An analysis of the background and development of regulations for the air transport of plutonium in the USA

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

McClure, J.D.; Luna, R.E.

1989-01-01

Several aspects of special packagings of plutonium for air transport should be recognized. The accident cases cited by Congressman Scheuer were incidents of local plutonium contamination in military aircraft accidents that had nuclear weapons on board. There is no disputing the occurrence of these military accidents but military weapon shipments were exempted from the provisions of the Scheuer amendment. There have been no recorded civilian aircraft crashes involving plutonium dispersal although there have been civilian aircraft crashes that were severe. Shortly after the introduction of the amendment by Mr. Scheuer on June 20, 1975, there was a serious aircraft crashmore » at JFK International. In his remarks to the House on July 24, 1975 Mr. Scheuer called attention to this event. The NRC originally opposed the provisions of the Scheuer amendment but with the passing of the amendment NRC compiled with its provisions. This led to the development of the plutonium air transport package PAT-1 in the US. The introduction of special rules for the air transport of plutonium into the US packaging regulations has been made them more severe than the provision of the international regulations, IAEA Safety Series 6. The IAEA is now discussing proposed regulations related to the air transport of plutonium. An additional legislative action was introduced the US in December 1987 which would require actual crash tests of packages intended for the air transport of plutonium, the Murkowski amendment. 13 refs.« less

75 FR 41850 - Amended Notice of Intent to Modify the Scope of the Surplus Plutonium Disposition Supplemental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-19

... immobilization). Also, DOE had identified a glass can-in-canister immobilization approach as its preferred... allow immobilization of some or all of the surplus plutonium in glass or ceramic material for disposal... in canisters to be filled with borosilicate glass containing intensely radioactive high-level waste...

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...

The benefits of an advanced fast reactor fuel cycle for plutonium management

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

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.

1996-12-31

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium andmore » long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.« less

Volatile Impurities in the Plutonium Immobilization Ceramic Wasteform

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

Cozzi, A.D.

1999-10-15

Approximately 18 of the 50 metric tons of plutonium identified for disposition contain significant quantities of impurities. A ceramic waste form is the chosen option for immobilization of the excess plutonium. The impurities associated with the stored plutonium have been identified (CaCl2, MgF2, Pb, etc.). For this study, only volatile species are investigated. The impurities are added individually. Cerium is used as the surrogate for plutonium. Three compositions, including the baseline composition, were used to verify the ability of the ceramic wasteform to accommodate impurities. The criteria for evaluation of the effect of the impurities were the apparent porosity andmore » phase assemblage of sintered pellets.« less

77 FR 1920 - Second Amended Notice of Intent To Modify the Scope of the Surplus Plutonium Disposition...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-12

... suitable for MOX fuel fabrication is disposal at the Waste Isolation Pilot Plant (WIPP) in New Mexico... Waste Processing Facility at SRS or disposal at the Waste Isolation Pilot Plant (WIPP) in New Mexico. On... are safety (criticality) limits on how much plutonium can be sent to the Defense Waste Processing...

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

Not Available

The Department of Energy (DOE) has contracted with Asea Brown Boveri-Combustion Engineering (ABB-CE) to provide information on the capability of ABB-CE`s System 80 + Advanced Light Water Reactor (ALWR) to transform, through reactor burnup, 100 metric tonnes (MT) of weapons grade plutonium (Pu) into a form which is not readily useable in weapons. This information is being developed as part of DOE`s Plutonium Disposition Study, initiated by DOE in response to Congressional action. This document Volume 2, provides a discussion of: Plutonium Fuel Cycle; Technology Needs; Regulatory Considerations; Cost and Schedule Estimates; and Deployment Strategy.

Reactive spark plasma synthesis of CaZrTi2O7 zirconolite ceramics for plutonium disposition

NASA Astrophysics Data System (ADS)

Sun, Shi-Kuan; Stennett, Martin C.; Corkhill, Claire L.; Hyatt, Neil C.

2018-03-01

Near single phase zirconolite ceramics, prototypically CaZrTi2O7, were fabricated by reactive spark plasma sintering (RSPS), from commercially available CaTiO3, ZrO2 and TiO2 reagents, after processing at 1200 °C for only 1 h. Ceramics were of theoretical density and formed with a controlled mean grain size of 1.9 ± 0.6 μm. The reducing conditions of RSPS afforded the presence of paramagnetic Ti3+, as demonstrated by EPR spectroscopy. Overall, this study demonstrates the potential for RSPS to be a disruptive technology for disposition of surplus separated plutonium stockpiles in ceramic wasteforms, given its inherent advantage of near net shape products and rapid throughput.

10 CFR 71.74 - Accident conditions for air transport of plutonium.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Accident conditions for air transport of plutonium. 71.74 Section 71.74 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE... in a manner that prevents any members or devices used to support the bar from contacting the package...

10 CFR 71.74 - Accident conditions for air transport of plutonium.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Accident conditions for air transport of plutonium. 71.74 Section 71.74 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE... in a manner that prevents any members or devices used to support the bar from contacting the package...

10 CFR 71.74 - Accident conditions for air transport of plutonium.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Accident conditions for air transport of plutonium. 71.74 Section 71.74 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE... in a manner that prevents any members or devices used to support the bar from contacting the package...

10 CFR 71.74 - Accident conditions for air transport of plutonium.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Accident conditions for air transport of plutonium. 71.74 Section 71.74 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE... in a manner that prevents any members or devices used to support the bar from contacting the package...

10 CFR 71.74 - Accident conditions for air transport of plutonium.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Accident conditions for air transport of plutonium. 71.74 Section 71.74 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE... in a manner that prevents any members or devices used to support the bar from contacting the package...

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

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

PAT-1 safety analysis report addendum.

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

Weiner, Ruth F.; Schmale, David T.; Kalan, Robert J.

2010-09-01

The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The purpose of this SAR Addendum is to incorporate plutonium (Pu) metal as a new payload for the PAT-1 package. The Pu metal is packed in an inner container (designated the T-Ampoule) that replaces the PC-1 inner container. The documentation and results from analysis contained in this addendum demonstrate that the replacement of the PC-1 and associated packaging material with the T-Ampoule and associated packaging withmore » the addition of the plutonium metal content are not significant with respect to the design, operating characteristics, or safe performance of the containment system and prevention of criticality when the package is subjected to the tests specified in 10 CFR 71.71, 71.73 and 71.74.« less

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

Singledecker, Steven John

The purpose of this document is to describe the waste stream from Z-Pinch Residual Waste Project that due to worker safety concerns and operational efficiency is a candidate for blending Transuranic and low level waste together and can be safely packaged as low-level waste consistent with DOE Order 435.1 requirements and NRC guidance 10 CFR 61.42. This waste stream consists of the Pu-ICE post-shot containment systems, including plutonium targets, generated from the Z Machine experiments requested by LANL and conducted by SNL/NM. In the past, this TRU waste was shipped back to LANL after Sandia sends the TRU data packagemore » to LANL to certify the characterization (by CCP), transport and disposition at WIPP (CBFO) per LANL MOU-0066. The Low Level Waste is managed, characterized, shipped and disposed of at NNSS by SNL/NM per Sandia MOU # 11-S-560.« less

Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

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

J.P. Nicot

The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the developmentmore » plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the WP. This calculation supports disposal criticality analysis and has been prepared in accordance with AP-3.12Q, Calculations (Ref. 49). This calculation uses results from Ref. 4 on actinide accumulation in the invert and more generally does reference heavily the cited calculation. In addition to the information provided in this calculation, the reader is referred to the cited calculation for a more thorough treatment of items applying to both the invert and fracture system such as the choice of the thermodynamic database, the composition of J-13 well water, tuff composition, dissolution rate laws, Pu(OH){sub 4} solubility and also for details on the source term composition. The flow conditions (seepage rate, water velocity in fractures) in the drift and the fracture system beneath initially referred to the TSPA-VA because this work was prepared before the release of the work feeding the TSPA-SR. Some new information feeding the TSPA-SR has since been included. Similarly, the soon-to-be-qualified thermodynamic database data0.ymp has not been released yet.« less

HB-Line Plutonium Oxide Data Collection Strategy

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

Watkins, R.; Varble, J.; Jordan, J.

2015-05-26

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

49 CFR 176.2 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-10-01

..., spillage, or other accident. INF cargo means packaged irradiated nuclear fuel, plutonium or high-level... Irradiated Nuclear Fuel, Plutonium and High-Level Radioactive Wastes on Board Ships” (INF Code) contained in...

Status of plutonium ceramic immobilization processes and immobilization forms

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

Ebbinghaus, B.B.; Van Konynenburg, R.A.; Vance, E.R.

1996-05-01

Immobilization in a ceramic followed by permanent emplacement in a repository or borehole is one of the alternatives currently being considered by the Fissile Materials Disposition Program for the ultimate disposal of excess weapons-grade plutonium. To make Pu recovery more difficult, radioactive cesium may also be incorporated into the immobilization form. Valuable data are already available for ceramics form R&D efforts to immobilize high-level and mixed wastes. Ceramics have a high capacity for actinides, cesium, and some neutron absorbers. A unique characteristic of ceramics is the existence of mineral analogues found in nature that have demonstrated actinide immobilization over geologicmore » time periods. The ceramic form currently being considered for plutonium disposition is a synthetic rock (SYNROC) material composed primarily of zirconolite (CaZrTi{sub 2}O{sub 7}), the desired actinide host phase, with lesser amounts of hollandite (BaAl{sub 2}Ti{sub 6}O{sub 16}) and rutile (TiO{sub 2}). Alternative actinide host phases are also being considered. These include pyrochlore (Gd{sub 2}Ti{sub 2}O{sub 7}), zircon (ZrSiO{sub 4}), and monazite (CePO{sub 4}), to name a few of the most promising. R&D activities to address important technical issues are discussed. Primarily these include moderate scale hot press fabrications with plutonium, direct loading of PuO{sub 2} powder, cold press and sinter fabrication methods, and immobilization form formulation issues.« less

SHIPMENT OF TWO DOE-STD-3013 CONTAINERS IN A 9977 TYPE B PACKAGE

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

Abramczyk, G.; Bellamy, S.; Loftin, B.

2011-06-06

The 9977 is a certified Type B Packaging authorized to ship uranium and plutonium in metal and oxide forms. Historically, the standard container for these materials has been the DOE-STD-3013 which was specifically designed for the long term storage of plutonium bearing materials. The Department of Energy has used the 9975 Packaging containing a single 3013 container for the transportation and storage of these materials. In order to reduce container, shipping, and storage costs, the 9977 Packaging is being certified for transportation and storage of two 3013 containers. The challenges and risks of this content and the 9977s ability tomore » meet the Code of Federal Regulations for the transport of these materials are presented.« less

Plutonium working group report on environmental, safety and health vulnerabilities associated with the department`s plutonium storage. Volume II, Appendix B, Part 9: Oak Ridge site site team report

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

NONE

1994-09-01

This report provides the input to and results of the Department of Energy (DOE) - Oak Ridge Operations (ORO) DOE Plutonium Environment, Safety and Health (ES & H) Vulnerability Assessment (VA) self-assessment performed by the Site Assessment Team (SAT) for the Oak Ridge National Laboratory (ORNL or X-10) and the Oak Ridge Y-12 Plant (Y-12) sites that are managed by Martin Marietta Energy Systems, Inc. (MMES). As initiated (March 15, 1994) by the Secretary of Energy, the objective of the VA is to identify and rank-order DOE-ES&H vulnerabilities associated for the purpose of decision making on the interim safe managementmore » and ultimate disposition of fissile materials. This assessment is directed at plutonium and other co-located transuranics in various forms.« less

Quarterly technical progress report, February 1, 1996--April 30, 1996

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

NONE

This report from the Amarillo National REsource Center for PLutonium provides research highlights and provides information regarding the public dissemination of information. The center is a a scientific resource for information regarding the issues of the storage, disposition, potential utilization and transport of plutonium, high explosives, and other hazardous materials generated from nuclear weapons dismantlement. The center responds to informational needs and interpretation of technical and scientific data raised by interested parties and advisory groups. Also, research efforts are carried out on remedial action programs and biological/agricultural studies.

FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

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

Greene, S.R.; Bevard, B.B.

1996-10-07

Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives.

Preliminary fabrication and characterisation of inert matrix and thoria fuels for plutonium disposition in light water reactors

NASA Astrophysics Data System (ADS)

Vettraino, F.; Magnani, G.; La Torretta, T.; Marmo, E.; Coelli, S.; Luzzi, L.; Ossi, P.; Zappa, G.

1999-08-01

The plutonium disposition is presently acknowledged as a most urgent issue at the world level. Inert matrix and thoria fuel concepts for Pu burning in LWRs show good potential in providing effective and ultimate solutions to this issue. In non-fertile (U-free) inert matrix fuel, plutonium oxide is diluted within inert oxides such as stabilised ZrO 2, Al 2O 3, MgO or MgAl 2O 4. Thoria addition, which helps improve neutronic characteristics of inert fuels, appears as a promising variant of U-free fuel. In the context of an R&D activity aimed at assessing the feasibility of the fuel concept above, simulated fuel pellets have been produced both from dry-powder metallurgy and the sol-gel route. Results show that they can be fabricated by matching basic nuclear grade specifications such as the required geometry, density and microstructure. Some characterisation testing dealing with thermo-physical properties, ion irradiation damage and solubility also have been started. Results from thermo-physical measurements at room temperature have been achieved. A main feature stemming from solubility testing outcomes is a very high chemical stability which should render the fuel strongly diversion resistant and suitable for direct final disposal in deep geological repository (once-through solution).

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

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

ULLAH, M K

2001-02-26

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

Screening study for evaluation of the potential for system 80+ to consume excess plutonium - Volume 1. Final report

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

Not Available

1994-04-30

As part of the U.S. effort to evaluate technologies offering solutions for the safe disposal or utilization of surplus nuclear materials, the fiscal year 1993 Energy and Water Appropriations legislation provided the Department of Energy (DOE) the necessary funds to conduct multi-phased studies to determine the technical feasibility of using reactor technologies for the triple mission of burning weapons grade plutonium, producing tritium for the existing smaller weapons stockpile, and generating commercial electricity. DOE limited the studies to five advanced reactor designs. Among the technologies selected is the ABB-Combustion Engineering (ABB-CE) System 80+. The DOE study, currently in Phase ID,more » is proceeding with a more detailed evaluation of the design`s capability for plutonium disposition.« less

Review of Excess Weapons Plutonium Disposition LLNL Contract Work in Russia-(English)

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

Jardine, L; Borisov, G B

This third meeting of the recently completed and ongoing Russian plutonium immobilization contract work was held at the State Education Center (SEC) in St. Petersburg on January 14-18, 2002. The meeting agenda is reprinted here as Appendix A and the attendance list as Appendix B. The meeting had 58 Russian participants from 21 Russian organizations, including the industrial sites (Mayak, Krasonayarsk-26, Tomsk), scientific institutes (VNIINM, KRI, VNIPIPT, RIAR), design organizations (VNIPIET and GSPI), universities (Nyzhny Novgorod, Urals Technical), Russian Academy of Sciences (Institute of Physical Chemistry or IPhCh, Institute of Ore-Deposit Geology, Petrography, Mineralogy, and Geochemistry or IGEM), Radon-Moscow, S&TCmore » Podol'osk, Kharkov-Ukraine, GAN-SEC-NRS and SNIIChM, the RF Ministry of Atomic Energy (Minatom) and Gosatomnadzor (GAN). This volume, published by LLNL, documents this third annual meeting. Forty-nine technical papers were presented by the Russian participants, and nearly all of these have been collected in this Proceedings. The two objectives for the meeting were to: (1) Bring together the Russian organizations, experts, and managers performing this contract work into one place for four days to review and discuss their work amongst each other. (2) Publish a meeting summary and proceedings of all the excellent Russian plutonium immobilization and other plutonium disposition contract work in one document so that the wide extent of the Russian immobilization activities are documented, referencable and available for others to use, as were the Proceedings of the two previous meetings. Attendees gave talks describing their LLNL contract work and submitted written papers documenting their contract work (in English and Russian), in both hard copy and on computer disks. Simultaneous translation into Russian and English was used for presentations made at the State Region Educational Center (SEC).« 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

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

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

Romano, T.

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is validmore » until October 1, 1999. After this date, an update or upgrade to this document is required.« less

Thermal Analysis on the Shipment of Russian Plutonium Fuel

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

Or, Chuen T; Skrabek, Emanuel A; Carpenter, Robert T

Paper presented at the 12th Symposium on Space Nuclear Power and Propulsion in Albuquerque, NM in January 1995. The Mound 9516 shipping package was designed for the shipment of Plutonium-238 fuel. One of the shipping configurations is the Russian Pu-238 powder can. Computer models using SINDA were created to predict the temperatures of the package under normal conditions of transport (NCT: 38oC ambient temperature), under hypothetical accident conditions (HAC: engulfed in fire for 30 minutes), and inside a standard cargo container. Pressure increases inside the package due to the expansion of the trapped gases and helium gas generation from isotopemore » decay were also analyzed. There is a duplicate copy and also a copy in the ESD Files.« less

Advanced research workshop: nuclear materials safety

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

Jardine, L J; Moshkov, M M

The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on themore » storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of nuclear experience on a common objectiveÑthe safe and secure storage and disposition of excess fissile nuclear materials.« less

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2012 CFR

2012-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2014 CFR

2014-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2010 CFR

2010-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2013 CFR

2013-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, M.L.; Rosenstein, R.G.

1998-10-13

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

Mox fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-05-15

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

2001-07-17

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

MOX fuel arrangement for nuclear core

DOEpatents

Kantrowitz, Mark L.; Rosenstein, Richard G.

1998-01-01

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

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.

Non-proliferation, safeguards, and security for the fissile materials disposition program immobilization alternatives

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

Duggan, R.A.; Jaeger, C.D.; Tolk, K.M.

1996-05-01

The Department of Energy is analyzing long-term storage and disposition alternatives for surplus weapons-usable fissile materials. A number of different disposition alternatives are being considered. These include facilities for storage, conversion and stabilization of fissile materials, immobilization in glass or ceramic material, fabrication of fissile material into mixed oxide (MOX) fuel for reactors, use of reactor based technologies to convert material into spent fuel, and disposal of fissile material using geologic alternatives. This paper will focus on how the objectives of reducing security and proliferation risks are being considered, and the possible facility impacts. Some of the areas discussed inmore » this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threats, and (4) potential proliferation, safeguards, and security issues and impacts on the facilities. Issues applicable to all of the possible disposition alternatives will be discussed in this paper. However, particular attention is given to the plutonium immobilization alternatives.« less

The 9th international symposium on the packaging and transportation of radioactive materials

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

None

1989-06-01

This three-volume document contains the papers and poster sessions presented at the symposium. Volume 3 contains 87 papers on topics such as structural codes and benchmarking, shipment of plutonium by air, spent fuel shipping, planning, package design and risk assessment, package testing, OCRWN operations experience and regulations. Individual papers were processed separately for the data base. (TEM)

21 CFR 25.30 - General.

Code of Federal Regulations, 2010 CFR

2010-04-01

... of recalls. (d) Destruction or disposition of any FDA-regulated article condemned after seizure or... if the method of destruction or disposition of the article, including packaging material, is in... grants for statistical and epidemiological studies, surveys and inventories, literature searches, and...

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

Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

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

Laurinat, James E.; Askew, Neal M.; Hensel, Steve J.

2013-03-21

Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation ofmore » hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.« less

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

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

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

Plutonium immobilization in glass and ceramics

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

Knecht, D.A.; Murphy, W.M.

1996-05-01

The Materials Research Society Nineteenth Annual Symposium on the Scientific Basis for Nuclear Waste Management was held in Boston on November 27 to December 1, 1995. Over 150 papers were presented at the Symposium dealing with all aspects of nuclear waste management and disposal. Fourteen oral sessions and on poster session included a Plenary session on surplus plutonium dispositioning and waste forms. The proceedings, to be published in April, 1996, will provide a highly respected, referred compilation of the state of scientific development in the field of nuclear waste management. This paper provides a brief overview of the selected Symposiummore » papers that are applicable to plutonium immobilization and plutonium waste form performance. Waste forms that were described at the Symposium cover most of the candidate Pu immobilization options under consideration, including borosilicate glass with a melting temperature of 1150 {degrees}C, a higher temperature (1450 {degrees}C) lanthanide glass, single phase ceramics, multi-phase ceramics, and multi-phase crystal-glass composites (glass-ceramics or slags). These Symposium papers selected for this overview provide the current status of the technology in these areas and give references to the relevant literature.« less

Preparation of plutonium-bearing ceramics via mechanically activated precursor

NASA Astrophysics Data System (ADS)

Chizhevskaya, S. V.; Stefanovsky, S. V.

2000-07-01

The problem of excess weapons plutonium disposition is suggested to be solved by means of its incorporation in stable ceramics with high chemical durability and radiation resistivity. The most promising host phases for plutonium as well as uranium and neutron poisons (gadolinium, hafnium) are zirconolite, pyrochlore, zircon, zirconia [1,2], and murataite [3]. Their production requires high temperatures and a fine-grained homogeneous precursor to reach final waste form with high quality and low leachability. Currently various routes to homogeneous products preparation such as sol-gel technology, wet-milling, and grinding in a ball or planetary mill are used. The best result demonstrates sol-gel technology but this route is very complicated. An alternative technology for preparation of ceramic precursors is the treatment of the oxide batch with high mechanical energy [4]. Such a treatment produces combination of mechanical (fine milling with formation of various defects, homogenization) and chemical (split bonds with formation of active centers—free radicals, ion-radicals, etc.) effects resulting in higher reactivity of the activated batch.

Sandia Review of High Bridge Associates Report: Comparison of Plutonium Disposition Alternatives: WIPP Diluted Plutonium Storage and MOX Fuel Irradiation

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

Shoemaker, Paul E.; Hardin, Ernest; Park, HeeHo Daniel

The subject report from High Bridge Associates (HBA) was issued on March 2, 2016, in reaction to a U.S. Department of Energy (DOE) program decision to pursue down-blending of surplus Pu and geologic disposal at the Waste Isolation Pilot Plant (WIPP). Sandia National Laboratories was requested by the DOE to review the technical arguments presented in the HBA report. Specifically, this review is organized around three technical topics: criticality safety, radiological release limits, and thermal impacts. Questions raised by the report pertaining to legal and regulatory requirements, safeguards and security, international agreements, and costing of alternatives, are beyond the scopemore » of this review.« less

Science and technology review, April 1997

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

Upadhye, R.

1997-04-01

This month's issue has the following articles: (1) The Laboratory in the News; (2) Commentary by Tom Isaacs--Shaping Nuclear Materials Policy; (3) Dealing with a Dangerous Surplus from the Cold War--Since the end of the Cold War, the Laboratory has been spearheading studies on the disposition of surplus weapons plutonium; (4) Volcanoes: A Peek into Our Planet's Plumbing; and (5) Optical Networks: The Wave of the Future.

27 CFR 19.773 - Daily record of wholesale liquor dealer and taxpaid storeroom operations.

Code of Federal Regulations, 2010 CFR

2010-04-01

... the receipt and disposition of distilled spirits and wines shall contain all data necessary... distilled spirits involved (proof and proof gallons if in packages, wine gallons or liters and proof if in bottles); (6) The package identification or serial numbers of packages involved; (7) The name of the...

Validation of COG10 and ENDFB6R7 on the Auk Workstation for General Application to Plutonium Systems

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

Percher, Catherine G

2011-08-08

The COG 10 code package1 on the Auk workstation is now validated with the ENBFB6R7 neutron cross section library for general application to plutonium (Pu) systems by comparison of the calculated keffective to the expected keffective of several relevant experimental benchmarks. This validation is supplemental to the installation and verification of COG 10 on the Auk workstation2.

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

Radiological analysis of plutonium glass batches with natural/enriched boron

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

Rainisch, R.

2000-06-22

The disposition of surplus plutonium inventories by the US Department of Energy (DOE) includes the immobilization of certain plutonium materials in a borosilicate glass matrix, also referred to as vitrification. This paper addresses source terms of plutonium masses immobilized in a borosilicate glass matrix where the glass components include both natural boron and enriched boron. The calculated source terms pertain to neutron and gamma source strength (particles per second), and source spectrum changes. The calculated source terms corresponding to natural boron and enriched boron are compared to determine the benefits (decrease in radiation source terms) for to the use ofmore » enriched boron. The analysis of plutonium glass source terms shows that a large component of the neutron source terms is due to (a, n) reactions. The Americium-241 and plutonium present in the glass emit alpha particles (a). These alpha particles interact with low-Z nuclides like B-11, B-10, and O-17 in the glass to produce neutrons. The low-Z nuclides are referred to as target particles. The reference glass contains 9.4 wt percent B{sub 2}O{sub 3}. Boron-11 was found to strongly support the (a, n) reactions in the glass matrix. B-11 has a natural abundance of over 80 percent. The (a, n) reaction rates for B-10 are lower than for B-11 and the analysis shows that the plutonium glass neutron source terms can be reduced by artificially enriching natural boron with B-10. The natural abundance of B-10 is 19.9 percent. Boron enriched to 96-wt percent B-10 or above can be obtained commercially. Since lower source terms imply lower dose rates to radiation workers handling the plutonium glass materials, it is important to know the achievable decrease in source terms as a result of boron enrichment. Plutonium materials are normally handled in glove boxes with shielded glass windows and the work entails both extremity and whole-body exposures. Lowering the source terms of the plutonium batches will make the handling of these materials less difficult and will reduce radiation exposure to operating workers.« less

10 CFR 71.64 - Special requirements for plutonium air shipments.

Code of Federal Regulations, 2011 CFR

2011-01-01

... so that under the tests specified in— (1) Section 71.74 (“Accident conditions for air transport of plutonium”)— (i) The containment vessel would not be ruptured in its post-tested condition, and the package... 10 mSv/h (1 rem/h) at a distance of 1 m (40 in) from the surface of the package in its post-tested...

Disposition of Chicago Pile 5 (CP-5) Converter Tubes in the 10-160B Cask

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

Pancake, Daniel C.; Rock, Cynthia

This paper will focus on the unique characterization, packaging, and transportation issues associated with the disposition of the two CP-5 Converter Tube assemblies from Argonne National Laboratory. The converter tubes were constructed of combinations of HEU and alloys of zirconium, and were part of the original research facilities attached to the CP-5 reactor during operating evolutions. These assemblies were heavily irradiated during their operational lifetime, and were segregated from the balance of irradiated test specimens when the reactor was deactivated and slated for Decontamination and Demolition (D&D). In addition, the substantial contribution of fissile material to the assemblies’ inventory mademore » the potential disposition pathways extremely challenging. As a result, these items became part of Argonne’s legacy “nuclear footprint”, and were added to the Nuclear Footprint Reduction Project scope for disposition. The Project was responsible for the size reduction and characterization of these items, as well as the ultimate disposition. After negotiating a disposal pathway for these tubes, there were significant transportation issues that required a small team to overcome, in order to successfully ship these items to the Nevada National Security Site (NNSS). The Project team at Argonne, technical support from transportation specialists, licensing support from the 10-160B license owner, the Savanah River National Lab (SRNL) Packaging Certification Team (PCT, and the DOE EM-33 staff contributed to license and safety analysis report amendments that eventually authorized the shipment of the material. The paper will identify the organizations, and the specific actions, required to successfully make three “one of a kind” shipments of irradiated test specimen material. This will include the unique packaging configurations, contents modification for the cask license (via the Amendment process), criticality evaluations, and associated review and approval processes.« less

27 CFR 19.601 - Finished products records.

Code of Federal Regulations, 2011 CFR

2011-04-01

... Finished products records. (a) Bottling and packaging. A proprietor must maintain daily transaction records and a daily summary record of the kind and quantity of finished products bottled or packaged within... daily records of the disposition of finished products from the processing account at the distilled...

21 CFR 111.87 - Who conducts a material review and makes a disposition decision?

Code of Federal Regulations, 2010 CFR

2010-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Requirement to Establish a Production and Process Control System § 111.87 Who conducts a material review and makes a disposition... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Who conducts a material review and makes a...

High Integrity Can Design Interfaces

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

Shaber, E.L.

1998-08-01

The National Spent Nuclear Fuel Program is chartered with facilitating the disposition of DOE-owned spent nuclear fuel to allow disposal at a geologic repository. This is done through coordination with the repository program and by assisting DOE Site owners of SNF with needed information, standardized requirements, packaging approaches, etc. The High Integrity Can (HIC) will be manufactured to provide a substitute or barrier enhancement for normal fuel geometry and cladding. The can would be nested inside the DOE standardized canister which is designed to interface with the repository waste package. The HIC approach may provide the following benefits over typicalmore » canning approaches for DOE SNF. (a) It allows ready calculation and management of criticality issues for miscellaneous. (b) It segments and further isolates damaged or otherwise problem materials from normal SNF in the repository package. (c) It provides a very long term corrosion barrier. (d) It provides an extra internal pressure barrier for particulates, gaseous fission products, hydrogen, and water vapor. (e) It delays any potential release of fission products to the repository environment. (f) It maintains an additional level of fuel geometry control during design basis accidents, rock-fall, and seismic events. (g) When seal welded, it could provide the additional containment required for shipments involving plutonium content in excess of 20 Ci. (10 CFR 71.63.b) if integrated with an appropriate cask design. Long term corrosion protection is central to the HIC concept. The material selected for the HIC (Hastelloy C-22) has undergone extensive testing for repository service. The most severe theoretical interactions between iron, repository water containing chlorides and other repository construction materials have been tested. These expected chemical species have not been shown capable of corroding the selected HIC material. Therefore, the HIC should provide a significant barrier to DOE SNF dispersal long after most commercial SNF has degraded and begun moving into the repository environment.« less

21 CFR 111.370 - What requirements apply to rejected dietary supplements?

Code of Federal Regulations, 2013 CFR

2013-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Production and Process Control... disposition any dietary supplement that is rejected and unsuitable for use in manufacturing, packaging, or... 21 Food and Drugs 2 2013-04-01 2013-04-01 false What requirements apply to rejected dietary...

21 CFR 111.370 - What requirements apply to rejected dietary supplements?

Code of Federal Regulations, 2014 CFR

2014-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Production and Process Control... disposition any dietary supplement that is rejected and unsuitable for use in manufacturing, packaging, or... 21 Food and Drugs 2 2014-04-01 2014-04-01 false What requirements apply to rejected dietary...

21 CFR 111.370 - What requirements apply to rejected dietary supplements?

Code of Federal Regulations, 2011 CFR

2011-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Production and Process Control... disposition any dietary supplement that is rejected and unsuitable for use in manufacturing, packaging, or... 21 Food and Drugs 2 2011-04-01 2011-04-01 false What requirements apply to rejected dietary...

21 CFR 111.370 - What requirements apply to rejected dietary supplements?

Code of Federal Regulations, 2012 CFR

2012-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Production and Process Control... disposition any dietary supplement that is rejected and unsuitable for use in manufacturing, packaging, or... 21 Food and Drugs 2 2012-04-01 2012-04-01 false What requirements apply to rejected dietary...

24 CFR 291.304 - Bidding process.

Code of Federal Regulations, 2012 CFR

2012-04-01

... DEVELOPMENT HUD-OWNED PROPERTIES DISPOSITION OF HUD-ACQUIRED SINGLE FAMILY PROPERTY Sale of HUD-Held Single... HUD in accordance with instructions in the bid package for a particular sale. (b) Effect of bid. By... bid package. Along with the bid, the bidder must submit an executed copy of the Loan Sale Agreement...

24 CFR 291.304 - Bidding process.

Code of Federal Regulations, 2014 CFR

2014-04-01

... DEVELOPMENT HUD-OWNED PROPERTIES DISPOSITION OF HUD-ACQUIRED SINGLE FAMILY PROPERTY Sale of HUD-Held Single... HUD in accordance with instructions in the bid package for a particular sale. (b) Effect of bid. By... bid package. Along with the bid, the bidder must submit an executed copy of the Loan Sale Agreement...

