petit, Thierry; Reich, Tobias; Hennig, Christoph; D'Acapito, Francesco; Hazemann, Jean Louis; Proux, Olivier

2003-01-01

Mixed oxide (MOX) fuel is usually considered as a solid solution formed by uranium and plutonium dioxides. Nevertheless, some physico-chemical properties of (U 1- y, Pu y)O 2 samples manufactured under industrial conditions showed anomalies in the domain of plutonium contents ranging between 3 and 15 at.%. Cerium is commonly used as an inactive analogue of plutonium in preliminary studies on MOX fuels. Extended X-ray Absorption Fine Structure (EXAFS) measurements performed at the European Synchrotron Radiation Facility (ESRF) at the cerium and uranium edges on (U 1- y, Ce y)O 2 samples are presented and discussed. They confirmed on an atomic scale the formation of an ideal solid solution for cerium concentrations ranging between 0 and 50 at.%.

SHIELDING AND DETECTOR RESPONSE CALCULATIONS PERTAINING TO CATEGORY 1 QUANTITIES OF PLUTONIUM AND HAND-HELD PLASTIC SCINTILLATORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Couture, A.

2013-06-07

Nuclear facilities sometimes use hand-held plastic scintillator detectors to detect attempts to divert special nuclear material in situations where portal monitors are impractical. MCNP calculations have been performed to determine the neutron and gamma radiation field arising from a Category I quantity of weapons-grade plutonium in various shielding configurations. The shields considered were composed of combinations of lead and high-density polyethylene such that the mass of the plutonium plus shield was 22.7 kilograms. Monte-Carlo techniques were also used to determine the detector response to each of the shielding configurations. The detector response calculations were verified using field measurements of high-,more » medium-, and low- energy gamma-ray sources as well as a Cf-252 neutron source.« less

EXTRACTION OF TETRAVALENT PLUTONIUM VALUES WITH METHYL ETHYL KETONE, METHYL ISOBUTYL KETONE ACETOPHENONE OR MENTHONE

DOEpatents

Seaborg, G.T.

1961-08-01

A process is described for extracting tetravalent plutonium from an aqueous acid solution with methyl ethyl ketone, methyl isobutyl ketone, or acetophenone and with the extraction of either tetravalent or hexavalent plutonium into menthone. (AEC)

H CANYON PROCESSING IN CORRELATION WITH FH ANALYTICAL LABS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weinheimer, E.

2012-08-06

Management of radioactive chemical waste can be a complicated business. H Canyon and F/H Analytical Labs are two facilities present at the Savannah River Site in Aiken, SC that are at the forefront. In fact H Canyon is the only large-scale radiochemical processing facility in the United States and this processing is only enhanced by the aid given from F/H Analytical Labs. As H Canyon processes incoming materials, F/H Labs provide support through a variety of chemical analyses. Necessary checks of the chemical makeup, processing, and accountability of the samples taken from H Canyon process tanks are performed at themore » labs along with further checks on waste leaving the canyon after processing. Used nuclear material taken in by the canyon is actually not waste. Only a small portion of the radioactive material itself is actually consumed in nuclear reactors. As a result various radioactive elements such as Uranium, Plutonium and Neptunium are commonly found in waste and may be useful to recover. Specific processing is needed to allow for separation of these products from the waste. This is H Canyon's specialty. Furthermore, H Canyon has the capacity to initiate the process for weapons-grade nuclear material to be converted into nuclear fuel. This is one of the main campaigns being set up for the fall of 2012. Once usable material is separated and purified of impurities such as fission products, it can be converted to an oxide and ultimately turned into commercial fuel. The processing of weapons-grade material for commercial fuel is important in the necessary disposition of plutonium. Another processing campaign to start in the fall in H Canyon involves the reprocessing of used nuclear fuel for disposal in improved containment units. The importance of this campaign involves the proper disposal of nuclear waste in order to ensure the safety and well-being of future generations and the environment. As processing proceeds in the fall, H Canyon will have a substantial number of samples being sent to F/H Labs. All analyses of these samples are imperative to safe and efficient processing. The important campaigns to occur would be impossible without feedback from analyses such as chemical makeup of solutions, concentrations of dissolution acids and nuclear material, as well as nuclear isotopic data. The necessity of analysis for radiochemical processing is evident. Processing devoid of F/H Lab's feedback would go against the ideals of a safety-conscious and highly accomplished processing facility such as H Canyon.« less

PLUTONIUM RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

DOEpatents

Moore, R.H.

1962-04-10

A process of recovering plutonium from neutronbombarded uranium fuel by dissolving the fuel in equimolar aluminum chloride-potassium chloride; heating the mass to above 700 deg C for decomposition of plutonium tetrachloride to the trichloride; extracting the plutonium trichloride into a molten salt containing from 40 to 60 mole % of lithium chloride, from 15 to 40 mole % of sodium chloride, and from 0 to 40 mole % of potassium chloride or calcium chloride; and separating the layer of equimolar chlorides containing the uranium from the layer formed of the plutonium-containing salt is described. (AEC)

13. VIEW OF THE MOLTEN SALT EXTRACTION LINE. THE MOLTEN ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

13. VIEW OF THE MOLTEN SALT EXTRACTION LINE. THE MOLTEN SALT EXTRACTION PROCESS WAS USED TO PURIFY PLUTONIUM BY REMOVING AMERICIUM, A DECAY BY-PRODUCT OF PLUTONIUM. (1/98) - Rocky Flats Plant, Plutonium Fabrication, Central section of Plant, Golden, Jefferson County, CO

METHOD AND MEANS FOR ELECTROLYTIC PURIFICATION OF PLUTONIUM

DOEpatents

Bjorklund, C.W.; Benz, R.; Maraman, W.J.; Leary, J.A.; Walsh, K.A.

1960-02-01

The technique of electrodepositing pure plutonium from a fused salt electrolyte of PuCl/sub 3/ and aixati metal halides is described. When an iron cathode is used, the plutonium deposit alloys therewith in the liquid state at the 400 to 600 deg C operating temperature, such liquid being allowed to drip through holes in the cathode and collect in a massive state in a tantallum cup. The process is adaptable to continuous processing by the use of depleted plutonium fuel as the anode: good to excellent separation from fission products is obtained with a Pu--Fe "fission" anode containing representative fractions of Ce, Ru, Zr, La, Mo, and Nb.

SEPARATION PROCESS USING COMPLEXING AND ADSORPTION

DOEpatents

Spedding, J.H.; Ayers, J.A.

1958-06-01

An adsorption process is described for separating plutonium from a solution of neutron-irradiated uranium containing ions of a compound of plutonium and other cations. The method consists of forming a chelate complex compound with plutoniunn ions in the solution by adding a derivative of 8- hydroxyquinoline, which derivative contains a sulfonic acid group, and adsorbing the remaining cations from the solution on a cation exchange resin, while the complexed plutonium remains in the solution.

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen Joseph; Fugate, Michael Lynn; Trellue, Holly Renee

2016-10-28

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly tomore » confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).« less

Ferric ion as a scavenging agent in a solvent extraction process

DOEpatents

Bruns, Lester E.; Martin, Earl C.

1976-01-01

Ferric ions are added into the aqueous feed of a plutonium scrap recovery process that employs a tributyl phosphate extractant. Radiolytic degradation products of tributyl phosphate such as dibutyl phosphate form a solid precipitate with iron and are removed from the extraction stages via the waste stream. Consequently, the solvent extraction characteristics are improved, particularly in respect to minimizing the formation of nonstrippable plutonium complexes in the stripping stages. The method is expected to be also applicable to the partitioning of plutonium and uranium in a scrap recovery process.

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

NASA Astrophysics Data System (ADS)

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.

Proliferation resistance design of a plutonium cycle (Proliferation Resistance Engineering Program: PREP)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sorenson, R.J.; Roberts, F.P.; Clark, R.G.

1979-01-19

This document describes the proliferation resistance engineering concepts developed to counter the threat of proliferation of nuclear weapons in an International Fuel Service Center (IFSC). The basic elements of an International Fuel Service Center are described. Possible methods for resisting proliferation such as processing alternatives, close-coupling of facilities, process equipment layout, maintenance philosophy, process control, and process monitoring are discussed. Political and institutional issues in providing proliferation resistance for an International Fuel Service Center are analyzed. The conclusions drawn are (1) use-denial can provide time for international response in the event of a host nation takeover. Passive use-denial is moremore » acceptable than active use-denial, and acceptability of active-denial concepts is highly dependent on sovereignty, energy dependence and economic considerations; (2) multinational presence can enhance proliferation resistance; and (3) use-denial must be nonprejudicial with balanced interests for governments and/or private corporations being served. Comparisons between an IFSC as a national facility, an IFSC with minimum multinational effect, and an IFSC with maximum multinational effect show incremental design costs to be less than 2% of total cost of the baseline non-PRE concept facility. The total equipment acquisition cost increment is estimated to be less than 2% of total baseline facility costs. Personnel costs are estimated to increase by less than 10% due to maximum international presence. 46 figures, 9 tables.« less

PROCESS FOR SEPARATING PLUTONIUM BY REPEATED PRECIPITATION WITH AMPHOTERIC HYDROXIDE CARRIERS

DOEpatents

Faris, B.F.

1960-04-01

A multiple carrier precipitation method is described for separating and recovering plutonium from an aqueous solution. The hydroxide of an amphoteric metal is precipitated in an aqueous plutonium-containing solution. This precipitate, which carries plutonium, is then separated from the supernatant liquid and dissolved in an aqueous hydroxide solution, forming a second plutonium- containing solution. lons of an amphoteric metal which forms an insoluble hydroxide under the conditions existing in this second solution are added to the second solution. The precipitate which forms and which carries plutonium is separated from the supernatant liquid. Amphoteric metals which may be employed are aluminum, bibmuth, copper, cobalt, iron, lanthanum, nickel, and zirconium.

PROCESS FOR SEPARATION OF HEAVY METALS

DOEpatents

Duffield, R.B.

1958-04-29

A method is described for separating plutonium from aqueous acidic solutions of neutron-irradiated uranium and the impurities associated therewith. The separation is effected by adding, to the solution containing hexavalent uranium and plutonium, acetate ions and the ions of an alkali metal and those of a divalent metal and thus forming a complex plutonium acetate salt which is carried by the corresponding complex of uranium, such as sodium magnesium uranyl acetate. The plutonium may be separated from the precipitated salt by taking the same back into solution, reducing the plutonium to a lower valent state on reprecipitating the sodium magnesium uranyl salt, removing the latter, and then carrying the plutonium from ihe solution by means of lanthanum fluoride.

PROCESSING ALTERNATIVES FOR DESTRUCTION OF TETRAPHENYLBORATE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lambert, D; Thomas Peters, T; Samuel Fink, S

Two processes were chosen in the 1980's at the Savannah River Site (SRS) to decontaminate the soluble High Level Waste (HLW). The In Tank Precipitation (ITP) process (1,2) was developed at SRS for the removal of radioactive cesium and actinides from the soluble HLW. Sodium tetraphenylborate was added to the waste to precipitate cesium and monosodium titanate (MST) was added to adsorb actinides, primarily uranium and plutonium. Two products of this process were a low activity waste stream and a concentrated organic stream containing cesium tetraphenylborate and actinides adsorbed on monosodium titanate (MST). A copper catalyzed acid hydrolysis process wasmore » built to process (3, 4) the Tank 48H cesium tetraphenylborate waste in the SRS's Defense Waste Processing Facility (DWPF). Operation of the DWPF would have resulted in the production of benzene for incineration in SRS's Consolidated Incineration Facility. This process was abandoned together with the ITP process in 1998 due to high benzene in ITP caused by decomposition of excess sodium tetraphenylborate. Processing in ITP resulted in the production of approximately 1.0 million liters of HLW. SRS has chosen a solvent extraction process combined with adsorption of the actinides to decontaminate the soluble HLW stream (5). However, the waste in Tank 48H is incompatible with existing waste processing facilities. As a result, a processing facility is needed to disposition the HLW in Tank 48H. This paper will describe the process for searching for processing options by SRS task teams for the disposition of the waste in Tank 48H. In addition, attempts to develop a caustic hydrolysis process for in tank destruction of tetraphenylborate will be presented. Lastly, the development of both a caustic and acidic copper catalyzed peroxide oxidation process will be discussed.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rudisill, T.S.

1999-08-11

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 precipitationmore » 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 ay of plutonium-241) in the dissolved precipitate, a value consistent with the recovery of europium, the americium surrogate.In a subsequent experiment, the plutonium solubility following an oxalate precipitation to simulate the preparation of a slurry feed for a batch melter was 21 mg/mL at 35 degrees C. The increase in solubility compared to the value measured during the pretreatment experiment was attributed to the increased nitrate concentration and ensuing increase in plutonium complexation. The solubility of the plutonium following a precipitant wash with 0.1M oxalic acid was unchanged. The recovery of plutonium from the precipitate slurry was greater than 97 percent allowing an estimation that approximately 92 percent of the plutonium in Tank 17.1 will report to the glass. The behavior of the lanthanides and soluble metal impurities was consistent with the behavior seen during the pretreatment experiment. A trace level material balance showed that 99.9 percent of the americium w as recovered from the precipitate slurry. The overall recovery of americium from the pretreatment and feed preparation processes was greater than 97 percent, which was consistent with the measured recovery of the europium surrogate.« less

Minimizing Glovebox Glove Breaches, Part III: Deriving Service Lifetimes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cournoyer, M.E.; Wilson, K.V.; Maestas, M.M.

At the Los Alamos Plutonium Facility, various isotopes of plutonium along with other actinides are handled in a glove box environment. Weapons-grade plutonium consists mainly in Pu-239. Pu-238 is another isotope used for heat sources. The Pu-238 is more aggressive regarding gloves due to its higher alpha-emitting characteristic ({approx}300 times more active than Pu-239), which modifies the change-out intervals for gloves. Optimization of the change-out intervals for gloves is fundamental since Nuclear Materials Technology (NMT) Division generates approximately 4 m{sup 3}/yr of TRU waste from the disposal of glovebox gloves. To reduce the number of glovebox glove failures, the NMTmore » Division pro-actively investigates processes and procedures that minimize glove failures. Aging studies have been conducted that correlate changes in mechanical (physical) properties with degradation chemistry. This present work derives glovebox glove change intervals based on mechanical data of thermally aged Hypalon{sup R}, and Butasol{sup R} glove samples. Information from this study represent an important baseline in gauging the acceptable standards for polymeric gloves used in a laboratory glovebox environment and will be used later to account for possible presence of dose-rate or synergistic effects in 'combined-environment'. In addition, excursions of contaminants into the operator's breathing zone and excess exposure to the radiological sources associated with unplanned breaches in the glovebox are reduced. (authors)« less

Plutonium recovery from organic materials

DOEpatents

Deaton, R.L.; Silver, G.L.

1973-12-11

A method is described for removing plutonium or the like from organic material wherein the organic material is leached with a solution containing a strong reducing agent such as titanium (III) (Ti/sup +3None)/, chromium (II) (Cr/ sup +2/), vanadium (II) (V/sup +2/) ions, or ferrous ethylenediaminetetraacetate (EDTA), the leaching yielding a plutonium-containing solution that is further processed to recover plutonium. The leach solution may also contain citrate or tartrate ion. (Official Gazette)

SEPARATION OF PLUTONIUM FROM AQUEOUS SOLUTIONS BY ION-EXCHANGE

DOEpatents

Schubert, J.

1958-06-01

A process is described for the separation of plutonium from an aqueous solution of a plutonium salt, which comprises adding to the solution an acid of the group consisting of sulfuric acid, phosphoric acid, and oxalic acid, and mixtures thereof to provide an acid concentration between 0.0001 and 1 M, contacting the resultant solution with a synthetic organic anion exchange resin, and separating the aqueous phase and the resin which contains the plutonium.

AMINE EXTRACTION OF PLUTONIUM FROM NITRIC ACID SOLUTIONS LOADING AND STRIPPING EXPERIMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, A.S.

1961-01-19

Information is presented on a suitable amine processing system for plutonium nitrate. Experiments with concentrated plutonium nitrate solutions show that trilaurylamine (TLA) - xylene solvent systems did not form a second organic phase. Experiments are also reported with tri-noctylamine (TnOA)-xylene and TLA-Amsco - octyl alcohol. Two organic phases appear in both these systems at high plutonium nitrate concentrations. Data are tabulated from loading and stripping experiments. (J.R.D.)

Investigation of criticality safety control infraction data at a nuclear facility

DOE PAGES

Cournoyer, Michael E.; Merhege, James F.; Costa, David A.; ...

2014-10-27

Chemical and metallurgical operations involving plutonium and other nuclear materials account for most activities performed at the LANL's Plutonium Facility (PF-4). The presence of large quantities of fissile materials in numerous forms at PF-4 makes it necessary to maintain an active criticality safety program. The LANL Nuclear Criticality Safety (NCS) Program provides guidance to enable efficient operations while ensuring prevention of criticality accidents in the handling, storing, processing and transportation of fissionable material at PF-4. In order to achieve and sustain lower criticality safety control infraction (CSCI) rates, PF-4 operations are continuously improved, through the use of Lean Manufacturing andmore » Six Sigma (LSS) business practices. Employing LSS, statistically significant variations (trends) can be identified in PF-4 CSCI reports. In this study, trends have been identified in the NCS Program using the NCS Database. An output metric has been developed that measures ADPSM Management progress toward meeting its NCS objectives and goals. Using a Pareto Chart, the primary CSCI attributes have been determined in order of those requiring the most management support. Data generated from analysis of CSCI data help identify and reduce number of corresponding attributes. In-field monitoring of CSCI's contribute to an organization's scientific and technological excellence by providing information that can be used to improve criticality safety operation safety. This increases technical knowledge and augments operational safety.« less

RECOVERY OF PLUTONIUM BY CARRIER PRECIPITATION

DOEpatents

Goeckermann, R.H.

1961-04-01

A process is given for recovering plutonium from an aqueous nitric acid zirconium-containing solution of an acidity between 0.2 and 1 N by adding fluoride anions (1.5 to 5 mg/l) and precipitating the plutonium with an excess of hydrogen peroxide at from 53 to 65 deg C.

ELECTRODEPOSITION OF PLUTONIUM

DOEpatents

Wolter, F.J.

1957-09-10

A process of electrolytically recovering plutonium from dilute aqueous solutions containing plutonium ions comprises electrolyzing the solution at a current density of about 0.44 ampere per square centimeter in the presence of an acetate-sulfate buffer while maintaining the pH of the solution at substantially 5 and using a stirred mercury cathode.

Joint Actinide Shock Physics Experimental Research - JASPER

ScienceCinema

None

2018-01-16

Commonly known as JASPER the Joint Actinide Shock Physics Experimental Research facility is a two stage light gas gun used to study the behavior of plutonium and other materials under high pressures, temperatures, and strain rates.

Method for dissolving delta-phase plutonium

DOEpatents

Karraker, David G.

1992-01-01

A process for dissolving plutonium, and in particular, delta-phase plutonium. The process includes heating a mixture of nitric acid, hydroxylammonium nitrate (HAN) and potassium fluoride to a temperature between 40.degree. and 70.degree. C., then immersing the metal in the mixture. Preferably, the nitric acid has a concentration of not more than 2M, the HAN approximately 0.66M, and the potassium fluoride 0.1M. Additionally, a small amount of sulfamic acid, such as 0.1M can be added to assure stability of the HAN in the presence of nitric acid. The oxide layer that forms on plutonium metal may be removed with a non-oxidizing acid as a pre-treatment step.

PROCESS FOR THE SEPARATION OF HEAVY METALS

DOEpatents

Gofman, J.W.; Connick, R.E.; Wahl, A.C.

1959-01-27

A method is presented for thc separation of plutonium from uranium and the fission products with which it is associated. The method is based on the fact that hexavalent plutonium forms an insoluble complex precipitate with sodium acetate, as does the uranyl ion, while reduced plutonium is not precipitated by sodium acetate. Several embodiments are shown, e.g., a solution containing plutonium and uranium in the hexavalent state may be contacted with sodium acetate causing the formation of a sodium uranyl acetate precipitate which carries the plutonium values while the fission products remain in solution. If the original solution is treated with a reducing agent, so that the plutonium is reduced while the uranium remains in the hexavalent state, and sodium and acetate ions are added, the uranium will precipitutc while the plutonium remains in solution effecting separation of the Pu from urarium.

DISSOLUTION OF LANTHANUM FLUORIDE PRECIPITATES

DOEpatents

Fries, B.A.

1959-11-10

A plutonium separatory ore concentration procedure involving the use of a fluoride type of carrier is presented. An improvement is given in the derivation step in the process for plutonium recovery by carrier precipitation of plutonium values from solution with a lanthanum fluoride carrier precipitate and subsequent derivation from the resulting plutonium bearing carrier precipitate of an aqueous acidic plutonium-containing solution. The carrier precipitate is contacted with a concentrated aqueous solution of potassium carbonate to effect dissolution therein of at least a part of the precipitate, including the plutonium values. Any remaining precipitate is separated from the resulting solution and dissolves in an aqueous solution containing at least 20% by weight of potassium carbonate. The reacting solutions are combined, and an alkali metal hydroxide added to a concentration of at least 2N to precipitate lanthanum hydroxide concomitantly carrying plutonium values.

High-Temperature Oxidation of Plutonium Surrogate Metals and Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sparks, Joshua C.; Krantz, Kelsie E.; Christian, Jonathan H.

The Plutonium Management and Disposition Agreement (PMDA) is a nuclear non-proliferation agreement designed to remove 34 tons of weapons-grade plutonium from Russia and the United States. While several removal options have been proposed since the agreement was first signed in 2000, processing the weapons-grade plutonium to mixed-oxide (MOX) fuel has remained the leading candidate for achieving the goals of the PMDA. However, the MOX program has received its share of criticisms, which causes its future to be uncertain. One alternative pathway for plutonium disposition would involve oxidizing the metal followed by impurity down blending and burial in the Waste Isolationmore » Pilot Plant (WIPP) in Carlsbad, New Mexico. This pathway was investigated by use of a hybrid microwave and a muffle furnace with Fe and Al as surrogate materials. Oxidation occurred similarly in the microwave and muffle furnace; however, the microwave process time was significantly faster.« less

Bi-Modal Model for Neutron Emissions from PuO{sub 2} and MOX Holdup

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menlove, Howard; Lafleur, Adrienne

2015-07-01

The measurement of uranium and plutonium holdup in plants during process activity and for decommissioning is important for nuclear safeguards and material control. The amount of plutonium and uranium holdup in glove-boxes, pipes, ducts, and other containers has been measured for several decades using both neutron and gamma-ray techniques. For the larger containers such as hot cells and glove-boxes that contain processing equipment, the gamma-ray techniques are limited by self-shielding in the sample as well as gamma absorption in the equipment and associated shielding. The neutron emission is more penetrating and has been used extensively to measure the holdup formore » the large facilities such as the MOX processing and fabrication facilities in Japan and Europe. In some case the totals neutron emission rates are used to determine the holdup mass and in other cases the coincidence rates are used such as at the PFPF MOX fabrication plant in Japan. The neutron emission from plutonium and MOX has 3 primary source terms: 1) Spontaneous fission (SF) from the plutonium isotopes, 2) The (α,n) reactions from the plutonium alpha particle emission reacting with the oxygen and other impurities, and 3) Neutron multiplication (M) in the plutonium and uranium as a result of neutrons created by the first two sources. The spontaneous fission yield per gram is independent of thickness, whereas, the above sources 2) and 3) are very dependent on the thickness of the deposit. As the effective thickness of the deposit becomes thin relative to the alpha particle range, the (α,n) reactions and neutrons from multiplication (M) approach zero. In any glove-box, there will always be two primary modes of holdup accumulation, namely direct powder contact and non-contact by air dispersal. These regimes correspond to surfaces in the glove-box that have come into direct contact with the process MOX powder versus surface areas that have not had direct contact with the powder. The air dispersal of PuO{sub 2} particles has been studied for several decades by health physicists, because the primary health hazard of plutonium is breathing the airborne particles. The air dispersal mechanism results from the smaller particles in the top layer of powder that are lifted into the air by the electrostatic charge buildup from the alpha decay process, and the air convection carries the particles to new more distant locations. If there is open plutonium powder in a glove-box, the surfaces at more distant locations will become contaminated over time. The range of an alpha particle in a solid or powder is a function of the particle energy, the material density, and the atomic number A of the material. The average energy of a plutonium alpha particle is ∼5.2 MeV and the range in air is ∼37 mm. The range in other materials can be estimated via the Bragg-Kleenman equation. For plutonium, A is 94, and the typical density for a single particle is ∼11.5 g/cm{sup 3}, but for a powder, the density would be less because of the air packing fraction. The significance of the small diameter is that the range of the alpha particle is ∼50 μm for powder density 2.5 and significantly less for a single particle with density 11.5, so the thin deposit of separate small particles will have a greatly reduced (α,n) yield. The average alpha transit length to the surface in the isolated MOX particle would be < 2.5 μm; whereas, the range of the alpha particle is much longer. Thus, most of the alpha particles would escape from the MOX particle and be absorbed by the walls and air. The air dispersal particles will have access to a large surface area that includes the walls, whereas, the powder contact surface area will be orders of magnitude smaller. Thus, the vast majority of the glove-box surface area does not produce the full (α,n) reaction neutron yield, even from the O{sub 2} in the PuO{sub 2} as well as any impurity contamination such as H{sub 2}O. To obtain a more quantitative estimate of the neutron (α,n) yields as a function of holdup deposit thickness, we have used MCNPX calculations to estimate the absorption of alpha particles in PuO{sub 2} holdup deposits. The powder thickness was varied from 0.1 μm to 5000 μm and the alpha particle escape probability was calculated. As would be expected, as the thickness approaches zero, the escape probability approaches 1.0, and as the thickness gets much greater than the alpha particle range (∼50 μm), the escape probability becomes small. Typically, the neutron holdup calibration measurement are performed using sealed containers of thick MOX that has all 3 sources of neutrons [SF, (α,n), and M], and no significant impurities. Thus, the calibration counting rates need to include corrections for M and (α,n) yields that are different for the holdup compared with the calibration samples. If totals neutron counting is used for the holdup measurements, the variability of the (α,n) term needs to be considered.« less

Plutonium in the atmosphere: A global perspective.

PubMed

Thakur, P; Khaing, H; Salminen-Paatero, S

2017-09-01

A number of potential source terms have contributed plutonium isotopes to the atmosphere. The atmospheric nuclear weapon tests conducted between 1945 and 1980 and the re-entry of the burned SNAP-9A satellite in 1964, respectively. It is generally believed that current levels of plutonium in the stratosphere are negligible and compared with the levels generally found at surface-level air. In this study, the time trend analysis and long-term behavior of plutonium isotopes ( 239+240 Pu and 238 Pu) in the atmosphere were assessed using historical data collected by various national and international monitoring networks since 1960s. An analysis of historical data indicates that 239+240 Pu concentration post-1984 is still frequently detectable, whereas 238 Pu is detected infrequently. Furthermore, the seasonal and time-trend variation of plutonium concentration in surface air followed the stratospheric trends until the early 1980s. After the last Chinese test of 1980, the plutonium concentrations in surface air dropped to the current levels, suggesting that the observed concentrations post-1984 have not been under stratospheric control, but rather reflect the environmental processes such as resuspension. Recent plutonium atmospheric air concentrations data show that besides resuspension, other environmental processes such as global dust storms and biomass burning/wildfire also play an important role in redistributing plutonium in the atmosphere. Copyright © 2017 Elsevier Ltd. All rights reserved.

PRECIPITATION METHOD OF SEPARATION OF NEPTUNIUM

DOEpatents

Magnusson, L.B.

1958-07-01

A process is described for the separation of neptunium from plutonium in an aqueous solution containing neptunium ions in a valence state not greater than +4, plutonium ioms in a valence state not greater than +4, and sulfate ions. The Process consists of adding hypochlorite ions to said solution in order to preferentially oxidize the neptunium and then adding lanthanum ions and fluoride ions to form a precipitate of LaF/sub 3/ carrying the plutonium, and thereafter separating the supernatant solution from the precipitate.

SCAVENGER AND PROCESS OF SCAVENGING

DOEpatents

Olson, C.M.

1960-04-26

Carrier precipitation processes are given for the separation and recovery of plutonium from aqueous acidic solutions containing plutonium and fission products. Bismuth phosphate is precipitated in the acidic solution while plutonlum is maintained in the hexavalent oxidation state. Preformed, uncalcined, granular titanium dioxide is then added to the solution and the fission product-carrying bismuth phosphate and titanium dioxide are separated from the resulting mixture. Fluosilicic acid, which dissolves any remaining titanium dioxide particles, is then added to the purified plutonium-containing solution.

Plutonium (TRU) transmutation and {sup 233}U production by single-fluid type accelerator molten-salt breeder (AMSB)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Furukawa, Kazuo; Kato, Yoshio; Chigrinov, Sergey E.

1995-09-15

For practical/industrial disposition of Pu(TRU) by accelerator facility, not only physical soundness and safety but also the following technological rationality should be required: (1) few R and D items including radiation damage, heat removal and material compatibility: (2) few operation/maintenance/processing works; (3) few reproduction of radioactivity; (4) effective energy production in parallel. This will be achieved by the new modification of Th-fertilizing Single-Fluid type Accelerator Molten-Salt Breeder (AMSB), by which a global nuclear energy strategy for next century might be prepared.

Plutonium (TRU) transmutation and {sup 233}U production by single-fluid type accelerator molten-salt breeder (AMSB)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Furukaw, Kazuo; Kato, Yoshio; Chigrinov, Sergey E.

1995-10-01

For practical/industrial disposition of Pu(TRU) by accelerator facility, not only physical soundness and safety but also the following technological rationality should be required: (1) few R&D items including radiation damage, heat removal and material compatibility; (2) few operation/maintenance/processing works: (3) few reproduction of radioactivity; (4) effective energy production in parallel. This will be achieved by the new modification of Th-fertilizing Single-Fluid type Accelerator Molten-Salt Breeder (AMSB), by which a global nuclear energy strategy for next century might be prepared.

13. Elevations, 233S, U.S. Atomic Energy Commission, Hanford Works, General ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

13. Elevations, 233-S, U.S. Atomic Energy Commission, Hanford Works, General Electric Company, Dwg. No. H-2-7203, 1956. - Reduction-Oxidation Complex, Plutonium Concentration Facility, 200 West Area, Richland, Benton County, WA

Interactive Video: Why Trainers Are Tuning In.

ERIC Educational Resources Information Center

Broderick, Richard

1982-01-01

The uses of interactive video are explored through various case studies. They include cardiopulmonary resuscitation training for the Dallas American Heart Association, Ford Motor Company dealership training, employee training at the Los Angeles Plutonium Facility, and others. (CT)

A Clear Success for International Transport of Plutonium and MOX Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blachet, L.; Jacot, P.; Bariteau, J.P.

2006-07-01

An Agreement between the United States and Russia to eliminate 68 metric tons of surplus weapons-grade plutonium provided the basis for the United States government and its agency, the Department of Energy (DOE), to enter into contracts with industry leaders to fabricate mixed oxide (MOX) fuels (a blend of uranium oxide and plutonium oxide) for use in existing domestic commercial reactors. DOE contracted with Duke, COGEMA, Stone and Webster (DCS), a limited liability company comprised of Duke Energy, COGEMA Inc. and Stone and Webster to design a Mixed Oxide Fuel Fabrication Facility (MFFF) which would be built and operated atmore » the DOE Savannah River Site (SRS) near Aiken, South Carolina. During this same time frame, DOE commissioned fabrication and irradiation of lead test assemblies in one of the Mission Reactors to assist in obtaining NRC approval for batch implementation of MOX fuel prior to the operations phase of the MFFF facility. On February 2001, DOE directed DCS to initiate a pre-decisional investigation to determine means to obtain lead assemblies including all international options for manufacturing MOX fuels. This lead to implementation of the EUROFAB project and work was initiated in earnest on EUROFAB by DCS on November 7, 2003. (authors)« less

PLATINUM HEXAFLUORIDE AND METHOD OF FLUORINATING PLUTONIUM CONTAINING MIXTURES THERE-WITH

DOEpatents

Malm, J.G.; Weinstock, B.; Claassen, H.H.

1959-07-01

The preparation of platinum hexafluoride and its use as a fluorinating agent in a process for separating plutonium from fission products is presented. According to the invention, platinum is reacted with fluorine gas at from 900 to 1100 deg C to form platinum hexafluoride. The platinum hexafluoride is then contacted with the plutonium containing mixture at room temperature to form plutonium hexafluoride which is more volatile than the fission products fluorides and therefore can be isolated by distillation.

SOLVENT EXTRACTION PROCESS FOR PLUTONIUM

DOEpatents

Seaborg, G.T.

1959-04-14

The separation of plutonium from aqueous inorganic acid solutions by the use of a water immiscible organic extractant liquid is described. The plutonium must be in the oxidized state, and the solvents covered by the patent include nitromethane, nitroethane, nitropropane, and nitrobenzene. The use of a salting out agents such as ammonium nitrate in the case of an aqueous nitric acid solution is advantageous. After contacting the aqueous solution with the organic extractant, the resulting extract and raffinate phases are separated. The plutonium may be recovered by any suitable method.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, Jon; Hayes, Steven; Walters, L. C.

This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO 2 and UO 2-PuO 2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availabilitymore » are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.« less

Characterization of 618-11 solid waste burial ground, disposed waste, and description of the waste generating facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hladek, K.L.

1997-10-07

The 618-11 (Wye or 318-11) burial ground received transuranic (TRTJ) and mixed fission solid waste from March 9, 1962, through October 2, 1962. It was then closed for 11 months so additional burial facilities could be added. The burial ground was reopened on September 16, 1963, and continued operating until it was closed permanently on December 31, 1967. The burial ground received wastes from all of the 300 Area radioactive material handling facilities. The purpose of this document is to characterize the 618-11 solid waste burial ground by describing the site, burial practices, the disposed wastes, and the waste generatingmore » facilities. This document provides information showing that kilogram quantities of plutonium were disposed to the drum storage units and caissons, making them transuranic (TRU). Also, kilogram quantities of plutonium and other TRU wastes were disposed to the three trenches, which were previously thought to contain non-TRU wastes. The site burial facilities (trenches, caissons, and drum storage units) should be classified as TRU and the site plutonium inventory maintained at five kilograms. Other fissile wastes were also disposed to the site. Additionally, thousands of curies of mixed fission products were also disposed to the trenches, caissons, and drum storage units. Most of the fission products have decayed over several half-lives, and are at more tolerable levels. Of greater concern, because of their release potential, are TRU radionuclides, Pu-238, Pu-240, and Np-237. TRU radionuclides also included slightly enriched 0.95 and 1.25% U-231 from N-Reactor fuel, which add to the fissile content. The 618-11 burial ground is located approximately 100 meters due west of Washington Nuclear Plant No. 2. The burial ground consists of three trenches, approximately 900 feet long, 25 feet deep, and 50 feet wide, running east-west. The trenches constitute 75% of the site area. There are 50 drum storage units (five 55-gallon steel drums welded together) buried in three rows in the northeast comer. In addition, five eight-foot diameter caissons are located at the west end of the center row of the drum storage units. Initially, wastes disposed to the caissons and drum storage units were from the 325 and 327 building hot cells. Later, a small amount of remote-handled (RH) waste from the 309 building Plutonium Recycle Test Reactor (PRTR) cells, and the newly built 324 building hot cells, was disposed at the site.« less

URANOUS IODATE AS A CARRIER FOR PLUTONIUM

DOEpatents

Miller, D.R.; Seaborg, G.T.; Thompson, S.G.