24 CFR 291.304 - Bidding process.

Code of Federal Regulations, 2010 CFR

2010-04-01

... DEVELOPMENT HUD-OWNED PROPERTIES DISPOSITION OF HUD-ACQUIRED SINGLE FAMILY PROPERTY Sale of HUD-Held Single... HUD in accordance with instructions in the bid package for a particular sale. (b) Effect of bid. By... bid package. Along with the bid, the bidder must submit an executed copy of the Loan Sale Agreement...

21 CFR 111.370 - What requirements apply to rejected dietary supplements?

Code of Federal Regulations, 2010 CFR

2010-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Production and Process Control... disposition any dietary supplement that is rejected and unsuitable for use in manufacturing, packaging, or... 21 Food and Drugs 2 2010-04-01 2010-04-01 false What requirements apply to rejected dietary...

77 FR 56241 - Board Meeting; October 17, 2012; Idaho Falls, ID

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-12

.... Nuclear Waste Technical Review Board will meet to discuss DOE work on packaging, transporting, and...) plans for the packaging, transportation, and disposition of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). Among the topics that will be discussed are current activities being undertaken by...

Superconducting RF Linacs Driving Subcritical Reactors for Profitable Disposition of Surplus Weapons-grade Plutonium

NASA Astrophysics Data System (ADS)

Cummings, Mary Anne; Johnson, Rolland

Acceptable capital and operating costs of high-power proton accelerators suitable for profitable commercial electric-power and process-heat applications have been demonstrated. However, studies have pointed out that even a few hundred trips of an accelerator lasting a few seconds would lead to unacceptable thermal stresses as each trip causes fission to be turned off in solid fuel structures found in conventional reactors. The newest designs based on the GEM*STAR concept take such trips in stride by using molten-salt fuel, where fuel pin fatigue is not an issue. Other aspects of the GEM*STAR concept which address all historical reactor failures include an internal spallation neutron target and high temperature molten salt fuel with continuous purging of volatile radioactive fission products such that the reactor contains less than a critical mass and almost a million times fewer volatile radioactive fission products than conventional reactors. GEM*STAR is a reactor that without redesign will burn spent nuclear fuel, natural uranium, thorium, or surplus weapons material. It will operate without the need for a critical core, fuel enrichment, or reprocessing making it an excellent candidate for export. As a first application, the design for a pilot plant is described for the profitable disposition of surplus weapons-grade plutonium by using process heat to produce green diesel fuel for the Department of Defense (DOD) from natural gas and renewable carbon.

Teaching Old Packaging New Tricks - 12593

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

England, Jeffery L.; Shuler, James M.

2012-07-01

Waste disposition campaigns have been an industry and government focus area since the mid- 1970's. With increased focus on this issue, and a lot of hard work, most waste packaging and transportation issues have been addressed. The material has been successfully shipped and dis-positioned. DOE has successfully de-inventoried materials from multiple sites to meet material consolidation, footprint reduction, nonproliferation, and regulatory obligations with cost savings from reduced maintenance and regulatory compliance. There has been a wide range of certified shipping packagings for the transportation of hazardous materials to meet most of the waste needs. The remaining materials are problematic, generallymore » low volume, and do not meet the certified content of the existing inventory of packaging. Designing, testing and certifying new packaging designs can be a long and expensive process and for small volumes of material it is cost prohibitive. One very cost effective option is to lease and use a certified packaging to overpack waste containers. There are many robust certified packagings available with the capability to envelope the waste content. The capability to use inner containers, inside the current fleet of certified casks or packaging, to address specific content problems of additional shielding (e.g., U-233) or containment (e.g., sodium bonded nuclear material) has successfully expanded the capability for timely cost effective shipment of unique contents. This option has been used successfully in the NAC-LWT, T-3 and other packagings. (authors)« less

21 CFR 111.113 - What quality control operations are required for a material review and disposition decision?

Code of Federal Regulations, 2010 CFR

2010-04-01

... MANUFACTURING PRACTICE IN MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS... that adulterates or may lead to adulteration of the component, dietary supplement, or packaging, or... or batches of a dietary supplement; or (5) A dietary supplement is returned. (b)(1) When there is a...

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

Progress in the Assessment of Waste-forms for the Immobilisation of UK Civil Plutonium

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

Harrison, M.T.; Scales, C.R.; Maddrell, E.R.

The alternatives for the disposition of the UK's civil plutonium stocks are currently being investigated by Nexia Solutions Ltd. on behalf of the Nuclear Decommissioning Authority (NDA). A number of scenarios are currently being considered depending on the strategic requirements of the UK. The two main disposition options are: re-use as MOX (Mixed Oxide) fuel in reactors, or immobilisation in the event of any material being declared surplus to requirements. The amount of Pu which will require immobilisation will depend on future UK nuclear strategy, along with the extent of any stocks deemed unsuitable for re-use. However, it is likelymore » that some portion will have to be immobilised and therefore three credible waste-forms are under consideration; ceramic, glass and 'immobilisation' MOX. These are currently being developed and assessed in a systematic programme that involves periodic evaluation against a range of criteria. In this way, by down-selecting on the basis of robust and technical review, the most appropriate option for immobilising surplus civil plutonium in the UK can be recommended. The latest results from the immobilisation experimental programme are presented following the de-selection of the least favourable glass and ceramic candidates. The main criteria for this decision were waste loading, durability, processability, criticality and proliferation resistance. In addition, the durability of unirradiated MOX fuel is being examined to determine its potential as a wasteform for Pu, and recent leach test data is discussed. The current evaluation comprises not only a comparison of the relevant physical properties of the various waste-forms, but also key processing parameters, e.g. glass viscosity and melter technology, ceramic fabrication routes, and criticality issues. Other important aspects of the long-term behaviour of the waste-forms under consideration in a potential repository environment, such as radiation damage, criticality control and the properties of any neutron poisons present, are also included. (authors)« less

Argillite And Crystalline Disposal Research: Accomplishments And Path-Forward.

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

McMahon, Kevin A.; Jove-Colon, Carlos F.; Wang, Yifeng

The intention of this document is to provide a path-forward for research and development (R&D) for two host rock media-specific (argillite and crystalline) disposal research work packages within the Used Fuel Disposition Campaign (UFDC). The two work packages, Argillite Disposal R&D and Crystalline Disposal R&D, support the achievement of the overarching mission and objectives of the Department of Energy Office of Nuclear Energy Fuel Cycle Technologies Program. These two work packages cover many of the fundamental technical issues that will have multiple implications to other disposal research work packages by bridging knowledge gaps to support the development of the safetymore » case. The path-forward begins with the assumption of target dates that are set out in the January 2013 DOE Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (http://energy.gov/downloads/strategy-management-and-disposal-used-nuclear-fuel-and-high-levelradioactive- waste). The path-forward will be maintained as a living document and will be updated as needed in response to available funding and the progress of multiple R&D tasks in the Used Fuel Disposition Campaign and the Fuel Cycle Technologies Program. This path forward is developed based on the report of “Used Fuel Disposition Campaign Disposal Research and Development Roadmap (FCR&D-USED- 2011-000065 REV0)” (DOE, 2011). This document delineates the goals and objectives of the UFDC R&D program, needs for generic disposal concept design, and summarizes the prioritization of R&D issues.« less

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

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

Actinides in the Geosphere

NASA Astrophysics Data System (ADS)

Runde, Wolfgang; Neu, Mary P.

Since the 1950s actinides have been used to benefit industry, science, health, and national security. The largest industrial application, electricity generation from uranium and thorium fuels, is growing worldwide. Thus, more actinides are being mined, produced, used and processed than ever before. The future of nuclear energy hinges on how these increasing amounts of actinides are contained in each stage of the fuel cycle, including disposition. In addition, uranium and plutonium were built up during the Cold War between the United States and the Former Soviet Union for defense purposes and nuclear energy.

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.

Savannah River Site Eastern Transportation Hub: A Concept For a DOE Eastern Packaging, Staging and Maintenance Center - 13143

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

England, Jeffery L.; Adams, Karen; Maxted, Maxcine

2013-07-01

The Department of Energy (DOE) is working to de-inventory sites and consolidate hazardous materials for processing and disposal. The DOE administers a wide range of certified shipping packages for the transport of hazardous materials to include Special Nuclear Material (SNM), radioactive materials, sealed sources and radioactive wastes. A critical element to successful and safe transportation of these materials is the availability of certified shipping packages. There are over seven thousand certified packagings (i.e., Type B/Type AF) utilized within the DOE for current missions. The synergistic effects of consolidated maintenance, refurbishment, testing, certification, and costing of these services would allow formore » efficient management of the packagings inventory and to support anticipated future in-commerce shipping needs. The Savannah River Site (SRS) receives and ships radioactive materials (including SNM) and waste on a regular basis for critical missions such as consolidated storage, stabilization, purification, or disposition using H-Canyon and HB-Line. The Savannah River National Laboratory (SRNL) has the technical capability and equipment for all aspects of packaging management. SRS has the only active material processing facility in the DOE complex and is one of the sites of choice for nuclear material consolidation. SRS is a logical location to perform maintenance and periodic testing of the DOE fleet of certified packagings. This initiative envisions a DOE Eastern Packaging Staging and Maintenance Center (PSMC) at the SRS and a western hub at the Nevada National Security Site (NNSS), an active DOE Regional Disposal Site. The PSMC's would be the first place DOE would go to meet their radioactive packaging needs and the primary locations projects would go to disposition excess packaging for beneficial reuse. These two hubs would provide the centralized management of a packaging fleet rather than the current approach to design, procure, maintain and dispose of packagings on a project-by-project basis. This initiative provides significant savings in packaging costs and acceleration of project schedules. In addition to certified packaging, the PSMC would be well suited for select designs of 7A Type A packaging and Industrial Packaging. (authors)« less

Selection of Russian Plutonium Beryllium Sources for Inclusion in the Nuclear Mateirals Information Program Archive

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

Narlesky, Joshua E; Padilla, Dennis D; Watts, Joe

2009-01-01

Throughout the 1960s and 1970s, the former Soviet Union produced and exported Plutonium-Beryllium (PuBe) neutron sources to various Eastern European countries. The Russian sources consist of an intermetallic compound of plutonium and beryllium encapsulated in an inner welded, sealed capsule and consisting of a body and one or more covers. The amount of plutonium in the sources ranges from 0.002 g up to 15 g. A portion of the sources was originally exported to East Germany. A portion of these sources were acquired by Los Alamos National Laboratory (LANL) in the late 1990s for destruction in the Offsite Source Recoverymore » Program. When the OSRP was canceled, the remaining 88 PuBe neutron sources were packaged and stored in a 55-gal drum at T A-55. This storage configuration is no longer acceptable for PuBe sources, and the sources must either be repackaged or disposed of. Repackaging would place the sources into Hagan container, and depending on the dose rates, some sources may be packaged individually increasing the footprint and cost of storage. In addition, each source will be subject to leak-checking every six months. Leaks have already been detected in some of the sources, and due to the age of these sources, it is likely that additional leaks may be detected over time, which will increase the overall complexity of handling and storage. Therefore, it was decided that the sources would be disposed of at the Waste Isolation Pilot Plant (WIPP) due to the cost and labor associated with continued storage at TA-55. However, the plutonium in the sources is of Russian origin and needs to be preserved for research purposes. Therefore, it is important that a representative sample of the sources retained and archived for future studies. This report describes the criteria used to obtain a representative sample of the sources. Nine Russian PuBe neutron sources have been selected out of a collection of 77 sources for inclusion in the NMIP archive. Selection criteria were developed so that the largest sources that are representative of the collection are included. One representative source was chosen for every 20 sources in the collection, and effort was made to preserve sources unique to the collection. In total, four representative sources and five unique sources were selected for the archive. The archive samples contain 40 grams of plutonium with an isotopic composition similar to that of weapon grade material and three grams of plutonium with an isotopic composition similar to that of reactor grade plutonium.« less

Nightshade Prototype Experiments (Silverleaf)

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

Danielson, Jeremy; Bauer, Amy L.

2016-12-23

The Red Sage campaign is a series of subcritical dynamic plutonium experiments designed to measure ejecta. Nightshade, the first experiments in Red Sage scheduled for fiscal year 2019, will measure the amount of ejecta emission into vacuum from a double-­shocked plutonium surface. To address the major technical risks in Nightshade, a Level 2 milestone was developed for fiscal year 2016. Silverleaf, a series of four experiments, was executed at the Los Alamos National Laboratory in July and August 2016 to demonstrate a prototype of the Nightshade package and to satisfy this Level 2 milestone. This report is documentation that Redmore » Sage Level 2 milestone requirements were successfully met.« less

Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Huber, Patrick

2017-09-01

We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

DISPOSITION PATHS FOR ROCKY FLATS GLOVEBOXES: EVALUATING OPTIONS

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

Lobdell, D.; Geimer, R.; Larsen, P.

2003-02-27

The Kaiser-Hill Company, LLC has the responsibility for closure activities at the Rocky Flats Environmental Technology Site (RFETS). One of the challenges faced for closure is the disposition of radiologically contaminated gloveboxes. Evaluation of the disposition options for gloveboxes included a detailed analysis of available treatment capabilities, disposal facilities, and lifecycle costs. The Kaiser-Hill Company, LLC followed several processes in determining how the gloveboxes would be managed for disposition. Currently, multiple disposition paths have been chosen to accommodate the needs of the varying styles and conditions of the gloveboxes, meet the needs of the decommissioning team, and to best managemore » lifecycle costs. Several challenges associated with developing a disposition path that addresses both the radiological and RCRA concerns as well as offering the most cost-effective solution were encountered. These challenges included meeting the radiological waste acceptance criteria of available disposal facilities, making a RCRA determination, evaluating treatment options and costs, addressing void requirements associated with disposal, and identifying packaging and transportation options. The varying disposal facility requirements affected disposition choices. Facility conditions that impacted decisions included radiological and chemical waste acceptance criteria, physical requirements, and measurement for payment options. The facility requirements also impacted onsite activities including management strategies, decontamination activities, and life-cycle cost.« less

Contribution of water vapor pressure to pressurization of plutonium dioxide storage containers

NASA Astrophysics Data System (ADS)

Veirs, D. Kirk; Morris, John S.; Spearing, Dane R.

2000-07-01

Pressurization of long-term storage containers filled with materials meeting the US DOE storage standard is of concern.1,2 For example, temperatures within storage containers packaged according to the standard and contained in 9975 shipping packages that are stored in full view of the sun can reach internal temperatures of 250 °C.3 Twenty five grams of water (0.5 wt.%) at 250 °C in the storage container with no other material present would result in a pressure of 412 psia, which is limited by the amount of water. The pressure due to the water can be substantially reduced due to interactions with the stored material. Studies of the adsorption of water by PuO2 and surface interactions of water with PuO2 show that adsorption of 0.5 wt.% of water is feasible under many conditions and probable under high humidity conditions.4,5,6 However, no data are available on the vapor pressure of water over plutonium dioxide containing materials that have been exposed to water.

Amarillo National Resource Center for plutonium. Work plan progress report, November 1, 1995--January 31, 1996

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

Cluff, D.

1996-04-01

The Center operates under a cooperative agreement between DOE and the State of Texas and is directed and administered by an education consortium. Its programs include developing peaceful uses for the materials removed from dismantled weapons, studying effects of nuclear materials on environment and public health, remedying contaminated soils and water, studying storage, disposition, and transport of Pu, HE, and other hazardous materials removed from weapons, providing research and counsel to US in carrying out weapons reductions in cooperation with Russia, and conducting a variety of education and training programs.

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.

Development of Metallic Magnetic Calorimeters for Nuclear Safeguards Applications

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

Bates, Cameron Russell

2015-03-11

Many nuclear safeguards applications could benefit from high-resolution gamma-ray spectroscopy achievable with metallic magnetic calorimeters. This dissertation covers the development of a system for these applications based on gamma-ray detectors developed at the University of Heidelberg. It demonstrates new calorimeters of this type, which achieved an energy resolution of 45.5 eV full-width at half-maximum at 59.54 keV, roughly ten times better than current state of the art high purity germanium detectors. This is the best energy resolution achieved with a gamma-ray metallic magnetic calorimeter at this energy to date. In addition to demonstrating a new benchmark in energy resolution, anmore » experimental system for measuring samples with metallic magnetic calorimeters was constructed at Lawrence Livermore National Laboratory. This system achieved an energy resolution of 91.3 eV full-width at half-maximum at 59.54 keV under optimal conditions. Using this system it was possible to characterize the linearity of the response, the count-rate limitations, and the energy resolution as a function of temperature of the new calorimeter. With this characterization it was determined that it would be feasible to measure 242Pu in a mixed isotope plutonium sample. A measurement of a mixed isotope plutonium sample was performed over the course of 12 days with a single two-pixel metallic magnetic calorimeter. The relative concentration of 242Pu in comparison to other plutonium isotopes was determined by direct measurement to less than half a percent accuracy. This is comparable with the accuracy of the best-case scenario using traditional indirect methods. The ability to directly measure the relative concentration of 242Pu in a sample could enable more accurate accounting and detection of indications of undeclared activities in nuclear safeguards, a better constraint on source material in forensic samples containing plutonium, and improvements in verification in a future plutonium disposition treaty.« less

Used fuel disposition in crystalline rocks

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

Wang, Y.; Hadgu, Teklu; Kalinina, Elena Arkadievna

The U.S. Department of Energy Office of Nuclear Energy, Office of Fuel Cycle Technology established the Used Fuel Disposition Campaign (UFDC) in fiscal year 2010 (FY10) to conduct the research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel and high level nuclear waste. The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media.

10 CFR 71.64 - Special requirements for plutonium air shipments.

Code of Federal Regulations, 2012 CFR

2012-01-01

... requirements of §§ 71.41 through 71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in— (1) Section 71.74 (“Accident conditions for air transport of plutonium”)— (i) The containment vessel would not be ruptured in its post-tested condition, and the package...

10 CFR 71.64 - Special requirements for plutonium air shipments.

Code of Federal Regulations, 2013 CFR

2013-01-01

... requirements of §§ 71.41 through 71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in— (1) Section 71.74 (“Accident conditions for air transport of plutonium”)— (i) The containment vessel would not be ruptured in its post-tested condition, and the package...

10 CFR 71.64 - Special requirements for plutonium air shipments.

Code of Federal Regulations, 2010 CFR

2010-01-01

... requirements of §§ 71.41 through 71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in— (1) Section 71.74 (“Accident conditions for air transport of plutonium”)— (i) The containment vessel would not be ruptured in its post-tested condition, and the package...

10 CFR 71.64 - Special requirements for plutonium air shipments.

Code of Federal Regulations, 2014 CFR

2014-01-01

... requirements of §§ 71.41 through 71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in— (1) Section 71.74 (“Accident conditions for air transport of plutonium”)— (i) The containment vessel would not be ruptured in its post-tested condition, and the package...

Anisotropic expansion of Pu through the α-β-γ phase transitions while under radial compressive stress

NASA Astrophysics Data System (ADS)

Spearing, Dane R.; Veirs, D. Kirk; Prenger, F. Coyne

2000-07-01

In support of the new US Department of Energy (DOE) standard for the stabilization, packaging, and storage of plutonium bearing materials, the effects of the volume expansion associated with the α/β and β/γ transformations of Pu metal on the integrity of a stainless steel storage container were examined.

Improvement of calculation method for electrical parameters of short network of ore-thermal furnaces

NASA Astrophysics Data System (ADS)

Aliferov, A. I.; Bikeev, R. A.; Goreva, L. P.

2017-10-01

The paper describes a new calculation method for active and inductive resistance of split interleaved current leads packages in ore-thermal electric furnaces. The method is developed on basis of regression analysis of dependencies of active and inductive resistances of the packages on their geometrical parameters, mutual disposition and interleaving pattern. These multi-parametric calculations have been performed with ANSYS software. The proposed method allows solving split current lead electrical parameters minimization and balancing problems for ore-thermal furnaces.

Device for providing high-intensity ion or electron beam

DOEpatents

McClanahan, Edwin D.; Moss, Ronald W.

1977-01-01

A thin film of a low-thermionic-work-function material is maintained on the cathode of a device for producing a high-current, low-pressure gas discharge by means of sputter deposition from an auxiliary electrode. The auxiliary electrode includes a surface with a low-work-function material, such as thorium, uranium, plutonium or one of the rare earth elements, facing the cathode but at a disposition and electrical potential so as to extract ions from the gas discharge and sputter the low-work-function material onto the cathode. By continuously replenishing the cathode film, high thermionic emissions and ion plasmas can be realized and maintained over extended operating periods.

Hydrothermal Synthesis and Crystal Structures of Actinide Compounds

NASA Astrophysics Data System (ADS)

Runde, Wolfgang; Neu, Mary P.

Since the 1950s actinides have been used to benefit industry, science, health, and national security. The largest industrial application, electricity generation from uranium and thorium fuels, is growing worldwide. Thus, more actinides are being mined, produced, used and processed than ever before. The future of nuclear energy hinges on how these increasing amounts of actinides are contained in each stage of the fuel cycle, including disposition. In addition, uranium and plutonium were built up during the Cold War between the United States and the Former Soviet Union for defense purposes and nuclear energy. These stockpiles have been significantly reduced in the last decade.

Nuclear waste forms for actinides

PubMed Central

Ewing, Rodney C.

1999-01-01

The disposition of actinides, most recently 239Pu from dismantled nuclear weapons, requires effective containment of waste generated by the nuclear fuel cycle. Because actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on risk assessments of geologic repositories. Thus, demonstrable, long-term chemical and mechanical durability are essential properties of waste forms for the immobilization of actinides. Mineralogic and geologic studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely materials science approach. The “mineralogic approach” is illustrated by a discussion of zircon as a phase for the immobilization of excess weapons plutonium. PMID:10097054

Regulatory fire test requirements for plutonium air transport packages : JP-4 or JP-5 vs. JP-8 aviation fuel.

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

Figueroa, Victor G.; Lopez, Carlos; Nicolette, Vernon F.

2010-10-01

For certification, packages used for the transportation of plutonium by air must survive the hypothetical thermal environment specified in 10CFR71.74(a)(5). This regulation specifies that 'the package must be exposed to luminous flames from a pool fire of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes.' This regulation was developed when jet propellant (JP) 4 and 5 were the standard jet fuels. However, JP-4 and JP-5 currently are of limited availability in the United States of America. JP-4 is very hard to obtain as it is not used much anymore. JP-5 may be easier to getmore » than JP-4, but only through a military supplier. The purpose of this paper is to illustrate that readily-available JP-8 fuel is a possible substitute for the aforementioned certification test. Comparisons between the properties of the three fuels are given. Results from computer simulations that compared large JP-4 to JP-8 pool fires using Sandia's VULCAN fire model are shown and discussed. Additionally, the Container Analysis Fire (CAFE) code was used to compare the thermal response of a large calorimeter exposed to engulfing fires fueled by these three jet propellants. The paper then recommends JP-8 as an alternate fuel that complies with the thermal environment implied in 10CFR71.74.« less

7 CFR 1124.10 - Producer-handler.

Code of Federal Regulations, 2012 CFR

2012-01-01

... and Orders; Milk), DEPARTMENT OF AGRICULTURE MILK IN THE PACIFIC NORTHWEST MARKETING AREA Order... which total route disposition and packaged sales of fluid milk products to other plants during the month... (b)(1) of this section necessary to produce all Class I milk handled (excluding receipts from...

7 CFR 1124.10 - Producer-handler.

Code of Federal Regulations, 2014 CFR

2014-01-01

... AND ORDERS; MILK), DEPARTMENT OF AGRICULTURE MILK IN THE PACIFIC NORTHWEST MARKETING AREA Order... which total route disposition and packaged sales of fluid milk products to other plants during the month... (b)(1) of this section necessary to produce all Class I milk handled (excluding receipts from...

7 CFR 1124.10 - Producer-handler.

Code of Federal Regulations, 2013 CFR

2013-01-01

... AND ORDERS; MILK), DEPARTMENT OF AGRICULTURE MILK IN THE PACIFIC NORTHWEST MARKETING AREA Order... which total route disposition and packaged sales of fluid milk products to other plants during the month... (b)(1) of this section necessary to produce all Class I milk handled (excluding receipts from...

7 CFR 1131.10 - Producer-handler.

Code of Federal Regulations, 2012 CFR

2012-01-01

... and Orders; Milk), DEPARTMENT OF AGRICULTURE MILK IN THE ARIZONA MARKETING AREA Order Regulating... route disposition and packaged sales of fluid milk products to other plants during the month does not... (b)(1) of this section necessary to produce all Class I milk handled (excluding receipts from...

7 CFR 1131.10 - Producer-handler.

Code of Federal Regulations, 2014 CFR

2014-01-01

... AND ORDERS; MILK), DEPARTMENT OF AGRICULTURE MILK IN THE ARIZONA MARKETING AREA Order Regulating... route disposition and packaged sales of fluid milk products to other plants during the month does not... (b)(1) of this section necessary to produce all Class I milk handled (excluding receipts from...

7 CFR 1131.10 - Producer-handler.

Code of Federal Regulations, 2013 CFR

2013-01-01

... AND ORDERS; MILK), DEPARTMENT OF AGRICULTURE MILK IN THE ARIZONA MARKETING AREA Order Regulating... route disposition and packaged sales of fluid milk products to other plants during the month does not... (b)(1) of this section necessary to produce all Class I milk handled (excluding receipts from...

Crystalline and Crystalline International Disposal Activities

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

Viswanathan, Hari S.; Chu, Shaoping; Reimus, Paul William

2015-12-21

This report presents the results of work conducted between September 2014 and July 2015 at Los Alamos National Laboratory in the crystalline disposal and crystalline international disposal work packages of the Used Fuel Disposition Campaign (UFDC) for DOE-NE’s Fuel Cycle Research and Development program.

24 CFR 291.301 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-04-01

... DEVELOPMENT HUD-OWNED PROPERTIES DISPOSITION OF HUD-ACQUIRED SINGLE FAMILY PROPERTY Sale of HUD-Held Single... apply: Bid package means the documents prepared for bidders in a mortgage loan sale, and includes the following: An Executive Summary containing information on FHA single family mortgage loan sales and...

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

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

Remote-Reading Safety and Safeguards Surveillance System for 3013 Containers

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

Lechelt, W. M.; Skorpik, J. R.; Silvers, K. L.

2002-02-26

At Hanford's Plutonium Finishing Plant (PFP), plutonium oxide is being loaded into stainless steel containers for long-term storage on the Hanford Site. These containers consist of two weld-sealed stainless steel cylinders nested one within the other. A third container holds the plutonium within the inner cylinder. This design meets the U.S. Department of Energy (DOE) storage standard, DOE-STD- 3013-2000, which anticipates a 50-year storage lifetime. The 3013 standard also requires a container surveillance program to continuously monitor pressure and to assure safeguards are adequate. However, the configuration of the container system makes using conventional measurement and monitoring methods difficult. Tomore » better meet the 3013 monitoring requirements, a team from Fluor Hanford (who manages the PFP), Pacific Northwest National Laboratory (PNNL), and Vista Engineering Technologies, LLC, developed a safer, cost-efficient, remote PFP 3013 container surveillance system. This new surveillance system is a combination of two successfully deployed technologies: (1) a magnetically coupled pressure gauge developed by Vista Engineering and (2) a radio frequency (RF) tagging device developed by PNNL. This system provides continuous, 100% monitoring of critical parameters with the containers in place, as well as inventory controls. The 3013 container surveillance system consists of three main elements: (1) an internal magnetic pressure sensor package, (2) an instrument pod (external electronics package), and (3) a data acquisition storage and display computer. The surveillance system described in this paper has many benefits for PFP and DOE in terms of cost savings and reduced personnel exposure. In addition, continuous safety monitoring (i.e., internal container pressure and temperature) of every container is responsible nuclear material stewardship and fully meets and exceeds DOE's Integrated Surveillance Program requirements.« less

21 CFR 111.535 - Under this subpart N, what records must you make and keep?

Code of Federal Regulations, 2014 CFR

2014-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Returned Dietary Supplements... material review and disposition decision on a returned dietary supplement; (3) The results of any testing... dietary supplement meets product specifications established in accordance with § 111.70(e). ...

21 CFR 111.535 - Under this subpart N, what records must you make and keep?

Code of Federal Regulations, 2013 CFR

2013-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Returned Dietary Supplements... material review and disposition decision on a returned dietary supplement; (3) The results of any testing... dietary supplement meets product specifications established in accordance with § 111.70(e). ...

21 CFR 111.535 - Under this subpart N, what records must you make and keep?

Code of Federal Regulations, 2012 CFR

2012-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Returned Dietary Supplements... material review and disposition decision on a returned dietary supplement; (3) The results of any testing... dietary supplement meets product specifications established in accordance with § 111.70(e). ...

21 CFR 111.535 - Under this subpart N, what records must you make and keep?

Code of Federal Regulations, 2011 CFR

2011-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Returned Dietary Supplements... material review and disposition decision on a returned dietary supplement; (3) The results of any testing... dietary supplement meets product specifications established in accordance with § 111.70(e). ...

21 CFR 111.535 - Under this subpart N, what records must you make and keep?

Code of Federal Regulations, 2010 CFR

2010-04-01

... MANUFACTURING, PACKAGING, LABELING, OR HOLDING OPERATIONS FOR DIETARY SUPPLEMENTS Returned Dietary Supplements... material review and disposition decision on a returned dietary supplement; (3) The results of any testing... dietary supplement meets product specifications established in accordance with § 111.70(e). ...

27 CFR 31.203 - Possession of used liquor bottles.

Code of Federal Regulations, 2014 CFR

2014-04-01

... emptied or elsewhere, including disposition for purposes that will result in the bottles being rendered... collectors' items or for other purposes not involving the packaging of any product for sale; (b) Any person may possess, offer for sale, or sell unusual or distinctive bottles for purposes not involving the...

27 CFR 31.203 - Possession of used liquor bottles.

Code of Federal Regulations, 2010 CFR

2010-04-01

... emptied or elsewhere, including disposition for purposes that will result in the bottles being rendered... collectors' items or for other purposes not involving the packaging of any product for sale; (b) Any person may possess, offer for sale, or sell unusual or distinctive bottles for purposes not involving the...

27 CFR 31.203 - Possession of used liquor bottles.

Code of Federal Regulations, 2013 CFR

2013-04-01

... emptied or elsewhere, including disposition for purposes that will result in the bottles being rendered... collectors' items or for other purposes not involving the packaging of any product for sale; (b) Any person may possess, offer for sale, or sell unusual or distinctive bottles for purposes not involving the...

27 CFR 31.203 - Possession of used liquor bottles.

Code of Federal Regulations, 2012 CFR

2012-04-01

... emptied or elsewhere, including disposition for purposes that will result in the bottles being rendered... collectors' items or for other purposes not involving the packaging of any product for sale; (b) Any person may possess, offer for sale, or sell unusual or distinctive bottles for purposes not involving the...

27 CFR 31.203 - Possession of used liquor bottles.

Code of Federal Regulations, 2011 CFR

2011-04-01

... emptied or elsewhere, including disposition for purposes that will result in the bottles being rendered... collectors' items or for other purposes not involving the packaging of any product for sale; (b) Any person may possess, offer for sale, or sell unusual or distinctive bottles for purposes not involving the...

Some less conventional options for plutonium disposal

NASA Astrophysics Data System (ADS)

Stoll, Dr. Wolfgang, Prof.

2000-07-01

Disposition of weapons Pu (W-Pu) aims at the replacement of military access restrictions by inherent longlasting technical barriers to make the return into the weapons state difficult and not rewarding anymore. At the time of the NAS-study in 1994, two ways were perceived to be mature and selected: Fissioning of W-Pu as LWR-MOX and the disposal in a vitrified radionuclide-spiked form.1 Both options since have been questioned for equality, met different acceptance at both superpowers and showed slow progress. A criterion to measure disarmament would be the amount of W-Pu in the different proliferation resistant forms, multiplied by the effort needed for each form to return to weapons quality.