1959-12-15

A process is described for precipitating plutonium on a uranous iodate carrier from an aqueous acid solution conA plutonium solution more concentrated than the original solution can then be obtained by oxidizing the uranium to the hexavalent state and dissolving the precipitate, after separating the latter from the original solution, by means of warm nitric acid.

BASIC PEROXIDE PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM CONTAMINANTS

DOEpatents

Seaborg, G.T.; Perlman, I.

1959-02-10

A process is described for the separation from each other of uranyl values, tetravalent plutonium values and fission products contained in an aqueous acidic solution. First the pH of the solution is adjusted to between 2.5 and 8 and hydrogen peroxide is then added to the solution causing precipitation of uranium peroxide which carries any plutonium values present, while the fission products remain in solution. Separation of the uranium and plutonium values is then effected by dissolving the peroxide precipitate in an acidic solution and incorporating a second carrier precipitate, selective for plutonium. The plutonium values are thus carried from the solution while the uranium remains flissolved. The second carrier precipitate may be selected from among the group consisting of rare earth fluorides, and oxalates, zirconium phosphate, and bismuth lihosphate.

Design and fabrication of 55-gallon drum shuffler standards

DOE Office of Scientific and Technical Information (OSTI.GOV)

Long, S.M.; Hsue, F.; Hoth, C.

1994-08-01

To analyze waste with varying levels of nuclear material, suitable standards are needed to calibrate analytical instrumentation. At the Los Alamos Plutonium Facility, the authors have designed and fabricated a single drum standard for a passive-active neutron counter (shuffler). The standard is modified simply by adding or subtracting plutonium of uranium cylinders to adapt to a range of nuclear material. The plutonium or uranium oxide was placed into small cylindrical containers (1-in. diameter by 5-in. long) and diluted with diatomaceous earth. The cylinders were welded closed and removed from the glove box environment without any external contamination. The containers weremore » leak tested and then placed on a segmented gamma scanner to assure homogeneous distribution of the nuclear material. The cylinders are now placed into the drum to achieve the needed ranges for calibration of the instruments.« less

7. VIEW OF AIRLOCK ENTRY. AIRLOCK DOUBLE DOORS WERE USED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

7. VIEW OF AIRLOCK ENTRY. AIRLOCK DOUBLE DOORS WERE USED TO KEEP ATMOSPHERES CONFINED TO SPECIFIC AREAS. (6/29/78) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

Effects of Aging on PuO2∙xH2O Particle Size in Alkaline Solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Delegard, Calvin H.

Between 1944 and 1989, 54.5 metric tons of the United States’ weapons-grade plutonium and an additional 12.9 metric tons of fuel-grade plutonium were produced and separated from irradiated fuel at the Hanford Site. Acidic high-activity wastes containing around 600 kg of plutonium were made alkaline and discharged to underground storage tanks from separations, isolation, and recycle processes to yield average plutonium concentration of about 0.003 grams per liter (or ~0.0002 wt%) in the ~200 million liter tank waste volume. The plutonium is largely associated with low-solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g.,more » iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of plutonium hydrous oxide, PuO2∙xH2O, could undergo sufficient crystal growth through dissolution and reprecipitation in the alkaline tank waste to potentially become separable from neutron absorbing constituents by settling or sedimentation. Thermodynamic considerations and laboratory studies of systems chemically analogous to tank waste show that the plutonium formed in the alkaline tank waste by precipitation through neutralization from acid solution probably entered as 2–4-nm PuO2∙xH2O crystallite particles that, because of their low solubility and opposition from radiolytic processes, grow from that point at exceedingly slow rates, thus posing no risk of physical segregation.« less

Safeguards by Design Challenge

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alwin, Jennifer Louise

The International Atomic Energy Agency (IAEA) defines Safeguards as a system of inspection and verification of the peaceful uses of nuclear materials as part of the Nuclear Nonproliferation Treaty. IAEA oversees safeguards worldwide. Safeguards by Design (SBD) involves incorporation of safeguards technologies, techniques, and instrumentation during the design phase of a facility, rather that after the fact. Design challenge goals are the following: Design a system of safeguards technologies, techniques, and instrumentation for inspection and verification of the peaceful uses of nuclear materials. Cost should be minimized to work with the IAEA’s limited budget. Dose to workers should always bemore » as low are reasonably achievable (ALARA). Time is of the essence in operating facilities and flow of material should not be interrupted significantly. Proprietary process information in facilities may need to be protected, thus the amount of information obtained by inspectors should be the minimum required to achieve the measurement goal. Then three different design challenges are detailed: Plutonium Waste Item Measurement System, Marine-based Modular Reactor, and Floating Nuclear Power Plant (FNPP).« less

URANIUM DECONTAMINATION WITH RESPECT TO ZIRCONIUM

DOEpatents

Vogler, S.; Beederman, M.

1961-05-01

A process is given for separating uranium values from a nitric acid aqueous solution containing uranyl values, zirconium values and tetravalent plutonium values. The process comprises contacting said solution with a substantially water-immiscible liquid organic solvent containing alkyl phosphate, separating an organic extract phase containing the uranium, zirconium, and tetravalent plutonium values from an aqueous raffinate, contacting said organic extract phase with an aqueous solution 2M to 7M in nitric acid and also containing an oxalate ion-containing substance, and separating a uranium- containing organic raffinate from aqueous zirconium- and plutonium-containing extract phase.

Volatile molecule PuO 3 observed from subliming plutonium dioxide

NASA Astrophysics Data System (ADS)

Ronchi, C.; Capone, F.; Colle, J. Y.; Hiernaut, J. P.

2000-06-01

Mass spectrometric measurements of effusing vapours over PuO 2 and (U, Pu)O 2 indicate the presence of volatile PuO 3 (g) molecules. The formation of plutonium trioxide vapour is due to a chemical process involving oxygen adsorbed during oxidation of the sample. Although in the examined samples, the fraction of trioxide effusing in vacuo was of the order of 0.02 ppm of the plutonium content, under steady-state oxidation conditions it has been shown that the process can have a relevant effect on the sublimation rate of the dioxide.

DEACTIVATION AND DECOMMISSIONING ENVIRONMENTAL STRATEGY FOR THE PLUTONIUM FINISHING PLANT COMPLEX, HANFORD NUCLEAR RESERVATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hopkins, A.M.; Heineman, R.; Norton, S.

Maintaining compliance with environmental regulatory requirements is a significant priority in successful completion of the Plutonium Finishing Plant (PFP) Nuclear Material Stabilization (NMS) Project. To ensure regulatory compliance throughout the deactivation and decommissioning of the PFP complex, an environmental regulatory strategy was developed. The overall goal of this strategy is to comply with all applicable environmental laws and regulations and/or compliance agreements during PFP stabilization, deactivation, and eventual dismantlement. Significant environmental drivers for the PFP Nuclear Material Stabilization Project include the Tri-Party Agreement; the Resource Conservation and Recovery Act of 1976 (RCRA); the Comprehensive Environmental Response, Compensation and Liability Actmore » of 1980 (CERCLA); the National Environmental Policy Act of 1969 (NEPA); the National Historic Preservation Act (NHPA); the Clean Air Act (CAA), and the Clean Water Act (CWA). Recent TPA negotiation s with Ecology and EPA have resulted in milestones that support the use of CERCLA as the primary statutory framework for decommissioning PFP. Milestones have been negotiated to support the preparation of Engineering Evaluations/Cost Analyses for decommissioning major PFP buildings. Specifically, CERCLA EE/CA(s) are anticipated for the following scopes of work: Settling Tank 241-Z-361, the 232-Z Incinerator, , the process facilities (eg, 234-5Z, 242, 236) and the process facility support buildings. These CERCLA EE/CA(s) are for the purpose of analyzing the appropriateness of the slab-on-grade endpoint Additionally, agreement was reached on performing an evaluation of actions necessary to address below-grade structures or other structures remaining after completion of the decommissioning of PFP. Remaining CERCLA actions will be integrated with other Central Plateau activities at the Hanford site.« less

METHOD FOR SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY SOLVENT EXTRACTION

DOEpatents

Seaborg, G.T.; Blaedel, W.J.; Walling, M.T. Jr.

1960-08-23

A process is given for separating from each other uranium, plutonium, and fission products in an aqueous nitric acid solution by the so-called Redox process. The plutonium is first oxidized to the hexavalent state, e.g., with a water-soluble dichromate or sodium bismuthate, preferably together with a holding oxidant such as potassium bromate. potassium permanganate, or an excess of the oxidizing agent. The solution is then contacted with a water-immiscible organic solvent, preferably hexone. whereby uranium and plutonium are extracted while the fission products remain in the aqueous solution. The separated organic phase is then contacted with an aqueous solution of a reducing agent, with or without a holding reductant (e.g., with a ferrous salt plus hydrazine or with ferrous sulfamate), whereby plutonium is reduced to the trivalent state and back- extracted into the aqueous solution. The uranium may finally be back-extracted from the organic solvent (e.g., with a 0.1 N nitric acid).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robertson, Janeen Denise

In a mortality study of white males who had worked at the Rocky Flats Nuclear Weapons Plant between 1952 and 1979, an increased number of deaths from benign and unspecified intracranial neoplasms was found. A case-control study nested within this cohort investigated the hypothesis that an association existed between brain tumor death and exposure to either internally deposited plutonium or external ionizing radiation. There was no statistically significant association found between estimated radiation exposure from internally deposited plutonium and the development of brain tumors. Exposure by job or work area showed no significant difference between the cohort and the controlmore » groups. An update of the study found elevated risk estimates for (1) all lymphopoietic neoplasms, and (2) all causes of death in employees with body burdens greater than or equal to two nanocuries of plutonium. There was an excess of brain tumors for the entire cohort. Similar cohort studies conducted on worker populations from other plutonium handling facilities have not yet shown any elevated risks for brain tumors. Historically, the Rocky Flats Nuclear Weapons Plant used large quantities of chemicals in their production operations. The use of solvents, particularly carbon tetrachloride, was unique to Rocky Flats. No investigation of the possible confounding effects of chemical exposures was done in the initial studies. The objectives of the present study are to (1) investigate the history of chemical use at the Rocky Flats facility; (2) locate and analyze chemical monitoring information in order to assess employee exposure to the chemicals that were used in the highest volume; and (3) determine the feasibility of establishing a chemical exposure assessment model that could be used in future epidemiology studies.« less

SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

DOEpatents

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 RECOVERY FROM NEUTRON-BOMBARDED URANIUM FUEL

DOEpatents

Moore, R.H.

1964-03-24

A process of recovering plutonium from fuel by dissolution in molten KAlCl/sub 4/ double salt is described. Molten lithium chloride plus stannous chloride is added to reduce plutonium tetrachloride to the trichloride, which is dissolved in a lithium chloride phase while the uranium, as the tetrachloride, is dissolved in a double-salt phase. Separation of the two phases is discussed. (AEC)

PROCESSING OF NEUTRON-IRRADIATED URANIUM

DOEpatents

Hopkins, H.H. Jr.

1960-09-01

An improved "Purex" process for separating uranium, plutonium, and fission products from nitric acid solutions of neutron-irradiated uranium is offered. Uranium is first extracted into tributyl phosphate (TBP) away from plutonium and fission products after adjustment of the acidity from 0.3 to 0.5 M and heating from 60 to 70 deg C. Coextracted plutonium, ruthenium, and fission products are fractionally removed from the TBP by three scrubbing steps with a 0.5 M nitric acid solution of ferrous sulfamate (FSA), from 3.5 to 5 M nitric acid, and water, respectively, and the purified uranium is finally recovered from the TBP by precipitation with an aqueous solution of oxalic acid. The plutonium in the 0.3 to 0.5 M acid solution is oxidized to the tetravalent state with sodium nitrite and extracted into TBP containing a small amount of dibutyl phosphate (DBP). Plutonium is then back-extracted from the TBP-DBP mixture with a nitric acid solution of FSA, reoxidized with sodium nitrite in the aqueous strip solution obtained, and once more extracted with TBP alone. Finally the plutonium is stripped from the TBP with dilute acid, and a portion of the strip solution thus obtained is recycled into the TBPDBP for further purification.

Removal of plutonium from hepatic tissue

DOEpatents

Lindenbaum, Arthur; Rosenthal, Marcia W.

1979-01-01

A method is provided for removing plutonium from hepatic tissues by introducing into the body and blood stream a solution of the complexing agent DTPA and an adjunct thereto. The adjunct material induces aberrations in the hepatic tissue cells and removes intracellularly deposited plutonium which is normally unavailable for complexation with the DTPA. Once the intracellularly deposited plutonium has been removed from the cell by action of the adjunct material, it can be complexed with the DTPA present in the blood stream and subsequently removed from the body by normal excretory processes.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, John P.; Miller, William E.

1989-01-01

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

DOEpatents

Ackerman, J.P.; Miller, W.E.

1987-11-05

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

Transuranic Contamination in Sediment and Groundwater at the U.S. DOE Hanford Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cantrell, Kirk J.

2009-08-20

A review of transuranic radionuclide contamination in sediments and groundwater at the DOE’s Hanford Site was conducted. The review focused primarily on plutonium-239/240 and americium-241; however, other transuranic nuclides were discussed as well, including neptunium-237, plutonium-238, and plutonium-241. The scope of the review included liquid process wastes intentionally disposed to constructed waste disposal facilities such as trenches and cribs, burial grounds, and unplanned releases to the ground surface. The review did not include liquid wastes disposed to tanks or solid wastes disposed to burial grounds. It is estimated that over 11,800 Ci of plutonium-239, 28,700 Ci of americium-241, and 55more » Ci of neptunium-237 have been disposed as liquid waste to the near surface environment at the Hanford Site. Despite the very large quantities of transuranic contaminants disposed to the vadose zone at Hanford, only minuscule amounts have entered the groundwater. Currently, no wells onsite exceed the DOE derived concentration guide for plutonium-239/240 (30 pCi/L) or any other transuranic contaminant in filtered samples. The DOE derived concentration guide was exceeded by a small fraction in unfiltered samples from one well (299-E28-23) in recent years (35.4 and 40.4 pCi/L in FY 2006). The primary reason that disposal of these large quantities of transuranic radionuclides directly to the vadose zone at the Hanford Site has not resulted in widespread groundwater contamination is that under the typical oxidizing and neutral to slightly alkaline pH conditions of the Hanford vadose zone, transuranic radionuclides (plutonium and americium in particular) have a very low solubility and high affinity for surface adsorption to mineral surfaces common within the Hanford vadose zone. Other important factors are the fact that the vadose zone is typically very thick (hundreds of feet) and the net infiltration rate is very low due to the desert climate. In some cases where transuranic radionuclides have been co-disposed with acidic liquid waste, transport through the vadose zone for considerable distances has occurred. For example, at the 216-Z-9 Crib, plutonium-239 and americium-241 have moved to depths in excess of 36 m (118 ft) bgs. Acidic conditions increase the solubility of these contaminants and reduce adsorption to mineral surfaces. Subsequent neutralization of the acidity by naturally occurring calcite in the vadose zone (particularly in the Cold Creek unit) appears to have effectively stopped further migration. The vast majority of transuranic contaminants disposed to the vadose zone on the Hanford Site (10,200 Ci [86%] of plutonium-239; 27,900 Ci [97%] of americium-241; and 41.8 Ci [78%] of neptunium-237) were disposed in sites within the PFP Closure Zone. This closure zone is located within the 200 West Area (see Figures 1.1 and 3.1). Other closure zones with notably high quantities of transuranic contaminant disposal include the T Farm Zone with 408 Ci (3.5%) plutonium-239, the PUREX Zone with 330 Ci (2.8%) plutonium-239, 200-W Ponds Zone with 324 Ci (2.8%) plutonium-239, B Farm Zone with 183 Ci (1.6%) plutonium-239, and the REDOX Zone with 164 Ci (1.4%) plutonium 239. Characterization studies for most of the sites reviewed in the document are generally limited. The most prevalent characterization methods used were geophysical logging methods. Characterization of a number of sites included laboratory analysis of borehole sediment samples specifically for radionuclides and other contaminants, and geologic and hydrologic properties. In some instances, more detailed research level studies were conducted. Results of these studies were summarized in the document.« less

Plutonium recovery from spent reactor fuel by uranium displacement

DOEpatents

Ackerman, John P.

1992-01-01

A process 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.

12. Architectural Floor Plans, 233S, U.S. Atomic Energy Commission, Hanford ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

12. Architectural Floor Plans, 233-S, U.S. Atomic Energy Commission, Hanford Atomic Products Operations, General Electric Company, Dwg. H-2-30464, 1956. - Reduction-Oxidation Complex, Plutonium Concentration Facility, 200 West Area, Richland, Benton County, WA

11. Architectural ELevations & Sections, 233S, U.S. Atomic Energy Commission, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

11. Architectural ELevations & Sections, 233-S, U.S. Atomic Energy Commission, Hanford Atomic Products Operations, General Electric Company, Dwg. No. H-2-30465, 1956. - Reduction-Oxidation Complex, Plutonium Concentration Facility, 200 West Area, Richland, Benton County, WA

REDUCTION IN Pu RECOVERY PROCESSES

DOEpatents

Ritter, D.M.; Black, R.P.S.

1959-09-29

A method is described for reducing plutonium from the hexavalent to the tetravalent state in a carrier precipitation process for separating plutonium and nuclear fission products. In accordance with the invention oxalate ions are incorporated in the hexavalent plutoniumcontaining solution prior to a step of precipitating lanthanum fluoride in the solution.

PLUTONIUM PURIFICATION PROCESS EMPLOYING THORIUM PYROPHOSPHATE CARRIER

DOEpatents

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.

An Update on the Status of the Supply of Plutonium-238 for Future NASA Missions

NASA Astrophysics Data System (ADS)

Wham, R. M.

2016-12-01

For more than five decades, Radioisotope Power Systems (RPSs) have enabled space missions to operate in locations where the Sun's intensity is too weak, obscured, or otherwise inadequate for solar power or other conventional power‒generation technologies. The natural decay heat (0.57 W/g) from the radioisotope, plutonium-238 (238Pu), provides the thermal energy source used by an RPS to generate electricity for operation of instrumentation, as well as heat to keep key subsystems warm for missions such as Voyagers 1 and 2, the Cassini mission to Saturn, the New Horizons flyby of Pluto, and the Mars Curiosity rover which were sponsored by the National Aeronautics and Space Administration (NASA). Plutonium-238 is produced by irradiation of neptunium-237 in a nuclear reactor a relatively high neutron flux. The United States has not produced new quantities of 238Pu since the early 1990s. RPS‒powered missions have continued since then using existing 238Pu inventory managed by the U.S. Department of Energy (DOE), including material purchased from Russia. A new domestic supply is needed to ensure the continued availability of RPSs for future NASA missions. NASA and DOE are currently executing a project to reestablish a 238Pu supply capability using its existing facilities and reactors, which are much smaller than the large-scale production reactors and processing canyon equipment used previously. The project is led by the Oak Ridge National Laboratory (ORNL). Target rods, containing NpO2, will be fabricated at ORNL and irradiated in the ORNL High Flux Isotope Reactor and the Advanced Test Reactor at Idaho National Laboratory. Irradiated targets will be processed in chemical separations at the ORNL Radiochemical Engineering Center to recover the plutonium product and unconverted neptunium for recycle. The 238PuO2 product will be shipped to Los Alamos National Laboratory for fabrication of heat source pellets. Key activities, such as transport of the neptunium to ORNL, irradiation of neptunium, and chemical processing to recover the newly generated 238Pu, have begun and have been demonstrated with the initial amounts (50-100 g) produced. Product samples have been shipped to LANL for evaluation, including chemical impurity analysis. This paper will provide an overview of the approach to the project and its progress to date.

Integrating workplace exposure databases for occupational medicine services and epidemiologic studies at a former nuclear weapons facility.

PubMed

Ruttenber, A J; McCrea, J S; Wade, T D; Schonbeck, M F; LaMontagne, A D; Van Dyke, M V; Martyny, J W

2001-02-01

We outline methods for integrating epidemiologic and industrial hygiene data systems for the purpose of exposure estimation, exposure surveillance, worker notification, and occupational medicine practice. We present examples of these methods from our work at the Rocky Flats Plant--a former nuclear weapons facility that fabricated plutonium triggers for nuclear weapons and is now being decontaminated and decommissioned. The weapons production processes exposed workers to plutonium, gamma photons, neutrons, beryllium, asbestos, and several hazardous chemical agents, including chlorinated hydrocarbons and heavy metals. We developed a job exposure matrix (JEM) for estimating exposures to 10 chemical agents in 20 buildings for 120 different job categories over a production history spanning 34 years. With the JEM, we estimated lifetime chemical exposures for about 12,000 of the 16,000 former production workers. We show how the JEM database is used to estimate cumulative exposures over different time periods for epidemiological studies and to provide notification and determine eligibility for a medical screening program developed for former workers. We designed an industrial hygiene data system for maintaining exposure data for current cleanup workers. We describe how this system can be used for exposure surveillance and linked with the JEM and databases on radiation doses to develop lifetime exposure histories and to determine appropriate medical monitoring tests for current cleanup workers. We also present time-line-based graphical methods for reviewing and correcting exposure estimates and reporting them to individual workers.

10. Architectural Door Details & Plot Plan, 233S, U.S. Atomic ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Architectural Door Details & Plot Plan, 233-S, U.S. Atomic Energy Commission, Hanford Atomic Products Operations, General Electric Company, Dwg. No. H-2-30469, 1956. - Reduction-Oxidation Complex, Plutonium Concentration Facility, 200 West Area, Richland, Benton County, WA

Looking inside plutonium

Science.gov Websites

Los Alamos National Laboratory Search Site submit About Mission Business Newsroom Publications Los Innovation in New Mexico Los Alamos Collaboration for Explosives Detection (LACED) SensorNexus Exascale Computing Project (ECP) User Facilities Center for Integrated Nanotechnologies (CINT) Los Alamos Neutron

The Manhattan Project

Science.gov Websites

Short History of Oak Ridge National Laboratory (1943 - 1993) Los Alamos, New Mexico * Selected as Atomic Weapons Laboratory * Site Selection * History @ Los Alamos Hanford, Washington * Selected as Plutonium Production Facility * History of the Hanford Site 1943 - 1990 Chicago, Illinois * Promethean Boldness at

8. Front (east) side of incinerator and glove boxes. Ash ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

8. Front (east) side of incinerator and glove boxes. Ash canning hood to the left, combustion chamber in the middle, incinerator hood to the right. Looking west. - Plutonium Finishing Plant, Waste Incinerator Facility, 200 West Area, Richland, Benton County, WA

Space Science

NASA Image and Video Library

1997-07-18

Jet Propulsion Research Lab (JPL) workers use a borescope to verify the pressure relief device bellow's integrity on a radioisotope thermoelectric generator (RTG) that has been installed on the Cassini spacecraft in the Payload Hazardous Servicing Facility. The activity is part of the mechanical and electrical verification testing of RTGs during prelaunch processing. RTGs use heat from the natural decay of plutonium to generate electrical power. The three RTGs on Cassini will enable the spacecraft to operate far from the Sun where solar power systems are not feasible. They will provide electrical power to Cassini on it seven year trip to the Saturnian system and during its four year mission at Saturn.

FUSED SALT PROCESS FOR RECOVERY OF VALUES FROM USED NUCLEAR REACTOR FUELS

DOEpatents

Moore, R.H.

1960-08-01

A process is given for recovering plutonium from a neutron-irradiated uranium mass (oxide or alloy) by dissolving the mass in an about equimolar alkali metalaluminum double chloride, adding aluminum metal to the mixture obtained at a temperature of between 260 and 860 deg C, and separating a uranium-containing metal phase and a plutonium-chloride- and fission-product chloridecontaining salt phase. Dissolution can be expedited by passing carbon tetrachloride vapors through the double salt. Separation without reduction of plutonium from neutron- bombarded uranium and that of cerium from uranium are also discussed.

CARBONATE METHOD OF SEPARATION OF TETRAVALENT PLUTONIUM FROM FISSION PRODUCT VALUES

DOEpatents

Duffield, R.B.; Stoughton, R.W.

1959-02-01

It has been found that plutonium forms an insoluble precipitate with carbonate ion when the carbonate ion is present in stoichiometric proportions, while an excess of the carbonate ion complexes plutonium and renders it soluble. A method for separating tetravalent plutonium from lanthanum-group rare earths has been based on this discovery, since these rare earths form insoluble carbonates in approximately neutral solutions. According to the process the pH is adjusted to between 5 and 7, and approximately stoichiometric amounts of carbonate ion are added to the solution causing the formation of a precipitate of plutonium carbonate and the lanthanum-group rare earth carbonates. The precipitate is then separated from the solution and contacted with a carbonate solution of a concentration between 1 M and 3 M to complex and redissolve the plutonium precipitate, and thus separate it from the insoluble rare earth precipitate.

River Corridor Cleanup Contract Fiscal Year 2006 Detailed Work Plan: D4 Project/Reactor ISS Closure Projects Field Remediation Project Waste Operations Project End State and Final Closure Project Mission/General Support, Volume 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Project Integration

2005-09-26

The Hanford Site contains many surplus facilities and waste sites that remain from plutonium production activities. These contaminated facilities and sites must either be stabilized and maintained, or removed, to prevent the escape of potentially hazardous contaminants into the environment and exposure to workers and the public.

Fire hazard analysis for Plutonium Finishing Plant complex

DOE Office of Scientific and Technical Information (OSTI.GOV)

MCKINNIS, D.L.

1999-02-23

A fire hazards analysis (FHA) was performed for the Plutonium Finishing Plant (PFP) Complex at the Department of Energy (DOE) Hanford site. The scope of the FHA focuses on the nuclear facilities/structures in the Complex. The analysis was conducted in accordance with RLID 5480.7, [DOE Directive RLID 5480.7, 1/17/94] and DOE Order 5480.7A, ''Fire Protection'' [DOE Order 5480.7A, 2/17/93] and addresses each of the sixteen principle elements outlined in paragraph 9.a(3) of the Order. The elements are addressed in terms of the fire protection objectives stated in paragraph 4 of DOE 5480.7A. In addition, the FHA also complies with WHC-CM-4-41,more » Fire Protection Program Manual, Section 3.4 [1994] and WHC-SD-GN-FHA-30001, Rev. 0 [WHC, 1994]. Objectives of the FHA are to determine: (1) the fire hazards that expose the PFP facilities, or that are inherent in the building operations, (2) the adequacy of the fire safety features currently located in the PFP Complex, and (3) the degree of compliance of the facility with specific fire safety provisions in DOE orders, related engineering codes, and standards.« less

DISSOLUTION OF PLUTONIUM CONTAINING CARRIER PRECIPITATE BY CARBONATE METATHESIS AND SEPARATION OF SULFIDE IMPURITIES THEREFROM BY SULFIDE PRECIPITATION

DOEpatents

Duffield, R.B.

1959-07-14

A process is described for recovering plutonium from foreign products wherein a carrier precipitate of lanthanum fluoride containing plutonium is obtained and includes the steps of dissolving the carrier precipitate in an alkali metal carbonate solution, adding a soluble sulfide, separating the sulfide precipitate, adding an alkali metal hydroxide, separating the resulting precipitate, washing, and dissolving in a strong acid.

Update on the Department of Energy's 1994 plutonium vulnerability assessment for the plutonium finishing plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

HERZOG, K.R.

1999-09-01

A review of the environmental, safety, and health vulnerabilities associated with the continued storage of PFP's inventory of plutonium bearing materials and other SNM. This report re-evaluates the five vulnerabilities identified in 1994 at the PFP that are associated with SNM storage. This new evaluation took a more detailed look and applied a risk ranking process to help focus remediation efforts.

METHOD FOR DISSOLVING LANTHANUM FLUORIDE CARRIER FOR PLUTONIUM

DOEpatents

Koshland, D.E. Jr.; Willard, J.E.

1961-08-01

A method is described for dissolving lanthanum fluoride precipitates which is applicable to lanthanum fluoride carrier precipitation processes for recovery of plutonium values from aqueous solutions. The lanthanum fluoride precipitate is contacted with an aqueous acidic solution containing dissolved zirconium in the tetravalent oxidation state. The presence of the zirconium increases the lanthanum fluoride dissolved and makes any tetravalent plutonium present more readily oxidizable to the hexavalent state. (AEC)

METHOD OF PREPARING PLUTONIUM TETRAFLUORIDE

DOEpatents

Beede, R.L.; Hopkins, H.H. Jr.

1959-11-17

C rystalline plutonium tetrafluoride is precipitated from aqueous up to 1.6 N mineral acid solutions of a plutorium (IV) salt with fluosilicic acid anions, preferably at room temperature. Hydrogen fluoride naay be added after precipitation to convert any plutonium fluosilicate to the tetrafluoride and any silica to fluosilicic acid. This process results in a purer product, especially as to iron and aluminum, than does the precipitation by the addition of hydrogen fluoride.

Plutonium recovery from spent reactor fuel by uranium displacement

DOEpatents

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.

Dealing With Russian Tactical Nuclear Weapons

DTIC Science & Technology

2004-01-01

the plants, facilities, and equipment necessary to safely secure and store weapons pits (the plutonium “triggers” at the center of a thermonuclear ... bomb , and hence the most critical piece) proffered by either party. Furthermore, the statute requires the IAEA to establish control over the pits until

Uranium and Thorium

ERIC Educational Resources Information Center

Finch, Warren I.

1978-01-01

The results of President Carter's policy on non-proliferation of nuclear weapons are expected to slow the growth rate in energy consumption, put the development of the breeder reactor in question, halt plans to reprocess and recycle uranium and plutonium, and expand facilities to supply enriched uranium. (Author/MA)

Digital pile-up rejection for plutonium experiments with solution-grown stilbene

NASA Astrophysics Data System (ADS)

Bourne, M. M.; Clarke, S. D.; Paff, M.; DiFulvio, A.; Norsworthy, M.; Pozzi, S. A.

2017-01-01

A solution-grown stilbene detector was used in several experiments with plutonium samples including plutonium oxide, mixed oxide, and plutonium metal samples. Neutrons from different reactions and plutonium isotopes are accompanied by numerous gamma rays especially by the 59-keV gamma ray of 241Am. Identifying neutrons correctly is important for nuclear nonproliferation applications and makes neutron/gamma discrimination and pile-up rejection necessary. Each experimental dataset is presented with and without pile-up filtering using a previously developed algorithm. The experiments were simulated using MCNPX-PoliMi, a Monte Carlo code designed to accurately model scintillation detector response. Collision output from MCNPX-PoliMi was processed using the specialized MPPost post-processing code to convert neutron energy depositions event-by-event into light pulses. The model was compared to experimental data after pulse-shape discrimination identified waveforms as gamma ray or neutron interactions. We show that the use of the digital pile-up rejection algorithm allows for accurate neutron counting with stilbene to within 2% even when not using lead shielding.

Plutonium Decontamination of Uranium using CO2 Cleaning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blau, M

A concern of the Department of Energy (DOE) Environmental Management (EM) and Defense Programs (DP), and of the Los Alamos National Laboratory (LANL) and the Lawrence Livermore National Laboratory (LLNL), is the disposition of thousands of legacy and recently generated plutonium (Pu)-contaminated, highly enriched uranium (HEU) parts. These parts take up needed vault space. This presents a serious problem for LLNL, as site limit could result in the stoppage of future weapons work. The Office of Fissile Materials Disposition (NN-60) will also face a similar problem as thousands of HEU parts will be created with the disassembly of site-return pitsmore » for plutonium recovery when the Pit Disassembly and Conversion Facility (PDCF) at the Savannah River Site (SRS) becomes operational. To send HEU to the Oak Ridge National Laboratory and the Y-12 Plant for disposition, the contamination for metal must be less than 20 disintegrations per minute (dpm) of swipable transuranic per 100 cm{sup 2} of surface area or the Pu bulk contamination for oxide must be less than 210 parts per billion (ppb). LANL has used the electrolytic process on Pu-contaminated HEU weapon parts with some success. However, this process requires that a different fixture be used for every configuration; each fixture cost approximately $10K. Moreover, electrolytic decontamination leaches the uranium metal substrate (no uranium or plutonium oxide) from the HEU part. The leaching rate at the uranium metal grain boundaries is higher than that of the grains and depends on the thickness of the uranium oxide layer. As the leaching liquid flows past the HEU part, it carries away plutonium oxide contamination and uranium oxide. The uneven uranium metal surface created by the leaching becomes a trap for plutonium oxide contamination. In addition, other DOE sites have used CO{sub 2} cleaning for Pu decontamination successfully. In the 1990's, the Idaho National Engineering Laboratory investigated this technology and showed that CO{sub 2} pellet blasting (or CO{sub 2} cleaning) reduced both fixed and smearable contamination on tools. In 1997, LLNL proved that even tritium contamination could be removed from a variety of different matrices using CO{sub 2}cleaning. CO{sub 2} cleaning is a non-toxic, nonconductive, nonabrasive decontamination process whose primary cleaning mechanisms are: (1) Impact of the CO{sub 2} pellets loosens the bond between the contaminant and the substrate. (2) CO{sub 2} pellets shatter and sublimate into a gaseous state with large expansion ({approx}800 times). The expanding CO{sub 2} gas forms a layer between the contaminant and the substrate that acts as a spatula and peels off the contaminant. (3) Cooling of the contaminant assists in breaking its bond with the substrate. Thus, LLNL conducted feasibility testing to determine if CO{sub 2} pellet blasting could remove Pu contamination (e.g., uranium oxide) from uranium metal without abrading the metal matrix. This report contains a summary of events and the results of this test.« less

Analysis of plutonium isotope ratios including 238Pu/239Pu in individual U-Pu mixed oxide particles by means of a combination of alpha spectrometry and ICP-MS.

PubMed

Esaka, Fumitaka; Yasuda, Kenichiro; Suzuki, Daisuke; Miyamoto, Yutaka; Magara, Masaaki

2017-04-01

Isotope ratio analysis of individual uranium-plutonium (U-Pu) mixed oxide particles contained within environmental samples taken from nuclear facilities is proving to be increasingly important in the field of nuclear safeguards. However, isobaric interferences, such as 238 U with 238 Pu and 241 Am with 241 Pu, make it difficult to determine plutonium isotope ratios in mass spectrometric measurements. In the present study, the isotope ratios of 238 Pu/ 239 Pu, 240 Pu/ 239 Pu, 241 Pu/ 239 Pu, and 242 Pu/ 239 Pu were measured for individual Pu and U-Pu mixed oxide particles by a combination of alpha spectrometry and inductively coupled plasma mass spectrometry (ICP-MS). As a consequence, we were able to determine the 240 Pu/ 239 Pu, 241 Pu/ 239 Pu, and 242 Pu/ 239 Pu isotope ratios with ICP-MS after particle dissolution and chemical separation of plutonium with UTEVA resins. Furthermore, 238 Pu/ 239 Pu isotope ratios were able to be calculated by using both the 238 Pu/( 239 Pu+ 240 Pu) activity ratios that had been measured through alpha spectrometry and the 240 Pu/ 239 Pu isotope ratios determined through ICP-MS. Therefore, the combined use of alpha spectrometry and ICP-MS is useful in determining plutonium isotope ratios, including 238 Pu/ 239 Pu, in individual U-Pu mixed oxide particles. Copyright © 2016 Elsevier B.V. All rights reserved.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, Xiangdong; Einziger, Robert E.

1997-01-01

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-08-12

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-01-28

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

NASA Astrophysics Data System (ADS)

Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

2016-08-01

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel assemblies. Based on analysis of modern engineering solutions on SNF regeneration, it has been concluded that new reprocessing technologies should meet the ecological safety requirements, provide a more extensive use of the resource base of nuclear engineering, allow the production of valuable and trace elements on an industrial scale, and decrease radioactive waste release.