Preparation and Characterization of a Master Blend of Plutonium Oxide for the 3013 Large Scale Shelf-Life Surveillance Project

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

Gillispie, Obie William; Worl, Laura Ann; Veirs, Douglas Kirk

A mixture of chlorine-containing, impure plutonium oxides has been produced and has been given the name Master Blend. This large quantity of well-characterized chlorinecontaining material is available for use in the Integrated Surveillance and Monitoring Program for shelf-life experiments. It is intended to be representative of materials packaged to meet DOE-STD-3013.1 The Master Blend contains a mixture of items produced in Los Alamos National Laboratory’s (LANL) electro-refining pyrochemical process in the late 1990s. Twenty items were crushed and sieved, calcined to 800ºC for four hours, and blended multiple times. This process resulted in four batches of Master Blend. Calorimetry andmore » density data on material from the four batches indicate homogeneity.« less

Surveillance and Monitoring Program Full-Scale Experiments to Evaluate the Potential for Corrosion in 3013 Containers

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

Narlesky, Joshua Edward; Berg, John M.; Duque, Juan

A set of six long-term, full-scale experiments were initiated to determine the type and extent of corrosion that occurs in 3013 containers packaged with chloride-bearing plutonium oxide materials. The materials were exposed to a high relative humidity environment representative of actual packaging conditions for the materials in storage. The materials were sealed in instrumented, inner 3013 containers with corrosion specimens designed to test the corrosiveness of the environment inside the containers under various conditions. This report focuses on initial loading conditions that are used to establish a baseline to show how the conditions change throughout the storage lifetime of themore » containers.« less

Destructive examination of shipping package 9975-02644

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

Daugherty, W. L.

Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02644 as part of a comprehensive SRS surveillance program for plutonium material stored in the K-Area Complex (KAC). During the field surveillance inspection of this package in KAC, three non-conforming conditions were noted: the axial gap of 1.389 inch exceeded the 1 inch maximum criterion, the exposed height of the lead shield was greater than the 4.65 inch maximum criterion, and the difference between the upper assembly inside height and the exposed height of the lead shield was less than the 0.425 inch minimum criterion. All threemore » of these observations relate to axial shrinkage of the lower fiberboard assembly. In addition, liquid water (condensation) was observed on the interior of the drum lid, the thermal blanket and the air shield.« less

Assessment of Residual Stresses in 3013 Inner and Outer Containers and Teardrop Samples

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

Stroud, Mary Ann; Prime, Michael Bruce; Veirs, Douglas Kirk

2015-12-08

This report is an assessment performed by LANL that examines packaging for plutonium-bearing materials and the resilience of its design. This report discusses residual stresses in the 3013 outer, the SRS/Hanford and RFETS/LLNL inner containers, and teardrop samples used in studies to assess the potential for SCC in 3013 containers. Residual tensile stresses in the heat affected zones of the closure welds are of particular concern.

Used Fuel Disposal in Crystalline Rocks. FY15 Progress Report

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

Wang, Yifeng

2015-08-20

The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media. Chapter headings are as follows: Fuel matrix degradation model and its integration with performance assessments, Investigation of thermal effects on the chemical behavior of clays, Investigation of uranium diffusion and retardation in bentonite, Long-term diffusion of U(VI) in bentonite: dependence on density, Sorption and desorption of plutonium by bentonite, Dissolution of plutonium intrinsic colloids in the presence of clay and asmore » a function of temperature, Laboratory investigation of colloid-facilitated transport of cesium by bentonite colloids in a crystalline rock system, Development and demonstration of discrete fracture network model, Fracture continuum model and its comparison with discrete fracture network model.« less

Johnston Atoll Plutonium Cleanup Project. Contract bridge report. Technical report, 1 June-5 August 1993

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

Hellier, C.L.; Doane, R.W.

1995-04-01

This report continues the documentation of the operation of TMA/Eberline`s Segmented Gate System technology for removing mixed plutonium and americium contamination at DNA`s Johnston Atoll site. Contaminated feed is conveyed under arrays of radiation detectors coupled with sophisticated computer software developed by Eberline Instrument Corporation. Segmented gates (chutes) on pneumatically-driven pistons move forward when contamination is detected to remove only the contaminated portion from the main flow of feed material. Only about one pint of contaminant is removed during each diversion event. At the JA site, a 98% volume reduction has been achieved, with the remediated soil cleaned to DNA`smore » criteria for release for unrestricted use of 500 Bq/kg total transuranic alpha contamination and no hot particles of greater than 5000 Becquerrels. The low level waste concentrate is expected to be packaged for shipment to an approved defense waste disposal site.« less

A Comparison Framework for Reactor Anti-Neutrino Detectors in Near-Field Nuclear Safeguards Applications

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

Mendenhall, M.; Bowden, N.; Brodsky, J.

Electron anti-neutrino ( e) detectors can support nuclear safeguards, from reactor monitoring to spent fuel characterization. In recent years, the scientific community has developed multiple detector concepts, many of which have been prototyped or deployed for specific measurements by their respective collaborations. However, the diversity of technical approaches, deployment conditions, and analysis techniques complicates direct performance comparison between designs. We have begun development of a simulation framework to compare and evaluate existing and proposed detector designs for nonproliferation applications in a uniform manner. This report demonstrates the intent and capabilities of the framework by evaluating four detector design concepts, calculatingmore » generic reactor antineutrino counting sensitivity, and capabilities in a plutonium disposition application example.« less

Summary of International Waste Management Programs (LLNL Input to SNL L3 MS: System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW)

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

Greenberg, Harris R.; Blink, James A.; Halsey, William G.

2011-08-11

The Used Fuel Disposition Campaign (UFDC) within the Department of Energy’s Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation’s spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. This Lessons Learned task is part of a multi-laboratory effort, with this LLNL report providing input to a Level 3 SNL milestone (System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW). The work package number is: FTLL11UF0328; the work package title is: Technical Bases / Lessons Learned;more » the milestone number is: M41UF032802; and the milestone title is: “LLNL Input to SNL L3 MS: System-Wide Integration and Site Selection Concepts for Future Disposition Options for HLW”. The system-wide integration effort will integrate all aspects of waste management and disposal, integrating the waste generators, interim storage, transportation, and ultimate disposal at a repository site. The review of international experience in these areas is required to support future studies that address all of these components in an integrated manner. Note that this report is a snapshot of nuclear power infrastructure and international waste management programs that is current as of August 2011, with one notable exception. No attempt has been made to discuss the currently evolving world-wide response to the tragic consequences of the earthquake and tsunami that devastated Japan on March 11, 2011, leaving more than 15,000 people dead and more than 8,000 people missing, and severely damaging the Fukushima Daiichi nuclear power complex. Continuing efforts in FY 2012 will update the data, and summarize it in an Excel spreadsheet for easy comparison and assist in the knowledge management of the study cases.« less

Safety evaluation for packaging transportation of equipment for tank 241-C-106 waste sluicing system

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

Calmus, D.B.

1994-08-25

A Waste Sluicing System (WSS) is scheduled for installation in nd waste storage tank 241-C-106 (106-C). The WSS will transfer high rating sludge from single shell tank 106-C to double shell waste tank 241-AY-102 (102-AY). Prior to installation of the WSS, a heel pump and a transfer pump will be removed from tank 106-C and an agitator pump will be removed from tank 102-AY. Special flexible receivers will be used to contain the pumps during removal from the tanks. After equipment removal, the flexible receivers will be placed in separate containers (packagings). The packaging and contents (packages) will be transferredmore » from the Tank Farms to the Central Waste Complex (CWC) for interim storage and then to T Plant for evaluation and processing for final disposition. Two sizes of packagings will be provided for transferring the equipment from the Tank Farms to the interim storage facility. The packagings will be designated as the WSSP-1 and WSSP-2 packagings throughout the remainder of this Safety Evaluation for Packaging (SEP). The WSSP-1 packagings will transport the heel and transfer pumps from 106-C and the WSSP-2 packaging will transport the agitator pump from 102-AY. The WSSP-1 and WSSP-2 packagings are similar except for the length.« less

PU/SS EUTECTIC ASSESSMENT IN 9975 PACKAGINGS IN A STORAGE FACILITY DURING EXTENDED FIRE

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

Gupta, N.

2012-03-26

In a radioactive material (RAM) packaging, the formation of eutectic at the Pu/SS (plutonium/stainless steel) interface is a serious concern and must be avoided to prevent of leakage of fissile material to the environment. The eutectic temperature for the Pu/SS is rather low (410 C) and could seriously impact the structural integrity of the containment vessel under accident conditions involving fire. The 9975 packaging is used for long term storage of Pu bearing materials in the DOE complex where the Pu comes in contact with the stainless steel containment vessel. Due to the serious consequences of the containment breach atmore » the eutectic site, the Pu/SS interface temperature is kept well below the eutectic formation temperature of 410 C. This paper discusses the thermal models and the results for the extended fire conditions (1500 F for 86 minutes) that exist in a long term storage facility and concludes that the 9975 packaging Pu/SS interface temperature is well below the eutectic temperature.« less

Procurement of a fully licensed radioisotope thermoelectric generator transportation system

NASA Astrophysics Data System (ADS)

Adkins, Harold E.; Bearden, Thomas E.

The present transportation system for radioisotope thermoelectric generators and heater units is being developed to comply with all applicable U.S. DOT regulations, including a doubly-contained 'bell jar' concept for the required double-containment of plutonium. Modifications in handling equipment and procedures are entailed by this novel packaging design, and will affect high-capacity forklifts, overhead cranes, He-backfilling equipment, etc. Attention is given to the design constraints involved, and to the Federal procurement process.

Temperature environment for 9975 packages stored in KAC

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

Daugherty, W. L.

Plutonium materials are stored in the K Area Complex (KAC) in shipping packages, typically the 9975 shipping package. In order to estimate realistic degradation rates for components within the shipping package (i.e. the fiberboard overpack and O-ring seals), it is necessary to understand actual facility temperatures, which can vary daily and seasonally. Relevant facility temperature data available from several periods throughout its operating history have been reviewed. The annual average temperature within the Crane Maintenance Area has ranged from approximately 70 to 74 °F, although there is significant seasonal variation and lesser variation among different locations within the facility. Themore » long-term average degradation rate for 9975 package components is very close to that expected if the component were to remain continually at the annual average temperature. This result remains valid for a wide range of activation energies (which describes the variation in degradation rate as the temperature changes), if the activation energy remains constant over the seasonal range of component temperatures. It is recommended that component degradation analyses and service life estimates incorporate these results. Specifically, it is proposed that future analyses assume an average facility ambient air temperature of 94 °F. This value is bounding for all packages, and includes margin for several factors such as increased temperatures within the storage arrays, the addition of more packages in the future, and future operational changes.« less

Nuclear waste disposal—pyrochlore (A2B2O7): Nuclear waste form for the immobilization of plutonium and "minor" actinides

NASA Astrophysics Data System (ADS)

Ewing, Rodney C.; Weber, William J.; Lian, Jie

2004-06-01

During the past half-century, the nuclear fuel cycle has generated approximately 1400 metric tons of plutonium and substantial quantities of the "minor" actinides, such as Np, Am, and Cm. The successful disposition of these actinides has an important impact on the strategy for developing advanced nuclear fuel cycles, weapons proliferation, and the geologic disposal of high-level radioactive waste. During the last decade, there has been substantial interest in the use of the isometric pyrochlore structure-type, A2B2O7, for the immobilization of actinides. Most of the interest has focused on titanate-pyrochlore because of its chemical durability; however, these compositions experience a radiation-induced transition from the crystalline-to-aperiodic state due to radiation damage from the alpha-decay of actinides. Depending on the actinide concentration, the titanate pyrochlore will become amorphous in less than 1000 years of storage. Recently, systematic ion beam irradiations of a variety of pyrochlore compositions has revealed that many zirconate pyrochlores do not become amorphous, but remain crystalline as a defect fluorite structure-type due to disordering of the A- and B-site cations. The zirconate pyrochlores will remain crystalline even to very high doses, greater than 100 displacements per atom. Systematic experimental studies of actinide-doped and ion beam-irradiated pyrochlore, analyses of natural U- and Th-bearing pyrochlore, and simulations of the energetics of the disordering process now provide a rather detailed understanding of the structural and chemical controls on the response of pyrochlore to radiation. These results provide a solid basis for predicting the behavior and durability of pyrochlore used to immobilize plutonium.

SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

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

NORTON SH

2010-02-23

CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material.more » The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford nuclear reservation. Workers had to pass through metal detectors when they arrived at the plant and materials leaving the plant had to be scanned for security reasons. Whereas other high-security nuclear materials were shipped from the PFP to Savannah River, S.C. as part ofa Department of Energy (DOE) program to consolidate weapons-grade plutonium, it was determined that the SIF should remain onsite pending disposition to a national repository. Nevertheless, the SIF still requires a high level of security that the PFP complex has always provided. With the 60-year PFP mission of producing and storing plutonium concluded, the environmental cleanup plans for Hanford call for the demolition of the 63-building PFP complex. Consequently, if the SIF remained at PFP it not only would have interfered with the environmental cleanup plans, but would have required $100 million in facility upgrades to meet increased national security requirements imposed after the 9/11 terrorist attacks. A new smaller and more cost-effective area was needed to store this material, which led to the SIF Project. Once the SIF project was successfully completed and the SIF was safely removed from PFP, the existing Protected Area at PFP could be removed, and demolition could proceed more quickly without being encumbered by restrictive security requirements that an active Protected Area requires. The lightened PFP security level brought by safely removing and storing the SIF would also yield lowered costs for deactivation and demolition, as well as reduce overall life-cycle costs.« less

Summary of FY 17 Assessments Sandia National Laboratories: Evaluation of FY16 SNL FCT M2 Milestone Deliverables

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

Appel, Gordon John

This report is the milestone deliverable M4FT-17SN111102091 “Summary of Assessments Performed FY17 by SNL QA POC” for work package FT-17SN11110209 titled “Quality Assurance – SNL”. This report summarizes the FY17 assessment performed on Fuel Cycle Technologies / Spent Fuel and Waste Disposition efforts.

Johnston Atoll Plutonium Cleanup Project, plant modification and operation. Volume 1. Annual report option year 2. Technical report, 1 October 1992-24 May 1993

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

Moroney, K.S.; Moroney, J.D.; Johnson, N.R.

1995-04-01

This report continues the documentation of the operation of TMA/Eberline`s Segmented Gate System technology for removing mixed plutonium and americium contamination at DNA`s Johnston Atoll site. Contaminated feed is conveyed under arrays of radiation detectors coupled with sophisticated computer software developed by Eberline Instrument Corporation. Segmented gates (chutes) on pneumatically-driven pistons move forward when contamination is detected to remove only the contaminated portion from the main flow of feed material. Only about one pint of contaminant is removed during each diversion event. At the JA site, a 98% volume reduction has been achieved, with the remediated soil cleaned to DNA`smore » criteria for release for unrestricted use of 500 Bq/kg total tnansuranic alpha contamination and no hot particles of greater than 5000 Becquerrels. The low level waste concentrate is expected to be packaged for shipment to an approved defense waste disposal site.« less

Johnston Atoll Plutonium Cleanup Project. Plant modification and operation. Volume 2. Annual report option year 2. Appendix A. Technical report, 1 October 1992-24 May 1993

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

Moroney, K.S.; Moroney, J.D.; Johnson, N.R.

1995-04-01

This report continues the documentation of the operation of TMA/Eberline`s Segmented Gate System technology for removing mixed plutonium and americium contamination at DNA`s Johnston Atoll site. Contaminated feed is conveyed under arrays of radiation detectors coupled with sophisticated computer software developed by Eberline Instrument Corporation. Segmented gates (chutes) on pneumatically-driven pistons move forward when contamination is detected to remove only the contaminated portion from the main flow of feed material. Only about one pint of contaminant is removed during each diversion event. At the JA site, a 98% volume reduction has been achieved, with the remediated soil cleaned to DNA`smore » criteria for release for unrestricted use of 500 Bq/kg total transuranic alpha contamination and no hot particles of greater than 5000 Becquerrels. The low level waste concentrate is expected to be packaged for shipment to an approved defense waste disposal site.« less

Johnston Atoll Plutonium Cleanup Project, plant modification and operation. Volume 3. Annual report option year 2. Appendices b through h. Technical report, 1 October 1992-24 May 1993

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

Moroney, K.S.; Moroney, J.D.; Johnson, N.R.

1995-04-01

This report continues the documentation of the operation of TMA/Eberline`s Segmented Gate System technology for removing mixed plutonium and americium contamination at DNA`s Johnston Atoll site. Contaminated feed is conveyed under arrays of radiation detectors coupled with sophisticated computer software developed by Eberline Instrument Corporation. Segmented gates (chutes) on pneumatically-driven pistons move forward when contamination is detected to remove only the contaminated portion from the main flow of feed material. Only about one pint of contaminant is removed during each diversion event. At the JA site, a 98% volume reduction has been achieved, with the remediated soil cleaned to DNA`smore » criteria for release for unrestricted use of 500 Bq/kg total transuranic alpha contamination and no hot particles of greater than 5000 Becquerrels. The low level waste concentrate is expected to be packaged for shipment to an approved defense waste disposal site.« less

Rupture loop annex ion exchange RLAIX vault deactivation

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

Ham, J.E.; Harris, D.L., Westinghouse Hanford

This engineering report documents the deactivation, stabilization and final conditions of the Rupture Loop Annex Ion Exchange (RLAIX) Vault located northwest of the 309 Building`s Plutonium Recycle Test Reactor (PRTR). Twelve ion exchange columns, piping debris, and column liquid were removed from the vault, packaged and shipped for disposal. The vault walls and floor were decontaminated, and portions of the vault were painted to fix loose contamination. Process piping and drains were plugged, and the cover blocks and rain cover were installed. Upon closure,the vault was empty, stabilized, isolated.

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

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

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

N /A

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign andmore » domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.« less

Thermal Analysis of the Mound One Kilowatt Package

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

Or, Chuen T.

The Mound One Kilowatt (1 KW) package was designed for the shipment of plutonium (Pu-238) with not more than 1 kW total heat dissipation. To comply with regulations, the Mound 1 kW package has to pass all the requirements under Normal Conditions of Transport (NCT; 38 degrees C ambient temperature) and Hypothetical Accident Conditions (HAC; package engulfed in fire for 30 minutes). Analytical and test results were presented in the Safety Analysis Report for Packaging (SARP) for the Mound 1 kW package, revision 1, April 1991. Some issues remained unresolved in that revision. In March 1992, Fairchild Space and Defensemore » Corporation was commissioned by the Department of Energy to perform the thermal analyses. 3-D thermal models were created to perform the NCT and HAC analyses. Four shipping configurations in the SARP revision 3 were analyzed. They were: (1) The GPHS graphite impact shell (GIS) in the threaded product can (1000 W total heat generation); (2) The fueled clads in the welded product can (1000 W total heat generation); (3) The General Purpose Heat Source (GPHS) module (750 W total heat generation); and (4) The Multi-Hundred Watt (MHW) spheres (810 W total heat generation). Results from the four cases show that the GIS or fuel clad in the product can is the worse case. The temperatures predicted under NCT and HAC in all four cases are within the design limits. The use of helium instead of argon as cover gas provides a bigger safety margin. There is a duplicate copy.« less

Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

NASA Astrophysics Data System (ADS)

Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

2018-01-01

In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

Used Fuel Disposition in Crystalline Rocks: FY16 Progress Report

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

Wang, Y.; Hadgu, Teklu; Kalinina, Elena Arkadievna

The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media. FY16 continued to be a successful year in both experimental and modeling arenas in evaluation of used fuel disposal in crystalline rocks. The work covers a wide range of research topics identified in the R&D plan.

DISSOLVED CONCENTRATION LIMITS OF RADIOACTIVE ELEMENTS

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

P. Bernot

The purpose of this study is to evaluate dissolved concentration limits (also referred to as solubility limits) of elements with radioactive isotopes under probable repository conditions, based on geochemical modeling calculations using geochemical modeling tools, thermodynamic databases, field measurements, and laboratory experiments. The scope of this activity is to predict dissolved concentrations or solubility limits for elements with radioactive isotopes (actinium, americium, carbon, cesium, iodine, lead, neptunium, plutonium, protactinium, radium, strontium, technetium, thorium, and uranium) relevant to calculated dose. Model outputs for uranium, plutonium, neptunium, thorium, americium, and protactinium are provided in the form of tabulated functions with pH andmore » log fCO{sub 2} as independent variables, plus one or more uncertainty terms. The solubility limits for the remaining elements are either in the form of distributions or single values. Even though selection of an appropriate set of radionuclides documented in Radionuclide Screening (BSC 2002 [DIRS 160059]) includes actinium, transport of Ac is not modeled in the total system performance assessment for the license application (TSPA-LA) model because of its extremely short half-life. Actinium dose is calculated in the TSPA-LA by assuming secular equilibrium with {sup 231}Pa (Section 6.10); therefore, Ac is not analyzed in this report. The output data from this report are fundamental inputs for TSPA-LA used to determine the estimated release of these elements from waste packages and the engineered barrier system. Consistent modeling approaches and environmental conditions were used to develop solubility models for the actinides discussed in this report. These models cover broad ranges of environmental conditions so they are applicable to both waste packages and the invert. Uncertainties from thermodynamic data, water chemistry, temperature variation, and activity coefficients have been quantified or otherwise addressed.« less

Spent Nuclear Fuel Disposition

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

Wagner, John C.

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

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

Daugherty, W. L.

Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02019 as part of a comprehensive SRS surveillance program for plutonium material stored in the K-Area Complex (KAC). During the field surveillance inspection of this package in KAC, two non-conforming conditions were noted: the axial gap of 1.577 inch exceeded the 1 inch maximum criterion, and two areas of dried glue residue were noted on the upper fiberboard subassembly. This package was subsequently transferred to SRNL for more detailed inspection and destructive examination. In addition to the conditions noted in KAC, the following conditions were noted: -more » Numerous small spots of corrosion were observed along the bottom edge of the drum. - In addition to the smeared glue residue on the upper fiberboard subassembly, there was also a small dark stain. - Mold was present on the side and bottom of the lower fiberboard subassembly. Dark stains from elevated moisture content were also present in these areas. - A dark spot with possible light corrosion was observed on the primary containment vessel flange, and corresponding rub marks were observed on the secondary containment vessel ID. - The fiberboard thermal conductivity in the radial orientation was above the specified range. When the test was repeated with slightly lower moisture content, the result was acceptable. The moisture content for both tests was within a range typical of other packages in storage. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement within the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.« less

Re-evaluation of Moisture Controls During ARIES Oxide Processing, Packaging and Characterization

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

Karmiol, Benjamin; Wayne, David Matthew

DOE-STD-3013 [1] requires limiting the relative humidity (RH) in the glovebox during processing of the oxide product for specific types of plutonium oxides. This requirement is mandated in order to limit corrosion of the stainless steel containers by deliquescence of chloride salts if present in the PuO2. DOE-STD-3013 also specifies the need to limit and monitor internal pressure buildup in the 3013 containers due to the potential for the generation of free H2 and O2 gas from the radiolysis of surfaceadsorbed water. DOE-STD-3013 requires that the oxide sample taken for moisture content verification be representative of the stabilized material inmore » the 3013 container. This is accomplished by either limiting the time between sampling and packaging, or by control of the glovebox relative humidity (%RH). This requirement ensures that the sample is not only representative, but also conservative from the standpoint of moisture content.« less

Enabling More than Moore: Accelerated Reliability Testing and Risk Analysis for Advanced Electronics Packaging

NASA Technical Reports Server (NTRS)

Ghaffarian, Reza; Evans, John W.

2014-01-01

For five decades, the semiconductor industry has distinguished itself by the rapid pace of improvement in miniaturization of electronics products-Moore's Law. Now, scaling hits a brick wall, a paradigm shift. The industry roadmaps recognized the scaling limitation and project that packaging technologies will meet further miniaturization needs or ak.a "More than Moore". This paper presents packaging technology trends and accelerated reliability testing methods currently being practiced. Then, it presents industry status on key advanced electronic packages, factors affecting accelerated solder joint reliability of area array packages, and IPC/JEDEC/Mil specifications for characterizations of assemblies under accelerated thermal and mechanical loading. Finally, it presents an examples demonstrating how Accelerated Testing and Analysis have been effectively employed in the development of complex spacecraft thereby reducing risk. Quantitative assessments necessarily involve the mathematics of probability and statistics. In addition, accelerated tests need to be designed which consider the desired risk posture and schedule for particular project. Such assessments relieve risks without imposing additional costs. and constraints that are not value added for a particular mission. Furthermore, in the course of development of complex systems, variances and defects will inevitably present themselves and require a decision concerning their disposition, necessitating quantitative assessments. In summary, this paper presents a comprehensive view point, from technology to systems, including the benefits and impact of accelerated testing in offsetting risk.

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.

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.

9975 SHIPPING PACKAGE LIFE EXTENSION SURVEILLANCE PROGRAM RESULTS SUMMARY

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

Daugherty, W.; Dunn, K.; Hackney, B.

2011-01-06

Results from the 9975 Surveillance Program at the Savannah River Site (SRS) are summarized for justification to extend the life of the 9975 packages currently stored in the K-Area Materials Storage (KAMS) facility from 10 years to 15 years. This justification is established with the stipulation that surveillance activities will continue throughout this extended time to ensure the continued integrity of the 9975 materials of construction and to further understand the currently identified degradation mechanisms. The current 10 year storage life was developed prior to storage. A subsequent report was later used to extend the qualification of the 9975 shippingmore » packages for 2 years for shipping plus 10 years for storage. However the qualification for the storage period was provided by the monitoring requirements of the Storage and Surveillance Program. This report summarizes efforts to determine a new safe storage limit for the 9975 shipping package based on the surveillance data collected since 2005 when the surveillance program began. KAMS is a zero-release facility that depends upon containment by the 9975 to meet design basis storage requirements. Therefore, to confirm the continued integrity of the 9975 packages while stored in KAMS, a 9975 Storage and Surveillance Program was implemented alongside the DOE required Integrated Surveillance Program (ISP) for 3013 plutonium-bearing containers. The 9975 Storage and Surveillance Program performs field surveillance as well as accelerated aging tests to ensure any degradation due to aging, to the extent that could affect packaging performance, is detected in advance of such degradation occurring in the field. The Program has demonstrated that the 9975 package has a robust design that can perform under a variety of conditions. As such the primary emphasis of the on-going 9975 Surveillance Program is an aging study of the 9975 Viton(reg.sign) GLT containment vessel O-rings and the Celotex(reg.sign) fiberboard thermal insulation at bounding conditions of radiation and elevated temperatures. Other materials of construction, however, are also discussed.« less

Disposal Of Irradiated Cadmium Control Rods From The Plumbrook Reactor Facility

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

Posivak, E.J.; Berger, S.R.; Freitag, A.A.

2008-07-01

Innovative mixed waste disposition from NASA's Plum Brook Reactor Facility was accomplished without costly repackaging. Irradiated characteristic hardware with contact dose rates as high as 8 Sv/hr was packaged in a HDPE overpack and stored in a Secure Environmental Container during earlier decommissioning efforts, awaiting identification of a suitable pathway. WMG obtained regulatory concurrence that the existing overpack would serve as the macro-encapsulant per 40CFR268.45 Table 1.C. The overpack vent was disabled and the overpack was placed in a stainless steel liner to satisfy overburden slumping requirements. The liner was sealed and placed in shielded shoring for transport to themore » disposal site in a US DOT Type A cask. Disposition via this innovative method avoided cost, risk, and dose associated with repackaging the high dose irradiated characteristic hardware. In conclusion: WMG accomplished what others said could not be done. Large D and D contractors advised NASA that the cadmium control rods could only be shipped to the proposed Yucca mountain repository. NASA management challenged MOTA to find a more realistic alternative. NASA and MOTA turned to WMG to develop a methodology to disposition the 'hot and nasty' waste that presumably had no path forward. Although WMG lead a team that accomplished the 'impossible', the project could not have been completed with out the patient, supportive management by DOE-EM, NASA, and MOTA. (authors)« less

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

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

G. Radulesscu; J.S. Tang

The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container alongmore » with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this analysis is to support Site Recommendation reports and to assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the Development Plan ''Design Analysis for the Defense High-Level Waste Disposal Container'' (CRWMS M&O 2000c) with no deviations from the plan.« 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.

LEVERAGING AGING MATERIALS DATA TO SUPPORT EXTENSION OF TRANSPORTATION SHIPPING PACKAGES SERVICE LIFE

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

Dunn, K.; Bellamy, S.; Daugherty, W.

Nuclear material inventories are increasingly being transferred to interim storage locations where they may reside for extended periods of time. Use of a shipping package to store nuclear materials after the transfer has become more common for a variety of reasons. Shipping packages are robust and have a qualified pedigree for performance in normal operation and accident conditions but are only certified over an approved transportation window. The continued use of shipping packages to contain nuclear material during interim storage will result in reduced overall costs and reduced exposure to workers. However, the shipping package materials of construction must maintainmore » integrity as specified by the safety basis of the storage facility throughout the storage period, which is typically well beyond the certified transportation window. In many ways, the certification processes required for interim storage of nuclear materials in shipping packages is similar to life extension programs required for dry cask storage systems for commercial nuclear fuels. The storage of spent nuclear fuel in dry cask storage systems is federally-regulated, and over 1500 individual dry casks have been in successful service up to 20 years in the US. The uncertainty in final disposition will likely require extended storage of this fuel well beyond initial license periods and perhaps multiple re-licenses may be needed. Thus, both the shipping packages and the dry cask storage systems require materials integrity assessments and assurance of continued satisfactory materials performance over times not considered in the original evaluation processes. Test programs for the shipping packages have been established to obtain aging data on materials of construction to demonstrate continued system integrity. The collective data may be coupled with similar data for the dry cask storage systems and used to support extending the service life of shipping packages in both transportation and storage.« less

Redox bias in loss of ignition moisture measurement for relatively pure plutonium-bearing oxide materials.

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

Eller, P. G.; Stakebake, J. L.; Cooper, T. D.

2001-01-01

This paper evaluates potential analytical bias in application of the Loss on Ignition (LOI) technique for moisture measurement to relatively pure (plutonium assay of 80 wt.% or higher) oxides containing uranium that have been stabilized according to stabilization and storage standard DOE-STD-3013-2000 (STD-3013). An immediate application is to Rocky Flats (RF) materials derived from highgrade metal hydriding separations subsequently treated by multiple calcination cycles. Specifically evaluated are weight changes due to oxidatiodreduction of multivalent impurity oxides that could mask true moisture equivalent content measurement. Process knowledge and characterization of materials representing complex-wide materials to be stabilized and packaged according tomore » STD-3013, and particularly for the immediate RF target stream, indicate that oxides of uranium, iron and gallium are the only potential multivalent constituents expected to be present above 0.5 wt.%. The evaluation shows that of these constituents, with few exceptions, only uranium oxides can be present at a sufficient level to produce weight gain biases significant with respect to the LO1 stability test. In general, these formerly high-value, high-actinide content materials are reliably identifiable by process knowledge and measurement. Si&icant bias also requires that UO1 components remain largely unoxidized after calcination and are largely converted to U30s clsning LO1 testing at only slightly higher temperatures. Based on wellestablished literature, it is judged unlikely that this set of conditions will be realized in practice. We conclude that it is very likely that LO1 weight gain bias will be small for the immediate target RF oxide materials containing greater than 80 wt.% plutonium plus a much smaller uranium content. Recommended tests are in progress to confum these expectations and to provide a more authoritative basis for bounding LO1 oxidatiodreduction biases. LO1 bias evaluation is more difficult for lower purity materials and for fuel-type uranium-plutonium oxides. However, even in these cases testing may show that bias effects are manageable.« less

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

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.

Effects of Laser Etching on the Corrosion Susceptibility of SAVY 4000 and Hagan Containers

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

Hyer, Holden Christopher; Duque, Juan; Smith, Paul Herrick

Since the late 1990’s, the Hagan container was used as the primary container for packaging of plutonium-bearing materials. The Hagan design consisted of a threaded closure, a Viton® ORing, a carbon-carbon filter, and a 304L stainless steel (SS) body. Over the years, Hagans have shown vulnerability in their design [1]. In 2008, The Department of Energy (DOE) issued DOE M 441.1-1, Nuclear Material Packaging Manual, which detailed an approach to obtain highconfidence in containers by including specific design requirements, material contents and an approach to determine life span from said contents, and surveillance techniques [2]. In response to both themore » vulnerability issues with the Hagan and DOE M 441.1-1, the SAVY 4000 container with its twist style closure, Viton® O-Ring, Fiberfrax-Gortex filter, and annealed 316L SS body, was designed as the replacement for Hagan containers, but only for a short term lifespan of 5 years [1]. However, both the Hagan and SAVY 4000 are being pushed to maintain a lifespan of 40 years. Therefore, proper confidence must be placed on each component of each container to last a minimum of 40 years. So far, the biggest concern found during surveillance of these containers is corrosion and the potential for failure by corrosion. One concern is that the containers fail due to stress corrosion cracking (SCC), especially around the weld between the collar and the body as welds leave residual stresses. One advantage the SAVY 4000 has is that the body is annealed, but its weld is still susceptible as it was welded after annealing [3, 4]. Moreover, 316L SS is known to have a higher pitting resistance (pits are a precursor to SCC and can also lead to extensive failure of the material), than 304L SS [4]. During recent surveillance activities, two SAVY 4000’s containing Solution Assay Instrument (SAI) solutions were opened. The SAI SAVY 4000’s contained plutonium (Pu) in 3M HCl solution in plastic volumetric flasks placed inside of polyethylene bags. Historically, a SAVY 4000 containing an SAI solution is packaged for 3 weeks, however, these particular containers were not reopened until 14 months later. When opened, brown-red corrosion product was found all over the inside of the container as shown in Figure 1.« less

Neutralization of Plutonium and Enriched Uranium Solutions Containing Gadolinium as a Neutron Poison

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

BRONIKOWSKI, MG.

2004-04-01

Materials currently being dissolved in the HB-Line Facility will result in an accumulated solution containing an estimated uranium:plutonium (U:Pu) ratio of 4.3:1 and an 235U enrichment estimated at 30 per cent The U:Pu ratio and the enrichment are outside the evaluated concentration range for disposition to high level waste (HLW) using gadolinium (Gd) as a neutron poison. To confirm that the solution generated during the current HB-Line dissolving campaign can be poisoned with Gd, neutralized and discarded to the Savannah River Site (SRS) high level waste (HLW) system without undue nuclear safety concerns the caustic precipitation of surrogate solutions wasmore » examined. Experiments were performed with a U/Pu/Gd solution representative of the HB-Line estimated concentration ratio and also a U/Gd solution. Depleted U was used in the experiments as the enrichment of the U will not affect the chemical behavior during neutralization, but will affect the amount of Gd added to the solution. Settling behavior of the neutralized solutions was found to be comparable to previous studies. The neutralized solutions mixed easily and had expected densities of typical neutralized waste. The neutralized solids were found to be homogeneous and less than 20 microns in size. Partially neutralized solids were more amorphous than the fully neutralized solids. Based on the results of these experiments, Gd was found to be a viable poison for neutralizing a U/Pu/Gd solution with a U:Pu mass ratio of 4.3:1 thus extending the U:Pu mass ratio from the previously investigated 0-3:1 to 4.3:1. However, further work is needed to allow higher U concentrations or U:Pu ratios greater than investigated in this work.« less

Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

NASA Astrophysics Data System (ADS)

Tucker, Lucas Powelson

This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

The NPG 7120.5A Electronic Review Process

NASA Technical Reports Server (NTRS)

McBrayer, Robert; Ives, Mark

1998-01-01

The use of electronics to review a document is well within the technical realm of today's state-of-the-art workplace. File servers and web site interaction are common tools for many NASA employees. The electronic comment processing described here was developed for the NPG 7120.5A review to augment the existing NASA Online Directives Information System (NODIS). The NODIS system is NASA's official system for formal review, approval and storage of NASA Directives. The electronic review process worked so well that NASA and other agencies may want to consider it as one of our "best practices." It was participatory decision making at its very best, a process that attracted dozens of very good ideas to improve the document as well as the way we can be managing projects far more effectively. The revision of NPG 7120.5A has significant implications for the way all elements of the Agency accomplish program and project management. Therefore, the review of NPG 7120.5A was an Agencywide effort with high visibility, heavy participation and a short schedule. The level of involvement created interest in supplementing the formal NODIS system with a system to collect comments efficiently and to allow the Centers and Codes to review and consolidate their comments into the official system in a short period of time. In addition, the Program Management Council Working Group (PMCWG), responsible for the revision of the document and the disposition of official comments, needed an electronic system to manage the disposition of comments, obtain PMCWG consensus on each disposition, and coordinate the disposition with the appropriate Headquarters Code that had submitted the official comment. The combined NASA and contractor talents and resources provided a system that supplemented the NODIS system and its operating personnel to produce a thorough review and approval of NPG 7120.5A on April 3, 1998, 7.5 months from the start of the process. The original six-month schedule is indicated. All milestones occurred on time, except for completion of comment disposition, which required an additional 30 days. Approval of the document occurred sixteen days after completion of the "Purple Package."

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.

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)

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 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.

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.

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.

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.

Packaging, Transportation and Recycling of NPP Condenser Modules - 12262

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

Polley, G.M.

2012-07-01

Perma-Fix was awarded contract from Energy Northwest for the packaging, transportation and disposition of the condenser modules, water boxes and miscellaneous metal, combustibles and water generated during the 2011 condenser replacement outage at the Columbia Generating Station. The work scope was to package the water boxes and condenser modules as they were removed from the facility and transfer them to the Perma-Fix Northwest facility for processing, recycle of metals and disposition. The condenser components were oversized and overweight (the condenser modules weighed ∼102,058 kg [225,000 lb]) which required special equipment for loading and transport. Additional debris waste was packaged inmore » inter-modals and IP-1 boxes for transport. A waste management plan was developed to minimize the generation of virtually any waste requiring landfill disposal. The Perma-Fix Northwest facility was modified to accommodate the ∼15 m [50-ft] long condenser modules and equipment was designed and manufactured to complete the disassembly, decontamination and release survey. The condenser modules are currently undergoing processing for free release to a local metal recycler. Over three millions pounds of metal will be recycled and over 95% of the waste generated during this outage will not require land disposal. There were several elements of this project that needed to be addressed during the preparation for this outage and the subsequent packaging, transportation and processing. - Staffing the project to support 24/7 generation of large components and other wastes. - The design and manufacture of the soft-sided shipping containers for the condenser modules that measured ∼15 m X 4 m X 3 m [50 ft X 13 ft X 10 ft] and weighed ∼102,058 kg [225,000 lbs] - Developing a methodology for loading the modules into the shipping containers. - Obtaining a transport vehicle for the modules. - Designing and modifying the processing facility. - Movement of the modules at the processing facility. If any of these issues were not adequately resolved prior to the start of the outage, costly delays would result and the re-start of the power plant could be impacted. The main focus of this project was to find successful methods for keeping this material out of the landfills and preserving the natural resources. In addition, this operation provided a significant cost savings to the public utility by minimizing landfill disposal. The onsite portion of the project has been completed without impact to the overall outage schedule. By the date of presentation, the majority of the waste from the condenser replacement project will have been processed and recycled. The goals for this project included helping Energy Northwest maintain the outage schedule, package and characterize waste compliantly, perform transportation activities in compliance with 49CFR (Ref-1), and minimize the waste disposal volume. During this condenser replacement project, over three millions pounds of waste was generated, packaged, characterized and transported without injury or incident. It is anticipated that 95% of the waste generated during this project will not require landfill disposal. All of the waste is scheduled to be processed, decontaminated and recycled by June of 2012. (authors)« 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.

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.

Mixed Oxide Fresh Fuel Package Auxiliary Equipment

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

Yapuncich, F.; Ross, A.; Clark, R.H.

2008-07-01

The United States Department of Energy's National Nuclear Security Administration (NNSA) is overseeing the construction the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) on the Savannah River Site. The new facility, being constructed by NNSA's contractor Shaw AREVA MOX Services, will fabricate fuel assemblies utilizing surplus plutonium as feedstock. The fuel will be used in designated commercial nuclear reactors. The MOX Fresh Fuel Package (MFFP), which has recently been licensed by the Nuclear Regulatory Commission (NRC) as a type B package (USA/9295/B(U)F-96), will be utilized to transport the fabricated fuel assemblies from the MFFF to the nuclear reactors. It wasmore » necessary to develop auxiliary equipment that would be able to efficiently handle the high precision fuel assemblies. Also, the physical constraints of the MFFF and the nuclear power plants require that the equipment be capable of loading and unloading the fuel assemblies both vertically and horizontally. The ability to reconfigure the load/unload evolution builds in a large degree of flexibility for the MFFP for the handling of many types of both fuel and non fuel payloads. The design and analysis met various technical specifications including dynamic and static seismic criteria. The fabrication was completed by three major fabrication facilities within the United States. The testing was conducted by Sandia National Laboratories. The unique design specifications and successful testing sequences will be discussed. (authors)« 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.

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.

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.

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.

GTA Welding Research and Development for Plutonium Containment

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

Sessions, C.E.

2002-02-21

This paper discusses the development of two welding systems that are used to contain actinide metals and oxides for long term storage. The systems are termed the bagless transfer system (BTS) and the outer container welder (OCW) system. The BTS is so named because it permits the containment of actinides without a polymeric package (i.e., bag). The development of these two systems was directed by Department of Energy Standard 3013, hereafter referred to as DOE 3013. This document defines the product and container requirements. In addition, it references national codes and standards for leak rates, ANSI N14.5, and design, Americanmore » Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section VIII (BandPVC).« 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.

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.

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.

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.

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

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.

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.

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

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

Limiting Surface Density Method Adapted to Large Arrays of Heterogeneous Shipping Packages with Nonlinear Responses

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

Stover, Tracy E.; Baker, James S.; Ratliff, Michael D.

The classic Limiting Surface Density (LSD) method is an empirical calculation technique for analyzing and setting mass limits for fissile items in storage arrays. LSD is a desirable method because it can reduce or eliminate the need for lengthy detailed Monte Carlo models of storage arrays. The original (or classic) method was developed based on idealized arrays of bare spherical metal items in air-spaced cubic units in a water-reflected cubic array. In this case, the geometric and material-based surface densities were acceptably correlated by linear functions. Later updates to the method were made to allow for concrete reflection rather thanmore » water, cylindrical masses rather than spheres, different material forms, and noncubic arrays. However, in the intervening four decades since those updates, little work has been done to update the method, especially for use with contemporary highly heterogeneous shipping packages that are noncubic and stored in noncubic arrays. In this work, the LSD method is reevaluated for application to highly heterogeneous shipping packages for fissile material. The package modeled is the 9975 shipping package, currently the primary package used to store fissile material at Savannah River Site’s K-Area Complex. The package is neither cubic nor rectangular but resembles nested cylinders of stainless steel, lead, aluminum, and Celotex. The fissile content is assumed to be a cylinder of plutonium metal. The packages may be arranged in arrays with both an equal number of packages per side (package cubic) and an unequal number of packages per side (noncubic). The cubic arrangements are used to derive the 9975-specific material and geometry constants for the classic linear form LSD method. The linear form of the LSD, with noncubic array adjustment, is applied and evaluated against computational models for these packages to determine the critical unit fissile mass. Sensitivity equations are derived from the classic method, and these are also used to make projections of the critical unit fissile mass. It was discovered that the heterogeneous packages have a nonlinear surface density versus critical mass relationship compared to the acceptably linear response of bare spherical fissile masses. Methodology is developed to address the nonlinear response. In so doing, the solution to the nonlinear LSD method becomes decoupled from the critical mass of a single unit, adding to its flexibility. The ability of the method to predict changes in neutron multiplication due to perturbations in a parameter is examined to provide a basis for analyzing upset conditions. In conclusion, a full rederivation of the classic LSD method from diffusion theory is also included as this was found to be lacking in the available literature.« less

Limiting Surface Density Method Adapted to Large Arrays of Heterogeneous Shipping Packages with Nonlinear Responses

DOE PAGES

Stover, Tracy E.; Baker, James S.; Ratliff, Michael D.; ...

2018-03-02

The classic Limiting Surface Density (LSD) method is an empirical calculation technique for analyzing and setting mass limits for fissile items in storage arrays. LSD is a desirable method because it can reduce or eliminate the need for lengthy detailed Monte Carlo models of storage arrays. The original (or classic) method was developed based on idealized arrays of bare spherical metal items in air-spaced cubic units in a water-reflected cubic array. In this case, the geometric and material-based surface densities were acceptably correlated by linear functions. Later updates to the method were made to allow for concrete reflection rather thanmore » water, cylindrical masses rather than spheres, different material forms, and noncubic arrays. However, in the intervening four decades since those updates, little work has been done to update the method, especially for use with contemporary highly heterogeneous shipping packages that are noncubic and stored in noncubic arrays. In this work, the LSD method is reevaluated for application to highly heterogeneous shipping packages for fissile material. The package modeled is the 9975 shipping package, currently the primary package used to store fissile material at Savannah River Site’s K-Area Complex. The package is neither cubic nor rectangular but resembles nested cylinders of stainless steel, lead, aluminum, and Celotex. The fissile content is assumed to be a cylinder of plutonium metal. The packages may be arranged in arrays with both an equal number of packages per side (package cubic) and an unequal number of packages per side (noncubic). The cubic arrangements are used to derive the 9975-specific material and geometry constants for the classic linear form LSD method. The linear form of the LSD, with noncubic array adjustment, is applied and evaluated against computational models for these packages to determine the critical unit fissile mass. Sensitivity equations are derived from the classic method, and these are also used to make projections of the critical unit fissile mass. It was discovered that the heterogeneous packages have a nonlinear surface density versus critical mass relationship compared to the acceptably linear response of bare spherical fissile masses. Methodology is developed to address the nonlinear response. In so doing, the solution to the nonlinear LSD method becomes decoupled from the critical mass of a single unit, adding to its flexibility. The ability of the method to predict changes in neutron multiplication due to perturbations in a parameter is examined to provide a basis for analyzing upset conditions. In conclusion, a full rederivation of the classic LSD method from diffusion theory is also included as this was found to be lacking in the available literature.« 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 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

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.

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

Radojcic, Riko; Nowak, Matt; Nakamoto, Mark

The status of the development of a Design-for-Stress simulation flow that captures the stress effects in packaged 3D-stacked Si products like integrated circuits (ICs) using advanced via-middle Through Si Via technology is outlined. The next set of challenges required to proliferate the methodology and to deploy it for making and dispositioning real Si product decisions are described here. These include the adoption and support of a Process Design Kit (PDK) that includes the relevant material properties, the development of stress simulation methodologies that operate at higher levels of abstraction in a design flow, and the development and adoption of suitablemore » models required to make real product reliability decisions.« less

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.

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.

Plutonium Critical Mass Curve Comparison to Mass at Upper Subcritical Limit (USL) Using Whisper

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

Alwin, Jennifer Louise; Zhang, Ning

Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the MCNP ® Monte Carlo radiation transport package. Standard approaches to validation rely on the selection of benchmarks based upon expert judgment. Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application or set of applications being analyzed. Using these benchmarks, Whisper computes a calculational margin. Whisper attempts to quantify the margin of subcriticality (MOS) from errors in software and uncertainties in nuclear data. The combination of the Whisper-derived calculational margin and MOS comprise the baseline upper subcritical limit (USL), tomore » which an additional margin may be applied by the nuclear criticality safety analyst as appropriate to ensure subcriticality. A series of critical mass curves for plutonium, similar to those found in Figure 31 of LA-10860-MS, have been generated using MCNP6.1.1 and the iterative parameter study software, WORM_Solver. The baseline USL for each of the data points of the curves was then computed using Whisper 1.1. The USL was then used to determine the equivalent mass for plutonium metal-water system. ANSI/ANS-8.1 states that it is acceptable to use handbook data, such as the data directly from the LA-10860-MS, as it is already considered validated (Section 4.3 4) “Use of subcritical limit data provided in ANSI/ANS standards or accepted reference publications does not require further validation.”). This paper attempts to take a novel approach to visualize traditional critical mass curves and allows comparison with the amount of mass for which the k eff is equal to the USL (calculational margin + margin of subcriticality). However, the intent is to plot the critical mass data along with USL, not to suggest that already accepted handbook data should have new and more rigorous requirements for validation.« 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.

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.

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

A database system for characterization of munitions items in conventional ammunition demilitarization stockpiles

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

Chun, K.C.; Chiu, S.Y.; Ditmars, J.D.

1994-05-01

The MIDAS (Munition Items Disposition Action System) database system is an electronic data management system capable of storage and retrieval of information on the detailed structures and material compositions of munitions items designated for demilitarization. The types of such munitions range from bulk propellants and small arms to projectiles and cluster bombs. The database system is also capable of processing data on the quantities of inert, PEP (propellant, explosives and pyrotechnics) and packaging materials associated with munitions, components, or parts, and the quantities of chemical compounds associated with parts made of PEP materials. Development of the MIDAS database system hasmore » been undertaken by the US Army to support disposition of unwanted ammunition stockpiles. The inventory of such stockpiles currently includes several thousand items, which total tens of thousands of tons, and is still growing. Providing systematic procedures for disposing of all unwanted conventional munitions is the mission of the MIDAS Demilitarization Program. To carry out this mission, all munitions listed in the Single Manager for Conventional Ammunition inventory must be characterized, and alternatives for resource recovery and recycling and/or disposal of munitions in the demilitarization inventory must be identified.« less

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.

Nuclear targets within the project of solving CHAllenges in Nuclear DAta

NASA Astrophysics Data System (ADS)

Sibbens, Goedele; Moens, André; Vanleeuw, David; Lewis, David; Aregbe, Yetunde

2017-09-01

In the frame of the European Commission funded integrated project CHANDA (solving CHAllenges in Nuclear DAta) the importance of nuclear target preparation for the accurateness and reliability of experimental nuclear data is set in a dedicated work package (WP3). The global aim of WP3 is the development of a network for nuclear target preparation and characterization, enabling to coordinate the target production corresponding to the experimental requirements. Therefore, a set of tasks within the work package needs to be followed. Primarily, an inventory of target related facilities and radioisotope providers was created. In the next step a priority list of target requests was made in agreement with the target user considering the technical specification, the scheduled experiments and the availability of the target laboratories. A set of target requests has been assigned to the Target Preparation laboratory of the European Commission - Joint Research Centre - Directorate G (EC-JRC.G.2) in Geel, Belgium. This contribution gives an overview of the nuclear targets that are produced within the CHANDA project. The equipment and techniques available for the preparation and characterization of uranium, plutonium and neptunium layers with an areal density ranging from 60 to 205 μg cm-2 will be emphasized.

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.

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.

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

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.

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

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.

Decontaminating the DOE-STD-3013 Inner Container to Meet 10-CFR-835 Appendix D Requirements

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

Martinez, H.E.; Nelson, T.O.; Rivera, Y.M.

The United States Department of Energy (DOE) has published a standard that specifies the criteria for preparation and packaging of plutonium metals and oxides for safe long-term storage (DOE-STD-3013-96). This standard is followed for the packaging of materials resulting from the disassembly of nuclear weapons at Los Alamos National Laboratory under the Advanced Retirement and Integrated Extraction System (ARIES) project. Declassified plutonium metal or oxide material from the ARES project is packaged into doubly contained and welded type 304L stainless steel containers that comply with the DOE standard. The 3013-96 standard describes requirements for maximum contamination limits on the outermore » surface of the sealed inner container. These limits are 500 dpm per 100 cm2 for direct measurements and 20 dpm per 100 cm2 for removable contamination. For containers filled, welded, and handled inside a highly contaminated glovebox line, these limits are difficult to obtain. Simple handling within the line is demonstrated to contaminate surfaces from 10,000 to 10,000,000 dpm alpha per 100 cm2. To routinely achieve contamination levels below the maximum contamination levels specified by the 3013-96 standard within a processing operation, a decontamination step must be included. In the ARIES line, this decontamination step is an electrolytic process that produces a controlled uniform etch of the container surfaces. Decontamination of the 3013-96 compliant ARIES inner container is well demonstrated. Within 30 to 50 minutes electrolysis time, tixed contamination is reduced to hundreds of dpm generally occurring only at electrode contact points and welds. Removable contamination is routinely brought to non-detectable levels. The total process time for the cycle (includes electrolysis, rinse, and dry stages) is on the order of 1.5 to 2 hours per container. The ARIES inner container decontamination system highly automated and consists of a plumbing loop, electronic controls and process monitors, and a decontamination chamber or "fixture". The tixture is situated like an air lock between a contaminated and an uncontaminated section of a processing glovebox. The welded and leak tested container is placed into the fixture through a door on the contaminated side and the electrolysis process is run, including rinse and dry cycles. The container is then removed through a second door into the uncontaminated side where it is monitored for surface alpha contamination, leak checked, and reweighed.« 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.

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.

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

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.

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.

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)

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.

PLUTONIUM-THORIUM ALLOYS

DOEpatents

Schonfeld, F.W.

1959-09-15

New plutonium-base binary alloys useful as liquid reactor fuel are described. The alloys consist of 50 to 98 at.% thorium with the remainder plutonium. The stated advantages of these alloys over unalloyed plutonium for reactor fuel use are easy fabrication, phase stability, and the accompanying advantuge of providing a means for converting Th/sup 232/ into U/sup 233/.

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.

Isotope ratio analysis of individual sub-micrometer plutonium particles with inductively coupled plasma mass spectrometry.

PubMed

Esaka, Fumitaka; Magara, Masaaki; Suzuki, Daisuke; Miyamoto, Yutaka; Lee, Chi-Gyu; Kimura, Takaumi

2010-12-15

Information on plutonium isotope ratios in individual particles is of great importance for nuclear safeguards, nuclear forensics and so on. Although secondary ion mass spectrometry (SIMS) is successfully utilized for the analysis of individual uranium particles, the isobaric interference of americium-241 to plutonium-241 makes difficult to obtain accurate isotope ratios in individual plutonium particles. In the present work, an analytical technique by a combination of chemical separation and inductively coupled plasma mass spectrometry (ICP-MS) is developed and applied to isotope ratio analysis of individual sub-micrometer plutonium particles. The ICP-MS results for individual plutonium particles prepared from a standard reference material (NBL SRM-947) indicate that the use of a desolvation system for sample introduction improves the precision of isotope ratios. In addition, the accuracy of the (241)Pu/(239)Pu isotope ratio is much improved, owing to the chemical separation of plutonium and americium. In conclusion, the performance of the proposed ICP-MS technique is sufficient for the analysis of individual plutonium particles. Copyright © 2010 Elsevier B.V. All rights reserved.

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.

Variations in the concentration of plutonium, strontium-90 and total alpha-emitters in human teeth collected within the British Isles.

PubMed

O'Donnell, R G; Mitchell, P I; Priest, N D; Strange, L; Fox, A; Henshaw, D L; Long, S C

1997-08-18

Concentrations of plutonium-239, plutonium-240, strontium-90 and total alpha-emitters have been measured in children's teeth collected throughout Great Britain and Ireland. The concentrations of plutonium and strontium-90 were measured in batched samples, each containing approximately 50 teeth, using low-background radiochemical methods. The concentrations of total alpha-emitters were determined in single teeth using alpha-sensitive plastic track detectors. The results showed that the average concentrations of total alpha-emitters and strontium-90 were approximately one to three orders of magnitude greater than the equivalent concentrations of plutonium-239,240. Regression analyses indicated that the concentrations of plutonium, but not strontium-90 or total alpha-emitters, decreased with increasing distance from the Sellafield nuclear fuel reprocessing plant-suggesting that this plant is a source of plutonium contamination in the wider population of the British Isles. Nevertheless, the measured absolute concentrations of plutonium (mean = 5 +/- 4 mBq kg-1 ash wt.) were so low that they are considered to present an insignificant radiological hazard.

Plutonium from Above-Ground Nuclear Tests in Milk Teeth: Investigation of Placental Transfer in Children Born between 1951 and 1995 in Switzerland

PubMed Central

Froidevaux, Pascal; Haldimann, Max

2008-01-01

Background Occupational risks, the present nuclear threat, and the potential danger associated with nuclear power have raised concerns regarding the metabolism of plutonium in pregnant women. Objective We measured plutonium levels in the milk teeth of children born between 1951 and 1995 to assess the potential risk that plutonium incorporated by pregnant women might pose to the radiosensitive tissues of the fetus through placenta transfer. Methods We used milk teeth, whose enamel is formed during pregnancy, to investigate the transfer of plutonium from the mother’s blood plasma to the fetus. We measured plutonium using sensitive sector field inductively coupled plasma mass spectrometry techniques. We compared our results with those of a previous study on strontium-90 (90Sr) released into the atmosphere after nuclear bomb tests. Results Results show that plutonium activity peaks in the milk teeth of children born about 10 years before the highest recorded levels of plutonium fallout. By contrast, 90Sr, which is known to cross the placenta barrier, manifests differently in milk teeth, in accordance with 90Sr fallout deposition as a function of time. Conclusions These findings demonstrate that plutonium found in milk teeth is caused by fallout that was inhaled around the time the milk teeth were shed and not from any accumulation during pregnancy through placenta transfer. Thus, plutonium may not represent a radiologic risk for the radiosensitive tissues of the fetus. PMID:19079728

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

METHOD OF SEPARATING PLUTONIUM

DOEpatents

Heal, H.G.

1960-02-16

BS>A method of separating plutonium from aqueous nitrate solutions of plutonium, uranium. and high beta activity fission products is given. The pH of the aqueous solution is adjusted between 3.0 to 6.0 with ammonium acetate, ferric nitrate is added, and the solution is heated to 80 to 100 deg C to selectively form a basic ferric plutonium-carrying precipitate.

PLUTONIUM AND ITS METALLURGY. A STAGE IN ITS DEVELOPMENT: THE INTERNATIONAL CONFERENCE ON THE METALLURGY OF PLUTONIUM (GRENOBLE, APRIL 1960) (in French)

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

Grison, E.

1961-01-01

A discussion is given on physical properties of plutonium, allotropic variations; kinetics of transformation; electrica; and magnetic properties; and electronic structure of the external layers of the atom. Plutonium can be used only as nuclear fuel; it is very expensive and toxic. (auth)

Siegfried S. Hecker, Plutonium, and Nonproliferation

Science.gov Websites

controversy involving the stability of certain structures (or phases) in plutonium alloys near equilibrium Cold War is Over. What Now?, DOE Technical Report, April, 1995 6th US-Russian Pu Science Workshop * Aging of Plutonium and Its Alloys * A Tale of Two Diagrams * Plutonium and Its Alloys-From Atoms to

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.

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)

SEPARATION OF RUTHENIUM FROM AQUEOUS SOLUTIONS

DOEpatents

Callis, C.F.; Moore, R.L.

1959-09-01

>The separation of ruthenium from aqueous solutions containing uranium plutonium, ruthenium, and fission products is described. The separation is accomplished by providing a nitric acid solution of plutonium, uranium, ruthenium, and fission products, oxidizing plutonium to the hexavalent state with sodium dichromate, contacting the solution with a water-immiscible organic solvent, such as hexone, to extract plutonyl, uranyl, ruthenium, and fission products, reducing with sodium ferrite the plutonyl in the solvent phase to trivalent plutonium, reextracting from the solvent phase the trivalent plutonium, ruthenium, and some fission products with an aqueous solution containing a salting out agent, introducing ozone into the aqueous acid solution to oxidize plutonium to the hexavalent state and ruthenium to ruthenium tetraoxide, and volatizing off the ruthenium tetraoxide.

Pyrochemical recovery of plutonium from calcium fluoride reduction slag

DOEpatents

Christensen, D.C.

A pyrochemical method of recovering finely dispersed plutonium metal from calcium fluoride reduction slag is claimed. The plutonium-bearing slag is crushed and melted in the presence of at least an equimolar amount of calcium chloride and a few percent metallic calcium. The calcium chloride reduces the melting point and thereby decreases the viscosity of the molten mixture. The calcium reduces any oxidized plutonium in the mixture and also causes the dispersed plutonium metal to coalesce and settle out as a separate metallic phase at the bottom of the reaction vessel. Upon cooling the mixture to room temperature, the solid plutonium can be cleanly separated from the overlying solid slag, with an average recovery yield on the order of 96 percent.

[Practical advices in choosing local anesthesia tools in dentistry. Management of carpule's quality in local anesthesia in dentistry].

PubMed

Kuzin, A V

2014-01-01

The equipment for local anesthesia is described in this article. Practical recommendations for the selection of the injection needle length, size, bevel type is given. Using dental needle for local anesthesia should be guided by the "one injection - one needle" rule, as a needle tends to deform by even the slightest contact with jawbone. Some of the shortcomings of carpule quality may be detected before use: signs of cup corrosion, the presence of sediment, air bubbles, rubber plunger disposition. In the case of such defects being identified all the package should not be used. The use of such carpule in clinical practice is unsafe.

Microdistribution and Long-Term Retention of 239Pu (NO3)4 in the Respiratory Tracts of an Acutely Exposed Plutonium Worker and Experimental Beagle Dogs

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

Nielsen, Christopher E.; Wilson, Dulaney A.; Brooks, Antone L.

The long-term retention of inhaled soluble forms of plutonium raises concerns as to the potential health effects in persons working in nuclear energy or the nuclear weapons program. The distributions of long-term retained inhaled plutonium-nitrate [239Pu (NO3)4] deposited in the lungs of an accidentally exposed nuclear worker (Human Case 0269) and in the lungs of experimentally exposed beagle dogs with varying initial lung depositions were determined via autoradiographs of selected histological lung, lymph node, trachea, and nasal turbinate tissue sections. These studies showed that both the human and dogs had a non-uniform distribution of plutonium throughout the lung tissue. Fibroticmore » scar tissue effectively encapsulated a portion of the plutonium and prevented its clearance from the body or translocation to other tissues and diminished dose to organ parenchyma. Alpha radiation activity from deposited plutonium in Human Case 0269 was observed primarily along the sub-pleural regions while no alpha activity was seen in the tracheobronchial lymph nodes of this individual. However, relatively high activity levels in the tracheobronchial lymph nodes of the beagles indicated the lymphatic system was effective in clearing deposited plutonium from the lung tissues. In both the human case and beagle dogs, the appearance of retained plutonium within the respiratory tract was inconsistent with current biokinetic models of clearance for soluble forms of plutonium. Bound plutonium can have a marked effect on the dose to the lungs and subsequent radiation exposure has the potential increase in cancer risk.« less

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.

Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

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

Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

2015-12-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are keptmore » open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

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

Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

2016-01-01

At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement formore » extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).« less

QUANTITATIVE PLUTONIUM MICRODISTRIBUTION IN BONE TISSUE OF VERTEBRA FROM A MAYAK WORKER

PubMed Central

Lyovkina, Yekaterina V.; Miller, Scott C.; Romanov, Sergey A.; Krahenbuhl, Melinda P.; Belosokhov, Maxim V.

2010-01-01

The purpose was to obtain quantitative data on plutonium microdistribution in different structural elements of human bone tissue for local dose assessment and dosimetric models validation. A sample of the thoracic vertebra was obtained from a former Mayak worker with a rather high plutonium burden. Additional information was obtained on occupational and exposure history, medical history, and measured plutonium content in organs. Plutonium was detected in bone sections from its fission tracks in polycarbonate film using neutron-induced autoradiography. Quantitative analysis of randomly selected microscopic fields on one of the autoradiographs was performed. Data included fission fragment tracks in different bone tissue and surface areas. Quantitative information on plutonium microdistribution in human bone tissue was obtained for the first time. From these data, quantitative relationship of plutonium decays in bone volume to decays on bone surface in cortical and trabecular fractions were defined as 2.0 and 0.4, correspondingly. The measured quantitative relationship of decays in bone volume to decays on bone surface does not coincide with recommended models for the cortical bone fraction by the International Commission on Radiological Protection. Biokinetic model parameters of extrapulmonary compartments might need to be adjusted after expansion of the data set on quantitative plutonium microdistribution in other bone types in human as well as other cases with different exposure patterns and types of plutonium. PMID:20838087

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.