Simulation of uranium and plutonium oxides compounds obtained in plasma

NASA Astrophysics Data System (ADS)

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.

PROCESS OF REDUCING PLUTONIUM TO TETRAVALENT TRIVALENT STATE

DOEpatents

Mastick, D.F.

1960-05-10

The reduction of hexavalent and tetravalert plutonium ions to the trivalent state in strong nitric acid can be accomplished with hydrogen peroxide. The trivalent state may be stabilized as a precipitate by including oxalate or fluoride ions in the solution. The acid should be strong to encourage the reduction from the plutonyl to the trivalent state (and discourage the opposed oxidation reaction) and prevent the precipitation of plutonium peroxide, although the latter may be digested by increasing the acid concentration. Although excess hydrogen peroxide will oxidize plutonlum to the plutonyl state, complete reduction is insured by gently warming the solution to break down such excess H/ sub 2/O/sub 2/. The particular advantage of hydrogen peroxide as a reductant lies in the precipitation technique, where it introduces no contaminating ions. The process is adaptable to separate plutonium from uranium and impurities by proper adjustment of the sequence of insoluble anion additions and the hydrogen peroxide addition.

Determination of the 240Pu/ 239Pu atomic ratio in soils from Palomares (Spain) by low-energy accelerator mass spectrometry

NASA Astrophysics Data System (ADS)

Chamizo, E.; García-León, M.; Synal, H.-A.; Suter, M.; Wacker, L.

2006-08-01

In 1966, the nuclear fuel of two thermonuclear bombs was released over the Spanish region of Palomares, due to a B52 bomber accident during a refuelling operation. Since then, much effort has been made to assess its impact to the different environmental compartments of this area in South-East Spain, mostly by measuring the 239+240Pu activity concentration and the 238Pu/239+240Pu activity ratio. Nevertheless, these measurements do not give enough information on the problem. In order to recognize unambiguously small traces of the weapon-grade plutonium released in the accident, the ratio of the two major isotopes of plutonium, 240Pu/239Pu, has to be determined. In this work, this ratio has been measured in low- and high-activity samples from Palomares by means of low-energy accelerator mass spectrometry (AMS). That way, we will show the potential of the new generation of compact AMS facilities in terms of plutonium characterization at ultra-trace levels.

Engineering study of 50 miscellaneous inactive underground radioactive waste tanks located at the Hanford Site, Washington

DOE Office of Scientific and Technical Information (OSTI.GOV)

Freeman-Pollard, J.R.

1994-03-02

This engineering study addresses 50 inactive underground radioactive waste tanks. The tanks were formerly used for the following functions associated with plutonium and uranium separations and waste management activities in the 200 East and 200 West Areas of the Hanford Site: settling solids prior to disposal of supernatant in cribs and a reverse well; neutralizing acidic process wastes prior to crib disposal; receipt and processing of single-shell tank (SST) waste for uranium recovery operations; catch tanks to collect water that intruded into diversion boxes and transfer pipeline encasements and any leakage that occurred during waste transfer operations; and waste handlingmore » and process experimentation. Most of these tanks have not been in use for many years. Several projects have, been planned and implemented since the 1970`s and through 1985 to remove waste and interim isolate or interim stabilize many of the tanks. Some tanks have been filled with grout within the past several years. Responsibility for final closure and/or remediation of these tanks is currently assigned to several programs including Tank Waste Remediation Systems (TWRS), Environmental Restoration and Remedial Action (ERRA), and Decommissioning and Resource Conservation and Recovery Act (RCRA) Closure (D&RCP). Some are under facility landlord responsibility for maintenance and surveillance (i.e. Plutonium Uranium Extraction [PUREX]). However, most of the tanks are not currently included in any active monitoring or surveillance program.« less

METHOD OF MAINTAINING PLUTONIUM IN A HIGHER STATE OF OXIDATION DURING PROCESSING

DOEpatents

Thompson, S.G.; Miller, D.R.

1959-06-30

This patent deals with the oxidation of tetravalent plutonium contained in an aqueous acid solution together with fission products to the hexavalent state, prior to selective fission product precipitation, by adding to the solution bismuthate or ceric ions as the oxidant and a water-soluble dichromate as a holding oxidant. Both oxidant and holding oxidant are preferably added in greater than stoichiometric quantities with regard to the plutonium present.

SEPARATION OF PLUTONIUM VALUES FROM OTHER METAL VALUES IN AQUEOUS SOLUTIONS BY SELECTIVE COMPLEXING AND ADSORPTION

DOEpatents

Beaton, R.H.

1960-06-28

A process is given for separating tri- or tetravalent plutonium from fission products in an aqueous solution by complexing the fission products with oxalate, tannate, citrate, or tartrate anions at a pH value of at least 2.4 (preferably between 2.4 and 4), and contacting a cation exchange resin with the solution whereby the plutonium is adsorbed while the complexed fission products remain in solution.

Aqueous Chloride Operations Overview: Plutonium and Americium Purification/Recovery

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gardner, Kyle Shelton; Kimball, David Bryan; Skidmore, Bradley Evan

These are a set of slides intended for an information session as part of recruiting activities at Brigham Young University. It gives an overview of aqueous chloride operations, specifically on plutonium and americium purification/recovery. This presentation details the steps taken perform these processes, from plutonium size reduction, dissolution, solvent extraction, oxalate precipitation, to calcination. For americium recovery, it details the CLEAR (chloride extraction and actinide recovery) Line, oxalate precipitation and calcination.

PURIFICATION OF PLUTONIUM USING A CERIUM PRECIPITATE AS A CARRIER FOR FISSION PRODUCTS

DOEpatents

Faris, B.F.; Olson, C.M.

1961-07-01

Bismuth phosphate carrier precipitation processes are described for the separation of plutonium from fission products wherein in at least one step bismuth phosphate is precipitated in the presence of hexavalent plutonium thereby carrying a portion of the fission products from soluble plu tonium values. In this step, a cerium phosphate precipitate is formed in conjunction with the bismuth phosphate precipitate, thereby increasing the amount of fission products removed from solution.

6. VIEW OF BUILDING 371 EXTERIOR WALL CONSTRUCTION DETAIL. BUILDING ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

6. VIEW OF BUILDING 371 EXTERIOR WALL CONSTRUCTION DETAIL. BUILDING CONSTRUCTION WAS HARDENED TO WITHSTAND THE FORCES IMPOSED BY A DESIGN-BASIS EARTHQUAKE OR TORNADO. (7/1/74) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

13. VIEW OF VACUUM CHAMBER AND WELDING EQUIPMENT IN MODULE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

13. VIEW OF VACUUM CHAMBER AND WELDING EQUIPMENT IN MODULE E. PARTS WERE WELDED UNDER A VACUUM TO PREVENT CORROSION. (11/6/73) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

18. VIEW OF THE SECOND FLOOR OF BUILDING 707. AIR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

18. VIEW OF THE SECOND FLOOR OF BUILDING 707. AIR EXHAUST FANS ARE USED TO MAINTAIN PRESSURE DIFFERENTIALS WITHIN THE BUILDING. (5/70) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

METHOD OF RECOVERING PLUTONIUM VALUES FROM AQUEOUS SOLUTIONS BY CARRIER PRECIPITATION

DOEpatents

James, R.A.; Thompson, S.G.

1959-11-01

A process is presented for pretreating aqueous nitric acid- plutonium solutions containing a small quantity of hydrazine that has formed as a decomposition product during the dissolution of neutron-bombarded uranium in nitric acid and that impairs the precipitation of plutonium on bismuth phosphate. The solution is digested with alkali metal dichromate or potassium permanganate at between 75 and 100 deg C; sulfuric acid at approximately 75 deg C and sodium nitrate, oxaiic acid plus manganous nitrate, or hydroxylamine are added to the solution to secure the plutonium in the tetravalent state and make it suitable for precipitation on BiPO/sub 4/.

Decrease the Number of Glovebox Glove Breaches and Failures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hurtle, Jackie C.

2013-12-24

Los Alamos National Laboratory (LANL) is committed to the protection of the workers, public, and environment while performing work and uses gloveboxes as engineered controls to protect workers from exposure to hazardous materials while performing plutonium operations. Glovebox gloves are a weak link in the engineered controls and are a major cause of radiation contamination events which can result in potential worker exposure and localized contamination making operational areas off-limits and putting programmatic work on hold. Each day of lost opportunity at Technical Area (TA) 55, Plutonium Facility (PF) 4 is estimated at $1.36 million. Between July 2011 and Junemore » 2013, TA-55-PF-4 had 65 glovebox glove breaches and failures with an average of 2.7 per month. The glovebox work follows the five step safety process promoted at LANL with a decision diamond interjected for whether or not a glove breach or failure event occurred in the course of performing glovebox work. In the event that no glove breach or failure is detected, there is an additional decision for whether or not contamination is detected. In the event that contamination is detected, the possibility for a glove breach or failure event is revisited.« less

PREPARATION OF PuF$sub 3$

DOEpatents

Benz, R.

1964-03-01

A process for preparing pure plutonium trifluoride is described in which a refractory plutonium compound is contacted with ammonium fluoride in a closed container at a pressure of at least 10 atmospheres and a temperature of about 550 deg C. (AEC)

Modeling of selected ceramic processing parameters employed in the fabrication of 238PuO 2 fuel pellets

DOE PAGES

Brockman, R. A.; Kramer, D. P.; Barklay, C. D.; ...

2011-10-01

Recent deep space missions utilize the thermal output of the radioisotope plutonium-238 as the fuel in the thermal to electrical power system. Since the application of plutonium in its elemental state has several disadvantages, the fuel employed in these deep space power systems is typically in the oxide form such as plutonium-238 dioxide ( 238PuO 2). As an oxide, the processing of the plutonium dioxide into fuel pellets is performed via ''classical'' ceramic processing unit operations such as sieving of the powder, pressing, sintering, etc. Modeling of these unit operations can be beneficial in the understanding and control of processingmore » parameters with the goal of further enhancing the desired characteristics of the 238PuO 2 fuel pellets. A finite element model has been used to help identify the time-temperature-stress profile within a pellet during a furnace operation taking into account that 238PuO 2 itself has a significant thermal output. The results of the modeling efforts will be discussed.« less

PLUTONIUM CARRIER METATHESIS WITH ORGANIC REAGENT

DOEpatents

Thompson, S.G.

1958-07-01

A method is described for converting a plutonium containing bismuth phosphate carrier precipitate Into a compositton more readily soluble in acid. The method consists of dissolving the bismuth phosphate precipitate in an aqueous solution of alkali metal hydroxide, and adding one of a certaia group of organic compounds, e.g., polyhydric alcohols or a-hydrorycarboxylic acids. The mixture is then heated causiing formation of a bismuth hydroxide precipitate containing plutonium which may be readily dissolved in nitric acid for further processing.

16. VIEW OF THE STATIONARY OPERATING ENGINEER CONTROL PANEL INSTALLATION. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

16. VIEW OF THE STATIONARY OPERATING ENGINEER CONTROL PANEL INSTALLATION. THE PANEL CONTROLS AIR-HANDLING EQUIPMENT AND AIR PRESSURE WITHIN THE BUILDING. (10/6/69) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

CONCENTRATION AND DECONTAMINATION OF SOLUTIONS CONTAINING PLUTONIUM VALUES BY BISMUTH PHOSPHATE CARRIER PRECIPITATION METHODS

DOEpatents

Seaborg, G.T.; Thompson, S.G.

1960-08-23

A process is given for isolating plutonium present in the tetravalent state in an aqueous solution together with fission products. First, the plutonium and fission products are coprecipitated on a bismuth phosphate carrier. The precipitate obtained is dissolved, and the plutonium in the solution is oxidized to the hexavalent state (with ceric nitrate, potassium dichromate, Pb/ sub 3/O/sub 4/, sodium bismuthate and/or potassium dichromate). Thereafter a carrier for fission products is added (bismuth phosphate, lanthanum fluoride, ceric phosphate, bismuth oxalate, thorium iodate, or thorium oxalate), and the fission-product precipitation can be repeated with one other of these carriers. After removal of the fission-product-containing precipitate or precipitates. the plutonium in the supernatant is reduced to the tetravalent state (with sulfur dioxide, hydrogen peroxide. or sodium nitrate), and a carrier for tetravalent plutonium is added (lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ceric phosphate, thorium iodate, thorium oxalate, bismuth oxalate, or niobium pentoxide). The plutonium-containing precipitate is then dissolved in a relatively small volume of liquid so as to obtain a concentrated solution. Prior to dissolution, the bismuth phosphate precipitates first formed can be metathesized with a mixture of sodium hydroxide and potassium carbonate and plutonium-containing lanthanum fluorides with alkali-metal hydroxide. In the solutions formed from a plutonium-containing lanthanum fluoride carrier the plutonium can be selectively precipitated with a peroxide after the pH was adjusted preferably to a value of between 1 and 2. Various combinations of second, third, and fourth carriers are discussed.

Tracking of Nuclear Production using Indigenous Species: Final LDRD Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alam, Todd Michael; Alam, Mary Kathleen; McIntyre, Sarah K.

Our LDRD research project sought to develop an analytical method for detection of chemicals used in nuclear materials processing. Our approach is distinctly different than current research involving hardware-based sensors. By utilizing the response of indigenous species of plants and/or animals surrounding (or within) a nuclear processing facility, we propose tracking 'suspicious molecules' relevant to nuclear materials processing. As proof of concept, we have examined TBP, tributylphosphate, used in uranium enrichment as well as plutonium extraction from spent nuclear fuels. We will compare TBP to the TPP (triphenylphosphate) analog to determine the uniqueness of the metabonomic response. We show thatmore » there is a unique metabonomic response within our animal model to TBP. The TBP signature can further be delineated from that of TPP. We have also developed unique methods of instrumental transfer for metabonomic data sets.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pool, Karl N.; Minette, Michael J.; Wahl, Jon H.

On September 28, 2015, debris collected from the PRF (236-Z) canyon floor, Pan J, was observed to exhibit chemical reaction. The material had been transferred from the floor pan to a collection tray inside the canyon the previous Friday. Work in the canyon was stopped to allow Industrial Hygiene to perform monitoring of the material reaction. Canyon floor debris that had been sealed out was sequestered at the facility, a recovery plan was developed, and drum inspections were initiated to verify no additional reactions had occurred. On October 13, in-process drums containing other Pan J material were inspected and showedmore » some indication of chemical reaction, limited to discoloration and degradation of inner plastic bags. All Pan J material was sealed back into the canyon and returned to collection trays. Based on the high airborne levels in the canyon during physical debris removal, ETGS (Encapsulation Technology Glycerin Solution) was used as a fogging/lock-down agent. On October 15, subject matter experts confirmed a reaction had occurred between nitrates (both Plutonium Nitrate and Aluminum Nitrate Nonahydrate (ANN) are present) in the Pan J material and the ETGS fixative used to lower airborne radioactivity levels during debris removal. Management stopped the use of fogging/lock-down agents containing glycerin on bulk materials, declared a Management Concern, and initiated the Potential Inadequacy in the Safety Analysis determination process. Additional drum inspections and laboratory analysis of both reacted and unreacted material are planned. This report compiles the results of many different sample analyses conducted by the Pacific Northwest National Laboratory on samples collected from the Plutonium Reclamation Facility (PRF) floor pans by the CH2MHill’s Plateau Remediation Company (CHPRC). Revision 1 added Appendix G that reports the results of the Gas Generation Rate and methodology. The scope of analyses requested by CHPRC includes the determination of common anions, gamma spectrometry, metals, corrosivity, organics and alpha spectrometry (note: alpha spectrometry was cancelled during the performance of this work with concurrence from CHPRC). Results may help elucidate the components that led to the unexpected reaction in the canyon as well as inform the radiological and hazardous characteristics. The specific anions, gamma emitters, organics and metals requested by CHPRC are provided in the analytical reports sections. The individual analyses were conducted under the Plutonium Finishing Plant (PFP) Floor Pan Evaluation Project Quality Assurance Project Plan (PFP Floor Pan Evaluation QAPP, Revision 0.) developed by PNNL specifically for this project. The final reports for each analysis set are included in this compilation of the results. Each package was reviewed under the PFP Floor Pan Evaluation Project Quality Assurance Project Plan so no additional reviews were conducted in this compilation task. The Gas Generation Rates in Appendix G were conducted under the PNNL “How Do I…” quality assurance program and were NOT conducted under the Plutonium Finishing Plant (PFP) Floor Pan Evaluation Project Quality Assurance Project Plan (PFP Floor Pan Evaluation QAPP, Revision 0.).« less

Transportability Class of Americium in K Basin Sludge under Ambient and Hydrothermal Processing Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Delegard, Calvin H.; Schmitt, Bruce E.; Schmidt, Andrew J.

2006-08-01

This report establishes the technical bases for using a ''slow uptake'' instead of a ''moderate uptake'' transportability class for americium-241 (241Am) for the K Basin Sludge Treatment Project (STP) dose consequence analysis. Slow uptake classes are used for most uranium and plutonium oxides. A moderate uptake class has been used in prior STP analyses for 241Am based on the properties of separated 241Am and its associated oxide. However, when 241Am exists as an ingrown progeny (and as a small mass fraction) within plutonium mixtures, it is appropriate to assign transportability factors of the predominant plutonium mixtures (typically slow) to themore » Am241. It is argued that the transportability factor for 241Am in sludge likewise should be slow because it exists as a small mass fraction as the ingrown progeny within the uranium oxide in sludge. In this report, the transportability class assignment for 241Am is underpinned with radiochemical characterization data on K Basin sludge and with studies conducted with other irradiated fuel exposed to elevated temperatures and conditions similar to the STP. Key findings and conclusions from evaluation of the characterization data and published literature are summarized here. Plutonium and 241Am make up very small fractions of the uranium within the K Basin sludge matrix. Plutonium is present at about 1 atom per 500 atoms of uranium and 241Am at about 1 atom per 19000 of uranium. Plutonium and americium are found to remain with uranium in the solid phase in all of the {approx}60 samples taken and analyzed from various sources of K Basin sludge. The uranium-specific concentrations of plutonium and americium also remain approximately constant over a uranium concentration range (in the dry sludge solids) from 0.2 to 94 wt%, a factor of {approx}460. This invariability demonstrates that 241Am does not partition from the uranium or plutonium fraction for any characterized sludge matrix. Most of the K Basin sludge characterization data is derived spent nuclear fuel corroded within the K Basins at 10-15?C. The STP process will place water-laden sludges from the K Basin in process vessels at {approx}150-180 C. Therefore, published studies with other irradiated (uranium oxide) fuel were examined. From these studies, the affinity of plutonium and americium for uranium in irradiated UO2 also was demonstrated at hydrothermal conditions (150 C anoxic liquid water) approaching those proposed for the STP process and even for hydrothermal conditions outside of the STP operating envelope (e.g., 150 C oxic and 100 C oxic and anoxic liquid water). In summary, by demonstrating that the chemical and physical behavior of 241Am in the sludge matrix is similar to that of the predominant species (uranium and for the plutonium from which it originates), a technical basis is provided for using the slow uptake transportability factor for 241Am that is currently used for plutonium and uranium oxides. The change from moderate to slow uptake for 241Am could reduce the overall analyzed dose consequences for the STP by more than 30%.« less

SOLVENT EXTRACTION PROCESS FOR SEPARATING URANIUM AND PLUTONIUM FROM AQUEOUS ACIDIC SOLUTIONS OF NEUTRON IRRADIATED URANIUM

DOEpatents

Bruce, F.R.

1962-07-24

A solvent extraction process was developed for separating actinide elements including plutonium and uranium from fission products. By this method the ion content of the acidic aqueous solution is adjusted so that it contains more equivalents of total metal ions than equivalents of nitrate ions. Under these conditions the extractability of fission products is greatly decreased. (AEC)

PROCESS FOR EXTRACTING NEPTUNIUM AND PLUTONIUM FROM NITRIC ACID SOLUTIONS OF SAME CONTAINING URANYL NITRATE WITH A TERTIARY AMINE

DOEpatents

Sheppard, J.C.

1962-07-31

A process of selectively extracting plutonium nitrate and neptunium nitrate with an organic solution of a tertiary amine, away from uranyl nitrate present in an aqueous solution in a maximum concentration of 1M is described. The nitric acid concentration is adjusted to about 4M and nitrous acid is added prior to extraction. (AEC)

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2017-12-01

It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

METHOD FOR REMOVING CONTAMINATION FROM PRECIPITATES

DOEpatents

Stahl, G.W.

1959-01-01

An improvement in the bismuth phosphate carrier precipitation process is presented for the recovery and purification of plutonium. When plutonium, in the tetravalent state, is carried on a bismuth phosphate precipitate, amounts of centain of the fission products are carried along with the plutonium. The improvement consists in washing such fission product contaminated preeipitates with an aqueous solution of ammonium hydrogen fluoride. since this solution has been found to be uniquely effective in washing fission production contamination from the bismuth phosphate precipitate.

LAB-SCALE DEMONSTRATION OF PLUTONIUM PURIFICATION BY ANION EXCHANGE, PLUTONIUM (IV) OXALATE PRECIPITATION, AND CALCINATION TO PLUTONIUM OXIDE TO SUPPORT THE MOX FEED MISSION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crowder, M.; Pierce, R.

2012-08-22

H-Canyon and HB-Line are tasked with the production of PuO{sub 2} from a feed of plutonium metal. The PuO{sub 2} will provide feed material for the MOX Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, the solution will be transferred to HB-Line for purification by anion exchange. Subsequent unit operations include Pu(IV) oxalate precipitation, filtration and calcination to form PuO{sub 2}. This report details the results from SRNL anion exchange, precipitation, filtration, calcination, and characterization tests, as requested by HB-Line1 and described in the task plan. This study involved an 80-g batch of Pu and employed testmore » conditions prototypical of HB-Line conditions, wherever feasible. In addition, this study integrated lessons learned from earlier anion exchange and precipitation and calcination studies. H-Area Engineering selected direct strike Pu(IV) oxalate precipitation to produce a more dense PuO{sub 2} product than expected from Pu(III) oxalate precipitation. One benefit of the Pu(IV) approach is that it eliminates the need for reduction by ascorbic acid. The proposed HB-Line precipitation process involves a digestion time of 5 minutes after the time (44 min) required for oxalic acid addition. These were the conditions during HB-line production of neptunium oxide (NpO{sub 2}). In addition, a series of small Pu(IV) oxalate precipitation tests with different digestion times were conducted to better understand the effect of digestion time on particle size, filtration efficiency and other factors. To test the recommended process conditions, researchers performed two nearly-identical larger-scale precipitation and calcination tests. The calcined batches of PuO{sub 2} were characterized for density, specific surface area (SSA), particle size, moisture content, and impurities. Because the 3013 Standard requires that the calcination (or stabilization) process eliminate organics, characterization of PuO{sub 2} batches monitored the presence of oxalate by thermogravimetric analysis-mass spectrometry (TGA-MS). To use the TGA-MS for carbon or oxalate content, some method development will be required. However, the TGA-MS is already used for moisture measurements. Therefore, SRNL initiated method development for the TGA-MS to allow quantification of oxalate or total carbon. That work continues at this time and is not yet ready for use in this study. However, the collected test data can be reviewed later as those analysis tools are available.« less

Application of laser induced breakdown spectroscopy (LIBS) instrumentation for international safeguards

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barefield Ii, James E; Clegg, Samuel M; Lopez, Leon N

2010-01-01

Advanced methodologies and improvements to current measurements techniques are needed to strengthen the effectiveness and efficiency of international safeguards. This need was recognized and discussed at a Technical Meeting on 'The Application of Laser Spectrometry Techniques in IAEA Safeguards' held at IAEA headquarters (September 2006). One of the principal recommendations from that meeting was the need to pursue the development of novel complementary access instrumentation based on Laser Induced Breakdown Spectroscopy (UBS) for the detection of gaseous and solid signatures and indicators of nuclear fuel cycle processes and associated materials'. Pursuant to this recommendation the Department of Safeguards (SG) undermore » the Division of Technical Support (SGTS) convened the 'Experts and Users Advisory Meeting on Laser Induced Breakdown Spectroscopy (LIBS) for Safeguards Applications' also held at IAEA headquarters (July 2008). This meeting was attended by 12 LlBS experts from the Czech Republic, the European Commission, France, the Republic of South Korea, the United States of America, Germany, the United Kingdom of Great Britain, Canada, and Northern Ireland. Following a presentation of the needs of the IAEA inspectors, the LIBS experts agreed that needs as presented could be partially or fully fulfilled using LIBS instrumentation. Inspectors needs were grouped into the following broad categories: (1) Improvements to in-field measurements/environmental sampling; (2) Monitoring status of activities in Hot Cells; (3) Verify status of activity at a declared facility via process monitoring; and (4) Need for pre-screening of environmental samples before analysis. The primary tool employed by the IAEA to detect undeclared processes and activities at special nuclear material facilities and sites is environmental sampling. One of the objectives of the Next Generation Safeguards Initiative (NGSI) Program Plan calls for the development of advanced tools and methodologies to detect and analyze undeclared processing or production of special nuclear material. Los Alamos National Laboratory is currently investigating potential uses of LIBS for safeguards applications, including (1) a user-friendly man-portable LIBS system to characterize samples in real to near-real time (typical analysis time are on the order of minutes) across a wide range of elements in the periodic table from hydrogen up to heavy elements like plutonium and uranium, (2) a LIBS system that can be deployed in harsh environments such as hot cells and glove boxes providing relative compositional analysis of process streams for example ratios like Cm/Up and Cm/U, (3) an inspector field deployable system that can be used to analyze the elemental composition of microscopic quantities of samples containing plutonium and uranium, and (4) a high resolution LIBS system that can be used to determine the isotopic composition of samples containing for example uranium, plutonium... etc. In this paper, we will describe our current development and performance testing results for LIBS instrumentation both in a fixed lab and measurements in field deployable configurations.« less

The role of troublesome components in plutonium vitrification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Hong; Vienna, J.D.; Peeler, D.K.

1996-05-01

One option for immobilizing surplus plutonium is vitrification in a borosilicate glass. Two advantages of the glass form are (1) high tolerance to feed variability and, (2) high solubility of some impurity components. The types of plutonium-containing materials in the United States inventory include: pits, metals, oxides, residues, scrap, compounds, and fuel. Many of them also contain high concentrations of carbon, chloride, fluoride, phosphate, sulfate, and chromium oxide. To vitrify plutonium-containing scrap and residues, it is critical to understand the impact of each component on glass processing and chemical durability of the final product. This paper addresses glass processing issuesmore » associated with these troublesome components. It covers solubility limits of chlorine, fluorine, phosphate, sulfate, and chromium oxide in several borosilicate based glasses, and the effect of each component on vitrification (volatility, phase segregation, crystallization, and melt viscosity). Techniques (formulation, pretreatment, removal, and/or dilution) to mitigate the effect of these troublesome components are suggested.« less

THE INTEGRATION OF A PROPOSED ZONE CLOSURE APPROACH FOR THE PLUTONIUM FINISHING PLANT (PFP) DECOMMISSIONING & THE PFP ZONE HANFORD SITE WASHINGTON

DOE Office of Scientific and Technical Information (OSTI.GOV)

HOPKINS, A.M.

2005-02-23

The Plutonium Finishing Plant (PFP) and associated processing facilities are located in the 200 area of the Hanford Site in Eastern Washington. This area is part of what is now called the Central Plateau. In order to achieve closure of the contaminated facilities and waste sites at Hanford on the Central Plateau (CP), a geographic re-districting of the area into zones has been proposed in the recently published Plan for Central Plateau Closure. One of the 22 zones proposed in the Central Plateau encompasses the PFP and ancillary facilities. Approximately eighty six buildings are included in the PFP Zone. Thismore » paper addresses the approach for the closure of the PFP Zone within the Central Plateau. The PFP complex of buildings forms the bulk of the structures in the PFP Zone. For closure of the above-grade portion of structures within the PFP complex, the approach is to remove them to a state called ''slab-on-grade'' per the criteria contained in PFP End Point Criteria document and as documented in action memoranda. For below-grade portions of the structures (such as below-grade rooms, pipe trenches and underground ducts), the approach is to remove as much residual contamination as practicable and to fill the void spaces with clean fill material such as sand, grout, or controlled density fill. This approach will be modified as planning for the waste sites progresses to ensure that the actions of the PFP decommissioning projects do not negatively impact future planned actions under the CERCLA. Cribs, settling tanks, septic tanks and other miscellaneous below-grade void spaces will either be cleaned to the extent practicable and filled or will be covered with an environmental barrier as determined by further studies and CERCLA decision documents. Currently, between two and five environmental barriers are proposed to be placed over waste sites and remaining building slabs in the PFP Zone.« less

1. VIEW OF A PORTION OF THE HYDRIDE PROCESSING LABORATORY. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

1. VIEW OF A PORTION OF THE HYDRIDE PROCESSING LABORATORY. OPERATIONS IN THE GLOVE BOX IN THE BACKGROUND OF THE PHOTOGRAPH INCLUDED HYDRIDING OF PLUTONIUM AND HYDRIDE SEPARATION. IN THE FOREGROUND, THE VACUUM MONITOR CONTROL PANEL MEASURED TEMPERATURES WITHIN THE GLOVEBOX. THE CENTER CONTROL PANEL REGULATED THE FURNACE INSIDE THE GLOVE BOX USED IN THE HYDRIDING PROCESSES. THIS EQUIPMENT WAS ESSENTIAL TO THE HYDRIDING PROCESS, AS WELL AS OTHER GLOVE BOX OPERATIONS. - Rocky Flats Plant, Plutonium Laboratory, North-central section of industrial area at 79 Drive, Golden, Jefferson County, CO

Application of Compton-suppressed self-induced XRF to spent nuclear fuel measurement

NASA Astrophysics Data System (ADS)

Park, Se-Hwan; Jo, Kwang Ho; Lee, Seung Kyu; Seo, Hee; Lee, Chaehun; Won, Byung-Hee; Ahn, Seong-Kyu; Ku, Jeong-Hoe

2017-11-01

Self-induced X-ray fluorescence (XRF) is a technique by which plutonium (Pu) content in spent nuclear fuel can be directly quantified. In the present work, this method successfully measured the plutonium/uranium (Pu/U) peak ratio of a pressurized water reactor (PWR)'s spent nuclear fuel at the Korea atomic energy research institute (KAERI)'s post irradiation examination facility (PIEF). In order to reduce the Compton background in the low-energy X-ray region, the Compton suppression system additionally was implemented. By use of this system, the spectrum's background level was reduced by a factor of approximately 2. This work shows that Compton-suppressed selfinduced XRF can be effectively applied to Pu accounting in spent nuclear fuel.

JPRS Report, Proliferation Issues

DTIC Science & Technology

1993-06-07

Ruta Skatikaite; RESPUBLIKA, 19 May 93] ................................................................................ 20 Radioactive Beryllium...nuclear fuel will be transported around 2000 to a reprocessing facility in [Yi] IAEA surveillance cameras are said to take four photos Tokai, Ibaraki...Comparing these two methods, the method of extracting May 93 pp 342-346. plutonium is similar to carrying a backpack to transport goods, while

12. VIEW OF THE INSPECTION MODULE (MODULE D). THE GLOVE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

12. VIEW OF THE INSPECTION MODULE (MODULE D). THE GLOVE BOX IN THE FOREFRONT OF THE PHOTOGRAPH CONTAINS A DRILL PRESS; OTHER GLOVE BOXES ARE USED FOR PARTS INSPECTION. (5/70) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

14. VIEW OF THE OUTSIDE OF A GLOVE BOX THAT ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

14. VIEW OF THE OUTSIDE OF A GLOVE BOX THAT CONTAINS ELECTROREFINING EQUIPMENT. ELECTROREFINING WAS ONE OF THE PROCESSES USED TO PURIFY PLUTONIUM THAT DID NOT MEET PURITY SPECIFICATIONS. (10/25/66) - Rocky Flats Plant, Plutonium Fabrication, Central section of Plant, Golden, Jefferson County, CO

PLUTONIUM FINISHING PLANT (PFP) 241-Z LIQUID WASTE TREATMENT FACILITY DEACTIVATION AND DEMOLITION

DOE Office of Scientific and Technical Information (OSTI.GOV)

JOHNSTON GA

2008-01-15

Fluor Hanford, Inc. (FH) is proud to submit the Plutonium Finishing Plant (PFP) 241-Z liquid Waste Treatment Facility Deactivation and Demolition (D&D) Project for consideration by the Project Management Institute as Project of the Year for 2008. The decommissioning of the 241-Z Facility presented numerous challenges, many of which were unique with in the Department of Energy (DOE) Complex. The majority of the project budget and schedule was allocated for cleaning out five below-grade tank vaults. These highly contaminated, confined spaces also presented significant industrial safety hazards that presented some of the most hazardous work environments on the Hanford Site.more » The 241-Z D&D Project encompassed diverse tasks: cleaning out and stabilizing five below-grade tank vaults (also called cells), manually size-reducing and removing over three tons of process piping from the vaults, permanently isolating service utilities, removing a large contaminated chemical supply tank, stabilizing and removing plutonium-contaminated ventilation ducts, demolishing three structures to grade, and installing an environmental barrier on the demolition site . All of this work was performed safely, on schedule, and under budget. During the deactivation phase of the project between November 2005 and February 2007, workers entered the highly contaminated confined-space tank vaults 428 times. Each entry (or 'dive') involved an average of three workers, thus equaling approximately 1,300 individual confined -space entries. Over the course of the entire deactivation and demolition period, there were no recordable injuries and only one minor reportable skin contamination. The 241-Z D&D Project was decommissioned under the provisions of the 'Hanford Federal Facility Agreement and Consent Order' (the Tri-Party Agreement or TPA), the 'Resource Conservation and Recovery Act of 1976' (RCRA), and the 'Comprehensive Environmental Response, Compensation, and Liability Act of 1980' (CERCLA). The project completed TPA Milestone M-083-032 to 'Complete those activities required by the 241-Z Treatment and Storage Unit's RCRA Closure Plan' four years and seven months ahead of this legally enforceable milestone. In addition, the project completed TPA Milestone M-083-042 to 'Complete transition and dismantlement of the 241-2 Waste Treatment Facility' four years and four months ahead of schedule. The project used an innovative approach in developing the project-specific RCRA closure plan to assure clear integration between the 241-Z RCRA closure activities and ongoing and future CERCLA actions at PFP. This approach provided a regulatory mechanism within the RCRA closure plan to place segments of the closure that were not practical to address at this time into future actions under CERCLA. Lessons learned from th is approach can be applied to other closure projects within the DOE Complex to control scope creep and mitigate risk. A paper on this topic, entitled 'Integration of the 241-Z Building D and D Under CERCLA with RCRA Closure at the PFP', was presented at the 2007 Waste Management Conference in Tucson, Arizona. In addition, techniques developed by the 241-Z D&D Project to control airborne contamination, clean the interior of the waste tanks, don and doff protective equipment, size-reduce plutonium-contaminated process piping, and mitigate thermal stress for the workers can be applied to other cleanup activities. The project-management team developed a strategy utilizing early characterization, targeted cleanup, and close coordination with PFP Criticality Engineering to significantly streamline the waste- handling costs associated with the project . The project schedule was structured to support an early transition to a criticality 'incredible' status for the 241-Z Facility. The cleanup work was sequenced and coordinated with project-specific criticality analysis to allow the fissile material waste being generated to be managed in a bulk fashion, instead of individual waste packages. This approach negated the need for real-time assay of individual waste packages, greatly improving the efficiency of the cleanup operation. The cleanup and stabilization of the 241-2 Liquid Effluent Treatment Facility reduced radiological risks to the environment and Hanford site workers. It was recognized as a success by regulatory agencies, the media, the DOE-client, and stakeholders. The 241-Z D&D Project demonstrated management excellence in adapting to significant changes in project direction, fostered a safety culture that amassed impressive results on this high-hazard job, maintained excellent communications with the client and stakeholders, and developed and implemented unique cleanup techniques.« less

Fluorination process using catalyst

DOEpatents

Hochel, Robert C.; Saturday, Kathy A.