PLUTONIUM-ZIRCONIUM ALLOYS

DOEpatents

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

1960-08-30

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

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.

Characterization studies and indicated remediation methods for plutonium contaminated soils at the Nevada Test Site

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

Murarik, T.M.; Wenstrand, T.K.; Rogers, L.A.

An initial soil characterization study was conducted to help identify possible remediation methods to remove plutonium from the Nevada Test Site and Tonapah Test Range surface soils. Results from soil samples collected across various isopleths at five sites indicate that the size-fraction distribution patterns of plutonium remain similar to findings from the Nevada Applied Ecology Group (NAEG) (1970's). The plutonium remains in the upper 10--15 cm of soils, as indicated in previous studies. Distribution of fine particles downwind'' of ground zero at each site is suggested. Whether this pattern was established immediately after each explosion or this resulted from post-shotmore » wind movement of deposited material is unclear. Several possible soil treatment scenarios are presented. Removal of plutonium from certain size fractions of the soils would alleviate the sites of much of the plutonium burden. However, the nature of association of plutonium with soil components will determine which remediation methods will most likely succeed.« less

Characterization studies and indicated remediation methods for plutonium contaminated soils at the Nevada Test Site

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

Murarik, T.M.; Wenstrand, T.K.; Rogers, L.A.

An initial soil characterization study was conducted to help identify possible remediation methods to remove plutonium from the Nevada Test Site and Tonapah Test Range surface soils. Results from soil samples collected across various isopleths at five sites indicate that the size-fraction distribution patterns of plutonium remain similar to findings from the Nevada Applied Ecology Group (NAEG) (1970`s). The plutonium remains in the upper 10--15 cm of soils, as indicated in previous studies. Distribution of fine particles ``downwind`` of ground zero at each site is suggested. Whether this pattern was established immediately after each explosion or this resulted from post-shotmore » wind movement of deposited material is unclear. Several possible soil treatment scenarios are presented. Removal of plutonium from certain size fractions of the soils would alleviate the sites of much of the plutonium burden. However, the nature of association of plutonium with soil components will determine which remediation methods will most likely succeed.« less

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.

PREPARATION OF PLUTONIUM

DOEpatents

Kolodney, M.

1959-07-01

Methods are presented for the electro-deposition of plutonium from fused mixtures of plutonium halides and halides of the alkali metals and alkaline earth metals. Th salts, preferably chlorides and with the plutonium prefer ably in the trivalent state, are placed in a refractory crucible such as tantalum or molybdenam and heated in a non-oxidizing atmosphere to 600 to 850 deg C, the higher temperatatures being used to obtain massive plutonium and the lower for the powder form. Electrodes of graphite or non reactive refractory metals are used, the crucible serving the cathode in one apparatus described in the patent.

30. VIEW OF A GLOVEBOX LINE USED IN PLUTONIUM OPERATIONS. ...

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

30. VIEW OF A GLOVEBOX LINE USED IN PLUTONIUM OPERATIONS. SAFETY AND HEALTH CONCERNS WERE OF MAJOR IMPORTANCE AT THE PLANT, BECAUSE OF THE RADIOACTIVE NATURE OF THE MATERIALS USED. PLUTONIUM GIVES OFF ALPHA AND BETA PARTICLES, GAMMA PROTONS, NEUTRONS, AND IS ALSO PYROPHORIC. AS A RESULT, PLUTONIUM OPERATIONS ARE PERFORMED UNDER CONTROLLED CONDITIONS THAT INCLUDE CONTAINMENT, FILTERING, SHIELDING, AND CREATING AN INERT ATMOSPHERE. PLUTONIUM WAS HANDLED WITHIN GLOVEBOXES THAT WERE INTERCONNECTED AND RAN SEVERAL HUNDRED FEET IN LENGTH (5/5/70). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

Rapid Method for Sodium Hydroxide/Sodium Peroxide Fusion ...

EPA Pesticide Factsheets

Technical Fact Sheet Analysis Purpose: Qualitative analysis Technique: Alpha spectrometry Method Developed for: Plutonium-238 and plutonium-239 in water and air filters Method Selected for: SAM lists this method as a pre-treatment technique supporting analysis of refractory radioisotopic forms of plutonium in drinking water and air filters using the following qualitative techniques: • Rapid methods for acid or fusion digestion • Rapid Radiochemical Method for Plutonium-238 and Plutonium 239/240 in Building Materials for Environmental Remediation Following Radiological Incidents. Summary of subject analytical method which will be posted to the SAM website to allow access to the method.

US Department of Energy Plutonium Stabilization and Immobilization Workshop, December 12-14, 1995: Final proceedings

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

NONE

1996-05-01

The purpose of the workshop was to foster communication within the technical community on issues surrounding stabilization and immobilization of the Department`s surplus plutonium and plutonium- contaminated wastes. The workshop`s objectives were to: build a common understanding of the performance, economics and maturity of stabilization and immobilization technologies; provide a system perspective on stabilization and immobilization technology options; and address the technical issues associated with technologies for stabilization and immobilization of surplus plutonium and plutonium- contaminated waste. The papers presented during this workshop have been indexed separately.

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.

Plutonium controversy

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

Richmond, C.R.

1980-01-01

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

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.

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.

MCNP Parametric Studies of Plutonium Metal and Various Interstitial Moderating Materials

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

Glazener, Natasha; Kamm, Ryan James

2017-03-31

Nuclear Criticality Safety (NCS) has performed calculations evaluating the effect of different interstitial materials on 5.0-kg of plutonium metal. As with all non-fissionable interstitials, the results here illustrate that it requires significant quantities of oil to be intimately mixed with plutonium, reflected by a thick layer of full-density water, to achieve the same reactivity as that of solid plutonium metal.

SEPARATION OF PLUTONIUM IONS FROM SOLUTION BY ADSORPTION ON ZIRCONIUM PYROPHOSPHATE

DOEpatents

Stoughton, R.W.

1961-01-31

A method is given for separating plutonium in its reduced, phosphate- insoluble state from other substances. It involves contacting a solution containing the plutonium with granular zirconium pyrophosphate.

Tank 241-AZ-101 criticality assessment resulting from pump jet mixing: Sludge mixing simulation

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

Onishi, Y.; Recknagle, K.

Tank 241-AZ-101 (AZ-101) is one of 28 double-shell tanks located in the AZ farm in the Hanford Site`s 200 East Area. The tank contains a significant quantity of fissile materials, including an estimated 9.782 kg of plutonium. Before beginning jet pump mixing for mitigative purposes, the operations must be evaluated to demonstrate that they will be subcritical under both normal and credible abnormal conditions. The main objective of this study was to address a concern about whether two 300-hp pumps with four rotating 18.3-m/s (60-ft/s) jets can concentrate plutonium in their pump housings during mixer pump operation and cause amore » criticality. The three-dimensional simulation was performed with the time-varying TEMPEST code to determine how much the pump jet mixing of Tank AZ-101 will concentrate plutonium in the pump housing. The AZ-101 model predicted that the total amount of plutonium within the pump housing peaks at 75 g at 10 simulation seconds and decreases to less than 10 g at four minutes. The plutonium concentration in the entire pump housing peaks at 0.60 g/L at 10 simulation seconds and is reduced to below 0.1 g/L after four minutes. Since the minimum critical concentration of plutonium is 2.6 g/L, and the minimum critical plutonium mass under idealized plutonium-water conditions is 520 g, these predicted maximums in the pump housing are much lower than the minimum plutonium conditions needed to reach a criticality level. The initial plutonium maximum of 1.88 g/L still results in safety factor of 4.3 in the pump housing during the pump jet mixing operation.« less

Complications Associated with Long-Term Disposition of Newly-Generated Transuranic Waste: A National Laboratory Perspective

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

B.J. Orchard; L.A. Harvego; T.L. Carlson

The Idaho National Laboratory (INL) is a multipurpose national laboratory delivering specialized science and engineering solutions for the U.S. Department of Energy (DOE). Sponsorship of INL was formally transferred to the DOE Office of Nuclear Energy, Science and Technology (NE) by Secretary Spencer Abraham in July 2002. The move to NE, and designation as the DOE lead nuclear energy laboratory for reactor technology, supports the nation’s expanding nuclear energy initiatives, placing INL at the center of work to develop advanced Generation IV nuclear energy systems; nuclear energy/hydrogen coproduction technology; advanced nuclear energy fuel cycle technologies; and providing national security answersmore » to national infrastructure needs. As a result of the Laboratory’s NE mission, INL generates both contact-handled and remote-handled transuranic (TRU) waste from ongoing operations. Generation rates are relatively small and fluctuate based on specific programs and project activities being conducted; however, the Laboratory will continue to generate TRU waste well into the future in association with the NE mission. Currently, plans and capabilities are being established to transfer INL’s contact-handled TRU waste to the Advanced Mixed Waste Treatment Plant (AMWTP) for certification and disposal to the Waste Isolation Pilot Plant (WIPP). Remote-handled TRU waste is currently placed in storage at the Materials and Fuels Complex (MFC). In an effort to minimize future liabilities associated with the INL NE mission, INL is evaluating and assessing options for the management and disposition of all its TRU waste on a real-time basis at time of generation. This paper summarizes near-term activities to minimize future re handling of INL’s TRU waste, as well as, potential complications associated with the long-term disposition of newly-generated TRU waste. Potential complications impacting the disposition of INL newly-generated TRU waste include, but are not limited to: 1) required remote-handled TRU packaging configuration(s) vs. current facility capabilities, 2) long-term NE mission activities, 3) WIPP certification requirements, and 4) budget considerations.« less

Plutonium and americium separation from salts

DOEpatents

Hagan, Paul G.; Miner, Frend J.

1976-01-01

Salts or materials containing plutonium and americium are dissolved in hydrochloric acid, heated, and contacted with an alkali metal carbonate solution to precipitate plutonium and americium carbonates which are thereafter readily separable from the solution.

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.

WET METHOD OF PREPARING PLUTONIUM TRIBROMIDE

DOEpatents

Davidson, N.R.; Hyde, E.K.

1958-11-11

S> The preparation of anhydrous plutonium tribromide from an aqueous acid solution of plutonium tetrabromide is described, consisting of adding a water-soluble volatile bromide to the tetrabromide to provide additional bromide ions sufficient to furnish an oxidation-reduction potential substantially more positive than --0.966 volt, evaporating the resultant plutonium tribromides to dryness in the presence of HBr, and dehydrating at an elevated temperature also in the presence of HBr.

Spectrophotometers for plutonium monitoring in HB-line

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

Lascola, R. J.; O'Rourke, P. E.; Kyser, E. A.

2016-02-12

This report describes the equipment, control software, calibrations for total plutonium and plutonium oxidation state, and qualification studies for the instrument. It also provides a detailed description of the uncertainty analysis, which includes source terms associated with plutonium calibration standards, instrument drift, and inter-instrument variability. Also included are work instructions for instrument, flow cell, and optical fiber setup, work instructions for routine maintenance, and drawings and schematic diagrams.

Plutonium radiation surrogate

DOEpatents

Frank, Michael I [Dublin, CA

2010-02-02

A self-contained source of gamma-ray and neutron radiation suitable for use as a radiation surrogate for weapons-grade plutonium is described. The source generates a radiation spectrum similar to that of weapons-grade plutonium at 5% energy resolution between 59 and 2614 keV, but contains no special nuclear material and emits little .alpha.-particle radiation. The weapons-grade plutonium radiation surrogate also emits neutrons having fluxes commensurate with the gamma-radiation intensities employed.

A GREEN'S FUNCTION APPROACH FOR DETERMINING DOSE RATES FOR SMALL GRAM QUANTITIES IN SHIPPING PACKAGINGS

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

Nathan, S.

The Small Gram Quantity (SGQ) concept is based on the understanding that small amounts of hazardous materials, in this case radioactive materials (RAM), are significantly less hazardous than large amounts of the same materials. This paper describes a methodology designed to estimate an SGQ for several neutron and gamma emitting isotopes that can be shipped in a package in compliance with 10 CFR Part 71 external radiation level limits regulations. The neutron and photon sources were calculated using both ORIGEN-S and RASTA. The response from a unit source in each neutron and photon group was calculated using MCNP5 with eachmore » unshielded and shielded container configuration. Effects of self-shielding on both neutron and photon response were evaluated by including either plutonium oxide or iron in the source region for the case with no shielded container. For the cases of actinides mixed with light elements, beryllium is the bounding light element. The added beryllium (10 to 90 percent of the actinide mass) in the cases studied represents between 9 and 47 percent concentration of the total mixture mass. For beryllium concentrations larger than 50 percent, the increase in the neutron source term and dose rate tend to increase at a much lower rate than at concentrations lower than 50%. The intimately mixed actinide-beryllium form used in these models is very conservative and thus the limits presented in this report are practical bounds on the mass that can be safely shipped. The calculated dose rate from one gram of each isotope was then used to determin the maximum amount of a single isotope that could be shipped in the Model 9977 Package (or packagings having the same or larger external dimensions as well as similar structural materials) and have the external radiation level within the regulatory dose limits at the surface of the package. The estimates of the mass limits presented would also serve as conservative limits for both the Models 9975 and 9978 packages. If a package contains a mixture of isotopes, the acceptability for shipment can be determined by a sum of fractions approach. It should be noted that the SGQ masses presented in this report represent limits that would comply with the external radiation limits under 10CFR Part 71. They do not necessarily bound lower limits that may be required to comply with other factors such as heat load of the package.« less

PLUTONIUM ALLOYS CONTAINING CONTROLLED AMOUNTS OF PLUTONIUM ALLOTROPES OBTAINED BY APPLICATION OF HIGH PRESSURES

DOEpatents

Elliott, R.O.; Gschneidner, K.A. Jr.

1962-07-10

A method of making stabilized plutonium alloys which are free of voids and cracks and have a controlled amount of plutonium allotropes is described. The steps include adding at least 4.5 at.% of hafnium, indium, or erbium to the melted plutonium metal, homogenizing the resulting alloy at a temperature of 450 deg C, cooling to room temperature, and subjecting the alloy to a pressure which produces a rapid increase in density with a negligible increase in pressure. The pressure required to cause this rapid change in density or transformation ranges from about 800 to 2400 atmospheres, and is dependent on the alloying element. (AEC)

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.

The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

NASA Astrophysics Data System (ADS)

Blandinskiy, V. Yu.

2014-12-01

This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

METHOD OF SEPARATING TETRAVALENT PLUTONIUM VALUES FROM CERIUM SUB-GROUP RARE EARTH VALUES

DOEpatents

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

1959-02-01

A method is presented for separating plutonium from the cerium sub-group of rare earths when both are present in an aqueous solution. The method consists in adding an excess of alkali metal carbonate to the solution, which causes the formation of a soluble plutonium carbonate precipitate and at the same time forms an insoluble cerium-group rare earth carbonate. The pH value must be adjusted to bctween 5.5 and 7.5, and prior to the precipitation step the plutonium must be reduced to the tetravalent state since only tetravalent plutonium will form the soluble carbonate complex.

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.

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.

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.

Advances in containment methods and plutonium recovery strategies that led to the structural characterization of plutonium(IV) tetrachloride tris-diphenylsulfoxide, PuCl 4(OSPh 2) 3

DOE PAGES

Schrell, Samantha K.; Boland, Kevin Sean; Cross, Justin Neil; ...

2017-01-18

In an attempt to further advance the understanding of plutonium coordination chemistry, we report a robust method for recycling and obtaining plutonium aqueous stock solutions that can be used as a convenient starting material in plutonium synthesis. This approach was used to prepare and characterize plutonium(IV) tetrachloride tris-diphenylsulfoxide, PuCl 4(OSPh 2) 3, by single crystal X-ray diffraction. The PuCl 4(OSPh 2) 3 compound represents a rare example of a 7-coordinate plutonium(IV) complex. Structural characterization of PuCl 4(OSPh 2) 3 by X-ray diffraction utilized a new containment method for radioactive crystals. The procedure makes use of epoxy, polyimide loops, and amore » polyester sheath to provide a robust method for safely containing and easily handling radioactive samples. Lastly, the described procedure is more user friendly than traditional containment methods that employ fragile quartz capillary tubes. Additionally, moving to polyester, instead of quartz, lowers the background scattering from the heavier silicon atoms.« less

JOWOG 22/2 - Actinide Chemical Technology (July 9-13, 2012)

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

Jackson, Jay M.; Lopez, Jacquelyn C.; Wayne, David M.

2012-07-05

The Plutonium Science and Manufacturing Directorate provides world-class, safe, secure, and reliable special nuclear material research, process development, technology demonstration, and manufacturing capabilities that support the nation's defense, energy, and environmental needs. We safely and efficiently process plutonium, uranium, and other actinide materials to meet national program requirements, while expanding the scientific and engineering basis of nuclear weapons-based manufacturing, and while producing the next generation of nuclear engineers and scientists. Actinide Process Chemistry (NCO-2) safely and efficiently processes plutonium and other actinide compounds to meet the nation's nuclear defense program needs. All of our processing activities are done in amore » world class and highly regulated nuclear facility. NCO-2's plutonium processing activities consist of direct oxide reduction, metal chlorination, americium extraction, and electrorefining. In addition, NCO-2 uses hydrochloric and nitric acid dissolutions for both plutonium processing and reduction of hazardous components in the waste streams. Finally, NCO-2 is a key team member in the processing of plutonium oxide from disassembled pits and the subsequent stabilization of plutonium oxide for safe and stable long-term storage.« less

Results Of Routine Strip Effluent Hold Tank, Decontaminated Salt Solution Hold Tank, And Caustic Wash Tank Samples From Modular Caustic-Side Solvent Extraction Unit During Macrobatch 4 Operations

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

Peters, T. B.; Fink, S. D.

Strip Effluent Hold Tank (SEHT), Decontaminated Salt Solution Hold Tank (DSSHT), and Caustic Wash Tank (CWT) samples from several of the ?microbatches? of Integrated Salt Disposition Project (ISDP) Salt Batch (?Macrobatch?) 4 have been analyzed for {sup 238}Pu, {sup 90}Sr, {sup 137}Cs, and by inductively-coupled plasma emission spectroscopy (ICPES). Furthermore, samples from the CWT have been analyzed by a variety of methods to investigate a decline in the decontamination factor (DF) of the cesium observed at MCU. The results indicate good decontamination performance within process design expectations. While the data set is sparse, the results of this set and themore » previous set of results for Macrobatch 3 samples indicate generally consistent operations. There is no indication of a disruption in plutonium and strontium removal. The average cesium DF and concentration factor (CF) for samples obtained from Macrobatch 4 are slightly lower than for Macrobatch 3, but still well within operating parameters. The DSSHT samples show continued presence of titanium, likely from leaching of the monosodium titanate in Actinide Removal Process (ARP).« less

Safety features of subcritical fluid fueled systems

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

Bell, C.R.

1995-10-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitativemore » in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.« less

Performance of NDA techniques on a vitrified waste form

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

Hurd, J.R.; Veazey, G.W.; Prettyman, T.H.

1997-11-01

Rocky Flats Environmental Technology Site (RFETS) is currently considering the use of vitrified transuranic (TRU)-waste forms for the final disposition of several waste materials. To date, however, little nondestructive assay (NDA) data have been acquired in the general NDA community to assist in this endeavor. This paper describes the efforts to determine constraints and operating parameters for using NDA instrumentation on vitrified waste. The present study was conducted on a sample composed of a plutonium-contaminated ash, similar to that found in the RFETS inventory, and a borosilicate-based glass. The vitrified waste item was fabricated at Los Alamos National Laboratory (LANL)more » using methods and equipment similar to those being proposed by RFETS to treat their ash material. The focus of this study centered on the segmented gamma scanner (SGS) with 1/2-inch collimation, a technique that is presently available at RFETS. The accuracy and precision of SGS technology was evaluated, with particular attention to bias issues involving matrix geometry, homogeneity, and attenuation. Tomographic gamma scanning was utilized in the determination of the waste form homogeneity. A thermal neutron technique was also investigated and comparisons made with the gamma results.« less

K Basins sludge removal temporary sludge storage tank system

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

Mclean, M.A.

1997-06-12

Shipment of sludge from the K Basins to a disposal site is now targeted for August 2000. The current path forward for sludge disposal is shipment to Tank AW-105 in the Tank Waste Remediation System (TWRS). Significant issues of the feasibility of this path exist primarily due to criticality concerns and the presence of polychlorinated biphenyls (PCBS) in the sludge at levels that trigger regulation under the Toxic Substance Control Act. Introduction of PCBs into the TWRS processes could potentially involve significant design and operational impacts to both the Spent Nuclear Fuel and TWRS projects if technical and regulatory issuesmore » related to PCB treatment cannot be satisfactorily resolved. Concerns of meeting the TWRS acceptance criteria have evolved such that new storage tanks for the K Basins sludge may be the best option for storage prior to vitrification of the sludge. A reconunendation for the final disposition of the sludge is scheduled for June 30, 1997. To support this decision process, this project was developed. This project provides a preconceptual design package including preconceptual designs and cost estimates for the temporary sludge storage tanks. Development of cost estimates for the design and construction of sludge storage systems is required to help evaluate a recommendation for the final disposition of the K Basin sludge.« less

PLUTONIUM-CERIUM ALLOY

DOEpatents

Coffinberry, A.S.

1959-01-01

An alloy is presented for use as a reactor fuel. The binary alloy consists essentially of from about 5 to 90 atomic per cent cerium and the balance being plutonium. A complete phase diagram for the cerium--plutonium system is given.

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.

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

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.)

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)

THE CHEMICAL ANALYSIS OF TERNARY ALLOYS OF PLUTONIUM WITH MOLYBDENUM AND URANIUM

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

Phillips, G.; Woodhead, J.; Jenkins, E.N.

1958-09-01

It is shown that the absorptiometric determination of molybdenum as thiocyanate may be used in the presence of plutonium. Molybdenum interferes with previously published methods for determining uranium and plutonium but conditlons have been established for its complete removal by solvent extraction of the compound with alpha -benzoin oxime. The previous methods for uranium and plutonium are satisfactory when applied to the residual aqueous phase following this solvent extraction. (auth)

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.

SEPARATION OF PLUTONIUM VALUES FROM URANIUM AND FISSION PRODUCT VALUES

DOEpatents

Maddock, A.G.; Booth, A.H.

1960-09-13

Separation of plutonium present in small amounts from neutron irradiated uranium by making use of the phenomenon of chemisorption is described. Plutonium in the tetravalent state is chemically absorbed on a fluoride in solid form. The steps for the separation comprise dissolving the irradiated uranium in nitric acid, oxidizing the plutonium in the resulting solution to the hexavalent state, adding to the solution a soluble calcium salt which by the common ion effect inhibits dissolution of the fluoride by the solution, passing the solution through a bed or column of subdivided calcium fluoride which has been sintered to about 8OO deg C to remove the chemisorbable fission products, reducing the plutonium in the solution thus obtained to the tetravalent state, and again passing the solution through a similar bed or column of calcium fluoride to selectively absorb the plutonium, which may then be recovered by treating the calcium fluoride with a solution of ammonium oxalate.

Using Biomolecules to Separate Plutonium

NASA Astrophysics Data System (ADS)

Gogolski, Jarrod

Used nuclear fuel has traditionally been treated through chemical separations of the radionuclides for recycle or disposal. This research considers a biological approach to such separations based on a series of complex and interdependent interactions that occur naturally in the human body with plutonium. These biological interactions are mediated by the proteins serum transferrin and the transferrin receptor. Transferrin to plutonium in vivo and can deposit plutonium into cells after interacting with the transferrin receptor protein at the cell surface. Using cerium as a non-radioactive surrogate for plutonium, it was found that cerium(IV) required multiple synergistic anions to bind in the N-lobe of the bilobal transferrin protein, creating a conformation of the cerium-loaded protein that would be unable to interact with the transferrin receptor protein to achieve a separation. The behavior of cerium binding to transferrin has contributed to understanding how plutonium(IV)-transferrin interacts in vivo and in biological separations.

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.

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.

Amarillo National Resource Center for Plutonium. Quarterly technical progress report, May 1, 1997--July 31, 1997

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

NONE

Progress summaries are provided from the Amarillo National Center for Plutonium. Programs include the plutonium information resource center, environment, public health, and safety, education and training, nuclear and other material studies.

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.

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

69. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING SOUTHWEST THROUGH ...

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

69. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING SOUTHWEST THROUGH DOOR-WAY INTO PLUTONIUM STORAGE AREA. - Loring Air Force Base, Weapons Storage Area, Northeastern corner of base at northern end of Maine Road, Limestone, Aroostook County, ME

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.

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.

NON-CORROSIVE PLUTONIUM FUEL SYSTEMS

DOEpatents

Coffinberry, A.S.; Waber, J.T.

1962-10-23

An improved plutonium reactor liquid fuel is described for utilization in a nuclear reactor having a tantalum fuel containment vessel. The fuel consists of plutonium and a diluent such as iron, cobalt, nickel, cerium, cerium-- iron, cerium--cobalt, cerium--nickel, and cerium--copper, and an additive of carbon and silicon. The carbon and silicon react with the tantalum container surface to form a coating that is self-healing and prevents the corrosive action of liquid plutonium on the said tantalum container. (AEC)

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.

Plutonium and americium in the foodchain lichen-reindeer-man

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

Jaakkola, T.; Hakanen, M.; Keinonen, M.

1977-01-01

The atmospheric nuclear tests have produced a worldwide fallout of transuranium elements. In addition to plutonium measurable concentrations of americium are to be found in terrestrial and aquatic environments. The metabolism of plutonium in reindeer was investigated by analyzing plutonium in liver, bone, and lung collected during 1963-1976. To determine the distribution of plutonium in reindeer all tissues of four animals of different ages were analyzed. To estimate the uptake of plutonium from the gastrointestinal tract in reindeer, the tissue samples of elk were also analyzed. Elk which is of the same genus as reindeer does not feed on lichenmore » but mainly on deciduous plants, buds, young twigs, and leaves of trees and bushes. The composition of its feed corresponds fairly well to that of reindeer during the summer. Studies on behaviour of americium along the foodchain lichen-reindeer-man were started by determining the Am-241 concentrations in lichen and reindeer liver. The Am-241 results were compared with those of Pu-239,240. The plutonium contents of the southern Finns, whose diet does not contain reindeer tissues, were determined by analyzing autopsy tissue samples (liver, lung, and bone). The southern Finns form a control group to the Lapps consuming reindeer tissues. Plutonium analyses of the placenta, blood, and tooth samples of the Lapps were performed.« less

METHOD OF REDUCING PLUTONIUM COMPOUNDS

DOEpatents

Johns, I.B.

1958-06-01

A method is described for reducing plutonium compounds in aqueous solution from a higher to a lower valence state. This reduction of valence is achieved by treating the aqueous solution of higher valence plutonium compounds with hydrogen in contact with an activated platinum catalyst.

71. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING NORTHEAST INTO ...

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

71. INTERIOR, BUILDING 272 (PLUTONIUM STORAGE BUILDING) LOOKING NORTHEAST INTO PLUTONIUM STORAGE ROOM SHOWING CUBICLES FOR STORAGE. - Loring Air Force Base, Weapons Storage Area, Northeastern corner of base at northern end of Maine Road, Limestone, Aroostook County, ME

Plutonium Isotopes in the Terrestrial Environment at the Savannah River Site, USA. A Long-Term Study

DOE PAGES

Armstrong, Christopher R.; Nuessle, Patterson R.; Brant, Heather A.; ...

2015-01-16

This work presents the findings of a long term plutonium study at Savannah River Site (SRS) conducted between 2003 and 2013. Terrestrial environmental samples were obtained at Savannah River National Laboratory (SRNL) in A-area. Plutonium content and isotopic abundances were measured over this time period by alpha spectrometry and three stage thermal ionization mass spectrometry (3STIMS). Here we detail the complete sample collection, radiochemical separation, and measurement procedure specifically targeted to trace plutonium in bulk environmental samples. Total plutonium activities were determined to be not significantly above atmospheric global fallout. However, the 238Pu/ 239+240Pu activity ratios attributed to SRS aremore » above atmospheric global fallout ranges. The 240Pu/ 239Pu atom ratios are reasonably consistent from year to year and are lower than fallout, while the 242Pu/ 239Pu atom ratios are higher than fallout values. Overall, the plutonium signatures obtained in this study reflect a mixture of weapons-grade, higher burn-up, and fallout material. This study provides a blue print for long term low level monitoring of plutonium in the environment.« less

Electronic structure, phase transitions and diffusive properties of elemental plutonium

NASA Astrophysics Data System (ADS)

Setty, Arun; Cooper, B. R.

2003-03-01

We present a SIC-LDA-LMTO based study of the electronic structure of the delta, alpha and gamma phases of plutonium, and also of the alpha and gamma phases of elemental cerium. We find excellent agreement with the experimental densities and magnetic properties [1]. Furthermore, detailed studies of the computational densities of states for delta plutonium, and comparison with the experimental photoemission spectrum [2], provide evidence for the existence of an unusual fluctuating valence state. Results regarding the vacancy formation and self-diffusion in delta plutonium will be presented. Furthermore, a study of interface diffusion between plutonium and steel (technologically relevant in the storage of spent fuel) or other technologically relevant alloys will be included. Preliminary results regarding gallium stabilization of delta plutonium, and of plutonium alloys will be presented. [1] M. Dormeval et al., private communication (2001). [2] A. J. Arko, J. J. Joyce, L. Morales, J. Wills, and J. Lashley et. al., Phys. Rev. B, 62, 1773 (2000). [3] B. R. Cooper et al, Phil. Mag. B 79, 683 (1999); B.R. Cooper, Los Alamos Science 26, 106 (2000)); B.R. Cooper, A.K. Setty and D.L.Price, to be published.

Radiation damage and annealing in plutonium tetrafluoride

NASA Astrophysics Data System (ADS)

McCoy, Kaylyn; Casella, Amanda; Sinkov, Sergey; Sweet, Lucas; McNamara, Bruce; Delegard, Calvin; Jevremovic, Tatjana

2017-12-01

A sample of plutonium tetrafluoride that was separated prior to 1966 at the Hanford Site in Washington State was analyzed at the Pacific Northwest National Laboratory (PNNL) in 2015 and 2016. The plutonium tetrafluoride, as received, was an unusual color and considering the age of the plutonium, there were questions about the condition of the material. These questions had to be answered in order to determine the suitability of the material for future use or long-term storage. Therefore, thermogravimetric/differential thermal analysis and X-ray diffraction evaluations were conducted to determine the plutonium's crystal structure, oxide content, and moisture content; these analyses reported that the plutonium was predominately amorphous and tetrafluoride, with an oxide content near ten percent. Freshly fluorinated plutonium tetrafluoride is known to be monoclinic. During the initial thermogravimetric/differential thermal analyses, it was discovered that an exothermic event occurred within the material near 414 °C. X-ray diffraction analyses were conducted on the annealed tetrafluoride. The X-ray diffraction analyses indicated that some degree of recrystallization occurred in conjunction with the 414 °C event. The following commentary describes the series of thermogravimetric/differential thermal and X-ray diffraction analyses that were conducted as part of this investigation at PNNL.

Determination of plutonium isotopes (238,239,240Pu) and strontium (90Sr) in seafood using alpha spectrometry and liquid scintillation spectrometry.