1985-01-01

A process for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3, AgF.sub.2 and NiF.sub.2, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF.sub.3 and AgF.sub.2, whereby the fluorination is significantly enhanced.

Fluorination process using catalysts

DOEpatents

Hochel, R.C.; Saturday, K.A.

1983-08-25

A process is given for converting an actinide compound selected from the group consisting of uranium oxides, plutonium oxides, uranium tetrafluorides, plutonium tetrafluorides and mixtures of said oxides and tetrafluorides, to the corresponding volatile actinide hexafluoride by fluorination with a stoichiometric excess of fluorine gas. The improvement involves conducting the fluorination of the plutonium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/, AgF/sub 2/ and NiF/sub 2/, whereby the fluorination is significantly enhanced. The improvement also involves conducting the fluorination of one of the uranium compounds in the presence of a fluoride catalyst selected from the group consisting of CoF/sub 3/ and AgF/sub 2/, whereby the fluorination is significantly enhanced.

Electrolysis of plutonium nitride in LiCl-KCl eutectic melts

NASA Astrophysics Data System (ADS)

Shirai, O.; Iwai, T.; Shiozawa, K.; Suzuki, Y.; Sakamura, Y.; Inoue, T.

2000-01-01

The electrolysis of plutonium nitride, PuN, was investigated in the LiCl-KCl eutectic salt with 0.54 wt% PuCl 3 at 773 K in order to understand the dissolution of PuN at the anode and the deposition of metal at the cathode from the viewpoint of the application of a pyrochemical process to nitride fuel cycle. It was found from cyclic voltammetry that the electrochemical dissolution of PuN began nearly at the theoretically evaluated potential and this reaction was irreversible. Several grams of plutonium metal were successfully recovered at the molybdenum electrode as a deposit with a current efficiency of about 90%, although some fractions of the deposited plutonium often fell from the molybdenum electrode.

Radioactive waste management and plutonium recovery within the context of the development of nuclear energy in Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kushnikov, V.

1996-05-01

The Russian strategy for radioactive waste and plutonium management is based on the concept of the closed fuel cycle that has been adopted in Russia, and, to a great degree, falls under the jurisdiction of the existing Russian nuclear energy structures. From its very beginning, Russian atomic energy policy was based on finding the most effective method of developing the new fuel direction with the maximum possible utilization of the energy potential from the fission of heavy atoms and the achievement of fuel self-sufficiency through the recycling of secondary fuel. Although there can be no doubt about the importance ofmore » economic considerations (for the future), concerns for the safety of the environment are currently of the utmost importance. In this context, spent NPP fuel can be viewed as a waste to be buried only if there is persuasive evidence that such an approach is both economically and environmentally sound. The production of I GW of energy per year is accompanied by the accumulation of up to 800-1000 kg of highly radioactive fission products and approximately 250 kg of plutonium. Currently, spent fuel from the VVER 100 and the RBNK reactors contains approximately 25 tons of plutonium. There is an additional 30 tons of fuel-grade plutonium in the form of purified oxide, separated from spent fuels used in VVER440 reactors and other power production facilities, as well as approximately 100 tons of weapons-grade plutonium from dismantled warheads. The spent fuel accumulates significant amounts of small actinoids - neptunium americium, and curium. Science and technology have not yet found technical solutions for safe and secure burial of non-reprocessed spent fuel with such a broad range of products, which are typically highly radioactive and will continue to pose a threat for hundreds of thousands of years.« less

A CHEMICAL METHOD OF TREATING FISSIONABLE MATERIAL

DOEpatents

Olson, C.M.

1959-09-01

One step of a process for separating plutonium from uranium and fission products is presented. A nitric acid solution containing these constituents is treated with formic acid to reduce simultaneously the plutonium to a valence state of not greater than +4 and destroy and eliminate the excess nitric acid.

SULFIDE METHOD PLUTONIUM SEPARATION

DOEpatents

Duffield, R.B.

1958-08-12

A process is described for the recovery of plutonium from neutron irradiated uranium solutions. Such a solution is first treated with a soluble sullide, causing precipitation of the plutoniunn and uraniunn values present, along with those impurities which form insoluble sulfides. The precipitate is then treated with a solution of carbonate ions, which will dissolve the uranium and plutonium present while the fission product sulfides remain unaffected. After separation from the residue, this solution may then be treated by any of the usual methods, such as formation of a lanthanum fluoride precipitate, to effect separation of plutoniunn from uranium.

The instrumental method of plutonium determination

NASA Astrophysics Data System (ADS)

Knyazev, B. B.; Kazachevskiy, I. V.; Solodukhin, V. P.; Lukashenko, S. N.; Knatova, M. K.; Kashirskiy, V. V.

2003-01-01

A method of direct instrumental determination of plutonium isotopes in soil samples is described. For the method a special program of spectra processing and activity calculation had to be prepared. The detection limit of 239+240Pu in absence of interfering radiation is about 200 Bq/kg (by 3.3σ criteria). Examples are given of the method application for the study of radionuclide soil composition in separate objects of Semipalatinsk Nuclear Test Site (SNTS). It is shown that for different objects under study the correlation degree between plutonium and americium activities may change rather substantially.

Dehydration of plutonium or neptunium trichloride hydrate

DOEpatents

Foropoulos, Jr., Jerry; Avens, Larry R.; Trujillo, Eddie A.

1992-01-01

A process of preparing anhydrous actinide metal trichlorides of plutonium or neptunium by reacting an aqueous solution of an actinide metal trichloride selected from the group consisting of plutonium trichloride or neptunium trichloride with a reducing agent capable of converting the actinide metal from an oxidation state of +4 to +3 in a resultant solution, evaporating essentially all the solvent from the resultant solution to yield an actinide trichloride hydrate material, dehydrating the actinide trichloride hydrate material by heating the material in admixture with excess thionyl chloride, and recovering anhydrous actinide trichloride is provided.

Dehydration of plutonium or neptunium trichloride hydrate

DOEpatents

Foropoulos, J. Jr.; Avens, L.R.; Trujillo, E.A.

1992-03-24

A process is described for preparing anhydrous actinide metal trichlorides of plutonium or neptunium by reacting an aqueous solution of an actinide metal trichloride selected from the group consisting of plutonium trichloride or neptunium trichloride with a reducing agent capable of converting the actinide metal from an oxidation state of +4 to +3 in a resultant solution, evaporating essentially all the solvent from the resultant solution to yield an actinide trichloride hydrate material, dehydrating the actinide trichloride hydrate material by heating the material in admixture with excess thionyl chloride, and recovering anhydrous actinide trichloride.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searchedmore » and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.« less

Processing plutonium-contaminated soil on Johnston Atoll

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moroney, K.; Moroney, J. III; Turney, J.

1994-07-01

This article describes a cleanup project to process plutonium- and americium-contaminated soil on Johnston Atoll for volume reduction. Thermo Analytical`s (TMA`s) segmented gate system (SGS) for this remedial operation has been in successful on-site operation since 1992. Topics covered include the basis for development, a description of the Johnston Atoll; the significance of results; the benefits of the technology; applicability to other radiologically contaminated sites. 7 figs., 1 tab.

Cygnus Performance in Subcritical Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. Corrow, M. Hansen, D. Henderson, S. Lutz, C. Mitton, et al.

2008-02-01

The Cygnus Dual Beam Radiographic Facility consists of two identical radiographic sources with the following specifications: 4-rad dose at 1 m, 1-mm spot size, 50-ns pulse length, 2.25-MeV endpoint energy. The facility is located in an underground tunnel complex at the Nevada Test Site. Here SubCritical Experiments (SCEs) are performed to study the dynamic properties of plutonium. The Cygnus sources were developed as a primary diagnostic for these tests. Since SCEs are single-shot, high-value events - reliability and reproducibility are key issues. Enhanced reliability involves minimization of failure modes through design, inspection, and testing. Many unique hardware and operational featuresmore » were incorporated into Cygnus to insure reliability. Enhanced reproducibility involves normalization of shot-to-shot output also through design, inspection, and testing. The first SCE to utilize Cygnus, Armando, was executed on May 25, 2004. A year later, April - May 2005, calibrations using a plutonium step wedge were performed. The results from this series were used for more precise interpretation of the Armando data. In the period February - May 2007 Cygnus was fielded on Thermos, which is a series of small-sample plutonium shots using a one-dimensional geometry. Pulsed power research generally dictates frequent change in hardware configuration. Conversely, SCE applications have typically required constant machine settings. Therefore, while operating during the past four years we have accumulated a large database for evaluation of machine performance under highly consistent operating conditions. Through analysis of this database Cygnus reliability and reproducibility on Armando, Step Wedge, and Thermos is presented.« less

Joint US/Russian Studies of Population Exposures Resulting from Nuclear Production Activities in the Southern Urals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Napier, Bruce A.

2014-01-01

Beginning in 1948, the Soviet Union initiated a program for production of nuclear materials for a weapons program. The first facility for production of plutonium was constructed in the central portion of the country east of the southern Ural Mountains, about halfway between the major industrial cities of Ekaterinburg and Chelyabinsk. The facility now known as the Mayak Production Association and its associated town, now known as Ozersk, were built to irradiate uranium in reactors, separate the resulting plutonium in reprocessing plants, and prepare plutonium metal. The rush to production, coupled with inexperience in handling radioactive materials, lead to largemore » radiation exposures, not only to the workers in the facilities, but also to the surrounding public. Fuel processing started with no controls on releases, and fuel dissolution and accidents in reactors resulted in release of about 37 PBq (1015 Bq) of 131I between 1948 and 1967. Designed disposals of low- and intermediate-level liquid radioactive wastes, and accidental releases via cooling water from tank farms of high-level liquid radioactive wastes, into the small Techa River caused significant contamination and exposures to residents of numerous small riverside villages downstream of the site. Discovery of the magnitude of the aquatic contamination in late 1951 caused revisions to the waste handling regimes, but not before over 200 PBq of radionuclides (with large contributions of 90Sr and 137Cs) were released. Liquid wastes were diverted to tiny Lake Karachay (which today holds over 4 EBq); cooling water was stopped in the tank farms. In 1957, one of the tanks in the tank farm overheated and exploded; over 70 PBq, disproportionately 90Sr, was blown over a large area to the northeast of the site; a large area was contaminated and many villages evacuated. This area today is known as the East Urals Radioactive Trace (EURT). Each of these releases was significant; together they have created a group of cohorts unrivaled in the world for their chronic, low-dose-rate radiation exposure. The 26,000 workers at Mayak were highly exposed to external gamma and inhaled plutonium. A cohort of individuals raised as children in Ozersk is under evaluation for their exposures to radioiodine. The Techa River Cohort consists of over 30,000 people who were born before the start of exposure in 1949 and lived along the Techa River. The Techa River Offspring Cohort consists of about 21,000 persons born to one or more exposed parents of this group - many of whom also lived along the contaminated river. The EURT Cohort consists of about 18,000 people who were evacuated from the EURT soon after the 1957 explosion and another 8000 who remained. These groups together are the focus of dose reconstruction and epidemiological studies funded by the US, Russia, and the European Union to address the question “Are doses delivered at low dose rates as effective in producing health effects as the same doses delivered at high dose rates?”« less

Accumulation, organ distribution, and excretion kinetics of ²⁴¹Am in Mayak Production Association workers.

PubMed

Suslova, Klara G; Sokolova, Alexandra B; Efimov, Alexander V; Miller, Scott C

2013-03-01

Americium-241 (²⁴¹Am) is the second most significant radiation hazard after ²³⁹Pu at some of the Mayak Production Association facilities. This study summarizes current data on the accumulation, distribution, and excretion of americium compared with plutonium in different organs from former Mayak PA workers. Americium and plutonium were measured in autopsy and bioassay samples and correlated with the presence or absence of chronic disease and with biological transportability of the aerosols encountered at different workplaces. The relative accumulation of ²⁴¹Am was found to be increasing in the workers over time. This is likely from ²⁴¹Pu that increases with time in reprocessed fuel and from the increased concentrations of ²⁴¹Am and ²⁴¹Pu in inhaled alpha-active aerosols. While differences were observed in lung retention with exposures to different industrial compounds with different transportabilities (i.e., dioxide and nitrates), there were no significant differences in lung retention between americium and plutonium within each transportability group. In the non-pulmonary organs, the highest ratios of ²⁴¹Am/²⁴¹Am + SPu were observed in the skeleton. The relative ratios of americium in the skeleton versus liver were significantly greater than for plutonium. The relative amounts of americium and plutonium found in the skeleton compared with the liver were even greater in workers with documented chronic liver diseases. Excretion rates of ²⁴¹Am in ‘‘healthy’’ workers were estimated using bioassay and autopsy data. The data suggest that impaired liver function leads to reduced hepatic ²⁴¹Am retention, leading to increased ²⁴¹Am excretion.

Radiocesium in the Savannah River Site environment.

PubMed

Carlton, W H; Murphy, C E; Evans, A G

1994-09-01

The Savannah River Site has produced plutonium, tritium, and other special nuclear materials for national defense, other government programs, and some civilian purposes. Radiocesium, a waste product, has been released to the environment during the operation of five reactors, two radio-chemical processing facilities, and a high-level waste storage system. During the period 1955-1989, 130 GBq of 137Cs was released to the atmosphere and 2.2 x 10(4) GBq was released to site streams and ponds. Approximately 65% of the latter remained on the site. The maximum individual effective dose equivalent at the site boundary was estimated to be 3.3 microSv from atmospheric releases and 600 microSv from liquid releases. The 80-km population dose was 1.6 person-Sv.

3. VIEW OF CHAINVEYOR. AN ENCLOSED CHAIN CONVEYOR CONNECTED GLOVE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

3. VIEW OF CHAINVEYOR. AN ENCLOSED CHAIN CONVEYOR CONNECTED GLOVE BOXES WITHIN AND BETWEEN MODULAR WORK AREAS. LEADED GLOVES WERE AFFIXED TO PORTS ALONG THE CHAINVEYOR PATHWAY TO ALLOW OPERATOR ACCESS. (1/25/93) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, Chris; Willis, William; Carter, Robert

2013-07-01

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for usemore » as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)« less

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

METHODS OF PREPARATION OF ELEMENT 95

DOEpatents

Seaborg, G.T.; James, R.A.

1962-07-17

A process of making americium by bombarding plutonium or uranium with neutrons or deuterons and aging the mass for decay of the plutonium formed to americium is described. The americium may then be separated by dissolving the mass in aqueous acid and carrier precipitation of the americium, especially on lanthanum or cerous fluoride. (AEC)

A Graphical Examination of Uranium and Plutonium Fissility

ERIC Educational Resources Information Center

Reed, B. Cameron

2008-01-01

The issue of why only particular isotopes of uranium and plutonium are suitable for use in nuclear weapons is analyzed with the aid of graphs and semiquantitative discussions of parameters such as excitation energies, fission barriers, reaction cross-sections, and the role of processes such as [alpha]-decay and spontaneous fission. The goal is to…

Method for aqueous radioactive waste treatment

DOEpatents

Bray, L.A.; Burger, L.L.

1994-03-29

Plutonium, strontium, and cesium found in aqueous waste solutions resulting from nuclear fuel processing are removed by contacting the waste solutions with synthetic zeolite incorporating up to about 5 wt % titanium as sodium titanate in an ion exchange system. More than 99.9% of the plutonium, strontium, and cesium are removed from the waste solutions. 3 figures.

Method for aqueous radioactive waste treatment

DOEpatents

Bray, Lane A.; Burger, Leland L.

1994-01-01

Plutonium, strontium, and cesium found in aqueous waste solutions resulting from nuclear fuel processing are removed by contacting the waste solutions with synthetic zeolite incorporating up to about 5 wt % titanium as sodium titanate in an ion exchange system. More than 99.9% of the plutonium, strontium, and cesium are removed from the waste solutions.

77 FR 44222 - Notice of Availability of the Draft Surplus Plutonium Disposition Supplemental Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-27

... announcing the dates, times and locations for public hearings to receive comments on the Draft SPD... additional options for pit disassembly and conversion, which entails processing plutonium metal components to... practicable. DOE will hold public hearings on the Draft SPD Supplemental EIS; the dates, times and locations...

SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

DOEpatents

Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

1962-11-13

A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

Investigation of a “Sharps” Incident

DOE PAGES

Cournoyer, Michael Edward; Trujillo, Stanley; Schreiber, Stephen Bruce

2016-08-03

Special nuclear material research, process development, technology demonstration, and manufacturing capabilities are provided at the Los Alamos National Laboratory Plutonium Facility. Engineered barriers provide the most effective protection from radioactive and hazardous materials. The Worker Safety Security Team augments these passive safety feature by investigating incidents to identify appropriate prevention and mitigation measures. “Learning Teams” facilitate employee feedback loop and integration toward process improvement. Here, this article reports an investigation of a “Sharps” incident and reviews a case study of a technician that cuts his left thumb while making a gasket. Causal analysis of the sharps incident uncovered contributing factorsmore » that created the environment in which the incident occurred. Finally, latent organizational conditions that created error-likely situations or weakened defenses were identified and controlled. Effective improvements that reduce the probability or consequence of similar sharps incidents were implemented.« less

Progress on 241Am Production for Use in Radioisotope Power Systems

NASA Astrophysics Data System (ADS)

Baker, S. R.; Bell, K. J.; Brown, J.; Carrigan, C.; Carrott, M. J.; Gregson, C.; Clough, M.; Maher, C. J.; Mason, C.; Rhodes, C. J.; Rice, T. G.; Sarsfield, M. J.; Stephenson, K.; Taylor, R. J.; Tinsley, T. P.; Woodhead, D. A.; Wiss, T.

2014-08-01

Electrical power sources used in outer planet missions are a key enabling technology for data acquisition and communications. Power sources generate electricity from the thermal energy from alpha decay of the radioisotope 238Pu via thermo-electric conversion. Production of 238Pu requires specialist facilities including a nuclear reactor and reprocessing plants that are expensive to build and operate, so naturally, a more economical alternative is attractive to the industry. Within Europe 241Am is a feasible alternative to 238Pu that can provide a heat source for radioisotope thermoelectric generators (RTGs) and radioisotope heating units (RHUs). As a daughter product of 241Pu decay, 241Am is present at 1000s kg levels within the UK civil plutonium stockpile.A chemical separation process is required to extract the 241Am in a pure form and this paper describes such a process, successfully developed to the proof of concept stage.

Evaluation and development plan of NRTA measurement methods for the Rokkasho Reprocessing Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, T.K.; Hakkila, E.A.; Flosterbuer, S.F.

Near-real-time accounting (NRTA) has been proposed as a safeguards method at the Rokkasho Reprocessing Plant (RRP), a large-scale commercial boiling water and pressurized water reactors spent-fuel reprocessing facility. NRTA for RRP requires material balance closures every month. To develop a more effective and practical NRTA system for RRP, we have evaluated NRTA measurement techniques and systems that might be implemented in both the main process and the co-denitration process areas at RRP to analyze the concentrations of plutonium in solutions and mixed oxide powder. Based on the comparative evaluation, including performance, reliability, design criteria, operation methods, maintenance requirements, and estimatedmore » costs for each possible measurement method, recommendations for development were formulated. This paper discusses the evaluations and reports on the recommendation of the NRTA development plan for potential implementation at RRP.« less

Design and fabrication of a glovebox for the Plasma Hearth Process radioactive bench-scale system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wahlquist, D.R.

This paper presents some of the design considerations and fabrication techniques for building a glovebox for the Plasma Hearth Process (PHP) radioactive bench-scale system. The PHP radioactive bench-scale system uses a plasma torch to process a variety of radioactive materials into a final vitrified waste form. The processed waste will contain plutonium and trace amounts of other radioactive materials. The glovebox used in this system is located directly below the plasma chamber and is called the Hearth Handling Enclosure (HHE). The HHE is designed to maintain a confinement boundary between the processed waste and the operator. Operations that take placemore » inside the HHE include raising and lowering the hearth using a hydraulic lift table, transporting the hearth within the HHE using an overhead monorail and hoist system, sampling and disassembly of the processed waste and hearth, weighing the hearth, rebuilding a hearth, and sampling HEPA filters. The PHP radioactive bench-scale system is located at the TREAT facility at Argonne National Laboratory-West in Idaho Falls, Idaho.« less

Separation by solvent extraction

DOEpatents

Holt, Jr., Charles H.

1976-04-06

17. A process for separating fission product values from uranium and plutonium values contained in an aqueous solution, comprising adding an oxidizing agent to said solution to secure uranium and plutonium in their hexavalent state; contacting said aqueous solution with a substantially water-immiscible organic solvent while agitating and maintaining the temperature at from -1.degree. to -2.degree. C. until the major part of the water present is frozen; continuously separating a solid ice phase as it is formed; separating a remaining aqueous liquid phase containing fission product values and a solvent phase containing plutonium and uranium values from each other; melting at least the last obtained part of said ice phase and adding it to said separated liquid phase; and treating the resulting liquid with a new supply of solvent whereby it is practically depleted of uranium and plutonium.

METHOD OF SEPARATING NEPTUNIUM

DOEpatents

Seaborg, G.T.

1961-10-24

plutonium in an aqueous solution containing sulfate ions. The process consists of contacting the solution with an alkali metal bromate, digesting the resulting mixture at 15 to 25 deg C for a period of time not more than that required to oxidize the neptunium, adding lanthanum ions and fluoride ions, and separating the plutonium-containing precipitate thus formed from the supernatant solution. (AEC)

10 CFR 140.107 - Appendix G-Form of indemnity agreement with licensees processing plutonium for use in plutonium...

Code of Federal Regulations, 2012 CFR

2012-01-01

... policies as proof of financial protection. 140.107 Section 140.107 Energy NUCLEAR REGULATORY COMMISSION... specified in Item 2 of the Attachment and in the form of the nuclear energy liability insurance policy... liability of any Nuclear Energy Liability-Property Insurance Association policy designated in Item 5 of the...

10 CFR 140.107 - Appendix G-Form of indemnity agreement with licensees processing plutonium for use in plutonium...

Code of Federal Regulations, 2013 CFR

2013-01-01

... policies as proof of financial protection. 140.107 Section 140.107 Energy NUCLEAR REGULATORY COMMISSION... specified in Item 2 of the Attachment and in the form of the nuclear energy liability insurance policy... liability of any Nuclear Energy Liability-Property Insurance Association policy designated in Item 5 of the...

10 CFR 140.107 - Appendix G-Form of indemnity agreement with licensees processing plutonium for use in plutonium...

Code of Federal Regulations, 2014 CFR

2014-01-01

... policies as proof of financial protection. 140.107 Section 140.107 Energy NUCLEAR REGULATORY COMMISSION... specified in Item 2 of the Attachment and in the form of the nuclear energy liability insurance policy... liability of any Nuclear Energy Liability-Property Insurance Association policy designated in Item 5 of the...

10 CFR 140.107 - Appendix G-Form of indemnity agreement with licensees processing plutonium for use in plutonium...

Code of Federal Regulations, 2010 CFR

2010-01-01

... policies as proof of financial protection. 140.107 Section 140.107 Energy NUCLEAR REGULATORY COMMISSION... specified in Item 2 of the Attachment and in the form of the nuclear energy liability insurance policy... liability of any Nuclear Energy Liability-Property Insurance Association policy designated in Item 5 of the...

10 CFR 140.107 - Appendix G-Form of indemnity agreement with licensees processing plutonium for use in plutonium...

Code of Federal Regulations, 2011 CFR

2011-01-01

... policies as proof of financial protection. 140.107 Section 140.107 Energy NUCLEAR REGULATORY COMMISSION... specified in Item 2 of the Attachment and in the form of the nuclear energy liability insurance policy... liability of any Nuclear Energy Liability-Property Insurance Association policy designated in Item 5 of the...

On the equilibrium isotopic composition of the thorium-uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2016-12-01

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium-uranium-plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

On the equilibrium isotopic composition of the thorium–uranium–plutonium fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. Ye., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

2016-12-15

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium–uranium–plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

Tritium and plutonium in waters from the Bering and Chukchi Seas

USGS Publications Warehouse

Landa, E.R.; Beals, D.M.; Halverson, J.E.; Michel, R.L.; Cefus, G.R.

1999-01-01

During the summer of 1993, seawater in the Bering and Chukchi Seas was sampled on a joint Russian-American cruise [BERPAC] of the RV Okean to determine concentrations of tritium, 239Pu and 240Pu. Concentrations of tritium were determined by electrolytic enrichment and liquid scintilation counting. Tritium levels ranged up to 420 mBq L-1 showed no evidence of inputs other than those attribute atmospheric nuclear weapons testing. Plutonium was recovered from water samples by ferric hydroxide precipitation, and concentrations were determined by thermal ionization mass spectrometry. 239+240Pu concentrations ranged from <1 to 5.5 [mu]Bq L-1. These concentrations are lower than those measured in water samples from other parts of the ocean during the mid-1960's to the late 1980's. The 240Pu:239Pu ratios, although associated with large uncertainties, suggest that most of the plutonium is derived from world-wide fallout. As points of comparison, the highest concentrations of tritium and plutonium observed here were about five orders of magnitude lower than the maximum permissible concentrations allowed in water released to the off-site environs from licensed nuclear facilities in the United States. This study and others sponsored by the International Atomic Energy Agency and the Office of Naval Research's Arctic Nuclear Waste Assessment Program are providing data for the assessment of potential radiological impacts in the Arctic regions associated with nuclear waste disposal by the former Soviet Union.

Cleaning up the Legacy of the Cold War: Plutonium Oxides and the Role of Synchrotron Radiation Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, David Lewis

2015-01-21

The deceptively simple binary formula of AnO 2 belies an incredibly complex structural nature, and propensity to form mixed-valent, nonstoichiometric phases of composition AnO 2±x. For plutonium, the very formation of PuO 2+x has challenged a long-established dogma, and raised fundamental questions for long-term storage and environmental migration. This presentation covers two aspects of Los Alamos synchrotron radiation studies of plutonium oxides: (1) the structural chemistry of laboratory-prepared AnO 2+x systems (An = U, Pu; 0 ≤ x ≤ 0.25) determined through a combination of x-ray absorption fine structure spectroscopy (XAFS) and x-ray scattering of laboratory prepared samples; and (2)more » the application of synchrotron radiation towards the decontamination and decommissioning of the Rocky Flats Environmental Technology Site. Making the case for particle transport mechanisms as the basis of plutonium and americium mobility, rather than aqueous sorption-desorption processes, established a successful scientific basis for the dominance of physical transport processes by wind and water. The scientific basis was successful because it was in agreement with general theory on insolubility of PuO 2 in oxidation state IV, results of ultrafiltration analyses of field water/sediment samples, XAFS analyses of soil, sediment, and concrete samples, and was also in general agreement with on-site monitoring data. This understanding allowed Site contractors to rapidly move to application of soil erosion and sediment transport models as the means of predicting plutonium and americium transport, which led to design and application of site-wide soil erosion control technology to help control downstream concentrations of plutonium and americium in streamflow.« less

Routine inspection effort required for verification of a nuclear material production cutoff convention

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dougherty, D.; Fainberg, A.; Sanborn, J.

On 27 September 1993, President Clinton proposed {open_quotes}... a multilateral convention prohibiting the production of highly enriched uranium or plutonium for nuclear explosives purposes or outside of international safeguards.{close_quotes} The UN General Assembly subsequently adopted a resolution recommending negotiation of a non-discriminatory, multilateral, and internationally and effectively verifiable treaty (hereinafter referred to as {open_quotes}the Cutoff Convention{close_quotes}) banning the production of fissile material for nuclear weapons. The matter is now on the agenda of the Conference on Disarmament, although not yet under negotiation. This accord would, in effect, place all fissile material (defined as highly enriched uranium and plutonium) produced aftermore » entry into force (EIF) of the accord under international safeguards. {open_quotes}Production{close_quotes} would mean separation of the material in question from radioactive fission products, as in spent fuel reprocessing, or enrichment of uranium above the 20% level, which defines highly enriched uranium (HEU). Facilities where such production could occur would be safeguarded to verify that either such production is not occurring or that all material produced at these facilities is maintained under safeguards.« less

Gamma-ray imaging assay of cells 3-5 of the east cell line in the 235-F plutonium fuel form facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brand, A. D.; Aucott, T. J.; Diprete, D. P.

In August and September, 2016, scientists from the Savannah River National Laboratory (SRNL) took a series of gamma-ray imaging measurements through the cell windows in front of Cells 3-5 on the east line of the Plutonium Fuel Form (PuFF) Facility using an electrically cooled, high-purity germanium detector. A Germanium Gamma Ray Imager (GeGI) was utilized since it allowed for the location from which the radiation was being emitted to be identified by incoming gamma-ray energy. This measurement technique provided a tool which allowed for the relative concentration of Pu-238 to be mapped throughout each cell. The mapping and new assaymore » data were then used to update the model used in an assay discussed in a 2014 report (SRNL-STI-2014-00629) and to calculate a more accurate value for the holdup in each of the cells [1]. Note that the mapping and new assay data did not replace the previous assay data in the model. Rather, the mapping and new assay data provided additional details on source distribution, which supplemented the previous assay data.« less

ACTUAL WASTE TESTING OF GYCOLATE IMPACTS ON THE SRS TANK FARM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martino, C.

2014-05-28

Glycolic acid is being studied as a replacement for formic acid in the Defense Waste Processing Facility (DWPF) feed preparation process. After implementation, the recycle stream from DWPF back to the high-level waste Tank Farm will contain soluble sodium glycolate. Most of the potential impacts of glycolate in the Tank Farm were addressed via a literature review and simulant testing, but several outstanding issues remained. This report documents the actual-waste tests to determine the impacts of glycolate on storage and evaporation of Savannah River Site high-level waste. The objectives of this study are to address the following: Determine the extentmore » to which sludge constituents (Pu, U, Fe, etc.) dissolve (the solubility of sludge constituents) in the glycolate-containing 2H-evaporator feed. Determine the impact of glycolate on the sorption of fissile (Pu, U, etc.) components onto sodium aluminosilicate solids. The first objective was accomplished through actual-waste testing using Tank 43H and 38H supernatant and Tank 51H sludge at Tank Farm storage conditions. The second objective was accomplished by contacting actual 2H-evaporator scale with the products from the testing for the first objective. There is no anticipated impact of up to 10 g/L of glycolate in DWPF recycle to the Tank Farm on tank waste component solubilities as investigated in this test. Most components were not influenced by glycolate during solubility tests, including major components such as aluminum, sodium, and most salt anions. There was potentially a slight increase in soluble iron with added glycolate, but the soluble iron concentration remained so low (on the order of 10 mg/L) as to not impact the iron to fissile ratio in sludge. Uranium and plutonium appear to have been supersaturated in 2H-evaporator feed solution mixture used for this testing. As a result, there was a reduction of soluble uranium and plutonium as a function of time. The change in soluble uranium concentration was independent of added glycolate concentration. The change in soluble plutonium content was dependent on the added glycolate concentration, with higher levels of glycolate (5 g/L and 10 g/L) appearing to suppress the plutonium solubility. The inclusion of glycolate did not change the dissolution of or sorption onto actual-waste 2H-evaporator pot scale to an extent that will impact Tank Farm storage and concentration. The effects that were noted involved dissolution of components from evaporator scale and precipitation of components onto evaporator scale that were independent of the level of added glycolate.« less

The North Korean nuclear dilemma.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hecker, Siegfried S.

2004-01-01

The current nuclear crisis, the second one in ten years, erupted when North Korea expelled international nuclear inspectors in December 2002, then withdrew from the Nuclear Nonproliferation Treaty (NPT), and claimed to be building more nuclear weapons with the plutonium extracted from the spent fuel rods heretofore stored under international inspection. These actions were triggered by a disagreement over U.S. assertions that North Korea had violated the Agreed Framework (which froze the plutonium path to nuclear weapons to end the first crisis in 1994) by clandestinely developing uranium enrichment capabilities providing an alternative path to nuclear weapons. With Stanford Universitymore » Professor John Lewis and three other Americans, I was allowed to visit the Yongbyon Nuclear Center on Jan. 8, 2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. On the basis of our visit, we were not able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. However, based on the capabilities we saw, we must assume that North Korea has the capability to produce a crude nuclear device. On the matter of uranium enrichment programs, our host categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' The denials were not convincing at the time and since then have proven to be quite hollow by the revelations of A.Q. Khan's nuclear black market activities. There is no easy solution to the nuclear crisis in North Korea. A military strike to eliminate the nuclear facilities was never very attractive and now has been overcome by events. The principal threat is posed by a stockpile of nuclear weapons and weapons-grade plutonium. We have no way of finding where either may be hidden. A diplomatic solution remains the only path forward, but it has proven elusive. All sides have proclaimed a nuclear weapons-free Korean Peninsula as the end goal. The U.S. Government has chosen to negotiate with North Korea by means of the six-party talks. It has very clearly outlined its position of insisting on complete, verifiable, irreversible dismantlement of all North Korean nuclear programs. North Korea has offered several versions of 're-freezing' its plutonium program while still denying a uranium enrichment program. It has insisted on simultaneous and reciprocal steps to a final solution. Regardless of which diplomatic path is chosen, the scientific challenges of eliminating the North Korean nuclear weapons programs (and its associated infrastructure) in a safe, secure, and verifiable manner are immense. The North Korean program is considerably more complex and developed than the fledgling Iraqi program of 1991 and Libyan program of 2004. It is more along the lines, but more complex than that of South Africa in the early 1990s. Actions taken or not taken by the North Koreans at their nuclear facilities during the course of the ongoing diplomatic discussions are key to whether or not the nuclear program can be eliminated safely and securely, and they will greatly influence the price tag for such operations. Moreover, they will determine whether or not one can verify complete elimination. Hence, cooperation of the North Koreans now and during the dismantlement and elimination stages is crucial. Technical discussions among specialists, perhaps within the framework of the working groups of the six-party talks, could be very productive in setting the stage for an effective, verifiable elimination of North Korea's nuclear weapons program.« less

Plutonium interaction studies with the Mont Terri Opalinus Clay isolate Sporomusa sp. MT-2.99: changes in the plutonium speciation by solvent extractions.

PubMed

Moll, Henry; Cherkouk, Andrea; Bok, Frank; Bernhard, Gert

2017-05-01

Since plutonium could be released from nuclear waste disposal sites, the exploration of the complex interaction processes between plutonium and bacteria is necessary for an improved understanding of the fate of plutonium in the vicinity of such a nuclear waste disposal site. In this basic study, the interaction of plutonium with cells of the bacterium, Sporomusa sp. MT-2.99, isolated from Mont Terri Opalinus Clay, was investigated anaerobically (in 0.1 M NaClO 4 ) with or without adding Na-pyruvate as an electron donor. The cells displayed a strong pH-dependent affinity for Pu. In the absence of Na-pyruvate, a strong enrichment of stable Pu(V) in the supernatants was discovered, whereas Pu(IV) polymers dominated the Pu oxidation state distribution on the biomass at pH 6.1. A pH-dependent enrichment of the lower Pu oxidation states (e.g., Pu(III) at pH 6.1 which is considered to be more mobile than Pu(IV) formed at pH 4) was observed in the presence of up to 10 mM Na-pyruvate. In all cases, the presence of bacterial cells enhanced removal of Pu from solution and accelerated Pu interaction reactions, e.g., biosorption and bioreduction.

Safe disposal of surplus plutonium

NASA Astrophysics Data System (ADS)

Gong, W. L.; Naz, S.; Lutze, W.; Busch, R.; Prinja, A.; Stoll, W.

2001-06-01

About 150 tons of weapons grade and weapons usable plutonium (metal, oxide, and in residues) have been declared surplus in the USA and Russia. Both countries plan to convert the metal and oxide into mixed oxide fuel for nuclear power reactors. Russia has not yet decided what to do with the residues. The US will convert residues into a ceramic, which will then be over-poured with highly radioactive borosilicate glass. The radioactive glass is meant to provide a deterrent to recovery of plutonium, as required by a US standard. Here we show a waste form for plutonium residues, zirconia/boron carbide (ZrO 2/B 4C), with an unprecedented combination of properties: a single, radiation-resistant, and chemically durable phase contains the residues; billion-year-old natural analogs are available; and criticality safety is given under all conceivable disposal conditions. ZrO 2/B 4C can be disposed of directly, without further processing, making it attractive to all countries facing the task of plutonium disposal. The US standard for protection against recovery can be met by disposal of the waste form together with used reactor fuel.

Adaptation of the ICRP publication 66 respiratory tract model to data on plutonium biokinetics for Mayak workers.

PubMed

Khokhryakov, V F; Suslova, K G; Vostrotin, V V; Romanov, S A; Eckerman, K F; Krahenbuhl, M P; Miller, S C

2005-02-01

The biokinetics of inhaled plutonium were analyzed using compartment models representing their behavior within the respiratory tract, the gastrointestinal tract, and in systemic tissues. The processes of aerosol deposition, particle transport, absorption, and formation of a fixed deposit in the respiratory tract were formulated in the framework of the Human Respiratory Tract Model described in ICRP Publication 66. The values of parameters governing absorption and formation of the fixed deposit were established by fitting the model to the observations in 530 autopsy cases. The influence of smoking on mechanical clearance of deposited plutonium activity was considered. The dependence of absorption on the aerosol transportability, as estimated by in vitro methods (dialysis), was demonstrated. The results of this study were compared to those obtained from an earlier model of plutonium behavior in the respiratory tract, which was based on the same set of autopsy data. That model did not address the early phases of respiratory clearance and hence underestimated the committed lung dose by about 25% for plutonium oxides. Little difference in lung dose was found for nitrate forms.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maxwell, Sherrod L.; Culligan, Brian K.; Hutchison, Jay B.

A new rapid fusion method for the determination of plutonium in large rice samples has been developed at the Savannah River National Laboratory (Aiken, SC, USA) that can be used to determine very low levels of plutonium isotopes in rice. The recent accident at Fukushima Nuclear Power Plant in March, 2011 reinforces the need to have rapid, reliable radiochemical analyses for radionuclides in environmental and food samples. Public concern regarding foods, particularly foods such as rice in Japan, highlights the need for analytical techniques that will allow very large sample aliquots of rice to be used for analysis so thatmore » very low levels of plutonium isotopes may be detected. The new method to determine plutonium isotopes in large rice samples utilizes a furnace ashing step, a rapid sodium hydroxide fusion method, a lanthanum fluoride matrix removal step, and a column separation process with TEVA Resin cartridges. The method can be applied to rice sample aliquots as large as 5 kg. Plutonium isotopes can be determined using alpha spectrometry or inductively-coupled plasma mass spectrometry (ICP-MS). The method showed high chemical recoveries and effective removal of interferences. The rapid fusion technique is a rugged sample digestion method that ensures that any refractory plutonium particles are effectively digested. The MDA for a 5 kg rice sample using alpha spectrometry is 7E-5 mBq g{sup -1}. The method can easily be adapted for use by ICP-MS to allow detection of plutonium isotopic ratios.« less

CSER 01-008 Canning of Thermally Stabilized Plutonium Oxide Powder in PFP Glovebox HC-21A

DOE Office of Scientific and Technical Information (OSTI.GOV)

ERICKSON, D.G.

This document presents the analysis performed to support the canning operation in HC-21A. Most of the actual analysis was performed for the operation in HC-18M and HA-20MB, and is documented in HNF-2707 Rev I a (Erickson 2001a). This document will reference Erickson (2001a) as necessary to support the operation in HC-21A. The plutonium stabilization program at the Plutonium Finishing Plant (PFP) uses heat to convert plutonium-bearing materials into dry powder that is chemically stable for long term storage. The stabilized plutonium is transferred into one of several gloveboxes for the canning process, Gloveboxes HC-18M in Room 228'2, HA-20MB in Roommore » 235B, and HC-21A in Room 230B are to be used for this process. This document presents the analysis performed to support the canning operation in HC-21A. Most of the actual analysis was performed for the operation in HC-I8M and HA-20MB, and is documented in HNF-2707 Rev l a (Erickson 2001a). This document will reference Erickson (2001a) as necessary to support the operation in HC-21A. Evaluation of this operation included normal, base cases, and contingencies. The base cases took the normal operations for each type of feed material and added the likely off-normal events. Each contingency is evaluated assuming the unlikely event happens to the conservative base case. Each contingency was shown to meet the double contingency requirement. That is, at least two unlikely, independent, and concurrent changes in process conditions are required before a criticality is possible.« less

Recovery of 238PuO2 by Molten Salt Oxidation Processing of 238PuO2 Contaminated Combustibles (Part II)

NASA Astrophysics Data System (ADS)

Remerowski, Mary Lynn; Dozhier, C.; Krenek, K.; VanPelt, C. E.; Reimus, M. A.; Spengler, D.; Matonic, J.; Garcia, L.; Rios, E.; Sandoval, F.; Herman, D.; Hart, R.; Ewing, B.; Lovato, M.; Romero, J. P.

2005-02-01

Pu-238 heat sources are used to fuel radioisotope thermoelectric generators (RTG) used in space missions. The demand for this fuel is increasing, yet there are currently no domestic sources of this material. Much of the fuel is material reprocessed from other sources. One rich source of Pu-238 residual material is that from contaminated combustible materials, such as cheesecloth, ion exchange resins and plastics. From both waste minimization and production efficiency standpoints, the best solution is to recover this material. One way to accomplish separation of the organic component from these residues is a flameless oxidation process using molten salt as the matrix for the breakdown of the organic to carbon dioxide and water. The plutonium is retained in the salt, and can be recovered by dissolution of the carbonate salt in an aqueous solution, leaving the insoluble oxide behind. Further aqueous scrap recovery processing is used to purify the plutonium oxide. Recovery of the plutonium from contaminated combustibles achieves two important goals. First, it increases the inventory of Pu-238 available for heat source fabrication. Second, it is a significant waste minimization process. Because of its thermal activity (0.567 W per gram), combustibles must be packaged for disposition with much lower amounts of Pu-238 per drum than other waste types. Specifically, cheesecloth residues in the form of pyrolyzed ash (for stabilization) are being stored for eventual recovery of the plutonium.

Recovery of 238PuO2 by Molten Salt Oxidation Processing of 238PuO2 Contaminated Combustibles (Part II)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remerowski, Mary Lynn; Dozhier, C.; Krenek, K.

2005-02-06

Pu-238 heat sources are used to fuel radioisotope thermoelectric generators (RTG) used in space missions. The demand for this fuel is increasing, yet there are currently no domestic sources of this material. Much of the fuel is material reprocessed from other sources. One rich source of Pu-238 residual material is that from contaminated combustible materials, such as cheesecloth, ion exchange resins and plastics. From both waste minimization and production efficiency standpoints, the best solution is to recover this material. One way to accomplish separation of the organic component from these residues is a flameless oxidation process using molten salt asmore » the matrix for the breakdown of the organic to carbon dioxide and water. The plutonium is retained in the salt, and can be recovered by dissolution of the carbonate salt in an aqueous solution, leaving the insoluble oxide behind. Further aqueous scrap recovery processing is used to purify the plutonium oxide. Recovery of the plutonium from contaminated combustibles achieves two important goals. First, it increases the inventory of Pu-238 available for heat source fabrication. Second, it is a significant waste minimization process. Because of its thermal activity (0.567 W per gram), combustibles must be packaged for disposition with much lower amounts of Pu-238 per drum than other waste types. Specifically, cheesecloth residues in the form of pyrolyzed ash (for stabilization) are being stored for eventual recovery of the plutonium.« less

Analysis of IAEA Environmental Samples for Plutonium and Uranium by ICP/MS in Support Of International Safeguards

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, Orville T.; Olsen, Khris B.; Thomas, May-Lin P.

2008-05-01

A method for the separation and determination of total and isotopic uranium and plutonium by ICP-MS was developed for IAEA samples on cellulose-based media. Preparation of the IAEA samples involved a series of redox chemistries and separations using TRU® resin (Eichrom). The sample introduction system, an APEX nebulizer (Elemental Scientific, Inc), provided enhanced nebulization for a several-fold increase in sensitivity and reduction in background. Application of mass bias (ALPHA) correction factors greatly improved the precision of the data. By combining the enhancements of chemical separation, instrumentation and data processing, detection levels for uranium and plutonium approached high attogram levels.

Independent Verification Survey of the Clean Coral Storage Pile at the Johnston Atoll Plutonium Contaminated Soil Remediation Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson-Nichols, M.J.; Egidi, P.V.; Roemer, E.K.

2000-09-01

f I The Oak Ridge National Laboratory (ORNL) Environmental Technology Section conducted an independent verification (IV) survey of the clean storage pile at the Johnston Atoll Plutonium Contaminated Soil Remediation Project (JAPCSRP) from January 18-25, 1999. The goal of the JAPCSRP is to restore a 24-acre area that was contaminated with plutonium oxide particles during nuclear testing in the 1960s. The selected remedy was a soil sorting operation that combined radiological measurements and mining processes to identify and sequester plutonium-contaminated soil. The soil sorter operated from about 1990 to 1998. The remaining clean soil is stored on-site for planned beneficialmore » use on Johnston Island. The clean storage pile currently consists of approximately 120,000 m3 of coral. ORNL conducted the survey according to a Sampling and Analysis Plan, which proposed to provide an IV of the clean pile by collecting a minimum number (99) of samples. The goal was to ascertain wi th 95% confidence whether 97% of the processed soil is less than or equal to the accepted guideline (500-Bq/kg or 13.5-pCi/g) total transuranic (TRU) activity.« less

KSC-97PC1070

NASA Image and Video Library

1997-07-18

Jet Propulsion Laboratory (JPL) workers use a borescope to verify pressure relief device bellows integrity on a radioisotope thermoelectric generator (RTG) which has been installed on the Cassini spacecraft in the Payload Hazardous Servicing Facility. The activity is part of the mechanical and electrical verification testing of RTGs during prelaunch processing. RTGs use heat from the natural decay of plutonium to generate electric power. The three RTGs on Cassini will enable the spacecraft to operate far from the Sun where solar power systems are not feasible. They will provide electrical power to Cassini on its 6.7-year trip to the Saturnian system and during its four-year mission at Saturn. The Cassini mission is scheduled for an Oct. 6 launch aboard a Titan IVB/Centaur expendable launch vehicle. Cassini is built and managed for NASA by JPL

FRACTIONAL DISTILLATION SEPARATION OF PLUTONIUM VALUES FROM LIGHT ELEMENT VALUES

DOEpatents

Cunningham, B.B.

1957-12-17

A process is described for removing light element impurities from plutonium. It has been found that plutonium contaminated with impurities may be purified by converting the plutonium to a halide and purifying the halide by a fractional distillation whereby impurities may be distilled from the plutonium halide. A particularly effective method includes the step of forming a lower halide such as the trior tetrahalide and distilling the halide under conditions such that no decomposition of the halide occurs. Molecular distillation methods are particularly suitable for this process. The apparatus may comprise an evaporation plate with means for heating it and a condenser surface with means for cooling it. The condenser surface is placed at a distance from the evaporating surface less than the mean free path of molecular travel of the material being distilled at the pressure and temperature used. The entire evaporating system is evacuated until the pressure is about 10/sup -4/ millimeters of mercury. A high temperuture method is presented for sealing porous materials such as carbon or graphite that may be used as a support or a moderator in a nuclear reactor. The carbon body is subjected to two surface heats simultaneously in an inert atmosphere; the surface to be sealed is heated to 1500 degrees centigrade; and another surface is heated to 300 degrees centigrade, whereupon the carbon vaporizes and flows to the cooler surface where it is deposited to seal that surface. This method may be used to seal a nuclear fuel in the carbon structure.

In-line Kevlar filters for microfiltration of transuranic-containing liquid streams.

PubMed

Gonzales, G J; Beddingfield, D H; Lieberman, J L; Curtis, J M; Ficklin, A C

1992-06-01

The Department of Energy Rocky Flats Plant has numerous ongoing efforts to minimize the generation of residue and waste and to improve safety and health. Spent polypropylene liquid filters held for plutonium recovery, known as "residue," or as transuranic mixed waste contribute to storage capacity problems and create radiation safety and health considerations. An in-line process-liquid filter made of Kevlar polymer fiber has been evaluated for its potential to: (1) minimize filter residue, (2) recover economically viable quantities of plutonium, (3) minimize liquid storage tank and process-stream radioactivity, and (4) reduce potential personnel radiation exposure associated with these sources. Kevlar filters were rated to less than or equal to 1 mu nominal filtration and are capable of reducing undissolved plutonium particles to more than 10 times below the economic discard limit, however produced high back-pressures and are not yet acid resistant. Kevlar filters performed independent of loaded particles serving as a sieve. Polypropylene filters removed molybdenum particles at efficiencies equal to Kevlar filters only after loading molybdenum during recirculation events. Kevlars' high-efficiency microfiltration of process-liquid streams for the removal of actinides has the potential to reduce personnel radiation exposure by a factor of 6 or greater, while simultaneously achieving a reduction in the generation of filter residue and waste by a factor of 7. Insoluble plutonium may be recoverable from Kevlar filters by incineration.

Transuranic solid waste management programs. Progress report, July--December 1975

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-09-01

Progress is reported for three transuranic solid waste management programs funded at the Los Alamos Scientific Laboratory (LASL) by the Energy Research and Development Administration (ERDA) Division of Fuel Cycle and Production (NFCP). Under the Transuranic Waste Research and Development Program, continued studies have shown the potential attractiveness of fiber drums as an acceptable substitute for the current mild steel storage containers. Various fire retardants have been evaluated, with one indicating significant ability to inhibit fire propagation. Continued radiolysis studies, under laboratory and field conditions, continue to reaffirm earlier LASL results indicating no significant hazard from radiolytic reactions, assuming nomore » change in current allowable loadings. Care must be exercised to differentiate between radiolytic and chemical reactions. Other efforts have identified a modification of chemical processing to reduce the amounts of plutonium requiring retrievable storage. Studies are also in progress to enhance the sensitivity of the LASL MEGAS assay system. The Transuranic-Contaminated Solid Waste Treatment Development Facility building was 72 percent complete as of December 31, 1975, which is in accord with the existing schedule. Procurement of process components is also on schedule. Certain modifications to the facility have been made, and various pre-facility experiments on waste container handling and processing have been completed. The program for the Evaluation of Transuranic-Contaminated Radioactive Waste Disposal Areas continued development of various computer modules for simulation of radionuclide transport within the biosphere. In addition, program staff contributed to an ERDA document on radioactive waste management through the preparation of a report on burial of radioactive waste at ERDA-contractor and commercial sites.« less

Ceramification: A plutonium immobilization process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rask, W.C.; Phillips, A.G.

1996-05-01

This paper describes a low temperature technique for stabilizing and immobilizing actinide compounds using a combination process/storage vessel of stainless steel, in which measured amounts of actinide nitrate solutions and actinide oxides (and/or residues) are systematically treated to yield a solid article. The chemical ceramic process is based on a coating technology that produces rare earth oxide coatings for defense applications involving plutonium. The final product of this application is a solid, coherent actinide oxide with process-generated encapsulation that has long-term environmental stability. Actinide compounds can be stabilized as pure materials for ease of re-use or as intimate mixtures withmore » additives such as rare earth oxides to increase their degree of proliferation resistance. Starting materials for the process can include nitrate solutions, powders, aggregates, sludges, incinerator ashes, and others. Agents such as cerium oxide or zirconium oxide may be added as powders or precursors to enhance the properties of the resulting solid product. Additives may be included to produce a final product suitable for use in nuclear fuel pellet production. The process is simple and reduces the time and expense for stabilizing plutonium compounds. It requires a very low equipment expenditure and can be readily implemented into existing gloveboxes. The process is easily conducted with less associated risk than proposed alternative technologies.« less

Bioprocessing of a stored mixed liquid waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wolfram, J.H.; Rogers, R.D.; Finney, R.

1995-12-31

This paper describes the development and results of a demonstration for a continuous bioprocess for mixed waste treatment. A key element of the process is an unique microbial strain which tolerates high levels of aromatic solvents and surfactants. This microorganism is the biocatalysis of the continuous flow system designed for the processing of stored liquid scintillation wastes. During the past year a process demonstration has been conducted on commercial formulation of liquid scintillation cocktails (LSC). Based on data obtained from this demonstration, the Ohio EPA granted the Mound Applied Technologies Lab a treatability permit allowing the limited processing of actualmore » mixed waste. Since August 1994, the system has been successfully processing stored, {open_quotes}hot{close_quotes} LSC waste. The initial LSC waste fed into the system contained 11% pseudocumene and detectable quantities of plutonium. Another treated waste stream contained pseudocumene and tritium. Data from this initial work shows that the hazardous organic solvent, and pseudocumene have been removed due to processing, leaving the aqueous low level radioactive waste. Results to date have shown that living cells are not affected by the dissolved plutonium and that 95% of the plutonium was sorbed to the biomass. This paper discusses the bioprocess, rates of processing, effluent, and the implications of bioprocessing for mixed waste management.« less

The Alliance of Advanced Process Control and Accountability – A Future Safeguards-By-Design Tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lumetta, Gregg J.; Bresee, James C.; Paviet, Patricia D.

For any chemical separation process producing a valuable product, a material balance is an important process control measurement. That is particularly true for the separation of actinides from irradiated nuclear fuel, not only for their intrinsic value but also because an incomplete material balance may indicate diversion for unauthorized use. The DOE Office of Nuclear Energy is currently carrying out at the Pacific Northwest National Laboratory an experimental measurement of how well and with what precision current technologies can implement near real-time actinide material balances. This measurement effort is called the CoDCon project. It involves the separation of a productmore » with a 70/30 uranium/plutonium mass ratio. Initial tests will use dissolved fuel simulants prepared with pure uranium and plutonium nitrates at the same input ratios as irradiated fuel. Subsequent testing with actual irradiated fuel would be done to verify the results obtained with simulants. The experiments will use advanced on-line instrumentation supported by dynamic process models. Since accountability uncertainties could mask diversions, the aim of the project is not only to measure present-day capabilities but also, through sensitivity analyses, to identify those measurements with the greatest potential for overall material-balance improvements. The latter results will help identify priorities for future fuel cycle R&D programs. Advanced separations process control and material accountability technologies thus have a common goal: to provide the best tools available for safeguards-by-design [defined by the International Atomic Energy Agency (IAEA) as the integration of the design of a new nuclear facility through planning, construction, operation and decommissioning]. Since the potential domestic use of CoDCon results may be later than their possible foreign applications, arrangements may be feasible for possible bilateral or multinational cooperation in the CoDCon project.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paton, Ian

The Rocky Flats Environmental Technology Site (RFETS) is a Department of Energy facility located approximately 16 miles northwest of Denver, Colorado. Processing and fabrication of nuclear weapons components occurred at Rocky Flats from 1952 through 1989. Operations at the Site included the use of several radionuclides, including plutonium-239/240 (Pu), americium-241 (Am), and various uranium (U) isotopes, as well as several types of chlorinated solvents. The historic operations resulted in legacy contamination, including contaminated facilities, process waste lines, buried wastes and surface soil contamination. Decontamination and removal of buildings at the site was completed in late 2005, culminating more than tenmore » years of active environmental remediation work. The Corrective Action Decision/Record of Decision was subsequently approved in 2006, signifying regulatory approval and closure of the site. The use of RFETS as a National Wildlife Refuge is scheduled to be in full operation by 2012. To develop a plan for remediating different types of radionuclide contaminants present in the RFETS environment required understanding the different environmental transport pathways for the various actinides. Developing this understanding was the primary objective of the Actinide Migration Evaluation (AME) project. Findings from the AME studies were used in the development of RFETS remediation strategies. The AME project focused on issues of actinide behavior and mobility in surface water, groundwater, air, soil and biota at RFETS. For the purposes of the AME studies, actinide elements addressed included Pu, Am, and U. The AME program, funded by DOE, brought together personnel with a broad range of relevant expertise in technical investigations. The AME advisory panel identified research investigations and approaches that could be used to solve issues related to actinide migration at the Site. An initial step of the AME was to develop a conceptual model to provide a qualitative description of the relationships among potential actinide sources and transport pathways at RFETS. One conceptual model was developed specifically for plutonium and americium, because of their similar geochemical and transport properties. A separate model was developed for uranium because of its different properties and mobility in the environment. These conceptual models were guidelines for quantitative analyses described in the RFETS Pathway Analysis Report, which used existing data from the literature as well as site-specific analyses, including field, laboratory and modeling studies to provide quantitative estimates of actinide migration in the RFETS environment. For pathways where more than one method was used to estimate offsite loads for a specific pathway, the method yielding the highest estimated off-site was used for comparison purposes. For all actinides studied, for pre-remediation conditions, air and surface water were identified to be the dominant transport mechanisms. The estimated annual airborne plutonium-239/240 load transported off site exceeded the surface water load by roughly a factor of 40. However, despite being the largest transport pathway, airborne radionuclide concentrations at the monitoring location with the highest measurements during the period studied were less than two percent of the allowable 10 milli-rem standard governing DOE facilities. Estimated actinide loads for other pathways were much less. Shallow groundwater was approximately two orders of magnitude lower, or 1/100 of the load conveyed in surface water. The estimated biological pathway load for plutonium was approximately five orders of magnitude less, or 1/100,000, of the load estimated for surface-water. The pathway analysis results were taken into consideration during subsequent remediation activities that occurred at the site. For example, when the 903 Pad area was remediated to address elevated concentrations of Pu and Am in the surface soil, portable tent structures were constructed to prevent wind and water erosion from occurring while remediation activities took place. Following remediation of the 903 Pad and surrounding area, coconut erosion blankets were installed to mitigate erosion effects while vegetation was reestablished [2]. These measures were effective tools to address the primary transport mechanisms identified, coupling the scientific understanding of the site with the remediation strategy.« less

Technical viability and development needs for waste forms and facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pegg, I.; Gould, T.

1996-05-01

The objective of this breakout session was to provide a forum to discuss technical issues relating to plutonium-bearing waste forms and their disposal facilities. Specific topics for discussion included the technical viability and development needs associated with the waste forms and/or disposal facilities. The expected end result of the session was an in-depth (so far as the limited time would allow) discussion of key issues by the session participants. The session chairs expressed allowance for, and encouragement of, alternative points of view, as well as encouragement for discussion of any relevant topics not addressed in the paper presentations. It wasmore » not the intent of this session to recommend or advocate any one technology over another.« less

Industrial research for transmutation scenarios

NASA Astrophysics Data System (ADS)

Camarcat, Noel; Garzenne, Claude; Le Mer, Joël; Leroyer, Hadrien; Desroches, Estelle; Delbecq, Jean-Michel

2011-04-01

This article presents the results of research scenarios for americium transmutation in a 22nd century French nuclear fleet, using sodium fast breeder reactors. We benchmark the americium transmutation benefits and drawbacks with a reference case consisting of a hypothetical 60 GWe fleet of pure plutonium breeders. The fluxes in the various parts of the cycle (reactors, fabrication plants, reprocessing plants and underground disposals) are calculated using EDF's suite of codes, comparable in capabilities to those of other research facilities. We study underground thermal heat load reduction due to americium partitioning and repository area minimization. We endeavor to estimate the increased technical complexity of surface facilities to handle the americium fluxes in special fuel fabrication plants, americium fast burners, special reprocessing shops, handling equipments and transport casks between those facilities.

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1993-12-01

This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose reduction activities, with a focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, testmore » and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and aocelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts. An author index and a subject index are provided to facilitate use. Both indices contain the abstract numbers from previous volumes, as well as the current volume. Information that the reader feels might be included in the next volume of this bibliography should be submitted to the BNL ALARA Center.« less

Fabrication of thorium bearing carbide fuels

DOEpatents

Gutierrez, Rueben L.; Herbst, Richard J.; Johnson, Karl W. R.

1981-01-01

Thorium-uranium carbide and thorium-plutonium carbide fuel pellets have been fabricated by the carbothermic reduction process. Temperatures of 1750.degree. C. and 2000.degree. C. were used during the reduction cycle. Sintering temperatures of 1800.degree. C. and 2000.degree. C. were used to prepare fuel pellet densities of 87% and >94% of theoretical, respectively. The process allows the fabrication of kilogram quantities of fuel with good reproducibility of chemicals and phase composition. Methods employing liquid techniques that form carbide microspheres or alloying-techniques which form alloys of thorium-uranium or thorium-plutonium suffer from limitation on the quantities processed of because of criticality concerns and lack of precise control of process conditions, respectively.

Status summary of chemical processing development in plutonium-238 supply program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, Emory D.; Benker, Dennis; Wham, Robert M.

This document summarizes the status of development of chemical processing in the Plutonium-238 Supply Program (PSP) near the end of Demonstration 1. The objective of the PSP is “to develop, demonstrate, and document a production process that meets program objectives and to prepare for its operation” (Frazier et al. 2016). Success in the effort includes establishing capability using the current infrastructure to produce Np targets for irradiation in Department of Energy research reactors, chemically processing the irradiated targets to separate and purify the produced Pu and transferring the PuO 2 product to Los Alamos National Laboratory (LANL) at an averagemore » rate of 1.5 kg/y.« less

Falling Behind: International Scrutiny of the Peaceful Atom

DTIC Science & Technology

2008-02-01

confused with critical masses .5 Significant quantity values currently in use by the IAEA are given in Table 1. In a previous Natural Resources... actinides and fission products would not add significantly to the plutonium mass , the state could divert the spiked plutonium to a small clandestine...and manufacturing processes and should not be 152 confused with critical masses . Significant quantities are used in establishing the quantity

The measurement of U(VI) and Np(IV) mass transfer in a single stage centrifugal contactor

NASA Astrophysics Data System (ADS)

May, I.; Birkett, E. J.; Denniss, I. S.; Gaubert, E. T.; Jobson, M.

2000-07-01

BNFL currently operates two reprocessing plants for the conversion of spent nuclear fuel into uranium and plutonium products for fabrication into uranium oxide and mixed uranium and plutonium oxide (MOX) fuels. To safeguard the future commercial viability of this process, BNFL is developing novel single cycle flowsheets that can be operated in conjunction with intensified centrifugal contactors.

IMPROVEMENT UPON THE CARRIER PRECIPITATION OF PLUTONIUM IONS FROM NITRIC ACID SOLUTIONS

DOEpatents

James, R.A.; Thompson, S.G.

1958-12-23

A process is reported for improving the removal of plutonlum by carrier precipitation by the addition of nitrite ions to a nitrlc acid solutlon of neutronirradiated unanium so as to destroy any hydrazine that may be present in the solution since the hydrazine tends to complex the tetravalent plutonium and prevents removal by the carrier precipltate, such as bismuth phospbate.

Next Generation Safeguards Initiative research to determine the Pu mass in spent fuel assemblies: Purpose, approach, constraints, implementation, and calibration

NASA Astrophysics Data System (ADS)

Tobin, S. J.; Menlove, H. O.; Swinhoe, M. T.; Schear, M. A.

2011-10-01

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy has funded a multi-lab/multi-university collaboration to quantify the plutonium mass in spent nuclear fuel assemblies and to detect the diversion of pins from them. The goal of this research effort is to quantify the capability of various non-destructive assay (NDA) technologies as well as to train a future generation of safeguards practitioners. This research is "technology driven" in the sense that we will quantify the capabilities of a wide range of safeguards technologies of interest to regulators and policy makers; a key benefit to this approach is that the techniques are being tested in a unified manner. When the results of the Monte Carlo modeling are evaluated and integrated, practical constraints are part of defining the potential context in which a given technology might be applied. This paper organizes the commercial spent fuel safeguard needs into four facility types in order to identify any constraints on the NDA system design. These four facility types are the following: future reprocessing plants, current reprocessing plants, once-through spent fuel repositories, and any other sites that store individual spent fuel assemblies (reactor sites are the most common facility type in this category). Dry storage is not of interest since individual assemblies are not accessible. This paper will overview the purpose and approach of the NGSI spent fuel effort and describe the constraints inherent in commercial fuel facilities. It will conclude by discussing implementation and calibration of measurement systems. This report will also provide some motivation for considering a couple of other safeguards concepts (base measurement and fingerprinting) that might meet the safeguards need but not require the determination of plutonium mass.

Survey of glass plutonium contents and poison selection

DOE Office of Scientific and Technical Information (OSTI.GOV)

Plodinec, M.J.; Ramsey, W.G.; Ellison, A.J.G.

1996-05-01

If plutonium and other actinides are to be immobilized in glass, then achieving high concentrations in the glass is desirable. This will lead to reduced costs and more rapid immobilization. However, glasses with high actinide concentrations also bring with them undersirable characteristics, especially a greater concern about nuclear criticality, particularly in a geologic repository. The key to achieving a high concentration of actinide elements in a glass is to formulate the glass so that the solubility of actinides is high. At the same time, the glass must be formulated so that the glass also contains neutron poisons, which will preventmore » criticality during processing and in a geologic repository. In this paper, the solubility of actinides, particularly plutonium, in three types of glasses are discussed. Plutonium solubilities are in the 2-4 wt% range for borosilicate high-level waste (HLW) glasses of the type which will be produced in the US. This type of glass is generally melted at relatively low temperatures, ca. 1150{degrees}C. For this melting temperature, the glass can be reformulated to achieve plutonium solubilities of at least 7 wt%. This low melting temperature is desirable if one must retain volatile cesium-137 in the glass. If one is not concerned about cesium volatility, then glasses can be formulated which can contain much larger amounts of plutonium and other actinides. Plutonium concentrations of at least 15 wt% have been achieved. Thus, there is confidence that high ({ge}5 wt%) concentrations of actinides can be achieved under a variety of conditions.« less

Anthropogenic plutonium-244 in the environment: Insights into plutonium’s longest-lived isotope

DOE PAGES

Armstrong, Christopher R.; Brant, Heather A.; Nuessle, Patterson R.; ...

2016-02-22

Owing to the rich history of heavy element production in the unique high flux reactors that operated at the Savannah River Site, USA (SRS) decades ago, trace quantities of plutonium with highly unique isotopic characteristics still persist today in the SRS terrestrial environment. Development of an effective sampling, processing, and analysis strategy enables detailed monitoring of the SRS environment, revealing plutonium isotopic compositions, e.g., 244Pu, that reflect the unique legacy of plutonium production at SRS. This work describes the first long-term investigation of anthropogenic 244Pu occurrence in the environment. Environmental samples, consisting of collected foot borne debris, were taken atmore » SRS over an eleven year period, from 2003 to 2014. Separation and purification of trace plutonium was carried out followed by three stage thermal ionization mass spectrometry (3STIMS) measurements for plutonium isotopic content and isotopic ratios. Furthermore, significant 244Pu was measured in all of the years sampled with the highest amount observed in 2003. The 244Pu content, in femtograms (fg = 10 –15 g) per gram, ranged from 0.31 fg/g to 44 fg/g in years 2006 and 2003 respectively. In all years, the 244Pu/ 239Pu atom ratios were significantly higher than global fallout, ranging from 0.003 to 0.698 in years 2014 and 2003 respectively.« less

Anthropogenic plutonium-244 in the environment: Insights into plutonium’s longest-lived isotope

PubMed Central

Armstrong, Christopher R.; Brant, Heather A.; Nuessle, Patterson R.; Hall, Gregory; Cadieux, James R.

2016-01-01

Owing to the rich history of heavy element production in the unique high flux reactors that operated at the Savannah River Site, USA (SRS) decades ago, trace quantities of plutonium with highly unique isotopic characteristics still persist today in the SRS terrestrial environment. Development of an effective sampling, processing, and analysis strategy enables detailed monitoring of the SRS environment, revealing plutonium isotopic compositions, e.g., 244Pu, that reflect the unique legacy of plutonium production at SRS. This work describes the first long-term investigation of anthropogenic 244Pu occurrence in the environment. Environmental samples, consisting of collected foot borne debris, were taken at SRS over an eleven year period, from 2003 to 2014. Separation and purification of trace plutonium was carried out followed by three stage thermal ionization mass spectrometry (3STIMS) measurements for plutonium isotopic content and isotopic ratios. Significant 244Pu was measured in all of the years sampled with the highest amount observed in 2003. The 244Pu content, in femtograms (fg = 10−15 g) per gram, ranged from 0.31 fg/g to 44 fg/g in years 2006 and 2003 respectively. In all years, the 244Pu/239Pu atom ratios were significantly higher than global fallout, ranging from 0.003 to 0.698 in years 2014 and 2003 respectively. PMID:26898531

The use of nuclear data in the field of nuclear fuel recycling

NASA Astrophysics Data System (ADS)

Martin, Julie-Fiona; Launay, Agnès; Grassi, Gabriele; Binet, Christophe; Lelandais, Jacques; Lecampion, Erick

2017-09-01

AREVA NC La Hague facility is the first step of the nuclear fuel recycling process implemented in France. The processing of the used fuel is governed by high standards of criticality-safety, and strong expectations on the quality of end-products. From the received used fuel assemblies, the plutonium and the uranium are extracted for further energy production purposes within the years following the reprocessing. Furthermore, the ultimate waste - fission products and minor actinides on the one hand, and hulls and end-pieces on the other hand - is adequately packaged for long term disposal. The used fuel is therefore separated into very different materials, and time scales which come into account may be longer than in some other nuclear fields of activity. Given the variety of the handled nuclear materials, as well as the time scales at stake, the importance given to some radionuclides, and hence to the associated nuclear data, can also be specific to the AREVA NC La Hague plant. A study has thus been led to identify a list of the most important radionuclides for the AREVA NC La Hague plant applications, relying on the running constraints of the facility, and the end-products expectations. The activities at the AREVA NC La Hague plant are presented, and the methodology to extract the most important radionuclides for the reprocessing process is detailed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cournoyer, Michael E; George, Gerald L; Dodge, Robert L

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 the use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Radiation shielding is commonly used to protect the glovebox worker from unintentional direct and secondary radiation exposure, while working with plutonium-238 and plutonium-239. In these environments, low-energy photons, i.e., those less than 250 keY, are encountered.more » Shielding glove box gloves are traditionally composed of lead-based materials, but these are now considered hazardous waste. This has prompted the development of new, nonhazardous- shielding gJovebox gloves. No studies, however, have investigated the effectiveness of these new glovebox gloves. We examined both leaded and nonhazardous- shielding glovebox gloves and compared their attenuation effectiveness over the energy range of interest at TA-55. All measurements are referenced to lead sheets, allowing direct comparisons to the common industry standard of 0.1 mm lead equivalent material. The attenuation properties of both types of glovebox gloves vary with energy, making it difficult for manufacturers to claim lead equivalency across the entire energy range used at TA-55. The positions of materials' photon energy absorption edges, which are particularly important to improved attenuation performance, depending upon the choice of radiation energy range, are discussed. This effort contributes to the Los Alamos National Laboratory Continuous Improvement Program by improving the efficiency, cost effectiveness, and formality of glovebox operations.« less

Natural radionuclide and plutonium content in Black Sea bottom sediments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strezov, A.; Stoilova, T.; Yordanova, I.