PubMed

Shin, Choonshik; Choi, Hoon; Kwon, Hye-Min; Jo, Hye-Jin; Kim, Hye-Jeong; Yoon, Hae-Jung; Kim, Dong-Sul; Kang, Gil-Jin

2017-10-01

The present study was carried out to survey the levels of plutonium isotopes ( 238 , 239 , 240 Pu) and strontium ( 90 Sr) in domestic seafood in Korea. In current, regulatory authorities have analyzed radionuclides, such as 134 Cs, 137 Cs and 131 I, in domestic and imported food. However, people are concerned about contamination of other radionuclides, such as plutonium and strontium, in food. Furthermore, people who live in Korea have much concern about safety of seafood. Accordingly, in this study, we have investigated the activity concentrations of plutonium and strontium in seafood. For the analysis of plutonium isotopes and strontium, a rapid and reliable method developed from previous study was used. Applicability of the test method was verified by examining recovery, minimum detectable activity (MDA), analytical time, etc. Total 40 seafood samples were analyzed in 2014-2015. As a result, plutonium isotopes ( 238 , 239 , 240 Pu) and strontium ( 90 Sr) were not detected or below detection limits in seafood. The detection limits of plutonium isotopes and strontium-90 were 0.01 and 1 Bq/kg, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

33 Shafts Category of Transuranic Waste Stored Below Ground within Area G

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

Hargis, Kenneth Marshall; Monk, Thomas H

This report compiles information to support the evaluation of alternatives and analysis of regulatory paths forward for the 33 shafts. The historical information includes a form completed by waste generators for each waste package (Reference 6) that included a waste description, estimates of Pu-239 and uranium-235 (U-235) based on an accounting technique, and calculations of mixed fission products (MFP) based on radiation measurements. A 1979 letter and questionnaire (Reference 7) provides information on waste packaging of hot cell waste and the configuration of disposal shafts as storage in the 33 Shafts was initiated. Tables of data by waste package weremore » developed during a review of historical documents that was performed in 2005 (Reference 8). Radiological data was coupled with material-type data to estimate the initial isotopic content of each waste package and an Oak Ridge National Laboratory computer code was used to calculate 2009 decay levels. Other sources of information include a waste disposal logbook for the 33 shafts (Reference 9), reports that summarize remote-handled waste generated at the CMR facility (Reference 10) and placement of waste in the 33 shafts (Reference 11), a report on decommissioning of the LAMPRE reactor (Reference 12), interviews with an employee and manager involved in placing waste in the 33 shafts (References 13 and 14), an interview with a long-time LANL employee involved in waste operations (Reference 15), a 2002 plan for disposition of remote-handled TRU waste (Reference 16), and photographs obtained during field surveys of several shafts in 2007. The WIPP Central Characterization Project (CCP) completed an Acceptable Knowledge (AK) summary report for 16 canisters of remote-handled waste from the CMR Facility that contains information relevant to the 33 Shafts on hot-cell operations and timeline (Reference 17).« 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)

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...

3. AERIAL VIEW, LOOKING SOUTH, OF BUILDING 371 BASEMENT UNDER ...

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

3. AERIAL VIEW, LOOKING SOUTH, OF BUILDING 371 BASEMENT UNDER CONSTRUCTION. THE BASEMENT HOUSES HEATING, VENTILATION, AND AIR CONDITIONING EQUIPMENT AND MECHANICAL UTILITIES, THE UPPER PART OF THE PLUTONIUM STORAGE VAULT AND MAINTENANCE BAY, AND SMALL PLUTONIUM PROCESSING AREAS. THE BASEMENT LEVEL IS DIVIDED INTO NEARLY EQUAL NORTH AND SOUTH PARTS BY THE UPPER PORTION OF THE PLUTONIUM STORAGE VAULT. (10/7/74) - Rocky Flats Plant, Plutonium Recovery Facility, Northwest portion of Rocky Flats Plant, Golden, Jefferson County, CO

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.

Uranium daughter growth must not be neglected when adjusting plutonium materials for assay and isotopic contents

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

Marsh, S.F.; Spall, W.D.; Abernathey, R.M.

1976-11-01

Relationships are provided to compute the decreasing plutonium content and changing isotopic distribution of plutonium materials for the radioactive decay of /sup 238/Pu, /sup 239/Pu, /sup 240/Pu and /sup 242/Pu to long-lived uranium daughters and of /sup 241/Pu to /sup 241/Am. This computation is important to the use of plutonium reference materials to calibrate destructive and nondestructive methods for assay and isotopic measurements, as well as to accountability inventory calculations.

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.

Actinide-contaminated Skin: Comparing Decontamination Efficacy of Water, Cleansing Gels, and DTPA Gels.

PubMed

Tazrart, A; Bolzinger, M A; Lamart, S; Coudert, S; Angulo, J F; Jandard, V; Briançon, S; Griffiths, N M

2018-07-01

Skin contamination by alpha-emitting actinides is a risk to workers during nuclear fuel production and reactor decommissioning. Also, the list of items for potential use in radiological dispersal devices includes plutonium and americium. The actinide chemical form is important and solvents such as tributyl phosphate, used to extract plutonium, can influence plutonium behavior. This study investigated skin fixation and efficacy of decontamination products for these actinide forms using viable pig skin in the Franz cell diffusion system. Commonly used or recommended decontamination products such as water, cleansing gel, diethylenetriamine pentaacetic acid, or octadentate hydroxypyridinone compound 3,4,3-LI(1,2-HOPO), as well as diethylenetriamine pentaacetic acid hydrogel formulations, were tested after a 2-h contact time with the contaminant. Analysis of skin samples demonstrated that more plutonium nitrate is bound to skin as compared to plutonium-tributyl phosphate, and fixation of americium to skin was also significant. The data show that for plutonium-tributyl phosphate all the products are effective ranging from 80 to 90% removal of this contaminant. This may be associated with damage to the skin by this complex and suggests a mechanical/wash-out action rather than chelation. For removal of americium and plutonium, both Trait Rouge cleansing gel and diethylenetriamine pentaacetic acid are better than water, and diethylenetriamine pentaacetic acid hydrogel is better than Osmogel. The different treatments, however, did not significantly affect the activity in deeper skin layers, which suggests a need for further improvement of decontamination procedures. The new diethylenetriamine pentaacetic acid hydrogel preparation was effective in removing americium, plutonium, and plutonium-tributyl phosphate from skin; such a formulation offers advantages and thus merits further assessment.

Sources of plutonium in the atmosphere and stratosphere-troposphere mixing

PubMed Central

Hirose, Katsumi; Povinec, Pavel P.

2015-01-01

Plutonium isotopes have primarily been injected to the stratosphere by the atmospheric nuclear weapon tests and the burn-up of the SNAP-9A satellite. Here we show by using published data that the stratospheric plutonium exponentially decreased with apparent residence time of 1.5 ± 0.5 years, and that the temporal variations of plutonium in surface air followed the stratospheric trends until the early 1980s. In the 2000s, plutonium and its isotope ratios in the atmosphere varied dynamically, and sporadic high concentrations of 239,240Pu reported for the lower stratospheric and upper tropospheric aerosols may be due to environmental events such as the global dust outbreaks and biomass burning. PMID:26508010

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

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

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 OF SEPARATION OF PLUTONIUM FROM CARRIER PRECIPITATES

DOEpatents

Dawson, I.R.

1959-09-22

The recovery of plutonium from fluoride carrier precipitates is described. The precipitate is dissolved in zirconyl nitrate, ferric nitrate, aluminum nitrate, or a mixture of these complexing agents, and the plutonium is then extracted from the aqueous solution formed with a water-immiscible organic solvent.

EXTRACTION METHOD FOR SEPARATING URANIUM, PLUTONIUM, AND FISSION PRODUCTS FROM COMPOSITIONS CONTAINING SAME

DOEpatents

Seaborg, G.T.

1957-10-29

Methods for separating plutonium from the fission products present in masses of neutron irradiated uranium are reported. The neutron irradiated uranium is first dissolved in an aqueous solution of nitric acid. The plutonium in this solution is present as plutonous nitrate. The aqueous solution is then agitated with an organic solvent, which is not miscible with water, such as diethyl ether. The ether extracts 90% of the uraryl nitrate leaving, substantially all of the plutonium in the aqueous phase. The aqueous solution of plutonous nitrate is then oxidized to the hexavalent state, and agitated with diethyl ether again. In the ether phase there is then obtained 90% of plutonium as a solution of plutonyl nitrate. The ether solution of plutonyl nitrate is then agitated with water containing a reducing agent such as sulfur dioxide, and the plutonium dissolves in the water and is reduced to the plutonous state. The uranyl nitrate remains in the ether. The plutonous nitrate in the water may be recovered by precipitation.

Plutonium release from Fukushima Daiichi fosters the need for more detailed investigations

NASA Astrophysics Data System (ADS)

Schneider, Stephanie; Walther, Clemens; Bister, Stefan; Schauer, Viktoria; Christl, Marcus; Synal, Hans-Arno; Shozugawa, Katsumi; Steinhauser, Georg

2013-10-01

The contamination of Japan after the Fukushima accident has been investigated mainly for volatile fission products, but only sparsely for actinides such as plutonium. Only small releases of actinides were estimated in Fukushima. Plutonium is still omnipresent in the environment from previous atmospheric nuclear weapons tests. We investigated soil and plants sampled at different hot spots in Japan, searching for reactor-borne plutonium using its isotopic ratio 240Pu/239Pu. By using accelerator mass spectrometry, we clearly demonstrated the release of Pu from the Fukushima Daiichi power plant: While most samples contained only the radionuclide signature of fallout plutonium, there is at least one vegetation sample whose isotope ratio (0.381 +/- 0.046) evidences that the Pu originates from a nuclear reactor (239+240Pu activity concentration 0.49 Bq/kg). Plutonium content and isotope ratios differ considerably even for very close sampling locations, e.g. the soil and the plants growing on it. This strong localization indicates a particulate Pu release, which is of high radiological risk if incorporated.

Plutonium release from the 903 pad at Rocky Flats.

PubMed

Mongan, T R; Ripple, S R; Winges, K D

1996-10-01

The Colorado Department of Public Health and Environment (CDH) sponsored a study to reconstruct contaminant doses to the public from operations at the Rocky Flats nuclear weapons plant. This analysis of the accidental release of plutonium from the area known as the 903 Pad is part of the CDH study. In the 1950's and 1960's, 55-gallon drums of waste oil contaminated with plutonium, and uranium were stored outdoors at the 903 Pad. The drums corroded, leaking contaminated oil onto soil subsequently carried off-site by the wind. The plutonium release is estimated using environmental data from the 1960's and 1970's and an atmospheric transport model for fugitive dust. The best estimate of total plutonium release to areas beyond plant-owned property is about 0.26 TBq (7 Ci). Off-site airborne concentrations and deposition of plutonium are estimated for dose calculation purposes. The best estimate of the highest predicted off-site effective dose is approximately 72 microSv (7.2 mrem).

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

Radioecology of natural systems. Fifteenth annual progress report, August 1, 1976--July 31, 1977. [Plutonium transport in terrestrial ecosystems at Rocky Flats Plant with emphasis on biological effects on mule deer and coyotes

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

Whicker, F.W.

1977-08-01

This report summarizes project activities during the period August 1, 1976 through July 31, 1977. Four major areas of effort are reported, namely plutonium behavior in a terrestrial ecosystem at Rocky Flats, mule deer and coyote studies at Rocky Flats, ecological consequences of transuranics in the terrestrial environment, and lead geochemistry of an alpine lake ecosystem. Much of the first area of effort involved the synthesis of data and preparation of manuscripts, although some new data are reported on plutonium levels in small mammals, plant uptake of plutonium from contaminated soil, and plutonium deposition rates on macroplot 1. The mulemore » deer studies generated a substantial body of new information which will permit quantitative assessment of plutonium dispersion by deer that utilize contaminated areas. These studies involve population dynamics, movement and use patterns, food habits, ingestion rates of contaminated soil and vegetation and plutonium burdens of deer tissues. A related study of coyote food habits in summer at Rocky Flats is reported. A manuscript dealing with the question of ecological effects of transuranics was prepared. This manuscript incorporates data from Rocky Flats on characteristics of natural populations which occupy ecologically similar areas having differing levels of plutonium contamination. The lead geochemistry studies continued to generate new data but the data are not yet reported.« less

Managing Waste Inventory and License Limits at the Perma-Fix Northwest Facility to Meet CH2M Hill Plateau Remediation Company (CHPRC) American Recovery and Reinvestment Act (ARRA) Deliverables - 12335

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

Moak, Don J.; Grondin, Richard L.; Triner, Glen C.

CH2M Hill Plateau Remediation Company (CHRPC) is a prime contractor to the U.S. Department of Energy (DOE) focused on the largest ongoing environmental remediation project in the world at the DOE Hanford Site Central Plateau, i.e. the DOE Hanford Plateau Remediation Contract. The East Tennessee Materials and Energy Corporation (M and EC); a wholly owned subsidiary of Perma-Fix Environmental Services, Inc. (PESI), is a small business team member to CHPRC. Our scope includes project management; operation and maintenance of on-site storage, repackaging, treatment, and disposal facilities; and on-site waste management including waste receipt from generators and delivery to on-site andmore » off-site treatment, storage, and disposal facilities. As part of this scope, M and EC staffs the centralized Waste Support Services organization responsible for all waste characterization and acceptance required to support CHPRC and waste generators across the Hanford Site. At the time of the CHPRC contract award (August 2008) slightly more than 9,000 cubic meters (m{sup 3}) of legacy waste was defined as 'no-path-forward waste'. A significant portion of this waste (7,650 m{sup 3}) comprised wastes with up to 50 grams of special nuclear materials (SNM) in oversized packages recovered during retrieval operations and large glove boxes removed from the Plutonium Finishing Plant (PFP). Through a collaborative effort between the DOE, CHPRC, and Perma-Fix Environmental Services, Inc. (PESI), pathways for these problematic wastes were developed that took advantage of commercial treatment capabilities at a nearby vendor facility, Perma-Fix Northwest (PFNW). In the spring of 2009, CHPRC initiated a pilot program under which they began shipping large package, low gram suspect TRU (<15 g SNM per container), and large package contact and remote handled MLLW to the off-site PFNW facility for treatment. PFNW is restricted by the SNM limits set for the total quantity of SNM allowed at the facility in accordance with the facility's radioactive materials license(s) (RML). While both CHPRC and PFNW maintain waste databases to track all waste movements, it became evident early in the process that a tool was needed that married the two systems to better track SNM inventories and sequence waste from the point of generation, through the PFNW facility, and back to the Hanford site for final disposition. This tool, known as the Treatment Integration and Planning Tool (TIPT), has become a robust planning tool that provides real-time data to support compliant and efficient waste generation, transportation, treatment, and disposition. TIPT is developing into the next generation tool that will change the way in which legacy wastes, retrieval wastes and decontamination and decommissioning operations are conducted on the Plateau Remediation Contract (PRC). The real value of the TIPT is its predictive capability. It allows the W and FMP to map out optimal windows for processing waste through the PFNW facility, or through any process that is in some way resource limited. It allows project managers to identify and focus on problem areas before shipments are affected. It has been modified for use in broader applications to predict turnaround times and identify windows of opportunity for processing higher gram wastes through PFNW and to allow waste generators, site-wide, to accurately predict scope, cost, and schedule for waste generation to optimize processing and eliminate storage, double handling, and related costs and unnecessary safety risks. The TIPT addresses the years old problem of how to effectively predict not only what needs to be done, but when. 'When' is the key planning parameter that has been ignored by the generator and processor for many years, but has proven to be the most important parameter for both parties. While further refinement is a natural part of any development process, the current improvements on the TIPT have shown that prediction is a powerful consideration. Even in lean times expected for the foreseeable future, the improved TIPT continues to play a central role in managing our way through those times to assure facilities remain viable and available. It is recommended that other major remediation projects and waste processing facilities incorporate a tool such as TIPT to improve customer-commercial supplier communications and better optimization of resources. (authors)« less

COMPLEX FLUORIDES OF PLUTONIUM AND AN ALKALI METAL

DOEpatents

Seaborg, G.T.

1960-08-01

A method is given for precipitating alkali metal plutonium fluorides. such as KPuF/sub 5/, KPu/sub 2/F/sub 9/, NaPuF/sub 5/, and RbPuF/sub 5/, from an aqueous plutonium(IV) solution by adding hydrogen fluoride and alkali-metal- fluoride.

DELTA PHASE PLUTONIUM ALLOYS

DOEpatents

Cramer, E.M.; Ellinger, F.H.; Land. C.C.

1960-03-22

Delta-phase plutonium alloys were developed suitable for use as reactor fuels. The alloys consist of from 1 to 4 at.% zinc and the balance plutonium. The alloys have good neutronic, corrosion, and fabrication characteristics snd possess good dimensional characteristics throughout an operating temperature range from 300 to 490 deg C.

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.

EPA Method: Rapid Radiochemical Method for Americium-241, Radium-226, Plutonium-238/-239, Radiostronium, and Isotopic Uranium in Water for Environmental Restoration Following Homeland Security Events

EPA Pesticide Factsheets

SAM lists this method for the qualitative determination of Americium-241, Radium-226, Plutonium-238, Plutonium-239 and isotopic uranium in drinking water samples using alpha spectrometry and radiostrontium using beta counting.

METHOD FOR OBTAINING PLUTONIUM METAL FROM ITS TRICHLORIDE

DOEpatents

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

1962-08-14

A method was developed for obtaining plutonium metal by direct reduction of plutonium chloride, without the use of a booster, using calcium and lanthamum as a reductant, the said reduction being carried out at temperature in the range of 700 to 850 deg C and at about atmospheric pressure. (AEC)

MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

DOEpatents

Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

1962-06-26

A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

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.

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.

Rapid Method for Sodium Hydroxide Fusion of Concrete and ...

EPA Pesticide Factsheets

Technical Fact Sheet Analysis Purpose: Qualitative analysis Technique: Alpha spectrometry Method Developed for: Americium-241, plutonium-238, plutonium-239, radium-226, strontium-90, uranium-234, uranium-235 and uranium-238 in concrete and brick samples Method Selected for: SAM lists this method for qualitative analysis of americium-241, plutonium-238, plutonium-239, radium-226, strontium-90, uranium-234, uranium-235 and uranium-238 in concrete or brick building materials. Summary of subject analytical method which will be posted to the SAM website to allow access to the method.

METHOD OF SEPARATING PLUTONIUM FROM LANTHANUM FLUORIDE CARRIER

DOEpatents

Watt, G.W.; Goeckermann, R.H.

1958-06-10

An improvement in oxidation-reduction type methods of separating plutoniunn from elements associated with it in a neutron-irradiated uranium solution is described. The method relates to the separating of plutonium from lanthanum ions in an aqueous 0.5 to 2.5 N nitric acid solution by 'treating the solution, at room temperature, with ammonium sulfite in an amount sufficient to reduce the hexavalent plutonium present to a lower valence state, and then treating the solution with H/sub 2/O/sub 2/ thereby forming a tetravalent plutonium peroxide precipitate.

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.

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

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/.

SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS

DOEpatents

Nicholls, C.M.; Wells, I.; Spence, R.

1959-10-13

The separation of uranium and plutonium from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in nitric acid to provide an aqueous solution 3N in nitric acid. The fission products of the solution are extruded by treating the solution with dibutyl carbitol substantially 1.8N in nitric acid. The organic solvent phase is separated and neutralized with ammonium hydroxide and the plutonium reduced with hydroxylamine base to the trivalent state. Treatment of the mixture with saturated ammonium nitrate extracts the reduced plutonium and leaves the uranium in the organic solvent.

Radiation from plutonium 238 used in space applications

NASA Technical Reports Server (NTRS)

Keenan, T. K.; Vallee, R. E.; Powers, J. A.

1972-01-01

The principal mode of the nuclear decay of plutonium 238 is by alpha particle emission at a rate of 17 curies per gram. Gamma radiation also present in nuclear fuels arises primarily from the nuclear de-excitation of daughter nuclei as a result of the alpha decay of plutonium 238 and reactor-produced impurities. Plutonium 238 has a spontaneous fission half life of 4.8 x 10 to the 10th power years. Neutrons associated with this spontaneous fission are emitted at a rate of 28,000 neutrons per second per gram. Since the space fuel form of plutonium 238 is the oxide pressed into a cermet with molybdenum, a contribution to the neutron emission rate arises from (alpha, n) reactions with 0-17 and 0-18 which occur in natural oxygen.

Evaluating ligands for use in polymer ligand film (PLF) for plutonium and uranium extraction

DOE PAGES

Rim, Jung H.; Peterson, Dominic S.; Armenta, Claudine E.; ...

2015-05-08

We describe a new analyte extraction technique using Polymer Ligand Film (PLF). PLFs were synthesized to perform direct sorption of analytes onto its surface for direct counting using alpha spectroscopy. The main focus of the new technique is to shorten and simplify the procedure for chemically isolating radionuclides for determination through a radiometric technique. 4'(5')-di-t-butylcyclohexano 18-crown-6 (DtBuCH 18C 6) and 2-ethylhexylphosphonic acid (HEH[EHP]) were examined for plutonium extraction. Di(2-ethyl hexyl) phosphoric acid (HDEHP) were examined for plutonium and uranium extraction. DtBuCH 18C 6 and HEH[EHP] were not effective in plutonium extraction. HDEHP PLFs were effective for plutonium but not formore » uranium.« 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.

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.

Structures of plutonium coordination compounds: A review of past work, recent single crystal x-ray diffraction results, and what we're learning about plutonium coordination chemistry

NASA Astrophysics Data System (ADS)

Neu, M. P.; Matonic, J. H.; Smith, D. M.; Scott, B. L.

2000-07-01

The compounds we have isolated and characterized include plutonium(III) and plutonium(IV) bound by ligands with a range of donor types and denticity (halide, phosphine oxide, hydroxamate, amine, sulfide) in a variety of coordination geometries. For example, we have obtained the first X-ray structure of Pu(III) complexed by a soft donor ligand. Using a "one pot" synthesis beginning with Pu metal strips and iodine in acetonitrile and adding trithiacyclononane we isolated the complex, PuI3(9S3)(MeCN)2 (Figure 1). On the other end of the coordination chemistry spectrum, we have obtained the first single crystal structure of the Pu(IV) hexachloro anion (Figure 2). Although this species has been used in plutonium purification via anion exchange chromatography for decades, the bond distances and exact structure were not known. We have also characterized the first plutonium-biomolecule complex, Pu(IV) bound by the siderophore desferrioxamine E.In this presentation we will review the preparation, structures, and importance of previously known coordination compounds and of those we have recently isolated. We will show the coordination chemistry of plutonium is rich and varied, well worth additional exploration.

Radiation damage and annealing in plutonium tetrafluoride

DOE PAGES

McCoy, Kaylyn; Casella, Amanda; Sinkov, Sergey; ...

2017-08-03

A sample of plutonium tetrafluoride that was separated prior to 1966 at the Hanford Site in Washington State was analyzed at the Pacific Northwest National Laboratory (PNNL) in 2015 and 2016. The plutonium tetrafluoride, as received, was an unusual color and considering the age of the plutonium, there were questions about the condition of the material. These questions had to be answered in order to determine the suitability of the material for future use or long-term storage. Therefore, thermogravimetric/differential thermal analysis and X-ray diffraction evaluations were conducted to determine the plutonium's crystal structure, oxide content, and moisture content; these analysesmore » reported that the plutonium was predominately amorphous and tetrafluoride, with an oxide content near ten percent. Freshly fluorinated plutonium tetrafluoride is known to be monoclinic. And during the initial thermogravimetric/differential thermal analyses, it was discovered that an exothermic event occurred within the material near 414 °C. X-ray diffraction analyses were conducted on the annealed tetrafluoride. The X-ray diffraction analyses indicated that some degree of recrystallization occurred in conjunction with the 414 °C event. This commentary describes the series of thermogravimetric/differential thermal and X-ray diffraction analyses that were conducted as part of this investigation at PNNL.« 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

Neutronics calculations on the impact of burnable poisons to safety and non-proliferation aspects of inert matrix fuel

NASA Astrophysics Data System (ADS)

Pistner, C.; Liebert, W.; Fujara, F.

2006-06-01

Inert matrix fuels (IMF) with plutonium may play a significant role to dispose of stockpiles of separated plutonium from military or civilian origin. For reasons of reactivity control of such fuels, burnable poisons (BP) will have to be used. The impact of different possible BP candidates (B, Eu, Er and Gd) on the achievable burnup as well as on safety and non-proliferation aspects of IMF are analyzed. To this end, cell burnup calculations have been performed and burnup dependent reactivity coefficients (boron worth, fuel temperature and moderator void coefficient) were calculated. All BP candidates were analyzed for one initial BP concentration and a range of different initial plutonium-concentrations (0.4-1.0 g cm-3) for reactor-grade plutonium isotopic composition as well as for weapon-grade plutonium. For the two most promising BP candidates (Er and Gd), a range of different BP concentrations was investigated to study the impact of BP concentration on fuel burnup. A set of reference fuels was identified to compare the performance of uranium-fuels, MOX and IMF with respect to (1) the fraction of initial plutonium being burned, (2) the remaining absolute plutonium concentration in the spent fuel and (3) the shift in the isotopic composition of the remaining plutonium leading to differences in the heat and neutron rate produced. In the case of IMF, the remaining Pu in spent fuel is unattractive for a would be proliferator. This underlines the attractiveness of an IMF approach for disposal of Pu from a non-proliferation perspective.

Comparisons of the skeletal locations of putative plutonium-induced osteosarcomas in humans with those in beagle dogs and with naturally occurring tumors in both species.

PubMed

Miller, Scott C; Lloyd, Ray D; Bruenger, Fred W; Krahenbuhl, Melinda P; Polig, Erich; Romanov, Sergey A

2003-11-01

Osteosarcomas occur from exposures to bone-seeking, alpha-particle-emitting isotopes, particularly plutonium. The skeletal distribution of putative 239Pu-induced osteosarcomas reported in Mayak Metallurgical and Radiochemical Plutonium Plant workers is compared with those observed in canine studies, and these are compared with distributions of naturally occurring osteosarcomas in both species. In the Mayak workers, 29% and 71% of the osteosarcomas were in the peripheral and central skeleton, respectively, with the spine having the most tumors (36%). An almost identical distribution of plutonium-induced osteosarcomas was reported for dogs injected with 239Pu as young adults. This distribution of osteosarcomas is quite different from the distributions of naturally occurring osteosarcomas for both species. In the Cooperative Osteosarcoma Study Group in humans (1,736 osteosarcomas from all ages), over 91% of the tumors occurred in the peripheral skeleton. In the Mayo Clinic group of older individuals (>40 years old), over 60% of the osteosarcomas appeared in the peripheral skeleton. The distribution of naturally occurring osteosarcomas in the canine is similar to that in the adult human. The similarities of the distributions of plutonium-associated osteosarcomas in the Mayak workers with those found in experimental studies suggest that many of the reported osteosarcomas may have been associated with plutonium exposures. These results also support the experimental paradigm that plutonium osteosarcomas have a preference for well vascularized cancellous bone sites. These sites have a greater initial deposition of plutonium, but also greater turnover due to elevated bone remodeling rates.

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

NASA Astrophysics Data System (ADS)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

PLUTONIUM-CERIUM-COPPER ALLOYS

DOEpatents

Coffinberry, A.S.

1959-05-12

A low melting point plutonium alloy useful as fuel is a homogeneous liquid metal fueled nuclear reactor is described. Vessels of tungsten or tantalum are useful to contain the alloy which consists essentially of from 10 to 30 atomic per cent copper and the balance plutonium and cerium. with the plutontum not in excess of 50 atomic per cent.

10 CFR 71.88 - Air transport of plutonium.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for...

10 CFR 71.88 - Air transport of plutonium.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for...

10 CFR 71.88 - Air transport of plutonium.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for...

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

10. VIEW OF CALCINER IN ROOM 146148. THE CALCINER HEATED ...

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

10. VIEW OF CALCINER IN ROOM 146-148. THE CALCINER HEATED PLUTONIUM PEROXIDE TO CONVERT IT TO PLUTONIUM OXIDE. THE PROCESS REMOVED RESIDUAL WATER AND NITRIC ACID LEAVING A DRY, POWDERED PRODUCT. (4/29/65) - Rocky Flats Plant, Plutonium Recovery & Fabrication Facility, North-central section of plant, Golden, Jefferson County, CO

10 CFR 71.88 - Air transport of plutonium.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for...

Removal of plutonium and americium from alkaline waste solutions

DOEpatents

Schulz, Wallace W.

1979-01-01

High salt content, alkaline waste solutions containing plutonium and americium are contacted with a sodium titanate compound to effect removal of the plutonium and americium from the alkaline waste solution onto the sodium titanate and provide an effluent having a radiation level of less than 10 nCi per gram alpha emitters.

10 CFR 71.88 - Air transport of plutonium.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Air transport of plutonium. 71.88 Section 71.88 Energy... Controls and Procedures § 71.88 Air transport of plutonium. (a) Notwithstanding the provisions of any..., whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for...

PREPARATION OF HALIDES OF PLUTONIUM

DOEpatents

Garner, C.S.; Johns, I.B.

1958-09-01

A dry chemical method is described for preparing plutonium halides, which consists in contacting plutonyl nitrate with dry gaseous HCl or HF at an elevated temperature. The addition to the reaction gas of a small quantity of an oxidizing gas or a reducing gas will cause formation of the tetra- or tri-halide of plutonium as desired.

SEPARATION OF FISSION PRODUCT VALUES FROM THE HEXAVALENT PLUTONIUM BY CARRIER PRECIPITATION

DOEpatents

Davies, T.H.

1959-12-15

An improved precipitation of fission products on bismuth phosphate from an aqueous mineral acid solution also containing hexavalent plutonium by incorporating, prior to bismuth phosphate precipitation, from 0.05 to 2.5 grams/ liter of zirconium phosphate, niobium oxide. and/or lanthanum fluoride is described. The plutonium remains in solution.

Enhanced ionization efficiency in TIMS analyses of plutonium and americium using porous ion emitters

DOE PAGES

Baruzzini, Matthew L.; Hall, Howard L.; Watrous, Matthew G.; ...

2016-12-05

Investigations of enhanced sample utilization in thermal ionization mass spectrometry (TIMS) using porous ion emitter (PIE) techniques for the analyses of trace quantities of americium and plutonium were performed. Repeat ionization efficiency (i.e., the ratio of ions detected to atoms loaded on the filament) measurements were conducted on sample sizes ranging from 10–100 pg for americium and 1–100 pg for plutonium using PIE and traditional (i.e., a single, zone-refined rhenium, flat filament ribbon with a carbon ionization enhancer) TIMS filament sources. When compared to traditional filaments, PIEs exhibited an average boost in ionization efficiency of ~550% for plutonium and ~1100%more » for americium. A maximum average efficiency of 1.09% was observed at a 1 pg plutonium sample loading using PIEs. Supplementary trials were conducted using newly developed platinum PIEs to analyze 10 pg mass loadings of plutonium. As a result, platinum PIEs exhibited an additional ~134% boost in ion yield over standard PIEs and ~736% over traditional filaments at the same sample loading level.« 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

Real-time monitoring of plutonium content in uranium-plutonium alloys

DOEpatents

Li, Shelly Xiaowei; Westphal, Brian Robert; Herrmann, Steven Douglas

2015-09-01

A method and device for the real-time, in-situ monitoring of Plutonium content in U--Pu Alloys comprising providing a crucible. The crucible has an interior non-reactive to a metallic U--Pu alloy within said interior of said crucible. The U--Pu alloy comprises metallic uranium and plutonium. The U--Pu alloy is heated to a liquid in an inert or reducing atmosphere. The heated U--Pu alloy is then cooled to a solid in an inert or reducing atmosphere. As the U--Pu alloy is cooled, the temperature of the U--Pu alloy is monitored. A solidification temperature signature is determined from the monitored temperature of the U--Pu alloy during the step of cooling. The amount of Uranium and the amount of Plutonium in the U--Pu alloy is then determined from the determined solidification temperature signature.

Plutonium segregation in glassy aerodynamic fallout from a nuclear weapon test

DOE PAGES

Holliday, K. S.; Dierken, J. M.; Monroe, M. L.; ...