1996-01-01

The content of uranium, thorium, radium, lead, polonium, and plutonium in bottom sediments and algae from two locations at the Bulgarian Black Sea coast have been determined. Some parent:progeny ratios for evaluation of the geochemical behavior of the nuclides have been estimated as well. The extractable and total uranium and thorium are determined by two separate radiochemical procedures to differentiate the more soluble chemical forms of the elements and to estimate the potential hazard for the biosphere and for humans. No distinct seasonal variation as well as no significant change in total and extractable uranium (also for {sup 226}Ra) contentmore » is observed. The same is valid for extractable thorium while the total thorium content in the first two seasons is slightly higher. Our data show that {sup 210}Po content is accumulated more in the sediments than {sup 210}Pb, and the evaluated disequilibria suggest that the two radionuclides belong to more recent sediment layers deposited in the slime samples compared to the silt ones for the different seasons. The obtained values for plutonium are in the lower limits of the data cited in literature, which is quite clear as there are no plutonium discharge facilities at the Bulgarian Black Sea coast. The obtained values for the activity ratio {sup 238}Pu: {sup 239+240}Pu are higher for Bjala sediments compared to those of Kaliakra. The ratio values are out of the variation range for the global contamination with weapon tests fallout plutonium which is probably due to Chernobyl accident contribution. The dependence of natural radionuclide content on the sediment type as well as the variation of nuclide accumulation for two types of algae in two sampling locations for five consecutive seasons is evaluated. No serious contamination with natural radionuclides in the algae is observed. 38 refs., 6 figs., 7 tabs.« less

Combined radiochemical procedure for determination of plutonium, americium and strontium-90 in the soil samples from SNTS

NASA Astrophysics Data System (ADS)

Kazachevskii, I. V.; Lukashenko, S. N.; Chumikov, G. N.; Chakrova, E. T.; Smirin, L. N.; Solodukhin, V. P.; Khayekber, S.; Berdinova, N. M.; Ryazanova, L. A.; Bannyh, V. I.; Muratova, V. M.

1999-01-01

The results of combined radiochemical procedure for the determination of plutonium, americium and90Sr (via measurement of90Y) in the soil samples from SNTS are presented. The processes of co-precipitation of these nuclides with calcium fluoride in the strong acid solutions have been investigated. The conditions for simultaneous separation of americium and yttrium using extraction chromatography have been studied. It follows from analyses of real soil samples that the procedure developed provides the chemical recovery of plutonium and yttrium in the range of 50-95% and 60-95%, respectively. The execution of the procedure requires 3.5 working days including a sample decomposition study.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Veirs, Douglas Kirk; Stroud, Mary Ann; Berg, John M.

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 amore » 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.« less

ESTIMATION OF INTERNAL EXPOSURE TO URANIUM WITH UNCERTAINTY FROM URINALYSIS DATA USING THE InDEP COMPUTER CODE

PubMed Central

Anderson, Jeri L.; Apostoaei, A. Iulian; Thomas, Brian A.

2015-01-01

The National Institute for Occupational Safety and Health (NIOSH) is currently studying mortality in a cohort of 6409 workers at a former uranium processing facility. As part of this study, over 220 000 urine samples were used to reconstruct organ doses due to internal exposure to uranium. Most of the available computational programs designed for analysis of bioassay data handle a single case at a time, and thus require a significant outlay of time and resources for the exposure assessment of a large cohort. NIOSH is currently supporting the development of a computer program, InDEP (Internal Dose Evaluation Program), to facilitate internal radiation exposure assessment as part of epidemiological studies of both uranium- and plutonium-exposed cohorts. A novel feature of InDEP is its batch processing capability which allows for the evaluation of multiple study subjects simultaneously. InDEP analyses bioassay data and derives intakes and organ doses with uncertainty estimates using least-squares regression techniques or using the Bayes’ Theorem as applied to internal dosimetry (Bayesian method). This paper describes the application of the current version of InDEP to formulate assumptions about the characteristics of exposure at the study facility that were used in a detailed retrospective intake and organ dose assessment of the cohort. PMID:22683620

Closure Report for Corrective Action Unit 415: Project 57 No. 1 Plutonium Dispersion (NTTR) Nevada Test and Training Range, Nevada, Revision 0 with ROTC-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cabble, Kevin J.; Boehlecke, Robert F.

This Closure Report (CR) presents information supporting the closure of Corrective Action Unit (CAU) 415: Project 57 No. 1 Plutonium Dispersion, which is located on Range 4808A of the Nevada Test and Training Range (NTTR). This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; U.S. Department of Energy (DOE), Environmental Management; U.S. Department of Defense; and DOE, Legacy Management. CAU 415 comprises one corrective action site (CAS): NAFR-23-02, Pu Contaminated Soil. The purpose of this CR is to provide justification and documentation supporting the recommendationmore » that no further corrective action is needed for CAU 415 based on the implementation of the corrective action of Closure in Place.« less

Nuclear Archeology in a Bottle: Evidence of Pre-Trinity U.S. Weapons Activities from a Waste Burial Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schwantes, Jon M.; Douglas, Matthew; Bonde, Steven E.

2009-02-15

During World War II, the Hanford Site in Washington was chosen for plutonium production. In 2004, a bottle containing a sample of plutonium was recovered from a Hanford waste trench. Isotopic age dating indicated the sample was separated from the fuel pellet 64 ±2.8 years earlier. Detectable products of secondary nuclear reactions, such as 22Na, proved useful as 1) a detectable analog for alpha emitting actinides, 2) an indicator of sample splitting, and 3) a measure of the time since sample splitting. The sample origin was identified as the X-10 reactor, Oak Ridge, TN. Corroborated by historical documents, we concludedmore » this sample was part of the first batch of Pu separated at T-Plant, Hanford, the world’s first industrial-scale reprocessing facility, on December 9, 1944.« less

Overview of reductants utilized in nuclear fuel reprocessing/recycling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paviet-Hartmann, P.; Riddle, C.; Campbell, K.

2013-07-01

The most widely used reductant to partition plutonium from uranium in the Purex process was ferrous sulfamate, other alternates were proposed such as hydrazine-stabilized ferrous nitrate or uranous nitrate, platinum catalyzed hydrogen, and hydrazine, hydroxylamine salts. New candidates to replace hydrazine or hydroxylamine nitrate (HAN) are pursued worldwide. They may improve the performance of the industrial Purex process towards different operations such as de-extraction of plutonium and reduction of the amount of hydrazine which will limit the formation of hydrazoic acid. When looking at future recycling technologies using hydroxamic ligands, neither acetohydroxamic acid (AHA) nor formohydroxamic acid (FHA) seem promisingmore » because they hydrolyze to give hydroxylamine and the parent carboxylic acid. Hydroxyethylhydrazine, HOC{sub 2}H{sub 4}N{sub 2}H{sub 3} (HEH) is a promising non-salt-forming reductant of Np and Pu ions because it is selective to neptunium and plutonium ions at room temperature and at relatively low acidity, it could serve as a replacement of HAN or AHA for the development of a novel used nuclear fuel recycling process.« less

Development of an alternate pathway for materials destined for disposition to WIPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ayers, Georgette Y; Mckerley, Bill; Veazey, Gerald W

2010-01-01

The Los Alamos National Laboratory currently has an inventory of process residues that may be viable candidates for disposition to the Waste Isolation Pilot Project (WIPP) located at Carlsbad, New Mexico. A recent 'Attractiveness Level D' exemption allows for the discard of specified intractable materials regardless of the percent plutonium. However, the limits with respect to drum loadings must be met. Cementation is a key component of the aqueous nitrate flowsheet and serves as a 'bleed-off' stream for impurities separated from the plutonium during processing operations. The main 'feed' to the cementation operations are the 'bottoms' from the evaporation process.more » In the majority of cases, the cemented bottoms contain less than the allowed amount per drum for WIPP acceptance. This project would expand the route to WIPP for items that have no defined disposition path, are difficult to process, have been through multiple passes, have no current recovery operations available to recover the plutonium and that are amenable to cementation. This initial work will provide the foundation for a full scale disposition pathway of the candidate materials. Once the pathway has been expanded and a cementation matrix developed, routine discard activities will be initiated.« less

Immobilization of plutonium from solutions on porous matrices by the method of high temperature sorption

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nardova, A.K.; Filippov, E.A.; Glagolenko, Y.B.

1996-05-01

This report presents the results of investigations of plutonium immobilization from solutions on inorganic matrices with the purpose of producing a solid waste form. High-temperature sorption is described which entails the adsorption of radionuclides from solutions on porous, inorganic matrices, as for example silica gel. The solution is brought to a boil with additional thermal process (calcination) of the saturated granules.

242-A Evaporator/plutonium uranium extraction (PUREX) effluent treatment facility (ETF) nonradioactive air emission test report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hill, J.S., Westinghouse Hanford

1996-05-10

This report shows the methods used to test the stack gas outlet concentration and emission rate of Volatile Organic Compounds as Total Non-Methane Hydrocarbons in parts per million by volume,grams per dry standard cubic meter, and grams per minute from the PUREX ETF stream number G6 on the Hanford Site. Test results are shown in Appendix B.1.

Report on the FY 1986 Activities of the Defense Science Study Group. Volume 1.

DTIC Science & Technology

1987-05-01

Reactors Improved Techniques for Wavefront Sensing and Correction in Adaptive Optics Hypervelocity Launchers Underground Facilities 0 Automated...oceanography and sound propagation in partially coherent media such as the turbulent ocean. There are large fixed arrays such as the Sound Surveillance System...Aircraft Continuous Patrol Aircraft Miscellaneous Studies Review of the Plutonium Special Isotope Separation Program of the DOE 4r Fusion Fission Hybrid

Hazardous Waste Surveys of Two Army Installations and an Army Hospital.

DTIC Science & Technology

1980-08-01

232 Nickel-63 Uranium-238 Plutonium-239 Polonium - 210 6 Army Medical Treatment Facilities: General Administration Army Regulation (AR) 40-2, 42A peren...Categories 10 2 Waste Matrix 14 3 Search Format 16 4 Field Sanitation Unit Personal Health Supplies 19 5 Company Vehicle Maintenance Supplies...increasing industrialization of society, coupled with an equally increasing environmental and health safety awareness, has created a long list of wastes

AMS of the Minor Plutonium Isotopes

NASA Astrophysics Data System (ADS)

Steier, P.; Hrnecek, E.; Priller, A.; Quinto, F.; Srncik, M.; Wallner, A.; Wallner, G.; Winkler, S.

2013-01-01

VERA, the Vienna Environmental Research Accelerator, is especially equipped for the measurement of actinides, and performs a growing number of measurements on environmental samples. While AMS is not the optimum method for each particular plutonium isotope, the possibility to measure 239Pu, 240Pu, 241Pu, 242Pu and 244Pu on the same AMS sputter target is a great simplification. We have obtained a first result on the global fallout value of 244Pu/239Pu = (5.7 ± 1.0) × 10-5 based on soil samples from Salzburg prefecture, Austria. Furthermore, we suggest using the 242Pu/240Pu ratio as an estimate of the initial 241Pu/239Pu ratio, which allows dating of the time of irradiation based solely on Pu isotopes. We have checked the validity of this estimate using literature data, simulations, and environmental samples from soil from the Salzburg prefecture (Austria), from the shut down Garigliano Nuclear Power Plant (Sessa Aurunca, Italy) and from the Irish Sea near the Sellafield nuclear facility. The maximum deviation of the estimated dates from the expected ages is 6 years, while relative dating of material from the same source seems to be possible with a precision of less than 2 years. Additional information carried by the minor plutonium isotopes may allow further improvements of the precision of the method.

Caustic Precipitation of Plutonium Using Gadolinium as the Neutron Poison for Disposition to High Level Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bronikowski, M.G.

2002-06-24

Nuclear Materials Management Division (NMMD) has proposed that up to 100 kg of the plutonium (Pu) solutions stored in H-Canyon be precipitated with a nuclear poison and dispositioned to H-Area Tank Farm. The use of gadolinium (Gd) as the poison would greatly reduce the number of additional glass logs resulting from this disposition. This report summarizes the characteristics of the precipitation process and addresses criticality concerns in the Nuclear Criticality Safety Evaluation. No problems were found with the nature of the precipitate or the neutralization process.

Isotopic fractionation studies of uranium and plutonium using porous ion emitters as thermal ionization mass spectrometry sources

DOE PAGES

Baruzzini, Matthew L.; Hall, Howard L.; Spencer, Khalil J.; ...

2018-04-22

Investigations of the isotope fractionation behaviors of plutonium and uranium reference standards were conducted employing platinum and rhenium (Pt/Re) porous ion emitter (PIE) sources, a relatively new thermal ionization mass spectrometry (TIMS) ion source strategy. The suitability of commonly employed, empirically developed mass bias correction laws (i.e., the Linear, Power, and Russell's laws) for correcting such isotope ratio data was also determined. Corrected plutonium isotope ratio data, regardless of mass bias correction strategy, were statistically identical to that of the certificate, however, the process of isotope fractionation behavior of plutonium using the adopted experimental conditions was determined to be bestmore » described by the Power law. Finally, the fractionation behavior of uranium, using the analytical conditions described herein, is also most suitably modeled using the Power law, though Russell's and the Linear law for mass bias correction rendered results that were identical, within uncertainty, to the certificate value.« less

TRANSURANIC ELEMENT, COMPOSITION THEREOF, AND METHODS FOR PRODUCING SEPARATING AND PURIFYING SAME

DOEpatents

Wahl, A.C.

1961-09-19

A process of separating plutonium from fission products contained in an aqueous solution is described. Plutonium, in the tri- or tetravalent state, and the fission products are coprecipitated on lanthanum fluoride, lanthanum oxalate, cerous fluoride, cerous phosphate, ceric iodate, zirconyl phosphate, thorium iodate, or thorium fluoride. The precipitate is dissolved in acid, and the plutonium is oxidized to the hexavalent state. The fission products are selectively precipitated on a carrier of the above group but different from that used for the coprecipitation. The plutonium in the solution, after removal of the fission product precipitate, is reduced to at least the tetravalent state and precipitated on lanthanum fluoride, lanthanum phosphate, lanthanum oxalate, lanthanum hydroxide, cerous fluoride, cerous phosphate, cerous oxalate, cerous hydroxide, ceric iodate, zirconyl phosphate, zirconyl iodate, zirconium hydroxide, thorium fluoride, thorium oxalate, thorium iodate, thorium peroxide, uranium iodate, uranium oxalate, or uranium peroxide, again using a different carrier than that used for the precipitation of the fission products.

Isotopic fractionation studies of uranium and plutonium using porous ion emitters as thermal ionization mass spectrometry sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baruzzini, Matthew L.; Hall, Howard L.; Spencer, Khalil J.

Investigations of the isotope fractionation behaviors of plutonium and uranium reference standards were conducted employing platinum and rhenium (Pt/Re) porous ion emitter (PIE) sources, a relatively new thermal ionization mass spectrometry (TIMS) ion source strategy. The suitability of commonly employed, empirically developed mass bias correction laws (i.e., the Linear, Power, and Russell's laws) for correcting such isotope ratio data was also determined. Corrected plutonium isotope ratio data, regardless of mass bias correction strategy, were statistically identical to that of the certificate, however, the process of isotope fractionation behavior of plutonium using the adopted experimental conditions was determined to be bestmore » described by the Power law. Finally, the fractionation behavior of uranium, using the analytical conditions described herein, is also most suitably modeled using the Power law, though Russell's and the Linear law for mass bias correction rendered results that were identical, within uncertainty, to the certificate value.« less

An independent evaluation of plutonium body burdens in populations near Los Alamos Laboratory using human autopsy data.

PubMed

Gaffney, Shannon H; Donovan, Ellen P; Shonka, Joseph J; Le, Matthew H; Widner, Thomas E

2013-06-01

In the mid-1940s, the United States began producing atomic weapon components at the Los Alamos National Laboratory (LANL). In an attempt to better understand historical exposure to nearby residents, this study evaluates plutonium activity in human tissue relative to residential location and length of time at residence. Data on plutonium activity in the lung, vertebrae, and liver of nearby residents were obtained during autopsies as a part of the Los Alamos Tissue Program. Participant residential histories and the distance from each residence to the primary plutonium processing buildings at LANL were evaluated in the analysis. Summary statistics, including Student t-tests and simple regressions, were calculated. Because the biological half-life of plutonium can vary significantly by organ, data were analyzed separately by tissue type (lung, liver, vertebrae). The ratios of plutonium activity (vertebrae:liver; liver:lung) were also analyzed in order to evaluate the importance of timing of exposure. Tissue data were available for 236 participants who lived in a total of 809 locations, of which 677 were verified postal addresses. Residents of Los Alamos were found to have higher plutonium activities in the lung than non-residents. Further, those who moved to Los Alamos before 1955 had higher lung activities than those who moved there later. These trends were not observed with the liver, vertebrae, or vertebrae:liver and liver:lung ratio data, however, and should be interpreted with caution. Although there are many limitations to this study, including the amount of available data and the analytical methods used to analyze the tissue, the overall results indicate that residence (defined as the year that the individual moved to Los Alamos) may have had a strong correlation to plutonium activity in human tissue. This study is the first to present the results of Los Alamos Autopsy Program in relation to residential status and location in Los Alamos. Copyright © 2012 Elsevier GmbH. All rights reserved.

Preserving Plutonium-244 as a National Asset

DOE Office of Scientific and Technical Information (OSTI.GOV)

Patton, Bradley D; Alexander, Charles W; Benker, Dennis

Plutonium-244 (244 Pu) is an extremely rare and long-lived isotope of plutonium with a half-life of 80 million years. Measureable amounts of 244 Pu are found in neither reactor-grade nor weapons-grade plutonium. Production of this isotope requires a very high thermal flux to permit the two successive neutron captures that convert 242 Pu to 243 Pu to 244 Pu, particularly given the short (about 5 hour) half-life of 243 Pu. Such conditions simply do not exist in plutonium production processes. Therefore, 244 Pu is ideal for precise radiochemical analyses measuring plutonium material properties and isotopic concentrations in items containing plutonium.more » Isotope dilution mass spectrometry is about ten times more sensitive when using 244 Pu rather than 242 Pu for determining plutonium isotopic content. The isotope can also be irradiated in small quantities to produce superheavy elements. The majority of the existing global inventory of 244 Pu is contained in the outer housing of Mark-18A targets at the Savannah River Site (SRS). The total inventory is about 20 grams of 244 Pu in about 400 grams of plutonium distributed among the 65 targets. Currently, there are no specific plans to preserve these targets. Although the cost of separating and preserving this material would be considerable, it is trivial in comparison to new production costs. For all practical purposes, the material is irreplaceable, because new production would cost billions of dollars and require a series of irradiation and chemical separation cycles spanning up to 50 years. This paper will discuss a set of options for overcoming the significant challenges to preserve the 244 Pu as a National Asset: (1) the need to relocate the material from SRS in a timely manner, (2) the need to reduce the volume of material to the extent possible for storage, and (3) the need to establish an operational capability to enrich the 244 Pu in significant quantities. This paper suggests that if all the Mark-18A plutonium is separated, it would occupy a small volume and would be inexpensive to store while an enrichment capability is developed. Very small quantities could be enriched in existing mass separators to support critical needs.« less

Recovery of fissile materials from nuclear wastes

DOEpatents

Forsberg, Charles W.

1999-01-01

A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

Characterization of the Old Hydrofracture Facility (OHF) waste tanks located at ORNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keller, J.M.; Giaquinto, J.M.; Meeks, A.M.

1997-04-01

The Old Hydrofracture Facility (OHF) is located in Melton Valley within Waste Area Grouping (WAG) 5 and includes five underground storage tanks (T1, T2, T3, T4, and T9) ranging from 13,000 to 25,000 gal. capacity. During the period of 1996--97 there was a major effort to re-sample and characterize the contents of these inactive waste tanks. The characterization data summarized in this report was needed to address waste processing options, examine concerns dealing with the performance assessment (PA) data for the Waste Isolation Pilot Plant (WIPP), evaluate the waste characteristics with respect to the waste acceptance criteria (WAC) for WIPPmore » and Nevada Test Site (NTS), address criticality concerns, and to provide the data needed to meet DOT requirements for transporting the waste. This report discusses the analytical characterization data collected on both the supernatant and sludge samples taken from three different locations in each of the OHF tanks. The isotopic data presented in this report supports the position that fissile isotopes of uranium ({sup 233}U and {sup 235}U) do not satisfy the denature ratios required by the administrative controls stated in the ORNL LLLW waste acceptance criteria (WAC). The fissile isotope of plutonium ({sup 239}Pu and {sup 241}Pu) are diluted with thorium far above the WAC requirements. In general, the OHF sludge was found to be hazardous (RCRA) based on total metal content and the transuranic alpha activity was well above the 100 nCi/g limit for TRU waste. The characteristics of the OHF sludge relative to the WIPP WAC limits for fissile gram equivalent, plutonium equivalent activity, and thermal power from decay heat were estimated from the data in this report and found to be far below the upper boundary for any of the remote-handled transuranic waste (RH-TRU) requirements for disposal of the waste in WIPP.« less

PROCESS FOR SEGREGATING URANIUM FROM PLUTONIUM AND FISSION-PRODUCT CONTAMINATION

DOEpatents

Ellison, C.V.; Runion, T.C.

1961-06-27

An aqueous nitric acid solution containing uranium, plutonium, and fission product values is contacted with an organic extractant comprised of a trialkyl phosphate and an organic diluent. The relative amounts of trialkyl phosphate and uranium values are controlled to achieve a concentration of uranium values in the organic extractant of at least 0.35 moles uranium per mole of trialkyl phosphate, thereby preferentially extracting uranium values into the organic extractant.

Summary of Plutonium-238 Production Alternatives Analysis Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

James Werner; Wade E. Bickford; David B. Lord

The Team implemented a two-phase evaluation process. During the first phase, a wide variety of past and new candidate facilities and processing methods were assessed against the criteria established by DOE for this assessment. Any system or system element selected for consideration as an alternative within the project to reestablish domestic production of Pu-238 must meet the following minimum criteria: Any required source material must be readily available in the United States, without requiring the development of reprocessing technologies or investments in systems to separate material from identified sources. It must be cost, schedule, and risk competitive with existing baselinemore » technology. Any identified facilities required to support the concept must be available to the program for the entire project life cycle (notionally 35 years, unless the concept is so novel as to require a shorter duration). It must present a solution that can generate at least 1.5 Kg of Pu-238 oxide per year, for at least 35 years. It must present a low-risk, near-term solution to the National Aeronautics and Space Administration’s urgent mission need. DOE has implemented this requirement by eliminating from project consideration any alternative with key technologies at less than Technology Readiness Level 5. The Team evaluated the options meeting these criteria using a more detailed assessment of the reasonable facility variations and compared them to the preferred option, which consists of target irradiation at the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), target fabrication and chemical separations processing at the ORNL Radiochemical Engineering Development Center, and neptunium 237 storage at the Materials and Fuels Complex at INL. This preferred option is consistent with the Records of Decision from the earlier National Environmental Policy Act (NEPA) documentation« less

PLUTONIUM-CERIUM-COBALT AND PLUTONIUM-CERIUM-NICKEL ALLOYS

DOEpatents

Coffinberry, A.S.

1959-08-25

>New plutonium-base teroary alloys useful as liquid reactor fuels are described. The alloys consist of 10 to 20 atomic percent cobalt with the remainder plutonium and cerium in any desired proportion, with the plutonium not in excess of 88 atomic percent; or, of from 10 to 25 atomic percent nickel (or mixture of nickel and cobalt) with the remainder plutonium and cerium in any desired proportion, with the plutonium not in excess of 86 atomic percent. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are a lower melting point and a wide range of permissible plutonium dilution.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Armstrong, Christopher R.; Brant, Heather A.; Nuessle, Patterson R.

Owing to the rich history of heavy element production in the unique high flux reactors that operated at the Savannah River Site, USA (SRS) decades ago, trace quantities of plutonium with highly unique isotopic characteristics still persist today in the SRS terrestrial environment. Development of an effective sampling, processing, and analysis strategy enables detailed monitoring of the SRS environment, revealing plutonium isotopic compositions, e.g., 244Pu, that reflect the unique legacy of plutonium production at SRS. This work describes the first long-term investigation of anthropogenic 244Pu occurrence in the environment. Environmental samples, consisting of collected foot borne debris, were taken atmore » SRS over an eleven year period, from 2003 to 2014. Separation and purification of trace plutonium was carried out followed by three stage thermal ionization mass spectrometry (3STIMS) measurements for plutonium isotopic content and isotopic ratios. Furthermore, significant 244Pu was measured in all of the years sampled with the highest amount observed in 2003. The 244Pu content, in femtograms (fg = 10 –15 g) per gram, ranged from 0.31 fg/g to 44 fg/g in years 2006 and 2003 respectively. In all years, the 244Pu/ 239Pu atom ratios were significantly higher than global fallout, ranging from 0.003 to 0.698 in years 2014 and 2003 respectively.« less

The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

DTIC Science & Technology

1969-12-01

a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

Defense Horizons. Number 38, January 2004. Dirty Bombs: The Threat Revisited

DTIC Science & Technology

2004-01-01

plutonium-238 (238Pu), americium - 241 (241Am), and cali- fornium-252 (252Cf). Types of Damage Deterministic Injuries. Radiation is said to cause...megasources” such as Russian radioisotope thermal generators ( RTGs ) and Gamma-Kolos seed irradiators. By far the most likely route for terrorist...facility or a business or residential district, not just open space . More efficient RDDs relying on other means to disseminate the same amount of

Solution In-Line Alpha Counter (SILAC) Instruction Manual-Version 4.00

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven M. Alferink; Joel E. Farnham; Malcolm M. Fowler

2002-06-01

The Solution In-Line Alpha Counter (SILAC) provides near real-time alpha activity measurements of aqueous solutions in gloveboxes located in the Plutonium Facility (TA-55) at Los Alamos National Laboratory (LANL). The SILAC detector and its interface software were first developed by Joel Farnham at LANL [1]. This instruction manual describes the features of the SILAC interface software and contains the schematic and fabrication instructions for the detector.

Japan’s Nuclear Future: Policy Debate, Prospects, and U.S. Interests

DTIC Science & Technology

2008-05-09

raised in particular over the construction of an industrial- scale reprocessing facility in Japan,. Additionally, fast breeder reactors also produce more...Nuclear Fuel Cycle Engineering Laboratories. 10 A fast breeder reactor is a fast neutron reactor that produces more plutonium than it consumes, which can...Japan Nuclear Fuel Limited (JNFL) has built and is currently running active testing on a large - scale commercial reprocessing plant at Rokkasho-mura

Where Radiobiology Began in Russia: A Physician’s Perspective

DTIC Science & Technology

2010-09-01

radiation incidents at the MPA and conduct studies of long-term health effects of radiation exposure on that population and also the general public along...White Archipelago”) (Gubarev, 2004): “Visiting a number of production facilities that worked with plutonium and polonium - 210 , I was struck by the...were established for major food products. These standards, of course, were higher than the Union-wide public health standards. They did not begin

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA; Volume 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1994-01-01

Promoting the exchange of information related to implementation of the As Low as Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report was prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health. It contains the fifth in a series of bibliographies on dose reduction at DOE facilities. The BNL ALARA Center was originally established in 1983 under the sponsorship of the Nuclear Regulatory Commission to monitor dose-reduction research and ALARA activities at nuclear power plants. This effort was expanded in 1988 by the DOE`s Office of Environment,more » Safety and Health, to include DOE nuclear facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and accelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts.« less

PEROXIDE PROCESS FOR SEPARATION OF RADIOACTIVE MATERIALS

DOEpatents

Seaborg, G.T.; Perlman, I.

1958-09-16

reduced state, from hexavalent uranium. It consists in treating an aqueous solution containing such uranium and plutonium ions with sulfate ions in order to form a soluble uranium sulfate complex and then treating the solution with a soluble thorium compound and a soluble peroxide compound in order to ferm a thorium peroxide carrier precipitate which carries down with it the plutonium peroxide present. During this treatment the pH of the solution must be maintained between 2 and 3.

Flowsheet Analysis of U-Pu Co-Crystallization Process as a New Reprocessing System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shunji Homma; Jun-ichi Ishii; Jiro Koga

2006-07-01

A new fuel reprocessing system by U-Pu co-crystallization process is proposed and examined by flowsheet analysis. This reprocessing system is based on the fact that hexavalent plutonium in nitric acid solution is co-crystallized with uranyl nitrate, whereas it is not crystallized when uranyl nitrate does not exist in the solution. The system consists of five steps: dissolution of spent fuel, plutonium oxidation, U-Pu co-crystallization as a co-decontamination, re-dissolution of the crystals, and U re-crystallization as a U-Pu separation. The system requires a recycling of the mother liquor from the U-Pu co-crystallization step and the appropriate recycle ratio is determined bymore » flowsheet analysis such that the satisfactory decontamination is achieved. Further flowsheet study using four different compositions of LWR spent fuels demonstrates that the constant ratio of plutonium to uranium in mother liquor from the re-crystallization step is achieved for every composition by controlling the temperature. It is also demonstrated by comparing to the Purex process that the size of the plant based on the proposed system is significantly reduced. (authors)« less

The behaviour of tributyl phosphate in an organic diluent

NASA Astrophysics Data System (ADS)

Leay, Laura; Tucker, Kate; Del Regno, Annalaura; Schroeder, Sven L. M.; Sharrad, Clint A.; Masters, Andrew J.

2014-09-01

Tributyl phosphate (TBP) is used as a complexing agent in the Plutonium Uranium Extraction (PUREX) liquid-liquid phase extraction process for recovering uranium and plutonium from spent nuclear reactor fuel. Here, we address the molecular and microstructure of the organic phases involved in the extraction process, using molecular dynamics to show that when TBP is mixed with a paraffinic diluent, the TBP self-assembles into a bi-continuous phase. The underlying self-association of TBP is driven by intermolecular interaction between its polar groups, resulting in butyl moieties radiating out into the organic solvent. Simulation predicts a TBP diffusion constant that is anomalously low compared to what might normally be expected for its size; experimental nuclear magnetic resonance (NMR) studies also indicate an extremely low diffusion constant, consistent with a molecular aggregation model. Simulation of TBP at an oil/water interface shows the formation of a bilayer system at low TBP concentrations. At higher concentrations, a bulk bi-continuous structure is observed linking to this surface bilayer. We suggest that this structure may be intimately connected with the surprisingly rapid kinetics of the interfacial mass transport of uranium and plutonium from the aqueous to the organic phase in the PUREX process.

Stability of zinc stearate under alpha irradiation in the manufacturing process of SFR nuclear fuels

NASA Astrophysics Data System (ADS)

Gracia, J.; Vermeulen, J.; Baux, D.; Sauvage, T.; Venault, L.; Audubert, F.; Colin, X.

2018-03-01

The manufacture of new fuels for sodium-cooled fast reactors (SFRs) will involve powders derived from recycling existing fuels in order to keep on producing electricity while saving natural resources and reducing the amount of waste produced by spent MOX fuels. Using recycled plutonium in this way will significantly increase the amount of 238Pu, a high energy alpha emitter, in the powders. The process of shaping powders by pressing requires the use of a solid lubricant, zinc stearate, to produce pellets with no defects compliant with the standards. The purpose of this study is to determine the impact of alpha radiolysis on this additive and its lubrication properties. Experiments were conducted on samples in contact with PuO2, as well as under external helium ion beam irradiation, in order to define the kinetics of radiolytic gas generation. The yield results relating to the formation of these gases (G0) show that the alpha radiation of plutonium can be simulated using external helium ion beam irradiation. The isotopic composition of plutonium has little impact on the yield. However, an increased yield was globally observed with increasing the mean linear energy transfer (LET). A radiolytic degradation process is proposed.

31. VIEW OF A WORKER HOLDING A PLUTONIUM 'BUTTON.' PLUTONIUM, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

31. VIEW OF A WORKER HOLDING A PLUTONIUM 'BUTTON.' PLUTONIUM, A MAN-MADE SUBSTANCE, WAS RARE. SCRAPS RESULTING FROM PRODUCTION AND PLUTONIUM RECOVERED FROM RETIRED NUCLEAR WEAPONS WERE REPROCESSED INTO VALUABLE PURE-PLUTONIUM METAL (9/19/73). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

10. AERIAL VIEW LOOKING NORTHWEST AT THE 400AREA COMPLEX. THIS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. AERIAL VIEW LOOKING NORTHWEST AT THE 400-AREA COMPLEX. THIS AREA OF THE PLANT MANUFACTURED NON-PLUTONIUM WEAPONS COMPONENTS FROM BERYLLIUM, DEPLETED URANIUM, AND STAINLESS STEEL. THE 400 - AREA ALSO INCLUDED A FACILITY FOR THE MODIFICATION OF SAFE SECURE TRANSPORT VEHICLES FOR SPECIAL NUCLEAR MATERIALS BEING SHIPPED TO AND FROM THE SITE. BUILDING 444, IN THE UPPER RIGHT EDGE OF THE PHOTOGRAPH, WAS THE ORIGINAL PLANT A. THE LARGE BUILDING IN THE TOP OF THE PHOTOGRAPH IS BUILDING 460, BUILT AS A STATE-OF-THE-ART STAINLESS STEEL MANUFACTURING FACILITY (6/27/95). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hoover, Andrew Scott; Bennett, D. A.; Croce, Mark Philip

In 2005 the LANL/NIST team used a single high-resolution microcalorimeter detector to measure the gamma-ray spectrum of a plutonium sample. After more than a decade of research and development on this topic, both the technology and our general understanding of its capabilities have advanced greatly, such that a progress review is now timely. We examine the scenario of a large-scale reprocessing plant and conclude that current non-destructive analysis (NDA) methods are inadequate to safeguard such a facility to the desired levels, leading to undesirable dependence on massspectrometry (MS) destructive analysis (DA). The development of microcalorimeter detectors is intended to closemore » the performance gap between NDA and DA methods to address the needs of nuclear facilities.« less

The Inglorious Death of Jumbo

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meade, Roger Allen

In the summer of 1944, J. Robert Oppenheimer and Los Alamos faced a crisis. An isotopic impurity in Plutonium rendered the metal unusable in a gun-assembled atomic bomb (i.e., Little Boy). Making this situation worse was a shortage of Uranium. The combination of these two problems threatened the entire wartime project. The answer to this dilemma, in part, was to develop a novel assembly method for Plutonium using the supersonic shock waves created by several tons of high explosives to compress a ball of Plutonium into a supercritical state. Since this method, implosion, was not much more than a theoreticalmore » construct, the Trinity test was devised to proof test the process. Given the speculative nature of implosion, Trinity was a gamble of sorts. If the test failed (i.e., little or no nuclear yield), the blast of the high explosives would scatter the scarce and expensive Plutonium over the surrounding desert. Since the probability of failure remained high into the early summer of 1945, some method of containing a failed nuclear explosion was needed. Jumbo was the answer.« less

Effect of temperature and radiation damage on the local atomic structure of elemental plutonium and related compounds

DOE PAGES

Booth, Corwin H.; Olive, Daniel Thomas

2016-10-26

This focused review provides an overview and a framework for understanding local structure in metallic plutonium (especially the metastable fcc δ-phase alloyed with Ga) as it relates to self-irradiation damage. Of particular concern is the challenge of understanding self-irradiation damage in plutonium-bearing materials where theoretical challenges of the unique involvement of the 5f electrons in bonding limit the efficacy of molecular dynamics simulations and experimental challenges of working with radioactive material have limited the ability to confirm the results of such simulations and to further push the field forward. The main concentration is on extended X-ray absorption fine-structure measurements ofmore » -phase Pu, but the scope is broadened to include certain studies on plutonium intermetallics and oxides insofar as they inform the physics of damage and healing processes in elemental Pu. Here, the studies reviewed here provide insight into lattice distortions and their production, damage annealing and defect migration, and the importance of understanding and controlling sample morphology when interpreting such experiments.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Veirs, Douglas Kirk; Stroud, Mary Ann; Berg, John M.

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 overmore » 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.« less

VITRIFICATION SYSTEM FOR THE TREATMENT OF PLUTONIUM-BEARING WASTE AT LOS ALAMOS NATIONAL LABORATORY

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. NAKAOKA; G. VEAZEY; ET AL

2001-05-01

A glove box vitrification system is being fabricated to process aqueous evaporator bottom waste generated at the Plutonium Facility (TA-55) at Los Alamos National Laboratory (LANL). The system will be the first within the U.S. Department of Energy Complex to routinely convert Pu{sup 239}-bearing transuranic (TRU) waste to a glass matrix for eventual disposal at the Waste Isolation Pilot Plant (WIPP). Currently at LANL, this waste is solidified in Portland cement. Radionuclide loading in the cementation process is restricted by potential radiolytic degradation (expressed as a wattage limit), which has been imposed to prevent the accumulation of flammable concentrations ofmore » H{sub 2} within waste packages. Waste matrixes with a higher water content (e.g., cement) are assigned a lower permissible wattage limit to compensate for their potential higher generation of H{sub 2}. This significantly increases the number of waste packages that must be prepared and shipped, thus driving up the costs of waste handling and disposal. The glove box vitrification system that is under construction will address this limitation. Because the resultant glass matrix produced by the vitrification process is non-hydrogenous, no H{sub 2} can be radiolytically evolved, and drums could be loaded to the maximum allowable limit of 40 watts. In effect, the glass waste form shifts the limiting constraint for loading disposal drums from wattage to the criticality limit of 200 fissile gram equivalents, thus significantly reducing the number of drums generated from this waste stream. It is anticipated that the number of drums generated from treatment of evaporator bottoms will be reduced by a factor of 4 annually when the vitrification system is operational. The system is currently undergoing non-radioactive operability testing, and will be fully operational in the year 2003.« less

Method of uranium reclamation from aqueous systems by reactive ion exchange. [US DOE patent application; anion exchange resin of copolymerized divinyl-benzene and styrene having quarternary ammonium groups and bicarbonate ligands

DOEpatents

Maya, L.

1981-11-05

A reactive ion exchange method for separation and recovery of values of uranium, neptunium, plutonium, or americium from substantially neutral aqueous systems of said metals comprises contacting said system with an effective amount of a basic anion exchange resin of copolymerized divinyl-benzene and styrene having quarternary ammonium groups and bicarbonate ligands to achieve nearly 100% sorption of said actinyl ion onto said resin and an aqueous system practically free of said actinyl ions. The method is operational over an extensive range of concentrations from about 10/sup -6/ M to 1.0 M actinyl ion and a pH range of about 4 to 7. The method has particulr application to treatment of waste streams from Purex-type nuclear fuel reprocessing facilities and hydrometallurgical processes involving U, Np, P, or Am.

SEPARATION OF NEPTUNIUM FROM PLUTONIUM BY CHLORINATION AND SUBLIMATION

DOEpatents

Fried, S.M.

1958-11-18

A process is described for separating neptunium from plutonium. The method consists in chlorinating a mixture of the oxides of Np and Pu by contacting the mixture with carbon tetrachloride at about 500 icient laborato C. ln this manner the Np is converted to the tetrachlorlde and the Pu converted to the trichloride. Since NpCl/sub 4/ is more latile than PuCl/sub 3/, the separation ls effected by vaporing sad subsequently condenslng the NpCl/sub 4/.

17. VIEW OF AIR LOCK ENTRY DOOR. BANKS OF AIR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

17. VIEW OF AIR LOCK ENTRY DOOR. BANKS OF AIR FILTERS ARE VISIBLE TO THE SIDES OF THE DOORS. THE BUILDING WAS DIVIDED INTO ZONES BY AIRLOCK DOORS AND AIR FILTERS. AIR PRESSURE DIFFERENTIALS WERE MAINTAINED IN THE ZONES, SUCH THAT AIRFLOW WAS PROGRESSIVELY TOWARD AREAS WITH THE HIGHEST POTENTIAL FOR CONTAMINATION. (9/24/91) - Rocky Flats Plant, Plutonium Manufacturing Facility, North-central section of Plant, just south of Building 776/777, Golden, Jefferson County, CO

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, M.H.

1981-01-09

Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

DOEpatents

Lloyd, Milton H.

1983-01-01

Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

Savannah River Site nuclear materials management plan FY 2017-2031

DOE Office of Scientific and Technical Information (OSTI.GOV)

Magoulas, V.

The purpose of the Nuclear Materials Management Plan (herein referred to as “this Plan”) is to integrate and document the activities required to disposition the legacy and/or surplus Enriched Uranium (EU) and Plutonium (Pu) and other nuclear materials already stored or anticipated to be received by facilities at the Department of Energy (DOE) Savannah River Site (SRS) as well as the activities to support the DOE Tritium mission. It establishes a planning basis for EU and Pu processing operations in Environmental Management Operations (EMO) facilities through the end of their program missions and for the tritium through the National Nuclearmore » Security Administration (NNSA) Defense Programs (DP) facilities. Its development is a joint effort among the Department of Energy - Savannah River (DOE-SR), DOE – Environmental Management (EM), NNSA Office of Material Management and Minimization (M3), NNSA Savannah River Field Office (SRFO), and the Management and Operations (M&O) contractor, Savannah River Nuclear Solutions, LLC (SRNS). Life-cycle program planning for Nuclear Materials Stabilization and Disposition and the Tritium Enterprise may use this Plan as a basis for the development of the nuclear materials disposition scope and schedule. This Plan assumes full funding to accomplish the required project and operations activities. It is recognized that some aspects of this Plan are pre decisional with regard to National Environmental Policy Act (NEPA); in such cases new NEPA actions will be required.« less

Solvent extraction system for plutonium colloids and other oxide nano-particles

DOEpatents

Soderholm, Lynda; Wilson, Richard E; Chiarizia, Renato; Skanthakumar, Suntharalingam

2014-06-03

The invention provides a method for extracting plutonium from spent nuclear fuel, the method comprising supplying plutonium in a first aqueous phase; contacting the plutonium aqueous phase with a mixture of a dielectric and a moiety having a first acidity so as to allow the plutonium to substantially extract into the mixture; and contacting the extracted plutonium with second a aqueous phase, wherein the second aqueous phase has a second acidity higher than the first acidity, so as to allow the extracted plutonium to extract into the second aqueous phase. The invented method facilitates isolation of plutonium polymer without the formation of crud or unwanted emulsions.

SEPARATION OF PLUTONIUM

DOEpatents

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.

Effect of Americium-241 Content on Plutonium Radiation Source Terms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rainisch, R.

1998-12-28

The management of excess plutonium by the US Department of Energy includes a number of storage and disposition alternatives. Savannah River Site (SRS) is supporting DOE with plutonium disposition efforts, including the immobilization of certain plutonium materials in a borosilicate glass matrix. Surplus plutonium inventories slated for vitrification include materials with elevated levels of Americium-241. The Am-241 content of plutonium materials generally reflects in-growth of the isotope due to decay of plutonium and is age-dependent. However, select plutonium inventories have Am-241 levels considerably above the age-based levels. Elevated levels of americium significantly impact radiation source terms of plutonium materials andmore » will make handling of the materials more difficult. Plutonium materials are normally handled in shielded glove boxes, and the work entails both extremity and whole body exposures. This paper reports results of an SRS analysis of plutonium materials source terms vs. the Americium-241 content of the materials. Data with respect to dependence and magnitude of source terms on/vs. Am-241 levels are presented and discussed. The investigation encompasses both vitrified and un-vitrified plutonium oxide (PuO2) batches.« less

Criticality safety strategy and analysis summary for the fuel cycle facility electrorefiner at Argonne National Laboratory West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mariani, R.D.; Benedict, R.W.; Lell, R.M.

1996-05-01

As part of the termination activities of Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory (ANL) West, the spent metallic fuel from EBR-II will be treated in the fuel cycle facility (FCF). A key component of the spent-fuel treatment process in the FCF is the electrorefiner (ER) in which the actinide metals are separated from the active metal fission products and the reactive bond sodium. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt, and refined uranium or uranium/plutonium products are deposited at cathodes. The criticality safety strategy and analysis for the ANLmore » West FCF ER is summarized. The FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. To show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOEs) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOEs, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that will verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality.« less

OPERATING THE WAND AND HERCULES PROTOTYPE SYSTEMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. GRUETZMACHER; ET AL

2001-01-01

Two prototype systems for low-density Green is Clean (GIC) waste at Los Alamos National Laboratory (LANL) have been in operation for three years at the Solid Waste Operation's (SWOs) non-destructive assay (NDA) building. The Waste Assay for Nonradioactive Disposal (WAND) and the High Efficiency Radiation Counters for Ultimate Low Emission Sensitivity (HERCULES) are used to verify the waste generator's acceptable knowledge (AK) that low-density waste is nonradioactive. GIC waste includes all non-regulated waste generated in radiological controlled areas (RCAs) that has been actively segregated as ''clean'' (i.e., nonradioactive) through the use of waste generator AK. GIC waste that is verifiedmore » clean can be disposed of at the Los Alamos County Landfill. It is estimated that 50-90% of the low-density room trash from RCAs at LANL might be free of contamination. To date, with pilot programs at five facilities at LANL, 3000 cubic feet of GIC waste has been verified clean by these two prototype systems. Both the WAND and HERCULES systems are highly sensitive measurement systems optimized to detect very small quantities of common LANL radionuclides. Both of the systems use a set of phoswich scintillation detectors in close proximity to the waste, which have the capability of detecting plutonium-239 concentrations below 3 pCi per gram of low density waste. Both systems detect low-energy x-rays and a broad range of gamma rays (10-2000 keV), while the WAND system also detects high energy beta particles (>100 keV). The WAND system consists of a bank of six shielded detectors which screen low density shredded waste or stacked sheets of paper moving under the detectors in a twelve inch swath on a conveyor belt. The WAND system was developed and tested at the Los Alamos Plutonium Facility in conjunction with instrument system designers from the Los Alamos Safeguards Science and Technology group. The HERCULES system consists of a bank of three shielded detectors which screen low-density waste in two cubic foot cardboard boxes or in bags sitting on a turntable. Waste that does not pass the verification process can be examined within the facility to determine the type and quantity of the contamination and its origin within a waste container. The paper discusses lessons learned that have helped generators improve their AK segregation.« less

Methods and Models of the Hanford Internal Dosimetry Program, PNNL-MA-860

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbaugh, Eugene H.; Bihl, Donald E.; Maclellan, Jay A.

2009-09-30

The Hanford Internal Dosimetry Program (HIDP) provides internal dosimetry support services for operations at the Hanford Site. The HIDP is staffed and managed by the Radiation and Health Technology group, within the Pacific Northwest National Laboratory (PNNL). Operations supported by the HIDP include research and development, the decontamination and decommissioning of facilities formerly used to produce and purify plutonium, and waste management activities. Radioelements of particular interest are plutonium, uranium, americium, tritium, and the fission and activation product radionuclides 137Cs, 90Sr, and 60Co. This manual describes the technical basis for the design of the routine bioassay monitoring program and formore » assessment of internal dose. The purposes of the manual are as follows: • Provide assurance that the HIDP derives from a sound technical base. • Promote the consistency and continuity of routine program activities. • Provide a historical record. • Serve as a technical reference for radiation protection personnel. • Aid in identifying and planning for future needs.« less

Actinide Oxidation State and O/M Ratio in Hypostoichiometric Uranium-Plutonium-Americium U0.750Pu0.246Am0.004O2-x Mixed Oxides.

PubMed

Vauchy, Romain; Belin, Renaud C; Robisson, Anne-Charlotte; Lebreton, Florent; Aufore, Laurence; Scheinost, Andreas C; Martin, Philippe M

2016-03-07

Innovative americium-bearing uranium-plutonium mixed oxides U1-yPuyO2-x are envisioned as nuclear fuel for sodium-cooled fast neutron reactors (SFRs). The oxygen-to-metal (O/M) ratio, directly related to the oxidation state of cations, affects many of the fuel properties. Thus, a thorough knowledge of its variation with the sintering conditions is essential. The aim of this work is to follow the oxidation state of uranium, plutonium, and americium, and so the O/M ratio, in U0.750Pu0.246Am0.004O2-x samples sintered for 4 h at 2023 K in various Ar + 5% H2 + z vpm H2O (z = ∼ 15, ∼ 90, and ∼ 200) gas mixtures. The O/M ratios were determined by gravimetry, XAS, and XRD and evidenced a partial oxidation of the samples at room temperature. Finally, by comparing XANES and EXAFS results to that of a previous study, we demonstrate that the presence of uranium does not influence the interactions between americium and plutonium and that the differences in the O/M ratio between the investigated conditions is controlled by the reduction of plutonium. We also discuss the role of the homogeneity of cation distribution, as determined by EPMA, on the mechanisms involved in the reduction process.

Characterization and source term assessments of radioactive particles from Marshall Islands using non-destructive analytical techniques

NASA Astrophysics Data System (ADS)

Jernström, J.; Eriksson, M.; Simon, R.; Tamborini, G.; Bildstein, O.; Marquez, R. Carlos; Kehl, S. R.; Hamilton, T. F.; Ranebo, Y.; Betti, M.

2006-08-01

Six plutonium-containing particles stemming from Runit Island soil (Marshall Islands) were characterized by non-destructive analytical and microanalytical methods. Composition and elemental distribution in the particles were studied with synchrotron radiation based micro X-ray fluorescence spectrometry. Scanning electron microscope equipped with energy dispersive X-ray detector and with wavelength dispersive system as well as a secondary ion mass spectrometer were used to examine particle surfaces. Based on the elemental composition the particles were divided into two groups: particles with pure Pu matrix, and particles where the plutonium is included in Si/O-rich matrix being more heterogenously distributed. All of the particles were identified as nuclear fuel fragments of exploded weapon components. As containing plutonium with low 240Pu/ 239Pu atomic ratio, less than 0.065, which corresponds to weapons-grade plutonium or a detonation with low fission yield, the particles were identified to originate from the safety test and low-yield tests conducted in the history of Runit Island. The Si/O-rich particles contained traces of 137Cs ( 239 + 240 Pu/ 137Cs activity ratio higher than 2500), which indicated that a minor fission process occurred during the explosion. The average 241Am/ 239Pu atomic ratio in the six particles was 3.7 × 10 - 3 ± 0.2 × 10 - 3 (February 2006), which indicated that plutonium in the different particles had similar age.

Method for dissolving plutonium dioxide

DOEpatents

Tallent, Othar K.

1978-01-01

The fluoride-catalyzed, non-oxidative dissolution of plutonium dioxide in HNO.sub.3 is significantly enhanced in rate by oxidizing dissolved plutonium ions. It is believed that the oxidation of dissolved plutonium releases fluoride ions from a soluble plutonium-fluoride complex for further catalytic action.

The Tiger Team Process in the Rebaselining of the Plutonium Finishing Plant (PFP)

DOE Office of Scientific and Technical Information (OSTI.GOV)

BAILEY, R.W.

2000-02-01

This paper will describe the integrated, teaming approach and planning process utilized by the Tiger Team in the development of the IPMP. This paper will also serve to document the benefits derived from this implementation process.

Concentration and purification of plutonium or thorium

DOEpatents

Hayden, John A.; Plock, Carl E.

1976-01-01

In this invention a first solution obtained from such as a plutonium/thorium purification process or the like, containing plutonium (Pu) and/or thorium (Th) in such as a low nitric acid (HNO.sub.3) concentration may have the Pu and/or Th separated and concentrated by passing an electrical current from a first solution having disposed therein an anode to a second solution having disposed therein a cathode and separated from the first solution by a cation permeable membrane, the Pu or Th cation permeating the cation membrane and forming an anionic complex within the second solution, and electrical current passage affecting the complex formed to permeate an anion membrane separating the second solution from an adjoining third solution containing disposed therein an anode, thereby effecting separation and concentration of the Pu and/or Th in the third solution.

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

Plutonium-related work and cause-specific mortality at the United States Department of Energy Hanford Site.

PubMed

Wing, Steve; Richardson, David; Wolf, Susanne; Mihlan, Gary

2004-02-01

Health effects of working with plutonium remain unclear. Plutonium workers at the United States Department of Energy (US-DOE) Hanford Site in Washington State, USA were evaluated for increased risks of cancer and non-cancer mortality. Periods of employment in jobs with routine or non-routine potential for plutonium exposure were identified for 26,389 workers hired between 1944 and 1978. Life table regression was used to examine associations of length of employment in plutonium jobs with confirmed plutonium deposition and with cause specific mortality through 1994. Incidence of confirmed internal plutonium deposition in all plutonium workers was 15.4 times greater than in other Hanford jobs. Plutonium workers had low death rates compared to other workers, particularly for cancer causes. Mortality for several causes was positively associated with length of employment in routine plutonium jobs, especially for employment at older ages. At ages 50 and above, death rates for non-external causes of death, all cancers, cancers of tissues where plutonium deposits, and lung cancer, increased 2.0 +/- 1.1%, 2.6 +/- 2.0%, 4.9 +/- 3.3%, and 7.1 +/- 3.4% (+/-SE) per year of employment in routine plutonium jobs, respectively. Workers employed in jobs with routine potential for plutonium exposure have low mortality rates compared to other Hanford workers even with adjustment for demographic, socioeconomic, and employment factors. This may be due, in part, to medical screening. Associations between duration of employment in jobs with routine potential for plutonium exposure and mortality may indicate occupational exposure effects. Copyright 2004 Wiley-Liss, Inc.

An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swinhoe, Martyn T; Tobin, Stephen J; Fensin, Mike L

2009-01-01

There are a variety of reasons for quantifying plutonium (Pu) in spent fuel. Below, five motivations are listed: (1) To verify the Pu content of spent fuel without depending on unverified information from the facility, as requested by the IAEA ('independent verification'). New spent fuel measurement techniques have the potential to allow the IAEA to recover continuity of knowledge and to better detect diversion. (2) To assure regulators that all of the nuclear material of interest leaving a nuclear facility actually arrives at another nuclear facility ('shipper/receiver'). Given the large stockpile of nuclear fuel at reactor sites around the world,more » it is clear that in the coming decades, spent fuel will need to be moved to either reprocessing facilities or storage sites. Safeguarding this transportation is of significant interest. (3) To quantify the Pu in spent fuel that is not considered 'self-protecting.' Fuel is considered self-protecting by some regulatory bodies when the dose that the fuel emits is above a given level. If the fuel is not self-protecting, then the Pu content of the fuel needs to be determined and the Pu mass recorded in the facility's accounting system. This subject area is of particular interest to facilities that have research-reactor spent fuel or old light-water reactor (LWR) fuel. It is also of interest to regulators considering changing the level at which fuel is considered self-protecting. (4) To determine the input accountability value at an electrochemical processing facility. It is not expected that an electrochemical reprocessing facility will have an input accountability tank, as is typical in an aqueous reprocessing facility. As such, one possible means of determining the input accountability value is to measure the Pu content in the spent fuel that arrives at the facility. (5) To fully understand the composition of the fuel in order to efficiently and safely pack spent fuel into a long-term repository. The NDA of spent fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument will be quantified for the full assembly library for measurements in three different media: air, water and borated water. The 12 NDA techniques being researched are the following: Delayed Gamma, Delayed Neutrons, Differential Die-Away, Lead Slowing Down Spectrometer, Neutron Multiplicity, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Passive Neutron Albedo Reactivity, Self-integration Neutron Resonance Densitometry, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection.« less

Collaborative Russian-US work in nuclear material protection, control and accounting at the Institute of Physics and Power Engineering. 2: Extension to additional facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuzin, V.V.; Pshakin, G.M.; Belov, A.P.

1996-12-31

During 1995, collaborative Russian-US nuclear material protection, control, and accounting (MPC and A) tasks at the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russia focused on improving the protection of nuclear materials at the BFS Fast Critical Facility. BFS has tens of thousands of fuel disks containing highly enriched uranium and weapons-grade plutonium that are used to simulate the core configurations of experimental reactors in two critical assemblies. Completed tasks culminated in demonstrations of newly implemented equipment (Russian and US) and methods that enhanced the MPC and A at BFS through computerized accounting, nondestructive inventory verification measurements, personnelmore » identification and access control, physical inventory taking, physical protection, and video surveillance. The collaborative work with US Department of Energy national laboratories is now being extended. In 1996 additional tasks to improve MPC and A have been implemented at BFS, the Technological Laboratory for Fuel Fabrication (TLFF) the Central Storage Facility (CSF), and for the entire site. The TLFF reclads BFS uranium metal fuel disks (process operations and transfers of fissile material). The CSF contains many different types of nuclear material. MPC and A at these additional facilities will be integrated with that at BFS as a prototype site-wide approach. Additional site-wide tasks encompass communications and tamper-indicating devices. Finally, new storage alternatives are being implemented that will consolidate the more attractive nuclear materials in a better-protected nuclear island. The work this year represents not just the addition of new facilities and the site-wide approach, but the systematization of the MPC and A elements that are being implemented as a first step and the more comprehensive ones planned.« less

Russian-US collaboration on implementation of the active well coincidence counter (AWCC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mozhajev, V.; Pshakin, G.; Stewart, J.

The feasibility of using a standard AWCC at the Obninsk IPPE has been demonstrated through active measurements of single UO{sub 2} (36% enriched) disks and through passive measurements of plutonium metal disks used for simulating reactor cores. The role of the measurements is to verify passport values assigned to the disks by the facility, and thereby facilitate the mass accountability procedures developed for the very large inventory of fuel disks at the facility. The AWCC is a very flexible instrument for verification measurements of the large variety of nuclear material items at the Obninsk IPPE and other Russian facilities. Futuremore » work at the IPPE will include calibration and verification measurements for other materials, both in individual disks and in multi-disk storage tubes; it will also include training in the use of the AWCC.« less

Johnston Atoll Plutonium Contaminated Soil Cleanup Project. 5th quarterly report, 1 August 94 to 31 October 1994. Technical report, 1 August-31 October 1994

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doane, R.W.; Grant, R.H.

1996-09-01

Thermo NUtech is the prime contractor for the Defense Nuclear Agency (DNA), responsible for the operation and maintenance of the Johnston Atoll plutonium Contaminated Soil Cleanup Project. During this production period, the Scope of Work included movement of soil to and from the plant, processing contaminated soil through the Segmented Gate System (SGS) and Soil Washing System, packaging of waste soil for shipment, identification and implementation of process improvements, data collection and validation, and compliance with all applicable regulations governing environmental safety and health. The SGS utilizes arrays of sensitive radiation detectors coupled with sophisticated computer software to segregate contaminatedmore » soil from a moving feed supply on conveyor belts. Contaminated soil is diverted to a `hot path` for plutonium particles greater than 5000 Becquerels or to a supplemental soil washing process designed to remove dispersed low leve%l contamination from a soil faction consisting of very small particles. Low to intermediate levels of contamination are removed from the soil to meet DNA`s criteria for unrestricted use of less than 500 Becquerels per kilogram of soil, with no hot particles. The low level concentrate is expected to be packaged for shipment to an approved defense waste disposal site.« less

20. VIEW OF THE INCINERATOR. DURING ROUTINE BUILDING OPERATIONS IN ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

20. VIEW OF THE INCINERATOR. DURING ROUTINE BUILDING OPERATIONS IN DECEMBER 1988, A HEAT PLUME WAS GENERATED THAT WAS REGISTERED ON FILM BY A PASSING AIRCRAFT. OFFICIALS WITH THE UNITED STATES ENVIRONMENTAL PROTECTION AGENCY (USEPA) BELIEVED THAT ILLEGAL OPERATIONS WERE BEING CONDUCTED. THE USEPA USED THIS OPPORTUNITY TO CONVINCE AUTHORITIES TO ISSUE A WARRANT TO ENTER THE ROCKY FLATS PLANT AND INVESTIGATE THE ALLEGATION. (4/98) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

Structural Characterization of the 216-Z-9 Crib Prior to Decontamination and Demolition Using Robot Crawler and High Resolution Photography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hopkins, A.; Sutter, C.; Klos, D.B.

2008-07-01

The Department of Energy, Richland Operations Office is preparing to conduct a Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) removal action for the decontamination and demolition of the above-grade mining structures and equipment at the 216-Z-9 Crib. An investigation of the condition of the mining complex was initiated to determine constraints necessary for safely conducting the removal of the buildings. While crib headspace chemical analysis and nondestructive analysis of the interior of the buildings was completed to address radiological contamination concerns, the primary concern regarding the removal of the above-grade structures located on the crib cover involves determining themore » loading capacity and structural integrity of the crib cover slab. Additional concerns included headspace gases and radionuclide contamination. Until the structural analysis was completed, loading limits on the crib cover had been restricted. Photographic documentation revealed the loss of protective tiles and acid resistant coating from the underside of the cover raising a question of concrete stability. The investigation relied heavily on the use of high resolution photography with high intensity lighting for photographic documentation of the underside of the crib cover, followed by structural analysis of the documentation by a team of qualified engineers. Deployment of a robot crawler with attached camera and positioning of a fixed camera were integral to this structural characterization effort. Results of the photographic documentation were of sufficient quality to allow for bounding decisions to be made regarding the loading of the crib cover while performing the demolition of the mining structures (glovebox, excavator, bucket) and the associated buildings. The 216-Z-9 Crib, also known as the 216 Z-9 Recuplex CAW (CA column waste) Waste Disposal Cavern, the Z-9 Trench and the Z-9 Crib was constructed as an engineered trench with an open area beneath a concrete slab. The crib is located near the Plutonium Finishing Plant (PFP) facility, at the Hanford Nuclear Reservation in Eastern Washington State. The crib was used as a disposal site for effluent chemical and radiological wastes from the recovery of uranium and plutonium through extraction or RECUPLEX process, a method that recovered uranium and plutonium from liquid and solid wastes and scraps from other PFP processes. During its operating life, from 1955 through 1962, the Z-9 Crib received liquid wastes totaling approximately four million liters, or one million gallons. Analyses of the crib soil in seven locations to a depth of up to two meters (six feet) beneath the crib floor indicated that the plutonium content of the crib soil ranged from 50 to 150 kg (the highest concentration measured was 34.5 g/L of soil). While performing the structural evaluation of the crib cover, additional characterization information was obtained on the radiological and chemical conditions of the crib and structures. (authors)« less

SEPARATION OF PLUTONIUM FROM LANTHANUM BY CHELATION-EXTRACTION

DOEpatents

James, R.A.; Thompson, S.G.

1958-12-01

Plutonium can be separated from a mixture of plutonlum and lanthanum in which the lanthanum to plutonium molal ratio ls at least five by adding the ammonium salt of N-nitrosoarylhydroxylamine to an aqueous solution having a pH between about 3 and 0.2 and containing the plutonium in a valence state of at least +3, to form a plutonium chelate compound of N-nitrosoarylhydroxylamine. The plutonium chelate compound may be recovered from the solution by extracting with an immiscible organic solvent such as chloroform.

Air transport of plutonium metal: content expansion initiative for the plutonium air transportable (PAT01) packaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caviness, Michael L; Mann, Paul T; Yoshimura, Richard H

2010-01-01

The National Nuclear Security Administration (NNSA) has submitted an application to the Nuclear Regulatory Commission (NRC) for the air shipment of plutonium metal within the Plutonium Air Transportable (PAT-1) packaging. The PAT-1 packaging is currently authorized for the air transport of plutonium oxide in solid form only. The INMM presentation will provide a limited overview of the scope of the plutonium metal initiative and provide a status of the NNSA application to the NRC.

The calculation of annual limits of intake for plutonium-239 in man using a bone model which allows for plutonium burial and recycling.

PubMed

Priest, N D; Hunt, B W

1979-05-01

Values of the annual limit of intake (ALI) for plutonium-239 in man have been calculated using committed dose equivalent limits as recommended by ICRP in Publication 26. The calculations were made using a multicompartment bone model which allows for plutonium burial and recycling in the skeleton. In one skeletal compartment, the growing surfaces of cortical bone, it is assumed that plutonium deposits are retained and are not subject to resorption or recycling. In the trabecular bone compartment plutonium is taken to be resorbed with either subsequent redeposition onto bone surfaces or retention in the bone marrow. ALIs for plutonium-239 have been calculated assuming a range of rates of bone accretion (0-32 micron yr-1), different amounts of plutonium retained in the marrow (0-60%) and a 20%, 45% or 70% deposition of plutonium in the skeleton from the blood. The calculations made using this bone model suggest that 750 Bq (20 nCi) is an appropriate ALI for the inhalation of class W and class Y plutonium compounds and that 830 kBq and 5 MBq (23 muCi and 136 muCi) are the appropriate ALIs for the ingestion of soluble and insoluble forms of plutonium respectively.

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Thermal properties of simulated Hanford waste glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodriguez, Carmen P.; Chun, Jaehun; Crum, Jarrod V.

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will vitrify the mixed hazardous wastes generated from 45 years of plutonium production. The molten glasses will be poured into stainless steel containers or canisters and subsequently quenched for storage and disposal. Such highly energy-consuming processes require precise thermal properties of materials for appropriate facility design and operations. Key thermal properties (heat capacity, thermal diffusivity, and thermal conductivity) of representative high-level and low-activity waste glasses were studied as functions of temperature in the range of 200 to 800°C (relevant to the cooling process), implementing simultaneous differential scanning calorimetry-thermal gravimetry (DSC-TGA), Xe-flashmore » diffusivity, pycnometry, and dilatometry. The study showed that simultaneous DSC-TGA would be a reliable method to obtain heat capacity of various glasses at the temperature of interest. Accurate thermal properties from this study were shown to provide a more realistic guideline for capacity and time constraint of heat removal process, in comparison to the design basis conservative engineering estimates. The estimates, though useful for design in the absence measured physical properties, can now be supplanted and the measured thermal properties can be used in design verification activities.« less

Radionuclide Basics: Plutonium

EPA Pesticide Factsheets

Plutonium (chemical symbol Pu) is a radioactive metal. Plutonium is considered a man-made element. Plutonium-239 is used to make nuclear weapons. Pu-239 and Pu-240 are byproducts of nuclear reactor operations and nuclear bomb explosions.

PROCESS OF PRODUCING Cm$sup 244$ AND Cm$sup 24$$sup 5$

DOEpatents

Manning, W.M.; Studier, M.H.; Diamond, H.; Fields, P.R.

1958-11-01

A process is presented for producing Cm and Cm/sup 245/. The first step of the process consists in subjecting Pu/sup 2339/ to a high neutron flux and subsequently dissolving the irradiated material in HCl. The plutonium is then oxidized to at least the tetravalent state and the solution is contacted with an anion exchange resin, causing the plutonium values to be absorbed while the fission products and transplutonium elements remain in the effluent solution. The effluent solution is then contacted with a cation exchange resin causing the transplutonium, values to be absorbed while the fission products remain in solution. The cation exchange resin is then contacted with an aqueous citrate solution and tbe transplutonium elements are thereby differentially eluted in order of decreasing atomic weight, allowing collection of the desired fractions.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1989-09-07

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.

A relativistic density functional study of the role of 5f electrons in atomic and molecular adsorptions on actinide surfaces

NASA Astrophysics Data System (ADS)

Huda, Muhammad Nurul

Atomic and molecular adsorptions of oxygen and hydrogen on actinide surfaces have been studied within the generalized gradient approximations to density functional theory (GGA-DFT). The primary goal of this work is to understand the details of the adsorption processes, such as chemisorption sites, energies, adsorption configurations and activation energies for dissociation of molecules; and the signature role of the plutonium 5f electrons. The localization of the 5f electrons remains one of central questions in actinides and one objective here is to understand the extent to which localizations plays a role in adsorption on actinide surfaces. We also investigated the magnetism of the plutonium surfaces, given the fact that magnetism in bulk plutonium is a highly controversial issue, and the surface magnetism of it is not a well explored territory. Both the non-spin-polarized and spin-polarized calculations have been performed to arrive at our conclusions. We have studied both the atomic and molecular hydrogen and oxygen adsorptions on plutonium (100) and (111) surfaces. We have also investigated the oxygen molecule adsorptions on uranium (100) surface. Comparing the adsorption on uranium and plutonium (100) surfaces, we have seen that O2 chemisorption energy for the most favorable adsorption site on uranium surface has higher chemisorption energy, 9.492 eV, than the corresponding plutonium site, 8.787 eV. Also degree of localization of 5f electrons is less for uranium surface. In almost all of the cases, the most favorable adsorption sites are found where the coordination numbers are higher. For example, we found center sites are the most favorable sites for atomic adsorptions. In general oxygen reacts more strongly with plutonium surface than hydrogen. We found that atomic oxygen adsorption energy on (100) surface is 3.613 eV more than that of the hydrogen adsorptions, considering only the most favorable site. This is also true for molecular adsorptions, as the oxygen molecules on both (100) and (111) plutonium surfaces dissociate almost spontaneously, whereas hydrogen needs some activation energy to dissociate. From spin-polarized calculations we found both (100) and (111) surfaces have the layer by layer alternating spin-magnetic behavior. In general adsorption of H2 and O2 do not change this behavior.

Dynamic Radioactive Source for Evaluating and Demonstrating Time-dependent Performance of Continuous Air Monitors.

PubMed

McLean, Thomas D; Moore, Murray E; Justus, Alan L; Hudston, Jonathan A; Barbé, Benoît

2016-11-01

Evaluation of continuous air monitors in the presence of a plutonium aerosol is time intensive, expensive, and requires a specialized facility. The Radiation Protection Services Group at Los Alamos National Laboratory has designed a Dynamic Radioactive Source, intended to replace plutonium aerosol challenge testing. The Dynamic Radioactive Source is small enough to be inserted into the sampler filter chamber of a typical continuous air monitor. Time-dependent radioactivity is introduced from electroplated sources for real-time testing of a continuous air monitor where a mechanical wristwatch motor rotates a mask above an alpha-emitting electroplated disk source. The mask is attached to the watch's minute hand, and as it rotates, more of the underlying source is revealed. The measured alpha activity increases with time, simulating the arrival of airborne radioactive particulates at the air sampler inlet. The Dynamic Radioactive Source allows the temporal behavior of puff and chronic release conditions to be mimicked without the need for radioactive aerosols. The new system is configurable to different continuous air monitor designs and provides an in-house testing capability (benchtop compatible). It is a repeatable and reusable system and does not contaminate the tested air monitor. Test benefits include direct user control, realistic (plutonium) aerosol spectra, and iterative development of continuous air monitor alarm algorithms. Data obtained using the Dynamic Radioactive Source has been used to elucidate alarm algorithms and to compare the response time of two commercial continuous air monitors.

Dynamic Radioactive Source for Evaluating and Demonstrating Time-dependent Performance of Continuous Air Monitors

DOE PAGES

McLean, Thomas D.; Moore, Murray E.; Justus, Alan L.; ...