2017-01-11

Our study combines electron microscopy equipped with energy dispersive spectroscopy to probe major element composition and autoradiography to map plutonium in order to examine the spatial relationships between plutonium and fallout composition in aerodynamic glassy fallout from a nuclear weapon test. We interrogated a sample set of 48 individual fallout specimens in order to reveal that the significant chemical heterogeneity of this sample set could be described compositionally with a relatively small number of compositional endmembers. Furthermore, high concentrations of plutonium were never associated with several endmember compositions and concentrated with the so-called mafic glass endmember. Our result suggests thatmore » it is the physical characteristics of the compositional endmembers and not the chemical characteristics of the individual component elements that govern the un-burnt plutonium distribution with respect to major element composition in fallout.« less

7 CFR 984.456 - Disposition of reserve walnuts and walnuts used for reserve disposition credit.

Code of Federal Regulations, 2010 CFR

2010-01-01

... reserve disposition credit. 984.456 Section 984.456 Agriculture Regulations of the Department of....456 Disposition of reserve walnuts and walnuts used for reserve disposition credit. (a) Beginning... walnuts in authorized outlets. (b) Any handler who desires to transfer disposition credit in excess of his...

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

Plutonium Oxidation State Distribution under Aerobic and Anaerobic Subsurface Conditions for Metal-Reducing Bacteria

NASA Astrophysics Data System (ADS)

Reed, D. T.; Swanson, J.; Khaing, H.; Deo, R.; Rittmann, B.

2009-12-01

The fate and potential mobility of plutonium in the subsurface is receiving increased attention as the DOE looks to cleanup the many legacy nuclear waste sites and associated subsurface contamination. Plutonium is the near-surface contaminant of concern at several DOE sites and continues to be the contaminant of concern for the permanent disposal of nuclear waste. The mobility of plutonium is highly dependent on its redox distribution at its contamination source and along its potential migration pathways. This redox distribution is often controlled, especially in the near-surface where organic/inorganic contaminants often coexist, by the direct and indirect effects of microbial activity. The redox distribution of plutonium in the presence of facultative metal reducing bacteria (specifically Shewanella and Geobacter species) was established in a concurrent experimental and modeling study under aerobic and anaerobic conditions. Pu(VI), although relatively soluble under oxidizing conditions at near-neutral pH, does not persist under a wide range of the oxic and anoxic conditions investigated in microbiologically active systems. Pu(V) complexes, which exhibit high chemical toxicity towards microorganisms, are relatively stable under oxic conditions but are reduced by metal reducing bacteria under anaerobic conditions. These facultative metal-reducing bacteria led to the rapid reduction of higher valent plutonium to form Pu(III/IV) species depending on nature of the starting plutonium species and chelating agents present in solution. Redox cycling of these lower oxidation states is likely a critical step in the formation of pseudo colloids that may lead to long-range subsurface transport. The CCBATCH biogeochemical model is used to explain the redox mechanisms and final speciation of the plutonium oxidation state distributions observed. These results for microbiologically active systems are interpreted in the context of their importance in defining the overall migration of plutonium in the subsurface.

Technical Review Report for the Mound 1KW Package Safety Analysis Report for Packaging Waiver for the Use of Modified Primary Containment Vessel (PCV)

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

West, M; Hafner, R

2008-05-05

This Technical Review Report (TRR) documents the review, performed by the Lawrence Livermore National Laboratory (LLNL) staff, at the request of the U.S. Department of Energy (DOE), on the Waiver for the Use of Modified Primary Containment Vessels (PCV). The waiver is to be used to support a limited number of shipments of fuel for the Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) Project in support of the National Aeronautics and Space Administration's (NASA's) Mars Science Laboratory (MSL) mission. Under the waiver, an inventory of existing national security PCVs will be converted to standard PCVs. Both types of PCVs are currently approvedmore » for use by the Office of Nuclear Energy. LLNL has previously reviewed the national security PCVs under Mound 1KW Package Safety Analysis Report for Packaging, Addendum No. 1, Revision c, dated June 2007 (Addendum 1). The safety analysis of the package is documented in the Safety Analysis Report for Packaging (SARP) for the Mound 1KW Package (i.e., the Mound 1KW SARP, or the SARP) where the standard PCVs have been reviewed by LLNL. The Mound 1KW Package is certified by DOE Certificate of Compliance (CoC) number USA/9516/B(U)F-85 for the transportation of Type B quantities of plutonium heat source material. The waiver requests an exemption, claiming safety equivalent to the requirements specified in 10 CFR 71.12, Specific Exemptions, and will lead to a letter amendment to the CoC. Under the waiver, the Office of Radioisotope Power Systems, NE-34, is seeking an exemption from 10 CFR 71.19(d)(1), Previously Approved Package,[5] which states: '(d) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided--(1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified in {section}71.71 and 71.73.' The LLNL staff had previously reviewed a request from Idaho National Laboratory (INL) to reconfigure national security PCVs to standard PCVs. With a nominal 50% reduction in both the height and the volume, the LLNL staff initially deemed the modifications to be significant, which would not be allowed under the provisions of 10 CFR 71.19(d)(1)--see above. As a follow-up, the DOE requested additional clarification from the Nuclear Regulatory Commission (NRC). The NRC concluded that the reconfiguration would be a new fabrication, and that an exemption to the regulations would be required to allow its use, as per the requirements specified in 10 CFR 71.19(c)(1), Previously Approved Package: '(c) A Type B(U) package, a Type B(M) package, or a fissile material package previously approved by the NRC with the designation '-85' in the identification number of the NRC CoC, may be used under the general license of {section}71.17 with the following additional conditions: (1) Fabrication of the package must be satisfactorily completed by December 31, 2006, as demonstrated by application of its model number in accordance with 71.85(c).' Although the preferred approach toward the resolution of this issue would be for the applicant to submit an updated SARP, the applicant has stated that the process of updating the Model Mound 1KW Package SARP is a work that is in progress, but that the updated SARP is not yet ready for submittal. The applicant has to provide a submittal, proving that the package meets the '-96' requirements of International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1, in order to fabricate approved packagings after December 31, 2006. The applicant has further stated that all other packaging features, as described in the currently approved Model Mound 1KW Package SARP, remain unchanged. This report documents the LLNL review of the waiver request. The specific review for each SARP Chapter is documented.« less

Estimation of Plutonium-240 Mass in Waste Tanks Using Ultra-Sensitive Detection of Radioactive Xenon Isotopes from Spontaneous Fission

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

Bowyer, Theodore W.; Gesh, Christopher J.; Haas, Daniel A.

This report details efforts to develop a technique which is able to detect and quantify the mass of 240Pu in waste storage tanks and other enclosed spaces. If the isotopic ratios of the plutonium contained in the enclosed space is also known, then this technique is capable of estimating the total mass of the plutonium without physical sample retrieval and radiochemical analysis of hazardous material. Results utilizing this technique are reported for a Hanford Site waste tank (TX-118) and a well-characterized plutonium sample in a laboratory environment.

PLUTONIUM ELECTROREFINING CELLS

DOEpatents

Mullins, L.J. Jr.; Leary, J.A.; Bjorklund, C.W.; Maraman, W.J.

1963-07-16

Electrorefining cells for obtaining 99.98% plutonium are described. The cells consist of an impure liquid plutonium anode, a molten PuCl/sub 3/-- alkali or alkaline earth metal chloanode, a molten PuCl/sub 3/-alkali or alkaline earth metal chloride electrolyte, and a nonreactive cathode, all being contained in nonreactive ceramic containers which separate anode from cathode by a short distance and define a gap for the collection of the purified liquid plutonium deposited on the cathode. Important features of these cells are the addition of stirrer blades on the anode lead and a large cathode surface to insure a low current density. (AEC)

PLUTONIUM SEPARATION METHOD

DOEpatents

Beaufait, L.J. Jr.; Stevenson, F.R.; Rollefson, G.K.

1958-11-18

The recovery of plutonium ions from neutron irradiated uranium can be accomplished by bufferlng an aqueous solutlon of the irradiated materials containing tetravalent plutonium to a pH of 4 to 7, adding sufficient acetate to the solution to complex the uranyl present, adding ferric nitrate to form a colloid of ferric hydroxide, plutonlum, and associated fission products, removing and dissolving the colloid in aqueous nitric acid, oxldizlng the plutonium to the hexavalent state by adding permanganate or dichromate, treating the resultant solution with ferric nitrate to form a colloid of ferric hydroxide and associated fission products, and separating the colloid from the plutonlum left in solution.

Development of first ever scanning probe microscopy capabilities for plutonium

NASA Astrophysics Data System (ADS)

Beaux, Miles F.; Cordoba, Miguel Santiago; Zocco, Adam T.; Vodnik, Douglas R.; Ramos, Michael; Richmond, Scott; Moore, David P.; Venhaus, Thomas J.; Joyce, Stephen A.; Usov, Igor O.

2017-04-01

Scanning probe microscopy capabilities have been developed for plutonium and its derivative compounds. Specifically, a scanning tunneling microscope and an atomic force microscope housed in an ultra-high vacuum system and an inert atmosphere glove box, respectively, were prepared for the introduction of small non-dispersible δ-Pu coupons. Experimental details, procedures, and preliminary imaging of δ-Pu coupons are presented to demonstrate the functionality of these new capabilities. These first of a kind capabilities for plutonium represent a significant step forward in the ability to characterize and understand plutonium surfaces with high spatial resolution.

Development of first ever scanning probe microscopy capabilities for plutonium

DOE PAGES

Beaux, Miles F.; Cordoba, Miguel Santiago; Zocco, Adam T.; ...

2017-04-01

Scanning probe microscopy capabilities have been developed for plutonium and its derivative compounds. Specifically, a scanning tunneling microscope and an atomic force microscope housed in an ultra-high vacuum system and an inert atmosphere glove box, respectively, were prepared for the introduction of small non-dispersible δ-Pu coupons. Experimental details, procedures, and preliminary imaging of δ-Pu coupons are presented to demonstrate the functionality of these new capabilities. In conclusion, these first of a kind capabilities for plutonium represent a significant step forward in the ability to characterize and understand plutonium surfaces with high spatial resolution.

Development of a Universal Canister for Disposal of High-Level Waste in Deep Boreholes.

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

Price, Laura L.; Gomberg, Steve

2015-11-01

The mission of the United States Department of Energy’s Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research. Some of the wastes that must be managed have been identified as good candidates for disposal in a deep borehole in crystalline rock. In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister-based system that can be used for handling these wastes during the disposition process (i.e., storage, transfer, transportation, and disposal)more » could facilitate the eventual disposal of these wastes. Development of specifications for the universal canister system will consider the regulatory requirements that apply to storage, transportation, and disposal of the capsules, as well as operational requirements and limits that could affect the design of the canister (e.g., deep borehole diameter). In addition, there are risks and technical challenges that need to be recognized and addressed as Universal Canister system specifications are developed. This paper provides an approach to developing specifications for such a canister system that is integrated with the overall efforts of the DOE’s Used Fuel Disposition Campaign's Deep Borehole Field Test and compatible with planned storage of potential borehole-candidate wastes.« less

Advance assessment for movement of Haz Cat 3 radioactive materials.

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

Vosburg, Susan K.

2010-04-01

The current packaging of most HC-3 radioactive materials at SNL/NM do not meet DOT requirements for offsite shipment. SNL/NM is transporting HC-3 quantities of radioactive materials from their storage locations in the Manzano Nuclear Facilities bunkers to facilities in TA-5 to be repackaged for offsite shipment. All transportation of HC-3 rad material by SNL/NM is onsite (performed within the confines of KAFB). Transport is performed only by the Regulated Waste/Nuclear Material Disposition Department. Part of the HC3T process is to provide the CAT with the following information at least three days prior to the move: (1) RFt-Request for transfer; (2)more » HC3T movement report; (3) Radiological survey; and (4) Transportation Route Map.« less

Evaluation of Used Fuel Disposition in Clay-Bearing Rock

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

Jove-Colon, Carlos F.; Weck, Philippe F.; Hammond, Glenn Edward

Deep geological disposal of nuclear waste in clay/shale/argillaceous rock formations has received much consideration given its desirable attributes such as isolation properties (low permeability), geochemically reduced conditions, slow diffusion, sorbtive mineralogy, and geologically widespread (Jové Colón et al., 2014). There is a wealth of gained scientific expertise on the behavior of clay/shale/ argillaceous rock given its focus in international nuclear waste repository programs that includes underground research laboratories (URLs) in Switzerland, France, Belgium, and Japan. Jové Colón et al. (2014) have described some of these investigative efforts in clay rock ranging from site characterization to research on the engineered barriermore » system (EBS). Evaluations of disposal options that include nuclear waste disposition in clay/shale/argillaceous rock have determined that this host media can accommodate a wide range of waste types. R&D work within the Used Fuel Disposition Campaign (UFDC) assessing thermal effects and fluid-mineral interactions for the disposition of heat-generating waste have so far demonstrated the feasibility for the EBS and clay host rock to withstand high thermal loads. This report represents the continuation of disposal R&D efforts on the advancement and refinement of coupled Thermal-Hydrological-Mechanical-Chemical (THMC), hydrothermal experiments on clay interactions, used fuel degradation (source term), and thermodynamic modeling and database development. The development and implementation of a clay/shale/argillite reference case described in Jové Colón et al. (2014) for FY15 will be documented in another report (Mariner et al. 2015) – only a brief description will be given here. This clay reference case implementation is the result of integration efforts between the GDSA PA and disposal in argillite work packages. The assessment of sacrificial zones in the EBS is being addressed through experimental work along with 1D reactive-transport and reaction path modeling. The focus of these investigations into the nature of sacrificial zones is to evaluate the chemical effects of heterogeneous chemical reactions at EBS interfaces. The difference in barrier material types and the extent of chemical reactions within these interfacial domains generates changes in mineral abundances. These mineralogical alterations also result in volume changes that, although small, could affect the interface bulk porosity. As in previous deliverables, this report is structured according to various national laboratory contributions describing R&D activities applicable to clay/shale/argillite media.« 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.

PLUTONIUM-URANIUM-TITANIUM ALLOYS

DOEpatents

Coffinberry, A.S.

1959-07-28

A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

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.

Nuclear Methods for Transmutation of Nuclear Waste: Problems, Perspextives, Cooperative Research - Proceedings of the International Workshop

NASA Astrophysics Data System (ADS)

Khankhasayev, Zhanat B.; Kurmanov, Hans; Plendl, Mikhail Kh.

1996-12-01

The Table of Contents for the full book PDF is as follows: * Preface * I. Review of Current Status of Nuclear Transmutation Projects * Accelerator-Driven Systems — Survey of the Research Programs in the World * The Los Alamos Accelerator-Driven Transmutation of Nuclear Waste Concept * Nuclear Waste Transmutation Program in the Czech Republic * Tentative Results of the ISTC Supported Study of the ADTT Plutonium Disposition * Recent Neutron Physics Investigations for the Back End of the Nuclear Fuel Cycle * Optimisation of Accelerator Systems for Transmutation of Nuclear Waste * Proton Linac of the Moscow Meson Factory for the ADTT Experiments * II. Computer Modeling of Nuclear Waste Transmutation Methods and Systems * Transmutation of Minor Actinides in Different Nuclear Facilities * Monte Carlo Modeling of Electro-nuclear Processes with Nonlinear Effects * Simulation of Hybrid Systems with a GEANT Based Program * Computer Study of 90Sr and 137Cs Transmutation by Proton Beam * Methods and Computer Codes for Burn-Up and Fast Transients Calculations in Subcritical Systems with External Sources * New Model of Calculation of Fission Product Yields for the ADTT Problem * Monte Carlo Simulation of Accelerator-Reactor Systems * III. Data Basis for Transmutation of Actinides and Fission Products * Nuclear Data in the Accelerator Driven Transmutation Problem * Nuclear Data to Study Radiation Damage, Activation, and Transmutation of Materials Irradiated by Particles of Intermediate and High Energies * Radium Institute Investigations on the Intermediate Energy Nuclear Data on Hybrid Nuclear Technologies * Nuclear Data Requirements in Intermediate Energy Range for Improvement of Calculations of ADTT Target Processes * IV. Experimental Studies and Projects * ADTT Experiments at the Los Alamos Neutron Science Center * Neutron Multiplicity Distributions for GeV Proton Induced Spallation Reactions on Thin and Thick Targets of Pb and U * Solid State Nuclear Track Detector and Radiochemical Studies on the Transmutation of Nuclei Using Relativistic Heavy Ions * Experimental and Theoretical Study of Radionuclide Production on the Electronuclear Plant Target and Construction Materials Irradiated by 1.5 GeV and 130 MeV Protons * Neutronics and Power Deposition Parameters of the Targets Proposed in the ISTC Project 17 * Multicycle Irradiation of Plutonium in Solid Fuel Heavy-Water Blanket of ADS * Compound Neutron Valve of Accelerator-Driven System Sectioned Blanket * Subcritical Channel-Type Reactor for Weapon Plutonium Utilization * Accelerator Driven Molten-Fluoride Reactor with Modular Heat Exchangers on PB-BI Eutectic * A New Conception of High Power Ion Linac for ADTT * Pions and Accelerator-Driven Transmutation of Nuclear Waste? * V. Problems and Perspectives * Accelerator-Driven Transmutation Technologies for Resolution of Long-Term Nuclear Waste Concerns * Closing the Nuclear Fuel-Cycle and Moving Toward a Sustainable Energy Development * Workshop Summary * List of Participants

The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

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

Eccleston, G.W.; Menlove, H.O.; Abhold, M.

1998-12-31

The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

TERNARY ALLOY-CONTAINING PLUTONIUM

DOEpatents

Waber, J.T.

1960-02-23

Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

Microprobe Analysis of Pu-Ga Standards

DOE PAGES

Wall, Angélique D.; Romero, Joseph P.; Schwartz, Daniel

2017-08-04

In order to obtain quantitative analysis using an Electron Scanning Microprobe it is essential to have a standard of known composition. Most elemental and multi-elemental standards can be easily obtained from places like Elemental Scientific or other standards organizations that are NIST (National Institute of Standards and Technology) traceable. It is, however, more challenging to find standards for plutonium. Past work performed in our group has typically involved using the plutonium sample to be analysed as its own standard as long as all other known components of the sample have standards to be compared to [1,2,3]. Finally, this method worksmore » well enough, but this experiment was performed in order to develop a more reliable standard for plutonium using five samples of known chemistry of a plutonium gallium mix that could then be used as the main plutonium and gallium standards for future experiments.« less

Radiolysis of hexavalent plutonium in solutions of uranyl nitrate containing fission product simulants

NASA Astrophysics Data System (ADS)

Rance, Peter J. W.; Zilberman, B. Ya.; Akopov, G. A.

2000-07-01

The effect of the inherent radioactivity on the chemical state of plutonium ions in solution was recognized very shortly after the first macroscopic amounts of plutonium became available and early studies were conducted as part of the Manhattan Project. However, the behavior of plutonium ions, in nitric acid especially, has been found to be somewhat complex, so much so that a relatively modern summary paper included the comment that, "The vast amount of work carried out in nitric acid solutions can not be adequately summarized. Suffice it to say results in these solutions are plagued with irreproducibility and induction periods…" Needless to say, the presence of other ions in solution, as occurs when irradiated nuclear fuel is dissolved, further complicates matters. The purpose of the work described below was to add to the rather small amount of qualitative data available relating to the radiolytic behavior of plutonium in solutions of irradiated nuclear fuel.

Microprobe Analysis of Pu-Ga Standards

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

Wall, Angélique D.; Romero, Joseph P.; Schwartz, Daniel

In order to obtain quantitative analysis using an Electron Scanning Microprobe it is essential to have a standard of known composition. Most elemental and multi-elemental standards can be easily obtained from places like Elemental Scientific or other standards organizations that are NIST (National Institute of Standards and Technology) traceable. It is, however, more challenging to find standards for plutonium. Past work performed in our group has typically involved using the plutonium sample to be analysed as its own standard as long as all other known components of the sample have standards to be compared to [1,2,3]. Finally, this method worksmore » well enough, but this experiment was performed in order to develop a more reliable standard for plutonium using five samples of known chemistry of a plutonium gallium mix that could then be used as the main plutonium and gallium standards for future experiments.« less

Applications Where Snap is BPM for Radioactive Waste Assay

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

Miller, T.J.

2008-07-01

Historically, the Atomic Weapons Establishment (AWE) at Aldermaston in the United Kingdom (UK), has used a variety of assay techniques to measure the radioactive content of a diverse range of waste packages from decommissioning, operational and legacy sources. The regulator, the Environment Agency in the UK, places conditions and limits on AWE through an authorisation within the Radioactive Substances Act (RSA93). The conditions and limits require Best Practical Means (BPM) measurements to be used to demonstrate compliance with the authorisation. Hence, the assay technique employed needs to achieve a balance between risk of exposure, environmental considerations, technological considerations, health andmore » safety considerations and cost effectiveness, without being grossly disproportionate in terms of money, time or trouble. Recently published work has concluded that the Spectral Non-destructive Assay Platform (SNAP) assay system is BPM for Depleted Uranium (DU) waste assay at AWE (1) and low level plutonium in soft drummed waste, HEPA filters and soils (2-4). The purpose of this paper is to highlight other applications where SNAP represents BPM for radioactive waste assay. This has been done by intercomparison studies of SNAP with other assay techniques, such as Segmented Gamma Scanner (SGS) and Passive Neutron Coincidence Counter (PNCC). It has been concluded that, for a large range of waste packages encountered at AWE, SNAP is BPM. (author)« less

Dispositional and Situational Avoidance and Approach as Predictors of Physical Symptom Bother Following Breast Cancer Diagnosis

PubMed Central

Bauer, Margaret R.; Harris, Lauren N.; Wiley, Joshua F.; Crespi, Catherine M.; Krull, Jennifer L.; Weihs, Karen L.; Stanton, Annette L.

2016-01-01

Background Few studies examine whether dispositional approach and avoidance coping and stressor-specific coping strategies differentially predict physical adjustment to cancer-related stress. Purpose This study examines dispositional and situational avoidance and approach coping as unique predictors of the bother women experience from physical symptoms after breast cancer treatment, as well as whether situational coping mediates the prediction of bother from physical symptoms by dispositional coping. Method Breast cancer patients (N=460) diagnosed within the past 3 months completed self-report measures of dispositional coping at study entry and of situational coping and bother from physical symptoms every 6 weeks through 6 months. Results In multilevel structural equation modeling analyses, both dispositional and situational avoidance predict greater symptom bother. Dispositional, but not situational, approach predicts less symptom bother. Supporting mediation models, dispositional avoidance predicts more symptom bother indirectly through greater situational avoidance. Dispositional approach predicts less symptom bother through less situational avoidance. Conclusion Psychosocial interventions to reduce cancer-related avoidance coping are warranted for cancer survivors who are high in dispositional avoidance and/or low in dispositional approach. PMID:26769023

Dispositional and Situational Avoidance and Approach as Predictors of Physical Symptom Bother Following Breast Cancer Diagnosis.

PubMed

Bauer, Margaret R; Harris, Lauren N; Wiley, Joshua F; Crespi, Catherine M; Krull, Jennifer L; Weihs, Karen L; Stanton, Annette L

2016-06-01

Few studies examine whether dispositional approach and avoidance coping and stressor-specific coping strategies differentially predict physical adjustment to cancer-related stress. This study examines dispositional and situational avoidance and approach coping as unique predictors of the bother women experience from physical symptoms after breast cancer treatment, as well as whether situational coping mediates the prediction of bother from physical symptoms by dispositional coping. Breast cancer patients (N = 460) diagnosed within the past 3 months completed self-report measures of dispositional coping at study entry and of situational coping and bother from physical symptoms every 6 weeks through 6 months. In multilevel structural equation modeling analyses, both dispositional and situational avoidance predict greater symptom bother. Dispositional, but not situational, approach predicts less symptom bother. Supporting mediation models, dispositional avoidance predicts more symptom bother indirectly through greater situational avoidance. Dispositional approach predicts less symptom bother through less situational avoidance. Psychosocial interventions to reduce cancer-related avoidance coping are warranted for cancer survivors who are high in dispositional avoidance and/or low in dispositional approach.

PRECIPITATION METHOD OF SEPARATING PLUTONIUM FROM CONTAMINATING ELEMENTS

DOEpatents

Duffield, R.B.

1959-02-24

S>A method is described for separating plutonium, in a valence state of less than five, from an aqueous solution in which it is dissolved. The niethod consists in adding potassium and sulfate ions to such a solution while maintaining the solution at a pH of less than 7.1, and isolating the precipitate of potassium plutonium sulfate thus formed.

Density of Plutonium Turnings Generated from Machining Activities

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

Gonzales, John Robert; Vigil, Duane M.; Jachimowski, Thomas A.

The purpose of this project was to determine the density of plutonium (Pu) turnings generated from the range of machining activities, using both surrogate material and machined Pu turnings. Verify that 500 grams (g) of plutonium will fit in a one quart container using a surrogate equivalent volume and that 100 grams of Pu will fit in a one quart Savy container.

Airborne plutonium-239 and americium-241 concentrations measured from the 125-meter Hanford Meteorological Tower

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

Sehmel, G.A.

1978-01-01

Airborne plutonium-239 and americium-241 concentrations and fluxes were measured at six heights from 1.9 to 122 m on the Hanford meteorological tower. The data show that plutonium-239 was transported on nonrespirable and small particles at all heights. Airborne americium-241 concentrations on small particles were maximum at the 91 m height.

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)

SEPARATION OF PLUTONIUM FROM URANIUM AND FISSION PRODUCTS BY ADSORPTION

DOEpatents

Seaborg, G.T.; Willard, J.E.

1958-01-01

A method is presented for the separation of plutonium from solutions containing that element in a valence state not higher than 41 together with uranium ions and fission products. This separation is accomplished by contacting the solutions with diatomaceous earth which preferentially adsorbs the plutonium present. Also mentioned as effective for this adsorbtive separation are silica gel, filler's earth and alumina.

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/.

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.

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.

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.

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.

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

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

Photoemission Spectroscopy of Delta- Plutonium: Experimental Review

NASA Astrophysics Data System (ADS)

Tobin, J. G.

2002-03-01

The electronic structure of Plutonium, particularly delta- Plutonium, remains ill defined and without direct experimental verification. Recently, we have embarked upon a program of study of alpha- and delta- Plutonium, using synchrotron radiation from the Advanced Light Source in Berkeley, CA, USA [1]. This work is set within the context of Plutonium Aging [2] and the complexities of Plutonium Science [3]. The resonant photoemission of delta-plutonium is in partial agreement with an atomic, localized model of resonant photoemission, which would be consistent with a correlated electronic structure. The results of our synchrotron- based studies will be compared with those of recent laboratory- based works [4,5,6]. The talk will conclude with a brief discussion of our plans for the future, such as the performance of spin-resolving and dichroic photoemission measurements of Plutonium [7] and the development of single crystal ultrathin films of Plutonium. This work was performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under Contract No. W-7405-Eng-48. 1. J. Terry, R.K. Schulze, J.D. Farr, T. Zocco, K. Heinzelman, E. Rotenberg, D.K. Shuh, G. van der Laan, D.A. Arena, and J.G. Tobin, “5f Resonant Photoemission from Plutonium”, UCRL-JC-140782, Surf. Sci. Lett., accepted October 2001. 2. B.D. Wirth, A.J. Schwartz, M.J. Fluss, M.J. Caturla, M.A. Wall, and W.G. Wolfer, MRS Bulletin 26, 679 (2001). 3. S.S. Hecker, MRS Bulletin 26, 667 (2001). 4. T. Gouder, L. Havela, F. Wastin, and J. Rebizant, Europhys. Lett. 55, 705 (2001); MRS Bulletin 26, 684 (2001); Phys. Rev. Lett. 84, 3378 (2000). 5. A.J. Arko, J.J. Joyce, L. Morales, J. Wills, J. Lashley, F. Wastin, and J. Rebizant, Phys. Rev. B 62, 1773 (2000). 6. L.E. Cox, O. Eriksson, and B.R. Cooper, Phys. Rev. B 46, 13571 (1992). 7. J. Tobin, D.A. Arena, B. Chung, P. Roussel, J. Terry, R.K. Schulze, J.D. Farr, T. Zocco, K. Heinzelman, E. Rotenberg, and D.K. Shuh, “Photoelectron Spectroscopy of Plutonium at the Advanced Light Source”, UCRL-JC-145703, J. Nucl. Sci. Tech./ Proc. of Actinides 2001, submitted November 2001.

Disposal of spent fuel

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

Blomeke, J O; Ferguson, D E; Croff, A G

1978-01-01

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed betweenmore » the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom.« less

Review of the Scientific Understanding of Radioactive Waste at the U.S. DOE Hanford Site

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

Peterson, Reid A.; Buck, Edgar C.; Chun, Jaehun

This paper reviews the origin and chemical and rheological complexity of radioactive waste at the U.S. Department of Energy’s Hanford Site. The waste, stored in underground tanks, was generated via three distinct processes over decades of plutonium extraction operations. Although close records were kept of original waste disposition, tank-to-tank transfers and conditions that impede equilibrium complicate our understanding of the chemistry, phase composition, and rheology of the waste. Tank waste slurries comprise particles and aggregates from nano to micron scales, with varying densities, morphologies, heterogeneous compositions, and complicated responses to flow regimes and process conditions. Further, remnant or changing radiationmore » fields may affect the stability and rheology of the waste. These conditions pose challenges for transport through conduits or pipes to treatment plants for vitrification. Additionally, recalcitrant boehmite degrades glass quality and must be reduced prior to vitrification, but dissolves much more slowly than predicted given surface normalized rates. Existing empirical models based on ex situ experiments and observations lack true predictive capabilities. Recent advances in in situ microscopy, aberration corrected TEM, theoretical modeling across scales, and experimental methods for probing the physics and chemistry at mineral-fluid and mineral-mineral interfaces are being implemented to build robustly predictive physics-based models.« less

24 CFR 290.15 - Disposition plan.

Code of Federal Regulations, 2014 CFR

2014-04-01

... DEVELOPMENT HUD-OWNED PROPERTIES DISPOSITION OF MULTIFAMILY PROJECTS AND SALE OF HUD-HELD MULTIFAMILY MORTGAGES Disposition of Multifamily Projects § 290.15 Disposition plan. (a) In general. Before disposing of a HUD-owned multifamily housing project, HUD will develop an initial and a final disposition plan...

24 CFR 290.15 - Disposition plan.

Code of Federal Regulations, 2012 CFR

2012-04-01

... DEVELOPMENT HUD-OWNED PROPERTIES DISPOSITION OF MULTIFAMILY PROJECTS AND SALE OF HUD-HELD MULTIFAMILY MORTGAGES Disposition of Multifamily Projects § 290.15 Disposition plan. (a) In general. Before disposing of a HUD-owned multifamily housing project, HUD will develop an initial and a final disposition plan...

24 CFR 290.15 - Disposition plan.

Code of Federal Regulations, 2011 CFR

2011-04-01

... DEVELOPMENT HUD-OWNED PROPERTIES DISPOSITION OF MULTIFAMILY PROJECTS AND SALE OF HUD-HELD MULTIFAMILY MORTGAGES Disposition of Multifamily Projects § 290.15 Disposition plan. (a) In general. Before disposing of a HUD-owned multifamily housing project, HUD will develop an initial and a final disposition plan...

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

Thermal radiative and thermodynamic properties of solid and liquid uranium and plutonium carbides in the visible-near-infrared range

NASA Astrophysics Data System (ADS)

Fisenko, Anatoliy I.; Lemberg, Vladimir F.