2016-01-01

Evaluation of continuous air monitors in the presence of a plutonium aerosol is time intensive, expensive, and requires a specialized facility. The Radiation Protection Services Group at Los Alamos National Laboratory has designed a Dynamic Radioactive Source, intended to replace plutonium aerosol challenge testing. Furthermore, the Dynamic Radioactive Source is small enough to be inserted into the sampler filter chamber of a typical continuous air monitor. Time-dependent radioactivity is introduced from electroplated sources for real-time testing of a continuous air monitor where a mechanical wristwatch motor rotates a mask above an alpha-emitting electroplated disk source. The mask is attached tomore » the watch’s minute hand, and as it rotates, more of the underlying source is revealed. The alpha activity we measured increases with time, simulating the arrival of airborne radioactive particulates at the air sampler inlet. The Dynamic Radioactive Source allows the temporal behavior of puff and chronic release conditions to be mimicked without the need for radioactive aerosols. The new system is configurable to different continuous air monitor designs and provides an in-house testing capability (benchtop compatible). It is a repeatable and reusable system and does not contaminate the tested air monitor. Test benefits include direct user control, realistic (plutonium) aerosol spectra, and iterative development of continuous air monitor alarm algorithms. We also used data obtained using the Dynamic Radioactive Source to elucidate alarm algorithms and to compare the response time of two commercial continuous air monitors.« less

The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, I.

1987-01-01

The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from idealmore » solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.« less

Method of separating thorium from plutonium

DOEpatents

Clifton, David G.; Blum, Thomas W.

1984-01-01

A method of chemically separating plutonium from thorium. Plutonium and thorium to be separated are dissolved in an aqueous feed solution, preferably as the nitrate salts. The feed solution is acidified and sodium nitrite is added to the solution to adjust the valence of the plutonium to the +4 state. A chloride salt, preferably sodium chloride, is then added to the solution to induce formation of an anionic plutonium chloride complex. The anionic plutonium chloride complex and the thorium in solution are then separated by ion exchange on a strong base anion exchange column.

Method of separating thorium from plutonium

DOEpatents

Clifton, D.G.; Blum, T.W.

A method of chemically separating plutonium from thorium is claimed. Plutonium and thorium to be separated are dissolved in an aqueous feed solution, preferably as the nitrate salts. The feed solution is acidified and sodium nitrite is added to the solution to adjust the valence of the plutonium to the +4 state. A chloride salt, preferably sodium chloride, is then added to the solution to induce formation of an anionic plutonium chloride complex. The anionic plutonium chloride complex and the thorium in solution are then separated by ion exchange on a strong base anion exchange column.

Method of separating thorium from plutonium

DOEpatents

Clifton, D.G.; Blum, T.W.

1984-07-10

A method is described for chemically separating plutonium from thorium. Plutonium and thorium to be separated are dissolved in an aqueous feed solution, preferably as the nitrate salts. The feed solution is acidified and sodium nitrite is added to the solution to adjust the valence of the plutonium to the +4 state. A chloride salt, preferably sodium chloride, is then added to the solution to induce formation of an anionic plutonium chloride complex. The anionic plutonium chloride complex and the thorium in solution are then separated by ion exchange on a strong base anion exchange column.

Plutonium isotopic signatures in soils and their variation (2011-2014) in sediment transiting a coastal river in the Fukushima Prefecture, Japan.

PubMed

Jaegler, Hugo; Pointurier, Fabien; Onda, Yuichi; Hubert, Amélie; Laceby, J Patrick; Cirella, Maëva; Evrard, Olivier

2018-05-04

The Fukushima Daiichi Nuclear Power Plant (FDNPP) accident resulted in a significant release of radionuclides that were deposited on soils in Northeastern Japan. Plutonium was detected at trace levels in soils and sediments collected around the FDNPP. However, little is known regarding the spatial-temporal variation of plutonium in sediment transiting rivers in the region. In this study, plutonium isotopic compositions were first measured in soils (n = 5) in order to investigate the initial plutonium deposition. Then, plutonium isotopic compositions were measured on flood sediment deposits (n = 12) collected after major typhoon events in 2011, 2013 and 2014. After a thorough radiochemical purification, isotopic ratios ( 240 Pu/ 239 Pu, 241 Pu/ 239 Pu and 242 Pu/ 239 Pu) were measured with a Multi-Collector Inductively Coupled Mass Spectrometer (MC ICP-MS), providing discrimination between plutonium derived from global fallout, from atmospheric nuclear weapon tests, and plutonium derived from the FDNPP accident. Results demonstrate that soils with the most Fukushima-derived plutonium were in the main radiocaesium plume and that there was a variable mixture of plutonium sources in the flood sediment samples. Plutonium concentrations and isotopic ratios generally decreased between 2011 and 2014, reflecting the progressive erosion and transport of contaminated sediment in this coastal river during flood events. Exceptions to this general trend were attributed to the occurrence of decontamination works or the remobilisation of contaminated material during typhoons. The different plutonium concentrations and isotopic ratios obtained on three aliquots of a single sample suggest that the Fukushima-derived plutonium was likely borne by discrete plutonium-containing particles. In the future, these particles should be isolated and further characterized in order to better understand the fate of this long-lived radionuclide in the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

Locating trace plutonium in contaminated soil using micro-XRF imaging

DOE PAGES

Worley, Christopher G.; Spencer, Khalil J.; Boukhalfa, Hakim; ...

2014-06-01

Micro-X-ray fluorescence (MXRF) was used to locate minute quantities of plutonium in contaminated soil. Because the specimen had previously been prepared for analysis by scanning electron microscopy, it was coated with gold to eliminate electron beam charging. However, this significantly hindered efforts to detect plutonium by MXRF. The gold L peak series present in all spectra increased background counts. Plutonium signal attenuation by the gold coating and severe peak overlap from potassium in the soil prevented detection of trace plutonium using the Pu Mα peak. However, the 14.3 keV Pu Lα peak sensitivity was not optimal due to poor transmissionmore » efficiency through the source polycapillary optic, and the instrument silicon drift detector sensitivity quickly declines for peaks with energies above ~10 keV. Instrumental parameters were optimized (eg. using appropriate source filters) in order to detect plutonium. An X-ray beam aperture was initially used to image a majority of the specimen with low spatial resolution. A small region that appeared to contain plutonium was then imaged at high spatial resolution using a polycapillary optic. Small areas containing plutonium were observed on a soil particle, and iron was co-located with the plutonium. Zinc and titanium also appeared to be correlated with the plutonium, and these elemental correlations provided useful plutonium chemical state information that helped to better understand its environmental transport properties.« less

Aqueous Electrochemical Mechanisms in Actinide Residue Processing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morris, David E.; Burns, Carol J.; Smith, Wayne H.

2000-12-31

Plutonium and uranium residues (e.g., incinerator ash, combustibles, and sand/slag/crucibles) resulting from the purification and processing of nuclear materials constitute an enormous volume of ''lean'' processing waste and represent a significant fraction of the U. S. Department of Energy's (DOE) legacy waste from fifty years of nuclear weapons production activities. Much of this material is presently in storage at sites throughout the DOE weapons production complex (most notably Rocky Flats, Savannah River and Hanford) awaiting further processing and/or final disposition. The chemical and physical stability of much of this material has been called into question recently by the Defense Nuclearmore » Facility Safety Board (DNFSB) and resulted in the issuance of a mandate by the DNFSB to undertake a program to stabilize these materials [1]. The ultimate disposition for much of these materials is anticipated to be geologic repositories such as the proposed Waste Isolation Pilot Plant in New Mexico. However, in light of the mandate to stabilize existing residues and the probable concomitant increase in the volume of material to be disposed as a result of stabilization (e.g., from repackaging at lower residue densities), the projected storage volume for these wastes within anticipated geologic repositories will likely be exceeded simply to handle existing wastes. Additional processing of some of these residue waste streams to reduce radionuclide activity levels, matrix volume, or both is a potentially important strategy to achieve both stabilization and volume reduction so that the anticipated geologic repositories will provide adequate storage volume. In general, the plutonium and uranium that remains in solid residue materials exists in a very stable chemical form (e.g., as binary oxides), and the options available to remove the actinides are limited. However, there have been some demonstrated successes in this vain using aqueous phase electrochemical methods such as the Catalyzed Electrochemical Plutonium Oxide Dissolution (CEPOD) process pioneered by workers at Pacific Northwest National Laboratory in the mid-1970s [2]. The basis for most of these mediated electrochemical oxidation/reduction (MEO/R) processes is the generation of a dissolved electrochemical catalyst, such as Ag2+, which is capable of oxidizing or reducing solid-phase actinide species or actinide sorbates via 7 heterogeneous electron transfer to oxidation states that have significantly greater solubilities (e.g., PuO2(s) to PuO2 2+ (dissolved)). The solubilized actinide can then be recovered by ion exchange or other mechanisms. These aqueous electrochemical methods for residue treatment have been considered in many of the ''trade studies'' to evaluate options for stabilization of the various categories of residue materials. While some concerns generally arise (e.g., large secondary waste volumes could results since the process stream normally goes th rough anion exchange or precipitation steps to remove the actinide), the real utility and versatility of these methods should not be overlooked. They are low temperature, ambient pressure processes that operate in a non-corrosive environment. In principle, they can be designed to be highly selective for the actinides (i.e., no substrate degradation occurs), they can be utilized for many categories of residue materials with little or no modification in hardware or operating conditions, and they can conceivably be engineered to minimize secondary waste stream volume. However, some fundamental questions remain concerning the mechanisms through which these processes act, and how the processes might be optimized to maximize efficiency while minimizing secondary waste. In addition, given the success achieved to date on the limited set of residues, further research is merited to extend the range of applicability of these electrochemical methods to other residue and waste streams. The principal goal of the work described here is to develop a fundamental understanding of the heterogeneous electron transfer thermodynamics and kinetics that lie at the heart of the MEO/R processes for actinide solids and actinide species entrained in or surface-bound to residue substrates. This has been accomplished as described in detail below through spectroscopic characterization of actinide-bearing substrates and electrochemical investigations of electron transfer reactions between uranium- and plutonium- (or surrogates) bearing solids (dispersed actinide solid phases and actinides sorbed to inorganic and organic colloids) and polarizable electrode materials. In general, the actinide solids or substrate-supported species were chosen to represent relevant residue materials (e.g., incinerator ash, sand/slag/crucible, and combustibles).« less

Feasibility study of a magnetic fusion production reactor

NASA Astrophysics Data System (ADS)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.

Development of Novel Method for Rapid Extract of Radionuclides from Solution Using Polymer Ligand Film

NASA Astrophysics Data System (ADS)

Rim, Jung H.

Accurate and fast determination of the activity of radionuclides in a sample is critical for nuclear forensics and emergency response. Radioanalytical techniques are well established for radionuclides measurement, however, they are slow and labor intensive, requiring extensive radiochemical separations and purification prior to analysis. With these limitations of current methods, there is great interest for a new technique to rapidly process samples. This dissertation describes a new analyte extraction medium called Polymer Ligand Film (PLF) developed to rapidly extract radionuclides. Polymer Ligand Film is a polymer medium with ligands incorporated in its matrix that selectively and rapidly extract analytes from a solution. The main focus of the new technique is to shorten and simplify the procedure necessary to chemically isolate radionuclides for determination by alpha spectrometry or beta counting. Five different ligands were tested for plutonium extraction: bis(2-ethylhexyl) methanediphosphonic acid (H2DEH[MDP]), di(2-ethyl hexyl) phosphoric acid (HDEHP), trialkyl methylammonium chloride (Aliquat-336), 4,4'(5')-di-t-butylcyclohexano 18-crown-6 (DtBuCH18C6), and 2-ethylhexyl 2-ethylhexylphosphonic acid (HEH[EHP]). The ligands that were effective for plutonium extraction further studied for uranium extraction. The plutonium recovery by PLFs has shown dependency on nitric acid concentration and ligand to total mass ratio. H2DEH[MDP] PLFs performed best with 1:10 and 1:20 ratio PLFs. 50.44% and 47.61% of plutonium were extracted on the surface of PLFs with 1M nitric acid for 1:10 and 1:20 PLF, respectively. HDEHP PLF provided the best combination of alpha spectroscopy resolution and plutonium recovery with 1:5 PLF when used with 0.1M nitric acid. The overall analyte recovery was lower than electrodeposited samples, which typically has recovery above 80%. However, PLF is designed to be a rapid field deployable screening technique and consistency is more important than recovery. PLFs were also tested using blind quality control samples and the activities were accurately measured. It is important to point out that PLFs were consistently susceptible to analytes penetrating and depositing below the surface. The internal radiation within the body of PLF is mostly contained and did not cause excessive self-attenuation and peak broadening in alpha spectroscopy. The analyte penetration issue was beneficial in the destructive analysis. H2DEH[MDP] PLF was tested with environmental samples to fully understand the capabilities and limitations of the PLF in relevant environments. The extraction system was very effective in extracting plutonium from environmental water collected from Mortandad Canyon at Los Alamos National Laboratory with minimal sample processing. Soil samples were tougher to process than the water samples. Analytes were first leached from the soil matrixes using nitric acid before processing with PLF. This approach had a limitation in extracting plutonium using PLF. The soil samples from Mortandad Canyon, which are about 1% iron by weight, were effectively processed with the PLF system. Even with certain limitations of the PLF extraction system, this technique was able to considerably decrease the sample analysis time. The entire environmental sample was analyzed within one to two days. The decrease in time can be attributed to the fact that PLF is replacing column chromatography and electrodeposition with a single step for preparing alpha spectrometry samples. The two-step process of column chromatography and electrodeposition takes a couple days to a week to complete depending on the sample. The decrease in time and the simplified procedure make this technique a unique solution for application to nuclear forensics and emergency response. A large number of samples can be quickly analyzed and selective samples can be further analyzed with more sensitive techniques based on the initial data. The deployment of a PLF system as a screening method will greatly reduce a total analysis time required to gain meaningful isotopic data for the nuclear forensics application. (Abstract shortened by UMI.)

Area G Perimeter Surface-Soil and Single-Stage Water Sampling: Environmental Surveillance for Fiscal Years 1996 and 1997, Group ESH-19

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marquis Childs; Ron Conrad

1998-10-01

Area Gin Technical Area 54, has been the principal facility at Los Alamos National Laboratory for the storage and disposal of low-level, solid mixed, and transuranic radioactive waste since 1957. Soil samples were analyzed for tritium, isotopic plutonium, americium-241, and cesium-137. Thirteen metals-silver, arsenic, barium, beryllium, cadmium, chromium, mercury, nickel, lead, antimony, selenium, thallium and zinc-were analyzed on filtered-sediment fractions of the single-stage samples using standard analytical chemistry techniques. During the two years of sampling discussed in this report elevated levels of tritium (as high as 716,000 pCi/L) in soil were found for sampling sites adjacent to the tritium burialmore » shafts located on the south- central perimeter of Area G. Additionally, tritium concentrations in soil as high as 38,300 pCi/L were detected adjacent to the TRU pads in the northeast comer of Area G. Plutonium-238 activities in FY96 soils ranged from 0.001-2.866 pCi/g, with an average concentration of 0.336& 0.734 pCdg. Pu-238 activities in FY97 soils ranged from 0.002-4.890 pCi/g, with an average concentration of 0.437 & 0.928 pCdg. Pu-239 activities in FY96 soils ranged from 0.009 to 1.62 pCdg, with an average of 0.177- 0.297 pCdg. Pu-239 activities in FY97 soils ranged from 0.005 to 1.71 pCi/g, with an average of 0.290- 0.415 pCi/g. The locations of elevated plutonium readings were consistent with the history of plutonium disposal at Area G. The two areas of elevated Am-241 activity reflected the elevated activities found for plutonium, the average values for Am-241 on soils were 0.6-2.07 pCi/g, and 0.10-0.14 pCi/g respectively for samples collected in FY96 and FY97. CS-137 activities in soils had average values of 0.33 pCi/g, and 0.28 pCi/g respectively for samples collected in FY96 and 97. There was no perimeter area where soil concentrations of CS-137 were significantly elevated.« less

Source and long-term behavior of transuranic aerosols in the WIPP environment.

PubMed

Thakur, P; Lemons, B G

2016-10-01

Source and long-term behavior transuranic aerosols ((239+240)Pu, (238)Pu, and (241)Am) in the ambient air samples collected at and near the Waste Isolation Pilot Plant (WIPP) deep geologic repository site were investigated using historical data from an independent monitoring program conducted by the Carlsbad Environmental Monitoring and Research Center and an oversight monitoring program conducted by the management and operating contractor for WIPP at and near the facility. An analysis of historical data indicates frequent detections of (239+240)Pu and (241)Am, whereas (238)Pu is detected infrequently. Peaks in (239+240)Pu and (241)Am concentrations in ambient air generally occur from March to June timeframe, which is when strong and gusty winds in the area frequently give rise to blowing dust. Long-term measurements of plutonium isotopes (1985-2015) in the WIPP environment suggest that the resuspension of previously contaminated soils is likely the primary source of plutonium in the ambient air samples from WIPP and its vicinity. There is no evidence that WIPP is a source of environmental contamination that can be considered significant by any health-based standard.

Lymph node clearance of plutonium from subcutaneous wounds in beagles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dagle, G.E.

1973-08-01

The lymph node clearance of /sup 239/Pu O/sub 2/ administered as insoluble particles from subcutaneous implants was studied in adult beagles to simulate accidental contamination of hand wounds. External scintillation data were collected from the popliteal lymph nodes of each dog after 9.2 to 39.4 mu Ci of plutonium oxide was subcutaneously implanted into the left or right hind paws. The left hind paw was armputated 4 weeks after implantation to prevent continued deposition of plutonium oxide particles in the left popliteal lymph node. Groups of 3 dogs were sacrificed 4, 8, 16, and 32 weeks after plutonium implantation formore » histopathologic, electron microscopic, and radiochemical analysis of regional lymph nodes. An additional group of dogs received treatment with the chelating agent diethyenetriaminepentaacetic acid (DTPA). Plutonium rapidly accumulated in the popliteal lymph nodes after subcutaneous injection into the hind paw, and 1 to 10% of the implant dose was present in the popliteal lymph nodes at the time of necropsy. Histopathologic changes in the popliteal lymph nodes with plutonium particles were characterized primarily by reticular cell hyperplasia, increased numbers of macrophages, necrosis, and fibroplasia. Eventually, the plutonium particles became sequestered by scar tissue that often replaced the entire architecture of the lymph node. Light microscopic autoradiographs of the popliteal lymph nodes showed a time-related increase in number of alpha tracks per plutonium source. Electron microscopy showed that the plutonium particles were aggregated in phagolysosomes of macrophages. There was slight clearance of plutonium from the popliteal lymph nodes of dogs monitored for 32 weeks. The clearance of plutonium particles from the popliteal lymph nodes was associated with necrosis of macrophages. The external iliac lymph nodes contained fewer plutonium particles than the popliteal lymph nodes and histopathologic changes were less severe. The superficial inguinal lymph nodes of one dog contained appreciable amounts of plutonium. Treatment with diethylenetriaminepentaacetic acid (DTPA) did not have a measurable effect on the clearance of plutonium from the popliteal lymph nodes. (60 references) (auth)« less

CONCENTRATION OF Pu USING OXALATE TYPE CARRIER

DOEpatents

Ritter, D.M.; Black, R.P.S.

1960-04-19

A method is given for dissolving and reprecipitating an oxalate carrier precipitate in a carrier precipitation process for separating and recovering plutonium from an aqueous solution. Uranous oxalate, together with plutonium being carried thereby, is dissolved in an aqueous alkaline solution. Suitable alkaline reagents are the carbonates and oxulates of the alkali metals and ammonium. An oxidizing agent selected from hydroxylamine and hydrogen peroxide is then added to the alkaline solution, thereby oxidizing uranium to the hexavalent state. The resulting solution is then acidified and a source of uranous ions provided in the acidified solution, thereby forming a second plutoniumcarrying uranous oxalate precipitate.

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

PubMed

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.

WEST VALLEY DEMONSTRATION PROJECT ANNUAL SITE ENVIRONMENTAL REPORT CALENDAR YEAR 2002

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2003-09-12

This annual environmental monitoring report for the West Valley Demonstration Project (WVDP or Project) is published to inform those with interest about environmental conditions at the WVDP. In accordance with U.S. Department of Energy (DOE) Order 231.1, Environment, Safety, and Health Reporting, the report summarizes calendar year (CY) 2002 environmental monitoring data so as to describe the performance of the WVDP's environmental management system, confirm compliance with standards and regulations, and highlight important programs. In 2002, the West Valley Demonstration Project, the site of a DOE environmental cleanup activity operated by West Valley Nuclear Services Co. (WVNSCO), was in themore » final stages of stabilizing high-level radioactive waste (HLW) that remained at the site after commercial nuclear fuel reprocessing had been discontinued in the early 1970s. The Project is located in western New York State, about 30 miles south of Buffalo, within the New York State-owned Western New York Nuclear Service Center (WNYNSC). The WVDP is being conducted in cooperation with the New York State Energy Research and Development Authority (NYSERDA). Ongoing work activities at the WVDP during 2002 included: (1) completing HLW solidification and melter shutdown; (2) shipping low-level radioactive waste off-site for disposal; (3) constructing a facility where large high-activity components can be safely packaged for disposal; (4) packaging and removing spent materials from the vitrification facility; (5) preparing environmental impact statements for future activities; (6) removing as much of the waste left behind in waste tanks 8D-1 and 8D-2 as was reasonably possible; (7) removing storage racks, canisters, and debris from the fuel receiving and storage pool, decontaminating pool walls, and beginning shipment of debris for disposal; (8) ongoing decontamination in the general purpose cell and the process mechanical cell (also referred to as the head end cells); (9) planning for cleanup of waste in the plutonium purification cell (south) and extraction cell number 2 in the main plant; (10) ongoing characterization of facilities such as the waste tank farm and process cells; (11) monitoring the environment and managing contaminated areas within the Project facility premises; and (12) flushing and rinsing HLW solidification facilities.« less

Tabulated Neutron Emission Rates for Plutonium Oxide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shores, Erik Frederick

This work tabulates neutron emission rates for 80 plutonium oxide samples as reported in the literature. Plutonium-­238 and plutonium-­239 oxides are included and such emission rates are useful for scaling tallies from Monte Carlo simulations and estimating dose rates for health physics applications.

Method for the recovery of actinide elements from nuclear reactor waste

DOEpatents

Horwitz, E. Philip; Delphin, Walter H.; Mason, George W.

1979-01-01

A process for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of dihexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate solution of tetramethylammonium hydrogen oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values. Uranium values remain in the extractant and may be recovered with a phosphoric acid strip. The neptunium and plutonium values are recovered from the oxalate by adding sufficient nitric acid to destroy the complexing ability of the oxalate, forming a second feed, and contacting the second feed with a second extractant of tricaprylmethylammonium nitrate in an inert diluent whereby the neptunium and plutonium values are selectively extracted. The values are recovered from the extractant with formic acid.

Sorption/Desorption Interactions of Plutonium with Montmorillonite

NASA Astrophysics Data System (ADS)

Begg, J.; Zavarin, M.; Zhao, P.; Kersting, A. B.

2012-12-01

Plutonium (Pu) release to the environment through nuclear weapon development and the nuclear fuel cycle is an unfortunate legacy of the nuclear age. In part due to public health concerns over the risk of Pu contamination of drinking water, predicting the behavior of Pu in both surface and sub-surface water is a topic of continued interest. Typically it was assumed that Pu mobility in groundwater would be severely restricted, as laboratory adsorption studies commonly show that naturally occurring minerals can effectively remove plutonium from solution. However, evidence for the transport of Pu over significant distances at field sites highlights a relative lack of understanding of the fundamental processes controlling plutonium behavior in natural systems. At several field locations, enhanced mobility is due to Pu association with colloidal particles that serve to increase the transport of sorbed contaminants (Kersting et al., 1999; Santschi et al., 2002, Novikov et al., 2006). The ability for mineral colloids to transport Pu is in part controlled by its oxidation state and the rate of plutonium adsorption to, and desorption from, the mineral surface. Previously we have investigated the adsorption affinity of Pu for montmorillonite colloids, finding affinities to be similar over a wide range of Pu concentrations. In the present study we examine the stability of adsorbed Pu on the mineral surface. Pu(IV) at an initial concentration of 10-10 M was pre-equilibrated with montmorillonite in a background electrolyte at pH values of 4, 6 and 8. Following equilibration, aliquots of the suspensions were placed in a flow cell and Pu-free background electrolyte at the relevant pH was passed through the system. Flow rates were varied in order to investigate the kinetics of desorption and hence gain a mechanistic understanding of the desorption process. The flow cell experiments demonstrate that desorption of Pu from the montmorillonite surface cannot be modeled as a simple first order process. Furthermore, a pH dependence was observed, with less desorbed at pH 4 compared to pH 8. We suggest the pH dependence is likely controlled by reoxidation of Pu(IV) to Pu(V) and aqueous speciation. We will present models used to describe desorption behavior and discuss the implications for Pu transport. References: Kersting, A.B.; Efurd, D.W.; Finnegan, D.L.; Rokop, D.J.; Smith, D.K.; Thompson J.L. (1999) Migration of plutonium in groundwater at the Nevada Test Site, Nature, 397, 56-59. Novikov A.P.; Kalmykov, S.N.; Utsunomiya, S.; Ewing, R.C.; Horreard, F.; Merkulov, A.; Clark, S.B.; Tkachev, V.V.; Myasoedov, B.F. (2006) Colloid transport of plutonium in the far-field of the Mayak Production Association, Russia, Science, 314, 638-641. Santschi, P.H.; Roberts, K.; Guo, L. (2002) The organic nature of colloidal actinides transported in surface water environments. Environ. Sci. Technol., 36, 3711-3719. This work was funded by U. S. DOE Office of Biological & Environmental Sciences, Subsurface Biogeochemistry Research Program, and performed under the auspices of the U. S. Department of Energy by Lawrence Livermore National Security, LLC under Contract DE-AC52-07NA27344. LLNL-ABS-570161

1. VIEW, LOOKING NORTH, NORTH OF PORTAL 1, JUST INSIDE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

1. VIEW, LOOKING NORTH, NORTH OF PORTAL 1, JUST INSIDE THE PROTECTED AREA. ON THE LEFT SIDE OF THE PHOTOGRAPH IS BUILDING 709, THE COOLING TOWER FOR BUILDING 707, AND BEHIND BUILDING 709 IS BUILDING 707, THE NEWEST OF THE PLUTONIUM FABRICATION BUILDINGS. IN THE RIGHT FOREGROUND IS BUILDING 763, A BREEZEWAY FOR PEDESTRIANS. IN THE FAR LEFT OF THE PHOTOGRAPH ARE THE T750 TRAILERS AND BUILDING 750, THE PRODUCTION SUPPORT ENGINEERING FACILITY. - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

Studies in support of an SNM cutoff agreement: The PUREX exercise

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stanbro, W.D.; Libby, R.; Segal, J.

1995-07-01

On September 23, 1993, President Clinton, in a speech before the United Nations General Assembly, called for an international agreement banning the production of plutonium and highly enriched uranium for nuclear explosive purposes. A major element of any verification regime for such an agreement would probably involve inspections of reprocessing plants in Nuclear Nonproliferation Treaty weapons states. Many of these are large facilities built in the 1950s with no thought that they would be subject to international inspection. To learn about some of the problems that might be involved in the inspection of such large, old facilities, the Department ofmore » Energy, Office of Arms Control and Nonproliferation, sponsored a mock inspection exercise at the PUREX plant on the Hanford Site. This exercise examined a series of alternatives for inspections of the PUREX as a model for this type of facility at other locations. A series of conclusions were developed that can be used to guide the development of verification regimes for a cutoff agreement at reprocessing facilities.« less

Los Alamos County Fire Department LAFD: TA-55 PF-4 Facility Familiarization Tour, OJT 55260

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rutherford, Victor Stephen

Los Alamos National Laboratory (LANL) will conduct familiarization tours for Los Alamos County Fire Department (LAFD) personnel at the Plutonium Facility (PF-4) at Technical Area (TA)-55. These familiarization tours are official LANL business; the purpose of these tours is to orient the firefighters to the facility so that they can respond efficiently and quickly to a variety of emergency situations. This orientation includes the ingress and egress of the area and buildings, layout and organization of the facility, evacuation procedures and assembly points, and areas of concern within the various buildings at the facility. LAFD firefighters have the skills andmore » abilities to perform firefighting operations and other emergency response tasks that cannot be provided by other LANL personnel who have the required clearance level. This handout provides details of the information, along with maps and diagrams, to be presented during the familiarization tours. The handout will be distributed to the trainees at the time of the tour. A corresponding checklist will also be used as guidance during the familiarization tours to ensure that all required information is presented to LAFD personnel.« less

Instrumentation for studying binder burnout in an immobilized plutonium ceramic wasteform

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mitchell, M; Pugh, D; Herman, C

The Plutonium Immobilization Program produces a ceramic wasteform that utilizes organic binders. Several techniques and instruments were developed to study binder burnout on full size ceramic samples in a production environment. This approach provides a method for developing process parameters on production scale to optimize throughput, product quality, offgas behavior, and plant emissions. These instruments allow for offgas analysis, large-scale TGA, product quality observation, and thermal modeling. Using these tools, results from lab-scale techniques such as laser dilametry studies and traditional TGA/DTA analysis can be integrated. Often, the sintering step of a ceramification process is the limiting process step thatmore » controls the production throughput. Therefore, optimization of sintering behavior is important for overall process success. Furthermore, the capabilities of this instrumentation allows better understanding of plant emissions of key gases: volatile organic compounds (VOCs), volatile inorganics including some halide compounds, NO{sub x}, SO{sub x}, carbon dioxide, and carbon monoxide.« less

Support for HLW Direct Feed - Phase 2, VSL-15R3440-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matlack, K. S.; Pegg, I.; Joseph, I.

This report describes work performed to develop and test new glass and feed formulations originating from a potential flow-sheet for the direct vitrification of High Level Waste (HLW) with minimal or no pretreatment. In the HLW direct feed option that is under consideration for early operations at the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the pretreatment facility would be bypassed in order to support an earlier start-up of the vitrification facility. For HLW, this would mean that the ultrafiltration and caustic leaching operations that would otherwise have been performed in the pretreatment facility would either not be performedmore » or would be replaced by an interim pretreatment function (in-tank leaching and settling, for example). These changes would likely affect glass formulations and waste loadings and have impacts on the downstream vitrification operations. Modification of the pretreatment process may result in: (i) Higher aluminum contents if caustic leaching is not performed; (ii) Higher chromium contents if oxidative leaching is not performed; (iii) A higher fraction of supernate in the HLW feed resulting from the lower efficiency of in-tank washing; and (iv) A higher water content due to the likely lower effectiveness of in-tank settling compared to ultrafiltration. The HLW direct feed option has also been proposed as a potential route for treating HLW streams that contain the highest concentrations of fast-settling plutoniumcontaining particles, thereby avoiding some of the potential issues associated with such particles in the WTP Pretreatment facility [1]. In response, the work presented herein focuses on the impacts of increased supernate and water content on wastes from one of the candidate source tanks for the direct feed option that is high in plutonium.« less

Safe, Cost Effective Management of Inactive Facilities at the Savannah River Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Austin, W. E.; Yannitell, D. M.; Freeman, D. W.

The Savannah River Site is part of the U.S. Department of Energy complex. It was constructed during the early 1950s to produce basic materials (such as plutonium-239 and tritium) used in the production of nuclear weapons. The 310-square-mile site is located in South Carolina, about 12 miles south of Aiken, South Carolina, and about 15 miles southeast of Augusta, Georgia. Savannah River Site (SRS) has approximately 200 facilities identified as inactive. These facilities range in size and complexity from large nuclear reactors to small storage buildings. These facilities are located throughout the site including three reactor areas, the heavy watermore » plant area, the manufacturing area, and other research and support areas. Unlike DOE Closure Sites such as Hanford and Rocky Flats, SRS is a Project Completion Site with continuing missions. As facilities complete their defined mission, they are shutdown and transferred from operations to the facility disposition program. At the SRS, Facilities Decontamination and Decommissioning (FDD) personnel manage the disposition phase of a inactive facility's life cycle in a manner that minimizes life cycle cost without compromising (1) the health or safety of workers and the public or (2) the quality of the environment. The disposition phase begins upon completion of operations shutdown and extends through establishing the final end-state. FDD has developed innovative programs to manage their responsibilities within a constrained budget.« less

NON-AQUEOUS DISSOLUTION OF MASSIVE PLUTONIUM

DOEpatents

Reavis, J.G.; Leary, J.A.; Walsh, K.A.

1959-05-12

A method is presented for obtaining non-aqueous solutions or plutonium from massive forms of the metal. In the present invention massive plutonium is added to a salt melt consisting of 10 to 40 weight per cent of sodium chloride and the balance zinc chloride. The plutonium reacts at about 800 deg C with the zinc chloride to form a salt bath of plutonium trichloride, sodium chloride, and metallic zinc. The zinc is separated from the salt melt by forcing the molten mixture through a Pyrex filter.

OXIDATIVE METHOD OF SEPARATING PLUTONIUM FROM NEPTUNIUM

DOEpatents

Beaufait, L.J. Jr.

1958-06-10

A method is described of separating neptunium from plutonium in an aqueous solution containing neptunium and plutonium in valence states not greater than +4. This may be accomplished by contacting the solution with dichromate ions, thus oxidizing the neptunium to a valence state greater than +4 without oxidizing any substantial amount of plutonium, and then forming a carrier precipitate which carries the plutonium from solution, leaving the neptunium behind. A preferred embodiment of this invention covers the use of lanthanum fluoride as the carrier precipitate.

Continuous plutonium dissolution apparatus

DOEpatents

Meyer, F.G.; Tesitor, C.N.

1974-02-26

This invention is concerned with continuous dissolution of metals such as plutonium. A high normality acid mixture is fed into a boiler vessel, vaporized, and subsequently condensed as a low normality acid mixture. The mixture is then conveyed to a dissolution vessel and contacted with the plutonium metal to dissolve the plutonium in the dissolution vessel, reacting therewith forming plutonium nitrate. The reaction products are then conveyed to the mixing vessel and maintained soluble by the high normality acid, with separation and removal of the desired constituent. (Official Gazette)

23. AERIAL VIEW LOOKING SOUTHEAST AT THE PLUTONIUM OPERATION BUILDINGS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

23. AERIAL VIEW LOOKING SOUTHEAST AT THE PLUTONIUM OPERATION BUILDINGS 771, 776/777, AND 707. BUILDING 771, IN THE FOREGROUND, WAS BUILT IN 1952 TO HOUSE ALL PLUTONIUM OPERATIONS. BY 1956, BUILDING 771 WAS NO LONGER ADEQUATE FOR PRODUCTION DEMANDS. BUILDING 776/777, TO THE SOUTH OF BUILDING 771, WAS CONSTRUCTED TO HOUSE PLUTONIUM FABRICATION AND FOUNDRY OPERATIONS. PLUTONIUM RECOVERY REMAINED IN BUILDING 771. BY 1967, CONSTRUCTION ON BUILDING 707, TO THE SOUTH OF BUILDING 776/777, BEGAN AS PRODUCTION LEVELS CONTINUED TO EXPAND NECESSITATING THE NEED FOR ADDITIONAL PLUTONIUM FABRICATION SPACE (7/1/69). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

Stabilization and immobilization of military plutonium: A non-proliferation perspective

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leventhal, P.

1996-05-01

The Nuclear Control Institute welcomes this DOE-sponsored technical workshop on stabilization and immobilization of weapons plutonium (W Pu) because of the significant contribution it can make toward the ultimate non-proliferation objective of eliminating weapons-usable nuclear material, plutonium and highly enriched uranium (HEU), from world commerce. The risk of theft or diversion of these materials warrants concern, as only a few kilograms in the hands of terrorists or threshold states would give them the capability to build nuclear weapons. Military plutonium disposition questions cannot be addressed in isolation from civilian plutonium issues. The National Academy of Sciences has urged that {open_quotes}furthermore » steps should be taken to reduce the proliferation risks posed by all of the world`s plutonium stocks, military and civilian, separated and unseparated...{close_quotes}. This report discusses vitrification and a mixed oxide fuels option, and the effects of disposition choices on civilian plutonium fuel cycles.« less