2016-09-01

The knowledge of thermal radiative and thermodynamic properties of uranium and plutonium carbides under extreme conditions is essential for designing a new metallic fuel materials for next generation of a nuclear reactor. The present work is devoted to the study of the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides at their melting/freezing temperatures. The Stefan-Boltzmann law, total energy density, number density of photons, Helmholtz free energy density, internal energy density, enthalpy density, entropy density, heat capacity at constant volume, pressure, and normal total emissivity are calculated using experimental data for the frequency dependence of the normal spectral emissivity of liquid and solid uranium and plutonium carbides in the visible-near infrared range. It is shown that the thermal radiative and thermodynamic functions of uranium carbide have a slight difference during liquid-to-solid transition. Unlike UC, such a difference between these functions have not been established for plutonium carbide. The calculated values for the normal total emissivity of uranium and plutonium carbides at their melting temperatures is in good agreement with experimental data. The obtained results allow to calculate the thermal radiative and thermodynamic properties of liquid and solid uranium and plutonium carbides for any size of samples. Based on the model of Hagen-Rubens and the Wiedemann-Franz law, a new method to determine the thermal conductivity of metals and carbides at the melting points is proposed.

Dispositions as Virtues: The Complexity of the Construct

ERIC Educational Resources Information Center

Sockett, Hugh

2009-01-01

The value of conceptualizing the desirable dispositions of the teacher as virtues is illuminated through distinguishing such dispositions-as-virtues from other dispositions and from personality traits. Dispositions as virtues are qualities achieved by the individual's initiative, in the face of obstacles, and are intrinsically motivated. The…

Future Remains: Industrial Heritage at the Hanford Plutonium Works

NASA Astrophysics Data System (ADS)

Freer, Brian

This dissertation argues that U.S. environmental and historic preservation regulations, industrial heritage projects, history, and art only provide partial frameworks for successfully transmitting an informed story into the long range future about nuclear technology and its related environmental legacy. This argument is important because plutonium from nuclear weapons production is toxic to humans in very small amounts, threatens environmental health, has a half-life of 24, 110 years and because the industrial heritage project at Hanford is the first time an entire U.S. Department of Energy weapons production site has been designated a U.S. Historic District. This research is situated within anthropological interest in industrial heritage studies, environmental anthropology, applied visual anthropology, as well as wider discourses on nuclear studies. However, none of these disciplines is really designed or intended to be a completely satisfactory frame of reference for addressing this perplexing challenge of documenting and conveying an informed story about nuclear technology and its related environmental legacy into the long range future. Others have thought about this question and have made important contributions toward a potential solution. Examples here include: future generations movements concerning intergenerational equity as evidenced in scholarship, law, and amongst Native American groups; Nez Perce and Confederated Tribes of the Umatilla Indian Reservation responses to the Hanford End State Vision and Hanford's Canyon Disposition Initiative; as well as the findings of organizational scholars on the advantages realized by organizations that have a long term future perspective. While these ideas inform the main line inquiry of this dissertation, the principal approach put forth by the researcher of how to convey an informed story about nuclear technology and waste into the long range future is implementation of the proposed Future Remains clause, as originated by the author, by amendment to two U.S. federal laws: National Historic Preservation Act and Comprehensive Environmental Response, Compensation, and Liability Act. The dissertation provides a case study in public anthropology. The findings of the dissertation include recommendations whereby the Future Remains clause gives historic preservation and cultural resources a leading and ongoing role in facilitating real-time forward looking historical documentation at environmental restoration projects at United States National Priorities List (i.e., "Superfund") sites.

12 CFR 908.52 - Partial summary disposition.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 12 Banks and Banking 7 2010-01-01 2010-01-01 false Partial summary disposition. 908.52 Section 908... summary disposition. If the presiding officer determines that a party is entitled to summary disposition... presiding officer has determined that summary disposition is warranted will be addressed in the recommended...

Fuel bundle design for enhanced usage of plutonium fuel

DOEpatents

Reese, Anthony P.; Stachowski, Russell E.

1995-01-01

A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced.

Fuel bundle design for enhanced usage of plutonium fuel

DOEpatents

Reese, A.P.; Stachowski, R.E.

1995-08-08

A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced. 10 figs.

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.

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

Density functional theory study of defects in unalloyed δ-Pu

DOE PAGES

Hernandez, S. C.; Freibert, F. J.; Wills, J. M.

2017-03-19

Using density functional theory, we explore in this paper various classical point and complex defects within the face-centered cubic unalloyed δ-plutonium matrix that are potentially induced from self-irradiation. For plutonium only defects, the most energetically stable defect is a distorted split-interstitial. Gallium, the δ-phase stabilizer, is thermodynamically stable as a substitutional defect, but becomes unstable when participating in a complex defect configuration. Finally, complex uranium defects may thermodynamically exist as uranium substitutional with neighboring plutonium interstitial and stabilization of uranium within the lattice is shown via partial density of states and charge density difference plots to be 5f hybridization betweenmore » uranium and plutonium.« less

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.

Determination of plutonium in nitric acid solutions using energy dispersive L X-ray fluorescence with a low power X-ray generator

NASA Astrophysics Data System (ADS)

Py, J.; Groetz, J.-E.; Hubinois, J.-C.; Cardona, D.

2015-04-01

This work presents the development of an in-line energy dispersive L X-ray fluorescence spectrometer set-up, with a low power X-ray generator and a secondary target, for the determination of plutonium concentration in nitric acid solutions. The intensity of the L X-rays from the internal conversion and gamma rays emitted by the daughter nuclei from plutonium is minimized and corrected, in order to eliminate the interferences with the L X-ray fluorescence spectrum. The matrix effects are then corrected by the Compton peak method. A calibration plot for plutonium solutions within the range 0.1-20 g L-1 is given.

Preparation of high purity plutonium oxide for radiochemistry instrument calibration standards and working standards

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

Wong, A.S.; Stalnaker, N.D.

1997-04-01

Due to the lack of suitable high level National Institute of Standards and Technology (NIST) traceable plutonium solution standards from the NIST or commercial vendors, the CST-8 Radiochemistry team at Los Alamos National Laboratory (LANL) has prepared instrument calibration standards and working standards from a well-characterized plutonium oxide. All the aliquoting steps were performed gravimetrically. When a {sup 241}Am standardized solution obtained from a commercial vendor was compared to these calibration solutions, the results agreed to within 0.04% for the total alpha activity. The aliquots of the plutonium standard solutions and dilutions were sealed in glass ampules for long termmore » storage.« less

Density functional theory study of defects in unalloyed δ-Pu

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

Hernandez, S. C.; Freibert, F. J.; Wills, J. M.

Using density functional theory, we explore in this paper various classical point and complex defects within the face-centered cubic unalloyed δ-plutonium matrix that are potentially induced from self-irradiation. For plutonium only defects, the most energetically stable defect is a distorted split-interstitial. Gallium, the δ-phase stabilizer, is thermodynamically stable as a substitutional defect, but becomes unstable when participating in a complex defect configuration. Finally, complex uranium defects may thermodynamically exist as uranium substitutional with neighboring plutonium interstitial and stabilization of uranium within the lattice is shown via partial density of states and charge density difference plots to be 5f hybridization betweenmore » uranium and plutonium.« less

17 CFR 10.91 - Summary disposition.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 17 Commodity and Securities Exchanges 1 2010-04-01 2010-04-01 false Summary disposition. 10.91... Disposition Without Full Hearing § 10.91 Summary disposition. (a) Filing of motions, answers. Any party who... decision as a matter of law may move for a summary disposition in his favor of all or any part of the...

SPRAY CALCINATION REACTOR

DOEpatents

Johnson, B.M.

1963-08-20

A spray calcination reactor for calcining reprocessin- g waste solutions is described. Coaxial within the outer shell of the reactor is a shorter inner shell having heated walls and with open regions above and below. When the solution is sprayed into the irner shell droplets are entrained by a current of gas that moves downwardly within the inner shell and upwardly between it and the outer shell, and while thus being circulated the droplets are calcined to solids, whlch drop to the bottom without being deposited on the walls. (AEC) H03 H0233412 The average molecular weights of four diallyl phthalate polymer samples extruded from the experimental rheometer were redetermined using the vapor phase osmometer. An amine curing agent is required for obtaining suitable silver- filled epoxy-bonded conductive adhesives. When the curing agent was modified with a 47% polyurethane resin, its effectiveness was hampered. Neither silver nor nickel filler impart a high electrical conductivity to Adiprenebased adhesives. Silver filler was found to perform well in Dow-Corning A-4000 adhesive. Two cascaded hot-wire columns are being used to remove heavy gaseous impurities from methane. This purified gas is being enriched in the concentric tube unit to approximately 20% carbon-13. Studies to count low-level krypton-85 in xenon are continuing. The parameters of the counting technique are being determined. The bismuth isotopes produced in bismuth irradiated for polonium production are being determined. Preliminary data indicate the presence of bismuth207 and bismuth-210m. The light bismuth isotopes are probably produced by (n,xn) reactions bismuth-209. The separation of uranium-234 from plutonium-238 solutions was demonstrated. The bulk of the plutonium is removed by anion exchange, and the remainder is extracted from the uranium by solvent extraction techniques. About 99% of the plutonium can be removed in each thenoyltrifluoroacetone extraction. The viscosity, liquid density, and selfdiffusion coefficient for lanthanum, cerium, and praseodymium were determined. The investigation of phase relationships in the plutonium-cerium-copper ternary system was continued on samples containing a high concentration of copper. These analyses indicate that complete solid solution exists between the binary compounds CeCu/sub 2/ and PuCu/sub 2/, thus forming a quasi-binary system. The study of high temperature ceramic fuel materials has continued with the homogenization and microspheroidization of binary mixtures of plutonium dioxide and zirconium dioxide. Sintering a die-pressed pellet of the mixed powders for one hour at 1450 deg C was not sufficient to completely react the constituents. Complete homogenization was obtained when the pellet was melted in the plasma flame. In addition to the plutonium dioxide-zirconium dioxide microspheres, pure beryllium oxide microspheres were produced in the plasma torch. The electronic distribution functions for the 10% by weight PuO/sub 2/ dissolved in a silicate glass were determined. The plutonium-oxygen interaction at about 2.2A is less than the plutonium-oxygen distance for the 5% PuO/sub 2/. The decrease in the interionic distance is indicative of a stronger plutonium-oxygen association for the more concentrated composition. Potassium plutonium sulfate is being evaluated as a reagent to quantitatively separate plutonium from aqueous solutions. The compound containing two waters of hydration was prepared for thermogravimetric studies using analytically pure plutonium-239. Because of the stability of this compound, it is being evaluated as a calorimetric standard for plutonium-238. (auth)

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).

Second-order Kinetics of DTPA and Plutonium in Rat Plasma

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

Miller, Guthrie; Poudel, Deepesh; Klumpp, John Allan

We report that in 2008, Serandour et al. reported on their in vitro experiment involving rat plasma samples obtained after an intravenous intake of plutonium citrate. Different amounts of DTPA were added to the plasma samples and the percentage of low-molecular-weight plutonium measured. Only when the DTPA dosage was three orders of magnitude greater than the recommended 30 μmol/kg was 100% of the plutonium apparently in the form of chelate. These data were modeled assuming three competing chemical reactions with other molecules that bind with plutonium. Here, time-dependent second-order kinetics of these reactions are calculated, intended eventually to become partmore » of a complete biokinetic model of DTPA action on actinides in laboratory animals or humans. The probability distribution of the ratio of stability constants for the reactants was calculated using Markov Chain Monte Carlo. In conclusion, these calculations substantiate that the inclusion of more reactions is needed in order to be in agreement with known stability constants.« less

Long-term retrievability and safeguards for immobilized weapons plutonium in geologic storage

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

Peterson, P.F.

1996-05-01

If plutonium is not ultimately used as an energy source, the quantity of excess weapons plutonium (w-Pu) that would go into a US repository will be small compared to the quantity of plutonium contained in the commercial spent fuel in the repository, and the US repository(ies) will likely be only one (or two) locations out of many around the world where commercial spent fuel will be stored. Therefore excess weapons plutonium creates a small perturbation to the long-term (over 200,000 yr) global safeguard requirements for spent fuel. There are details in the differences between spent fuel and immobilized w-Pu wastemore » forms (i.e. chemical separation methods, utility for weapons, nuclear testing requirements), but these are sufficiently small to be unlikely to play a significant role in any US political decision to rebuild weapons inventories, or to change the long-term risks of theft by subnational groups.« less

Second-order Kinetics of DTPA and Plutonium in Rat Plasma

DOE PAGES

Miller, Guthrie; Poudel, Deepesh; Klumpp, John Allan; ...

2017-11-15

We report that in 2008, Serandour et al. reported on their in vitro experiment involving rat plasma samples obtained after an intravenous intake of plutonium citrate. Different amounts of DTPA were added to the plasma samples and the percentage of low-molecular-weight plutonium measured. Only when the DTPA dosage was three orders of magnitude greater than the recommended 30 μmol/kg was 100% of the plutonium apparently in the form of chelate. These data were modeled assuming three competing chemical reactions with other molecules that bind with plutonium. Here, time-dependent second-order kinetics of these reactions are calculated, intended eventually to become partmore » of a complete biokinetic model of DTPA action on actinides in laboratory animals or humans. The probability distribution of the ratio of stability constants for the reactants was calculated using Markov Chain Monte Carlo. In conclusion, these calculations substantiate that the inclusion of more reactions is needed in order to be in agreement with known stability constants.« less

Plutonium Immobilization Project System Design Description for Can Loading System

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

Kriikku, E.

2001-02-15

The purpose of this System Design Description (SDD) is to specify the system and component functions and requirements for the Can Loading System and provide a complete description of the system (design features, boundaries, and interfaces), principles of operation (including upsets and recovery), and the system maintenance approach. The Plutonium Immobilization Project (PIP) will immobilize up to 13 metric tons (MT) of U.S. surplus weapons usable plutonium materials.

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.

METHOD OF SEPARATING URANIUM, PLUTONIUM AND FISSION PRODUCTS BY BROMINATION AND DISTILLATION

DOEpatents

Jaffey, A.H.; Seaborg, G.T.

1958-12-23

The method for separation of plutonium from uranium and radioactive fission products obtained by neutron irradiation of uranlum consists of reacting the lrradiated material with either bromine, hydrogen bromide, alumlnum bromide, or sulfur and bromine at an elevated temperature to form the bromides of all the elements, then recovering substantlally pure plutonium bromide by dlstillatlon in combinatlon with selective condensatlon at prescribed temperature and pressure.

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 OF PREPARING METAL HALIDES

DOEpatents

Hendrickson, A.V.

1958-11-18

The conversion of plutonium halides from plutonium peroxide can be done by washing the peroxide with hydrogen peroxide, drying the peroxide, passing a dry gaseous hydrohalide over the surface of the peroxide at a temperature of about lOO icient laborato C until the reaction rate has stabillzed, and then ralsing the reaction temperature to between 400 and 600 icient laborato C until the conversion to plutonium halide is substantially complete.

REDUCTION OF PLUTONIUM TO Pu$sup +3$ BY SODIUM DITHIONITE IN POTASSIUM CARBONATE

DOEpatents

Miller, D.R.; Hoekstra, H.R.

1958-12-16

Plutonium values are reduced in an alkaline aqueous medlum to the trlvalent state by means of sodium dlthionite. Plutonlum values are also separated from normally assoclated contaminants by metathesizing a lanthanum fluoride carrier precipitate containing plutonium with a hydroxide solution, performing the metathesis in the presence of about 0.2 M sodium dithionite at a temperature of between 40 and 90 icient laborato C.

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)

Literature Review: Crud Formation at the Liquid/Liquid Interface of TBP-Based Solvent-Extraction Processes

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

Delegard, Calvin H.; Casella, Amanda J.

2016-09-30

This report summarizes the literature reviewed on crud formation at the liquid:liquid interface of solvent extraction processes. The review is focused both on classic PUREX extraction for industrial reprocessing, especially as practiced at the Hanford Site, and for those steps specific to plutonium purification that were used at the Plutonium Reclamation Facility (PRF) within the Plutonium Finishing Plant (PFP) at the Hanford Site.

RECOVERY OF Pu VALUES BY FLUORINATION AND FRACTIONATION

DOEpatents

Brown, H.S.; Webster, D.S.

1959-01-20

A method is presented for the concentration and recovery of plutonium by fluorination and fractionation. A metallic mass containing uranium and plutonium is heated to 250 C and contacted with a stream of elemental fluorine. After fluorination of the metallic mass, the rcaction products are withdrawn and subjected to a distillation treatment to separate the fluorination products of uranium and to obtain a residue containing the fluorination products of plutonium.

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.

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

DOEpatents

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

1996-01-01

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

Radiation damage and annealing in plutonium tetrafluoride

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

McCoy, Kaylyn; Casella, Amanda; Sinkov, Sergey

Plutonium tetrafluoride that was separated prior to 1966 at the Hanford Site in Washington State was analyzed at the Pacific Northwest National Laboratory (PNNL) in 2015 and 2016. The plutonium tetrafluoride, as received, was an off-normal color and considering the age of the plutonium, there were questions about the condition of the material. These questions had to be answered in order to determine the suitability of the material for future use or long-term storage. Therefore, Thermogravimetric/Differential Thermal Analysis and X-ray Diffraction evaluations were conducted to determine the plutonium’s crystal structure, oxide content, and moisture content; these analyses reported that themore » plutonium was predominately amorphous and tetrafluoride, with an oxide content near ten percent. Freshly fluorinated plutonium tetrafluoride is known to be monoclinic. During the initial Thermogravimetric/Differential Thermal analyses, it was discovered that an exothermic event occurred within the material near 414°C. X-ray Diffraction analyses were conducted on the annealed tetrafluoride. The X-ray Diffraction analyses indicated that some degree of recrystallization occurred in conjunction with the 414°C event. The following commentary describes the series of Thermogravimetric/Differential Thermal and X-ray Diffraction analyses that were conducted as part of this investigation at PNNL, in collaboration with the University of Utah Nuclear Engineering Program.« less

Resuspension studies in the Marshall Islands.

PubMed

Shinn, J H; Homan, D N; Robison, W L

1997-07-01

The contribution of inhalation exposure to the total dose for residents of the Marshall Islands was monitored at occasions of opportunity on several islands in the Bikini and Enewetak Atolls. To determine the long-term potential for inhalation exposure, and to understand the mechanisms of redistribution and personal exposure, additional investigations were undertaken on Bikini Island under modified and controlled conditions. Experiments were conducted to provide key parameters for the assessment of inhalation exposure from plutonium-contaminated dust aerosols: characterization of the contribution of plutonium in soil-borne aerosols as compared to sea spray and organic aerosols, determination of plutonium resuspension rates as measured by the meteorological flux-gradient method during extreme conditions of a bare-soil vs. a stabilized surface, determination of the approximate individual exposures to resuspended plutonium by traffic, and studies of exposures to individuals in different occupational environments simulated by personal air sampling of workers assigned to a variety of tasks. Enhancement factors (defined as ratios of the plutonium-activity of suspended aerosols relative to the plutonium-activity of the soil) were determined to be less than 1 (typically 0.4 to 0.7) in the undisturbed, vegetated areas, but greater than 1 (as high as 3) for the case studies of disturbed bare soil, roadside travel, and for occupational duties in fields and in and around houses.

Ultra-small plutonium oxide nanocrystals: an innovative material in plutonium science.

PubMed

Hudry, Damien; Apostolidis, Christos; Walter, Olaf; Janssen, Arne; Manara, Dario; Griveau, Jean-Christophe; Colineau, Eric; Vitova, Tonya; Prüssmann, Tim; Wang, Di; Kübel, Christian; Meyer, Daniel

2014-08-11

Apart from its technological importance, plutonium (Pu) is also one of the most intriguing elements because of its non-conventional physical properties and fascinating chemistry. Those fundamental aspects are particularly interesting when dealing with the challenging study of plutonium-based nanomaterials. Here we show that ultra-small (3.2±0.9 nm) and highly crystalline plutonium oxide (PuO2 ) nanocrystals (NCs) can be synthesized by the thermal decomposition of plutonyl nitrate ([PuO2 (NO3 )2 ]⋅3 H2 O) in a highly coordinating organic medium. This is the first example reporting on the preparation of significant quantities (several tens of milligrams) of PuO2 NCs, in a controllable and reproducible manner. The structure and magnetic properties of PuO2 NCs have been characterized by a wide variety of techniques (powder X-ray diffraction (PXRD), X-ray absorption fine structure (XAFS), X-ray absorption near edge structure (XANES), TEM, IR, Raman, UV/Vis spectroscopies, and superconducting quantum interference device (SQUID) magnetometry). The current PuO2 NCs constitute an innovative material for the study of challenging problems as diverse as the transport behavior of plutonium in the environment or size and shape effects on the physics of transuranium elements. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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

Environmental consequences of postulated plutonium releases from General Electric Company Vallecitos Nuclear Center, Vallecitos, California, as a result of severe natural phenomena

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

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

1980-11-01

Potential environmental consequences in terms of radiation dose to people are presented for postulated plutonium releases caused by severe natural phenomena at the General Electric Company Vallecitos Nuclear Center, Vallecitos, California. The severe natural phenomena considered are earthquakes, tornadoes, and high straight-line winds. Maximum 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 also given. The most likelymore » 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 earthquakes, and the 180-mph and 230-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 135-mph tornado are below the EPA proposed guidelines.« 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.

Improved plutonium identification and characterization results with NaI(Tl) detector using ASEDRA

NASA Astrophysics Data System (ADS)

Detwiler, R.; Sjoden, G.; Baciak, J.; LaVigne, E.

2008-04-01

The ASEDRA algorithm (Advanced Synthetically Enhanced Detector Resolution Algorithm) is a tool developed at the University of Florida to synthetically enhance the resolved photopeaks derived from a characteristically poor resolution spectra collected at room temperature from scintillator crystal-photomultiplier detector, such as a NaI(Tl) system. This work reports on analysis of a side-by-side test comparing the identification capabilities of ASEDRA applied to a NaI(Tl) detector with HPGe results for a Plutonium Beryllium (PuBe) source containing approximately 47 year old weapons-grade plutonium (WGPu), a test case of real-world interest with a complex spectra including plutonium isotopes and 241Am decay products. The analysis included a comparison of photopeaks identified and photopeak energies between the ASEDRA and HPGe detector systems, and the known energies of the plutonium isotopes. ASEDRA's performance in peak area accuracy, also important in isotope identification as well as plutonium quality and age determination, was evaluated for key energy lines by comparing the observed relative ratios of peak areas, adjusted for efficiency and attenuation due to source shielding, to the predicted ratios from known energy line branching and source isotopics. The results show that ASEDRA has identified over 20 lines also found by the HPGe and directly correlated to WGPu energies.

The plutonium isotopic composition of marine biota on Enewetak Atoll: a preliminary assessment.

PubMed

Hamilton, Terry F; Martinelli, Roger E; Kehl, Steven R; McAninch, Jeffrey E

2008-10-01

We have determined the level and distribution of gamma-emitting radionuclides, plutonium activity concentrations, and 240Pu/239Pu atom ratios in tissue samples of giant clam (Tridacna gigas and Hippopus hippopus), a top snail (Trochus nilaticas) and sea cucumber (Holothuria atra) collected from different locations around Enewetak Atoll. The plutonium isotopic measurements were performed using ultra-high sensitivity accelerator mass spectrometry (AMS). Elevated levels of plutonium were observed in the stomachs (includes the stomach lining) of Tridacna clam (0.62 to 2.98 Bq kg(-1), wet wt.), in the soft parts (edible portion) of top snails (0.25 to 1.7 Bq kg(-1)), wet wt.) and, to a lesser extent, in sea cucumber (0.015 to 0.22 Bq kg(-1), wet wt.) relative to muscle tissue concentrations in clam (0.006 to 0.021 Bq kg(-1), wet wt.) and in comparison with previous measurements of plutonium in fish. These data and information provide a basis for re-evaluating the relative significance of dietary intakes of plutonium from marine foods on Enewetak Atoll and, perhaps most importantly, demonstrate that discrete 240Pu239Pu isotope signatures might well provide a useful investigative tool to monitor source-term attribution and consequences on Enewetak Atoll. One potential application of immediate interest is to monitor and assess the health and ecological impacts of leakage of plutonium (as well as other radionuclides) from a low-level radioactive waste repository on Runit Island relative to background levels of fallout contamination in Enewetak Atoll lagoon.

Verification of Plutonium Content in PuBe Sources Using MCNP® 6.2.0 Beta with TENDL 2012 Libraries

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

Lockhart, Madeline Louise; McMath, Garrett Earl

Although the production of PuBe neutron sources has discontinued, hundreds of sources with unknown or inaccurately declared plutonium content are in existence around the world. Institutions have undertaken the task of assaying these sources, measuring, and calculating the isotopic composition, plutonium content, and neutron yield. The nominal plutonium content, based off the neutron yield per gram of pure 239Pu, has shown to be highly inaccurate. New methods of measuring the plutonium content allow a more accurate estimate of the true Pu content, but these measurements need verification. Using the TENDL 2012 nuclear data libraries, MCNP6 has the capability to simulatemore » the (α, n) interactions in a PuBe source. Theoretically, if the source is modeled according to the plutonium content, isotopic composition, and other source characteristics, the calculated neutron yield in MCNP can be compared to the experimental yield, offering an indication of the accuracy of the declared plutonium content. In this study, three sets of PuBe sources from various backgrounds were modeled in MCNP6 1.2 Beta, according to the source specifications dictated by the individuals who assayed the source. Verification of the source parameters with MCNP6 also serves as a means to test the alpha transport capabilities of MCNP6 1.2 Beta with TENDL 2012 alpha transport libraries. Finally, good agreement in the comparison would indicate the accuracy of the source parameters in addition to demonstrating MCNP's capabilities in simulating (α, n) interactions.« less

Verification of Plutonium Content in PuBe Sources Using MCNP® 6.2.0 Beta with TENDL 2012 Libraries

DOE PAGES

Lockhart, Madeline Louise; McMath, Garrett Earl

2017-10-26

Although the production of PuBe neutron sources has discontinued, hundreds of sources with unknown or inaccurately declared plutonium content are in existence around the world. Institutions have undertaken the task of assaying these sources, measuring, and calculating the isotopic composition, plutonium content, and neutron yield. The nominal plutonium content, based off the neutron yield per gram of pure 239Pu, has shown to be highly inaccurate. New methods of measuring the plutonium content allow a more accurate estimate of the true Pu content, but these measurements need verification. Using the TENDL 2012 nuclear data libraries, MCNP6 has the capability to simulatemore » the (α, n) interactions in a PuBe source. Theoretically, if the source is modeled according to the plutonium content, isotopic composition, and other source characteristics, the calculated neutron yield in MCNP can be compared to the experimental yield, offering an indication of the accuracy of the declared plutonium content. In this study, three sets of PuBe sources from various backgrounds were modeled in MCNP6 1.2 Beta, according to the source specifications dictated by the individuals who assayed the source. Verification of the source parameters with MCNP6 also serves as a means to test the alpha transport capabilities of MCNP6 1.2 Beta with TENDL 2012 alpha transport libraries. Finally, good agreement in the comparison would indicate the accuracy of the source parameters in addition to demonstrating MCNP's capabilities in simulating (α, n) interactions.« less

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

Quantitative determination of environmental levels of uranium, thorium and plutonium in bone by solvent extraction and alpha spectrometry

NASA Astrophysics Data System (ADS)

Singh, Narayani P.; Zimmerman, Carol J.; Lewis, Laura L.; Wrenn, McDonald E.

1984-06-01

Solvent extraction and alpha-spectrometry have been emplyed in the quantitative simultaneous determination of uranium. thorium and plutonium. The bone specimens, spiked with 232U, 229Th and 242Pu tracers, are wet ashed with HNO 3 followed by alternate additions of a new drops of HNO 3 and H 2O 2. Uranium is reduced to the tetravalent state with 200 mg SnCl 2 and 25 ml HI. Uranium, thorium and plutonium are then coprecipitated with calcium as oxalate, heated to 550°C, dissolved in 50 ml HCl, and the acidity adjusted to 10 M. Uranium and plutonium are extracted into a 20% tri-lauryl amine (TLA) solution in xylene, leaving thorium in the aqueous phase. Plutonium is first back-extracted from the TLA phase by shaking with a 1:1.5 volume of 0.05 M NH 4I in 8 M HCl, which reduces Pu(IV) to Pu(III). Uranium is then back-extracted with an equal volume of 0.1 M HCl. Thorium, which was left in the aqueous phase, is evaporated to dryness, dissolved in 4 M HNO 3, and the acidity adjusted to 4 M. Thorium is then extracted into 20% TLA solution in xylene pre-equilibrated with 4 M HNO 3, and back-extracted with 10 M HCl. Uranium, thorium, and plutonium are then electrodeposited separately onto platinum discs and counted by an alpha-spectrometer with a multi-channel analyzer and surface barrier silicon diodes. The mean recoveries of uranium, thorium, and plutonium in bovine, dog, and human bones were over 70%.

An Examination of the Relationship between a Candidate's Disposition Assessments from Admission to the Program to Completion of the Program

ERIC Educational Resources Information Center

Seay, Darolyn

2017-01-01

The research focused on clarifying the responsibility of the institution preparing these individuals including but not limited to: identifying dispositions of successful teachers, evaluating those dispositions in teacher education candidates going into education, developing dispositions standards, and evaluating candidate dispositions as they…

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.

Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Edwards, Geoffrey W R; Priest, Nicholas D

2014-11-01

The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low as 1 mSv. In addition, if this method is extended so that Pu is also measured, then the combined amount of Pu and Pu is sufficiently high in the thorium-plutonium fuel that a committed effective dose of 1 mSv would be measurable. However, the fraction of Pu and Pu in the other two fuels is sufficiently low that a 1 mSv dose would remain below the detection limit using this technique. Thus new methods, such as fecal measurements of Pu (or other alpha emitters), will be required to measure exposure to these new fuels.

Initial Radionuclide Inventories

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

Miller, H

The purpose of this analysis is to provide an initial radionuclide inventory (in grams per waste package) and associated uncertainty distributions for use in the Total System Performance Assessment for the License Application (TSPA-LA) in support of the license application for the repository at Yucca Mountain, Nevada. This document is intended for use in postclosure analysis only. Bounding waste stream information and data were collected that capture probable limits. For commercially generated waste, this analysis considers alternative waste stream projections to bound the characteristics of wastes likely to be encountered using arrival scenarios that potentially impact the commercial spent nuclearmore » fuel (CSNF) waste stream. For TSPA-LA, this radionuclide inventory analysis considers U.S. Department of Energy (DOE) high-level radioactive waste (DHLW) glass and two types of spent nuclear fuel (SNF): CSNF and DOE-owned (DSNF). These wastes are placed in two groups of waste packages: the CSNF waste package and the codisposal waste package (CDSP), which are designated to contain DHLW glass and DSNF, or DHLW glass only. The radionuclide inventory for naval SNF is provided separately in the classified ''Naval Nuclear Propulsion Program Technical Support Document'' for the License Application. As noted previously, the radionuclide inventory data presented here is intended only for TSPA-LA postclosure calculations. It is not applicable to preclosure safety calculations. Safe storage, transportation, and ultimate disposal of these wastes require safety analyses to support the design and licensing of repository equipment and facilities. These analyses will require radionuclide inventories to represent the radioactive source term that must be accommodated during handling, storage and disposition of these wastes. This analysis uses the best available information to identify the radionuclide inventory that is expected at the last year of last emplacement, currently identified as 2030 and 2033, depending on the type of waste. TSPA-LA uses the results of this analysis to decay the inventory to the year of repository closure projected for the year of 2060.« 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.

Exploding the myths about the fast breeder reactor

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

Burns, S.

1979-01-01

This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

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

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.

Heterogeneity Effects in Plutonium Contaminated Soil

DTIC Science & Technology

2009-03-01

masses up to one kilogram once the ratio of Americium - 241 (Am- 241 ) and plutonium concentrations was established (Rademacher, 2001). Alpha...with a sample number and tared weight with a non-smearing marker. A standard control was then set using a point source of Americium - 241 on an aluminum...During the fire the weapons grade plutonium (Pu- 239, Pu-240, and Pu- 241 ) ignited and was released into the surrounding area, due to both